Science.gov

Sample records for fuel safety criteria

  1. PWR-2 Blanket Fuel Assembly Removal Safety Basis Criteria Document

    SciTech Connect

    BUSHORE, R.P.

    2001-01-22

    This criteria document describes the proposed format, content, and schedule for the preparation of an amendment to the Interim Safety Basis for Solid Waste Facilities (T Plant) (ISB), (HNF-SD-WM-ISB-006), and to the T Plant Interim Operational Safety Requirements (IOSR) (''F-SD-WM-TSR-003). The amendments to these documents are intended to authorize removal of spent nuclear fuel (SNF) assemblies from the spent fuel pool in the Solid Waste Treatment Facility 221-T canyon for interim storage in the Canister Storage Building (CSB). The amendments will include a stand-alone safety assessment as well as revisions to these safety documents as needed to reflect the changes in work scope not currently authorized to accomplish the expected end-state of the Fuel Removal Project for the 221-T Facility.

  2. Laser Safety Inspection Criteria

    SciTech Connect

    Barat, K

    2005-02-11

    A responsibility of the Laser Safety Officer (LSO) is to perform laser safety audits. The American National Standard Z136.1 Safe use of Lasers references this requirement in several sections: (1) Section 1.3.2 LSO Specific Responsibilities states under Hazard Evaluation, ''The LSO shall be responsible for hazards evaluation of laser work areas''; (2) Section 1.3.2.8, Safety Features Audits, ''The LSO shall ensure that the safety features of the laser installation facilities and laser equipment are audited periodically to assure proper operation''; and (3) Appendix D, under Survey and Inspections, it states, ''the LSO will survey by inspection, as considered necessary, all areas where laser equipment is used''. Therefore, for facilities using Class 3B and or Class 4 lasers, audits for laser safety compliance are expected to be conducted. The composition, frequency and rigueur of that inspection/audit rests in the hands of the LSO. A common practice for institutions is to develop laser audit checklists or survey forms. In many institutions, a sole Laser Safety Officer (LSO) or a number of Deputy LSO's perform these audits. For that matter, there are institutions that request users to perform a self-assessment audit. Many items on the common audit list and the associated findings are subjective because they are based on the experience and interest of the LSO or auditor in particular items on the checklist. Beam block usage is an example; to one set of eyes a particular arrangement might be completely adequate, while to another the installation may be inadequate. In order to provide more consistency, the National Ignition Facility Directorate at Lawrence Livermore National Laboratory (NIF-LLNL) has established criteria for a number of items found on the typical laser safety audit form. These criteria are distributed to laser users, and they serve two broad purposes: first, it gives the user an expectation of what will be reviewed by an auditor, and second, it is an

  3. Laser Safety Inspection Criteria

    SciTech Connect

    Barat, K

    2005-06-13

    A responsibility of the Laser Safety Officer (LSO) is to perform laser audits. The American National Standard Z136.1 Safe Use of Lasers references this requirement through several sections. One such reference is Section 1.3.2.8, Safety Features Audits, ''The LSO shall ensure that the safety features of the laser installation facilities and laser equipment are audited periodically to assure proper operation''. The composition, frequency and rigor of that inspection/audit rests in the hands of the LSO. A common practice for institutions is to develop laser audit checklists or survey forms It is common for audit findings from one inspector or inspection to the next to vary even when reviewing the same material. How often has one heard a comment, ''well this area has been inspected several times over the years and no one ever said this or that was a problem before''. A great number of audit items, and therefore findings, are subjective because they are based on the experience and interest of the auditor to particular items on the checklist. Beam block usage, to one set of eyes might be completely adequate, while to another, inadequate. In order to provide consistency, the Laser Safety Office of the National Ignition Facility Directorate has established criteria for a number of items found on the typical laser safety audit form. The criteria are distributed to laser users. It serves two broad purposes; first, it gives the user an expectation of what will be reviewed by an auditor. Second, it is an opportunity to explain audit items to the laser user and thus the reasons for some of these items, such as labelling of beam blocks.

  4. Aerostructural safety factor criteria

    NASA Technical Reports Server (NTRS)

    Verderaime, V.

    1992-01-01

    The present modification of the conventional safety factor method for aircraft structures evaluation involves the expression of deterministic safety factors in probabilistic tolerance limit ratios; these are found to involve a total of three factors that control the interference of applied and resistive stress distributions. The deterministic expression is extended so that it may furnish a 'relative ultimate safety' index that encompasses all three distribution factors. Operational reliability is developed on the basis of the applied and the yield stress distribution interferences. Industry standards are suggested to be derivable from factor selections that are based on the consequences of failure.

  5. FFTF fuel systems design criteria

    SciTech Connect

    Dutt, D.S.; Baars, R.E.; Jackson, R.J.; Weber, J.W.

    1980-01-01

    The purpose of this paper is to first enumerate the design considerations that were given to the fuel system, then secondly, show how these design allowances, methods, and criteria compare to the subsequent irradiation data. This comparison will show that decisions made by the design team were generally correct and, if in error, tended to be conservative. The FFTF driver fuel assemblies addressed by this paper are composed of the duct, a spacer system, and 217 fuel pins. Each of these subcomponents is described as the criteria are discussed and important parameters noted.

  6. Ferrocyanide Safety Program: Safety criteria for ferrocyanide watch list tanks

    SciTech Connect

    Postma, A.K.; Meacham, J.E.; Barney, G.S.

    1994-01-01

    This report provides a technical basis for closing the ferrocyanide Unreviewed Safety Question (USQ) at the Hanford Site. Three work efforts were performed in developing this technical basis. The efforts described herein are: 1. The formulation of criteria for ranking the relative safety of waste in each ferrocyanide tank. 2. The current classification of tanks into safety categories by comparing available information on tank contents with the safety criteria; 3. The identification of additional information required to resolve the ferrocyanide safety issue.

  7. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  8. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  9. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  10. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  11. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  12. Gaseous-fuel safety assessment. Status report

    SciTech Connect

    Krupka, M.C.; Edeskuty, F.J.; Bartlit, J.R.; Williamson, K.D. Jr.

    1982-01-01

    The Los Alamos National Laboratory, in support of studies sponsored by the Office of Vehicle and Engine Research and Development in the US Department of Energy, has undertaken a safety assessment of selected gaseous fuels for use in light automotive transportation. The purpose is to put into perspective the hazards of these fuels relative to present day fuels and delineated criteria for their safe handling. Fuels include compressed and liquified natural gas (CNG and LNG), liquefied petroleum gas (LPG), and for reference gasoline and diesel. This paper is a program status report. To date, physicochemical property data and general petroleum and transportation information were compiled; basic hazards defined; alternative fuels were safety-ranked based on technical properties alone; safety data and vehicle accident statistics reviewed; and accident scenarios selected for further analysis. Methodology for such analysis is presently under consideration.

  13. Uncertainties in the effects of burnup and their impact on criticality safety licensing criteria

    SciTech Connect

    Carlson, R.W.; Fisher, L.E.

    1990-07-13

    Current criteria for criticality safety for spent fuel shipping and storage casks are conservative because no credit is permitted for the effects of burnup of the fuel inside the cask. Cask designs that will transport and store large numbers of fuel assemblies (20 or more) must devote a substantial part of their payload to criticality control measures if they are to meet this criteria. The Department of Energy is developing the data necessary to support safety analyses that incorporate the effects of burnup for the next generation of spent fuel shipping casks. The efforts described here are devoted to the development of acceptance criteria that will be the basis for accepting safety analyses. Preliminary estimates of the uncertainties of the effects of burnup have been developed to provide a basis for the consideration of critically safety criteria. The criticality safety margins in a spent fuel shipping or storage cask are dominated by the portions of a fuel assembly that are in low power regions of a reactor core, and the reactor operating conditions are very different from spent fuel storage or transport cask conditions. Consequently, the experience that has been gathered during years of reactor operation does not apply directly to the prediction of criticality safety margins for spent fuel shipping or storage casks. The preliminary estimates of the uncertainties presented in this paper must be refined by both analytical and empirical studies that address both the magnitude of the uncertainties and their interdependence. 9 refs., 5 figs.

  14. Fuel Safety Activities in Korea

    SciTech Connect

    Auh, Geun-Sun; Shin, A.D.; Lee, J.S.; Woo, S.W.; Ryu, Y.H.; Kim, Jun-Hwan; Kim, S.K.; Jeong, Y.H.

    2007-07-01

    The current regulatory requirements for fuel performance were based on earlier test data of fresh or low burnup Zircaloy fuels of less than 40 GWD/MTU. Most countries have not changed the current regulatory requirements even if they are actively investigating the high burnup and new cladding alloy effects. Korea agrees with commonly accepted international consensus that although there are technical issues requiring resolutions, these issues do not constitute immediate safety concerns. The high burnup fuel reactor performance experiences of Korea do not show any major problems even if there have been some burnup related fuel failures which are described in the paper. KINS has recommended the industry to have lower fuel failure rates than 1-2 per 50,000 fuel rods. A research project of High Burnup Fuel Safety Tests and Evaluations has started in 2002 under a joint cooperation of KAERI/KNFC/KEPRI and KINS to obtain performance results of high burnup fuel and to develop evaluation technologies of high burnup fuel safety issues. From 1998, KINS has closely monitored and actively participated in international activities such as OECD/NEA CABRI Water Loop Program to reflect on regulatory requirements if needed. KINS will closely monitor the high burnup fuel performances of Korea to strength the regulatory activities if needed. The research activities in Korea including of LOCA and RIA being performed at KAERI with active supports of the industry are summarized in the paper. (authors)

  15. Nuclear safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The purpose of this document is to define safety criteria which must be met to implement US safety policy for space fission reactors. These criteria provide the bases for decisions on the acceptability of specific mission and reactor design proposals. (JDH)

  16. 10 CFR 32.23 - Same: Safety criteria.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Same: Safety criteria. 32.23 Section 32.23 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.23 Same: Safety criteria. An applicant for a...

  17. 10 CFR 32.27 - Same: Safety criteria.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Same: Safety criteria. 32.27 Section 32.27 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.27 Same: Safety criteria. An applicant for a...

  18. 10 CFR 32.27 - Same: Safety criteria.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Same: Safety criteria. 32.27 Section 32.27 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.27 Same: Safety criteria. An applicant for a...

  19. 10 CFR 32.27 - Same: Safety criteria.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Same: Safety criteria. 32.27 Section 32.27 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.27 Same: Safety criteria. An applicant for a...

  20. 10 CFR 32.23 - Same: Safety criteria.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Same: Safety criteria. 32.23 Section 32.23 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.23 Same: Safety criteria. An applicant for a...

  1. 10 CFR 32.23 - Same: Safety criteria.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Same: Safety criteria. 32.23 Section 32.23 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.23 Same: Safety criteria. An applicant for a...

  2. 10 CFR 32.27 - Same: Safety criteria.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Same: Safety criteria. 32.27 Section 32.27 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.27 Same: Safety criteria. An applicant for a...

  3. 10 CFR 32.23 - Same: Safety criteria.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Same: Safety criteria. 32.23 Section 32.23 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.23 Same: Safety criteria. An applicant for a...

  4. 10 CFR 32.27 - Same: Safety criteria.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Same: Safety criteria. 32.27 Section 32.27 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.27 Same: Safety criteria. An applicant for a...

  5. 10 CFR 32.23 - Same: Safety criteria.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Same: Safety criteria. 32.23 Section 32.23 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.23 Same: Safety criteria. An applicant for a...

  6. Safety goals and functional performance criteria. [Advanced Reactor Design

    SciTech Connect

    Youngblood, R.W.; Pratt, W.T. ); Hardin, W.B. . Div. of Regulatory Applications)

    1990-01-01

    This report discusses a possible approach to the development of functional performance criteria to be applied to evolutionary LWR designs. Key safety functions are first identified; then, criteria are drawn up for each individual function, based on the premise that no single function's projected unreliability should be allowed to exhaust the safety goal frequencies. In the area of core damage prevention, functional criteria are cast in terms of necessary levels of redundancy and diversity of critical equipment. In the area of core damage mitigation (containment), functional performance criteria are cast with the aim of mitigating post-core-melt phenomena with sufficient assurance to eliminate major uncertainties in containment performance. 9 refs.

