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This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S&M) activities at the 221-B Canyon Building and ancillary support structures (B Plant). The document replaces BHI-010582, Documented Safety Analysis for the B-Plant Facility. The B Plant is non-operational, deactivated and undergoing long term S&M prior to decontamination and decommissioning (D&D). This DSA is compliant with 10 CFR 830, Nuclear Safety Management, Subpart B, ''Safety Basis Requirements.'' The DSA was developed in accordance with U.S. Department of Energy (DOE) standard DOE-STD-1120-98, Integration of Environment, Safety, and Health into Facility Disposition Activities (DOE 1998) per Table 2 of 10 CFR 830 Appendix A, DOE Richland Operation Office (RL) direction (02-ABD-0053, Fluor Hanford Nuclear Safety Basis Strategy and Criteria) for facilities in long term S&M, and RL Direction (02-ABD-0091, ''FHI Nuclear Safety Expectations for Nuclear Facilities in Surveillance and Maintenance''). A crosswalk was prepared to identify potential inconsistencies between the previous B Plant safety analysis and DOE-STD-1120-98 guidance. In general, the safety analysis met the criteria of DOE-STD-1120-98. Some format and content changes have been made, including incorporating recent facility modifications and updating the evaluation guidelines and control selection criteria in accordance with RL direction (02-ABD-0053). The facility fire hazard analysis (FHA) and Technical Safety Requirements (TSR) are appended to this DSA as an aid to the users, to minimize editorial redundancy, and to provide an efficient basis for update.
Plutonium Finishing Plant safety evaluation report
Not Available
The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE`s independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91.
Kansas City Plant Celebrates Safety Milestone
A gang of motorcycle riders arrived at the NNSA's Kansas City Plant on July 1 to help celebrate a significant safety achievement - working nearly five million hours, covering a one-year period without a lost-time injury. The bikers -- some of whom are plant employees -- represent Bikers Against Child Abuse, the local nonprofit selected to receive a $5,000 donation as part of the plant's safety achievement celebration. The organization was selected because it aligns with the plant's community outreach focus on Family Safety & Security and partnership with the plant's union members.
Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches
Steven R. Sherman
The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.
SAFETY AND HEALTH PROGRAM Including the Chemical Hygiene Plan
Evans, Paul G.
SAFETY AND HEALTH PROGRAM Including the Chemical Hygiene Plan Wisconsin Center for Applied, Technical Staff & Chemical Hygiene Officer email@example.com 262-2982 Lab Facility Website http..........................................................................................................3 CHEMICAL HYGIENE PLAN III. Work-site Analysis and Hazard Identification 3.1 Hazardous Chemical
Waste Isolation Pilot Plant Safety Analysis Report
The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.
South Ukraine NPP: Safety improvements through Plant Computer upgrade
Brenman, O. [Westinghouse Electric Company, 4350 Northern Pike, Monroeville, PA 15146 (United States); Chernyshov, M. A. [Westron, LLC, 1 Acad. Proskura St., Kharkiv 61070 (Ukraine); Denning, R. S. [Battelle, 505 King Ave, Columbus, OH 43201 (United States); Kolesov, S. A. [NAEK Energoatom, 3 Vetrov Str., Kiev, 01032 (Ukraine); Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I. [PO South Ukraine NPP, NAEK Energoatom, Mylolayv Region, 55000 (Ukraine); Trosman, G. [US Dept. of Energy, International Nuclear Safety Program, Washington, DC 20585 (United States)
This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)
Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants
Kisner, Roger A [ORNL; Mullens, James Allen [ORNL; Wilson, Thomas L [ORNL; Wood, Richard Thomas [ORNL; Korsah, Kofi [ORNL; Qualls, A L [ORNL; Muhlheim, Michael David [ORNL; Holcomb, David Eugene [ORNL; Loebl, Andy [ORNL
Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.
A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems
A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon through safety analy- sis is strongly mandated for safety-critical systems. Nuclear plant protection. INTRODUCTION Safety-critical systems (e.g. nuclear power plants and air- planes) require rigorous quality
Floating nuclear power plant safety assurance principles
Zvonarev, B.M.; Kuchin, N.L.; Sergeev, I.V.
In the north regions of the Russian federation and low density population areas, there is a real necessity for ecological clean energy small power sources. For this purpose, floating nuclear power plants, designed on the basis of atomic ship building engineering, are being conceptualized. It is possible to use the ship building plants for the reactor purposes. Issues such as radioactive waste management are described.
Safety Evaluation Report of the Waste Isolation Pilot Plant Contact Handled (CH) Waste Documented Safety Analysis
Washington TRU Solutions LLC
This Safety Evaluation Report (SER) documents the Department of Energy�s (DOE's) review of Revision 9 of the Waste Isolation Pilot Plant Contact Handled (CH) Waste Documented Safety Analysis, DOE/WIPP-95-2065 (WIPP CH DSA), and provides the DOE Approval Authority with the basis for approving the document. It concludes that the safety basis documented in the WIPP CH DSA is comprehensive, correct, and commensurate with hazards associated with CH waste disposal operations. The WIPP CH DSA and associated technical safety requirements (TSRs) were developed in accordance with 10 CFR 830, Nuclear Safety Management, and DOE-STD-3009-94, Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports.
Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant
Perkins, W.C.; Durant, W.S.; Dexter, A.H.
The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.
Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR
Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.
This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.
Some aspects of nuclear power plant safety under war conditions
Stritar, A.; Mavko, B.; Susnik, J.; Sarler, B. (Jozef Stefan Inst., Ljubljana (Slovenia))
In the summer of 1991, the Krsko nuclear power plant in Slovenia found itself in an area of military operations. This was probably the first commercial nuclear power plant to have been threatened by an attack by fighter jets. A number of never-before-asked questions had to be answered by the operating staff and supporting organizations. Some aspects of nuclear power plant safety under war conditions are described, such as the selection of the best plant operating state before the attack and the determination of plant system vulnerability and dose releases from the potentially damaged spent fuel in the spent-fuel pit. The best operating mode to which the plant should be brought before the attack is cold shutdown, and radiological consequences to the environment after the spent fuel is damaged and the water in the pit is lost are not very high. The problem of nuclear power plant safety under war conditions should be addressed in more detail in the future.
PLC-Based Safety Critical Software Development for Nuclear Power Plants
PLC-Based Safety Critical Software Development for Nuclear Power Plants Junbeom Yoo1 , Sungdeok Cha development technique for nuclear power plants'I&C soft- ware controllers. To improve software safety, we in developing safety-critical control software for a Korean nuclear power plant, and experience to date has been
Aging of safety class 1E transformers in safety systems of nuclear power plants
Roberts, E.W.; Edson, J.L.; Udy, A.C. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)
This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.
An approach for solvent selection in extractive distillation systems including safety considerations
Grossmann, Ignacio E.
of bioethanol, for which solvents that offer the best cost-safety compromise are identified. Keywords. Extractive distillation; Multiobjective optimization; Process safety; Solvents; Bioethanol 1. Introduction
Floating LNG plant will stress reliability and safety
Kinney, C.D.; Schulz, H.R.; Spring, W.
Mobil has developed a unique floating LNG plant design after extensive studies that set safety as the highest priority. The result is a production, storage and offloading platform designed to produce 6 million tons per year of LNG and up to 55,000 bpd of condensate from 1 Bcfd of feed gas. All production and off-loading equipment is supported by a square donut-shaped concrete hull, which is spread-moored. The hull contains storage tanks for 250,000 m{sup 3} of LNG, 6540,000 bbl of condensate and ballast water. Both LNG and condensate can be directly offloaded to shuttle tankers. Since the plant may be moved to produce from several different gas fields during its life, the plant and barge were designed to be generic. It can be used at any location in the Pacific Rim, with up to 15% CO{sub 2}, 100 ppm H{sub 2}S, 55 bbl/MMcf condensate and 650 ft water depth. It can be modified to handle other water depths, depending upon the environment. In addition, it is much more economical than an onshore grassroots LNG plant, with potential capital savings of 25% or more. The paper describes the machinery, meteorology and oceanography, and safety engineering.
Aging of Class 1E batteries in safety systems of nuclear power plants
Edson, J.L.; Hardin, J.E.
This report presents the results of a study of aging effects on safety-related batteries in nuclear power plants. The purpose is to evaluate the aging effects caused by operation within a nuclear facility and to evaluate maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant-Aging Research approach and investigates the materials used in battery construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes battery-failure events reported in various data bases, and evaluates recommended maintenance practices. Data bases that were analyzed included the NRC's Licensee Event Report system, the Institute for Nuclear Power Operations' Nuclear Plant Reliability Data System, the Oak Ridge National Laboratory's In-Plant Reliability Data System, and The S.M. Stoller Corporation's Nuclear Power Experience data base.
Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications
Heljanko, Keijo
C4.2 Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications and control (I&C) systems play a crucial role in the operation of nuclear power plants (NPP) and other safety is available. The use of model checking to verify two nuclear power plant related systems is described: an arc
Preliminary Safety Information Document for the Standard MHTGR. Volume 1, (includes latest Amendments)
With NRC concurrence, the Licensing Plan for the Standard HTGR describes an application program consistent with 10CFR50, Appendix O to support a US Nuclear Regulatory Commission (NRC) review and design certification of an advanced Standard modular High Temperature Gas-Cooled Reactor (MHTGR) design. Consistent with the NRC's Advanced Reactor Policy, the Plan also outlines a series of preapplication activities which have as an objective the early issuance of an NRC Licensability Statement on the Standard MHTGR conceptual design. This Preliminary Safety Information Document (PSID) has been prepared as one of the submittals to the NRC by the US Department of Energy in support of preapplication activities on the Standard MHTGR. Other submittals to be provided include a Probabilistic Risk Assessment, a Regulatory Technology Development Plan, and an Emergency Planning Bases Report.
Lab Safety/Hazardous Waste Training Persons (including faculty, staff and students) working in a lab and work-
Tennessee, University of
Lab Safety/Hazardous Waste Training Persons (including faculty, staff and students) working in a lab and work- ing with hazardous materials should receive annual training that address- es lab safety, personal protective equipment, storage, use, and disposal of hazardous materials, emergency procedures
Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program
Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.
The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.
Abstract Successful transformation of plant tissue using Agrobacterium relies on several factors including bacterial
Finer, John J.
Abstract Successful transformation of plant tissue using Agrobacterium relies on several factors including bacterial infection, host recognition, and transformation competency of the target tissue particle bombardment, Agrobacterium-mediated transformation of this tissue has not been demonstrated. We
Paris-Sud XI, Université de
by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe stateASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant
Underwater nuclear power plants: improved safety, environmental compatibility and efficiency
Galustov, K.Z.; Abadjyan, K.A.; Pavlov, A.B.
The further development of nuclear power engineering depends on the creation of a new generation of nuclear power plant (NPP) projects that have a high degree of safety. Decisions ensuring secure NPP exploitation must be based on the possibility of eliminating or localizing accidents. Using environmental properties to achieve secure NPP exploitation and accident elimination leads to suggest the construction of NPPs in water. An efficient way to provide energy to remote coastal areas is through use of floatable construction of prefabricated units. Floatable construction raises the quality of works, reduces expenditures on industrial facilities, and facilities building conditions in districts with extreme climatic conditions. A type of NPP that is situated on a shelf with the reactor compartment placed at the sea bottom is proposed. The underwater location of the reactor compartment on the fixed depth allows the natural water environment conditions of natural hydrostatic pressure, heat transfer and circulation to provide NPP safety. An example of new concept for power units with under-water localization of the reactor compartment is provided by the double-block NPP in a VVER reactor.
Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant
Durant, W.S.; Perkins, W.C.; Lee, R.; Stoddard, D.H.
The Safety Technology Group is developing methodology that can be used to assess the risk of operating a plant to reprocess spent nuclear fuel. As an early step in the methodology, a preliminary hazards analysis identifies safety-related incidents. In the absence of appropriate safety features, these incidents could lead to significant consequences and risk to onsite personnel or to the public. This report is a compilation of potential safety-related incidents that have been identified in studies at SRL and in safety analyses of various commercially designed reprocessing plants. It is an expanded revision of the version originally published as DP-1558, Published December 1980.
Campbell, Andrew T.
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Nuclear criticality safety evaluation of Spray Booth Operations in X-705, Portsmouth Gaseous Diffusion Plant
Sheaffer, M.K.; Keeton, S.C.
This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested.
Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR
Morris, F.A.; Hooper, R.L.
The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.
Engineering support services for the DOE/GRI coal gasification research program. Safety audits of pilot plants and PDU's
Bostwick, L.E.; Hubbard, D.A.; Lee, M.D.; Miller, G.R.; Bernard, D.M.
M.W. Kellogg (formerly Pullmann Kellogg) was requested by DOE to investigate and to evaluate normal and emergency operating procedures and the drawing record systems of the coal gasification pilot plants and process development units (PDU). The purpose of this Safety Audit was to identify deficiencies in operating policies or procedures which could lead to potential hazards. The evaluation of safety-related documentation at the pilot plants and PDU's was also included in the audit. The safety audit visits and meetings were conducted at the following research sites: Bell Aerosopace, BCR BI-GAS, Exxon, IGT Hygas/Peatgas, Rockwell International, and Westinghouse. Kellogg conducted the safety audits requested by DOE. These reviews show the developers as possessing very sincere, positive attitudes toward safety and as being committed to ongoing safety programs. Kellogg found that (in general) all of the developers: use written statements of objectives, operating procedures and check lists; have some form of formal safety training for operators; review equipment and procedural revisions with operators; and maintain timely and accurate drawing records.
Poinsot, Laurent
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Implementation of Recommendations from the One System Comparative Evaluation of the Hanford Tank Farms and Waste Treatment Plant Safety Bases
Garrett, Richard L. [Washington River Protection Systems, Richland, WA (United States); Niemi, Belinda J. [Washington River Protection Systems, Richland, WA (United States); Paik, Ingle K. [Washington River Protection Systems, Richland, WA (United States); Buczek, Jeffrey A. [AREVA Federal Services LLC (United States); Lietzow, J. [URS Professional Services (United States); McCoy, F. [AREVA Federal Services LLC (United States); Beranek, F. [URS Professional Services (United States); Gupta, M. [URS Professional Services (United States)
A Comparative Evaluation was conducted for One System Integrated Project Team to compare the safety bases for the Hanford Waste Treatment and Immobilization Plant Project (WTP) and Tank Operations Contract (TOC) (i.e., Tank Farms) by an Expert Review Team. The evaluation had an overarching purpose to facilitate effective integration between WTP and TOC safety bases. It was to provide One System management with an objective evaluation of identified differences in safety basis process requirements, guidance, direction, procedures, and products (including safety controls, key safety basis inputs and assumptions, and consequence calculation methodologies) between WTP and TOC. The evaluation identified 25 recommendations (Opportunities for Integration). The resolution of these recommendations resulted in 16 implementation plans. The completion of these implementation plans will help ensure consistent safety bases for WTP and TOC along with consistent safety basis processes. procedures, and analyses. and should increase the likelihood of a successful startup of the WTP. This early integration will result in long-term cost savings and significant operational improvements. In addition, the implementation plans lead to the development of eight new safety analysis methodologies that can be used at other U.S. Department of Energy (US DOE) complex sites where URS Corporation is involved.
Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant
Elder, H. K.
Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0.88 million, the annual maintenance and surveillance cost is estimated to be about $0.095 million, and deferred decontamination is estimated to cost about $6.50 million. Therefore, passive SAFSTOR for 10 years is estimated to cost $8.33 million in nondiscounted 1981 dollars. DECON with lagoon waste stabilization is estimated to cost about $4.59 million, with an annual cost of $0.011 million for long-term care. All of these estimates include a 25% contingency. Waste management costs for DECON, including the net cost of disposal of the solvent extraction lagoon wastes by shipping those wastes to a uranium mill for recovery of residual uranium, comprise about 38% of the total decommissioning cost. Disposal of lagoon waste at a commercial low-level waste burial ground is estimated to add $10.01 million to decommissioning costs. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year committed dose equivalent to members of the public from airborne releases during normal decommissioning activities is estimated to 'Je about 4.0 man-rem. Radiation doses to the public from accidents are found to be very low for all phases of decommissioning. Occupational radiation doses from normal decommissioning operations (excluding transport operations) are estimated to be about 79 man-rem for DECON and about 80 man-rem for passive SAFSTOR with 10 years of safe storage. Doses from DECON with lagoon waste stabilization are about the same as for DECON except there is less dose resulting from transportation of radioactive waste. The number of fatalities and serious lost-time injuries not related to radiation is found to be very small for all decommissioning alternatives. Comparison of the cost estimates shows that DECON with lagoon waste stabilization is the least expensive method. However, this alternative does not allow unrestricted release of the site. The cumulative cost of maintenance and surveillance and the higher cost of deferred decontamination makes passive SAFSTOR more expensive than DECON. Seve
Optimization of Industrial Refrigeration Plants: Including a Case Study at Stonyfield Farm Yogurt
Dixon, R.; McCowan, B.; Drake, L.; Epstein, G.; D'Antonio, M.; Moray, S.
controls and unloading (specifically in the case of screw compressors which do not unload linearly). A lower refrigerant temperature results in lower suction pressure and increased compressor power requirements. A lower condensing pressure, which is a...Optimization of Industrial Refrigeration Plants: Including a Case Study at Stonyfield Farm Yogurt Mark D�Antonio Satyen Moray Brian McCowan Gary Epstein VP Engineering Services Project Manager VP Technology & Development President Energy...
Individual plant examination program: Perspectives on reactor safety and plant performance. Part 1: Final summary report; Volume 1
This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.
Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2
This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.
Assessment of safety-critical software in nuclear power plants
Parnas, D.L.; Madey, J. [McMaster Univ., Hamilton, Ontario (Canada); Asmis, G.J.K. [Atomic Energy Control Board, Ottawa (Canada)
This article outlines an approach in the design, documentation, and evaluation of computer systems. This allows the use of software in many safety-critical applications because it enables the systematic comparison of the program behavior with the engineering specifications of the computer system. Many of the ideas in this article have been used by the Atomic Energy Control Board of Canada in its safety assessment of the software for the shutdown systems of the Darlington Station. The four main elements of this approach follow: (1) Formal Documentation of Software Requirements: Most of the details of a complex environment can be ignored by system implementers and reviewers if they are given a complete and precise statement of the behavioral requirements for the computer system. We describe five mathematical relations that specify the requirements for the software in a computerized control system. (2) Design and Documentation of the Module Structure: Complexity caused by interactions between separately written components can be reduced by applying Data Abstraction, Abstract Data Types, and Object-Oriented Programming if the interfaces are precisely and completely documented. (3) Program Function Documentation: Software executions are lengthy sequences of state changes described by algorithms. The effects of these executive sequences can be precisely specified documented with tabular presentations of the program functions. Also, large programs can be decomposed and presented at a collection of well-documented smaller programs. (4) Tripod Approach to Assessment: There are three basic approaches to the assessment of complex software products: (i) testing, (ii) systematic inspection, and (iii) certification of people and processes. Assessment of a complex system cannot depend on any one of these alone. The approach used on the Darlington shutdown software, which included systematic inspection as well as planned and statistically designed random testing, is outlined.
Pantex Plant Achieves Key Safety Milestone Ahead of Schedule...
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Achieves Key Safety Milestone Ahead of Schedule | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...
Concentration of Actinides in Plant Mounds at Safety Test Nuclear Sites in Nevada
David S. Shafer; Jenna Gommes
Plant mounds or blow-sand mounds are accumulations of soil particles and plant debris around large shrubs and are common features in deserts in the southwestern United States. Believed to be an important factor in their formation, the shrubs create surface roughness that causes wind-suspended particles to be deposited and resist further suspension. Shrub mounds occur in some plant communities on the Nevada Test Site, the Nevada Test and Training Range (NTTR), and Tonopah Test Range (TTR), including areas of surface soil contamination from past nuclear testing. In the 1970s as part of early studies to understand properties of actinides in the environment, the Nevada Applied Ecology Group (NAEG) examined the accumulation of isotopes of Pu, {sup 241}Am, and U in plant mounds at safety test sites. The NAEG studies found concentrations of these contaminants to be greater in shrub mounds than in the surrounding areas of desert pavement. For example, at Project 57 on the NTTR, it was estimated that 15 percent of the radionuclide inventory of the site was associated with shrub mounds, which accounted for 17 percent of the surface area of the site, a ratio of inventory to area of 0.85. At Clean Slate III at the TTR, 29 percent of the inventory was associated with approximately 32 percent of the site covered by shrub mounds, a ratio of 0.91. While the total inventory of radionuclides in intershrub areas was greater, the ratio of radionuclide inventory to area was 0.40 and 0.38, respectively, at the two sites. The comparison between the shrub mounds and adjacent desert pavement areas was made for only the top 5 cm since radionuclides at safety test sites are concentrated in the top 5 cm of intershrub areas. Not accounting for radionuclides associated with the shrub mounds would cause the inventory of contaminants and potential exposure to be underestimated. As part of its Environmental Restoration Soils Subproject, the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office has proposed that the majority of its contaminated soil 'Corrective Action Units', including the safety test sites, be closed by fencing and posting with administrative controls. The concentration of actinides in the shrub mounds has important implications for postclosure management of the safety test sites. Because resuspension factors at safety test sites can be three to four orders-of-magnitude higher than soil sites associated with atmospheric tests where criticality occurred, the shrub mounds are an important factor in stabilization of actinide contaminants. Loss of shrubs associated with mounds from fire or plant die-back from drought could cause radionuclides at these sites to become more prone to suspension and water erosion until the sites are stabilized. Alternatively, although shrub mounds are usually composed of predominantly fine sand size particles, smaller silt and clay size particles in them are often high in CaCO{sub 3} content. The CaCO{sub 3} may act as a cementing agent to limit erosion of the shrub mounds even if the vegetation cover is temporarily lost.
Technical safety appraisal: Buildings 776/777 Rocky Flats Plant
Field, H C
Buildings 776/777 at the Rocky Flats Plant are major components of the production complex at the plant site. They have been in operation since 1957. The operations taking place in the buildings are nuclear weapons production support, processing of weapons assemblies returned from Pantex, waste processing, research and development in support of production, special projects, and those generated by support groups, such as maintenance. The appraisal team identified nine deficiencies that it believed required prompt attention. DOE management for EH, the program office (Defense Programs), and the field office analyzed the information provided by the appraisal team and instituted compensatory measures for closer monitoring of contractor activities by knowledgeable DOE staff and staff from other sites. Concurrently, the contractor was requested to address both short-term and long-term remedial measures to correct the identified issues as well as the underlying problems. The contractor has provided his action plan, which is included. This plan was under evaluation by EH and the DOE program office at the time this report was prepared. In addressing the major areas of concern identified above, a well as the specific deficiencies identified by the appraisal team, the contractor and the field office are cautioned to search for the root causes for the problems and to direct corrective actions to those root causes rather than solely to the symptoms to assure the sustainability of the improvements being made. The results of prior TSAs led DOE to conclude that previous corrective actions were not sufficient in that a large number of the individual findings are recurrent. Pending completion of remedial actions over the next few months, enhanced DOE oversight of the contractor is warranted.
The impact of offsite factors on the safety performance of small nuclear power plants
Baranaev, Yu.D.; Viktorov, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation)
The results of an analysis of the influence of offsite factors on small nuclear power-plant (SNPP) safety performance during postulated severe accidents are presented. Given the plant locations in the immediate vicinity of residential areas and the impossibility of accomplishing the expeditious evacuation of the public, the risk caused by an SNPP severe accident may be considerably less than that for such an event in a large nuclear power plant. 3 refs., 3 figs., 5 tabs.
Safety provisions for UF{sub 6} handling in the design of a new UF{sub 6} conversion plant
Bannister, S.P. [British Nuclear Fuels plc, Preston (United Kingdom)
British Nuclear Fuels plc (BNFL) Fuel Division is currently undertaking the final design and construction of a new UF{sub 6} conversion plant at its production site at Springfields near Preston in the north of England. The Company has gained much experience in the handling of UF{sub 6} during operation of plants on site since 1961. The major hazard occurs during the liquefication cycle and the basis of the maximum credible incident scenario adopted for safety assessment and design purposes is discussed. This paper considers the design features which have been incorporated in the new plant to counter the hazards presented by the presence of UF{sub 6} in gaseous and liquid form and explains current thinking on operational procedures in areas of potential risk such as cylinder filling. The plant emergency response philosophy and systems are described and specific design provisions which have been included to satisfy the UK regulatory bodies are outlined in some detail.
Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants
Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)
This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.
lue light affects many aspects of plant growth and develop-ment. Plant blue-light responses include inhibition of
Lin, Chentao
B lue light affects many aspects of plant growth and develop- ment. Plant blue-light responses have identified three blue- light receptors: cryptochrome 1, cryptochrome 2 and phototropin, which of the cryptochromes Plant blue-light receptors have long been recognized from their action spectra to have roles
Explore the Genetic Frontier: 10 Steps to Safety: Regulating Plant Biotechnology
Vestal, Andy
. They depend heavily on universities for knowledge and guidance to set regulatory standards. As a result, the United States ranks among the world leaders in food and environmental safety. Food biotechnology products coming to market face stringent evaluation...DNA biotechnology-derived foods compares the new plant variety to its traditional counterpart to identify and evaluate differences that might pose safety or nutritional concerns. Are consumers eating large quantities of rDNA? No. The DNA introduced using r...
Entomology Department safety training for new employees Anyone who works in a laboratory (including the Bohart Museum)
Ishida, Yuko
the Bohart Museum) 1. General training a. Employee completes the online UC Laboratory Safety Fundamentals
Worker Involvement Improves Safety at Hanford Site's Plutonium Finishing Plant
Broader source: Energy.gov [DOE]
Employees at the Hanford site are working together to find new and innovative ways to stay safe at the Plutonium Finishing Plant, one of the site�s most complex decommissioning projects.
Luminant's Big Brown Plant wins for continuous improvement and safety programs
Peltier, R.
Staff from Luminant's Big Brown Plant accepted the PRB Coal Users' Group's top honour for innovative improvements to coal-handling systems and a sterling safety record. The numbers reveal their accomplishments: an average EFOR less than 4%, an availability factor averaging 90% for a plant that burns a lignite/PRB mix, and staff who worked more than 2.6 million man-hours since March 2000 without a lost-time injury. 13 photos., 1 tab.
Next Generation Nuclear Plant Structures, Systems, and Components Safety Classification White Paper
Pete Jordan
This white paper outlines the relevant regulatory policy and guidance for a risk-informed approach for establishing the safety classification of Structures, Systems, and Components (SSCs) for the Next Generation Nuclear Plant and sets forth certain facts for review and discussion in order facilitate an effective submittal leading to an NGNP Combined Operating License application under 10 CFR 52.
An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989
Mattson, Roger J.
This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.
Evaluation of station blackout accidents at nuclear power plants: Technical findings related to unresolved safety issue A-44: Final report
Not Available
''Station Blackout,'' which is the complete loss of alternating current (AC) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on AC power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the fequency of loss of offsite power; the probability that emergency or onsite AC power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of AC power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without AC power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events.
The Integrated Safety Management System Verification Enhancement Review of the Plutonium Finishing Plant (PFP)
The primary purpose of the verification enhancement review was for the DOE Richland Operations Office (RL) to verify contractor readiness for the independent DOE Integrated Safety Management System Verification (ISMSV) on the Plutonium Finishing Plant (PFP). Secondary objectives included: (1) to reinforce the engagement of management and to gauge management commitment and accountability; (2) to evaluate the ''value added'' benefit of direct public involvement; (3) to evaluate the ''value added'' benefit of direct worker involvement; (4) to evaluate the ''value added'' benefit of the panel-to-panel review approach; and, (5) to evaluate the utility of the review's methodology/adaptability to periodic assessments of ISM status. The review was conducted on December 6-8, 1999, and involved the conduct of two-hour interviews with five separate panels of individuals with various management and operations responsibilities related to PFP. A semi-structured interview process was employed by a team of five ''reviewers'' who directed open-ended questions to the panels which focused on: (1) evidence of management commitment, accountability, and involvement; and, (2) consideration and demonstration of stakeholder (including worker) information and involvement opportunities. The purpose of a panel-to-panel dialogue approach was to better spotlight: (1) areas of mutual reinforcement and alignment that could serve as good examples of the management commitment and accountability aspects of ISMS implementation, and, (2) areas of potential discrepancy that could provide opportunities for improvement. In summary, the Review Team found major strengths to include: (1) the use of multi-disciplinary project work teams to plan and do work; (2) the availability and broad usage of multiple tools to help with planning and integrating work; (3) senior management presence and accessibility; (4) the institutionalization of worker involvement; (5) encouragement of self-reporting and self-assessment by management; (6) the availability of multiple internal communication mechanisms; and, (7) the existence of overall facility-wide safety management goals as well as individualized project work team goals. Major opportunities for improvement identified include: (1) the enhancement of external communications relative to ISM; (2) the institutionalization of ISM-related performance agreements/incentives; (3) the strengthening of feedback loops; (4) fine-tuning the use of tools; and, (5) the formalization of good practices.
Nuclear Safety Risk Management in Refueling Outage of Qinshan Nuclear Power Plant
Meijing Wu; Guozhang Shen [Qinshan Nuclear power company (China)
The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclear safety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control. (authors)
Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant
This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation.
Environment, safety and Health Progress Assessment of the Rocky Flats Plant
Not Available
This report documents the result of the US Department of Energy`s (DOE) Environment, Safety and Health (ES&H) Progress Assessment of the DOE Rocky Flats Plant (RFP) in Golden, Colorado. The assessment, which was conducted during the period of May 17 through May 28, 1993, included a selective review of the ES&H management systems and programs of the responsible DOE Headquarters Program Offices (Defense Programs (DP) and Environmental Restoration and Waste Management (EM)), the DOE Rocky Flats Office (RFO), and the site contractor, EG&G Rocky Flats, Inc. (EG&G). Despite the near constant state of flux under which RFP has been required to operate, the Progress Assessment Team has concluded that significant progress has been made in correcting the deficiencies identified in the 1989 Assessment and in responding responsibly to regulations, and DOE directives and guidance that have been issued since that time. The Team concluded that the improvements have been concentrated in the activities associated with plutonium facilities and in regulatory driven programs. Much remains to be done with respect to implementing on a sitewide basis those management systems that anchor an organization`s pursuit of continuous ES&H improvement. Furthermore the Team concluded that the pace of improvement has been constrained by a combination of factors that have limited the site`s ability to manage change in the pursuit of sitewide ES&H excellence.
Management of the aging of critical safety-related concrete structures in light-water reactor plants
Naus, D.J.; Oland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))
The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs.
Use of artificial intelligence to enhance the safety of nuclear power plants
Uhrig, R.E.
In the operation of a nuclear power plant, the sheer magnitude of the number of process parameters and systems interactions poses difficulties for the operators, particularly during abnormal or emergency situations. Recovery from an upset situation depends upon the facility with which the available raw data can be converted into and assimilated as meaningful knowledge. Plant personnel are sometimes affected by stress and emotion, which may have varying degrees of influence on their performance. Expert systems can take some of the uncertainty and guesswork out of their decisions by providing expert advice and rapid access to a large information base. Application of artificial intelligence technologies, particularly expert systems, to control room activities in a nuclear power plant has the potential to reduce operator error and improve power plant safety and reliability. 12 refs.
Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework
Cappelli, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Gadomski, A. M. [ECONA, Centro Interuniversitario Elaborazione Cognitiva Sistemi Naturali e Artificiali, via dei Marsi 47, Rome (Italy); Sepiellis, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Wronikowska, M. W. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Poznan School of Social Sciences (Poland)
In the field of nuclear power plant (NPP) safety modeling, the perception of the role of socio-cognitive engineering (SCE) is continuously increasing. Today, the focus is especially on the identification of human and organization decisional errors caused by operators and managers under high-risk conditions, as evident by analyzing reports on nuclear incidents occurred in the past. At present, the engineering and social safety requirements need to enlarge their domain of interest in such a way to include all possible losses generating events that could be the consequences of an abnormal state of a NPP. Socio-cognitive modeling of Integrated Nuclear Safety Management (INSM) using the TOGA meta-theory has been discussed during the ICCAP 2011 Conference. In this paper, more detailed aspects of the cognitive decision-making and its possible human errors and organizational vulnerability are presented. The formal TOGA-based network model for cognitive decision-making enables to indicate and analyze nodes and arcs in which plant operators and managers errors may appear. The TOGA's multi-level IPK (Information, Preferences, Knowledge) model of abstract intelligent agents (AIAs) is applied. In the NPP context, super-safety approach is also discussed, by taking under consideration unexpected events and managing them from a systemic perspective. As the nature of human errors depends on the specific properties of the decision-maker and the decisional context of operation, a classification of decision-making using IPK is suggested. Several types of initial situations of decision-making useful for the diagnosis of NPP operators and managers errors are considered. The developed models can be used as a basis for applications to NPP educational or engineering simulators to be used for training the NPP executive staff. (authors)
A probabilistic safety analysis of UF{sub 6} handling at the Portsmouth Gaseous Diffusion Plant
Boyd, G.J.; Lewis, S.R.; Summitt, R.L. [Safety and Reliability Optimization Services (SAROS), Inc., Knoxville, TN (United States)
A probabilistic safety study of UF{sub 6} handling activities at the Portsmouth Gaseous Diffusion Plant has recently been completed. The analysis provides a unique perspective on the safety of UF{sub 6} handling activities. The estimated release frequencies provide an understanding of current risks, and the examination of individual contributors yields a ranking of important plant features and operations. Aside from the probabilistic results, however, there is an even more important benefit derived from a systematic modeling of all operations. The integrated approach employed in the analysis allows the interrelationships among the equipment and the required operations to be explored in depth. This paper summarizes the methods used in the study and provides an overview of some of the technical insights that were obtained. Specific areas of possible improvement in operations are described.
Introduction to the nuclear criticality safety evaluation of facility X-705, Portsmouth Gaseous Diffusion Plant
Sheaffer, M.K.; Keeton, S.C.
This report is the first in a series of documents that will evaluate nuclear criticality safety in the Decontamination and Recovery Facility, X-705, Portsmouth Gaseous Diffusion Plant. It provides an overview of the facility, categorizes its functions for future analysis, reviews existing NCS documentation, and explains the follow-on effort planned for X-705. A detailed breakdown of systems, subsystems, and operational areas is presented and cross-referenced to existing NCS documentation.
Use of probabilistic risk assessment in expert system usage for nuclear power plant safety
Uhrig, R.E.
The introduction of probability risk assessments (PRA's) to nuclear power plants in the Rasmussen Report (WASH-1400) gave us a means of evaluating the risk to the public associated with the operation of nuclear power plants, at least on a relative basis. While the choice of the ''source term'' and methodology in a PRA significantly influence the absolute probability and the consequences of core melt, comparison of two PRA calculations for two configurations of the same plant, carried out on a consistent basis, can be readily identify the increase in risk associated with going from one configuration of a plant to another by removing components or systems from service. This ratio of core melt probabilities (assuming no recovery of failed systems) obtained from two PRA calculations for different configurations was the criterion (called ''risk factor'') chosen as a basis for making a decision in an expert system as to what mitigating action, if any, would be taken to avoid a trip situation from developing. PRISIM was developed by JBF Associates of Knoxville under the sponsorship of the NRC as a system for Resident Inspectors at nuclear power plants to provide them with a relative safety status of the plant under all configurations. PRISIM calculated the risk factor---the ration of core melt probabilities of the plant under the current configuration relative to the normal configuration with all systems functioning---using an algorithm that emulates the results of the original PRA. It also presents time and core melt (assuming no recovery of systems or components).
The potential role of new technology for enhanced safety and performance of nuclear power plants through improved service maintenance
Achorn, Ted Glen
Refinements in the safety and performance of nuclear power plants must be made to maintain public confidence and ensure competitiveness with other power sources. The aircraft industry, US Navy, and other programs have ...
Safety analysis -- 200 Area Savannah River Plant, F-Canyon Operations. Supplement 4
Beary, M.M.; Collier, C.D.; Fairobent, L.A.; Graham, R.F.; Mason, C.L.; McDuffee, W.T.; Owen, T.L.; Walker, D.H.
The F-Canyon facility is located in the 200 Separations Area and uses the Purex process to recover plutonium from reactor-irradiated uranium. The irradiated uranium is normally in the form of solid or hollow cylinders called slugs. These slugs are encased in aluminum cladding and are sent to the F-Canyon from the Savannah River Plant (SRP) reactor areas or from the Receiving Basin for Offsite Fuels (RBOF). This Safety Analysis Report (SAR) documents an analysis of the F-Canyon operations and is an update to a section of a previous SAR. The previous SAR documented an analysis of the entire 200 Separations Area operations. This SAR documents an analysis of the F-Canyon and is one of a series of documents for the Separations Area as specified in the Savannah River Implementation Plans. A substantial amount of the information supporting the conclusions of this SAR is found in the Systems Analysis. Some F-Canyon equipment has been updated during the time between the Systems Analysis and this SAR and a complete description of this equipment is included in this report. The primary purpose of the analysis was to demonstrate that the F-Canyon can be operated without undue risk to onsite or offsite populations and to the environment. In this report, risk is defined as the expected frequency of an accident, multiplied by the resulting radiological consequence in person-rem. The units of risk for radiological dose are person-rem/year. Maximum individual exposure values have also been calculated and reported.
Worker Safety and Health Program for DOE (Including the National Nuclear Security Administration) Federal and Contractor Employees
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This document was developed to assist the Department of Energy (DOE or the Department) Federal and contractor employees in effectively developing, managing, and implementing a worker safety and health program. Cancels DOE G 440.1-1A and DOE G 440.1-8. Adm Chg 1, dated 3-22-13.
Worker Safety and Health Program for DOE (Including the National Nuclear Security Administration) Federal and Contractor Employees
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This document was developed to assist the Department of Energy (DOE or the Department) Federal and contractor employees in effectively developing, managing, and implementing a worker safety and health program. Cancels DOE G 440.1-1A and DOE G 440.1-8. Adm Chg 1, dated 3-22-13, cancels DOE G 440.1-1B.
Review guidelines on software languages for use in nuclear power plant safety systems. Final report
Hecht, H.; Hecht, M.; Graff, S.; Green, W.; Lin, D.; Koch, S.; Tai, A.; Wendelboe, D. [SoHaR, Inc., Beverly Hills, CA (United States)
Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada, C/C++, Programmable Logic Controller (PLC) Ladder Logic, International Electrotechnical Commission (IEC) Standard 1131-3 Sequential Function Charts, Pascal, and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.s
Inspection of Environment, Safety, and Health Management at the Waste Isolation Pilot Plant- Volume I, August 2002
Broader source: Energy.gov [DOE]
The Secretary of Energy�s Office of Independent Oversight and Performance Assurance (OA) conducted an inspection of environment, safety, and health (ES&H) and emergency management programs at the Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) in July and August 2002. The inspection was performed as a joint effort by the OA Office of Environment, Safety and Health Evaluations and the Office of Emergency Management Oversight. This volume discusses the results of the review of the WIPP ES&H programs. The results of the review of the WIPP emergency management program are discussed in Volume II of this report, and the combined results are discussed in a summary report. As discussed throughout this report, the ISM program at WIPP is generally effective. Although improvements are warranted in some areas, the current programs have contributed to overall effective ES&H performance and a good safety record at WIPP. Section 2 of this volume provides an overall discussion of the results of the review of the WIPP ES&H programs, including positive aspects and weaknesses. Section 3 provides OA�s conclusions regarding the overall effectiveness of CBFO and WTS management of the ES&H programs. Section 4 presents the ratings assigned as a result of this review. Appendix A provides supplemental information, including team composition. Appendix B identifies the specific finding that requires corrective action and follow-up. Appendix C presents the results of the review of selected guiding principles of ISM. Appendix D presents the results of the review of the CBFO and WTS feedback and continuous improvement processes. The results of the review of the application of the core functions of ISM for the selected WIPP activities are discussed in Appendix E.
Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants
O, J.M.; Higgins, J.; Stephen Fleger - NRC
The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.
Continuous Improvement and the Safety Case for the Waste Isolation Pilot Plant Geologic Repository - 13467
Van Luik, Abraham; Patterson, Russell; Nelson, Roger [US Department of Energy, Carlsbad Field Office, 4021 S. National parks Highway, Carlsbad, NM 88220 (United States)] [US Department of Energy, Carlsbad Field Office, 4021 S. National parks Highway, Carlsbad, NM 88220 (United States); Leigh, Christi [Sandia National Laboratories Carlsbad Operations, 4100 S. National parks Highway, Carlsbad, NM 88220 (United States)] [Sandia National Laboratories Carlsbad Operations, 4100 S. National parks Highway, Carlsbad, NM 88220 (United States)
The Waste Isolation Pilot Plant (WIPP) is a geologic repository 2150 feet (650 m) below the surface of the Chihuahuan desert near Carlsbad, New Mexico. WIPP permanently disposes of transuranic waste from national defense programs. Every five years, the U.S. Department of Energy (DOE) submits an application to the U.S. Environmental Protection Agency (EPA) to request regulatory-compliance re-certification of the facility for another five years. Every ten years, DOE submits an application to the New Mexico Environment Department (NMED) for the renewal of its hazardous waste disposal permit. The content of the applications made by DOE to the EPA for re-certification, and to the NMED for permit-renewal, reflect any optimization changes made to the facility, with regulatory concurrence if warranted by the nature of the change. DOE points to such changes as evidence for its having taken seriously its 'continuous improvement' operations and management philosophy. Another opportunity for continuous improvement is to look at any delta that may exist between the re-certification and re-permitting cases for system safety and the consensus advice on the nature and content of a safety case as being developed and published by the Nuclear Energy Agency's Integration Group for the Safety Case (IGSC) expert group. DOE at WIPP, with the aid of its Science Advisor and teammate, Sandia National Laboratories, is in the process of discerning what can be done, in a reasonably paced and cost-conscious manner, to continually improve the case for repository safety that is being made to the two primary regulators on a recurring basis. This paper will discuss some aspects of that delta and potential paths forward to addressing them. (authors)
Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results
Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.
The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs.
Pantex Plant final safety analysis report, Zone 4 magazines. Staging or interim storage for nuclear weapons and components: Issue D
Not Available
This Safety Analysis Report (SAR) contains a detailed description and evaluation of the significant environmental, safety, and health (ES&H) issues associated with the operations of the Pantex Plant modified-Richmond and steel arch construction (SAC) magazines in Zone 4. It provides (1) an overall description of the magazines, the Pantex Plant, and its surroundings; (2) a systematic evaluations of the hazards that could occur as a result of the operations performed in these magazines; (3) descriptions and analyses of the adequacy of the measures taken to eliminate, control, or mitigate the identified hazards; and (4) analyses of potential accidents and their associated risks.
Privatization of the gaseous diffusion plants and impacts on nuclear criticality safety administration
D`Aquila, D.M.; Holliday, R.T. [Lockheed Martin Utility Services, Inc., Piketon, OH (United States); Dean, J.C. [Lockheed Martin Utility Services, Inc., Paducah, KY (United States)
The Energy Policy Act of 1992 created the United States Enrichment Corporation (USEC) on July 1, 1993. The USEC is a government-owned business that leases those Gaseous Diffusion Plant (GDP) facilities at the Portsmouth, Ohio, and Paducah, Kentucky, sites from the U.S. Department of Energy (DOE) that are required for enriching uranium. Lockheed Martin Utility Services is the operating contractor for the USEC-leased facilities. The DOE has retained use of, and regulation over, some facilities and areas at the Portsmouth and Paducah sites for managing legacy wastes and environmental restoration activities. The USEC is regulated by the DOE, but is currently changing to regulation under the U.S. Nuclear Regulatory Commission (NRC). The USEC is also preparing for privatization of the uranium enrichment enterprise. These changes have significantly affected the nuclear criticality safety (NCS) programs at the sites.
Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report. Revision 3
Herborn, D.I.
Westinghouse Hanford Company (WHC) is the Integrating Contractor for the Hanford Waste Vitrification Plant (HWVP) Project, and as such is responsible for preparation of the HWVP Preliminary Safety Analysis Report (PSAR). The HWVP PSAR was prepared pursuant to the requirements for safety analyses contained in US Department of Energy (DOE) Orders 4700.1, Project Management System (DOE 1987); 5480.5, Safety of Nuclear Facilities (DOE 1986a); 5481.lB, Safety Analysis and Review System (DOE 1986b) which was superseded by DOE order 5480-23, Nuclear Safety Analysis Reports, for nuclear facilities effective April 30, 1992 (DOE 1992); and 6430.lA, General Design Criteria (DOE 1989). The WHC procedures that, in large part, implement these DOE requirements are contained in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals.
Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants
Chang, T.Y.
Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.
PRB coal safety design considerations for new greenfield plants: an EPCC's perspective
Brown, J.H. [Fluor Power (United States)
The article reviews the design and safety aspects to consider in a new greenfield Powder River Basin (PRB) coal-fired power plant such as the 200 MW TS Power Plant (TSPP) in Nevada that Fluor is working on as an engineering, procurement and construction contractor (EPCC). PRB coals can become fragmented and form coal dust that is highly volatile and easily self-ignited. Coal handling systems incorporate features to minimise dust, such as totally enclosed chute works, 'spoon drops' to reduce impact turbulence, and overflow hoods. Conveyors have extended skirtboards and tight clearances between the wear plates and the belts. Storage piles are designed to have high compaction to deprive oxygen and dust suppression monitor hydrants to minimise dust and assist in compaction. The coal silo filling bay is designed to minimise dust once the coal is crushed, and attention is paid to cleaning and lighting. The silos are designed to ensure mass flow to the feeder and incorporate a carbon monoxide monitor and an F-500 fire suppressant. 3 photos.
Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology
Kohut, P.; Epel, L.G.; Tutu, N.K. [and others
The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.
Rewriting the standard on the functional requirements for computers used in safety systems of nuclear power plants
Matras, J.R. [Science Applications International Corp., Las Vegas, NV (United States)
Because of the rapid development of state-of-the-art computer technology, a rewrite of ANSI/IEEE-ANS-7-4.3.2-1982, {open_quotes}Application Criteria for Digital Computer Systems of Nuclear Power Generating Stations,{close_quotes} was required. This article outlines the thrust of this rewrite, which is nearing the balloting process, and identifies standards and guidelines to be used in the development of a highly reliable to be used in the development of a highly reliable computer system. The rewrite activity has been in process approximately 2 yr and is a cooperative project of the American Nuclear Society (ANS) Nuclear Power Plant Standards Committee (NUPPSCO) and the Institute of Electrical and Electronics Engineers (IEEE) Nuclear Power Engineering Committee (NPEC). Because computer technology has progressed significantly since ANSI/IEEE-ANS-7-4.3.2-1982 was issued, the rewrite was a very interesting challenge to the work group. The primary difference between the 1982 version and the rewrite is that the 1982 version addressed the quality assurance aspects of the Quality criteria, which included the integration of hardware and software and subsequent verification and validation, whereas the rewrite, being a product of IEEE Std 603-1991, {open_quotes}IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations,{close_quotes} was written to establish the minimum requirements for computer systems (hardware, software, and interfaces) used in safety systems. This article presents an introduction to the scope of the rewrite, provides a brief comment on how the work group addressed the scope, and follows with details on how the work group addressed the scope and rewrite direction. 1 fig., 1 tab.
Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11
The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.
Independent Oversight Inspection of Environment, Safety, and Health Programs at the Pantex Plant, Technical Appendices, Volume II, February 2005
Broader source: Energy.gov [DOE]
The U.S. Department of Energy (DOE) Office of Independent Oversight and Performance Assurance (OA) conducted an inspection of environment, safety, and health (ES&H) programs at the DOE Pantex Plant during January and February 2005. The inspection was performed by the OA Office of Environment, Safety and Health Evaluations. OA reports to the Director of the Office of Security and Safety Performance Assurance, who reports directly to the Secretary of Energy. This volume of the report provides four technical appendices (C through F) containing detailed results of the OA review. Appendix C provides the results of the review of the application of the core functions of ISM for Pantex Plant work activities. Appendix D presents the results of the review of NNSA, PXSO, and BWXT feedback and continuous improvement processes and management systems. Appendix E presents the results of the review of essential safety system functionality, and Appendix F presents the results of the review of safety management of the selected focus areas.
Integrated Safety Management System Phase I Verification for the Plutonium Finishing Plant (PFP) [VOL 1 & 2
U.S. Department of Energy (DOE) Policy 450.4, Safety Management System Policy commits to institutionalizing an Integrated Safety Management System (ISMS) throughout the DOE complex as a means of accomplishing its missions safely. DOE Acquisition Regulation 970.5204-2 requires that contractors manage and perform work in accordance with a documented safety management system.
Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Pilot Plant
Brown, E.S.; Homer, G.B.; Spencer, S.G.; Shaber, C.R.
The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.
Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Research and Development Plant
Brown, E.S.; Homer, G.B.; Shaber, C.R.; Thurow, T.L.
The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.
Safety Culture at the WTP White Paper: Potential Attachment for Advice on Waste Treatment Plant Safety Culture
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's PossibleRadiation Protection245C Unlimited ReleaseWelcome ton nSafeguards andSafety29/2011 Page 1
What Does Self-Assessment of Safety Culture Look Like? Discussion from the Pantex Plant Perspective
Broader source: Energy.gov [DOE]
Presenter: Dr. Suzanne Helfinstine, Staff Engineer High Reliability Operations B&W Pantex Pantex Plant
Review of nuclear power plant safety cable aging studies with recommendations for improved approaches and for future work.
Gillen, Kenneth Todd; Bernstein, Robert
Many U. S. nuclear power plants are approaching 40 years of age and there is a desire to extend their life for up to 100 total years. Safety-related cables were originally qualified for nuclear power plant applications based on IEEE Standards that were published in 1974. The qualifications involved procedures to simulate 40 years of life under ambient power plant aging conditions followed by simulated loss of coolant accident (LOCA). Over the past 35 years or so, substantial efforts were devoted to determining whether the aging assumptions allowed by the original IEEE Standards could be improved upon. These studies led to better accelerated aging methods so that more confident 40-year lifetime predictions became available. Since there is now a desire to potentially extend the life of nuclear power plants way beyond the original 40 year life, there is an interest in reviewing and critiquing the current state-of-the-art in simulating cable aging. These are two of the goals of this report where the discussion is concentrated on the progress made over the past 15 years or so and highlights the most thorough and careful published studies. An additional goal of the report is to suggest work that might prove helpful in answering some of the questions and dealing with some of the issues that still remain with respect to simulating the aging and predicting the lifetimes of safety-related cable materials.
Risk-informed public safety policy for seismic events in the vicinity of a nuclear power plant
Afolayan Jejeloye, Olubukola
Nuclear Power Plants (NPPs) are potentially vulnerable to accidents, which can either be internally or externally initiated. External events include natural events like tornadoes, hurricanes, and earthquakes. The purpose ...
Guidelines for nuclear power plant safety issue prioritization information development. Supplement 2
Andrews, W.B.; Gallucci, R.H.V.; Konzek, G.J.; Heaberlin, S.W.; Fecht, B.A.; Allen, C.H.; Allen, R.D.; Bickford, W.E., Carbaugh, E.H.; Lewis, J.R.
This is the third in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 31 issues. Each issue was considered within the constraints of available information as of summer 1983, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, A Prioritization of Generic Safety Issues.
Guidelines for nuclear power plant safety issue prioritization information development. Supplement 4
Tabatabai, A.S.; Fecht, B.A.; Powers, T.B.; Bickford, W.E.; Andrews, W.B.; Gallucci, R.H.V.; Bian, S.H.; Daling, P.M.; Eschbach, E.J.; Allen, C.H.
This is the fifth in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 23 issues. Each issue was considered within the constraints of available information as of winter 1986, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, ''A Prioritization of Generic Safety Issues.''
Review guidelines for software languages for use in nuclear power plant safety systems: Final report. Revision 1
Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S. [SoHaR, Inc., Beverly Hills, CA (United States)
Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada83 and Ada95; C and C++; International Electrochemical Commission (IEC) Standard 1131-3 Ladder Logic, Sequential Function Charts, Structured Text, and Function Block Diagrams; Pascal; and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.
Systems interaction results from the digraph matrix analysis of the Watts Bar Nuclear Power Plant high pressure safety injection systems. Volume 2
Sacks, I.J.; Ashmore, B.C.; Alesso, H.P.
Spatial and functional coupling of nuclear power plant systems that lead to interdependencies are called Systems Interactions. At present, the US Nuclear Regulatory Commission (NRC) is investigating ways of integrating a systems interactions study with existing Probabilistic Risk Assessment efforts. One approach is based on graph-theoretic methods utilizing matrix representations of logic diagrams called Digraph Matrix Analysis (DMA). The objective in this report is to demonstrate the capabilities of Digraph Matrix Analysis to model an accident sequence (including front line systems, support systems and human actions) as a continuous, well-integrated logic model in order to identify and evaluate functional systems interactions. The selected accident sequence, loss of high pressure safety injection during an S1 LOCA, was modeled and qualitative and quantitative comparisons were made to WASH 1400 aand other studies.
Iron oxide aerosol experiments in steam-air atmospheres: NSPP (Nuclear Safety Pilot Plant) tests 501-505 and 511: Data record report
Adams, R.E.; Tobias, M.L.
This data record report summarizes the results from five tests involving Fe/sub 2/O/sub 3/ test aerosol in a steam-air environment and one test in a dry air environment. This research sponsored by the US Nuclear Regulatory Commission was conducted in the Nuclear Safety Pilot Plant at the Oak Ridge National Laboratory. The purpose of this project is to provide a data base on the behavior of aerosols in containment under conditions assumed to occur in postulated LWR accident sequences; this data base will provide experimental validation of aerosol behavioral codes under development. In the report a brief description is given of each test together with the results in the form of tables and graphs. Included are data on aerosol mass concentration, aerosol fallout and plateout rates, total mass fallout and plateout, aerosol particle size, vessel atmosphere pressure, vessel atmosphere temperatures, temperature gradients near the vessel wall, and steam condensation rates on the vessel wall.
Limestone concrete aerosol experiments in steam-air atmospheres: NSPP (Nuclear Safety Pilot Plant) Tests 521, 522, and 531: Data record report
Tobias, M.L.; Adams, R.E.
This data record report summarizes the results from two tests involving limestone concrete test aerosol in a steam-air environment and one test in a dry air environment. This research sponsored by the US Nuclear Regulatory Commission was conducted in the Nuclear Safety Pilot Plant at the Oak Ridge National Laboratory. The purpose of this project is to provide a data base on the behavior of aerosols in containment under conditions assumed to occur in postulated LWR accident sequences; this data base will provide experimental validation of aerosol behavioral codes under development. In the report a brief description is given of each test together with the results in the form of tables and graphs. Included are data on aerosol mass concentration, aerosol fallout and plateout rates, total mass fallout and plateout, aerosol particle size, vessel atmosphere pressure, vessel atmosphere temperatures, temperature gradients near the vessel wall, and steam condensation rates on the vessel wall.
The Safety Data Sheet, or SDS, is written or printed material used to convey the hazards of a hazardous chemical product. It contains 16 sections of important chemical information, including
The Safety Data Sheet, or SDS, is written or printed material used to convey the hazards of a hazardous chemical product. It contains 16 sections of important chemical information, including: Chemical characteristics; Physical and health hazards, including relevant exposure limits; Precautions for safe handling
Guidelines for nuclear-power-plant safety-issue-prioritization information development
Andrews, W.B.; Gallucci, R.H.V.; Heaberlin, S.W.; Bickford, W.E.; Konzek, G.J.; Strenge, D.L.; Smith, R.I.; Weakley, S.A.
Pacific Northwest Laboratory has developed a methodology, with examples, to calculate - to an approximation serviceable for prioritization purposes - the risk, dose and cost impacts of implementing resolutions to reactor safety issues. This report is an applications guide to issue-specific calculations. A description of the approach, mathematical models, worksheets and step-by-step examples are provided. Analysis using this method is intended to provide comparable results for many issues at a cost of two staff-weeks per issue. Results will be used by the NRC to support decisions related to issue priorities in allocation of resources to complete safety issue resolutions.
Guidelines for nuclear power plant safety issue prioritization information development. Supplement 5
Daling, P.M.; Lavender, J.C. [Pacific Northwest National Lab., Richland, WA (United States)
This is the sixth in a series of reports to document the development and use of a methodology developed by the Pacific Northwest Laboratory (PNL) to calculate, for prioritization purposes, the risk, dose, and cost impacts of implementing potential resolutions to reactor safety issues (see NUREG/CR-2800, Andrews, et al., 1983). This report contains the results of issue-specific analyses for 34 generic issues. Each issue was considered within the constraints of available information at the time the issues were examined and approximately 2 staff-weeks of labor. The results are referenced as one consideration in NUREG-0933, A Prioritization of Generic Safety Issues (Emrit, et al., 1983).
COMMISSIONDECISION Small Power Plant Exemption (06-SPPE-2)
............................................................................. 14 Transmission Line Safety & Nuisance...................................................... 15 to review and license proposals to construct and operate large electric power plants, includingCOMMISSIONDECISION Small Power Plant Exemption (06-SPPE-2) Imperial County Order No: 07
Occupational Safety Review of High Technology Facilities
Lee Cadwallader
This report contains reviews of operating experiences, selected accident events, and industrial safety performance indicators that document the performance of the major US DOE magnetic fusion experiments and particle accelerators. These data are useful to form a basis for the occupational safety level at matured research facilities with known sets of safety rules and regulations. Some of the issues discussed are radiation safety, electromagnetic energy exposure events, and some of the more widespread issues of working at height, equipment fires, confined space work, electrical work, and other industrial hazards. Nuclear power plant industrial safety data are also included for comparison.
Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1
This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met.
Assessment of inservice conditions of safety-related nuclear plant structures
Ashar, H.; Bagchi, G.
The report is a compilation from a number of sources of information related to the condition Of structures and civil engineering features at operating nuclear power plants in the United States. The most significant information came from the hands-on inspection of the six old plants (licensed prior to 1977) performed by the staff of the Civil Engineering and Geosciences Branch (ECGB) in the Division of Engineering of the Office of Nuclear Reactor Regulation. For the containment structures, most of the information related to the degraded conditions came from the licensees as part of the Licensing Event Report System (10 CFR 50.73), or as part of the requirement under limiting condition of operation of the plant-specific Technical Specifications. Most of the information related to the degradation of other Structures and civil engineering features was extracted from the industry survey, the reported incidents, and the plant visits. The report discusses the condition of the structures and civil engineering features at operating nuclear power plants and provides information that would help detect, alleviate, and correct the degraded conditions of the structures and civil engineering features.
Activities in Support of Continuing the Service of Nuclear Power Plant Safety-Related Concrete Structures
Naus, Dan J [ORNL
Nuclear power plant (NPP) concrete structures are described. In-service inspection and testing requirements in the U.S. are summarized. The license renewal process in the U.S. is outlined and its current status provided. Operating experience related to performance of the concrete structures is presented. Basic components of a program to manage aging of the concrete structures are identified and described: (1) Degradation mechanisms, damage models, and material performance; (2) Assessment and remediation: i.e., component selection, in- service inspection, non-destructive examinations, and remedial actions; and (3) Estimation of performance at present or some future point in time: i.e., application of structural reliability theory to the design and optimization of in-service inspection/maintenance strategies, and determination of the effects of degradation on plant risk. Finally, areas are noted where additional research would be of benefit to aging management of nuclear power plant concrete structures.
A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants
Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)
The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Order establishes facility and programmatic safety requirements for Department of Energy facilities, which includes nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards mitigation, and the System Engineer Program. Cancels DOE O 420.1A. DOE O 420.1B Chg 1 issued 4-19-10.
Power plant including an exhaust gas recirculation system for injecting recirculated exhaust gases in the fuel and compressed air of a gas turbine engine
Anand, Ashok Kumar; Nagarjuna Reddy, Thirumala Reddy; Shaffer, Jason Brian; York, William David
A power plant is provided and includes a gas turbine engine having a combustor in which compressed gas and fuel are mixed and combusted, first and second supply lines respectively coupled to the combustor and respectively configured to supply the compressed gas and the fuel to the combustor and an exhaust gas recirculation (EGR) system to re-circulate exhaust gas produced by the gas turbine engine toward the combustor. The EGR system is coupled to the first and second supply lines and configured to combine first and second portions of the re-circulated exhaust gas with the compressed gas and the fuel at the first and second supply lines, respectively.
NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: software require- ments, safety analysis, formal, the missiles, and the digital protection systems embed- ded in nuclear power plants. Obviously, safety method SOFTWARE SAFETY ANALYSIS OF DIGITAL PROTECTION SYSTEM REQUIREMENTS USING A QUALITATIVE FORMAL
Process safety management (OSHA) and process risk management (CAA) application. Application to a coke plant
Graeser, W.C.; Mentzer, W.P.
Risk Management Programs for Chemical Accidental Release Prevention is the name of the proposed rule for the RMP Risk Management Program. The RMP was written in response to several catastrophic releases of hazardous substances. The rule is applicable to facilities that store, process or use greater than threshold quantities of 62 listed flammable chemicals and another 100 listed toxic substances. Additionally, a Risk Management Plan is registered with the EPA, Chemical Safety and Hazardous Investigation Board, state governments and the local emergency planning commission. The Clean Air Act Amendments of 1990 (specifically Section 112r) required the EPA to develop a three phase Risk Management Plan for industry: prevention program; hazard assessment; and emergency response program. The Prevention Program closely follows the OSHA`s Process Safety Management Standard. The Hazard Assessment section requires facilities to develop plans for a worst case scenario. The Emergency Response section defines the steps the facility and each employee will take if a release occurs. This section also needs to be coordinated with the Local Emergency Planning Commission. These regulations are described using Clairton Works as an example of compliance.
Preliminary systems-interaction results from the Digraph Matrix Analysis of the Watts Bar Nuclear Power Plant safety-injection systems
Sacks, I.J.; Ashmore, B.C.; Champney, J.M.; Alesso, H.P.
This report provides preliminary results generated by a Digraph Matrix Analysis (DMA) for a Systems Interaction analysis performed on the Safety Injection System of the Tennessee Valley Authority Watts Bar Nuclear Power Plant. An overview of DMA is provided along with a brief description of the computer codes used in DMA.
PGDP (Paducah Gaseous Diffusion Plant)-UF/sub 6/ handling, sampling, analysis and associated QC/QA and safety related procedures
Harris, R.L. (comp.)
This document is a compilation of Paducah Gaseous Diffusion Plant procedures on UF/sub 6/ handling, sampling, and analysis, along with associated QC/QA and safety related procedures. It was assembled for transmission by the US Department of Energy to the Korean Advanced Energy Institute as a part of the US-Korea technical exchange program.
The (safety-related) heat exchangers aging management guideline for commercial nuclear power plants, and developments since 1994
Clauss, J.M.
The US Department of Energy (DOE), in cooperation with the Electric Power Research Institute (EPRI) and US nuclear power plant utilities, is preparing a series of aging management guidelines (AMGs) for commodity types of components (e.g., heat exchangers, electrical cable and terminations, pumps). Commodities are included in this series based on their importance to continued nuclear plant operation and license renewal. The AMGs contain a detailed summary of operating history, stressors, aging mechanisms, and various types of maintenance and surveillance practices that can be combined to create an effective aging management program. Each AMG is intended for use by the systems engineers and plant maintenance staff (i.e., an AMG is intended to be a hands-on technical document rather than a licensing document). The heat exchangers AMG, published in June 1994, includes the following information of interest to nondestructive examination (NDE) personnel: aging mechanisms determined to be non-significant for all applications; aging mechanisms determined to be significant for some applications; effective conventional programs for managing aging; and effective unconventional programs for managing aging. Since the AMG on heat exchangers was published four years ago, a brief review has been conducted to identify emerging regulatory issues, if any. The results of this review and lessons learned from the collective set of AMGs are presented.
Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 14
Tam, P.S.
Supplement No. 14 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation with additional information submitted by the applicant since Supplement No. 13 was issued, and matters that the staff had under review when Supplement No. 13 was issued.