  7. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  8. Nuclear-safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The policy of the United States for all US nuclear power sources in space is to ensure that the probability of release of radioactive material and the amounts released are such that an undue risk is not presented, considering the benefits of the mission. The objective of this document is to provide safety criteria which a mission/reactor designer can use to help ensure that the design is acceptable from a radiological safety standpoint. These criteria encompass mission design, reactor design, and radiological impact limitation requirements for safety, and the documentation required. They do not address terrestrial operations, occupational safety or system reliability except where the systems are important for radiological safety. Specific safety specifications based on these criteria shall also be generated and made part of contractual requirements.

  9. 76 FR 20070 - Commercial Space Transportation Safety Approval Performance Criteria

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-11

    ... Federal Aviation Administration Commercial Space Transportation Safety Approval Performance Criteria... received, a safety approval for the ability of its Space Training System: Model 400 (STS-400) to replicate....19 (a)(4). NASTAR's ] STS-400 suborbital space flight simulator (a multi-axis centrifuge) is...

  10. Discussions about safety criteria and guidelines for radioactive waste management.

    PubMed

    Yamamoto, Masafumi

    2011-07-01

    In Japan, the clearance levels for uranium-bearing waste have been established by the Nuclear Safety Commission (NSC). The criteria for uranium-bearing waste disposal are also necessary; however, the NSC has not concluded the discussion on this subject. Meanwhile, the General Administrative Group of the Radiation Council has concluded the revision of its former recommendation 'Regulatory exemption dose for radioactive solid waste disposal', the dose criteria after the institutional control period for a repository. The Standardization Committee on Radiation Protection in the Japan Health Physics Society (The Committee) also has developed the relevant safety criteria and guidelines for existing exposure situations, which are potentially applicable to uranium-bearing waste disposal. A new working group established by The Committee was initially aimed at developing criteria and guidelines specifically for uranium-bearing waste disposal; however, the aim has been shifted to broader criteria applicable to any radioactive wastes. PMID:21531746

  11. Explosive safety criteria at a Department of Energy contractor facility

    NASA Astrophysics Data System (ADS)

    Krach, F.

    1984-08-01

    Monsanto Research Corporation (MRC) operates the Mound facility in Miamisburg, Ohio, for the Department of Energy. Small explosive components are manufactured at MRC, and stringent explosive safety criteria have been developed for their manufacturing. The goals of these standards are to reduce employee injuries and eliminate fenceline impacts resulting from accidental detonations. The manner in which these criteria were developed and what DOD standards were incorporated into MRC's own design criteria are described. These design requirements are applicable to all new construction at MRC. An example of the development of the design of a Component Test Facility is presented to illustrate the application of the criteria.

  12. Sipping fuel and saving lives: increasing fuel economy withoutsacrificing safety

    SciTech Connect

    Gordon, Deborah; Greene, David L.; Ross, Marc H.; Wenzel, Tom P.

    2007-06-11

    The public, automakers, and policymakers have long worried about trade-offs between increased fuel economy in motor vehicles and reduced safety. The conclusion of a broad group of experts on safety and fuel economy in the auto sector is that no trade-off is required. There are a wide variety of technologies and approaches available to advance vehicle fuel economy that have no effect on vehicle safety. Conversely, there are many technologies and approaches available to advance vehicle safety that are not detrimental to vehicle fuel economy. Congress is considering new policies to increase the fuel economy of new automobiles in order to reduce oil dependence and reduce greenhouse gas emissions. The findings reported here offer reassurance on an important dimension of that work: It is possible to significantly increase the fuel economy of motor vehicles without compromising their safety. Automobiles on the road today demonstrate that higher fuel economy and greater safety can co-exist. Some of the safest vehicles have higher fuel economy, while some of the least safe vehicles driven today--heavy, large trucks and SUVs--have the lowest fuel economy. At an October 3, 2006 workshop, leading researchers from national laboratories, academia, auto manufacturers, insurance research industry, consumer and environmental groups, material supply industries, and the federal government agreed that vehicles could be designed to simultaneously improve safety and fuel economy. The real question is not whether we can realize this goal, but the best path to get there. The experts' studies reveal important new conclusions about fuel economy and safety, including: (1) Vehicle fuel economy can be increased without affecting safety, and vice versa; (2) Reducing the weight and height of the heaviest SUVs and pickup trucks will simultaneously increase both their fuel economy and overall safety; and (3) Advanced materials can decouple size from mass, creating important new possibilities for

  13. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  14. Safety evaluation of a hydrogen fueled transit bus

    SciTech Connect

    Coutts, D.A.; Thomas, J.K.; Hovis, G.L.; Wu, T.T.

    1997-12-31

    Hydrogen fueled vehicle demonstration projects must satisfy management and regulator safety expectations. This is often accomplished using hazard and safety analyses. Such an analysis has been completed to evaluate the safety of the H2Fuel bus to be operated in Augusta, Georgia. The evaluation methods and criteria used reflect the Department of Energy`s graded approach for qualifying and documenting nuclear and chemical facility safety. The work focused on the storage and distribution of hydrogen as the bus motor fuel with emphases on the technical and operational aspects of using metal hydride beds to store hydrogen. The safety evaluation demonstrated that the operation of the H2Fuel bus represents a moderate risk. This is the same risk level determined for operation of conventionally powered transit buses in the United States. By the same criteria, private passenger automobile travel in the United States is considered a high risk. The evaluation also identified several design and operational modifications that resulted in improved safety, operability, and reliability. The hazard assessment methodology used in this project has widespread applicability to other innovative operations and systems, and the techniques can serve as a template for other similar projects.

  15. Hydrogen and Gaseous Fuel Safety and Toxicity

    SciTech Connect

    Lee C. Cadwallader; J. Sephen Herring

    2007-06-01

    Non-traditional motor fuels are receiving increased attention and use. This paper examines the safety of three alternative gaseous fuels plus gasoline and the advantages and disadvantages of each. The gaseous fuels are hydrogen, methane (natural gas), and propane. Qualitatively, the overall risks of the four fuels should be close. Gasoline is the most toxic. For small leaks, hydrogen has the highest ignition probability and the gaseous fuels have the highest risk of a burning jet or cloud.

  16. Risk-based versus deterministic explosives safety criteria

    SciTech Connect

    Wright, R.E.

    1996-12-01

    The Department of Defense Explosives Safety Board (DDESB) is actively considering ways to apply risk-based approaches in its decision- making processes. As such, an understanding of the impact of converting to risk-based criteria is required. The objectives of this project are to examine the benefits and drawbacks of risk-based criteria and to define the impact of converting from deterministic to risk-based criteria. Conclusions will be couched in terms that allow meaningful comparisons of deterministic and risk-based approaches. To this end, direct comparisons of the consequences and impacts of both deterministic and risk-based criteria at selected military installations are made. Deterministic criteria used in this report are those in DoD 6055.9-STD, `DoD Ammunition and Explosives Safety Standard.` Risk-based criteria selected for comparison are those used by the government of Switzerland, `Technical Requirements for the Storage of Ammunition (TLM 75).` The risk-based criteria used in Switzerland were selected because they have been successfully applied for over twenty-five years.

  17. Fuel Fracture (Crumbling) Safety Impact (OCRWM)

    SciTech Connect

    DUNCAN, D.R.

    1999-12-08

    The safety impact of experimentally observed N Reactor fuel sample fracture and fragmentation is evaluated using an average reaction rate enhancement derived from data from thermo-gravimetric analysis (TGA) experiments on fuel samples. The enhanced reaction rates attributed to fragmentation were within the existing safety basis.

  18. Fuel Fracture (Crumbling) Safety Impacts (OCRWM)

    SciTech Connect

    DUNCAN, D.R.

    2000-01-24

    The safety impact of experimentally observed N Reactor fuel sample fracture and fragmentation is evaluated using an average reaction rate enhancement derived from data from thermo-gravimetric analysis (TGA) experiments on fuel samples. The enhanced reaction rates attributed to fragmentation were within the existing safety basis. Peer review comments for the Revision 0 version were incorporated.

  19. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    SciTech Connect

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  20. Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

    SciTech Connect

    Rechard, R.P.; Lord, M.E.; Stockman, C.T.; McCurley, R.D.

    1997-12-31

    The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE`s Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection.

  1. Safety-related operator actions: methodology for developing criteria

    SciTech Connect

    Kozinsky, E.J.; Gray, L.H.; Beare, A.N.; Barks, D.B.; Gomer, F.E.

    1984-03-01

    This report presents a methodology for developing criteria for design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base. It is the eleventh and final NUREG/CR Report on the Safety-Related Operator Actions Program, conducted by Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The operator performance data were developed from training simulator experiments involving operator responses to simulated scenarios of plant disturbances; from field data on events with similar scenarios; and from task analytic data. A conceptual model to integrate the data was developed and a computer simulation of the model was run, using the SAINT modeling language. Proposed is a quantitative predictive model of operator performance, the Operator Personnel Performance Simulation (OPPS) Model, driven by task requirements, information presentation, and system dynamics. The model output, a probability distribution of predicted time to correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria.

  2. Methodology for determining criteria for storing spent fuel in air

    SciTech Connect

    Reid, C.R.; Gilbert, E.R.

    1986-11-01

    Dry storage in an air atmosphere is a method being considered for spent light water reactor (LWR) fuel as an alternative to storage in an inert gas environment. However, methods to predict fuel integrity based on oxidation behavior of the fuel first must be evaluated. The linear cumulative damage method has been proposed as a technique for defining storage criteria. Analysis of limited nonconstant temperature data on nonirradiated fuel samples indicates that this approach yields conservative results for a strictly decreasing-temperature history. On the other hand, the description of damage accumulation in terms of remaining life concepts provides a more general framework for making predictions of failure. Accordingly, a methodology for adapting remaining life concepts to UO/sub 2/ oxidation has been developed at Pacific Northwest Laboratory. Both the linear cumulative damage and the remaining life methods were used to predict oxidation results for spent fuel in which the temperature was decreased with time to simulate the temperature history in a dry storage cask. The numerical input to the methods was based on oxidation data generated with nonirradiated UO/sub 2/ pellets. The calculated maximum allowable storage temperatures are strongly dependent on the temperature-time profile and emphasize the conservatism inherent in the linear cumulative damage model. Additional nonconstant temperature data for spent fuel are needed to both validate the proposed methods and to predict temperatures applicable to actual spent fuel storage.

  3. Safety analysis, risk assessment, and risk acceptance criteria

    SciTech Connect

    Jamali, K.; Stack, D.W.; Sullivan, L.H.; Sanzo, D.L.

    1997-08-01

    This paper discusses a number of topics that relate safety analysis as documented in the Department of Energy (DOE) safety analysis reports (SARs), probabilistic risk assessments (PRA) as characterized primarily in the context of the techniques that have assumed some level of formality in commercial nuclear power plant applications, and risk acceptance criteria as an outgrowth of PRA applications. DOE SARs of interest are those that are prepared for DOE facilities under DOE Order 5480.23 and the implementing guidance in DOE STD-3009-94. It must be noted that the primary area of application for DOE STD-3009 is existing DOE facilities and that certain modifications of the STD-3009 approach are necessary in SARs for new facilities. Moreover, it is the hazard analysis (HA) and accident analysis (AA) portions of these SARs that are relevant to the present discussions. Although PRAs can be qualitative in nature, PRA as used in this paper refers more generally to all quantitative risk assessments and their underlying methods. HA as used in this paper refers more generally to all qualitative risk assessments and their underlying methods that have been in use in hazardous facilities other than nuclear power plants. This discussion includes both quantitative and qualitative risk assessment methods. PRA has been used, improved, developed, and refined since the Reactor Safety Study (WASH-1400) was published in 1975 by the Nuclear Regulatory Commission (NRC). Much debate has ensued since WASH-1400 on exactly what the role of PRA should be in plant design, reactor licensing, `ensuring` plant and process safety, and a large number of other decisions that must be made for potentially hazardous activities. Of particular interest in this area is whether the risks quantified using PRA should be compared with numerical risk acceptance criteria (RACs) to determine whether a facility is `safe.` Use of RACs requires quantitative estimates of consequence frequency and magnitude.

  4. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  5. Safety features of subcritical fluid fueled systems

    SciTech Connect

    Bell, Charles R.

    1995-09-15

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

  6. Safety features of subcritical fluid fueled systems

    SciTech Connect

    Bell, C.R.

    1994-09-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

  7. Occupational safety and health criteria for responsible development of nanotechnology.

    PubMed

    Schulte, P A; Geraci, C L; Murashov, V; Kuempel, E D; Zumwalde, R D; Castranova, V; Hoover, M D; Hodson, L; Martinez, K F

    2014-01-01

    Organizations around the world have called for the responsible development of nanotechnology. The goals of this approach are to emphasize the importance of considering and controlling the potential adverse impacts of nanotechnology in order to develop its capabilities and benefits. A primary area of concern is the potential adverse impact on workers, since they are the first people in society who are exposed to the potential hazards of nanotechnology. Occupational safety and health criteria for defining what constitutes responsible development of nanotechnology are needed. This article presents five criterion actions that should be practiced by decision-makers at the business and societal levels-if nanotechnology is to be developed responsibly. These include (1) anticipate, identify, and track potentially hazardous nanomaterials in the workplace; (2) assess workers' exposures to nanomaterials; (3) assess and communicate hazards and risks to workers; (4) manage occupational safety and health risks; and (5) foster the safe development of nanotechnology and realization of its societal and commercial benefits. All these criteria are necessary for responsible development to occur. Since it is early in the commercialization of nanotechnology, there are still many unknowns and concerns about nanomaterials. Therefore, it is prudent to treat them as potentially hazardous until sufficient toxicology, and exposure data are gathered for nanomaterial-specific hazard and risk assessments. In this emergent period, it is necessary to be clear about the extent of uncertainty and the need for prudent actions. PMID:24482607

  8. Health, safety and environmental criteria for siting of laboratory facilities.

    PubMed

    Lees, P S; Corn, M

    1983-04-01

    The development of applicable criteria for assessing the suitability of a site for construction of full and partial containment laboratories for the analysis of unknown and highly toxic chemicals is described. The criteria, based on considerations of health, safety and environmental factors, are used to define critical considerations in site selection to minimize the risk to non-laboratory personnel and the surrounding environment. Criteria are synthesized from several sources using the assumption of a worst-case chemical release. Mechanical failures, human failures, critical events and social/legal limitations are investigated, as are the characteristics of a site which may limit construction of such a facility. A detailed description is made of the various types of laboratories and the types of samples analyzed in them. The final recommendations are summarized for five typical laboratory settings; they are based primarily on the potential impacts on people, property and natural resources. A single occupancy building in a rural setting is recommended as the most suitable site for a full containment laboratory. A single occupancy building in an industrial park setting is acceptable, while multiple occupancy buildings and sites which are more highly developed are unacceptable. Similar recommendations are made for partial containment and conventional laboratories. PMID:6858856

  9. Criteria for Modeling in LES of Multicomponent Fuel Flow

    NASA Technical Reports Server (NTRS)

    Bellan, Josette; Selle, Laurent

    2009-01-01

    A report presents a study addressing the question of which large-eddy simulation (LES) equations are appropriate for modeling the flow of evaporating drops of a multicomponent liquid in a gas (e.g., a spray of kerosene or diesel fuel in air). The LES equations are obtained from the direct numerical simulation (DNS) equations in which the solution is computed at all flow length scales, by applying a spatial low-pass filter. Thus, in LES the small scales are removed and replaced by terms that cannot be computed from the LES solution and instead must be modeled to retain the effect of the small scales into the equations. The mathematical form of these models is a subject of contemporary research. For a single-component liquid, there is only one LES formulation, but this study revealed that for a multicomponent liquid, there are two non-equivalent LES formulations for the conservation equations describing the composition of the vapor. Criteria were proposed for selecting the multicomponent LES formulation that gives the best accuracy and increased computational efficiency. These criteria were applied in examination of filtered DNS databases to compute the terms in the LES equations. The DNS databases are from mixing layers of diesel and kerosene fuels. The comparisons resulted in the selection of one of the multicomponent LES formulations as the most promising with respect to all criteria.

  10. 21 CFR 70.42 - Criteria for evaluating the safety of color additives.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Criteria for evaluating the safety of color additives. 70.42 Section 70.42 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL COLOR ADDITIVES Safety Evaluation § 70.42 Criteria for evaluating the safety of...

  11. Criteria for solid recovered fuels as a substitute for fossil fuels--a review.

    PubMed

    Beckmann, Michael; Pohl, Martin; Bernhardt, Daniel; Gebauer, Kathrin

    2012-04-01

    The waste treatment, particularly the thermal treatment of waste has changed fundamentally in the last 20 years, i.e. from facilities solely dedicated to the thermal treatment of waste to facilities, which in addition to that ensure the safe plant operation and fulfill very ambitious criteria regarding emission reduction, resource recovery and energy efficiency as well. Therefore this contributes to the economic use of raw materials and due to the energy recovered from waste also to the energy provision. The development described had the consequence that waste and solid recovered fuels (SRF) has to be evaluated based on fuel criteria as well. Fossil fuels - coal, crude oil, natural gas etc. have been extensively investigated due to their application in plants for energy conversion and also due to their use in the primary industry. Thereby depending on the respective processes, criteria on fuel technical properties can be derived. The methods for engineering analysis of regular fuels (fossil fuels) can be transferred only partially to SRF. For this reason methods are being developed or adapted to current analytical methods for the characterization of SRF. In this paper the possibilities of the energetic utilization of SRF and the characterization of SRF before and during the energetic utilization will be discussed. PMID:22467662

  12. 40 CFR 600.115-11 - Criteria for determining the fuel economy label calculation method.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... alternative-fuel vehicles, dual fuel vehicles when operating on the alternative fuel, plug-in hybrid electric... 40 Protection of Environment 30 2014-07-01 2014-07-01 false Criteria for determining the fuel... PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND GREENHOUSE GAS EXHAUST EMISSIONS OF...

  13. 40 CFR 600.115-11 - Criteria for determining the fuel economy label calculation method.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... alternative-fuel vehicles, dual fuel vehicles when operating on the alternative fuel, plug-in hybrid electric... 40 Protection of Environment 31 2012-07-01 2012-07-01 false Criteria for determining the fuel... PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND GREENHOUSE GAS EXHAUST EMISSIONS OF...

  14. 40 CFR 600.115-11 - Criteria for determining the fuel economy label calculation method.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... alternative-fuel vehicles, dual fuel vehicles when operating on the alternative fuel, plug-in hybrid electric... 40 Protection of Environment 31 2013-07-01 2013-07-01 false Criteria for determining the fuel... PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND GREENHOUSE GAS EXHAUST EMISSIONS OF...

  15. 49 CFR Appendix A to Part 385 - Explanation of Safety Audit Evaluation Criteria

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 5 2011-10-01 2011-10-01 false Explanation of Safety Audit Evaluation Criteria A Appendix A to Part 385 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS SAFETY FITNESS PROCEDURES Pt. 385,...

  16. 10 CFR 32.31 - Certain industrial devices containing byproduct material: Safety criteria.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Certain industrial devices containing byproduct material: Safety criteria. 32.31 Section 32.31 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO... Certain industrial devices containing byproduct material: Safety criteria. (a) An applicant for a...

  17. 10 CFR 32.31 - Certain industrial devices containing byproduct material: Safety criteria.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Certain industrial devices containing byproduct material: Safety criteria. 32.31 Section 32.31 Energy NUCLEAR REGULATORY COMMISSION SPECIFIC DOMESTIC LICENSES TO... Certain industrial devices containing byproduct material: Safety criteria. (a) An applicant for a...

  18. Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

    SciTech Connect

    Botsch, W.; Smalian, S.; Hinterding, P.

    2013-07-01

    casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

  19. Photovoltaic system criteria documents. Volume 5: Safety criteria for photovoltaic applications

    NASA Technical Reports Server (NTRS)

    Koenig, John C.; Billitti, Joseph W.; Tallon, John M.

    1979-01-01

    Methodology is described for determining potential safety hazards involved in the construction and operation of photovoltaic power systems and provides guidelines for the implementation of safety considerations in the specification, design and operation of photovoltaic systems. Safety verification procedures for use in solar photovoltaic systems are established.

  20. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-02

    ... criticality safety review procedures and acceptance criteria contained in NUREG-1536, Revision 1, ``Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,'' NUREG-1567, ``Standard Review Plan for Spent Fuel Dry Storage Facilities,'' and NUREG-1617, ``Standard Review Plan...

  1. Long range program plan for safety and fuel economy

    SciTech Connect

    Feirice, B.

    1983-11-01

    An analysis was made to determine which potential highway and motor vehicle safety activies were most deserving of further pursuit so as to obtain the greatest improvement in safety at the least cost. Criteria were established to guide in the rating and ranking of potential safety projects. The most significant national safety problems were identified and a description and schedule of the chosen safety projects to address those problems were prepared.

  2. 32 CFR 636.33 - Vehicle safety inspection criteria.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... above the vehicle manufacturer's designated height (49 CFR 570.8). (b) The criteria listed in paragraph... starburst or spider webbing effect greater than 3 inches by 3 inches. No opaque or solid material...

  3. 32 CFR 636.33 - Vehicle safety inspection criteria.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... above the vehicle manufacturer's designated height (49 CFR 570.8). (b) The criteria listed in paragraph... starburst or spider webbing effect greater than 3 inches by 3 inches. No opaque or solid material...

  4. 32 CFR 636.33 - Vehicle safety inspection criteria.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... above the vehicle manufacturer's designated height (49 CFR 570.8). (b) The criteria listed in paragraph... starburst or spider webbing effect greater than 3 inches by 3 inches. No opaque or solid material...

  5. 32 CFR 636.33 - Vehicle safety inspection criteria.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... above the vehicle manufacturer's designated height (49 CFR 570.8). (b) The criteria listed in paragraph... starburst or spider webbing effect greater than 3 inches by 3 inches. No opaque or solid material...

  6. 32 CFR 636.33 - Vehicle safety inspection criteria.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... above the vehicle manufacturer's designated height (49 CFR 570.8). (b) The criteria listed in paragraph... starburst or spider webbing effect greater than 3 inches by 3 inches. No opaque or solid material...

  7. Developing Criteria and Judgment of Safety for Crossing Streets with Gaps in Traffic.

    ERIC Educational Resources Information Center

    Sauerburger, Dona

    1999-01-01

    Discusses using the Timing Method for Assessing the Detection of Vehicles (TMAD) to help individuals with visual impairments develop the ability to judge their safety for crossing streets with no traffic control. Functional criteria for assessing risks are discussed. (CR)

  8. Structural deterministic safety factors selection criteria and verification

    NASA Technical Reports Server (NTRS)

    Verderaime, V.

    1992-01-01

    Though current deterministic safety factors are arbitrarily and unaccountably specified, its ratio is rooted in resistive and applied stress probability distributions. This study approached the deterministic method from a probabilistic concept leading to a more systematic and coherent philosophy and criterion for designing more uniform and reliable high-performance structures. The deterministic method was noted to consist of three safety factors: a standard deviation multiplier of the applied stress distribution; a K-factor for the A- or B-basis material ultimate stress; and the conventional safety factor to ensure that the applied stress does not operate in the inelastic zone of metallic materials. The conventional safety factor is specifically defined as the ratio of ultimate-to-yield stresses. A deterministic safety index of the combined safety factors was derived from which the corresponding reliability proved the deterministic method is not reliability sensitive. The bases for selecting safety factors are presented and verification requirements are discussed. The suggested deterministic approach is applicable to all NASA, DOD, and commercial high-performance structures under static stresses.

  9. Neutronics and safety characteristics of a 100% MOX fueled PWR using weapons grade plutonium

    SciTech Connect

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

    Preliminary neutronics and safety studies, pertaining to the feasibility of using 100% weapons grade mixed-oxide (MOX) fuel in an advanced PWR Westinghouse design are presented in this paper. The preliminary results include information on boron concentration, power distribution, reactivity coefficients and xenon and control rode worth for the initial and the equilibrium cycle. Important safety issues related to rod ejection and steam line break accidents and shutdown margin requirements are also discussed. No significant change from the commercial design is needed to denature weapons-grade plutonium under the current safety and licensing criteria.

  10. FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect

    C.E. Sanders

    2005-06-30

    The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the ''Q-list'' (BSC 2003, p. A-6

  11. SNF fuel retrieval sub project safety analysis document

    SciTech Connect

    BERGMANN, D.W.