Operating experience feedback report-reliability of safety-related steam turbine-driven standby pumps used in US commerical nuclear power plants
Boardman, J.R. [Nuclear Regulatory Commission, Washington, DC (United States)
Pump failure experience is collected by two primary means: (1) Licensee Event Reports, and (2) Nuclear Plant Reliability Data System failure reports. Certain safety-related turbine-driven standby pumps were identified by these data systems as experiencing significant ongoing repetitive failures of their turbine drivers, resulting in low reliability of the pump units. The root causes of identified failures were determined, and actions to preclude these repetitive failures were identified. 5 refs., 1 tab.
Baldwin, N.D.; Spooner, K.G.; Walkden, P. [British Nuclear Group Ltd, Daresbury, Warrington (United Kingdom)
In the United Kingdom there have been significant recent changes to the management of civil nuclear liabilities. With the formation in April 2005 of the Nuclear Decommissioning Authority (NDA), ownership of the civil nuclear licensed sites in the UK, including the Magnox Reactor Stations, passed to this new organisation. The NDAs mission is to seek acceleration of the nuclear clean up programme and deliver increased value for money and, consequently, are driving their contractors to seek more innovative ways of performing work. British Nuclear Group manages the UK Magnox stations under contract to the NDA. This paper summarises the approach being taken within its Reactor Sites business to work with suppliers to enhance working arrangements at sites, improve the delivery of decommissioning programmes and deliver improvements in safety and environmental performance. The UK Magnox stations are 1. generation gas-graphite reactors, constructed in the 1950's and 1960's. Two stations are currently still operating, three are shut-down undergoing defueling and the other five are being decommissioned. Despite the distractions of industry restructuring, an uncompromising policy of demanding improved performance in conjunction with improved safety and environmental standards has been adopted. Over the past 5 years, this policy has resulted in step-changes in performance at Reactor Sites, with increased electrical output and accelerated defueling and decommissioning. The improvements in performance have been mirrored by improvements in safety (DACR of 0 at 5 sites); environmental standards (reductions in energy and water consumption, increased waste recycling) and the overall health of the workforce (20% reduction in sickness absence). These achievements have, in turn, been recognised by external bodies, resulting in several awards, including: the world's first ISRS and IERS level 10 awards (Sizewell, 2006), the NUMEX plant maintenance award (Bradwell, 2006), numerous RoSPA awards at site and sector level and nomination, at Company level, for the RoSPA George Earle trophy for outstanding performance in Health and Safety (Reactor Sites, 2006). After 'setting the scene' and describing the challenges that the company has had to respond to, the paper explains how these improvements have been delivered. Specifically it explains the process that has been followed and the parts played by sites and suppliers to deliver improved performance. With the experience of already having transitioned several Magnox stations from operations to defueling and then to decommissioning, the paper describes the valuable experience that has been gained in achieving an optimum change process and maintaining momentum. (authors)
Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980
McCormack, K.E.; Gallaher, R.B.
This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).
Preliminary safety analysis report for project 89-GEB-610 Plutonium Finishing Plant instrumentation upgrade. Revision 1
Huber, T.E.
This document consists of an analysis of the MICON system upgrade. This project shall install a Micon Co. distributed process monitor and control system with Sparc Sun workstation operator interfaces. The Sparc workstations are housed in consoles custom designed to human factors specifications. The distributed control system (DCS) shall have the installed capacity to monitor and control all related instruments and equipment presently connected to the panels in the PFP Power Control Room 321A as listed in the input/output list. This also includes all devices monitored and controlled by the 2736-ZB Allen Bradley programmable logic controller. The system has since assumed the control and monitoring responsibilities for Projects B- 680H ``Low Level Waste Treatment Facility`` and C-031H ``PFP Liquid Effluent Facilities``. Part of the new en`s change area in Building 234-5ZA, Room 712, has been remodeled to house two consoles and one supervisor console. Local control units containing the microprocontrollers and the input/output interface circuit boards shall be wired to the instrumentation and controlled equipment. These units communicate with the Sparc workstations via a redundant data communications highway and shall be strategic, throughout the PFP facility. The DCS has already been purchased from Micon Co., located in Houston Texas, presently on site.
Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1
Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.
EA-0970: Environmental Safety and Health Analytical Laboratory Project No. 94-AA-01 Pantex Plant, Amarillo, Texas
Broader source: Energy.gov [DOE]
This EA evaluates the environmental impacts of the proposal to construct and operate an Environmental Safety and Health Analytical Laboratory and subsequent demolition of the existing Analytical...
Tiger Team assessment of the Pinellas Plant
Not Available
This Document contains findings identified during the Tiger Team Compliance Assessment of the Department of Energy's (DOE's) Pinellas Plant, Pinellas County, Florida. The assessment wa directed by the Department's Office of Environment, Safety, and Health (ES H) from January 15 to February 2, 1990. The Pinellas Tiger Team Compliance Assessment is comprehensive in scope. It covers the Environment Safety and Health, and Management areas and determines the plant's compliance with applicable Federal (including DOE), State, and local regulations and requirements.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
To establish facility safety requirements for the Department of Energy, including National Nuclear Security Administration. Cancels DOE O 420.1. Canceled by DOE O 420.1B.
Nuclear reactor safety heat transfer
Jones, O.C.
Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.
Fast reactor safety program. Progress report, January-March 1980
Not Available
The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development.
Simplifying documentation while approaching site closure: integrated health & safety plans as documented safety analysis
Brown, Tulanda
At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). By isolating any remediation activities that deal with Enriched Restricted Materials, the SBRs and PRs assure that the hazard categories of former nuclear facilities undergoing remediation remain less than Nuclear. These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D&D) of over 150 structures, including six major nuclear production plants. This paper presents the FCP method for maintaining safety basis documentation, using the D&D I-HASP as an example.
Good housekeeping: Safety and order in the scientific laboratory
Sims, B. H. (Benjamin H.)
Laboratory safety might not seem, at first, to be very profoundly related to scientific knowledge. Of course safety is a relatively trivial issue in many scientific settings, especially in comparison to the kind of safety concerns found, say, at a construction site or a chemical plant. However, as scientific work has come to involve more exotic chemicals, biological organisms, and forms of radiation, and generally become more industrial in character, safety has become more of a concern. This has occurred alongside a general expansion of government regulation of workplace safety during the 20thc entury, and a recent trend toward extending work lace safety efforts to new kinds of work, including administrative and professional tasks. As a result of these trends, scientists find that they are increasingly being held responsible for following safety regulations in their re{approx}earc
Contribution to the safety assessment of instrumentation and control software for nuclear power plants: Application to SPIN N4
Soubies, B.; Henry, J.Y.; LeMeur, M.; Elsensohn, O.; Boulc`h, J.
1300 MWe pressurized water reactors (PWRs), like the 1400 MWe reactors, operate with microprocessor-based safety systems. This is particularly the case for the Digital Integrated Protection System (SPIN), which trips the reactor in an emergency and sets in action the safeguard functions. The software used in these systems must therefore be highly dependable in the execution of their functions.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.
Broader source: Energy.gov [DOE]
The Nuclear Safety Division conducts a comprehensive nuclear power plant oversight review program of the nine reactors at the five nuclear power sites in Pennsylvania. It also monitors the...
Total safety: A new safety culture to integrate nuclear safety and operational safety
Saji, G. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan); Murphy, G.A. [ed.
The creation of a complete and thorough safety culture is proposed for the purpose of providing additional assurance about nuclear safety and improving the performance of nuclear power plants. The safety philosophy developed a combination of the former hardware-oriented nuclear safety approach and recent operational safety concepts. The improvement of the latter, after TMI-2 and Chernobyl, has been proven very effective in reducing the total risk associated with nuclear power plants. The first part of this article introduces a {open_quotes}total safety{close_quotes} concept. This extends the concept of {open_quotes}nuclear safety{close_quotes} and makes it closer to the public perception of safety. This concept is defined by means of a taxonomy of total safety. The second part of the article shows that total safety can be achieved by integrating it into a modern quality assurance (QA) system since it is tailored to make implementation into a framework of QA easier. The author believes that the outstanding success experienced by various industries as a result of introducing the modern QA system should lead to its application for ensuring the safety and performance of nuclear facilities. 15 refs., 3 figs.
Proceedings: Decommissioning, Decontamination, ALARA, and Worker Safety Workshop
This workshop on decontamination, ALARA, and worker safety was the sixth in a series initiated by EPRI to aid utility personnel in assessing the technologies for decommissioning nuclear power plants. The workshop focused on specific aspects of decommissioning related to the management of worker radiation exposure and safety. The information presented will help individual utilities assess benefits of programs in these areas for their projects, including their potential to reduce decommissioning costs.
Centrifuge enrichment plants. (Latest citations from the NTIS bibliographic database). Published Search
Not Available
The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are considered. (Contains a minimum of 172 citations and includes a subject term index and title list.)
Centrifuge enrichment plants. (Latest citations from the NTIS bibliographic database). Published Search
The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are considered. (Contains a minimum of 171 citations and includes a subject term index and title list.)
Not Available
Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.
Fast reactor safety: proceedings of the international topical meeting. Volume 1
Not Available
The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.
Plant maintenance and plant life extension issue, 2007
Agnihotri, Newal (ed.)
The focus of the March-April issue is on plant maintenance and plant life extension. Major articles/reports in this issue include: Three proposed COLs expected in 2007, by Dale E. Klein, U.S. Nuclear Regulatory Commission; Delivering behaviors that our customers value, by Jack Allen, Westinghouse Electric Company; Facilitating high-level and fuel waste disposal technologies, by Malcolm Gray, IAEA, Austria; Plant life management and long-term operation, by Pal Kovacs, OECD-NEA, France; Measuring control rod position, by R. Taymanov, K. Sapozhnikova, I. Druzhinin, D.I. Mendeleyev, Institue for Metrology, Russia; and, 'Modernization' means higher safety, by Svetlana Genova, Kozluduy NPP plc, Bulgaria.
Safety of nuclear power reactors in the former Eastern European countries
Chakraborty, S. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland)
This article discusses the safety of nuclear power plants in the former Eastern European countries (including the former Soviet Union). The current international design fabrication, construction, operation, safely, regulatory standards and practices, and ways to resolve plant problems are addressed in light of experience with the Western nuclear power development programs. 9 refs., 4 figs.
Nero, A.V.
average value for nuclear plants) aFinal Envir. Statement (Statement, Koshkonong Nuclear Plant, August 1976. U. S.rem; operation of the nuclear plants themselves only *Other
Calderon, L.M., Westinghouse Hanford
Management of Nuclear Safety, Industrial Safety, Industrial Hygiene, and Fire Protection programs, functions, and field support resources for Tank Waste Remediation Systems (TWRS) has, until recently, been centralized in TWRS Safety, under the Emergency, Safety, and Quality organization. Industrial hygiene technician services were also provided to support operational needs related to safety basis compliance. Due to WHC decentralization of safety and reengineering efforts in West Tank Farms, staffing and safety responsibilities have been transferred to the facilities. Under the new structure, safety personnel for TWRS are assigned directly to East Tank Farms, West Tank Farms, and a core Safety Group in TWRS Engineering. The Characterization Project Operations (CPO) safety organization will remain in tact as it currently exists. Personnel assigned to East Tank Farms, West Tank Farms, and CPO will perform facility-specific or project-specific duties and provide field implementation of programs. Those assigned to the core group will focus on activities having a TWRS-wide or programmatic focus. Hanford-wide activities will be the responsibility of the Safety Center of Expertise. In order to ensure an effective and consistent safety program for TWRS under the new organization program functions, goals, organizational structure, roles, responsibilities, and path forward must be clearly established. The purpose of the TWRS Safety Program Plan is to define the overall safety program, responsibilities, relationships, and communication linkages for safety personnel under the new structure. In addition, issues associated with reorganization transition are addressed, including training, project ownership, records management, and dissemination of equipment. For the purpose of this document ``TWRS Safety`` refers to all safety professionals and technicians (Industrial Safety, Industrial Hygiene, Fire Protection, and Nuclear Safety) within the TWRS organization, regardless of their location in the organization.
achieving industrial safety: Topics by E-print Network
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Safety culture refers to the attitudes, behaviors, and conditions that affect safety performance and often arises in discussions following incidents at nuclear power plants. As it...
Management of National Nuclear Power Programs for assured safety
Connolly, T.J. (ed.)
Topics discussed in this report include: nuclear utility organization; before the Florida Public Service Commission in re: St. Lucie Unit No. 2 cost recovery; nuclear reliability improvement and safety operations; nuclear utility management; training of nuclear facility personnel; US experience in key areas of nuclear safety; the US Nuclear Regulatory Commission - function and process; regulatory considerations of the risk of nuclear power plants; overview of the processes of reliability and risk management; management significance of risk analysis; international and domestic institutional issues for peaceful nuclear uses; the role of the Institute of Nuclear Power Operations (INPO); and nuclear safety activities of the International Atomic Energy Agency (IAEA).
and Communications Manager (951) 827-6303 email@example.com Beiwei Tu, MS, CIH, CSP Safety and Industrial Hygiene
Evolution of Safety Basis Documentation for the Fernald Site
Brown, T.; Kohler, S.; Fisk, P.; Krach, F.; Klein, B.
The objective of the Department of Energy's (DOE) Fernald Closure Project (FCP), in suburban Cincinnati, Ohio, is to safely complete the environmental restoration of the Fernald site by 2006. Over 200 out of 220 total structures, at this DOE plant site which processed uranium ore concentrates into high-purity uranium metal products, have been safely demolished, including eight of the nine major production plants. Documented Safety Analyses (DSAs) for these facilities have gone through a process of simplification, from individual operating Safety Analysis Reports (SARs) to a single site-wide Authorization Basis containing nuclear facility Bases for Interim Operations (BIOs) to individual project Auditable Safety Records (ASRs). The final stage in DSA simplification consists of project-specific Integrated Health and Safety Plans (I-HASPs) and Nuclear Health and Safety Plans (N-HASPs) that address all aspects of safety, from the worker in the field to the safety basis requirements preserving the facility/activity hazard categorization. This paper addresses the evolution of Safety Basis Documentation (SBD), as DSAs, from production through site closure.
Centrifuge enrichment plants. January 1970-October 1988 (Citations from the NTIS data base). Report for January 1970-October 1988
Not Available
This bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socio-economic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are briefly considered. (Contains 151 citations fully indexed and including a title list.)
Health and safety plan for the removal action at the former YS-860 Firing Ranges, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee
This health and safety plan sets forth the requirements and procedures to protect the personnel involved in the Lead Source Removal Project at the Former YS-86O Firing Ranges. This project will be conducted in a manner that ensures the protection of the safety and health of workers, the public, and the environment. The purpose of this removal action is to address lead contaminated soil and reduce a potential risk to human health and the environment. This site is an operable unit within the Upper East Fork Poplar Creek watershed. The removal action will contribute to early source actions within the watershed. The project will accomplish this through the removal of lead-contaminated soil in the target areas of the two small arms firing ranges. This plan covers the removal actions at the Former YS-86O Firing Ranges. These actions involve the excavation of lead-contaminated soils, the removal of the concrete trench and macadam (asphalt) paths, verification/confirmation sampling, grading and revegetation. The primary hazards include temperature extremes, equipment operation, noise, potential lead exposure, uneven and slippery working surfaces, and insects.
Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 15
Tam, P.S.
This report supplements the Safety Evaluation Report (SER), NUREG-0847 (June 1982), Supplement No. 1 (September 1982), Supplement No. 2 (January 1984), Supplement No. 3 (January 1985), Supplement No. 4 (March 1985), Supplement No. 5 (November 1990), Supplement No. 6 (April 1991), Supplement No. 7 (September 1991), Supplement No. 8 (January 1992), Supplement No. 9 (June 1992), Supplement No. 10 (October 1992), Supplement No. 11 (April 1993), Supplement No. 12 (October 1993), Supplement No. 13 (April 1994), and Supplement No. 14 (December 1994) issued by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by the Tennessee Valley Authority, as applicant and owner, for licenses to operate the Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). The facility is located in Rhea County, Tennessee, near the Watts Bar Dam on the Tennessee River. This supplement provides recent information regarding resolution of some of the outstanding and confirmatory items, and proposed license conditions identified in the SER.
Pinellas Plant annual site environmental report for calendar year 1995
Lockheed Martin Specialty Components, Inc., and the US Department of Energy are committed to successfully administering a high-quality Environmental, Safety and Health Program at the Pinellas Plant in Pinellas County, Florida. Part of this commitment includes accurately documenting and communicating to the Pinellas Plant stakeholders the results of the Pinellas Plant`s environmental compliance and monitoring activities. The Annual Site Environmental Report presents a comprehensive summary of the results of the Environmental Monitoring, Waste Management, and Environmental Restoration Programs at the Pinellas Plant for 1995. This report also includes the plant`s performance in the areas of compliance with applicable regulatory requirements and standards and identifies major Environmental, Safety and Health Program initiatives and accomplishments for 1995. As a result of the end of the Department of Energy`s Defense Programs mission (weapons production) on September 30, 1994, considerable changes at the Pinellas Plant are occurring. The Department of Energy`s Environmental Management is now the landlord of the Pinellas Plant to facilitate the plant`s new mission of transition to alternate use in support of economic development and safe shutdown. The Department of Energy sold the Pinellas Plant to the Pinellas County Industry Council in March 1995, and it is leasing back a portion of the plant through September 1997, to complete the safe shutdown and transition activities.
Kessler, S
With the implementation of DOE Order 420.1B, Facility Safety, and DOE-STD-3007-2007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities', a new requirement was imposed that all criticality safety controls be evaluated for inclusion in the facility Documented Safety Analysis (DSA) and that the evaluation process be documented in the site Criticality Safety Program Description Document (CSPDD). At the Hanford site in Washington State the CSPDD, HNF-31695, 'General Description of the FH Criticality Safety Program', requires each facility develop a linking document called a Criticality Control Review (CCR) to document performance of these evaluations. Chapter 5, Appendix 5B of HNF-7098, Criticality Safety Program, provided an example of a format for a CCR that could be used in lieu of each facility developing its own CCR. Since the Plutonium Finishing Plant (PFP) is presently undergoing Deactivation and Decommissioning (D&D), new procedures are being developed for cleanout of equipment and systems that have not been operated in years. Existing Criticality Safety Evaluations (CSE) are revised, or new ones written, to develop the controls required to support D&D activities. Other Hanford facilities, including PFP, had difficulty using the basic CCR out of HNF-7098 when first implemented. Interpretation of the new guidelines indicated that many of the controls needed to be elevated to TSR level controls. Criterion 2 of the standard, requiring that the consequence of a criticality be examined for establishing the classification of a control, was not addressed. Upon in-depth review by PFP Criticality Safety staff, it was not clear that the programmatic interpretation of criterion 8C could be applied at PFP. Therefore, the PFP Criticality Safety staff decided to write their own CCR. The PFP CCR provides additional guidance for the evaluation team to use by clarifying the evaluation criteria in DOE-STD-3007-2007. In reviewing documents used in classifying controls for Nuclear Safety, it was noted that DOE-HDBK-1188, 'Glossary of Environment, Health, and Safety Terms', defines an Administrative Control (AC) in terms that are different than typically used in Criticality Safety. As part of this CCR, a new term, Criticality Administrative Control (CAC) was defined to clarify the difference between an AC used for criticality safety and an AC used for nuclear safety. In Nuclear Safety terms, an AC is a provision relating to organization and management, procedures, recordkeeping, assessment, and reporting necessary to ensure safe operation of a facility. A CAC was defined as an administrative control derived in a criticality safety analysis that is implemented to ensure double contingency. According to criterion 2 of Section IV, 'Linkage to the Documented Safety Analysis', of DOESTD-3007-2007, the consequence of a criticality should be examined for the purposes of classifying the significance of a control or component. HNF-PRO-700, 'Safety Basis Development', provides control selection criteria based on consequence and risk that may be used in the development of a Criticality Safety Evaluation (CSE) to establish the classification of a component as a design feature, as safety class or safety significant, i.e., an Engineered Safety Feature (ESF), or as equipment important to safety; or merely provides defense-in-depth. Similar logic is applied to the CACs. Criterion 8C of DOE-STD-3007-2007, as written, added to the confusion of using the basic CCR from HNF-7098. The PFP CCR attempts to clarify this criterion by revising it to say 'Programmatic commitments or general references to control philosophy (e.g., mass control or spacing control or concentration control as an overall control strategy for the process without specific quantification of individual limits) is included in the PFP DSA'. Table 1 shows the PFP methodology for evaluating CACs. This evaluation process has been in use since February of 2008 and has proven to be simple and effective. Each control identified i
Nero, A.V.
in U. S. Conunercial Nuclear Power Plants", Report WASH-Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"
Nero, A.V.
Report LBL-5287. "Power Plant Reliability-Availability andConunercial Nuclear Power Plants", Report WASH-1400 (NUREG-Standards for Nuclear Power Plants," by A.V. Nero and Y.C.
Szilard, Ronaldo H; Smith, Curtis L
SUMMARY In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: 1. A value proposition (�why is this important?�) that will make the case for stakeholder�s use of the ASAP research and development (R&D) products. 2. An identification of likely end users and pathway to adoption of enhanced tools by the end-users. 3. A proposed set of practical and achievable �use case� demonstrations. 4. A proposed plan to address ASAP verification and validation (V&V) needs. 5. A proposed schedule for the multi-year ASAP.
National Renewable Energy Laboratory Hydrogen Safety Sensor Development
-of-plant components such as air compressors, and sensors and controls." #12;Safety Sensor Performance Goals
Recommended practice for fire protection for electric generating plants and high voltage direct current converter stations. 2005 ed.
The standard outlines fire safety recommendations for gas, oil, coal, and alternative fuel electric generating plants including high voltage direct current converter stations and combustion turbine units greater than 7500 hp used for electric generation. Provisions apply to both new and existing plants. The document provides fire prevention and fire protection recommendations for the: safety of construction and operating personnel; physical integrity of plant components; and continuity of plant operations. The 2005 edition includes revisions and new art that clarify existing provisions. 5 annexes.
Sessions include: Beginning Farmer and Rancher
Watson, Craig A.
Sessions include: � Beginning Farmer and Rancher � New Markets and Regulations � Food Safety � Good Bug, Bad Bug ID � Horticulture � Hydroponics � Livestock and Pastured Poultry � Mushrooms � Organic � Live animal exhibits � Saturday evening social, and � Local foods Florida Small Farms and Alternative
Investigation of the radiological safety concerns and medical history of the late Joseph T. Harding, former employee of the Paducah Gaseous Diffusion Plant
Vallario, E J
An ex-employee's claims that inadequate enforcement of radiation safety regulations allowed excess radiation exposure thereby causing his deteriorating health was not substantiated by a thorough investigation.
Westar's Lawrence Energy Center wins for not blinking on safety
Peltier, R.
It took Westar Energy eight years to upgrade the Lawrence Energy Center to burn Powder River Basin coal. Its zero lost-time accident record during the eight-year, million-man-hour project is a testament to Westar's commitment to workplace safety. The plant won the Powder River Basin Coal Users' Group plant of the year award for 2006. The article describes all the changes implemented at the plant, including replacing and upgrading controls for the belt conveyor, replacing the coal crushers, minimising dust and modifying coal bunkers, to cope with the increased volatility of Powder River Basin coal. Modifications were made to minimise slagging and fouling of boilers. 10 photos.
Quality Assurance Specialist (Safety Software Qualtiy Assurance)
Broader source: Energy.gov [DOE]
A successful candidate in this position performs safety software quality assurance oversight of all activities that support the Waste Isolation Pilot Plant (WIPP) compliance with environmental,...
Five-megawatt geothermal-power pilot-plant project
Not Available
This is a report on the Raft River Geothermal-Power Pilot-Plant Project (Geothermal Plant), located near Malta, Idaho; the review took place between July 20 and July 27, 1979. The Geothermal Plant is part of the Department of Energy's (DOE) overall effort to help commercialize the operation of electric power plants using geothermal energy sources. Numerous reasons were found to commend management for its achievements on the project. Some of these are highlighted, including: (a) a well-qualified and professional management team; (b) effective cost control, performance, and project scheduling; and (c) an effective and efficient quality-assurance program. Problem areas delineated, along with recommendations for solution, include: (1) project planning; (2) facility design; (3) facility construction costs; (4) geothermal resource; (5) drilling program; (6) two facility construction safety hazards; and (7) health and safety program. Appendices include comments from the Assistant Secretary for Resource Applications, the Controller, and the Acting Deputy Director, Procurement and Contracts Management.
Integrated Safety Management Policy
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The policy establishes DOE's expectation for safety, including integrated safety management that will enable the Department�s mission goals to be accomplished efficiently while ensuring safe operations at all departmental facilities and activities. Cancels DOE P 411.1, DOE P 441.1, DOE P 450.2A, DOE P 450.4, and DOE P 450.7
Foote Hydroelectric Plant spillway rehabilitation
Sowers, D.L. [Consumers Power Co., Jackson, MI (United States); Hasan, N.; Gertler, L.R. [Raytheon Infrastructures Services, New York, NY (United States)
In 1993 the spillway of the 9 MW Foote Hydroelectric Plant located on the AuSable River, near Oscoda, Michigan was rehabilitated. The Foote Plant, built in 1917, is owned and operated by Consumers Power Company. In the 76 years of continuous operation the spillway had deteriorated such that much of the concrete and associated structure needed to be replaced to assure safety of the structure. The hydro station includes an earth embankment with concrete corewall, a concrete spillway with three tainter gates and a log chute, a penstock structure and a steel and masonry powerhouse. The electric generation is by three vertical shaft units of 3,000 KW each. A plan of the plant with spillway and an elevation of the spillway section is shown. This paper describes the evaluation and repair of the plant spillway and associated structure.
Safety assured financial evaluation of maintenance
Erguina, Vera
Management decisions in complex industrial facilities usually consider both the economic and environmental aspects of the plant's performance. For nuclear power plants (NPPs), safety is also a very substantial issue. The objectives...
An Overview of the Safety Case for Small Modular Reactors
Ingersoll, Daniel T [ORNL] [ORNL
Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.
Radioisotope thermoelectric generator reliability and safety
Campbell, R.; Klein, J.
There are numerous occasions when a planetary mission requires energy in remote areas of the solar system. Anytime power is required much beyond Mars or the Asteroid Belts, solar power is not an option. The radioisotope thermoelectric generator (RTG) was developed for such a mission requirement. It is a relatively small and lightweight power source that can produce power under adverse conditions. Just this type of source has become the backbone of the power system for far outer plant exploration. Voyagers I and II are utilizing RTGs, which will soon power the Galileo spacecraft to Jupiter and the Ulysses spacecraft to study the solar poles. The paper discusses RTG operation including thermoelectric design, converter design, general-purpose heat source; RTG reliability including design, testing, experience, and launch approval; and RTG safety issues and methods of ensuring safety.
Safety First Safety Last Safety Always Personal fall-protection systems include a body harness (safe-
Minnesota, University of
did it that way doesn't make it right. Personal fall-protection equipment is the only thing between, general requirements 1910.269 Electric power generation, transmission, and distribution 1915.159 Personal
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The objective of this Order is to establish facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. The Order has Change 1 dated 11-16-95, Change 2 dated 10-24-96, and the latest Change 3 dated 11-22-00 incorporated. The latest change satisfies a commitment made to the Defense Nuclear Facilities Safety Board (DNFSB) in response to DNFSB recommendation 97-2, Criticality Safety.
Shaw, Bryan W.; Garcia, Monica L.
Protect yourself and your family from the deadly effects of carbon monoxide--a colorless, odorless poisonous gas. This publication describes the warning signs of carbon monoxide exposure and includes a home safety checklist....
Chapter 13 Employee Health and Safety 13.01 Safety Policy and Accident Reporting
Sheridan, Jennifer
Chapter 13 Employee Health and Safety 13.01 Safety Policy and Accident Reporting General Safety. If an accident occurs, this responsibility includes making an adequate investigation and taking necessary are responsible for following established safety procedures and using protective equipment. Safety and Accident
Safety First Safety Last Safety Always Summer in Minnesota means high humidity and sunny, hot
Minnesota, University of
Safety First Safety Last Safety Always Summer in Minnesota means high humidity and sunny, hot days. · Heat stroke is life threatening! Symptoms include high body temperature, red and dry skin, rapid before you get thirsty. Adequate fluid intake is the biggest key. Cool (not ice cold) water is the best
Preliminary design of the Carrisa Plains solar central receiver power plant. Volume III, Book 3. Appendices. Part 2
Lee, S. K.
The auxiliary heat transport systems of the Carrisa Plains Solar Power Plant (CPSPP) comprise facilities which are used to support plant operation and provide plant safety and maintenance. The facilities are the sodium purification system, argon cover gas system, sodium receiving and filling system, sodium-water reaction product receiving system, and safety and maintenance equipment. The functions of the facilities of the auxiliary system are described. Design requirements are established based on plant operating parameters. Descriptions are given on the system which will be adequate to perform the function and satisfy the requirements. Valve and equipment lists are included in the appendix.
Review of Nuclear Safety Culture at the Hanford Site Waste Treatment...
Office of Environmental Management (EM)
Review of Nuclear Safety Culture at the Hanford Site Waste Treatment and Immobilization Plant Project, October 2010 Review of Nuclear Safety Culture at the Hanford Site Waste...
Safety Analysis: Evaluation of Accident Risks in the Transporation of Hazardous Materials by Truck and Rail at the Savannah River Plant
Blanchard, A.
This report presents an analysis of the consequences and risks of accidents resulting from hazardous material transportation at the Savannah River Plant.
Centrifuge enrichment plants. (Latest citations from the NTIS data base). Published Search
Not Available
The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are briefly considered. (Contains a minimum of 169 citations and includes a subject term index and title list.)
Brain surgery breathes new life into aging plants
Makansi, J. [Pearl Street Inc. (United States)
Unlike managing the human aging process, extending the life of a power plant often includes brain surgery, modernizing its control and automation system. Lately, such retrofits range from wholesale replacing of existing controls to the addition of specific control elements that help optimize performance. Pending revisions to safety codes and cybersecurity issues also need to be considered. 4 figs.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and the System Engineer Program.Chg 1 incorporates the use of DOE-STD-1189-2008, Integration of Safety into the Design Process, mandatory for Hazard Category 1, 2 and 3 nuclear facilities. Cancels DOE O 420.1A.
Broader source: Energy.gov [DOE]
The DOE's Biological Safety Program provides a forum for the exchange of best practices, lessons learned, and guidance in the area of biological safety. This content is supported by the Biosurety Executive Team. The Biosurety Executive Team is a DOE-chartered group. The DOE Office of Worker Safety and Health Policy provides administrative support for this group. The group identifies biological safety-related issues of concern to the DOE and pursues solutions to issues identified.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. Cancels DOE 5480.7A, DOE 5480.24, DOE 5480.28 and Division 13 of DOE 6430.1A. Canceled by DOE O 420.1A.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. Cancels DOE O 420.1B, DOE G 420.1-2 and DOE G 420.1-3.
Independent Oversight Inspection, Waste Isolation Pilot Plant, Summary Report- August 2002
Broader source: Energy.gov [DOE]
Inspection of Environment, Safety, and Health and Emergency Management at the Waste Isolation Pilot Plant
Electric power monthly, September 1990. [Glossary included
Not Available
The purpose of this report is to provide energy decision makers with accurate and timely information that may be used in forming various perspectives on electric issues. The power plants considered include coal, petroleum, natural gas, hydroelectric, and nuclear power plants. Data are presented for power generation, fuel consumption, fuel receipts and cost, sales of electricity, and unusual occurrences at power plants. Data are compared at the national, Census division, and state levels. 4 figs., 52 tabs. (CK)
Verifying Therapy Safety Interlock System with SPIN Motlatsi Seotsanyana1
Geldenhuys, Jaco
, air traffic and space vehicle control systems, nuclear power plant safety control systems, medical as banking systems) to embedded environments (such as nuclear power plant safety control systemsVerifying Therapy Safety Interlock System with SPIN Motlatsi Seotsanyana1 , Jaco Geldenhuys2 1
Safety concept while managing radioactive waste formed as a result of decontamination measures at territories of Ukraine after Chernobyl power plant accident
Karamushka, V.P.; Chukhin, S.G.; Ostroborodov, V.V. [VNIPIPROMTECHNOLOGII, Moscow (Russian Federation)
This article addresses the decision-making processes that are involved in the management of radioactive wastes that were created as a result of the Chernobyl reactor accident. The authors propose a systematic approach to reach this goal. Radiation safety must be provided in order to provide protection for man and the environment at the present time and in the future.
Basic TRUEX process for Rocky Flats Plant
Leonard, R.A.; Chamberlain, D.B.; Dow, J.A.; Farley, S.E.; Nunez, L.; Regalbuto, M.C.; Vandegrift, G.F.
The Generic TRUEX Model was used to develop a TRUEX process flowsheet for recovering the transuranics (Pu, Am) from a nitrate waste stream at Rocky Flats Plant. The process was designed so that it is relatively insensitive to changes in process feed concentrations and flow rates. Related issues are considered, including solvent losses, feed analysis requirements, safety, and interaction with an evaporator system for nitric acid recycle.
Safety Activities on Safety-Critical Software for Reactor Protection System Gee-Yong Park1
Jee, Eunkyoung
of nuclear power plants. The various techniques applied to a safety analysis on the structures and systemsSafety Activities on Safety-Critical Software for Reactor Protection System Gee-Yong Park1 , Kee Institute, P.O.Box 105 Yuseong, Daejeon, 305-353 KOREA 2: Korea Advanced Institute of Science and Technology
Yen, W.W.S.
EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSINGEmergency Planning for Nuclear Power Plants Determination ofproposed nuclear power plants . . . . . . . . . � . . . .
WIPP-025, Rev. 0 Summary of Nuclear Criticality Safety
at the Waste Isolation Pilot Plant #12;SUMMARY OF NUCLEAR CRITICALITY SAFETY EVALUATION FOR SHIELDED CONTAINERS PLANT WIPP-025, REV. 0 AUGUST 2009 Summary of Nuclear Criticality Safety Evaluation for Shielded ISOLATION PILOT PLANT WIPP-025, REV. 0 AUGUST 2009 ES-1 Executive Summary This report summarizes the nuclear
Department of Energy Job Safety and Health Poster
Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site
Job Safety & Health It's the law EMPLOYEES And their Representatives have a right to: * Participate in the worksite safety and health program, including inspections of the site; *...
Internship Contract (Includes Practicum)
Thaxton, Christopher S.
Internship Contract (Includes Practicum) Student's name-mail: _________________________________________ Internship Agency Contact Agency Name: ____________________________________ Address-mail: __________________________________________ Location of Internship, if different from Agency: ________________________________________________ Copies
Pinellas Plant Annual Site Environmental Report for calendar year 1994
This report presents a comprehensive summary of the results of the Environmental Monitoring, Waste Management, and Environmental Restoration Programs at the Pinellas Plant, in Pinellas County, Florida for 1994. This report also includes the plant`s performance in the areas of compliance with applicable regulatory requirements and standards and identifies major Environmental, Safety and Health Program initiatives and accomplishments for 1994. As a result of the end of Department of Energy Defense Programs mission production on September 30, 1994, considerable changes at the Pinellas Plant occurred. These changes, which included transitioning the plant toward alternate use in support of economic development and safe shutdown, both increased and heightened Environmental, Safety and Health responsibilities. In December 1994, the Department of Energy announced it had reached an agreement to sell the Pinellas Plant to the Pinellas County Industry Council in March 1995. The plant is being leased back by the Department of Energy through September 1997 to complete safe shutdown, reconfiguration, transfer of equipment to other Department of Energy production facilities, and transition to commercial ventures. Permit modifications and transfers will be completed during 1995 to reflect the new ownership by the Pinellas County Industry Council and to include new tenants as needed.
Pump apparatus including deconsolidator
Sonwane, Chandrashekhar; Saunders, Timothy; Fitzsimmons, Mark Andrew
A pump apparatus includes a particulate pump that defines a passage that extends from an inlet to an outlet. A duct is in flow communication with the outlet. The duct includes a deconsolidator configured to fragment particle agglomerates received from the passage.
Structure/piping sensitivity studies for the US NRC Seismic Safety Margins Research Program. [PWR; BWR
Shieh, L.C.; O'Connell, W.J.; Johnson, J.J.
The Seismic Safety Margins Research Program (SSMRP) is a NRC-funded, multi-year program conducted by Lawrence Livermore National Laboratory (LLNL). One of the goals of the program is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis design process and explicitly includes the uncertainties inherent in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I, a probabilistic computational procedure was developed for the seismic safety assessment. In Phase II, sensitivity studies were performed, codes and models were improved, and analysis of the Zion plant was completed.
Nuclear power plant control room operator control and monitoring tasks
Bovell, C.R.; Beck, M.G. [Concord Associates, Inc., Knoxville, TN (United States); Carter, R.J. [Oak Ridge National Labs., TN (United States)
Oak Ridge National Laboratory is conducting a research project the purpose of which is to develop the technical bases for regulatory review criteria for use in evaluating the safety implications of human factors associated with the use of artificial intelligence and expert systems, and with advanced instrumentation and control (I and C) systems in nuclear power plants (NPP). This report documents the results from Task 8 of that project. The primary objectives of the task was to identify the scope and type of control and monitoring tasks now performed by control-room operators. Another purpose was to address the types of controls and safety systems needed to operate the nuclear plant. The final objective of Task 8 was to identify and categorize the type of information and displays/indicators required to monitor the performance of the control and safety systems. This report also discusses state-of-the-art controls and advanced display devices which will be available for use in control-room retrofits and in control room of future plants. The fundamental types of control and monitoring tasks currently conducted by operators can be divided into four classifications: function monitoring tasks, control manipulation tasks, fault diagnostic tasks, and administrative tasks. There are three general types of controls used in today`s NPPs, switches, pushbuttons, and analog controllers. Plant I and C systems include components to achieve a number of safety-related functions: measuring critical plant parameters, controlling critical plant parameters within safety limits, and automatically actuating protective devices if safe limits are exceeded. The types of information monitored by the control-room operators consist of the following parameters: pressure, fluid flow and level, neutron flux, temperature, component status, water chemistry, electrical, and process and area radiation. The basic types of monitoring devices common to nearly all NPP control rooms include: analog meters, graphic recorders, digital displays and counters, light indicators, visual and audio alarms, and cathode-ray tubes.
Molecular mechanisms of the plant heat stress response
Qu, Ai-Li; Ding, Yan-Fei; Jiang, Qiong [China Jiliang University, Xueyuan Road 258, Hangzhou 310018 (China)] [China Jiliang University, Xueyuan Road 258, Hangzhou 310018 (China); Zhu, Cheng, E-mail: email@example.com [China Jiliang University, Xueyuan Road 258, Hangzhou 310018 (China)] [China Jiliang University, Xueyuan Road 258, Hangzhou 310018 (China)
Highlights: ? This review elaborates the response networks of heat stress in plants. ? It elaborates proteins responding to heat stress in special physiological period. ? The proteins and pathways have formed a basic network of the heat stress response. ? Achievements of the various technologies are also combined. -- Abstract: High temperature has become a global concern, which seriously affects the growth and production of plants, particularly crops. Thus, the molecular mechanism of the heat stress response and breeding of heat-tolerant plants is necessary to protect food production and ensure crop safety. This review elaborates on the response networks of heat stress in plants, including the Hsf and Hsp response pathways, the response of ROS and the network of the hormones. In addition, the production of heat stress response elements during particular physiological periods of the plant is described. We also discuss the existing problems and future prospects concerning the molecular mechanisms of the heat stress response in plants.
Methanol production from eucalyptus wood chips. Attachment IV. Health and safety aspects of the eucalypt biomass to methanol energy system
Fishkind, H.H.
The basic eucalyptus-to-methanol energy process is described and possible health and safety risks are identified at all steps of the process. The toxicology and treatment for exposure to these substances are described and mitigating measures are proposed. The health and safety impacts and risks of the wood gasification/methanol synthesis system are compared to those of the coal liquefaction and conversion system. The scope of this report includes the health and safety risks of workers (1) in the laboratory and greenhouse, where eucalyptus seedlings are developed, (2) at the biomass plantation, where these seedlings are planted and mature trees harvested, (3) transporting these logs and chips to the refinery, (4) in the hammermill, where the logs and chips will be reduced to small particles, (5) in the methanol synthesis plant, where the wood particles will be converted to methanol, and (6) transporting and dispensing the methanol. Finally, the health and safety risks of consumers using methanol is discussed.
Fire Safety Committee Meeting Minutes- May 2014
Broader source: Energy.gov [DOE]
DOE Fire Safety Committee Meeting Minutes, May, 2014 Topics included discussions on Fire modeling, revisions to DOE regulations and other important items relating to DOE and Fire Safety Community.
Living Expenses (includes approximately
Maroncelli, Mark
& engineering programs All other programs Graduate: MBA/INFSY at Erie & Harrisburg (12 credits) Business Guarantee 3 (Does not include Dependents Costs4 ) Altoona, Berks, Erie, and Harrisburg 12-Month Estimated
Safety evaluation for packaging (onsite) product removal can containers
Boettger, J.S.
This safety evaluation for packaging allows the transport of nine Product Removal (PR) Cans with their Containers from the PUREX Facility to the Plutonium Finishing Plant.
FAQS Qualification Card � Criticality Safety
Broader source: Energy.gov [DOE]
A key element for the Department�s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).
LASER SAFETY POLICY Policy Statement
Vertes, Akos
LASER SAFETY POLICY Policy Statement Each department that acquires or operates lasers for use in the university's Laser Safety Manual. Reason for Policy/Purpose If improperly used or controlled, lasers can produce multiple injuries, including burns, blindness, and electrocution. This policy and the university
Houston, Paul L.
LASER SAFETY POLICY MANUAL ISSUED BY ENVIRONMENTAL HEALTH & SAFETY OFFICE OF RADIOLOGICAL SAFETY and GEORGIA TECH LASER SAFETY COMMITTEE July 1, 2010 Revised July 31, 2012 #12;Laser Safety Program 1-1 #12;Laser Safety Policy Manual TABLE OF CONTENTS 1. POLICY AND SCOPE
Maroncelli, Mark
ENVIRONMENTAL HEALTH AND SAFETY GENERAL SAFETY MANUAL May 10, 2002 #12;i Acknowledgements Environmental Health and Safety gratefully acknowledges the assistance provided by the University Safety Council extremely helpful. #12;ii Environmental Health and Safety General Safety Manual Table of Contents Section
Broader source: Energy.gov (indexed) [DOE]
in the documented safety analysis. BACKGROUND On March 11 , 2011 , the Fukushima Daiichi nuclear power station in Japan was damaged by a magnitude 9.0 earthquake and the...
Next Generation Nuclear Plant Defense-in-Depth Approach
Edward G. Wallace; Karl N. Fleming; Edward M. Burns
The purpose of this paper is to (1) document the definition of defense-in-depth and the pproach that will be used to assure that its principles are satisfied for the NGNP project and (2) identify the specific questions proposed for preapplication discussions with the NRC. Defense-in-depth is a safety philosophy in which multiple lines of defense and conservative design and evaluation methods are applied to assure the safety of the public. The philosophy is also intended to deliver a design that is tolerant to uncertainties in knowledge of plant behavior, component reliability or operator performance that might compromise safety. This paper includes a review of the regulatory foundation for defense-in-depth, a definition of defense-in-depth that is appropriate for advanced reactor designs based on High Temperature Gas-cooled Reactor (HTGR) technology, and an explanation of how this safety philosophy is achieved in the NGNP.
Enterprise Assessments Review, Waste Isolation Pilot Plant -...
Office of Environmental Management (EM)
December, 2014 Review of the Waste Isolation Pilot Plant Conduct of Maintenance Recovery Plan The Office of Nuclear Safety and Environmental Assessments, within the U.S. Department...
Enterprise Assessments Review, Waste Isolation Pilot Plant -...
December 2014 Review of the Waste Isolation Pilot Plant Recovery Plan for Operating Diesel Equipment with Available Underground Airflows. The Office of Nuclear Safety and...
Independent Oversight Review, Portsmouth Gaseous Diffusion Plant...
Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site
April 2013 Independent Oversight Review, Portsmouth Gaseous Diffusion Plant - April 2013 April 2013 Review of the Integrated Safety Management System Phase I Verification Review at...
Hanford Waste Treatment Plant Construction Quality Review
Broader source: Energy.gov (indexed) [DOE]
Safety and Health Evaluations Activity Report for the Hanford Waste Treatment Plant Construction Quality Review Dates of Activity 02142011 - 02172011 Report Preparer Joseph...
Packaging and Transportation Safety
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The order establishes safety requirements for the proper packaging and transportation of DOE, including NNSA, offsite shipments and onsite transfers of radioactive and other hazardous materials and for modal transportation. Cancels DOE O 460.1B, 5-14-10
Yen, W.W.S.
DENSITIES AROUND CALIFORNIA NUCLEAR POWER PLANT. le Iil _. .AROUND CALIFORNIA NUCLEAR POWER PLANTS Miles San OnofreIN CALIFORNIA The California Nuclear Power Plant Emergency
Nero, A.V.
of electric generating plants usefully begins with anmatters, a plant's position within the generating networkthe plant may be divided into a steam generating system and
Yen, W.W.S.
from the Rancho Seco nuclear plant was simulated, A total ofdistributions around the nuclear plant sites based on thegrowth surrounding nuclear plants after the issuance of the
Nero, A.V.
Removal Equipment (nuclear plant) Turbine Building ClosedCooling Water System (nuclear plant) SteamReheater (nuclear plant) Inspection Water Induction
Nero, jA.V.
surrounding a nuclear plant, and they are stronglylocation for a nuclear plant, but it is the measures thatand consequences of nuclear plant accidents and would match
Yen, W.W.S.
S. Commercial Nuclear Power Plants. WASH-1400. October 1975.Content of for Nuclear Power Plants. Regulatory Guide 1.101.PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSING PROCESS
Safety study application guide. Safety Analysis Report Update Program
Not Available
Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly {open_quotes}low{close_quotes}) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, {open_quotes}Technical Safety Requirements,{close_quotes} and 5480.23, {open_quotes}Nuclear Safety Analysis Reports.{close_quotes} A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis.
Nuclear explosive safety study process
Nuclear explosives by their design and intended use require collocation of high explosives and fissile material. The design agencies are responsible for designing safety into the nuclear explosive and processes involving the nuclear explosive. The methodology for ensuring safety consists of independent review processes that include the national laboratories, Operations Offices, Headquarters, and responsible Area Offices and operating contractors with expertise in nuclear explosive safety. A NES Study is an evaluation of the adequacy of positive measures to minimize the possibility of an inadvertent or deliberate unauthorized nuclear detonation, high explosive detonation or deflagration, fire, or fissile material dispersal from the pit. The Nuclear Explosive Safety Study Group (NESSG) evaluates nuclear explosive operations against the Nuclear Explosive Safety Standards specified in DOE O 452.2 using systematic evaluation techniques. These Safety Standards must be satisfied for nuclear explosive operations.
State-of-the-Art of Non-Destructive Testing Methods and Technologies for Application to Nuclear Power Plant Safety-Related Concrete Structures
Wiggenhauser, Dr. Herbert [Federal Institute for Materials Research and Testing (BAM)] [Federal Institute for Materials Research and Testing (BAM); Naus, Dan J [ORNL] [ORNL
The inspection of nuclear power plant concrete structures presents challenges different from conventional civil engineering structures. Wall thicknesses can be in excess of one meter and the structures often have increased steel reinforcement density with more complex detailing. The accessibility for any testing method may be limited due to the presence of liners and other components and there can be a number of penetrations or cast-in-place items present. The objective of the report is to present the state-of-the art of non-destructive testing methods and technologies for the inspection of thick, heavily-reinforced nuclear power plant concrete cross-sections with particular respect to: locating steel reinforcement and identification of its cover depth locating tendon ducts and identification of the condition of the grout materials detection of cracking, voids, delamination, and honeycombing in concrete structures detection of inclusions of different materials or voids adjacent to the concrete side of the containment liner methods capable of identification of corrosion occurrence on the concrete side of the containment liner
State-of-the-art of non-destructive testing methods and technologies for application to nuclear power plant safety-related concrete structures
Wiggenhauser, Herbert [BAM - Federal Institute for Materials Research and Testing, Berlin, D (Germany); Naus, Dan J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)
The inspection of nuclear power plant concrete structures presents challenges different from conventional civil engineering structures. Wall thicknesses can be in excess of one meter and the structures often have increased steel reinforcement density with more complex detailing. The accessibility for any testing method may be limited due to the presence of liners and other components and there can be a number of penetrations or cast-in-place items present. The objective of the report is to present the state-of-the art of non-destructive testing methods and technologies for the inspection of thick, heavily-reinforced nuclear power plant concrete cross-sections with particular respect to: �locating steel reinforcement and identification of its cover depth �locating tendon ducts and identification of the condition of the grout materials �detection of cracking, voids, delamination, and honeycombing in concrete structures �detection of inclusions of different materials or voids adjacent to the concrete side of the containment liner �methods capable of identification of corrosion occurrence on the concrete side of the containment liner.
Nuclear criticality safety engineer qualification program utilizing SAT
Baltimore, C.J.; Dean, J.C.; Henson, T.L. [Lockheed Martin Utility Services, Inc., Paducah, KY (United States)
As part of the privatization process of the U.S. uranium enrichment plants, the Paducah Gaseous Diffusion Plant (PGDP) and the Portsmouth Gaseous Diffusion Plant (PORTS) have been in transition from U.S. Department of Energy (DOE) regulatory oversight to U.S. Nuclear Regulatory Commission (NRC) oversight since July 1993. One of the focus areas of this transition has been training and qualification of plant personnel who perform tasks important to nuclear safety, such as nuclear criticality safety (NCS) engineers.
Interaction analysis method for the Hanford Waste Vitrification Plant
Grant, P.R.; Deshotels, R.L. [Fluor Daniel, Inc., Irvine, CA (United States); Van Katwijk, C. [Westinghouse Hanford Co., Richland, WA (United States)
In order to anticipate potential problems as early as possible during the design effort, a method for interaction analysis was developed to meet the specific hazards of the Hanford Waste Vitrification Plant (HWVP). The requirement for interaction analysis is given in DOE Order 6430.1B and DOE-STD-1021-92. The purpose of the interaction analysis is to ensure that non-safety class items will not fail in a manner that will adversely affect the ability of any safety class item to perform its safety function. In the HWVP there are few structures, equipment, or controls that are safety class. In addition to damage due to failure of non-safety class items as a result of natural phenomena, threats to HWVP safety class items include the following: room flooding from firewater, leakage of chemically reactive liquids, high-pressure gas impingement from leaking piping, rocket-type impact from broken pressurized gas cylinders, loss of control of mobile equipment, cryogenic liquid spill, fire, and smoke. The time needed to perform the interaction analysis is minimized by consolidating safety class items into segregated areas. Each area containing safety class items is evaluated, and any potential threat to the safety functions is noted. After relocation of safety class items is considered, items that pose a threat are generally upgraded to eliminate the threat to the safety class items. Upgrading is the preferred option when relocation is not possible. An example will illustrate the method and application in the phased design, procurement, and construction environment of the HWVP.
Safety evaluation for packaging CPC metal boxes
Romano, T.
This Safety Evaluation for Packaging (SEP) provides authorization for the use of Container Products Corporation (CPC) metal boxes, as described in this document, for the interarea shipment of radioactive contaminated equipment and debris for storage in the Central Waste Complex (CWC) or T Plant located in the 200 West Area. Authorization is granted until November 30, 1995. The CPC boxes included in this SEP were originally procured as US Department of Transportation (DOT) Specification 7A Type A boxes. A review of the documentation provided by the manufacturer revealed the documentation did not adequately demonstrate compliance to the 4 ft drop test requirement of 49 CFR 173.465(c). Preparation of a SEP is necessary to document the equivalent safety of the onsite shipment in lieu of meeting DOT packaging requirements until adequate documentation is received. The equivalent safety of the shipment is based on the fact that the radioactive contents consist of contaminated equipment and debris which are not dispersible. Each piece is wrapped in two layers of no less than 4 mil plastic prior to being placed in the box which has an additional 10 mil liner. Pointed objects and sharp edges are padded to prevent puncture of the plastic liner and wrapping.
Guaranteeing safety in spatially situated agents
Kohout, R.C.; Hendler, J.A. [Univ. of Maryland, College Park, MD (United States); Musliner, D.J. [Honeywell Technology Center, Minneapolis, MN (United States)
{open_quote}Mission-critical{close_quotes} systems, which include such diverse applications as nuclear power plant controllers, {open_quotes}fly-by-wire{close_quotes} airplanes, medical care and monitoring systems, and autonomous mobile vehicles, are characterized by the fact that system failure is potentially catastrophic. The high cost of failure justifies the expenditure of considerable effort at design-time in order to guarantee the correctness of system behavior. This paper examines the problem of guaranteeing safety in a well studied class of robot motion problems known as the {open_quotes}asteroid avoidance problem.{close_quotes} We establish necessary and sufficient conditions for ensuring safety in the simple version of this problem which occurs most frequently in the literature, as well as sufficient conditions for a more general and realistic case. In doing so, we establish functional relationships between the number, size and speed of obstacles, the robot`s maximum speed and the conditions which must be maintained in order to ensure safety.
SELF�MONITORING DISTRIBUTED MONITORING SYSTEM FOR NUCLEAR POWER PLANTS (PRELIMINARY VERSION) Aldo and identification are extremely important activities for the safety of a nuclear power plant. In particular inside huge and complex production plants. 1 INTRODUCTION Safety in nuclear power plants requires
Toolbox Safety Talk Ladder Safety
Pawlowski, Wojtek
Health & Safety for recordkeeping. Slips, trips, and falls constitute the majority of general industry and construction worker injuries. Falls cause 15% of all accidental deaths, and are second only to motor vehicle
Aging management of containment structures in nuclear power plants
Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [The Johns Hopkins Univ., Baltimore, MD (United States); Graves, H.L. III; Norris, W.E. [US Nuclear Regulatory Commission, Washington, DC (United States)
Research is being conducted by ORNL under US Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques. assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.
Next Generation Nuclear Plant GAP Analysis Report
Ball, Sydney J [ORNL; Burchell, Timothy D [ORNL; Corwin, William R [ORNL; Fisher, Stephen Eugene [ORNL; Forsberg, Charles W. [Massachusetts Institute of Technology (MIT); Morris, Robert Noel [ORNL; Moses, David Lewis [ORNL
As a follow-up to the phenomena identification and ranking table (PIRT) studies conducted recently by NRC on next generation nuclear plant (NGNP) safety, a study was conducted to identify the significant 'gaps' between what is needed and what is already available to adequately assess NGNP safety characteristics. The PIRT studies focused on identifying important phenomena affecting NGNP plant behavior, while the gap study gives more attention to off-normal behavior, uncertainties, and event probabilities under both normal operation and postulated accident conditions. Hence, this process also involved incorporating more detailed evaluations of accident sequences and risk assessments. This study considers thermal-fluid and neutronic behavior under both normal and postulated accident conditions, fission product transport (FPT), high-temperature metals, and graphite behavior and their effects on safety. In addition, safety issues related to coupling process heat (hydrogen production) systems to the reactor are addressed, given the limited design information currently available. Recommendations for further study, including analytical methods development and experimental needs, are presented as appropriate in each of these areas.
Reactor operation safety information document
Not Available
The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)
Controller Synthesis of Discrete Linear Plants Using MATTEO SLANINA
Sankaranarayanan, Sriram
Controller Synthesis of Discrete Linear Plants Using Polyhedra MATTEO SLANINA Stanford University controllers for linear discrete systems with disturbances. Given a plant description and a safety We study techniques for synthesizing synchronous controllers for affine plants with disturbances
In-Plant Reliability Data base for nuclear plant components. Interim report: diesel generators, batteries, chargers and inverters
Kahl, W.K.; Borkowski, R.J.
The objective of the In-Plant Reliability Data (IPRD) program is to develop a comprehensive, component-specific reliability data base for probabilistic risk assessment and for other statistical analyses relevant to component reliability evaluations. This document is the product of a pilot study that was undertaken to demonstrate the methodology and feasibility of applying IPRDS techniques to develop and analyze the reliability characteristics of key electrical components in five nuclear power plants. These electrical components include diesel generators, batteries, battery chargers and inverters. The sources used to develop the data base and produce the component failure rates and mean repair times were the plant equipment lists, plant drawings, maintenance work requests, Final Safety Analysis Reports (FSARs), and interviews with plant personnel. The data spanned approximately 33 reactor-years of commercial operation.
MHTGR (modular high-temperature gas-cooled reactor) control: A non-safety related system
Rodriguez, C.; Swart, F.
The modular high-temperature gas-cooled reactor (MHTGR) design meets stringent top-level safety regulatory criteria and user requirements that call for high plant availability and no disruption of the public's day to day activities during normal and off-normal operation of the plant. These requirements lead to a plant design that relies mainly on physical properties and passive design features to ensure plant safety regardless of operator actions, plus simplicity and automation to ensure high plant availability and lower cost of operations. The plant does not require safety-related operator actions, and it does not require the control room to be safety related.
Criteria for safety-related operator actions
Gray, L.H.; Haas, P.M.
The Safety-Related Operator Actions (SROA) Program was designed to provide information and data for use by NRC in assessing the performance of nuclear power plant (NPP) control room operators in responding to abnormal/emergency events. The primary effort involved collection and assessment of data from simulator training exercises and from historical records of abnormal/emergency events that have occurred in operating plants (field data). These data can be used to develop criteria for acceptability of the use of manual operator action for safety-related functions. Development of criteria for safety-related operator actions are considered.
Toolbox Safety Talk Welding & Metal Work Safety
Pawlowski, Wojtek
Toolbox Safety Talk Welding & Metal Work Safety Environmental Health & Safety Facilities Safety or harmful emission giving metals. Welding Safety When welding outside of a designated welding booth, ensure injury. Avoid welding on materials such as galvanized or stainless steel in order to minimize toxic fume
Assistant Manager for Safety and Environment
Broader source: Energy.gov [DOE]
A successful candidate in this position will provide high quality and timely services to fulfill environment, safety, health, and quality programs and responsibilities. This includes oversight of...
Nuclear Safety Regulatory Framework
Broader source: Energy.gov (indexed) [DOE]
overall Nuclear Safety Policy & ESH Goals Safety Basis Review and Approval In the DOE governance model, contractors responsible for the facility develop the safety basis and...
Broader source: Energy.gov [DOE]
This section establishes the Pipeline Safety Division within the Utility Regulatory Commission to administer federal pipeline safety standards and establish minimum state safety standards for...
A literature review of safety culture.
Cole, Kerstan Suzanne; Stevens-Adams, Susan Marie; Wenner, Caren A.
Workplace safety has been historically neglected by organizations in order to enhance profitability. Over the past 30 years, safety concerns and attention to safety have increased due to a series of disastrous events occurring across many different industries (e.g., Chernobyl, Upper Big-Branch Mine, Davis-Besse etc.). Many organizations have focused on promoting a healthy safety culture as a way to understand past incidents, and to prevent future disasters. There is an extensive academic literature devoted to safety culture, and the Department of Energy has also published a significant number of documents related to safety culture. The purpose of the current endeavor was to conduct a review of the safety culture literature in order to understand definitions, methodologies, models, and successful interventions for improving safety culture. After reviewing the literature, we observed four emerging themes. First, it was apparent that although safety culture is a valuable construct, it has some inherent weaknesses. For example, there is no common definition of safety culture and no standard way for assessing the construct. Second, it is apparent that researchers know how to measure particular components of safety culture, with specific focus on individual and organizational factors. Such existing methodologies can be leveraged for future assessments. Third, based on the published literature, the relationship between safety culture and performance is tenuous at best. There are few empirical studies that examine the relationship between safety culture and safety performance metrics. Further, most of these studies do not include a description of the implementation of interventions to improve safety culture, or do not measure the effect of these interventions on safety culture or performance. Fourth, safety culture is best viewed as a dynamic, multi-faceted overall system composed of individual, engineered and organizational models. By addressing all three components of safety culture, organizations have a better chance of understanding, evaluating, and making positive changes towards safety within their own organization.
Watson, Craig A.
Woods Safety SFRC UF 7/09 #12;Working alone #12;Poison Plants Poison ivy (Toxicodendron radicans) Poison oak (Toxicodendron pubescens) #12;Commonly Mistaken Plants Virginia creeper (Parthenocissus quinquefolia) Blackberry (Rubus spp.) Three leaflets let it be! #12;Poison Ivy/Oak Prevention: � Wear long
Balance of Plant Requirements for a Nuclear Hydrogen Plant
Bradley Ward
This document describes the requirements for the components and systems that support the hydrogen production portion of a 600 megawatt thermal (MWt) Next Generation Nuclear Plant (NGNP). These systems, defined as the "balance-of-plant" (BOP), are essential to operate an effective hydrogen production plant. Examples of BOP items are: heat recovery and heat rejection equipment, process material transport systems (pumps, valves, piping, etc.), control systems, safety systems, waste collection and disposal systems, maintenance and repair equipment, heating, ventilation, and air conditioning (HVAC), electrical supply and distribution, and others. The requirements in this document are applicable to the two hydrogen production processes currently under consideration in the DOE Nuclear Hydrogen Initiative. These processes are the sulfur iodide (S-I) process and the high temperature electrolysis (HTE) process. At present, the other two hydrogen production process - the hybrid sulfur-iodide electrolytic process (SE) and the calcium-bromide process (Ca-Br) -are under flow sheet development and not included in this report. While some features of the balance-of-plant requirements are common to all hydrogen production processes, some details will apply only to the specific needs of individual processes.
Health, Safety, and Environment Division
Wade, C [comp.] [comp.
The primary responsibility of the Health, Safety, and Environmental (HSE) Division at the Los Alamos National Laboratory is to provide comprehensive occupational health and safety programs, waste processing, and environmental protection. These activities are designed to protect the worker, the public, and the environment. Meeting these responsibilities requires expertise in many disciplines, including radiation protection, industrial hygiene, safety, occupational medicine, environmental science and engineering, analytical chemistry, epidemiology, and waste management. New and challenging health, safety, and environmental problems occasionally arise from the diverse research and development work of the Laboratory, and research programs in HSE Division often stem from these applied needs. These programs continue but are also extended, as needed, to study specific problems for the Department of Energy. The results of these programs help develop better practices in occupational health and safety, radiation protection, and environmental science.
McQuade, D. Tyler
use Integrated Safety Management here at the lab to reduce risk and work to improve the quality and safety of the work? #12;Integrated Safety Management Use (greater in size than wavelength) #12;Integrated Safety Management Remember, we
Online Safety Monitoring Using Safety Modes Jeremie Guiochet, David Powell and Etienne Baudin
Paris-Sud XI, Universit� de
: transportation [3], space [4], medical systems [5], [6], civil engineering [7], nuclear power plants [8- functional robotic systems. We present an approach and a formalization of the process for determining safety by the monitored system. In practice, each safety mode is related to one or several functional modes
Formal Management Review of the Safety Basis Calculations Noncompliance
Altenbach, T J
In Reference 1, LLNL identified a failure to adequately implement an institutional commitment concerning administrative requirements governing the documentation of Safety Basis calculations supporting the Documented Safety Analysis (DSA) process for LLNL Hazard Category 2 and Category 3 nuclear facilities. The AB Section has discovered that the administrative requirements of AB procedure AB-006, 'Safety Basis Calculation Procedure for Category 2 and 3 Nuclear Facilities', have not been uniformly or consistently applied in the preparation of Safety Basis calculations for LLNL Hazard Category 2 and 3 Nuclear Facilities. The SEP Associated Director has directed the AB Section to initiate a formal management review of the issue that includes, but is not necessarily limited to the following topics: (1) the basis establishing Ab-006 as a required internal procedure for Safety Basis calculations; (2) how requirements for Safety Basis calculations flow down in the institutional DSA process; (3) the extent to which affected Laboratory organizations have explicitly complied with the requirements of Procedure AB-006; (4) what alternative approaches LLNL organizations has used for Safety Basis calculations and how these alternate approaches compare with Procedure AB-006 requirements; and (5) how to reconcile Safety Basis calculations that were performed before Procedure AB-006 came into existence (i.e., August 2001). The management review2 also includes an extent-of-condition evaluation to determine how widespread the discovered issue is throughout Laboratory organizations responsible for operating nuclear facilities, and to determine if implementation of AB procedures other than AB-006 has been similarly affected. In Reference 2, Corrective Action 1 was established whereby the SEP Directorate will develop a plan for performing a formal management review of the discovered condition, including an extent-of condition evaluation. In Reference 3, a plan was provided to prepare a formal management review, satisfying Corrective Action 1. An AB-006 Working Group was formed,led by the AB Section, with representatives from the Nuclear Materials Technology Program (NMTP), the Radioactive and Hazardous Waste Management (RHWM) Division, and the Packaging and Transportation Safety (PATS) Program. The key action of this management review was for Working Group members to conduct an assessment of all safety basis calculations referenced in their respective DSAs. Those assessments were tasked to provide the following information: (1) list which safety basis calculations correctly follow AB-006 and therefore require no additional documentation; (2) identify and list which safety basis calculations do not strictly follow AB-006, these include NMTP Engineering Notes, Engineering Safety Notes, and calculations by organizations external to the nuclear facilities (such as Plant Engineering), subcontractor calculations, and other internally generated calculations. Each of these will be reviewed and listed on a memorandum with the facility manager's (or designee's) signature accepting that calculation for use in the DSA. If any of these calculations are lacking the signature of a technical reviewer, they must also be reviewed for technical content and that review documented per AB-006.