    1999-02-24

    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.

  12. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  13. Updating of Safety Criteria for Basic Diagnostic Indicators of Dam at the Sayano-Shushenskaya HPP

    SciTech Connect

    Gordon, L. A.; Skvortsova, A. E.

    2013-09-15

    Values of diagnostic indicators [K]-limitations placed on radial displacements and turn angles of horizontal sections of the dam - which are permitted for each upper-pool level within the range from 520 to 539 m are determined and proposed for inclusion in the Declaration of Safety. Empirical relationships used to develop safety criteria K1 and K2 are modified.

  14. Reassessment of the basis for NRC fuel damage criteria for reactivity transients

    SciTech Connect

    McCardell, R.K.

    1994-10-01

    The present basis for NRC Fuel Damage Criteria was obtained from experiments performed in the Special Power Excursion Reactor Test (SPERT) IV Reactor Capsule Driver Core (CDC) at the Idaho National Engineering Laboratory (INEL) between 1967 and 1970. Most of the CDC test fuel rods were previously unirradiated and the failure threshold for these unirradiated fuel rods was measured to be about 200 calories per gram of UO{sub 2} radially averaged fuel enthalpy at the axial peak.

  15. Fuel Systems Architecture (FSA) evaluation criteria and concept evaluation methodology

    NASA Technical Reports Server (NTRS)

    Hendershot, J. E.; Corban, R. R.; Stevenson, S. M.

    1991-01-01

    Consideration is given to two methods developed for the evaluation, screening, and ranking of concepts for Space Exploration Initiative vehicle propellant management systems. The methods selected for handling this multicriteria decision problem are based on the utility theory which transforms both qualitative and quantitative criteria into a nondimensional utility scale for comparison of dissimilar figures of merit. The development of the resultant FSA evaluation criteria and concept evaluation methodology is summarized.

  16. Nuclear energy with inherent safety: Change of outdated paradigm, criteria

    NASA Astrophysics Data System (ADS)

    Adamov, E. O.; Orlov, V. V.; Rachkov, V. I.; Slessarev, I. S.; Khomyakov, Yu. S.

    2015-12-01

    Modern nuclear power technology still has significant sources of risk, and, weak links, such as, a threat of severe accidents with catastrophic unpredictable consequences and damage to the population, proliferation of nuclear weapon-usable materials, risks of long-term storage of toxic radioactive waste, risks of loss of major investments in nuclear facilities and their construction, lack of fuel resources for the ambitious role of nuclear power in the competitive balance of energy. Each of these risks is important and almost independent, though the elimination of some of them does not significantly alter the overall assessment of nuclear power.

  17. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    SciTech Connect

    BENECKE, M.W.

    2000-09-06

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility.

  18. Safety engineering in handling fuels and lubricants in civil aviation

    NASA Astrophysics Data System (ADS)

    Protoereiskii, Aleksandr Stepanovich

    The book is concerned with methods of improving working conditions, work hygiene, safety engineering, and fire and explosion prevention during the storage and handling of petroleum products at fuel and lubricant storage facilities. The discussion covers methods of protection against static and atmospheric discharges, lightning protection, safety engineering in fuel and lubricant laboratories, and methods of fire prevention and fire extinction. Attention is also given to methods for administering first aid in case of accidents and poisoning.

  19. Safety criteria for flying E-sail through solar eclipse

    NASA Astrophysics Data System (ADS)

    Janhunen, Pekka; Toivanen, Petri

    2015-09-01

    The electric solar wind sail (E-sail) propellantless propulsion device uses long, charged metallic tethers to tap momentum from the solar wind to produce spacecraft propulsion. If flying through planetary or moon eclipse, the long E-sail tethers can undergo significant thermal contraction and expansion. Rapid shortening of the tether increases its tension due to inertia of the tether and a Remote Unit that is located on the tether tip (a Remote Unit is part of typical E-sail designs). We analyse by numerical simulation the conditions under which eclipse induced stresses are safe for E-sail tethers. We calculate the closest safe approach distances for Earth, Moon, Venus, Mars, Jupiter, Ceres and an exemplary 300 km main belt asteroid Interamnia for circular, parabolic and hyperbolic orbits. We find that any kind of eclipsing is safe beyond approximately 2.5 au distance, but for terrestrial planets safety depends on the parameters of the orbit. For example, for Mars the safe distance with 20 km E-sail tether lies between Phobos and Deimos orbits.

  20. Fuel supply shutdown facility interim operational safety requirements

    SciTech Connect

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-05-23

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance).

  1. Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    SciTech Connect

    Marshall, William BJ J; Wagner, John C

    2012-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

  2. Fuel Storage Facility Final Safety Analysis Report. Revision 1

    SciTech Connect

    Linderoth, C.E.

    1984-03-01

    The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.

  3. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  4. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  5. 49 CFR Appendix C to Part 236 - Safety Assurance Criteria and Processes

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 4 2013-10-01 2013-10-01 false Safety Assurance Criteria and Processes C Appendix C to Part 236 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RULES, STANDARDS, AND INSTRUCTIONS GOVERNING THE INSTALLATION, INSPECTION, MAINTENANCE, AND REPAIR...

  6. 49 CFR Appendix C to Part 236 - Safety Assurance Criteria and Processes

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 4 2014-10-01 2014-10-01 false Safety Assurance Criteria and Processes C Appendix C to Part 236 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RULES, STANDARDS, AND INSTRUCTIONS GOVERNING THE INSTALLATION, INSPECTION, MAINTENANCE, AND REPAIR...

  7. 14 CFR 414.19 - Technical criteria for reviewing a safety approval application.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Technical criteria for reviewing a safety approval application. 414.19 Section 414.19 Aeronautics and Space COMMERCIAL SPACE TRANSPORTATION, FEDERAL... customized by the manufacturer that intends to produce the system, system component, or part. The...

  8. Safety aspects of the IFR pyroprocess fuel cycle

    SciTech Connect

    Forrester, R.J.; Lineberry, M.J.; Charak, I.; Tessier, J.H.; Solbrig, C.W.; Gabor, J.D.

    1989-01-01

    This paper addresses the important safety considerations related to the unique Integral Fast Reactor (IFR) fuel cycle technology, the pyroprocess. Argonne has been developing the IFR since 1984. It is a liquid metal cooled reactor, with a unique metal alloy fuel, and it utilizes a radically new fuel cycle. An existing facility, the Hot Fuel Examination Facility-South (HFEF/S) is being modified and equipped to provide a complete demonstration of the fuel cycle. This paper will concentrate on safety aspects of the future HFEF/S operation, slated to begin late next year. HFEF/S is part of Argonne's complex of reactor test facilities located on the Idaho National Engineering Laboratory. HFEF/S was originally put into operation in 1964 as the EBR-II Fuel Cycle Facility (FCF) (Stevenson, 1987). From 1964--69 FCF operated to demonstrate an earlier and incomplete form of today's pyroprocess, recycling some 400 fuel assemblies back to EBR-II. The FCF mission was then changed to one of an irradiated fuels and materials examination facility, hence the name change to HFEF/S. The modifications consist of activities to bring the facility into conformance with today's much more stringent safety standards, and, of course, providing the new process equipment. The pyroprocess and the modifications themselves are described more fully elsewhere (Lineberry, 1987; Chang, 1987). 18 refs., 5 figs., 2 tabs.

  9. Safety analysis of irradiated nuclear fuel transportation container

    SciTech Connect

    Uspuras, E.; Rimkevicius, S.

    2007-07-01

    Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit I on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level. (authors)

  10. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    SciTech Connect

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O/sub 2/ fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO/sub 2/ fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O/sub 2/-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O/sub 2/-fueled BWR should perform similar to a UO/sub 2/-fueled BWR under all operating conditions. A (Pu/Th)O/sub 2/-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO/sub 2/-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths.

  11. 12 CFR 747.3003 - Review of order reclassifying a corporate credit union on safety and soundness criteria.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... credit union on safety and soundness criteria. 747.3003 Section 747.3003 Banks and Banking NATIONAL CREDIT UNION ADMINISTRATION REGULATIONS AFFECTING CREDIT UNIONS ADMINISTRATIVE ACTIONS, ADJUDICATIVE... corporate credit union on safety and soundness criteria. (a) Notice of proposed reclassification based...

  12. 12 CFR 747.3003 - Review of order reclassifying a corporate credit union on safety and soundness criteria.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... credit union on safety and soundness criteria. 747.3003 Section 747.3003 Banks and Banking NATIONAL CREDIT UNION ADMINISTRATION REGULATIONS AFFECTING CREDIT UNIONS ADMINISTRATIVE ACTIONS, ADJUDICATIVE... corporate credit union on safety and soundness criteria. (a) Notice of proposed reclassification based...

  13. Performance of a Fuel-Injection Spark-Ignition Engine Using a Hydrogenated Safety Fuel

    NASA Technical Reports Server (NTRS)

    Schey, Oscar W; Young, Alfred W

    1934-01-01

    This report presents the performance of a single-cylinder test engine using a hydrogenated safety fuel. The safety fuel has a flash point of 125 degrees f. (Cleveland open-dup method), which is high enough to remove most of the fire hazard, and an octane number of 95, which permits higher compression ratios to be used than are permissible with most undoped gasolines.

  14. Safety Issues with Hydrogen as a Vehicle Fuel

    SciTech Connect

    Cadwallader, Lee Charles; Herring, James Stephen

    1999-10-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of hydrogen as a vehicle fuel in automobiles. Several forms of hydrogen have been considered: gas, liquid, slush, and hydrides. The safety issues have been discussed, beginning with properties of hydrogen and the phenomenology of hydrogen combustion. Safety-related operating experiences with hydrogen vehicles have been summarized to identify concerns that must be addressed in future design activities and to support probabilistic risk assessment. Also, applicable codes, standards, and regulations pertaining to hydrogen usage and refueling have been identified and are briefly discussed. This report serves as a safety foundation for any future hydrogen safety work, such as a safety analysis or a probabilistic risk assessment.

  15. Safety Issues with Hydrogen as a Vehicle Fuel

    SciTech Connect

    L. C. Cadwallader; J. S. Herring

    1999-09-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of hydrogen as a vehicle fuel in automobiles. Several forms of hydrogen have been considered: gas, liquid, slush, and hydrides. The safety issues have been discussed, beginning with properties of hydrogen and the phenomenology of hydrogen combustion. Safety-related operating experiences with hydrogen vehicles have been summarized to identify concerns that must be addressed in future design activities and to support probabilistic risk assessment. Also, applicable codes, standards, and regulations pertaining to hydrogen usage and refueling have been identified and are briefly discussed. This report serves as a safety foundation for any future hydrogen safety work, such as a safety analysis or a probabilistic risk assessment.

  16. Double-clad nuclear-fuel safety rod

    DOEpatents

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  17. Acceptance criteria for the evaluation of Category 1 fuel cycle facility physical security plans

    SciTech Connect

    Dwyer, P.A.

    1991-10-01

    This NUREG document presents criteria developed from US Nuclear Regulatory Commission regulations for the evaluation of physical security plans submitted by Category 1 fuel facility licensees. Category 1 refers to those licensees who use or possess a formula quantity of strategic special nuclear material.

  18. Safety issues of dry fuel storage at RSWF

    SciTech Connect

    Clarksean, R.L.; Zahn, T.P.

    1995-02-01

    Safety issues associated with the dry storage of EBR-II spent fuel are presented and discussed. The containers for the fuel have been designed to prevent a leak of fission gases to the environment. The storage system has four barriers for the fission gases. These barriers are the fuel cladding, an inner container, an outer container, and the liner at the RSWF. Analysis has shown that the probability of a leak to the environment is much less than 10{sup {minus}6} per year, indicating that such an event is not considered credible. A drop accident, excessive thermal loads, criticality, and possible failure modes of the containers are also addressed.

  19. Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy

    SciTech Connect

    TRAWINSKI, B.J.

    2000-02-08

    The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system.

  20. Radiological criteria for the remediation of sites for spent fuel and radioactive waste storage in the Russian Northwest.