The safety implications of emerging software paradigms
Suski, G.J.; Persons, W.L.; Johnson, G.L.
This paper addresses some of the emerging software paradigms that may be used in developing safety-critical software applications. Paradigms considered in this paper include knowledge-based systems, neural networks, genetic algorithms, and fuzzy systems. It presents one view of the software verification and validation activities that should be associated with each paradigm. The paper begins with a discussion of the historical evolution of software verification and validation. Next, a comparison is made between the verification and validation processes used for conventional and emerging software systems. Several verification and validation issues for the emerging paradigms are discussed and some specific research topics are identified. This work is relevant for monitoring and control at nuclear power plants.
Nero, jA.V.
Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"Densities Surrounding Nuclear Power Plants," by A.V. Nero,
Yen, W.W.S.
Vol. 16, No.1, January- Power-Plant Sites. 1i February 1975,~ ties Surrounding Nuclear Power Plants. LBlr5921, LawrenceS. Commercial Nuclear Power Plants. WASH-1400. October 1975.
Nero, jA.V.
Standards for Nuclear Power Plants," by A.V. Nero and Y.C.Planning for Nuclear Power Plants in California," by W.W.S.Surrounding Nuclear Power Plants," by A.V. Nero, C.H.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
DOE-STD-1104 contains the Department's method and criteria for reviewing and approving nuclear facility's documented safety analysis (DSA). This review and approval formally document the basis for DOE, concluding that a facility can be operated safely in a manner that adequately protects workers, the public, and the environment. Therefore, it is appropriate to formally require implementation of the review methodology and criteria contained in DOE-STD-1104.
Bergman, Ulf C. (Malmoe, SE)
The safety valve contains a resilient gland to be held between a valve seat and a valve member and is secured to the valve member by a sleeve surrounding the end of the valve member adjacent to the valve seat. The sleeve is movable relative to the valve member through a limited axial distance and a gap exists between said valve member and said sleeve.
Spitalnik, J.
Safety Management has lately been considered by some Nuclear Regulatory agencies as the tool on which to concentrate their efforts to implement modern regulation structures, because Safety Culture was said to be difficult to monitor. However, Safety Culture can be assessed and monitored even if it is problematical to make Safety Culture the object of regulation. This paper stresses the feasibility and importance of Safety Culture Assessment based on self-assessment applications performed in several nuclear organizations in Latin America. Reasons and ownership for assessing Safety Culture are discussed, and relevant aspects considered for setting up and programming such an assessment are shown. Basic principles that were taken into account, as well as financial and human resources used in actual self-assessments are reviewed, including the importance of adequate statistical analyses and the necessity of proper feed-back of results. The setting up of action plans to enhance Safety Culture is the final step of the assessment program that once implemented will enable to establish a Safety Culture monitoring process within the organization.
www.nasa.gov WSTF SAFETY AND HEALTH CAPABILITIES
and Evaluation Contract (NTEC). The disciplines include system safety, industrial hygiene, industrial safety) Safety & Mission Assurance (S&MA) offices support all WSTF test activities and general industrial safety consultation to managers and supervisors on matters concerning industrial and test issues. They assist
Nuclear power plant control room operators' performance research
Gray, L.H.; Haas, P.M.
A research program is being conducted to provide information on the performance of nuclear power plant control room operators when responding to abnormal/emergency events in the plants and in full-scope training simulators. The initial impetus for this program was the need for data to assess proposed design criteria for the choice of manual versus automatic action for accomplishing safety-related functions during design basis accidents. The program also included studies of training simulator capabilities, of procedures and data for specifying and verifying simulator performance, and of methods and applications of task analysis.
Tingley, Joseph V.
Basin Center for Geothermal Energy at University of Nevada, Reno (UNR) 2 Nevada Geodetic LaboratoryStillwater Power Plant Wabuska Power Plant Casa Diablo Power Plant Glass Mountain Geothermal Area Lassen Geothermal Area Coso Hot Springs Power Plants Lake City Geothermal Area Thermo Geothermal Area
Safety aspects of the US advanced LMR (liquid metal reactor) design
Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S. (Argonne National Lab., IL (USA); General Electric Co., San Jose, CA (USA); Argonne National Lab., IL (USA); USDOE Assistant Secretary for Nuclear Energy, Washington, DC (USA))
The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs.
Safety and Security on Campus 2013 Annual Security and Fire Safety Report for the Minneapolis and St. Paul Campuses
Minnesota, University of
Safety and Security on Campus 2013 Annual Security and Fire Safety Report for the Minneapolis, active place. Safety and security are important for the thousands of us who live, work, and study here the Twin Cities. But looking out for our safety and security includes us, too. We all have a role to play
Safety Share from National Safety Council
Broader source: Energy.gov [DOE]
Slide Presentation by Joe Yanek, Fluor Government Group. National Safety Council Safety Share. The Campbell Institute is the �Environmental, Health and Safety (EHS) Center of Excellence� at the National Safety Council and provides a Forum for Leaders in EHS to exchange ideas and collaborate across industry sectors and organizational types.
Wedde, Horst F.
. In safety-critical systems (such as nuclear power plants, autonomous cooperation of robots in Outer Space-Critical / Safety-critical Systems, Functional/ Operational Integration 1. INTRODUCTION Safety-critical systems that are typically unpredictable, a very high amount of adaptability of system functions is demanded. safety
Westinghouse Small Modular Reactor balance of plant and supporting systems design
Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)
The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)
OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 2: Designs, Assessments, and Comparisons
Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C. D.; Laybourne, P.; Shillito, K. R.
This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability, economics, and technology development needs.
OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 1: Executive Summary & Overview
Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C.D.; Laybourne, P.; Shillito, K. R.
This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability economics, and technology development needs.
Autoclave nuclear criticality safety analysis
D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)
Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.
Safety culture management: The importance of organizational factors
Haber, S.B.; Shurberg, D.A. [Brookhaven National Lab., Upton, NY (United States); Jacobs, R. [Pennsylvania State Univ., University Park, PA (United States); Hofmann, D. [Purdue Univ., Lafayette, IN (United States)
The concept of safety culture has been used extensively to explain the underlying causes of performance based events, both positive and negative, across the nuclear industry. The work described in this paper represents several years of effort to identify, define and assess the organizational factors important to safe performance in nuclear power plants (NPPs). The research discussed in this paper is primarily conducted in support of the US Nuclear Regulatory Commission`s (NRC) efforts in understanding the impact of organizational performance on safety. As a result of a series of research activities undertaken by numerous NRC contractors, a collection of organizational dimensions has been identified and defined. These dimensions represent what is believed to be a comprehensive taxonomy of organizational elements that relate to the safe operation of nuclear power plants. Techniques were also developed by which to measure these organizational dimensions, and include structured interview protocols, behavioral checklists, and behavioral anchored rating scales (BARS). Recent efforts have focused on devising a methodology for the extraction of information related to the identified organizational dimensions from existing NRC documentation. This type of effort would assess the applicability of the organizational dimensions to existing NRC inspection and evaluation reports, refine the organizational dimensions previously developed so they are more relevant to the task of retrospective analysis, and attempt to rate plants based on the review of existing NRC documentation using the techniques previously developed for the assessment of organizational dimensions.
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Scienceand Requirements RecentlyElectronicResourcesjobsJuly throughR EMaterialsSafety, Security
Risk-Oriented Safety Evaluation of the CAREM-25 Prototype Reactor
Baron, Jorge H.; McLeod, Jorge E. Nunez; Rivera, Selva S
Construction of the CAREM-25 full-size prototype, a very low power nuclear power station [25 MW(electric)], is scheduled to begin in Argentina in 2001. The CAREM-25 is designed based on principles of inherent safety, passive safety functions, and ease of operation. This paper analyzes the safety philosophy from the point of view of risk by performing a level-III probabilistic safety assessment (PSA) of this prototype. The specific PSA steps are discussed, including a specially developed method to obtain representative initiating events, system analysis by fault trees, event development in event trees, plant and containment response analysis, containment event tree development, consequence calculations, and risk representation. The PSA results are presented and discussed in terms of their own values as well as in comparison to other PSA results performed for larger nuclear power plants (NPPs). The advantages of the CAREM-25 from the risk point of view are studied in terms of the effective reduction of both the probability of severe accident sequences and the potential consequences of such sequences (radiological and emergency preparedness impact). The risk point of view also provides a perspective to analyze the impact of several design modifications in order to further reduce the residual risk of the NPP. These design modifications, several of which have already been included in the prototype, are discussed and evaluated.
Safety culture in the nuclear power industry : attributes for regulatory assessment
Alexander, Erin L
Safety culture refers to the attitudes, behaviors, and conditions that affect safety performance and often arises in discussions following incidents at nuclear power plants. As it involves both operational and management ...
CRAD, Nuclear Safety Delegations for Documented Safety Analysis...
Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site
Safety Delegations for Documented Safety Analysis Approval - January 8, 2015 (EA CRAD 31-09, Rev. 0) CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval -...
Session and Paper Number: 245g Plant & Environmental Safety:Plant & Environmental Safety:Plant & Environmental Safety:Plant & Environmental Safety
Howat, Colin S. "Chip"
Laboratory Department of Chemical & Petroleum EngineeringDepartment of Chemical & Petroleum EngineeringDepartment of Chemical & Petroleum EngineeringDepartment of Chemical & Petroleum Engineering University of Kansas of ChemicaDepartment of ChemicaDepartment of ChemicaDepartment of Chemical & Petroleum Engineeringl
Topaz II preliminary safety assessment
Marshall, A.C. (Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States)); Standley, V. (Air Force Phillips Laboratory, Albuquerque, New Mexico 87110 (United States)); Voss, S.S. (Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)); Haskin, E. (Department of Chemical and Nuclear Engineering Department, Institute for Nuclear Power Studies, University of New Mexico, Albuquerque, New Mexico 87110 (United States))
The Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary safety assessment included a top level event tree, neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, and analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment, it appears that it will be possible to safely launch the Topaz II system in the U.S. with some possible system modifications. The principal system modifications will probably include design changes to preclude water flooded criticality and to assure intact reentry.
System Design and the Safety Basis
Ellingson, Darrel
The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination & decommissioning (D&D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities.
Safety evaluation of interim stabilization of non-stabilized single-shell watch list tanks
Stahl, S.M.
The report provides a summation of the status of safety issues associated with interim stabilization of Watch List SSTs (organic, ferrocyanide, and flammable gas), as extracted from recent safety analyses, including the Tank Farms Accelerated Safety Analysis efforts.
Nam Dinh; Ronaldo Szilard
The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system�s �loading� and its �capacity�, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.
Acceptable NSLS Safety Documentation
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Acceptable NSLS Safety Documentation Print NSLS users who have completed NSLS Safety Module must present a copy of one of the following documents to receive ALS 1001: Safety at the...
Orientation Visit to the Portsmouth Gaseous Diffusion Plant
Office of Environmental Management (EM)
Analysis (DSA) and Technical Safety Requirements (TSR) for Portsmouth Gaseous Diffusion Plant Category 2 Non-leased Facilities: X-345 Special Nuclear Material Storage Facility;...
Independent Activity Report, Pantex Plant - October 2011 | Department...
Broader source: Energy.gov (indexed) [DOE]
Plant Operational Awareness Site Visit HIAR PTX-2011-10-28 This Independent Activity Report documents an operational awareness activity conducted by Office of Health, Safety and...
Introduction to LNG vehicle safety. Topical report
Bratvold, D.; Friedman, D.; Chernoff, H.; Farkhondehpay, D.; Comay, C.
Basic information on the characteristics of liquefied natural gas (LNG) is assembled in this report to provide an overview of safety issues and practices for the use of LNG vehicles. This document is intended for those planning or considering the use of LNG vehicles, including vehicle fleet owners and operators, public transit officials and boards, local fire and safety officials, manufacturers and distributors, and gas industry officials. Safety issues and mitigation measures that should be considered for candidate LNG vehicle projects are addressed.
RADIATION SAFETY TRAINING MANUAL Radiation Safety Office
Sibille, Etienne
protection and the potential risks of ionizing radiation. Radiation Safety Office personnel provide.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 II. OVERVIEW OF REGULATIONS, PROTECTION STANDARDS, AND RADIATION SAFETY ORGANIZATION.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 V. BASIC RADIATION PROTECTION PRINCIPLES
Integrating Safety Assessment Methods using the Risk Informed Safety Margins Characterization (RISMC) Approach
Curtis Smith; Diego Mandelli
Safety is central to the design, licensing, operation, and economics of nuclear power plants (NPPs). As the current light water reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of systems, structures, and components (SSC) degradations or failures that initiate safety significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly �over-design� portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as �safety margin.� Historically, specific safety margin provisions have been formulated primarily based on engineering judgment backed by a set of conservative engineering calculations. The ability to better characterize and quantify safety margin is important to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development (R&D) in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the RISMC Pathway provides methods and tools that enable mitigation options known as margins management strategies. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. As the lead Department of Energy (DOE) Laboratory for this Pathway, the Idaho National Laboratory (INL) is tasked with developing and deploying methods and tools that support the quantification and management of safety margin and uncertainty.
Najmabadi, Farrokh
FUSION POWER PLANTS � GOALS AND TECHNOLOGICAL CHALLENGES Farrokh Najmabadi Dept. of Electrical for fusion power plants is given and their economic, safety, and environmental features are explored. Concep- tual design studies predict that fusion power plants will be capital intensive and will be used
Broader source: Energy.gov [DOE]
The Pennsylvania Department of Environmental Protection's Division of Dam Safety provides for the regulation and safety of dams and reservoirs throughout the Commonwealth in order to protect the...
Pipeline Safety (Pennsylvania)
Broader source: Energy.gov [DOE]
The Pennsylvania legislature has empowered the Public Utility Commission to direct and enforce safety standards for pipeline facilities and to regulate safety practices of certificated utilities...
SHSD Manager Safety Engineering Group Manager
Safety, Machine Shop Safety, Tier I Program, Traffic Safety S. Moss: Nuclear Criticality Safety G. Shepherd: Explosives Safety, Facility Authorization Basis, Nuclear Safety R. Travis: Readiness Evaluations
Environmental Health and Safety Fire and Life Safety Laboratory Assessment
Environmental Health and Safety Fire and Life Safety Laboratory Assessment PI or environmental concerns were identified. B. Items of safety or environmental concerns were identified. C. Uncorrected repeated safety or environmental items were identified. Safety Equipment # Compliance Items
Paris-Sud XI, Université de
SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS Piero Baraldi Chevalier EDF R&D Simulation and information Technologies for Power generation system Department 6, Quai Monitoring, Empirical Modeling, Power Plants, Safety Critical Nuclear Instrumentation, Autoassociative models
Risk-informed incident management for nuclear power plants
Smith, Curtis Lee, 1966-
Decision making as a part of nuclear power plant operations is a critical, but common, task. Plant management is forced to make decisions that may have safety and economic consequences. Formal decision theory offers the ...
EHSO TRAINING CLASSES Fire Safety Program Training
EHSO TRAINING CLASSES Fire Safety Program Training 1. Fire Safety (60 minutes) Instruction includes an actual fire eperience. 2. Fire Extinguisher Training (30 minutes) A practical demonstration on actual burnable liquid fires. This practical extinguisher training is a critical portion of the fire
FAQS Qualification Card- Aviation Safety Officer
Broader source: Energy.gov [DOE]
A key element for the Department�s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).
FAQS Qualification Card � Safety Software Quality Assurance
Broader source: Energy.gov [DOE]
A key element for the Department�s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).
FAQS Qualification Card � Senior Technical Safety Manager
Broader source: Energy.gov [DOE]
A key element for the Department�s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).
FAQS Qualification Card � Nuclear Explosive Safety Study
Broader source: Energy.gov [DOE]
A key element for the Department�s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).
2011 Annual Criticality Safety Program Performance Summary
Andrea Hoffman
The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The TSR limits fuel plate bundles to 1085 grams U-235, which is the maximum loading of an ATR fuel element. The overloaded fuel plate bundle contained 1097 grams U-235 and was assembled under an 1100 gram U-235 limit in 1982. In 2003, the limit was reduced to 1085 grams citing a new criticality safety evaluation for ATR fuel elements. The fuel plate bundle inventories were not checked for compliance prior to implementing the reduced limit. A subsequent review of the NMIS inventory did not identify further violations. Requirements Management - The INL Criticality Safety program is organized and well documented. The source requirements for the INL Criticality Safety Program are from 10 CFR 830.204, DOE Order 420.1B, Chapter III, 'Nuclear Criticality Safety,' ANSI/ANS 8-series Industry Standards, and DOE Standards. These source requirements are documented in LRD-18001, 'INL Criticality Safety Program Requirements Manual.' The majority of the criticality safety source requirements are contained in DOE Order 420.1B because it invokes all of the ANSI/ANS 8-Series Standards. DOE Order 420.1B also invokes several DOE Standards, including DOE-STD-3007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities.' DOE Order 420.1B contains requirements for DOE 'Heads of Field Elements' to approve the criticality safety program and specific elements of the program, namely, the qualification of criticality staff and the method for preparing criticality safety evaluations. This was accomplished by the approval of SAR-400, 'INL Standardized Nuclear Safety Basis Manual,' Chapter 6, 'Prevention of Inadvertent Criticality.' Chapter 6 of SAR-400 contains sufficient detail and/or reference to the specific DOE and contractor documents that adequately describe the INL Criticality Safety Program per the elements specified in DOE Order 420.1B. The Safety Evaluation Report for SAR-400 specifically recognizes that the approval of SAR-400 approves the INL Criticality Safety Program. No new source requirements were released in 2011. A revision to LRD-18001 is
Use of neural networks in the operation of nuclear power plants
Uhrig, R.E. (Tennessee Univ., Knoxville, TN (USA) Oak Ridge National Lab., TN (USA))
Application of neural networks to the operation of nuclear power plants is being investigated under a US Department of Energy sponsored program at the University of Tennessee. Projects include the feasibility of using neural networks for the following tasks: (a) diagnosing specific abnormal conditions, (b) detection of the change of mode of operation, (c) signal validation, (d) monitoring of check valves, (e) modeling of the plant thermodynamics, (f) emulation of core reload calculations, (g) analysis of temporal sequences in NRC's licensee event report,'' (h) monitoring of plant parameters, and (i) analysis of plant vibrations. Each of these projects and its status are described briefly in this article. the objective of each of these projects is to enhance the safety and performance of nuclear plants through the use of neural networks. 6 refs.
FPS 12.1 Life Safety, 5/24/2000
Broader source: Energy.gov [DOE]
The objective of this surveillance is for the Facility Representative to verify that conditions in the plant provide an appropriate level of life safety for workers conducting operations in the...
Nuclear Safety Workshop Agenda - Post Fukushima Initiatives and...
Initiatives and Results In response to the March 2011 accident at the Fukushima Daiichi nuclear power plant, Secretary Chu initiated a series of actions to review the safety of...
Aspects of environmental and safety analysis of fusion reactors
Kazimi, Mujid S.
This report summarizes the progress made between October 1976 and September 1977 in studies of some environmental and safety considerations in fusion reactor plants. A methodology to assess the admissible occurrence rate ...
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Plants & Animals Plants & Animals Plant and animal monitoring is performed to determine whether Laboratory operations are impacting human health via the food chain. February 2,...
Facility Safety (9-23-10)--Withdrawn
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Withdrawn, 5-19-2014--This approval includes revision of the three implementing Guides: DOE G 420.1-1, Nonreactor Nuclear Safety Design Criteria and Explosive Safety Criteria Guide for Use with DOE O 420.1, Facility Safety; DOE G 420.1-2, Guide for the Mitigation of Natural Phenomena Hazards for DOE Nuclear Facilities and NonNuclear Facilities; and DOE G 420.1-3, Implementation Guide for DOE Fire Protection and Emergency Services Programs for Use with DOE O 420.1B, Facility Safety
Pilot plant UF/sub 6/ to UF/sub 4/ test operations report
Bicha, W.J.; Fallings, M.; Gilbert, D.D.; Koch, G.E.; Levine, P.J.; McLaughlin, D.F.; Nuhfer, K.R.; Reese, J.C.
The FMPC site includes a plant designed for the reduction of uranium hexafluoride (UF/sub 6/) to uranium tetrafluoride (UF/sub 4/). Limited operation of the upgraded reduction facility began in August 1984 and continued through January 19, 1986. A reaction vessel ruptured on that date causing the plant operation to be shut down. The DOE conducted a Class B investigation with the findings of the investigation board issued in preliminary form in May 1986 and as a final recommendation in July 1986. A two-phase restart of the plant was planned and implemented. Phase I included implementing safety system modifications, changing reaction vessel temperature control strategy, and operating the reduction plant under an 8-week controlled test. The results of the test period are the subject of this report. 41 figs., 11 tabs.
Topaz II preliminary safety assessment
Marshall, A.C. (Sandia National Labs., Albuquerque, NM (United States)); Standley, V. (Air Force Phillips Laboratory, Albuquerque, NM (United States)); Voss, S.S. (Los Alamos National Lab., NM (United States)); Haskin, E. (New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering)
The Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz 11 space nuclear power system. A preliminary safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safely assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary safety assessment included a top level event tree, neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, and analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment, it appears that it will be possible to safely launch the Topaz II system in the US with some possible system modifications. The principal system modifications will probably include design changes to preclude water flooded criticality and to assure intact reentry.
Recommended research on LNG safety
Carpenter, H.J.; Gilmore, F.R.
The US Department of Energy (DOE) is conducting research on the safety and other environmental aspects of liquefied energy gases including liquefied natural gas (LNG). The effort reported here was conducted as part of the planning for further research into the safety aspects of transporting and storing LNG, with primary emphasis on public safety. Although the modern LNG industry has enjoyed excellent success in providing for safe operations, significant questions remain on the part of many, the expressions of which were intensified with the addition of marine-based LNG import terminals. Public safety with regard to large-scale importation of this fuel has received widespread attention in the US Congress, state legislatures, county and city governments, and from various individuals and public groups, with coverage in all the news media, including books published on the subject. The safety concerns have centered around the consequences to the public of a large spill of the cryogenic liquid from an ocean tanker or a larger storage tank, either of which might hold as much as 125,000 m/sup 3/ of LNG.
Tiger Team Assessment of the Paducah Gaseous Diffusion Plant
Not Available
This document contains findings and concerns identified during the Tiger Team Assessment of the Department of Energy's (DOE) Paducah Gaseous Diffusion Plant (PGDP) in Paducah, Kentucky. The assessment was directed by the Department's Office of Environment, Safety and Health (ES H) and was conducted from June 18 to July 20, 1990. The PGDP Tiger Team Assessment is comprehensive in scope. It covers the Environmental, Safety and Health (including OSHA Compliance), and Management areas and determines the site's compliance with applicable federal (including DOE), state, and local regulations and requirements. The objective of the assessment program is to provide the Secretary with information on the current ES H compliance status of DOE facilities, root causation for noncompliance, adequacy of DOE and site contractor ES H management programs, response actions to address the identified problem areas, and DOE-wide ES H compliance trends and root causes.
Tiger Team Assessment of the Paducah Gaseous Diffusion Plant
Not Available
This document contains findings and concerns identified during the Tiger Team Assessment of the Department of Energy's (DOE's) Paducah Gaseous Diffusion Plant (PGDP) in Paducah, Kentucky. The assessment was directed by the Department's Office of Environment, Safety and Health (ES H) and was conducted from June 18 to July 20, 1990. The PGDP Tiger Team Assessment is comprehensive in scope. It covers the Environmental, Safety and Health (including OSHA Compliance), and Management areas and determines the site's compliance with applicable federal (including DOE), state, and local regulations and requirements. The objective of the assessment program is to provide the Secretary with information on the current ES H compliance status of DOE facilities, root causation for noncompliance, adequacy of DOE and site contractor ES H management programs, response actions to address the identified problem areas, and DOE-wide ES H compliance trends and root causes. This volume contains appendices.
Pressure safety program Lawrence Livermore National Laboratory
Borzileri, C.; Traini, M.
The Lawrence Livermore National Laboratory (LLNL) is a Research and Development facility. Programs include research in: nuclear weapons, energy, environmental, biomedical, and other DOE funded programs. LLNL is managed by the University of California for the Department of Energy. Many research and development programs require the use of pressurized fluid systems. In the early 1960`s, courses were developed to train personnel to safely work with pressurized systems. These courses served as a foundation for the Pressure Safety Program. The Pressure Safety Program is administered by the Pressure Safety Manager through the Hazards Control Department, and responsibilities include: (1) Pressure Safety course development and training, (2) Equipment documentation, tracking and inspections/retests, (3) Formal and informal review of pressure systems. The program uses accepted codes and standards and closely follows the DOE Pressure Safety Guidelines Manual. This manual was developed for DOE by Lawrence Livermore National Laboratory. The DOE Pressure Safety Guidelines Manual defines five (5) basic elements which constitute this Pressure Safety Program. These elements are: (1) A Pressure Safety Manual, (2) A Safety Committee, (3) Personnel who are trained and qualified, (4) Documentation and accountability for each pressure vessel or system, (5) Control of the selection and the use of high pressure hardware.
A risk-informed approach to safety margins analysis
Curtis Smith; Diego Mandelli
The Risk Informed Safety Margins Characterization (RISMC) Pathway is a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. The model has been tested on the Advanced Test Reactor (ATR) at Idaho National Lab.
Westinghouse Hanford Company health and safety performance report
Rogers, L.
Topping the list of WHC Safety recognition during this reporting period is a commendation received from the National Safety Council (NSC). The NSC bestowed their Award of Honor upon WHC for significant reduction of incidence rates during CY 1995. The award is based upon a reduction of 48 % or greater in cases involving days away from work, a 30 % or greater reduction in the number of days away, and a 15% or greater reduction in the total number of occupational injuries and illnesses. (page 2-1). A DOE-HQ review team representing the Office of Envirorunent, Safety and Health (EH), visited the Hanford Site during several weeks of the quarter. Ile 40-member Safety Management Evaluation Team (SMET) assessed WHC in the areas of management responsibility, comprehensive requirements, and competence commensurate with responsibility. As part of their new approach to oversight, they focused on the existence of management systems and programs (comparable approach to VPP). Plant/project areas selected for review within WHC were PFP, B Plant/WESF, Tank Farms, and K-Basins (page 2-2). Effective safety meetings, prejob safety meetings, etc., are a cornerstone of any successful safety program. In an effort to improve the reporting of safety meetings, the Safety/Security Meeting Report form was revised. It now provides a mechanism for recording and tracking safety issues (page 2-4). WHC has experienced an increase in the occupational injury and illness incidence rates during the first quarter of CY 1996. Trends show this increase can be partially attributed to inattention to workplace activities due 0999to the uncertainty Hanford employees currently face with recent reduction of force, reorganization, and reengineering efforts (page 2-7). The cumulative CY 1995 lost/restricted workday case incidence rate for the first quarter of CY 1996 (1.28) is 25% below the DOE CY 1991-93 average (1.70). However, the incidence rate increased 24% from the CY 1995 rate of 1.03 (page 2-8). The reengineering of the Radiological Control organization has resulted in an increase in the availability of personnel to help facilities with the procurement and use of practical ALARA measures. In addition, there seems to be more awareness of the need to use ALARA, resulting in a significant increase in the number of calls received by the ALARA Program Office for help in solving radiological problems (page 1-3). The Figure 3-2-1 chart data includes WHC, BCSR, and ICF KH employee exposure. The first quarter CY 1996 results represent the exposure of 1,913 quarterly-badged employees and an average of 846 monthly-badged employees. There were three instances of potential loss of contamination control during the calendar quarter involving three workers where internal dosimetry follow-up was performed. No intakes of contamination were detected (page 3-6). There were five skin contaminations and 18 clothing contaminations reported during this quarter in all WHC-managed facilities/areas. This represents an improvement in performance compared to the first quarter of CY 1995 (page 3-9). A monthly average of 76 Radiological Problem Reports (RPR) was issued during the first quarter of CY 1996 for a total issuance of 227 RPRS. The monthly average for the same quarter in CY 1995 was 76; a total of 228 for the quarter (Figure 3-5-1). At the end of March, WHC Dosimetry was monitoring the following employees/sub- contractors: 631 monthly standard dosimeters, 251 monthly combination dosimeters, 1,386 quarterly standard dosimeters, 472 quarterly combination dosimeters, and 3,716 annual dosimeters. During this period, Dosimetry had 987 requests for changes to the frequency, and terminated 731 dosimetry records.
Hanford Waste Vitrification Plant technical manual
Larson, D.E. [ed.; Watrous, R.A.; Kruger, O.L. [and others
A key element of the Hanford waste management strategy is the construction of a new facility, the Hanford Waste Vitrification Plant (HWVP), to vitrify existing and future liquid high-level waste produced by defense activities at the Hanford Site. The HWVP mission is to vitrify pretreated waste in borosilicate glass, cast the glass into stainless steel canisters, and store the canisters at the Hanford Site until they are shipped to a federal geological repository. The HWVP Technical Manual (Manual) documents the technical bases of the current HWVP process and provides a physical description of the related equipment and the plant. The immediate purpose of the document is to provide the technical bases for preparation of project baseline documents that will be used to direct the Title 1 and Title 2 design by the A/E, Fluor. The content of the Manual is organized in the following manner. Chapter 1.0 contains the background and context within which the HWVP was designed. Chapter 2.0 describes the site, plant, equipment and supporting services and provides the context for application of the process information in the Manual. Chapter 3.0 provides plant feed and product requirements, which are primary process bases for plant operation. Chapter 4.0 summarizes the technology for each plant process. Chapter 5.0 describes the engineering principles for designing major types of HWVP equipment. Chapter 6.0 describes the general safety aspects of the plant and process to assist in safe and prudent facility operation. Chapter 7.0 includes a description of the waste form qualification program and data. Chapter 8.0 indicates the current status of quality assurance requirements for the Manual. The Appendices provide data that are too extensive to be placed in the main text, such as extensive tables and sets of figures. The Manual is a revision of the 1987 version.
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Changes include invoking DOE-STD-3009 and DOE-STD-1104 as requirements documents. NOTE: The two standards are not final documents but can be found in Technical Standards RevCom at https://www.standards.doe.gov/login.jsp. Login as an SME using your email address.
HTGR Dust Safety Issues and Needs for Research and Development
Paul W. Humrickhouse
This report presents a summary of high temperature gas-cooled reactor dust safety issues. It draws upon a literature review and the proceedings of the Very High Temperature Reactor Dust Assessment Meeting held in Rockville, MD in March 2011 to identify and prioritize the phenomena and issues that characterize the effect of carbonaceous dust on high temperature reactor safety. It reflects the work and input of approximately 40 participants from the U.S. Department of Energy and its National Labs, the U.S. Nuclear Regulatory Commission, industry, academia, and international nuclear research organizations on the topics of dust generation and characterization, transport, fission product interactions, and chemical reactions. The meeting was organized by the Idaho National Laboratory under the auspices of the Next Generation Nuclear Plant Project, with support from the U.S. Nuclear Regulatory Commission. Information gleaned from the report and related meetings will be used to enhance the fuel, graphite, and methods technical program plans that guide research and development under the Next Generation Nuclear Plant Project. Based on meeting discussions and presentations, major research and development needs include: generating adsorption isotherms for fission products that display an affinity for dust, investigating the formation and properties of carbonaceous crust on the inside of high temperature reactor coolant pipes, and confirming the predominant source of dust as abrasion between fuel spheres and the fuel handling system.