    PubMed

    Shandala, N K; Sneve, M K; Titov, A V; Smith, G M; Novikova, N Ya; Romanov, V V; Seregin, V A

    2008-12-01

    In the 1960s, two technical bases of the Northern Fleet were created in Northwest Russia, at Andreeva Bay in the Kola Peninsula and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, performing receipt and storage of radioactive waste and spent nuclear fuel, and are now designated sites of temporary storage (STSs). An analysis of the radiation situation at these sites demonstrates that substantial long-term remediation work will be required after the removal of the waste and spent nuclear fuel. Regulatory guidance is under development to support this work. Having in mind modern approaches to guaranteeing radiation safety, the primary regulatory focus is on a justification of dose constraints for determining acceptable residual contamination which might lead to exposure to workers and the public. For these sites, four principal options for remediation have been considered-renovation, conversion, conservation and liquidation. This paper describes a system of recommended dose constraints and derived control levels formulated for each option. The unconditional guarantee of long-term radioecological protection provides the basis for criteria development. Non-exceedance of these dose constraints and control levels implies compliance with radiological protection objectives related to the residual contamination. Dose reduction below proposed dose constraint values must also be carried out according to the optimisation principle. The developed criteria relate to the condition of the facilities and the STS areas after the termination of remediation activities. The proposed criteria for renovation, conversion, conservation and liquidation are entirely within the dose limits adopted in Russia for the management of man-made radiation sources, and are consistent with ICRP recommendations and national practice in other countries. The proposed criteria for STS remediation and new industrial (non-radiation-hazardous) facilities and buildings on

  1. MOX LTA Fuel Cycle Analyses: Nuclear and Radiation Safety

    SciTech Connect

    Pavlovitchev, A.M.

    2001-09-28

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  2. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  3. Spent Nuclear Fuel (SNF) Project Acceptance Criteria for Light Water Reactor Spent Fuel Storage System [OCRWM PER REV2

    SciTech Connect

    JOHNSON, D.M.

    2000-12-20

    As part of the decommissioning of the 324 Building Radiochemical Engineering Cells there is a need to remove commercial Light Water Reactor (LWR) spent nuclear fuel (SNF) presently stored in these hot cells. To enable fuel removal from the hot cells, the commercial LWR SNF will be packaged and shipped to the 200 Area Interim Storage Area (ISA) in a manner that satisfies site requirements for SNF interim storage. This document identifies the criteria that the 324 Building Radiochemical Engineering Cell Clean-out Project must satisfy for acceptance of the LWR SNF by the SNF Project at the 200 Area ISA. In addition to the acceptance criteria identified herein, acceptance is contingent on adherence to applicable Project Hanford Management Contract requirements and procedures in place at the time of work execution.

  4. Recommendations for ductile and brittle failure design criteria for ductile cast iron spent-fuel shipping containers

    SciTech Connect

    Schwartz, M.W.

    1984-04-01

    This report presents recommendations for establishing design and acceptance criteria for the ductile cast iron to be used for fabricating spent-fuel shipping casks. These recommendations address design criteria for preventing ductile failure, and acceptance criteria for preventing brittle fracture, based upon drop testing a flawed prototype cask.

  5. Criticality safety criteria for license review of low-level waste facilities

    SciTech Connect

    Hopper, C.M.; Odegaarden, R.H.; Parks, C.V.; Fox, P.B.

    1995-03-01

    The handling and burial of specified quantities of special nuclear material (SNM) at low-level-waste (LLW) facilities require a license from the Nuclear Regulatory Commission (NRC). With assistance from Oak Ridge National Laboratory (ORNL) staff, the NRC Office of Nuclear Material Safety and Safeguards, Low-Level-Waste and Decommissioning Projects Branch, has developed technical specifications for the nuclear criticality safety of {sup 235}U and {sup 239}Pu in LLW facilities. The objective of the development of these technical specifications was to establish a set of review criteria that are rigorously defensible that can be applied uniformly to all license applications, and that conservatively ensures that buried SNM will not pose a criticality safety concern.

  6. Multi-level multi-criteria analysis of alternative fuels for waste collection vehicles in the United States.

    PubMed

    Maimoun, Mousa; Madani, Kaveh; Reinhart, Debra

    2016-04-15

    Historically, the U.S. waste collection fleet was dominated by diesel-fueled waste collection vehicles (WCVs); the growing need for sustainable waste collection has urged decision makers to incorporate economically efficient alternative fuels, while mitigating environmental impacts. The pros and cons of alternative fuels complicate the decisions making process, calling for a comprehensive study that assesses the multiple factors involved. Multi-criteria decision analysis (MCDA) methods allow decision makers to select the best alternatives with respect to selection criteria. In this study, two MCDA methods, Technique for Order Preference by Similarity to Ideal Solution (TOPSIS) and Simple Additive Weighting (SAW), were used to rank fuel alternatives for the U.S. waste collection industry with respect to a multi-level environmental and financial decision matrix. The environmental criteria consisted of life-cycle emissions, tail-pipe emissions, water footprint (WFP), and power density, while the financial criteria comprised of vehicle cost, fuel price, fuel price stability, and fueling station availability. The overall analysis showed that conventional diesel is still the best option, followed by hydraulic-hybrid WCVs, landfill gas (LFG) sourced natural gas, fossil natural gas, and biodiesel. The elimination of the WFP and power density criteria from the environmental criteria ranked biodiesel 100 (BD100) as an environmentally better alternative compared to other fossil fuels (diesel and natural gas). This result showed that considering the WFP and power density as environmental criteria can make a difference in the decision process. The elimination of the fueling station and fuel price stability criteria from the decision matrix ranked fossil natural gas second after LFG-sourced natural gas. This scenario was found to represent the status quo of the waste collection industry. A sensitivity analysis for the status quo scenario showed the overall ranking of diesel and

  7. The maternal early warning criteria: a proposal from the national partnership for maternal safety.

    PubMed

    Mhyre, Jill M; D'Oria, Robyn; Hameed, Afshan B; Lappen, Justin R; Holley, Sharon L; Hunter, Stephen K; Jones, Robin L; King, Jeffrey C; D'Alton, Mary E

    2014-01-01

    Case reviews of maternal death have revealed a concerning pattern of delay in recognition of hemorrhage, hypertensive crisis, sepsis, venous thromboembolism, and heart failure. Early-warning systems have been proposed to facilitate timely recognition, diagnosis, and treatment for women developing critical illness. A multidisciplinary working group convened by the National Partnership for Maternal Safety used a consensus-based approach to define The Maternal Early Warning Criteria, a list of abnormal parameters that indicate the need for urgent bedside evaluation by a clinician with the capacity to escalate care as necessary in order to pursue diagnostic and therapeutic interventions. This commentary reviews the evidence supporting the use of early-warning systems, describes The Maternal Early Warning Criteria, and provides considerations for local implementation. PMID:25203897

  8. The maternal early warning criteria: a proposal from the national partnership for maternal safety.

    PubMed

    Mhyre, Jill M; DʼOria, Robyn; Hameed, Afshan B; Lappen, Justin R; Holley, Sharon L; Hunter, Stephen K; Jones, Robin L; King, Jeffrey C; DʼAlton, Mary E

    2014-10-01

    Case reviews of maternal death have revealed a concerning pattern of delay in recognition of hemorrhage, hypertensive crisis, sepsis, venous thromboembolism, and heart failure. Early-warning systems have been proposed to facilitate timely recognition, diagnosis, and treatment for women developing critical illness. A multidisciplinary working group convened by the National Partnership for Maternal Safety used a consensus-based approach to define The Maternal Early Warning Criteria, a list of abnormal parameters that indicate the need for urgent bedside evaluation by a clinician with the capacity to escalate care as necessary in order to pursue diagnostic and therapeutic interventions. This commentary reviews the evidence supporting the use of early-warning systems and describes The Maternal Early Warning Criteria, along with considerations for local implementation. PMID:25198266

  9. ENSURING ADEQUATE SAFETY WHEN USING HYDROGEN AS A FUEL

    SciTech Connect

    Coutts, D

    2007-01-22

    Demonstration projects using hydrogen as a fuel are becoming very common. Often these projects rely on project-specific risk evaluations to support project safety decisions. This is necessary because regulations, codes, and standards (hereafter referred to as standards) are just being developed. This paper will review some of the approaches being used in these evolving standards, and techniques which demonstration projects can implement to bridge the gap between current requirements and stakeholder desires. Many of the evolving standards for hydrogen-fuel use performance-based language, which establishes minimum performance and safety objectives, as compared with prescriptive-based language that prescribes specific design solutions. This is being done for several reasons including: (1) concern that establishing specific design solutions too early will stifle invention, (2) sparse performance data necessary to support selection of design approaches, and (3) a risk-adverse public which is unwilling to accept losses that were incurred in developing previous prescriptive design standards. The evolving standards often contain words such as: ''The manufacturer shall implement the measures and provide the information necessary to minimize the risk of endangering a person's safety or health''. This typically implies that the manufacturer or project manager must produce and document an acceptable level of risk. If accomplished using comprehensive and systematic process the demonstration project risk assessment can ease the transition to widespread commercialization. An approach to adequately evaluate and document the safety risk will be presented.

  10. Safety issues related to synthetic-fuels facilities

    NASA Astrophysics Data System (ADS)

    The design, siting, construction, operation, and decommissioning of coal gasification, coal liquefaction, and oil shale facilities could present safety risks both to synfuels workers and to the environmental system unless careful controls are exercised. Many of these hazards are expected to be similar to those associated with conventional mining, mineral processing, coking operations, and refining of petroleum. However, because of the chemical and physical properties of coal and shale and their products, the types of technologies to be employed, and the scales of their operations, it has been suggested that unconventional hazards may occur. Issues which summarize the deliverations and work of the Committee on Synthetic Fuels Facilities Safety in evaluating the various technologies and their potentials for unconventional safety hazards are described in Chapters 2 and 3 of the report.

  11. 12 CFR 747.2003 - Review of order reclassifying a credit union on safety and soundness criteria.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 12 Banks and Banking 7 2014-01-01 2014-01-01 false Review of order reclassifying a credit union on safety and soundness criteria. 747.2003 Section 747.2003 Banks and Banking NATIONAL CREDIT UNION... Corrective Action § 747.2003 Review of order reclassifying a credit union on safety and soundness...

  12. 12 CFR 747.2003 - Review of order reclassifying a credit union on safety and soundness criteria.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 12 Banks and Banking 6 2010-01-01 2010-01-01 false Review of order reclassifying a credit union on safety and soundness criteria. 747.2003 Section 747.2003 Banks and Banking NATIONAL CREDIT UNION... Corrective Action § 747.2003 Review of order reclassifying a credit union on safety and soundness...

  13. 12 CFR 747.2003 - Review of order reclassifying a credit union on safety and soundness criteria.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 12 Banks and Banking 6 2011-01-01 2011-01-01 false Review of order reclassifying a credit union on safety and soundness criteria. 747.2003 Section 747.2003 Banks and Banking NATIONAL CREDIT UNION... Corrective Action § 747.2003 Review of order reclassifying a credit union on safety and soundness...

  14. Preliminary safety criteria for organic watch list tanks at the Hanford site

    SciTech Connect

    Webb, A.B.; Stewart, J.L.; Turner, O.A.; Plys, M.G.; Malinovic, B.; Grigsby, J.M.; Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J.

    1995-11-01

    Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended.

  15. INPRO Assessment of an INS in the Area of Safety of Fuel Cycle Installations

    SciTech Connect

    Raj, B.; Busurin, Y.; Depisch, F.

    2006-07-01

    INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contribution to meeting the energy needs of the 21. century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. The manuals set out the information that needs to be assembled to perform an assessment and provide guidance on selecting the acceptance limits and, for a given INS, for determining the value of the indicators for comparison with the associated acceptance limits. Each manual also includes an example of a specific assessment to illustrate the guidance. This paper discusses the manual for performing an INPRO assessment in the area of safety of fuel cycle installations. The example, chosen solely for the purpose of illustrating the INPRO methodology, describes an assessment of an MOX fuel fabrication plant based on sol-gel technology and illustrates an assessment performed for an INS at an early stage of development. The safety issues and the assessment steps are presented in detail in the paper. (authors)

  16. Electronic Safety Resource Tools -- Supporting Hydrogen and Fuel Cell Commercialization

    SciTech Connect

    Barilo, Nick F.

    2014-09-29

    The Pacific Northwest National Laboratory (PNNL) Hydrogen Safety Program conducted a planning session in Los Angeles, CA on April 1, 2014 to consider what electronic safety tools would benefit the next phase of hydrogen and fuel cell commercialization. A diverse, 20-person team led by an experienced facilitator considered the question as it applied to the eight most relevant user groups. The results and subsequent evaluation activities revealed several possible resource tools that could greatly benefit users. The tool identified as having the greatest potential for impact is a hydrogen safety portal, which can be the central location for integrating and disseminating safety information (including most of the tools identified in this report). Such a tool can provide credible and reliable information from a trustworthy source. Other impactful tools identified include a codes and standards wizard to guide users through a series of questions relating to application and specific features of the requirements; a scenario-based virtual reality training for first responders; peer networking tools to bring users from focused groups together to discuss and collaborate on hydrogen safety issues; and a focused tool for training inspectors. Table ES.1 provides results of the planning session, including proposed new tools and changes to existing tools.

  17. Criteria for Corrosion Protection of Aluminum-Clad Spent Nuclear Fuel in Interim Wet Storage

    SciTech Connect

    Howell, J.P.

    1999-09-14

    Storage of aluminum-clad spent nuclear fuel at the Savannah River Site (SRS) and other locations in the U. S. and around the world has been a concern over the past decade because of the long time interim storage requirements in water. Pitting corrosion of production aluminum-clad fuel in the early 1990''s at SRS was attributed to less than optimum quality water and corrective action taken has resulted in no new pitting since 1994. The knowledge gained from the corrosion surveillance testing and other investigations at SRS over the past 8 years has provided an insight into factors affecting the corrosion of aluminum in relatively high purity water. This paper reviews some of the early corrosion issues related to aluminum-clad spent fuel at SRS, including fundamentals for corrosion of aluminum alloys. It updates and summarizes the corrosion surveillance activities supporting the future storage of over 15,000 research reactor fuel assemblies from countries over the world during the next 15-20 years. Criteria are presented for providing corrosion protection for aluminum-clad spent fuel in interim storage during the next few decades while plans are developed for a more permanent disposition.

  18. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  19. Development of water quality criteria for diesel fuel No. 2 for remediating contaminated groundwater

    SciTech Connect

    Kangas, M.J.; Proctor, D.M.; Trowbridge, K.R.

    1994-12-31

    Site-specific ambient water quality criteria (AWQC) were developed as benchmarks for back-calculating safe levels of diesel fuel No. 2 as a petroleum mixture in groundwater that could migrate to Fish Creek north of Butler, Indiana. Three types of AWQC were considered relevant according to State-modified US Environmental Protection Agency procedures: An Acute Aquatic Criterion (AAC); A Chronic Aquatic Criterion (CAC); and A Terrestrial Life Cycle Safe Concentration (TLSC). The AAC is the maximum concentration considered protective for aquatic life exposed in the zone of discharge-induced mixing and outside the zone of initial dilution. The remaining criteria applies to all areas of a stream outside the mixing zone. The CAC is intended to protect aquatic life from chronic toxic effects under a four-day average exposure. The TLSC is developed to protect terrestrial organisms that may experience a four-day average exposure to surface water as a result of consumption of aquatic organisms and water from the creek. Scientifically valid toxicological data on the water soluble fraction of diesel fuel and site-specific resident and surrogate species information were used for criterion development. An AAC of 11.4 mg/L was derived as the benchmark for back-calculating a safe level of diesel fuel in groundwater based on modeled groundwater and surface water flow from the spill area to the creek. Uncertainties and limitations of developing benchmark concentrations for mixtures are presented.

  20. Criteria for development of a database for safety evaluation of fragrance ingredients.

    PubMed

    Ford, R A; Domeyer, B; Easterday, O; Maier, K; Middleton, J

    2000-04-01

    Over 2000 different ingredients are used in the manufacture of fragrances. The majority of these ingredients have been used for many decades. Despite this long history of use, all of these ingredients need continued monitoring to ensure that each ingredient meets acceptable safety standards. As with other large databases of existing chemicals, fulfilling this need requires an organized approach to identify the most important potential hazards. One such approach, specifically considering the dermal route of exposure as the most relevant one for fragrance ingredients, has been developed. This approach provides a rational selection of materials for review and gives guidance for determining the test data that would normally be considered necessary for the elevation of safety under intended conditions of use. As a first step, the process takes into account the following criteria: quantity of use, consumer exposure, and chemical structure. These are then used for the orderly selection of materials for review with higher quantity, higher exposure, and the presence of defined structural alerts all contributing to a higher priority for review. These structural alerts along with certain exposure and volume limits are then used to develop guidelines for determining the quality and quantity of data considered necessary to support an adequate safety evaluation of the chosen materials, taking into account existing data on the substance itself as well as on closely related analogs. This approach can be considered an alternative to testing; therefore, it is designed to be conservative but not so much so as to require excessive effort when not justified. PMID:10854123

  1. Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

    DOE PAGESBeta

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; Scaglione, John M.

    2016-06-15

    We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

  2. A Criteria Standard for Conflict Resolution: A Vision for Guaranteeing the Safety of Self-Separation in NextGen

    NASA Technical Reports Server (NTRS)

    Munoz, Cesar; Butler, Ricky; Narkawicz, Anthony; Maddalon, Jeffrey; Hagen, George

    2010-01-01

    Distributed approaches for conflict resolution rely on analyzing the behavior of each aircraft to ensure that system-wide safety properties are maintained. This paper presents the criteria method, which increases the quality and efficiency of a safety assurance analysis for distributed air traffic concepts. The criteria standard is shown to provide two key safety properties: safe separation when only one aircraft maneuvers and safe separation when both aircraft maneuver at the same time. This approach is complemented with strong guarantees of correct operation through formal verification. To show that an algorithm is correct, i.e., that it always meets its specified safety property, one must only show that the algorithm satisfies the criteria. Once this is done, then the algorithm inherits the safety properties of the criteria. An important consequence of this approach is that there is no requirement that both aircraft execute the same conflict resolution algorithm. Therefore, the criteria approach allows different avionics manufacturers or even different airlines to use different algorithms, each optimized according to their own proprietary concerns.

  3. Safety aspects of fuel handling in IGCC and PFBC plants

    SciTech Connect

    Wilen, C.; Rautalin, A.

    1999-07-01

    Safety-technical characteristics of fuels, primarily biomass and, as reference, coal have been studied at VTT Energy since the year 1993. The work has related mainly to the development work of feeding and handling systems for pressurized gasification and combustion technology. This paper compares various pressurized system alternatives based primarily on lock-hopper feeding technology. A significant issue is how to arrange pressurization and sufficiently safe conditions. New alternatives to produce inert gas and the latest dust explosion suppression technology are assessed. New data on the safety-technical characteristics of renewable fuels, wastes, low-rank coals and mixtures of these, created in a research project funded by EC under the Joule 3 Programme are presented. Dust explosion testing was performed at initial pressures of up to 15 bar and temperatures of 150 C to simulate pressurized drying and handling of the biomass fuels. Inerting tests with nitrogen and flue gases were carried out to determine the requirements of non-explosive conditions. Very high explosion pressures and rates of pressure rise are measured at elevated initial pressures. The required level of inertization on dust explosions is dependent of the initial pressure and temperature. Safe operation would require an oxygen concentration of max 10 vol% in the surrounding atmosphere. The allowable oxygen concentration decreases with increasing initial temperature. For all powders tested this decrease was more or less the same, around 1--3 vol% per 100 C temperature rise. Suppression tests were performed in co-operation with Coal Technology Development Division of British Coal Corp. to assess the usability of this explosion protection method for biomass fuels in elevated conditions.

  4. 30 CFR 75.1903 - Underground diesel fuel storage facilities and areas; construction and safety precautions.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Underground diesel fuel storage facilities and areas; construction and safety precautions. 75.1903 Section 75.1903 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Diesel-Powered Equipment...

  5. 12 CFR 747.3003 - Review of order reclassifying a corporate credit union on safety and soundness criteria.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... the record nor the Uniform Rules of Practice and Procedure (12 CFR part 747) will apply to an informal... HEARINGS, RULES OF PRACTICE AND PROCEDURE, AND INVESTIGATIONS Issuance, Review and Enforcement of Orders... corporate credit union on safety and soundness criteria. (a) Notice of proposed reclassification based...

  6. 12 CFR 747.3003 - Review of order reclassifying a corporate credit union on safety and soundness criteria.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... the record nor the Uniform Rules of Practice and Procedure (12 CFR part 747) will apply to an informal... HEARINGS, RULES OF PRACTICE AND PROCEDURE, AND INVESTIGATIONS Issuance, Review and Enforcement of Orders... corporate credit union on safety and soundness criteria. (a) Notice of proposed reclassification based...

  7. Enhancing Credibility of Chemical Safety Studies: Emerging Consensus on Key Assessment Criteria

    PubMed Central

    Conrad, James W.; Becker, Richard A.

    2011-01-01

    Objectives We examined the extent to which consensus exists on the criteria that should be used for assessing the credibility of a scientific work, regardless of its funding source, and explored how these criteria might be implemented. Data sources Three publications, all presented at a session of the 2009 annual meeting of the Society for Risk Analysis, have proposed a range of criteria for evaluating the credibility of scientific studies. At least two other similar sets of criteria have recently been proposed elsewhere. Data extraction/synthesis In this article we review these criteria, highlight the commonalities among them, and integrate them into a list of 10 criteria. We also discuss issues inherent in any attempt to implement the criteria systematically. Conclusions Recommendations by many scientists and policy experts converge on a finite list of criteria for assessing the credibility of a scientific study without regard to funding source. These criteria should be formalized through a consensus process or a governmental initiative that includes discussion and pilot application of a system for reproducibly implementing them. Formal establishment of such a system should enable the debate regarding chemical studies to move beyond funding issues and focus on scientific merit. PMID:21163723

  8. Unreviewed safety question evaluation of 100 K West fuel canister gas and liquid sampling

    SciTech Connect

    Alwardt, L.D.

    1995-01-12

    The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation for the gas and liquid sampling activities associated with the fuel characterization program at the 100 K West (KW) fuel storage basin. The scope of this safety evaluation is limited to the movement of canisters between the main storage basin, weasel pit, and south loadout pit transfer channel (also known as the decapping station); gas and liquid sampling of fuel canisters in the weasel pit; mobile laboratory preliminary sample analysis in or near the 105 KW basin building; and the placement of sample containers in an approved shipping container. It was concluded that the activities and potential accident consequences associated with the gas and liquid sampling of 100 KW fuel canisters are bounded by the current safety basis documents and do not constitute an Unreviewed Safety Question.

  9. An advanced deterministic method for spent fuel criticality safety analysis

    SciTech Connect

    DeHart, M.D.

    1998-01-01

    Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte Carlo methods for criticality calculations. Monte Carlo has become popular because of its capability to model complex, non-orthogonal configurations or fissile materials, typical of real world problems. Over the last few years, however, interest in determinist transport methods has been revived, due shortcomings in the stochastic nature of Monte Carlo approaches for certain types of analyses. Specifically, deterministic methods are superior to stochastic methods for calculations requiring accurate neutron density distributions or differential fluxes. Although Monte Carlo methods are well suited for eigenvalue calculations, they lack the localized detail necessary to assess uncertainties and sensitivities important in determining a range of applicability. Monte Carlo methods are also inefficient as a transport solution for multiple pin depletion methods. Discrete ordinates methods have long been recognized as one of the most rigorous and accurate approximations used to solve the transport equation. However, until recently, geometric constraints in finite differencing schemes have made discrete ordinates methods impractical for non-orthogonal configurations such as reactor fuel assemblies. The development of an extended step characteristic (ESC) technique removes the grid structure limitations of traditional discrete ordinates methods. The NEWT computer code, a discrete ordinates code built upon the ESC formalism, is being developed as part of the SCALE code system. This paper will demonstrate the power, versatility, and applicability of NEWT as a state-of-the-art solution for current computational needs.

  10. Criticality safety considerations for MSRE fuel drain tank uranium aggregation

    SciTech Connect

    Hollenbach, D.F.; Hopper, C.M.