AVLIS Production Plant Preliminary Quality Assurance Plan and Assessment
Not Available
This preliminary Quality Assurance Plan and Assessment establishes the Quality Assurance requirements for the AVLIS Production Plant Project. The Quality Assurance Plan defines the management approach, organization, interfaces, and controls that will be used in order to provide adequate confidence that the AVLIS Production Plant design, procurement, construction, fabrication, installation, start-up, and operation are accomplished within established goals and objectives. The Quality Assurance Program defined in this document includes a system for assessing those elements of the project whose failure would have a significant impact on safety, environment, schedule, cost, or overall plant objectives. As elements of the project are assessed, classifications are provided to establish and assure that special actions are defined which will eliminate or reduce the probability of occurrence or control the consequences of failure. 8 figures, 18 tables.
Plan for addressing issues relating to oil shale plant siting
Noridin, J. S.; Donovan, R.; Trudell, L.; Dean, J.; Blevins, A.; Harrington, L. W.; James, R.; Berdan, G.
The Western Research Institute plan for addressing oil shale plant siting methodology calls for identifying the available resources such as oil shale, water, topography and transportation, and human resources. Restrictions on development are addressed: land ownership, land use, water rights, environment, socioeconomics, culture, health and safety, and other institutional restrictions. Descriptions of the technologies for development of oil shale resources are included. The impacts of oil shale development on the environment, socioeconomic structure, water availability, and other conditions are discussed. Finally, the Western Research Institute plan proposes to integrate these topics to develop a flow chart for oil shale plant siting. Western Research Institute has (1) identified relative topics for shale oil plant siting, (2) surveyed both published and unpublished information, and (3) identified data gaps and research needs. 910 refs., 3 figs., 30 tabs.
Nero, A.V.
Power Plant Reliability-Availability and State Regulation,"Report on Equipment Availability: Fossil and NuclearBasic Definitions* Availability: Reliability: Base Loading:
Waste Handling and Disposal Biological Safety
Pawlowski, Wojtek
plumbing services, EHS personnel wastewater treatment plant personnel, and the general public canWaste Handling and Disposal Biological Safety General Biosafety Practices (GBP) Why You Should Care on the next experiment. Are you working with r/sNA, biological toxins, human materials, needles, plasticware
Weston, Ken
affecting facility operations or equipment, including electrical, life safety, mechanical, or plumbing outages to any area in the NHMFL. Advance notice to affected areas shall be provided by the Facilities
2003 Pantex Plant Annual Illness and Injury Surveillance Report, Revised September 2007
U.S. Department of Energy, Office of Health, Safety and Security, Office of Illness and Injury Prevention Programs
Annual Illness and Injury Surveillance Program report for 2003 for the Pantex Plant. DOE is commited to assuring the health and safety of its workers. This includes the conduct of epidemiologic surveillance activities that provide an early warning system for health problems among workers. The IISP monitors illnesses and health conditions that result in an absence of workdays, occupational injuries and illnesses, and disabilities and deaths among current workers.
2003 Kansas City Plant Annual Illness and Injury Surveillance Report, Revised September 2007
U.S. Department of Energy, Office of Health, Safety and Security, Office of Illness and Injury Prevention Programs
Annual Illness and Injury Surveillance Program report for 2003 for the Kansas City Plant. The U.S. Department of Energy�s (DOE) commitment to assuring the health and safety of its workers includes the conduct of epidemiologic surveillance activities that provide an early warning system for health problems among workers. The IISP monitors illnesses and health conditions that result in an absence of workdays, occupational injuries and illnesses, and disabilities and deaths among current workers.
Cold Vacuum Drying (CVD) Facility Technical Safety Requirements
The Technical Safety Requirements (TSRs) for the Cold Vacuum Drying Facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation. Controls required for public safety, significant defense-in-depth, significant worker safety, and for maintaining radiological and toxicological consequences below risk evaluation guidelines are included.
AVESTAR Center for Operational Excellence of Electricity Generation Plants
Zitney, Stephen
To address industry challenges in attaining operational excellence for electricity generation plants, the U.S. Department of Energy�s (DOE) National Energy Technology Laboratory (NETL) has launched a world-class facility for Advanced Virtual Energy Simulation Training and Research (AVESTARTM). This presentation will highlight the AVESTARTM Center simulators, facilities, and comprehensive training, education, and research programs focused on the operation and control of high-efficiency, near-zero-emission electricity generation plants. The AVESTAR Center brings together state-of-the-art, real-time, high-fidelity dynamic simulators with full-scope operator training systems (OTSs) and 3D virtual immersive training systems (ITSs) into an integrated energy plant and control room environment. AVESTAR�s initial offering combines--for the first time--a �gasification with CO2 capture� process simulator with a �combined-cycle� power simulator together in a single OTS/ITS solution for an integrated gasification combined cycle (IGCC) power plant with carbon dioxide (CO2) capture. IGCC systems are an attractive technology option for power generation, especially when capturing and storing CO2 is necessary to satisfy emission targets. The AVESTAR training program offers a variety of courses that merge classroom learning, simulator-based OTS learning in a control-room operations environment, and immersive learning in the interactive 3D virtual plant environment or ITS. All of the courses introduce trainees to base-load plant operation, control, startups, and shutdowns. Advanced courses require participants to become familiar with coordinated control, fuel switching, power-demand load shedding, and load following, as well as to problem solve equipment and process malfunctions. Designed to ensure work force development, training is offered for control room and plant field operators, as well as engineers and managers. Such comprehensive simulator-based instruction allows for realistic training without compromising worker, equipment, and environmental safety. It also better prepares operators and engineers to manage the plant closer to economic constraints while minimizing or avoiding the impact of any potentially harmful, wasteful, or inefficient events. The AVESTAR Center is also used to augment graduate and undergraduate engineering education in the areas of process simulation, dynamics, control, and safety. Students and researchers gain hands-on simulator-based training experience and learn how the commercial-scale power plants respond dynamically to changes in manipulated inputs, such as coal feed flow rate and power demand. Students also analyze how the regulatory control system impacts power plant performance and stability. In addition, students practice start-up, shutdown, and malfunction scenarios. The 3D virtual ITSs are used for plant familiarization, walk-through, equipment animations, and safety scenarios. To further leverage the AVESTAR facilities and simulators, NETL and its university partners are pursuing an innovative and collaborative R&D program. In the area of process control, AVESTAR researchers are developing enhanced strategies for regulatory control and coordinated plant-wide control, including gasifier and gas turbine lead, as well as advanced process control using model predictive control (MPC) techniques. Other AVESTAR R&D focus areas include high-fidelity equipment modeling using partial differential equations, dynamic reduced order modeling, optimal sensor placement, 3D virtual plant simulation, and modern grid. NETL and its partners plan to continue building the AVESTAR portfolio of dynamic simulators, immersive training systems, and advanced research capabilities to satisfy industry�s growing need for training and experience with the operation and control of clean energy plants. Future dynamic simulators under development include natural gas combined cycle (NGCC) and supercritical pulverized coal (SCPC) plants with post-combustion CO2 capture. These dynamic simulators are targeted for us
Nuclear criticality safety department training implementation
Carroll, K.J.; Taylor, R.G.; Worley, C.A.
The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. The NCSD Qualification Program is described in Y/DD-694, Qualification Program, Nuclear Criticality Safety Department This document provides a listing of the roles and responsibilities of NCSD personnel with respect to training and details of the Training Management System (TMS) programs, Mentoring Checklists and Checksheets, as well as other documentation utilized to implement the program. This document supersedes Y/DD-696, Revision 2, dated 3/27/96, Training Implementation, Nuclear Criticality Safety Department. There are no backfit requirements associated with revisions to this document.
Subsystem fragility: Seismic Safety Margins Research Program (Phase I)
Kennedy, R. P.; Campbell, R. D.; Hardy, G.; Banon, H.
Seismic fragility levels of safety related equipment are developed for use in a seismic oriented Probabilistic Risk Assessment (PRA) being conducted as part of the Seismic Safety Margins Research Program (SSMRP). The Zion Nuclear Power Plant is being utilized as a reference plant and fragility descriptions are developed for specific and generic safety related equipment groups in Zion. Both equipment fragilities and equipment responses are defined in probabilistic terms to be used as input to the SSMRP event tree/fault tree models of the Zion systems. 65 refs., 14 figs., 11 tabs.
Wedde, Horst F.
. In safety-critical systems, (such as nuclear power plants, distributed cooperation of autonomous robotsNOVEL PRINCIPLES FOR DEVELOPING AND EVALUATING DISTRIBUTED SAFETY-CRITICAL SYSTEMS Horst F. Wedde, Jon A. Lind Informatik III University of Dortmund 44221 Dortmund / Germany Abstract Safety
Office of Nuclear Safety and Environmental Assessments | Department...
Office of Environmental Management (EM)
examine implementation of safety management programs during phases of the life cycle of a nuclear facility including operation, deactivation, decontamination, decommissioning and...
Environment, Safety, and Health Special Review, Department of...
Broader source: Energy.gov (indexed) [DOE]
approach outlined in DOE's Nanoscale Science Research Centers Approach to Nanoscale Environment, Safety, and Health, and other applicable requirements including 10 C.F.R. Part 851...
Boyer, Edmond
, Italy email@example.com 2 INET, Institute of Nuclear and New Energy Technology Tsinghua University, Beijing,100084, China ABSTRACT For licensing purposes, safety cases of Nuclear Power Plants (NPPs) must are traditionally performed for the verification of the safety performance of a Nuclear Power Plant (NPP) under
Nuclear criticality safety aspects of gaseous uranium hexafluoride (UF{sub 6}) in the diffusion cascade
Huffer, J.E. [Parallax, Inc., Atlanta, GA (United States)
This paper determines the nuclear safety of gaseous UF{sub 6} in the current Gaseous Diffusion Cascade and auxiliary systems. The actual plant safety system settings for pressure trip points are used to determine the maximum amount of HF moderation in the process gas, as well as the corresponding atomic number densities. These inputs are used in KENO V.a criticality safety models which are sized to the actual plant equipment. The ENO V.a calculation results confirm nuclear safety of gaseous UF{sub 6} in plant operations..
STARFIRE: a commercial tokamak fusion power plant study
Not Available
STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.
Coal Mine Safety Act (Virginia)
Broader source: Energy.gov [DOE]
This Act is the primary legislation pertaining to coal mine safety in Virginia. It contains information on safety rules, safety standards and required certifications for mine workers, prohibited...
Nuclear Safety Research and Development...
Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site
Nuclear Safety Research and Development Proposal Review and Prioritization Process and Criteria Nuclear Safety Research and Development Program Office of Nuclear Safety Office of...
Magnetic Field Safety Magnetic Field Safety
McQuade, D. Tyler
Magnetic Field Safety Training #12;Magnetic Field Safety Strong Magnetic Fields exist around energized magnets. High magnetic fields alone are a recognized hazard only for personnel with certain medical conditions such as pacemakers, magnetic implants, or embedded shrapnel. In addition, high magnetic
TA-55 Final Safety Analysis Report Comparison Document and DOE Safety Evaluation Report Requirements
Alan Bond
This document provides an overview of changes to the currently approved TA-55 Final Safety Analysis Report (FSAR) that are included in the upgraded FSAR. The DOE Safety Evaluation Report (SER) requirements that are incorporated into the upgraded FSAR are briefly discussed to provide the starting point in the FSAR with respect to the SER requirements.
Chapter 13 Employee Health and Safety 13.04 Safety Committees
Sheridan, Jennifer
compliance in hazardous waste management, environmental permits and other issues related to chemical safety involving use of hazardous biological materials including recombinant DNA for compliance with NIH guidelines and tuberculosis prevention are areas of special concern. Contacts can be made with these committees or with safety
Safety Issues Chemical Storage
Cohen, Robert E.
Safety Issues � Chemical Storage �Store in compatible containers that are in good condition to store separately. #12;Safety Issues � Flammable liquid storage -Store bulk quantities in flammable storage cabinets -UL approved Flammable Storage Refrigerators are required for cold storage � Provide
Broader source: Energy.gov [DOE]
The Office of Nuclear Safety establishes nuclear safety requirements and expectations for the Department to ensure protection of workers and the public from the hazards associated with nuclear operations with all Department operations.
Pipeline Safety Rule (Tennessee)
Broader source: Energy.gov [DOE]
The Pipeline Safety Rule simply states, "The Minimum Federal Safety Standards for the transportation of natural and other gas by pipeline (Title 49, Chapter 1, Part 192) as published in the Federal...
Broader source: Energy.gov [DOE]
The Dam Safety Division within the Department of the Environment is responsible for administering a dam safety program to regulate the construction, operation, and maintenance of dams to prevent...
Broader source: Energy.gov [DOE]
The Delaware Dam Safety Law was adopted in 2004 and provides the framework for proper design, construction, operation, maintenance, and inspection of dams in the interest of public health, safety,...
The AP1000{sup R} nuclear power plant innovative features for extended station blackout mitigation
Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)
Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)
Countries Gasoline Prices Including Taxes
Gasoline and Diesel Fuel Update (EIA)
Selected Countries (U.S. dollars per gallon, including taxes) Date Belgium France Germany Italy Netherlands UK US 51115 6.15 6.08 6.28 6.83 6.96 6.75 3.06 5415 6.14 6.06...
Sponsorship includes: Agriculture in the
Nebraska-Lincoln, University of
Sponsorship includes: · Agriculture in the Classroom · Douglas County Farm Bureau · Gifford Farm · University of Nebraska Agricultural Research and Development Center · University of Nebraska- Lincoln Awareness Coalition is to help youth, primarily from urban communities, become aware of agriculture
Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program
Not Available
Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.
Nuclear Explosive Safety Manual
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Manual provides supplemental details to support the requirements of DOE O 452.2D, Nuclear Explosive Safety.
DOE handbook electrical safety
Electrical Safety Handbook presents the Department of Energy (DOE) safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety guidance and information for DOE installations to effect a reduction or elimination of risks associated with the use of electrical energy. The objectives of this handbook are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.
Not Available
The Electrical Safety Guidelines prescribes the DOE safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety standards and guidance for DOE installations in order to affect a reduction or elimination of risks associated with the use of electrical energy. The objectives of these guidelines are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.
Department of Environmental Health & Safety Emergency Management
O'Toole, Alice J.
Management Fire & Life Safety Industrial Hygiene Laboratory Safety Occupational & General Safety Management Environmental Management Fire & Life Safety Industrial Hygiene Laboratory Safety Occupational Values A Note from the Director Environmental Management Fire & Life Safety Lab Safety & Industrial
Modular HTGR Safety Basis and Approach
Thomas Hicks
The Next Generation Nuclear Plant (NGNP) will be a licensed commercial high temperature gas-cooled reactor (HTGR) capable of producing electricity and/or high temperature process heat for industrial markets supporting a range of end-user applications. The NGNP Project has adopted the 10 CFR 52 Combined License (COL) process, as recommended in the NGNP Licensing Strategy - A Report to Congress, dated August 2008, as the foundation for the NGNP licensing strategy [DOE/NRC 2008]. Nuclear Regulatory Commission (NRC) licensing of the NGNP plant utilizing this process will demonstrate the efficacy for licensing future HTGRs for commercial industrial applications. This information paper is one in a series of submittals that address key generic issues of the priority licensing topics as part of the process for establishing HTGR regulatory requirements. This information paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach with the NRC staff and public stakeholders. The NGNP project does not expect to receive comments on this information paper because other white papers are addressing key generic issues of the priority licensing topics in greater detail.
Environmental Health & Safety Sub Department Name 480 Oak Rd, Stanford, CA 94305 T 650.723.0448 F 650.725.3468 DEPUTY DIRECTOR, ENVIRONMENTAL HEALTH AND SAFETY Exempt, Full-Time (100% FTE) Posted May 1, 2014 The Department of Environmental Health and Safety (EH&S) at Stanford University seeks
Earth Sciences Safety Handbook
Cambridge, University of
Report of Earth Sciences Departmental Safety Committee 2011 - 12 5 Chemical Safety 21 - 22 Chemical Waste Assessment Hire Vehicle Checklist Department Driving Protocol: Bullard vehicles 38 - 48 Electrical Safety 24 and outside adjacent to areas which present a particular fire hazard. Persons wishing to smoke are asked to do
Earth Sciences Safety Handbook
Cambridge, University of
Report of Earth Sciences Departmental Safety Committee 2012 - 13 5 Chemical Safety 21 - 22 Chemical Waste Assessment Hire Vehicle Checklist Department Driving Protocol: Bullard vehicles 38 - 48 Electrical Safety 24 and outside adjacent to areas which present a particular fire hazard. Persons wishing to smoke are asked to do
Leak before break application in French PWR plants under operation
Faidy, C. [EDF SEPTEN, Villeurbanne (France)
Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.
Waste Isolation Pilot Plant Annual Site Environmental Report for 2012
The purpose of the Waste Isolation Pilot Plant (WIPP) Annual Site Environmental Report for 2012 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1B, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to: Characterize site environmental management performance; Summarize environmental occurrences and responses reported during the calendar year; Confirm compliance with environmental standards and requirements; Highlight significant environmental accomplishments, including progress toward the DOE Environmental Sustainability Goals made through implementation of the WIPP Environmental Management System (EMS).
September 2013 Laboratory Safety Manual Section 7 -Safety Training
Wilcock, William
September 2013 Laboratory Safety Manual Section 7 - Safety Training UW Environmental Health and Safety Page 7-1 Section 7 - Safety Training Contents A. SAFETY TRAINING REQUIREMENTS ......................................................7-1 B. EH&S SAFETY TRAINING AND RECORDS ..............................................7-1 C
Center for Intermodal Transportation Safety
Fernandez, Eduardo
Center for Intermodal Transportation Safety and Security Panagiotis Scarlatos, Ph.D., Director Transportation Safety and Security #12;Center for Intermodal Transportation Safety and Security Partners #12 evacuations · Tracking systems for hazardous materials Center for Intermodal Transportation Safety
Alpay, S. Pamir
-3113 Emergency maintenance to report a water leak, electrical outage, non-working fume hood, etc. after normal Radiation and Laser Safety 19 Laser Safety 21 Compressed Gas and Cryogenic Safety 22 Electrical Safety 24
Verification of voltage/frequency requirement for emergency diesel generator in nuclear power plant using dynamic modeling
Hur, Jin-Suk; Roh, Myung- Sub [KEPCO International Nuclear Graduate School, 1456-1 Shinam-ri, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of)
One major cause of the plant shutdown is the loss of electrical power. The study is to comprehend the coping action against station blackout including emergency diesel generator, sequential loading of safety system and to ensure that the emergency diesel generator should meet requirements, especially voltage and frequency criteria using modeling tool. This paper also considered the change of the sequencing time and load capacity only for finding electrical design margin. However, the revision of load list must be verified with safety analysis. From this study, it is discovered that new load calculation is a key factor in EDG localization and in-house capability increase.
Development of the seismic input for use in the seismic safety margins research program
Bernreuter, D.L.; Chung, D.H.
This paper briefly outlines the overall systems approach being developed for the Seismic Safety Margins Research Program. The unique features of the approach being taken to reduce the uncertainty in the seismic input for this program are discussed. These unique features will include extensive use of expert opinion, earthquake rupture simulation studies and the way in which the seismic hazard is incorporated into the overall systems analysis. Some very preliminary results are also given for the Zion site which is the power plant chosen for analysis in Phase I of the program.
Seismic Safety Margins Research Program (Phase I). Project IV. Structural building response; Structural Building Response Review
Healey, J.J.; Wu, S.T.; Murga, M.
As part of the Phase I effort of the Seismic Safety Margins Research Program (SSMRP) being performed by the University of California Lawrence Livermore Laboratory for the US Nuclear Regulatory Commission, the basic objective of Subtask IV.1 (Structural Building Response Review) is to review and summarize current methods and data pertaining to seismic response calculations particularly as they relate to the objectives of the SSMRP. This material forms one component in the development of the overall computational methodology involving state of the art computations including explicit consideration of uncertainty and aimed at ultimately deriving estimates of the probability of radioactive releases due to seismic effects on nuclear power plant facilities.
Sandia National Laboratories: Transportation Safety
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Experimental Testing Phenomenological Modeling Risk and Safety Assessment Cyber-Based Vulnerability Assessments Uncertainty Analysis Transportation Safety Fire Science Human...
Pressure Safety Program Implementation at ORNL
Lower, Mark [ORNL; Etheridge, Tom [ORNL; Oland, C. Barry [XCEL Engineering, Inc.
The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According to 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply with pressure safety requirements in 10 CFR 851. It also describes actions taken to develop and implement ORNL�s Pressure Safety Program.
The LBB methodology application results performed on the safety related piping of NPP V-1 in Jaslovske Bohunice
Kupca, L.; Beno, P. [Nuclear Power Plants Research Institute, Trnava (Slovakia)
A broad overview of the leak before break (LBB) application to the Slovakian V-1 nuclear power plant is presented in the paper. LBB was applied to the primary cooling circuit and surge lines of both WWER 440 type units, and also used to assess the integrity of safety related piping in the feed water and main steam systems. Experiments and calculations performed included analyses of stresses, material mechanical properties, corrosion, fatigue damage, stability of heavy component supports, water hammer, and leak rates. A list of analysis results and recommendations are included in the paper.
Final environmental impact statement. Waste Isolation Pilot Plant
Not Available
This volume contains the appendices for the Final Environmental Impact Statement for the Waste Isolation Pilot Plant (WIPP). Alternative geologic environs are considered. Salt, crystalline rock, argillaceous rock, and tuff are discussed. Studies on alternate geologic regions for the siting of WIPP are reviewed. President Carter's message to Congress on the management of radioactive wastes and the findings and recommendations of the interagency review group on nuclear waste management are included. Selection criteria for the WIPP site including geologic, hydrologic, tectonic, physicochemical compatability, and socio-economic factors are presented. A description of the waste types and the waste processing procedures are given. Methods used to calculate radiation doses from radionuclide releases during operation are presented. A complete description of the Los Medanos site, including archaeological and historic aspects is included. Environmental monitoring programs and long-term safety analysis program are described. (DMC)
Automatic safety rod for reactors
Germer, John H. (San Jose, CA)
An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.
Ferrocyanide Safety Program rationale for removing six tanks from the safety watch list
Borsheim, G.L.
This report documents an in-depth study of single-shell tanks containing ferrocyanide wastes. Topics include: safety assessments, tank histories, supportive documentation about interim stabilization and planned remedial activities.
Safety control circuit for a neutronic reactor
Ellsworth, Howard C. (Richland, WA)
A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.
High-temperature gas-cooled reactor safety studies for the division of accident evaluation. Quarterly progress report, October 1-December 31, 1982
Ball, S.J.; Clapp, N.E. Jr.; Cleveland, J.C.; Conklin, J.C.; Harrington, R.M.; Lindemer, T.B.; Siman-Tov, I.
Work continued on high-temperature gas-cooled reactor safety code development, including both the Fort St. Vrain and the 2240-MW(t) lead plant versions of the three-dimensional core code ORECA, the BLAST steam generator code, and a simplified core model code called SCORE. Oak Ridge National Laboratory participated in the Nuclear Regulatory Commission siting study for the lead plant with three other laboratories. Investigations continued to determine the status of fission-product source-term methodology applicable to postulated severe accidents.
Overview BETTER BUILDINGS, BETTER PLANTS
are available to Better Plants Partners on a facilitated basis. These resources include free energy audits for qualifying facilities, energy analysis software tools, and...
Lock open mechanism for downhole safety valve
Shirk, S.H.
This patent describes a wireline operable lock open mechanism for use in a downhole safety valve for a subterranean well conduit having an axially shiftable valve member movable downwardly between a closed and an open position relative to the conduit bore and biased to the closed position. The safety valve including a tubular housing assemblage defining an upwardly facing no-go shoulder in its upper end and a downwardly facing locking shoulder in its lower end above the valve member.
Review of Public Safety in Viewpoint of Complex Networks
Gai Chengcheng; Weng Wenguo; Yuan Hongyong [Center for Public Safety Research, Tsinghua university, Department of Engineering Physics, Tsinghua university (China)
In this paper, a brief review of public safety in viewpoint of complex networks is presented. Public safety incidents are divided into four categories: natural disasters, industry accidents, public health and social security, in which the complex network approaches and theories are need. We review how the complex network methods was developed and used in the studies of the three kinds of public safety incidents. The typical public safety incidents studied by the complex network methods in this paper are introduced, including the natural disaster chains, blackouts on electric power grids and epidemic spreading. Finally, we look ahead to the application prospects of the complex network theory on public safety.
Fire Safety Tests for Cesium-Loaded Spherical Resorcinol Formaldehyde Resin: Data Summary Report
Kim, Dong-Sang; Schweiger, Michael J.; Peterson, Reid A.
A draft safety evaluation of the scenario for spherical resorcinol formaldehyde (SRF) resin fire inside the ion exchange column was performed by the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Fire Safety organization. The result of this draft evaluation suggested a potential change of the fire safety classification for the Cesium Ion Exchange Process System (CXP) emergency elution vessels, equipment, and piping. To resolve this question, the fire properties of the SRF resin were measured by Southwest Research Institute (SwRI) through a subcontract managed by Pacific Northwest National Laboratory (PNNL). The results of initial fire safety tests on the SRF resin were documented in a previous report (WTP-RPT-218). The present report summarizes the results of additional tests performed by SwRI on the cesium-loaded SRF resin. The efforts by PNNL were limited to summarizing the test results provided by SwRI into one consolidated data report. The as-received SwRI report is attached to this report in the Appendix A. Where applicable, the precision and bias of each test method, as given by each American Society for Testing and Materials (ASTM) standard procedure, are included and compared with the SwRI test results of the cesium-loaded SRF resin.
Fire Safety Tests for Spherical Resorcinol Formaldehyde Resin: Data Summary Report
Kim, Dong-Sang; Peterson, Reid A.; Schweiger, Michael J.
A draft safety evaluation of the scenario for spherical resorcinol-formaldehyde (SRF) resin fire inside the ion exchange column was performed by the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Fire Safety organization. The result of this draft evaluation suggested a potential change of the fire safety classification for the Cesium Ion Exchange Process System (CXP) emergency elution vessels, equipment, and piping, which may be overly bounding based on the fire performance data from the manufacturer of the ion exchange resin selected for use at the WTP. To resolve this question, the fire properties of the SRF resin were measured by Southwest Research Institute (SwRI), following the American Society for Testing and Materials (ASTM) standard procedures, through a subcontract managed by Pacific Northwest National Laboratory (PNNL). For some tests, the ASTM standard procedures were not entirely appropriate or practical for the SRF resin material, so the procedures were modified and deviations from the ASTM standard procedures were noted. This report summarizes the results of fire safety tests performed and reported by SwRI. The efforts by PNNL were limited to summarizing the test results provided by SwRI into one consolidated data report. All as-received SwRI reports are attached to this report in the Appendix. Where applicable, the precision and bias of each test method, as given by each ASTM standard procedure, are included and compared with the SwRI test results of the SRF resin.
Evidence of aging effects on certain safety-related components
Magleby, H.L.; Atwood, C.L.; MacDonald, P.E.; Edson, J.L.; Bramwell, D.L. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)
In response to interest shown by the Nuclear Energy Agency (NEA), Principal Working Group I (PWG- 1) of the Committee on the Safety of Nuclear Installations (CSNI) conducted a generic study on the effects of aging of active components in nuclear power plants. (This focus on active components is consistent with PWG-l`s mandate; passive components are primarily within the mandate of PWG-3.) Representatives from France, Sweden, Finland, Japan, the United States, and the United Kingdom participated in the study by submitting reports documenting aging studies performed in their countries. This report consists of summaries of those reports, along with a comparison of the various statistical analysis methods used in the studies. The studies indicate that with some exceptions, active components generally do not present a significant aging problem in nuclear power plants. Design criteria and effective preventative maintenance programs, including timely replacement of components, are effective in mitigating potential aging problems. However, aging studies (such as qualitative and statistical analyses of failure modes and maintenance data) are an important part of efforts to identify and solve potential aging problems. Solving these problems typically includes such strategies as replacing suspect components with improved components, and implementing improved maintenance programs.
An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility
P. Calderoni; P. Sharpe; M. Shimada
The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.
Safety of Decommissioning of Nuclear Facilities
Batandjieva, B.; Warnecke, E.; Coates, R. [International Atomic Energy Agency, Vienna (Austria)
Full text of publication follows: ensuring safety during all stages of facility life cycle is a widely recognised responsibility of the operators, implemented under the supervision of the regulatory body and other competent authorities. As the majority of the facilities worldwide are still in operation or shutdown, there is no substantial experience in decommissioning and evaluation of safety during decommissioning in majority of Member States. The need for cooperation and exchange of experience and good practices on ensuring and evaluating safety of decommissioning was one of the outcomes of the Berlin conference in 2002. On this basis during the last three years IAEA initiated a number of international projects that can assist countries, in particular small countries with limited resources. The main IAEA international projects addressing safety during decommissioning are: (i) DeSa Project on Evaluation and Demonstration of Safety during Decommissioning; (ii) R{sup 2}D{sup 2}P project on Research Reactors Decommissioning Demonstration Project; and (iii) Project on Evaluation and Decommissioning of Former Facilities that used Radioactive Material in Iraq. This paper focuses on the DeSa Project activities on (i) development of a harmonised methodology for safety assessment for decommissioning; (ii) development of a procedure for review of safety assessments; (iii) development of recommendations on application of the graded approach to the performance and review of safety assessments; and (iv) application of the methodology and procedure to the selected real facilities with different complexities and hazard potentials (a nuclear power plant, a research reactor and a nuclear laboratory). The paper also outlines the DeSa Project outcomes and planned follow-up activities. It also summarises the main objectives and activities of the Iraq Project and introduces the R{sup 2}D{sup 2} Project, which is a subject of a complementary paper.
Reactor Safety Research Programs
Edler, S. K.
This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.
Reactor Safety Research Programs
Dotson, CW
This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.
Generic small modular reactor plant design.
Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.
This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.
A Procedure for Determination of Degradation Acceptance Criteria for Structures and Passive Components in Nuclear Power Plants
Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y-S.; Hahm, D.; Choi, I-K.
The Korea Atomic Energy Research Institute (KAERI) has been collaborating with Brookhaven National Laboratory since 2007 to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). This collaboration program aims at providing technical support to a five-year KAERI research project, which includes three specific areas that are essential to seismic probabilistic risk assessment: (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. The understanding and assessment of age-related degradations of structures, systems, and components and their impact on plant safety is the major goal of this KAERI-BNL collaboration. Four annual reports have been published before this report as a result of the collaboration research.
Steam Pressure-Reducing Station Safety and Energy Efficiency Improvement Project
Lower, Mark D [ORNL; Christopher, Timothy W [ORNL; Oland, C Barry [ORNL
The Facilities and Operations (F&O) Directorate is sponsoring a continuous process improvement (CPI) program. Its purpose is to stimulate, promote, and sustain a culture of improvement throughout all levels of the organization. The CPI program ensures that a scientific and repeatable process exists for improving the delivery of F&O products and services in support of Oak Ridge National Laboratory (ORNL) Management Systems. Strategic objectives of the CPI program include achieving excellence in laboratory operations in the areas of safety, health, and the environment. Identifying and promoting opportunities for achieving the following critical outcomes are important business goals of the CPI program: improved safety performance; process focused on consumer needs; modern and secure campus; flexibility to respond to changing laboratory needs; bench strength for the future; and elimination of legacy issues. The Steam Pressure-Reducing Station (SPRS) Safety and Energy Efficiency Improvement Project, which is under the CPI program, focuses on maintaining and upgrading SPRSs that are part of the ORNL steam distribution network. This steam pipe network transports steam produced at the ORNL steam plant to many buildings in the main campus site. The SPRS Safety and Energy Efficiency Improvement Project promotes excellence in laboratory operations by (1) improving personnel safety, (2) decreasing fuel consumption through improved steam system energy efficiency, and (3) achieving compliance with applicable worker health and safety requirements. The SPRS Safety and Energy Efficiency Improvement Project being performed by F&O is helping ORNL improve both energy efficiency and worker safety by modifying, maintaining, and repairing SPRSs. Since work began in 2006, numerous energy-wasting steam leaks have been eliminated, heat losses from uninsulated steam pipe surfaces have been reduced, and deficient pressure retaining components have been replaced. These improvements helped ORNL reduce its overall utility costs by decreasing the amount of fuel used to generate steam. Reduced fuel consumption also decreased air emissions. These improvements also helped lower the risk of burn injuries to workers and helped prevent shrapnel injuries resulting from missiles produced by pressurized component failures. In most cases, the economic benefit and cost effectiveness of the SPRS Safety and Energy Efficiency Improvement Project is reflected in payback periods of 1 year or less.