    1997-03-01

    This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations. Since the salt was transferred to the FDTs in 1969, radiological and chemical reactions have been converting the uranium and fluorine in the salt to UF{sub 6} and free fluorine. Significant amounts of uranium (at least 3 kg) and fluorine have migrated out of the FDTs and into the off-gas system (OGS) and the auxiliary charcoal bed (ACB). The loss of uranium and fluorine from the salt changes the chemical properties of the salt sufficiently to possibly allow the reduction of the UF{sub 4} in the salt to uranium metal as the salt is remelted prior to removal. It has been postulated that up to 9 kg of the maximum 19.4 kg of uranium in one FDT could be reduced to metal and concentrated. This study shows that criticality becomes a concern when more than 5 kg of uranium concentrates to over 8 wt% of the salt in a favorable geometry.

  11. Safety analysis of B and W Standard PWR using thorium-based fuels

    SciTech Connect

    Uotinen, V.O.; Carroll, W.P.; Jones, H.M.; Toops, E.C.

    1980-06-01

    A study was performed to assess the safety and licenseability of the Babcock and Wilcox standard 205-fuel assembly PWR when it is fueled with three types of thoria-based fuels denatured (/sup 233/U//sup 238/U-Th)O/sub 2/, denatured (/sup 235//U/sup 238/U-Th)O/sub 2/, and (Th-Pu)O/sub 2/. Selected transients were analyzed using typical PWR safety analysis calculational methods. The results support the conclusion that it is feasible from a safety standpoint to utilize either of the denatured urania-thoria fuels in the standard B and W plant. In addition, it appears that the use of thoria-plutonia fuels would probably also be feasible. These tentative conclusions depend on a data that is more limited than that available for UO/sub 2/ fuels.

  12. 21 CFR 70.42 - Criteria for evaluating the safety of color additives.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... experiments to determine its safety, the Commissioner will advise a person who wishes to establish the safety of a color additive whether he believes the experiments planned will yield data adequate for...

  13. 21 CFR 70.42 - Criteria for evaluating the safety of color additives.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... experiments to determine its safety, the Commissioner will advise a person who wishes to establish the safety of a color additive whether he believes the experiments planned will yield data adequate for...

  14. 21 CFR 70.42 - Criteria for evaluating the safety of color additives.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... experiments to determine its safety, the Commissioner will advise a person who wishes to establish the safety of a color additive whether he believes the experiments planned will yield data adequate for...

  15. 21 CFR 70.42 - Criteria for evaluating the safety of color additives.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... experiments to determine its safety, the Commissioner will advise a person who wishes to establish the safety of a color additive whether he believes the experiments planned will yield data adequate for...

  16. Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

    SciTech Connect

    Lam, P.; Sindelar, R.

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  17. Development of flaw acceptance criteria for aging management of spent nuclear fuel multi-purpose canisters

    SciTech Connect

    Lam, Poh -Sang; Sindelar, Robert L.

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic in-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  18. Greenhouse Gas and Criteria Air Pollutant Emission Reductions from Forest Fuel Treatment Projects in Placer County, California

    NASA Astrophysics Data System (ADS)

    Saah, D. S.; Moritz, M.; Ganz, D. J.; Stine, P. A.; Moody, T.

    2010-12-01

    Years of successful fire suppression activities have left forests unnaturally dense, overstocked, and with high hazardous fuel loads. Wildfires, particularly those of high severity, may dramatically reduce carbon stocks and convert forested lands from carbon sinks to decades-long carbon sources . Forest resource managers are currently pursuing fuels reduction and mitigation strategies to reduce wildfire risk and maintain carbon stocks. These projects include selective thinning and removal of trees and brush to return forest ecosystems to more natural stocking levels, resulting in a more fire-resilient forest that in theory would retain higher carry capacity for standing above ground carbon. Resource managers are exploring the possibility of supporting these local forest management projects by offering greenhouse gas (GHG) offsets to project developers that require GHG emissions mitigation. Using robust field data, this research project modeled three types of carbon benefits that could be realized from forest management: 1. Fuels treatments in the study area were shown to reduce the GHG and Criteria Air Pollutant emissions from wildfires by decreasing the probability, extent, and severity of fires and the corresponding loss in forest carbon stocks; 2. Biomass utilization from fuel treatment was shown to reduce GHG and Criteria Air Pollutant emissions over the duration of the fuels treatment project compared to fossil fuel energy. 3. Management and thinning of forests in order to stimulate growth, resulting in more rapid uptake of atmospheric carbon and approaching a carbon carrying capacity stored in a forest ecosystem under prevailing environmental conditions and natural disturbance regimes.

  19. 49 CFR Appendix to Subpart H of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Non-North America-Domiciled...

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Evaluation Criteria for Non-North America-Domiciled Motor Carriers Appendix to Subpart H of Part 385... America-Domiciled Motor Carriers I. General (a) FMCSA will perform a safety audit of each non-North...-North America-domiciled carrier's basic safety management controls. (b) The safety audit is a review...

  20. Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

    SciTech Connect

    1980-10-01

    This document is arranged in three volumes and reports on progress in the Liquefied Gaseous Fuels (LGF) Safety and Environmental Control Assessment Program made in fiscal Year (FY)-1979 and early FY-1980. Volume 3 contains reports from 6 government contractors on LPG, anhydrous ammonia, and hydrogen energy systems. Report subjects include: simultaneous boiling and spreading of liquefied petroleum gas (LPG) on water; LPG safety research; state-of-the-art of release prevention and control technology in the LPG industry; ammonia: an introductory assessment of safety and environmental control information; ammonia as a fuel, and hydrogen safety and environmental control assessment.

  1. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2002-06-26

    Document presents analysis performed to demonstrate criticality safety of packaging spent PWR fuel assemblies currently located at the 324 Building into a NAC-1 cask. Interim storage of the cask is also documented.

  2. Applications of nuclear data covariances to criticality safety and spent fuel characterization

    SciTech Connect

    Williams, Mark L; Ilas, Germina; Marshall, William BJ J; Rearden, Bradley T

    2014-01-01

    Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

  3. Applications of Nuclear Data Covariances to Criticality Safety and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-04-01

    Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

  4. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  5. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    SciTech Connect

    Shedrow, C.B.

    1999-11-29

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected.

  6. 49 CFR Appendix C to Part 236 - Safety Assurance Criteria and Processes

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... improperly designed human-machine interface; installation and maintenance errors; and errors associated with... Applications: Specification and Demonstration of Reliability, Availability, Maintainability and Safety...

  7. 40 CFR 600.115-08 - Criteria for determining the fuel economy label calculation method for 2011 and later model year...

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy and Carbon-Related Exhaust Emission Regulations... criteria to determine if the derived 5-cycle method for determining fuel economy label values, as...

  8. 49 CFR Appendix C to Part 236 - Safety Assurance Criteria and Processes

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    .... The product design must sufficiently incorporate human factors engineering that is appropriate to the... principles when designing and demonstrating the safety of products covered by subpart H or I of this part. In... associated with the design principle not followed. (1) System safety under normal operating conditions....

  9. 49 CFR Appendix C to Part 236 - Safety Assurance Criteria and Processes

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... results of system safety analyses provided in the RSPP, PSP, PTCIP, PTCDP, and PTCSP documents as appropriate. An analysis performed under this appendix must: (1) Address each of the safety principles of... system operation. Hazards categorized as unacceptable, which are determined by hazard analysis, must...

  10. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect

    Durant, W.S.; Perkins, W.C.; Lee, R.; Stoddard, D.H.

    1982-05-20

    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.

  11. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect

    Perkins, W.C.; Durant, W.S.; Dexter, A.H.

    1980-12-01

    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.

  12. Criticality Safety and Sensitivity Analyses of PWR Spent Nuclear Fuel Repository Facilities

    SciTech Connect

    Maucec, Marko; Glumac, Bogdan

    2005-01-15

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based storage and dry transport containers under various loading patterns and moderating conditions. To comply with standard safety requirements, fresh 4.25% enriched nuclear fuel was assumed. The impact of potential optimum moderation due to water steam or foam formation as well as of different interpretations, of neutron multiplication through varying the system boundary conditions was elaborated. The simulations indicate that in the case of compact (all rack locations filled with fresh fuel) single or 'double tiering' loading, the supercriticality can occur under the conditions of enhanced neutron moderation, due to accidentally reduced density of cooling water. Under standard operational conditions the effective multiplication factor (k{sub eff}) of pool-based storage facility remains below the specified safety limit of 0.95. The nuclear safety requirements are fulfilled even when the fuel elements are arranged at a minimal distance, which can be initiated, for example, by an earthquake. The dry container in its recommended loading scheme with 26 fuel elements represents a safe alternative for the repository of fresh fuel. Even in the case of complete water flooding, the k{sub eff} remains below the specified safety level of 0.98. The criticality safety limit may however be exceeded with larger amounts of loaded fuel assemblies (i.e., 32). Additional Monte Carlo criticality safety analyses are scheduled to consider the 'burnup credit' of PWR spent nuclear fuel, based on the ongoing calculation of typical burnup activities.

  13. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    SciTech Connect

    Garvin, L.J.

    1995-11-01

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

  14. Management concepts and safety applications for nuclear fuel facilities

    SciTech Connect

    Eisner, H.; Scotti, R.S.; Delicate, W.S.

    1995-05-01

    This report presents an overview of effectiveness of management control of safety. It reviews several modern management control theories as well as the general functions of management and relates them to safety issues at the corporate and at the process safety management (PSM) program level. Following these discussions, structured technique for assessing management of the safety function is suggested. Seven modern management control theories are summarized, including business process reengineering, the learning organization, capability maturity, total quality management, quality assurance and control, reliability centered maintenance, and industrial process safety. Each of these theories is examined for-its principal characteristics and implications for safety management. The five general management functions of planning, organizing, directing, monitoring, and integrating, which together provide control over all company operations, are discussed. Under the broad categories of Safety Culture, Leadership and Commitment, and Operating Excellence, key corporate safety elements and their subelements are examined. The three categories under which PSM program-level safety issues are described are Technology, Personnel, and Facilities.

  15. 78 FR 28275 - Office of Commercial Space Transportation; Safety Approval Performance Criteria

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-14

    ... Federal Aviation Administration Office of Commercial Space Transportation; Safety Approval Performance... hypobaric chamber training for crew and space flight participants to experience and demonstrate knowledge of...), FAA Office of Commercial Space Transportation (AST), 800 Independence Avenue SW., Room 331,...

  16. 77 FR 58607 - Office of Commercial Space Transportation Safety Approval Performance Criteria

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-21

    ... Federal Aviation Administration Office of Commercial Space Transportation Safety Approval Performance..., Licensing and Evaluation Division (AST-200), FAA Office of Commercial Space Transportation (AST), 800... Space Transportation. BILLING CODE 4910-13-P...

  17. Potential health and safety impacts from distribution and storage of alcohol fuels

    SciTech Connect

    Rosenberg, S.E.; Gasper, J.R.

    1980-06-01

    This assessment includes three major sections. Section 1 is a synopsis of literature on the health and safety aspects of neat alcohols, alcohol-gasoline blends, and typical gasoline. Section 2 identifies the toxic properties of each fuel type and describes existing standards and regulations and suggests provisions for establishing others. Section 3 analyzes the major safety and health risks that would result from the increased use of each type of alcohol fuel. Potential accidents are described and their probable impacts on occupational and public populations are determined. An attempt was made to distill the important health and safety issues and to define gaps in our knowledge regarding alcohol fuels to highlight the further research needed to circumvent potential helth and safety problems.

  18. Packaging Strategies for Criticality Safety for "Other" DOE Fuels in a Repository

    SciTech Connect

    Larry L Taylor

    2004-06-01

    Since 1998, there has been an ongoing effort to gain acceptance of U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in the national repository. To accomplish this goal, the fuel matrix was used as a discriminating feature to segregate fuels into nine distinct groups. From each of those groups, a characteristic fuel was selected and analyzed for criticality safety based on a proposed packaging strategy. This report identifies and quantifies the important criticality parameters for the canisterized fuels within each criticality group to: (1) demonstrate how the “other” fuels in the group are bounded by the baseline calculations or (2) allow identification of individual type fuels that might require special analysis and packaging.

  19. Environmental and safety issues of the fusion fuel cycle

    SciTech Connect

    Crocker, J.G.