Hydrogen Technologies Safety Guide
Rivkin, C.; Burgess, R.; Buttner, W.
The purpose of this guide is to provide basic background information on hydrogen technologies. It is intended to provide project developers, code officials, and other interested parties the background information to be able to put hydrogen safety in context. For example, code officials reviewing permit applications for hydrogen projects will get an understanding of the industrial history of hydrogen, basic safety concerns, and safety requirements.
Oak Ridge National Laboratory Health and Safety Long-Range Plan: Fiscal years 1989--1995
Not Available
The health and safety of its personnel is the first concern of ORNL and its management. The ORNL Health and Safety Program has the responsibility for ensuring the health and safety of all individuals assigned to ORNL activities. This document outlines the principal aspects of the ORNL Health and Safety Long-Range Plan and provides a framework for management use in the future development of the health and safety program. Each section of this document is dedicated to one of the health and safety functions (i.e., health physics, industrial hygiene, occupational medicine, industrial safety, nuclear criticality safety, nuclear facility safety, transportation safety, fire protection, and emergency preparedness). Each section includes functional mission and objectives, program requirements and status, a summary of program needs, and program data and funding summary. Highlights of FY 1988 are included.
Hazardous-material accidents near nuclear power plants: an evaluation of analyses and approaches
Kot, C.A.; Lin, H.C.; van Erp, J.B.; Eichler, T.V.; Wiedermann, A.H.
The state of knowledge concerning postulated accidents involving offsite hazardous materials in the vicinity of nuclear power plants is critically evaluated. This effort is part of a study to analyze the potential effects of offsite hazards upon the safety of nuclear power plants and to develop a technical basis for the assessment of siting approaches. The evaluation includes consideration of data bases and statistics of hazardous materials and accidents involving them, deterministic aspects of possible material dispersion and threat environments, the susceptibility and vulnerability of vital plant systems, and a critical review of past licensing experience and regulatory practice with respect to these hazards. While many of the data bases and analysis methods exist for an adequate estimate of threat and plant response, this knowledge is not fully used and no comprehensive guidance has been developed. Siting of nuclear power plants relative to offsite hazardous materials is a risk based procedure that considers both probabilities and consequences of events that make up accident scenarios. In this context it appears feasible to improve the procedures vis-a-vis the perception of safety, economy of effort, and efficiency of implementation. A scenario dependent conditional risk approach is outlined as a possible means of improving the siting procedures.
Considerations in the evaluation of concrete structures for continued service in aged Nuclear Power Plants (NPPs)
Naus, D.; Marchbanks, M.; Oland, B.; Arndt, G.; Brown, T.
Currently, there are /approximately/119 commercial nuclear power plants (NPPs) in the US either under construction, operating at low-to-full power, or awaiting an operating license. Together, these units have a net generating capacity of /approximately/110 GW(e). Assuming no life extension of present facilities, the operating licenses for these plants will start to expire in the middle of the next decade with Yankee Rowe being the first plant to attain this status. Where it is noted that with no life extension of facilities, a potential loss of electrical generating capacity in excess of 75 GW(e) could occur during the time period 2006 to 2020 when the operating licenses of 80 to 90 NPPs are scheduled to expire. A potential timely and cost-effective solution to meeting future electricity demand, which has worked well for non-nuclear generating plants, is to extend the service life (operating licenses) of existing NPPs. Since the concrete components in these plants provide a vital safety function, any continued service considerations must include an in-depth assessment of the safety-related concrete structures. 7 refs.
Aviation Management and Safety
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
To establish a policy framework that will ensure safety, efficiency and effectiveness of government or contractor aviation operations. Cancels DOE O 440.2B.
Broader source: Energy.gov [DOE]
Dam safety in Florida is a shared responsibility among the Florida Department of Environmental Protection (FDEP), the regional water management districts, the United States Army Corps of Engineers ...
Broader source: Energy.gov (indexed) [DOE]
Battery Safety Testing Christopher J. Orendorff, Leigh Anna M. Steele, Josh Lamb, and Scott Spangler Sandia National Laboratories 2014 Energy Storage Annual Merit Review...
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
be continuously escorted by someone who has such training: The training consists of an eye exam, BNL general laser safety lecture, and formal ATF laser familiarization. Untrained...
Crow, W.
This document addresses safety concerns regarding the use of coiled tubing as it pertains to the preservation of personnel, environment and the wellbore.
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Safety Hazards of Batteries Battery technology is at the heart of much of our technological revolution. One of the most prevalent rechargeable batteries in use today is the...
Broader source: Energy.gov [DOE]
The Public Service Commission has the authority enact regulations pertaining to pipeline safety. These regulations address pipeline monitoring, inspections, enforcement, and penalties.
Intrastate Pipeline Safety (Minnesota)
Broader source: Energy.gov [DOE]
These regulations provide standards for gas and liquid pipeline maintenance and operating procedures, per the Federal Hazardous Liquid and Natural Gas Pipeline Safety Acts, and give the...
Pipeline Safety (South Dakota)
Broader source: Energy.gov [DOE]
The South Dakota Pipeline Safety Program, administered by the Public Utilities Commission, is responsible for regulating hazardous gas intrastate pipelines. Relevant legislation and regulations...
Broader source: Energy.gov [DOE]
This law authorizes the establishment of rules and regulations covering the design, fabrication, installation, inspection, testing and safety standards for installation, operation and maintenance...
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Manual describes DOE's explosives safety requirements applicable to operations involving the development, testing, handling, and processing of explosives or assemblies containing explosives.
Neural network recognition of nuclear power plant transients
Bartlett, E.B.; Danofsky, R.; Adams, J.; AlJundi, T.; Basu, A.; Dhanwada, C.; Kerr, J.; Kim, K.; Lanc, T.
The objective of this report is to describe results obtained during the first year of funding that will lead to the development of an artificial neural network (ANN) fault - diagnostic system for the real - time classification of operational transients at nuclear power plants. The ultimate goal of this three-year project is to design, build, and test a prototype diagnostic adviser for use in the control room or technical support center at Duane Arnold Energy Center (DAEC); such a prototype could be integrated into the plant process computer or safety - parameter display system. The adviser could then warn and inform plant operators and engineers of plant component failures in a timely manner. This report describes the work accomplished in the first of three scheduled years for the project. Included herein is a summary of the first year's results as, well as individual descriptions of each of the major topics undertaken by the researchers. Also included are reprints of the articles written under this funding as well as those that were published during the funded period.
Myths and representations in French nuclear history: The impact on decommissioning safety
Paris-Sud XI, Université de
Myths and representations in French nuclear history: The impact on decommissioning safety C. Martin. The decommissioning of many operational plants (whether because nuclear power is being withdrawn, or because plants accident at the Fukushima nuclear power plant has shown that in many countries the debate on the withdrawal
Proceedings of the international ANS/ENS topical meeting on thermal reactor safety
Not Available
This book presents the papers given at a conference on reactor safety. Topics considered at the conference included containment, fission product release, safety injection, reactor instrumentation, thermal-hydraulic analyses, reactor safety experiments, meltdowns, gas production rates, computerized simulation, occupational safety, human factors engineering, emergency plans, emergency core cooling systems, reactor licensing, after-heat removal, RHR systems, risk assessment, probability, fuel-cladding interactions, safety standards, and molten metal-water interactions.
Seismic Safety Margins Research Program: Phase II program plan (FY 83-FY 84)
Bohn, M.P.; Bernreuter, D.L.; Cover, L.E.; Johnson, J.J.; Shieh, L.C.; Shukla, S.N.; Wells, J.E.
The Seismic Safety Margins Research Program (SSMRP) is an NRC-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its goal is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from a commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis and design process and explicitly includes the uncertainties inherent in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. As currently planned, the SSMRP will be completed in September, 1984. This document presents the program plan for work to be done during the remainder of the program. In Phase I of the SSMRP, the necessary tools (both computer codes and data bases) for performing a detailed seismic risk analysis were identified and developed. Demonstration calculations were performed on the Zion Nuclear Power Plant. In the remainder of the program (Phase II) work will be concentrated on developing a simplified SSMRP methodology for routine probabilistic risk assessments, quantitative validation of the tools developed and application of the simplified methodology to a Boiling Water Reactor. (The Zion plant is a pressurized water reactor.) In addition, considerable effort will be devoted to making the codes and data bases easily accessible to the public.
Development of a Societal-Risk Goal for Nuclear Power Safety
Vicki Bier; Michael Corradini; Robert Youngblood; Caleb Roh; Shuji Liu
The safety-goal policy of the Nuclear Regulatory Commission (NRC) has never included a true societal-risk goal. The NRC did acknowledge that the original goal for the risk of latent cancer facilities �was an individual risk goal not related to the number of people involved,� and stated that �a true societal risk goal would place a limit on the aggregate number of people affected.� However, this limitation was never satisfactorily addressed. Moreover, the safety goal has historically focused primarily on fatalities and latent health effects, while experience with actual nuclear accidents has shown that societal disruption can be significant even in accidents that yield only small to modest numbers of fatalities. Therefore, we have evaluated the social disruption effects from severe reactor accidents as a basis to develop a societal-risk goal for nuclear power plants, considering both health effects and non-health concerns such as property damage and land interdiction. Our initial analysis considered six different nuclear power plant sites in the U.S. for Boiling Water Reactors and Pressurized Water Reactors. The accident sequences considered for these two reactor types were station blackout sequences (both short-term and long-term SBO) as well as an STSBO with RCIC failure for the BWR and a Steam Generator Tube Rupture for the PWR. The source term release was an input in a RASCAL calculation of the off-site consequences using actual site-based weather data for each of the six plant sites randomly selected over a two-year period. The source term release plumes were then compared to Geographical Information System data for each site to determine the population affected and that would need to be evacuated to meet current emergency preparedness regulations. Our results to date suggest that number of people evacuated to meet current protective action guidelines appears to be a good proxy for disruption -- and, unlike other measures of disruption, has the advantage of being relatively straightforward to calculate for a given accident scenario and a given geographical location and plant site. Revised safety goals taking into account the potential for societal disruption could in principle be applied to the current generation of nuclear plants, but could also be used in evaluating and siting new technologies, such as small modular light water reactors, advanced Gen-IV high-temperature reactors, as well as reactor designs with passive safety features such as filtered vented containments.
Savannah River Plant/Savannah River Laboratory radiation exposure report
Rogers, C.D.; Hyman, S.D.; Keisler, L.L. (Du Pont de Nemours (E.I.) and Co., Aiken, SC (USA). Savannah River Plant); Reeder, D.F.; Jolly, L.; Spoerner, M.T.; Schramm, G.R. (Du Pont de Nemours (E.I.) and Co., Aiken, SC (USA). Savannah River Lab.)
The protection of worker health and safety is of paramount concern at the Savannah River Site. Since the site is one of the largest nuclear sites in the nation, radiation safety is a key element in the protection program. This report is a compendium of the results in 1988 of the programs at the Savannah River Plant and the Savannah River Laboratory to protect the radiological health of employees. By any measure, the radiation protection performance at this site in 1988 was the best since the beginning of operations. This accomplishment was made possible by the commitment and support at all levels of the organizations to reduce radiation exposures to ALARA (As Low As Reasonably Achievable). The report provides detailed information about the radiation doses received by departments and work groups within these organizations. It also includes exposure data for recent years to allow Plant and Laboratory units to track the effectiveness of their ALARA efforts. Many of the successful practices and methods that reduced radiation exposure are described. A new goal for personnel contamination cases has been established for 1989. Only through continual and innovative efforts to minimize exposures can the goals be met. The radiation protection goals for 1989 and previous years are included in the report. 27 figs., 58 tabs.
Application of food and feed safety assessment principles to evaluate transgenic approaches to gene modulation in crops
Parrott, Wayne
safety of genetically modified (GM) crops has been assessed by the application of a set of internationally accepted procedures for evaluating the safety of GM crops. The goal of this paper is to review for conducting a safety assessment for GM crops that are developed by technologies that modify endogenous plant
Occupational Hygiene & Chemical Safety Division Department of Environmental Health & Safety
Machel, Hans
Occupational Hygiene & Chemical Safety Division Department of Environmental Health & Safety Risk all connections and fittings prior to start of anesthesia. Carefully pour Isoflurane from Environmental Health & Safety before re-entering the laboratory. REFERENCES 1. Procedure
Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report
OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.
Sandia Energy - Risk and Safety Assessment
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Risk and Safety Assessment Home Stationary Power Nuclear Fuel Cycle Nuclear Energy Safety Technologies Risk and Safety Assessment Risk and Safety AssessmentTara...
Office of Nuclear Facility Safety Programs
Broader source: Energy.gov [DOE]
The Office of Nuclear Facility Safety Programs establishes nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities.
Facility Safety - DOE Directives, Delegations, and Requirements
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Change, Safety, The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety,...
Cogeneration handbook for the petroleum refining industry. [Glossary included
Not Available
This Handbook deals only with industrial cogeneration, that is, simultaneous production of both heat and electricity at the industrial plant site. The cogenerator has the option of either selling all cogenerated power to the utility while simultaneously purchasing power to satisfy his plant demand, or directly supplying the plant demand with cogenerated power, thus displacing utility-supplied power. This Handbook provides the refinery plant manager or company energy coordinator with a framework for making a preliminary assessment of the feasibility and viability of cogeneration at a particular plant. The handbook is intended to provide an understanding of the potential of several standardized cogeneration systems, as well as their limitations. However, because the decision to cogenerate is very site specific, the handbook cannot provide all of the answers. It does attempt, however, to bring to light the major issues that should be addressed in the decision-making process. The decision of whether to cogenerate involves several considerations, including technical, economic, environmental, legal, and regulatory issues. Each of these issues is addressed separately in this handbook. In addition, a chapter is included on preparing a three-phase work statement, which is needed to guide the design of a cogeneration system. 39 figures, 37 tables.
Safety approaches for high power modular laser operation
Handren, R.T.
Approximately 20 years ago, a program was initiated at the Lawrence Livermore National Laboratory (LLNL) to study the feasibility of using lasers to separate isotopes of uranium and other materials. Of particular interest has been the development of a uranium enrichment method for the production of commercial nuclear power reactor fuel to replace current more expensive methods. The Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has progressed to the point where a plant-scale facility to demonstrate commercial feasibility has been built and is being tested. The U-AVLIS Program uses copper vapor lasers which pump frequency selective dye lasers to photoionize uranium vapor produced by an electron beam. The selectively ionized isotopes are electrostatically collected. The copper lasers are arranged in oscillator/amplifier chains. The current configuration consists of 12 chains, each with a nominal output of 800 W for a system output in excess of 9 kW. The system requirements are for continuous operation (24 h a day, 7 days a week) and high availability. To meet these requirements, the lasers are designed in a modular form allowing for rapid change-out of the lasers requiring maintenance. Since beginning operation in early 1985, the copper lasers have accumulated over 2 million unit hours at a >90% availability. The dye laser system provides approximately 2.5 kW average power in the visible wavelength range. This large-scale laser system has many safety considerations, including high-power laser beams, high voltage, and large quantities ({approximately}3000 gal) of ethanol dye solutions. The Laboratory`s safety policy requires that safety controls be designed into any process, equipment, or apparatus in the form of engineering controls. Administrative controls further reduce the risk to an acceptable level. Selected examples of engineering and administrative controls currently being used in the U-AVLIS Program are described.
Barrash, Warren
ANNUAL SECURITY AND FIRE SAFETY REPORT 2014 #12;2 Boise State University 2014 Annual Security and Fire Safety Report From the Vice President for Campus Operations and General Counsel At Boise State University, we are committed to providing a safe and secure environment for students, staff
Safety of Accelerator Facilities
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
To establish accelerator-specific safety requirements which, when supplemented by other applicable safety and health requirements, will serve to prevent injuries and illnesses associated with Department of Energy (DOE) or National Nuclear Security Administration (NNSA) accelerator operations. Cancels DOE O 420.2A. Certified 5-13-08. Canceled by DOE O 420.2C.
Safety of Accelerator Facilities
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The order defines accelerators and establishes accelerator specific safety requirements and approval authorities which, when supplemented by other applicable safety and health requirements, promote safe operations to ensure protection of workers, the public, and the environment. Cancels DOE O 420.2B.
Safety of Accelerator Facilities
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
To establish accelerator-specific safety requirements which, when supplemented by other applicable safety and health requirements, will serve to prevent injuries and illnesses associated with Department of Energy (DOE) or National Nuclear Security Administration (NNSA) accelerator operations. Cancels DOE O 420.2. Canceled by DOE O 420.2B.
Interdisciplinary Research and Training Program in the Plant Sciences
Wolk, C.P.
Research on plants continued. Topics include: Molecular basis of symbiotic plant-microbe interations; enzymatic mechanisms and regulation of plant cell wall biosynthesis; molecular mechanisms that regulate the expression of genes in plants; resistance of plants to environmental stress; studies on hormone biosynthesis and action; plant cell wall proteins; interaction of nuclear and organelle genomes; sensor transduction in plants; molecular mechanisms of trafficking in the plant cell; regulation of lipid metabolism; molecular bases of plant disease resistance mechanisms; biochemical and molecular aspects of plant pathogenesis; developmental biology of nitrogen-fixing cyanobacteria; environmental control of plant development and its relation to plant hormones.
Trends in HFE Methods and Tools and Their Applicability to Safety Reviews
O'Hara, J.M.; Plott, C.; Milanski, J.; Ronan, A.; Scheff, S.; Laux, L.; and Bzostek, J.
The U.S. Nuclear Regulatory Commission's (NRC) conducts human factors engineering (HFE) safety reviews of applicant submittals for new plants and for changes to existing plants. The reviews include the evaluation of the methods and tools (M&T) used by applicants as part of their HFE program. The technology used to perform HFE activities has been rapidly evolving, resulting in a whole new generation of HFE M&Ts. The objectives of this research were to identify the current trends in HFE methods and tools, determine their applicability to NRC safety reviews, and identify topics for which the NRC may need additional guidance to support the NRC's safety reviews. We conducted a survey that identified over 100 new HFE M&Ts. The M&Ts were assessed to identify general trends. Seven trends were identified: Computer Applications for Performing Traditional Analyses, Computer-Aided Design, Integration of HFE Methods and Tools, Rapid Development Engineering, Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. We assessed each trend to determine its applicability to the NRC's review by considering (1) whether the nuclear industry is making use of M&Ts for each trend, and (2) whether M&Ts reflecting the trend can be reviewed using the current design review guidance. We concluded that M&T trends that are applicable to the commercial nuclear industry and are expected to impact safety reviews may be considered for review guidance development. Three trends fell into this category: Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. The other trends do not need to be addressed at this time.
Porten, D.R.; Crowe, R.D.
The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall.
Occupational Hygiene & Chemical Safety Division Department of Environmental Health & Safety
Machel, Hans
Occupational Hygiene & Chemical Safety Division Department of Environmental Health & Safety Risk (most common � personal hygiene very important); d) storage � leaks; and e) waste � storage and disposal
KERENA safety concept in the context of the Fukushima accident
Zacharias, T.; Novotny, C.; Bielor, E. [AREVA NP GmbH, Paul-Gossen-Strasse 100, 91052 Erlangen (Germany)
Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basic physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)
Margin of Safety Definition and Examples Used in Safety Basis Documents and the USQ Process
Beaulieu, R. A.
The Nuclear Safety Management final rule, 10 CFR 830, provides an undefined term, margin of safety (MOS). Safe harbors listed in 10 CFR 830, Table 2, such as DOE?STD?3009 use but do not define the term. This lack of definition has created the need for the definition. This paper provides a definition of MOS and documents examples of MOS as applied in a U.S. Department of Energy (DOE) approved safety basis for an existing nuclear facility. If we understand what MOS looks like regarding Technical Safety Requirements (TSR) parameters, then it helps us compare against other parameters that do not involve a MOS. This paper also documents parameters that are not MOS. These criteria could be used to determine if an MOS exists in safety basis documents. This paper helps DOE, including the National Nuclear Security Administration (NNSA) and its contractors responsible for the safety basis improve safety basis documents and the unreviewed safety question (USQ) process with respect to MOS.
Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20
Not Available
This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.
Performance Evaluation of VeMAC Supporting Safety Applications in Vehicular Networks
Zhuang, Weihua
applications is to enhance the public safety standards and provide a safer environment for people on road. Safety Communications (VSC) project [9] is established by seven car manufacturers (including GM, BMW
CRAD, Facility Safety- Nuclear Facility Safety Basis
Broader source: Energy.gov [DOE]
A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Nuclear Facility Safety Basis.
CRAD, Facility Safety- Technical Safety Requirements
Broader source: Energy.gov [DOE]
A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Technical Safety Requirments (TSA).
COLUMBIA UNIVERSITY Radiation Safety Program
Jia, Songtao
COLUMBIA UNIVERSITY Radiation Safety Program Medical Center - T: 212-305-0303 F: 212 Psychiatric Institute Radiation Safety Office (Please complete this form within 24 hours and send a copy to your supervisor and The Radiation Safety Office) Your Name
Normalization of Process Safety Metrics
Wang, Mengtian
and organizational risks, there is an emerging need to evaluate the process safety implementation across an organization through measurements. Thus, the process safety metric is applied as a powerful tool that measures safety activities, status, and performance...
Digital controls for plant control systems at the Ulchin 5 and 6 NPP: Lessons for the U.S. NPP retrofits?
McCreary, T.; Stevens, J. [HF Controls Corp., 16650 Westgrove, Addison, TX 75001 (United States)
South Korea, through its nuclear operating arm, Korea Hydro and Nuclear Power, Ltd. (KHNP), has been one of the leading countries in the routine construction and operation of nuclear power plants. While standardizing on certain design approaches in the plant, KHNP continues to stress the need for the most reliable and cutting edge I and C technologies available for safety and non safety related NSSS and BOP functions. Its specifications are among the most challenging in the world with regard to the system performance and reliability requirements that must be met for the systems to be supplied. The Ulchin Units 5 and 6 (1000 MW each) nuclear power plants have been recently commissioned (2004 and 2005 respectively) with advanced programmable digital control systems in the NSSS and BOP (which includes both safety and non-safety applications). The systems supplied met what the authors believe are the most rigorous performance and functional specification criteria in use today. The authors will highlight a number of the specification requirements, the rationale for same and the plant experiences in the resultant supplied systems. As the existing U.S. nuclear fleet is faced with the prospect of upgrading legacy controls with digital controls, there are valuable lessons to be learned in the Ulchin 5 and 6 experiences that can be applied to retrofits and even new U.S. plant designs. Ulchin 5 and 6 is one of the success stories in the application of advanced digital technology to nuclear power plants. The specification focus that guided the selection and project execution for the digital controls supplied for Ulchin was on several key areas: a) a high reliable fault tolerant system, b)functional segmentation to minimize the failure effects on operation, c) validation and verification of the owner's engineer's application design (for regulatory compliance) as well as validation and verification of the system architecture and operation (to assure system performance met intended use); d)system performance criteria (system response time is critical to plant performance); e)global long term system support (HMI, control system components; communications); f)diversity (safety and non-safety; high integrity communication) The paper will review the key specification criteria, not only in context of a successful implementation for the Ulchin plant, but will also discuss applicability to the challenges of using digital controls in retrofit applications in U.S. nuclear power plants and in particular with regard to system performance and regulatory issues. (authors)
Seismic Safety Margins Research Program: a concluding look
Cummings, G.E.
The Seismic Safety Margins Research Program (SSMRP) was started in 1978 with the goal of developing tools and data bases to compute the probability of earthquake - caused radioactive release from commercial nuclear power plants. These tools and data bases were to help NRC to assess seismic safety at nuclear plants. The methodology to be used was finalized in 1982 and applied to the Zion Nuclear Power Station. The SSMRP will be completed this year with the development of a more simplified method of analysis and a demonstration of its use on Zion. This simplified method is also being applied to a boiling-water-reactor, LaSalle.
SNF fuel retrieval sub project safety analysis document
This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.
Expression of multiple proteins in transgenic plants
Vierstra, Richard D. (Madison, WI); Walker, Joseph M. (Madison, WI)
A method is disclosed for the production of multiple proteins in transgenic plants. A DNA construct for introduction into plants includes a provision to express a fusion protein of two proteins of interest joined by a linking domain including plant ubiquitin. When the fusion protein is produced in the cells of a transgenic plant transformed with the DNA construction, native enzymes present in plant cells cleave the fusion protein to release both proteins of interest into the cells of the transgenic plant. Since the proteins are produced from the same fusion protein, the initial quantities of the proteins in the cells of the plant are approximately equal.
Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model
Steven R. Sherman
The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant.
Potential application of neural networks to the operation of nuclear power plants
Uhrig, R.E. [University of Tennessee, Knoxville, TN (United States)]|[Oak Ridge National Laboratory, TN (United States)
The application of neural networks, a rapidly evolving technology used extensively in defense applications, to some of the problems of operating nuclear power plants is a logical complement to the expert systems currently being introduced in some of those plants. The potential applications of neural networks include, but are not limited to: (1) Diagnosing specific abnormal conditions. (2) Identifying nonlinear dynamics and transients. (3) Detecting the change of mode of operation. (4) Controlling temperature and pressure during start-up. (5) validating signals. (6) Plant-wide monitoring using autoassociative neural networks. (7) Monitoring of check valves. (8) Modeling the plant thermodynamics to increase efficiency. (9) Emulating core reload calculations. (10) Analyzing temporal sequences in the U.S. Nuclear Regulatory Commission Licensee Event Reports. (11) Monitoring plant parameters. (12) Analyzing vibrations in plants and rotating machinery. The work on such applications indicates that neural networks alone, or in conjunction with other advanced technologies, have the potential to enhance the safety, reliability, and operability of nuclear power plants. 36 refs.
Environmental Health and Safety Department
. Fire Safety, Radiation Safety and Hazardous Materials Facility are at other locations on campus Training Requirements If you work with: � Biological Materials � Chemical Agents � Radiological Materials
Safety Culture in Nuclear Installations
Broader source: Energy.gov [DOE]
IAEA-TECDOC-1329 Safety Culture in Nuclear Installations, Guidance for use in the Enhancement of Safety Culture, International Atomic Energy Agency IAEA, December 2002.
Industrial Safety | The Ames Laboratory
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
such as machine guarding, personal protective equipment (PPE), electrical safety, accident prevention and investigation, building design and code review, fire safety, and...
Gordon wins NNSA Safety Professional
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
in electrical safety at the Laboratory and across the DOE complex," said Industrial Hygiene and Safety manager Theresa Cull. "I am very pleased that NNSA has recognized Lloyd's...
Developed 2007 _____________________________ Environment, Health, & Safety _________ __________________
Eisen, Michael
_________ __________________ Training Program EHS 300~ Fiber optic Safety Course Syllabus Subject Category: Industrial Hygiene Course Alignment EH&S Website: Industrial Hygiene/Laser Safety Group - http
An overview of the US Department of Energy Plant Lifetime Improvement Program
Moonka, A.K. [Los Alamos National Lab., NM (United States); Harrison, D.L. [USDOE, Washington, DC (United States)
This paper provides a brief summary of the U.S. Department of Energy`s (USDOE`s) cooperative effort with the nuclear industry to develop technology to manage the effects of material degradation in systems, structures and components (SSCs) that impact plant safety or can significantly improve plant performance/economics and to establish and demonstrate the license renewal process. Also included are efforts to reduce decontamination/decommission costs, and reduce the uncertainty in long-term service-life decision making. During 1995, the Plant Lifetime Improvement (PLIM) Program was renamed the Commercial Operating Light Water Reactor (COLWR) Program activities are focused on sustaining the LWR option for domestic electricity generation by supporting operation of existing LWRs as long as they are safe, efficient, and economical. The status of the key projects is discussed in this paper.
Glutathione-S-conjugate transport in plants
Rea, Philip A. (Ardmore, PA); Lu, Yu-Ping (Havertown, PA); Li, Ze-Sheng (Prospect Park, PA)
The invention includes an isolated DNA encoding a plant GS-X pump polypeptide and an isolated preparation of a plant GS-X pump polypeptide. Also included is an isolated preparation of a nucleic acid which is antisense in orientation to a portion or all of a plant GS-X pump gene. The invention also includes a cells, vectors and transgenic plants having an isolated DNA encoding a plant GS-X pump and methods of use thereof. In addition, the invention relates to plant GS-X pump promoter sequences and the uses thereof.
Risk Informed Margins Management as part of Risk Informed Safety Margin Characterization
Curtis Smith
The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies.
Mirror Advanced Reactor Study (MARS). Final report. Volume 1-B. Commercial fusion electric plant
Donohue, M.L.; Price, M.E. (eds.)
Volume 1-B contains the following chapters: (1) blanket and reflector; (2) central cell shield; (3) central cell structure; (4) heat transport and energy conversion; (5) tritium systems; (6) cryogenics; (7) maintenance; (8) safety; (9) radioactivity, activation, and waste disposal; (10) instrumentation and control; (11) balance of plant; (12) plant startup and operation; (13) plant availability; (14) plant construction; and (15) economic analysis.
Safety Training for the Hydrogen Economy
Fassbender, Linda L.; Kinzey, Bruce R.; Akers, Bret M.
PNNL and the Volpentest Hazardous Materials Management and Emergency Response (HAMMER) Training and Education Center are helping to prepare emergency responders and permitting/code enforcement officials for their respective roles in the future Hydrogen Economy. Safety will be a critical component of the anticipated hydrogen transition. Public confidence goes hand in hand with perceived safety to such an extent that, without it, the envisioned transition is unlikely to occur. Stakeholders and the public must be reassured that hydrogen, although very different from gasoline and other conventional fuels, is no more dangerous. Ensuring safety in the hydrogen infrastructure will require a suitably trained emergency response force for containing the inevitable incidents as they occur, coupled with knowledgeable code officials to ensure that such incidents are kept to a minimum. PNNL and HAMMER are, therefore, designing a hydrogen safety training program, funded by DOE's Hydrogen, Fuel Cells, and Infrastructure Technologies Program, and modeled after the Occupational Safety and Health Administration�s multi-tiered approach to hazardous materials training. Capabilities under development at HAMMER include classroom and long-distance (i.e., satellite and internet broadcast) learning, as well as life-size, hands-on hydrogen burn props for �training as real as it gets.� This paper presents insights gained from the early emergency response hydrogen safety training courses held in 2005 and current plans for design and construction of a number of hydrogen burn props.
Radiation Safety Training Basic Radiation Safety Training for
Dai, Pengcheng
Radiation Safety Training Basic Radiation Safety Training for X-ray Users for Physics 461 & 462 Protocol Title: Basic Radiation Safety Training for X-ray Users Drafted By: Chris Millsaps, RSS Reviewers: ZB, TU, GS Purpose: To provide basic radiation safety training to the users of x-ray producing
Safety Criteria and Safety Lifecycle for Artificial Neural Networks
Kelly, Tim
Safety Criteria and Safety Lifecycle for Artificial Neural Networks Zeshan Kurd, Tim Kelly and Jim. The paper also presents a safety lifecycle for artificial neural networks. This lifecycle focuses, knowledge. INTRODUCTION Artificial neural networks (ANNs) are used in many safety-related applications
ES&H Safety Walk Around Check List
Knowles, David William
ES&H Safety Walk Around Check List for Managers Lab Work Environment (Labs & Storage Areas) Are all Waste x4644 Health Services x6266 Industrial & Shop Safety, including Machine Guarding x7170 Radiation shields or face shield) available for dispensing super cold liquids (cryogens like liquid nitrogen
Construction safety program for the National Ignition Facility, Appendix B
Cerruti, S.J.
This Appendix contains material from the LLNL Health and Safety Manual as listed below. For sections not included in this list, please refer to the Manual itself. The areas covered are: asbestos, lead, fire prevention, lockout, and tag program confined space traffic safety.
Implementation Guide for Use in Addressing Unreviewed Safety Question Requirements
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Guide provides information to assist in the implementation and interpretation of Title 10 Code of Federal Regulations (CFR) Part 830.203, Unreviewed Safety Question Process, of the Nuclear Safety Management Rules for applicable nuclear facilities owned or operated by the Department of Energy (DOE), including the National Nuclear Security Administration (NNSA). Canceled by DOE N 251.93.
Implementation Guide for Use in Addressing Unreviewed Safety Question Requirements
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Guide provides information to assist in implementation of Title 10 Code of Federal Regulations Part 830.203, "Unreviewed Safety Question Process," of the Nuclear Safety Management Rules for Category 1, 2, and 3 nuclear facilities owned or operated by the Department of Energy, including the National Nuclear Security Administration. Cancels DOE G 424.1-1.
Laser Safety Management Policy Statement ............................................................................................................1
Davidson, Fordyce A.
Laser Safety Management Policy Statement...........................................................2 Laser Users.............................................................................................................2 Unit Laser Safety Officer (ULSO
East Carolina University ENVIRONMENTAL SAFETY COMMITTEE
as workers' compensation, accident prevention, industrial hygiene, occupational safety, fire and life safety
Aviation Management and Safety
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
To establish a policy framework that will ensure safety, efficiency and effectiveness of government or contractor aviation operations. Cancels DOE O 440.2B. Admin Chg 1, dated 6-22-11, cancels DOE O 440.2C.
Dam Safety Regulation (Mississippi)
Broader source: Energy.gov [DOE]
The purpose of the Dam Safety Regulation is to ensure that all dams constructed in the state of Mississippi are permitted and thus do not potentially harm wildlife, water supplies and property. ...
Broader source: Energy.gov [DOE]
The purpose of the High Voltage Safety Act is to prevent injury to persons and property and interruptions of utility service resulting from accidental or inadvertent contact with high-voltage...
Packaging and Transportation Safety
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Establishes safety requirements for the proper packaging and transportation of offsite shipments and onsite transfers of hazardous materials andor modal transport. Cancels DOE 1540.2 and DOE 5480.3
Packaging and Transportation Safety
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Canceled by DOE 460.1A
Packaging and Transportation Safety
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Establishes safety requirements for the proper packaging and transportation of Department of Energy (DOE) offsite shipments and onsite transfers of hazardous materials and for modal transport. Cancels DOE O 460.1.
Promulgating Nuclear Safety Requirements
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
Applies to all Nuclear Safety Requirements Adopted by the Department to Govern the Conduct of its Nuclear Activities. Cancels DOE P 410.1. Canceled by DOE N 251.85.
Broader source: Energy.gov [DOE]
DOE solar reliability and safety research and development (R&D) focuses on testing photovoltaic (PV) modules, inverters, and systems for long-term performance, and helping investors, consumers,...
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, or successor directive, for routine and planned nuclear explosive operations (NEOs).
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs).
Complete Experiment Safety Documentation
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Safety Sheet If you did not submit a General User Proposal, you must submit an ESS one month prior to arrival at the ALS. 2. Biological, Radioactive, and Hazardous...
Broader source: Energy.gov [DOE]
North Carolina Administrative Code Title 15A, Subchapter 2K lays out further regulations for the design, approval, construction, maintenance, and inspection of dams to ensure public safety and...
Safety | Argonne National Laboratory
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Safety Argonne National Laboratory and the U.S. Department of Energy (DOE) are very concerned about the well-being of all employees. Students at the undergraduate and graduate...
Hutcheon, N. B.
Building codes are essentially sets of safety regulations in respect of structure, fire, and health. They were originally developed in response to frequently demonstrated hazards of structural collapse, catastrophic fires, ...
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
The directive provides supplemental details to support the requirements of DOE O 452.2C, Nuclear Explosive Safety, dated 6-12-06. Canceled by DOE M 452.2-1A.
Technical basis for evaluating electromagnetic and radio-frequency interference in safety-related I&C systems
Ewing, P.D.; Korsah, K. [Oak Ridge National Lab., TN (United States)
This report discusses the development of the technical basis for the control of upsets and malfunctions in safety-related instrumentation and control (I&C) systems caused by electromagnetic and radio-frequency interference (EMI/RFI) and power surges. The research was performed at the Oak Ridge National Laboratory (ORNL) and was sponsored by the USNRC Office of Nuclear Regulatory Research (RES). The motivation for research stems from the safety-related issues that need to be addressed with the application of advanced I&C systems to nuclear power plants. Development of the technical basis centered around establishing good engineering practices to ensure that sufficient levels of electromagnetic compatibility (EMC) are maintained between the nuclear power plant`s electronic and electromechanical systems known to be the source(s) of EMI/RFI and power surges. First, good EMC design and installation practices need to be established to control the impact of interference sources on nearby circuits and systems. These EMC good practices include circuit layouts, terminations, filtering, grounding, bonding, shielding, and adequate physical separation. Second, an EMI/RFI test and evaluation program needs to be established to outline the tests to be performed, the associated test methods to be followed, and carefully formulated acceptance criteria based on the intended environment to ensure that the circuit or system under test meets the recommended guidelines. Third, a program needs to be developed to perform confirmatory tests and evaluate the surge withstand capability (SWC) and of I&C equipment connected to or installed in the vicinity of power circuits within the nuclear power plant. By following these three steps, the design and operability of safety-related I&C systems against EMI/RFI and power surges can be evaluated, acceptance criteria can be developed, and appropriate regulatory guidance can be provided.
Construction safety program for the National Ignition Facility, Appendix A
Cerruti, S.J.
Topics covered in this appendix include: General Rules-Code of Safe Practices; 2. Personal Protective Equipment; Hazardous Material Control; Traffic Control; Fire Prevention; Sanitation and First Aid; Confined Space Safety Requirements; Ladders and Stairways; Scaffolding and Lift Safety; Machinery, Vehicles, and Heavy Equipment; Welding and Cutting-General; Arc Welding; Oxygen/Acetylene Welding and Cutting; Excavation, Trenching, and Shoring; Fall Protection; Steel Erection; Working With Asbestos; Radiation Safety; Hand Tools; Electrical Safety; Nonelectrical Work Performed Near Exposed High-Voltage Power-Distribution Equipment; Lockout/Tagout Requirements; Rigging; A-Cranes; Housekeeping; Material Handling and Storage; Lead; Concrete and Masonry Construction.
Reactor Safety Research: Semiannual report, July-December 1986
Not Available
Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.
Hearn, J; James Lightner, J
The objective of this paper is to share the Savannah River Site lessons learned on Safeguards and Security (S&S) program integration with K-Area Complex (KAC) safety basis. The KAC Documented Safety Analysis (DSA), is managed by the Washington Savannah River Company (WSRC), and the S&S program, managed by Wackenhut Services, Incorporated--Savannah River Site (WSI-SRS). WSRC and WSI-SRS developed a contractual arrangement to recognize WSI-SRS requirements in the KAC safety analysis. Design Basis Threat 2003 (DBT03) security upgrades required physical modifications and operational changes which included the availability of weapons which could potentially impact the facility safety analysis. The KAC DSA did not previously require explicit linkage to the S&S program to satisfy the safety analysis. WSI-SRS have contractual requirements with the Department of Energy (DOE) which are separate from WSRC contract requirements. The lessons learned will include a discussion on planning, analysis, approval of the controls and implementation issues.
R.J. Garrett
As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.
Hydraulic limits on maximum plant transpiration and the emergence of the safetyefficiency trade-off
Jackson, Robert B.
Hydraulic limits on maximum plant transpiration and the emergence of the safety�efficiency trade.12126 Key words: hydraulic limitation, safety� efficiency trade-off, soil�plant�atmosphere model, trait hydraulics constrain ecosystem productivity by setting physical limits to water transport and hence carbon
EM Health and Safety Plan Guidelines
Not Available
This document contains information about the Health and Safety Plan Guidelines. Topics discussed include: Regulatory framework; key personnel; hazard assessment; training requirements; personal protective equipment; extreme temperature disorders or conditions; medical surveillance; exposure monitoring/air sampling; site control; decontamination; emergency response/contingency plan; emergency action plan; confined space entry; and spill containment.
Radiation Safety Annual Refresher Training
Thomas, David D.
Radiation Safety Annual Refresher Training Radiation Protection Division Department of Environmental Health & Safety #12;Topics in Radiation Safety (applicable RPD Manual sections indicated) User;Topics in Radiation Safety (applicable RPD Manual sections indicated) User and Non-user topics Types
Toolbox Safety Talk Lead Awareness
Pawlowski, Wojtek
Toolbox Safety Talk Lead Awareness Environmental Health & Safety Facilities Safety & Health Section Health & Safety for recordkeeping. Lead based paint is commonly found in homes built before 1978 and many industrial paints today still contain lead. Lead overexposure is one of the leading causes of workplace
School of Forest Safety Training
Thomas, Andrew
) � Laboratories: � Material Safety Data Sheets (MSDS) for chemicals � Chemical hygiene plan (CHP) manual
Waste Isolation Pilot Plant Annual Site Environmental Report for 2010
The purpose of the Waste Isolation Pilot Plant (WIPP) Annual Site Environmental Report for 2010 (ASER) is to provide information required by U.S. Department of Energy (DOE) Order 231.1A, Environment, Safety, and Health Reporting. Specifically, the ASER presents summary environmental data to: (1) Characterize site environmental management performance. (2) Summarize environmental occurrences and responses reported during the calendar year. (3) Confirm compliance with environmental standards and requirements. (4) Highlight significant environmental accomplishments, including progress toward the DOE Environmental Sustainability Goals made through implementation of the WIPP Environmental Management System (EMS). The DOE Carlsbad Field Office (CBFO) and the management and operating contractor (MOC), Washington TRU Solutions LLC (WTS), maintain and preserve the environmental resources at the WIPP. DOE Order 231.1A; DOE Order 450.1A, Environmental Protection Program; and DOE Order 5400.5, Radiation Protection of the Public and the Environment, require that the affected environment at and near DOE facilities be monitored to ensure the safety and health of the public and workers, and preservation of the environment. This report was prepared in accordance with DOE Order 231.1A, which requires that DOE facilities submit an ASER to the DOE Headquarters Chief Health, Safety, and Security Officer. The WIPP Hazardous Waste Facility Permit Number NM4890139088-TSDF (Permit) further requires that the ASER be provided to the New Mexico Environment Department (NMED).
Fusion safety program Annual report, Fiscal year 1995
Longhurst, G.R.; Cadwallader, L.C.; Carmack, W.J. [and others
This report summarizes the major activities of the Fusion Safety Program in FY-95. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory, and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions. Among the technical areas covered in this report are tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate database development, and safety code development and application to fusion safety issues. Most of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and the technical support for commercial fusion facility conceptual design studies. A final activity described is work to develop DOE Technical Standards for Safety of Fusion Test Facilities.
Wedde, Horst F.
-critical systems (such as nuclear power plants, distributed cooperation of autonomous robots in Outer Space that are typi- cally unpredictable, a very high amount of adaptability of sys- tem functions is demanded. SafetyDISTRIBUTED REAL-TIME TASK MONITORING IN THE SAFETY-CRITICAL SYSTEM MELODY Horst F. Wedde, Jon A
TESEC 2001, Genova, Italy ADVANCED TECHNIQUES FOR SAFETY ANALYSIS APPLIED TO
Tronci, Enrico
TESEC 2001, Genova, Italy 1 ADVANCED TECHNIQUES FOR SAFETY ANALYSIS APPLIED TO THE GAS TURBINE for safety analysis of complex computer based systems. Such approaches are applied to the gas turbine control and electrical power supply of the centre of ENEA CR Casaccia. The plant is based on a small gas turbine and has
Tiger Team Assessment of the Pantex Plant, Amarillo, Texas
Not Available
This document contains the findings and associated root causes identified during the Tiger Team Assessment of the Department of Energy's (DOE) Pantex Plant in Amarillo, Texas. This assessment was conducted by the Department's Office of Environment, Safety and Health between October 2 and 31, 1989. The scope of the assessment of the Pantex Plant covered all areas of environment, safety and health (ES H) activities, including compliance with federal, state, and local regulations, requirements, permits, agreements, orders and consent decrees, and DOE ES H Orders. The assessment also included an evaluation of the adequacy of DOE and site contractor ES H management programs. The draft findings were submitted to the Office of Defense Programs, the Albuquerque Operations Office, the Amarillo Area Office, and regulatory agencies at the conclusion of the on-site assessment activities for review and comment on technical accuracy. Final modifications and any other appropriate changes have been incorporated in the final report. The Tiger Team Assessment of the Pantex Plant is part of the larger Tiger Team Assessment program which will encompass over 100 DOE operating facilities. The assessment program is part of a 10-point initiative announced by Secretary of Energy James D. Watkins on June 27, 1989, to strengthen environmental protection and waste management activities in the Department. The results of the program will provide the Secretary with information on the compliance status of DOE facilities with regard to ES H requirements, root causes for noncompliance, adequacy of DOE and site contractor ES H management programs, and DOE-wide ES H compliance trends.
Unreviewed safety question evaluation of 100K East and 100K West in-basin fuel characterization program activities
Alwardt, L.D.
The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation of the 105K East (KE) and 105K West (KW) in-basin activities associated with the fuel characterization program as described in the characterization shipping plan. The significant activities that are common to both 105 KE and 105 KW basins are the movement of canisters from their main basin storage locations (or potentially from the 105 KE Tech View Pit if a dump table is available) to the south loadout pit transfer channel, hydrogen generation testing in the single element fuel container, loading the single element fuel container into the shipping cask, loading of the shipping cask onto a flat-bed trailer, return of the test fuel elements or element pieces from the 327 facility, placement of the fuel elements back into Mark 2 canisters, and placement of the canisters in the main storage basin. Decapping of canisters in the south loadout pit transfer channel and re-encapsulation of canisters are activities specific to the 105 KW basin. The scope of this safety evaluation includes only those characterization fuel shipment activities performed in the 105 KE and 105 KW fuel storage basin structures up to installation of the overpack. The packaging safety evaluation report governs the shipment of the fuel elements. The K Basins Plant Review Committee has determined that the in-basin activities associated with the fuel characterization program fuel shipments are bounded by the current safety envelop and do not constitute an unreviewed safety question. This determination is documented on Attachment 1.
Safety Assessment for the Kozloduy National Disposal Facility in Bulgaria - 13507
Biurrun, E.; Haverkamp, B. [DBE TECHNOLOGY GmbH, Eschenstr. 55, D-31224 Peine (Germany)] [DBE TECHNOLOGY GmbH, Eschenstr. 55, D-31224 Peine (Germany); Lazaro, A.; Miralles, A. [Westinghouse Electric Spain SAR, Padilla 17, E-28006 Madrid (Spain)] [Westinghouse Electric Spain SAR, Padilla 17, E-28006 Madrid (Spain); Stefanova, I. [SERAW, 52 A Dimitrov Blvd, 6 Fl., 1797 Sofia (Bulgaria)] [SERAW, 52 A Dimitrov Blvd, 6 Fl., 1797 Sofia (Bulgaria)
Due to the early decommissioning of four Water-Water Energy Reactors (WWER) 440-V230 reactors at the Nuclear Power Plant (NPP) near the city of Kozloduy in Bulgaria, large amounts of low and intermediate radioactive waste will arise much earlier than initially scheduled. In or-der to manage the radioactive waste from the early decommissioning, Bulgaria has intensified its efforts to provide a near surface disposal facility at Radiana with the required capacity. To this end, a project was launched and assigned in international competition to a German-Spanish consortium to provide the complete technical planning including the preparation of the Intermediate Safety Assessment Report. Preliminary results of operational and long-term safety show compliance with the Bulgarian regulatory requirements. The long-term calculations carried out for the Radiana site are also a good example of how analysis of safety assessment results can be used for iterative improvements of the assessment by pointing out uncertainties and areas of future investigations to reduce such uncertainties in regard to the potential radiological impact. The computer model used to estimate the long-term evolution of the future repository at Radiana predicted a maximum total annual dose for members of the critical group, which is carried to approximately 80 % by C-14 for a specific ingestion pathway. Based on this result and the outcome of the sensitivity analysis, existing uncertainties were evaluated and areas for reasonable future investigations to reduce these uncertainties were identified. (authors)
A study of improved MHR 50/100 toward a higher level of inherent safety
Minatsuki, I.; Otani, T.; Shimizu, K. [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Oyama, S.; Tsukamoto, H. [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-Chome, Hyogo-ku, Kobe (Japan); Kunitomi, K. [Japan Atomic Energy Agency, 4002 Narita, Oarai, Ibaraki (Japan)
A new concept of the Mitsubishi small-sized High temperature gas-cooled modular Reactors (MHR-50/100) had been developed and published in papers. The study results of the first design concept show that the MHR-50/100 can achieve the inherent safety level set as a design target in case of water ingress during steam generator tube rupture accident. And more specifically, the reactor was shown to remain stable during long-term station black out (SBO) with protection of only passive devices during a depressurization accident and with additional motion of steam dump system during a water ingress accident. Recently greater requirements for safety of future nuclear plants including the MHR-50/100 have been expected. This study has thus made a key design improvement for the MHR-50/100 in order to secure the inherent safety aspect without reliance on active steam dump system in case of a water ingress accident. The innovative technologies listed below have been created and investigated to achieve the improved MHR-50/100 design; (1) Design improvement of steam generator, (2) Heat balance optimization of steam cycle, (3) Control system design of differential pressure between primary helium gas and water/steam, (4) Study on operation procedure during a water ingress accident. (authors)
Plant Breeding Program COLLEGE OF AGRICULTURAL
Bradford, Kent
Plant Breeding Program COLLEGE OF AGRICULTURAL AND ENVIRONMENTAL SCIENCES Office of the Dean Cereal Breeding Program 51 Acknowlegements 51 COLLEGE OF AGRICULTURAL AND ENVIRONMENTAL SCIENCES Office in production agriculture, which included plant breeding, was necessary for California farmers to thrive
Ding, Biao
This grant supported the Second International Conference on Plant Vascular Biology (PVB 2010) held July 24-28, 2010 on the campus of Ohio State University, Columbus, Ohio. Biao Ding (Ohio State University; OSU) and David Hannapel (Iowa State University; ISU) served as co-chairs of this conference. Biao Ding served as the local organizer. PVB is defined broadly here to include studies on the biogenesis, structure and function of transport systems in plants, under conditions of normal plant growth and development as well as of plant interactions with pathogens. The transport systems cover broadly the xylem, phloem, plasmodesmata and vascular cell membranes. The PVB concept has emerged in recent years to emphasize the integrative nature of the transport systems and approaches to investigate them.
Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2. International Agreement Report
Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P. [Asociacion Nuclear Asco, Barcelona (Spain)
The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ``Pressurizer spray valve faulty opening`` presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data.
Station Blackout: A case study in the interaction of mechanistic and probabilistic safety analysis
Curtis Smith; Diego Mandelli; Cristian Rabiti
The ability to better characterize and quantify safety margins is important to improved decision making about nuclear power plant design, operation, and plant life extension. As research and development (R&D) in the light-water reactor (LWR) Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a �station blackout� wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario.
Hybrid Wet/Dry Cooling for Power Plants (Presentation)
Kutscher, C.; Buys, A.; Gladden, C.
This presentation includes an overview of cooling options, an analysis of evaporative enhancement of air-cooled geothermal power plants, field measurements at a geothermal plant, a preliminary analysis of trough plant, and improvements to air-cooled condensers.
Environment, Safety, and Health Risk Assessment Program (ESHRAP)
Eide, Steven Arvid; Thomas Wierman
The Environment, Safety and Health Risk Assessment Program (ESHRAP) models human safety and health risk resulting from waste management and environmental restoration activities. Human safety and health risks include those associated with storing, handling, processing, transporting, and disposing of radionuclides and chemicals. Exposures to these materials, resulting from both accidents and normal, incident-free operation, are modeled. In addition, standard industrial risks (falls, explosions, transportation accidents, etc.) are evaluated. Finally, human safety and health impacts from cleanup of accidental releases of radionuclides and chemicals to the environment are estimated. Unlike environmental impact statements and safety analysis reports, ESHRAP risk predictions are meant to be best estimate, rather than bounding or conservatively high. Typically, ESHRAP studies involve risk predictions covering the entire waste management or environmental restoration program, including such activities as initial storage, handling, processing, interim storage, transportation, and final disposal. ESHRAP can be used to support complex environmental decision-making processes and to track risk reduction as activities progress.
Mullins, Dyche
for environmental management and applied research. Areas of expertise include industrial hygiene and biosafety (e.g., industrial hygiene and toxicology, non-radioactive chemical and laboratory safety, asbestos management
HAZARDOUS MATERIAL SAFETY Effective Date: January 1, 1992
Cui, Yan
to Hazardous Chemicals in Laboratories, as noted in Subject H. Laboratory Safety. Items in the CHP include hazardous waste (see sample CHP for definitions), it is subject to the RCRA generator rules which are found
Environment/Health/Safety (EHS): Laser Safety
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
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A realistic anticipated transient without scram evaluation of the Zorita nuclear power plant
Rebollo, L. (Union Fenosa, Madrid (Spain))
A best-estimate methodology for analysis of an anticipated transient without scram (ATWS) in a pressurized water reactor (PWR) is applied to the simulation of the passive response to postulated ATWS scenarios of the Jose Cabrera (Zorita) nuclear power plant (NPP) owned and operated by Union Fenosa, which is the only Westinghouse PWR with a single coolant loop. A justification of the calculation hypotheses is included. The results of the specific studies are evaluated, and the conclusion is that the intrinsic safety margins of the original design of the plant guarantees the integrity of the fuel, primary circuit, and containment, without the need to incorporate an automatic ATWS mitigation system. Finally, a suitable plant-specific prototype emergency operating procedure is designed that is substantially different from the previous Zorita NPP procedure and from the generic procedure applicable to multiloop plants. This procedure is validated by simulating the operator-plant interface by means of a validation matrix including the scenarios presenting the most adverse dynamic modes foreseeable.
Inertial fusion energy: A clearer view of the environmental and safety perspectives
Latkowski, J.F.
If fusion energy is to achieve its full potential for safety and environmental (S&E) advantages, the S&E characteristics of fusion power plant designs must be quantified and understood, and the resulting insights must be embodied in the ongoing process of development of fusion energy. As part of this task, the present work compares S&E characteristics of five inertial and two magnetic fusion power plant designs. For each design, a set of radiological hazard indices has been calculated with a system of computer codes and data libraries assembled for this purpose. These indices quantify the radiological hazards associated with the operation of fusion power plants with respect to three classes of hazard: accidents, occupational exposure, and waste disposal. The three classes of hazard have been qualitatively integrated to rank the best and worst fusion power plant designs with respect to S&E characteristics. From these rankings, the specific designs, and other S&E trends, design features that result in S&E advantages have been identified. Additionally, key areas for future fusion research have been identified. Specific experiments needed include the investigation of elemental release rates (expanded to include many more materials) and the verification of sequential charged-particle reactions. Improvements to the calculational methodology are recommended to enable future comparative analyses to represent more accurately the radiological hazards presented by fusion power plants. Finally, future work must consider economic effects. Trade-offs among design features will be decided not by S&E characteristics alone, but also by cost-benefit analyses. 118 refs., 35 figs., 35 tabs.
Not Available
The Department of Energy (DOE) policy requires that all DOE activities be conducted in a manner that protects the safety of the public and provides a safe and healthful workplace for employees. DOE has also prescribed that all personnel be protected in any explosives operation undertaken. The level of safety provided shall be at least equivalent to that of the best industrial practice. The risk of death or serious injury shall be limited to the lowest practicable minimum. DOE and contractors shall continually review their explosives operations with the aim of achieving further refinements and improvements in safety practices and protective features. This manual describes the Department's explosive safety requirements applicable to operations involving the development, testing, handling, and processing of explosives or assemblies containing explosives. It is intended to reflect the state-of-the-art in explosives safety. In addition, it is essential that applicable criteria and requirements for implementing this policy be readily available and known to those responsible for conducting DOE programs.
Student manual, Book 2: Orientation to occupational safety compliance in DOE
Colley, D.L.
This is a student hand-book an Occupational Safety Compliance in DOE. Topics include the following: Electrical; materials handling & storage; inspection responsibilities & procedures; general environmental controls; confined space entry; lockout/tagout; office safety, ergonomics & human factors; medical & first aid, access to records; construction safety; injury/illness reporting system; and accident investigation procedures.
The Goal Structuring Notation A Safety Argument Notation Tim Kelly and Rob Weaver
Kelly, Tim
industries (including defence, aerospace, nuclear and railways). Studying the safety standards and guidanceThe Goal Structuring Notation � A Safety Argument Notation Tim Kelly and Rob Weaver firstname.lastname@example.com, firstname.lastname@example.com Abstract In Europe, over recent years, the responsibility for ensuring system safety
Implementation Guide for Use in Addressing Unreviewed Safety Question Requirements
Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]
This Guide, including its attachments, provides information to assist in the implementation of Title 10 Code of Federal Regulations (CFR) Section 830.203, �Unreviewed Safety Question Process,� of the Nuclear Safety Management Rules for Category 1, 2, and 3 nuclear facilities owned or operated by the Department of Energy (DOE), including the National Nuclear Security Administration (NNSA). Cancels DOE G 424.1-1A. Admin Chg 1 dated 4-12-13.
Fiscal year 1996 well installation program summary, Y-12 Plant Oak Ridge, Tennessee
This report summarizes the well installation activities conducted during the federal fiscal year (FY) 1996 drilling program at the Oak Ridge Y-12 Plant, Oak Ridge Tennessee. Synopses of monitoring well construction/well development data, well location rationale, geological/hydrological observations, quality assurance/quality control methods, and health and safety monitoring are included. Two groundwater monitoring wells were installed during the FY 1996 drilling program. One of the groundwater monitoring wells was installed in the Lake Reality area and was of polyvinyl chloride screened construction. The other well, installed near the Ash Disposal Basin, was of stainless steel construction.
Evaluation of aged concrete structures for continued service in nuclear power plants
Naus, D.J.; Marchbanks, M.F.; Arndt, E.G.
Results are summarized of a study on concrete component aging and its significance relative to continued service of nuclear power plants (NPPs) beyond the initial period for which they were granted operating licenses. Progress is presented of a second study being conducted to identify and provide acceptance criteria for structural safety issues which the USNRC staff will need to address when applications are submitted for continued service of NPPs. Major activities under this program include: development of a materials property data base, establishment of structural component assessment and repair procedures, and development of a methodology for determination of structural reliability. 19 refs., 5 figs., 3 tabs.
Evaluation of aged concrete structures for continued service in nuclear power plants
Naus, D.J.; Marchbanks, M.F.; Arndt, E.G.
Results are summarized of a study on concrete component aging and its significance relative to continued service of nuclear power plants (NPPs) beyond the initial period for which they were granted operating licenses. Progress is presented of a second study being conducted to identify and provide acceptance criteria for structural safety issues which the USNRC staff will need to address when applications are submitted for continued service of NPPs. Major activities under this program include: development of a materials property data base, establishment of structural component assessment and repair procedures, and development of a methodology for determination of structural reliability.
Mechanisms in Plant Development
Hake, Sarah [USDA ARS Plant Gene Expression Center
This meeting has been held every other year for the past twenty-two years and is the only regularly held meeting focused specifically on plant development. Topics covered included: patterning in developing tissues; short and long distance signaling; differentiation of cell types; the role of epigenetics in development; evolution; growth.
Nuclear power plant fire protection: philosophy and analysis. [PWR; BWR
Berry, D. L.
This report combines a fire severity analysis technique with a fault tree methodology for assessing the importance to nuclear power plant safety of certain combinations of components and systems. Characteristics unique to fire, such as propagation induced by the failure of barriers, have been incorporated into the methodology. By applying the resulting fire analysis technique to actual conditions found in a representative nuclear power plant, it is found that some safety and nonsafety areas are both highly vulnerable to fire spread and impotant to overall safety, while other areas prove to be of marginal importance. Suggestions are made for further experimental and analytical work to supplement the fire analysis method.
Radiological Emergency Response Health and Safety Manual
D. R. Bowman
This manual was created to provide health and safety (H&S) guidance for emergency response operations. The manual is organized in sections that define each aspect of H and S Management for emergency responses. The sections are as follows: Responsibilities; Health Physics; Industrial Hygiene; Safety; Environmental Compliance; Medical; and Record Maintenance. Each section gives guidance on the types of training expected for managers and responders, safety processes and procedures to be followed when performing work, and what is expected of managers and participants. Also included are generic forms that will be used to facilitate or document activities during an emergency response. These ensure consistency in creating useful real-time and archival records and help to prevent the loss or omission of information.
Eagling, D.G. (ed.)
This guide provides managers with practical guidelines for administering a comprehensive earthquake safety program. The Guide is comprehensive with respect to earthquakes in that it covers the most important aspects of natural hazards, site planning, evaluation and rehabilitation of existing buildings, design of new facilities, operational safety, emergency planning, special considerations related to shielding blocks, non-structural elements, lifelines, fire protection and emergency facilities. Management of risk and liabilities is also covered. Nuclear facilities per se are not dealt with specifically. The principles covered also apply generally to nuclear facilities but the design and construction of such structures are subject to special regulations and legal controls.
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Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Scienceand Requirements RecentlyElectronicResourcesjobsJuly throughR EMaterials (CRM)SafetySafety
Hainz, L.C. [Hainz Engineering Services, Inc., Albany, OR (United States)
Electron Beam melting technology, along with other vacuum metallurgical technologies, requires special attention to safety involving operation and maintenance of the EB furnace and systems. Although the EB industry has been relatively accident free, the importance of safety awareness and compliance becomes increasingly important. It is very important to provide a safe work environment for employees and economically important to protect the equipment from damage and potential downtime. Safety and accident prevention directly affects overhead costs by keeping accident insurance rates at a minimum. Routine safety requirements will be reviewed and safety aspects requiring extra attention will be addressed. Safety improvements and experiences of furnace users will be shared as examples.
Digital Full-Scope Simulation of a Conventional Nuclear Power Plant Control Room, Phase 2: Installation of a Reconfigurable Simulator to Support Nuclear Plant Sustainability
Ronald L. Boring; Vivek Agarwal; Kirk Fitzgerald; Jacques Hugo; Bruce Hallbert
The U.S. Department of Energy�s Light Water Reactor Sustainability program has developed a control room simulator in support of control room modernization at nuclear power plants in the U.S. This report highlights the recent completion of this reconfigurable, full-scale, full-scope control room simulator buildout at the Idaho National Laboratory. The simulator is fully reconfigurable, meaning it supports multiple plant models developed by different simulator vendors. The simulator is full-scale, using glasstop virtual panels to display the analog control boards found at current plants. The present installation features 15 glasstop panels, uniquely achieving a complete control room representation. The simulator is also full-scope, meaning it uses the same plant models used for training simulators at actual plants. Unlike in the plant training simulators, the deployment on glasstop panels allows a high degree of customization of the panels, allowing the simulator to be used for research on the design of new digital control systems for control room modernization. This report includes separate sections discussing the glasstop panels, their layout to mimic control rooms at actual plants, technical details on creating a multi-plant and multi-vendor reconfigurable simulator, and current efforts to support control room modernization at U.S. utilities. The glasstop simulator provides an ideal testbed for prototyping and validating new control room concepts. Equally importantly, it is helping create a standardized and vetted human factors engineering process that can be used across the nuclear industry to ensure control room upgrades maintain and even improve current reliability and safety.
Nuclear Safety Information Dashboard | Department of Energy
Office of Environmental Management (EM)
Nuclear Safety Information Dashboard Nuclear Safety Information Dashboard The Nuclear Safety Information (NSI) Dashboard provides a new user interface to the Occurrence Reporting...
Safety Interlocks Group - Advanced Photon Source
Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)
Info APS Safety Page ESH Safety Manual Safety Interlocks Systems ACIS PSS FEEPS BLEPS UPS Division Links APS Organization Chart Beamlines Directory APS Engineering Support...