    1980-01-01

    This paper discusses the environmental and safety concerns inherent in the development of fusion energy, and the current Department of Energy programs seeking to: (1) develop safe and reliable techniques for tritium control; (2) reduce the quantity of activation products produced; and (3) provide designs to limit the potential for accidents that could result in release of radioactive materials. Because of the inherent safety features of fusion and the early start that has been made in safety problem recognition and solution, fusion should be among the lower risk technologies for generation of commercial power.

  20. 49 CFR Appendix A to Subpart E of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Mexico-Domiciled Motor...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 5 2011-10-01 2011-10-01 false Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Mexico-Domiciled Motor Carriers A Appendix A to Subpart E of Part 365 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION FEDERAL...

  1. TOPSIS multiple-criteria decision support analysis for material selection of metallic bipolar plates for polymer electrolyte fuel cell

    NASA Astrophysics Data System (ADS)

    Shanian, A.; Savadogo, O.

    Several kinds of metallic bipolar plates for PEMFCs are currently being developed in order to meet the demands of cost reduction, stack volume, lower weight and enhanced power density. This work shows an application of the Technique of ranking Preferences by Similarity to the Ideal Solution (TOPSIS) Multiple Attribute Decision Making (MADM) method for solving the material selection problem of metallic bipolar plates for polymer electrolyte fuel cell (PEFC), which often involves multiple and conflicting objectives. The proposed methodological tool can aid the material designer in the modeling and selection of suitable materials according to a set of predefined criteria. After introducing the theoretical background, a case study is presented for the material selection of a bipolar plate in a PEFC. A list of all possible choices, from the best to the worst materials, is obtained by taking into account all the material selection criteria, including the cost of production. A user-defined code in Mathematica has been developed to facilitate the implementation of the method. It was shown that the optimum value of each criterion is independent of other criteria values (i.e., no interaction is allowed). The proposed approach may be applied to other problems of material selection of fuel cell components.

  2. 76 FR 30232 - Office of Commercial Space Transportation Safety Approval Performance Criteria

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-24

    ... Federal Aviation Administration Office of Commercial Space Transportation Safety Approval Performance... associated with suborbital space flight. The reduced gravity levels are: --0.00 g 0.05 g for 17 continuous... Division (AST-200), FAA Office of Commercial Space Transportation (AST), 800 Independence Avenue, SW.,...

  3. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    SciTech Connect

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm (12 in).

  4. Criteria for the Research Institute for Fragrance Materials, Inc. (RIFM) safety evaluation process for fragrance ingredients.

    PubMed

    Api, A M; Belsito, D; Bruze, M; Cadby, P; Calow, P; Dagli, M L; Dekant, W; Ellis, G; Fryer, A D; Fukayama, M; Griem, P; Hickey, C; Kromidas, L; Lalko, J F; Liebler, D C; Miyachi, Y; Politano, V T; Renskers, K; Ritacco, G; Salvito, D; Schultz, T W; Sipes, I G; Smith, B; Vitale, D; Wilcox, D K

    2015-08-01

    The Research Institute for Fragrance Materials, Inc. (RIFM) has been engaged in the generation and evaluation of safety data for fragrance materials since its inception over 45 years ago. Over time, RIFM's approach to gathering data, estimating exposure and assessing safety has evolved as the tools for risk assessment evolved. This publication is designed to update the RIFM safety assessment process, which follows a series of decision trees, reflecting advances in approaches in risk assessment and new and classical toxicological methodologies employed by RIFM over the past ten years. These changes include incorporating 1) new scientific information including a framework for choosing structural analogs, 2) consideration of the Threshold of Toxicological Concern (TTC), 3) the Quantitative Risk Assessment (QRA) for dermal sensitization, 4) the respiratory route of exposure, 5) aggregate exposure assessment methodology, 6) the latest methodology and approaches to risk assessments, 7) the latest alternatives to animal testing methodology and 8) environmental risk assessment. The assessment begins with a thorough analysis of existing data followed by in silico analysis, identification of 'read across' analogs, generation of additional data through in vitro testing as well as consideration of the TTC approach. If necessary, risk management may be considered. PMID:25510979

  5. ASCO steam generators operating experience. Safety criteria for defect management and effectiveness of preventive measures

    SciTech Connect

    Toribio, E.L.

    1997-02-01

    ASCO NPP is a two W-PWR 930 Mwe Units. Each Unit is provided with three Westinghouse Model D3 steam generators which are of preheater type and Inconel 600 MA as tube material. The Secondary side was designed and erected with copper alloys. Unit I: 81.072 EFPH, and Unit II: 69.720 EFPH. The results of the Eddy Currents Inspections performed during the first refueling outage showed Denting at tube support plates and PWSCC at roll transition zone in Unit I and Denting in Unit II. Later inspections showed other types of damages, such as: (1) ODSCC at tube support plates intersections. (2) Circumferential cracks OD and ID at roll transition zone. (3) Wear at antivibration bars and preheater baffles level. Consequently, in order to limit the plugging rate, A.N. ASCO decided to license new plugging criteria in addition to the 40% depth criterion included in Technical Specification. The new licensing criteria and surveillance requirements, varying with tube zone, are explained in the paper.

  6. A Multi-Criteria Decision Analysis Model to Assess the Safety of Botanicals Utilizing Data on History of Use

    PubMed Central

    Neely, T.; Walsh-Mason, B.; Russell, P.; Horst, A. Van Der; O’Hagan, S.; Lahorkar, P.

    2011-01-01

    Botanicals (herbal materials and extracts) are widely used in traditional medicines throughout the world. Many have an extensive history of safe use over several hundreds of years. There is now a growing consumer interest in food and cosmetic products, which contain botanicals. There are many publications describing the safety assessment approaches for botanicals, based on the history of safe use. However, they do not define what constitutes a history of safe use, a decision that is ultimately a subjective one. The multi-criteria decision analysis (MCDA), is a model that has been developed, which assesses the safety of botanical ingredients using a history of use approach. The model evaluates the similarity of the botanical ingredient of interest to its historic counterpart – the comparator, the evidence supporting the history of use, and any evidence of concern. The assessment made is whether a botanical ingredient is as safe as its comparator botanical, which has a history of use. In order to establish compositional similarity between the botanical ingredient and its comparator, an analytical ‘similarity scoring’ approach has been developed. Applicability of the model is discussed with an example, Brahmi ( Bacopa monnieri). This evolution of the risk assessment of botanicals gives an objective, transparent, and transferable safety assessment approach. PMID:22025816

  7. Increasing the Fuel Economy and Safety of New Light-DutyVehicles

    SciTech Connect

    Wenzel, Tom; Ross, Marc

    2006-09-18

    One impediment to increasing the fuel economy standards forlight-duty vehicles is the long-standing argument that reducing vehiclemass to improve fuel economy will inherently make vehicles less safe.This technical paper summarizes and examines the research that is citedin support of this argument, and presents more recent research thatchallenges it. We conclude that the research claiming that lightervehicles are inherently less safe than heavier vehicles is flawed, andthat other aspects of vehicle design are more important to the on-roadsafety record of vehicles. This paper was prepared for a workshop onexperts in vehicle safety and fuel economy, organized by the William andFlora Hewlett Foundation, to discuss technologies and designs that can betaken to simultaneously improve vehicle safety and fuel economy; theworkshop was held in Washington DC on October 3, 2006.

  8. Direct-hydrogen-fueled proton-exchange-membrane fuel cell system for transportation applications. Hydrogen vehicle safety report

    SciTech Connect

    Thomas, C.E.

    1997-05-01

    This report reviews the safety characteristics of hydrogen as an energy carrier for a fuel cell vehicle (FCV), with emphasis on high pressure gaseous hydrogen onboard storage. The authors consider normal operation of the vehicle in addition to refueling, collisions, operation in tunnels, and storage in garages. They identify the most likely risks and failure modes leading to hazardous conditions, and provide potential countermeasures in the vehicle design to prevent or substantially reduce the consequences of each plausible failure mode. They then compare the risks of hydrogen with those of more common motor vehicle fuels including gasoline, propane, and natural gas.

  9. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  10. Task D: Hydrogen safety analysis

    SciTech Connect

    Swain, M.R.; Sievert, B.G.; Swain, M.N.

    1996-10-01

    This report covers two topics. The first is a review of codes, standards, regulations, recommendations, certifications, and pamphlets which address safety of gaseous fuels. The second is an experimental investigation of hydrogen flame impingement. Four areas of concern in the conversion of natural gas safety publications to hydrogen safety publications are delineated. Two suggested design criteria for hydrogen vehicle fuel systems are proposed. It is concluded from the experimental work that light weight, low cost, firewalls to resist hydrogen flame impingement are feasible.

  11. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  12. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  13. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128 Section 72.128 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL,...

  14. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  15. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  16. Inherent safety advantages of carbide fuel systems and technical issues regarding natural convection in LMRs

    SciTech Connect

    Barthold, W.P.

    1984-08-01

    The scope of work is to summarize inherent safety advantages that are unique to the use of a carbide based fuel system and to summarize the technical issues regarding natural convection flow in LMFBR cores. As discussed in this report, carbide fuel provides the designer with far greater flexibility than oxide fuel. Carbide fuel systems can be designed to eliminate major accident initiators. They turn quantitative advantages into a qualitative advantage. The author proposed to LANL a series of core design and component concepts that would greatly enhance the safety of carbide over oxide systems. This report cites a series of safety advantages which potentially exist for a carbide fuel system. Natural convection issues have not been given much attention in the past. Only during the last few years has this issue been addressed in some detail. Despite claims to the contrary by some of the LMR contractors, the author does not think that the natural convection phenomena is fully understood. Some of the approximations made in natural convection transient analyses have probably a greater impact on calculated transient temperatures than the effects under investigation. Only integral in-pile experimental data and single assembly out-of-pile detailed data are available for comparisons with analytical models and correlations. Especially for derated cores, the natural convection capability of a LMR should be far superior to that of a LWR. The author ranks the natural convection capability of the LMR as the most important inherent safety feature.

  17. Safety team assessments at NRC (Nuclear Regulatory Commission)-licensed fuel facilities

    SciTech Connect

    Sjoblom, G.L.

    1988-01-01

    Following the hydraulic rupture of a UF cylinder at the Sequoyah Fuels Facility on January 4, 1986, the US Nuclear Regulatory Commission's (NRC's) executive director for operations (EDO) established an augmented inspection team to investigate the accident. The investigation is reported in NUREG-1179. The EDO then formed a lessons-learned group to report on the action NRC might reasonably take to prevent similar accidents. The group's recommendations are reported in NUREG-1198. In addition, the EDO formed an independent materials safety regulation review study group (MSRRSG) to review the licensing and inspection program for NRC-licensed fuel cycle and materials facilities. During the same period of time that the MSRRSG report was being prepared and evaluated, the staff undertook an independent action to assess operational safety at each of the 12 major fuel facilities licensed by the NRC. The facilities included the 2 facilities producing uranium hexafluoride, the 7 facilities producing commercial nuclear reactor fuel, and the 3 facilities producing naval reactor fuel. The most important safety issues identified as needing attention by licensees were in the areas of fire protection, chemical hazards identification and mitigation, management controls or quality assurance, safety-related instrumentation and maintenance, and emergency preparedness.

  18. Additional Studies of the Criticality Safety of Failed Used Nuclear Fuel

    SciTech Connect

    Marshall, William BJ J; Wagner, John C

    2013-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories and specific configurations were evaluated to understand trends and quantify the consequences of worst-case potential reconfiguration progressions. These results will be summarized here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g., >20% keff). It can be concluded that the consequences of credible fuel failure configurations from ES or transportation following ES are manageable (e.g., <5% keff). The current work expands on these efforts and examines some modified scenarios and modified approaches to investigate the effectiveness of some techniques for reducing the calculated increase in keff. The areas included here are more realistic modeling of some assembly types and the effect of reconfiguration of some assemblies in the storage and transportation canister.

  19. The JRC-ITU approach to the safety of advanced nuclear fuel cycles

    SciTech Connect

    Fanghaenel, T.; Rondinella, V.V.; Somers, J.; Konings, R.; Erdmann, N.; Uffelen, P. van; Glatz, J.P.

    2013-07-01

    The JRC-ITU safety studies of advanced fuels and cycles adopt two main axes. First the full exploitation of still available and highly relevant knowledge and samples from past fuel preparation and irradiation campaigns (complementing the limited number of ongoing programmes). Secondly, the shift of focus from simple property measurement towards the understanding of basic mechanisms determining property evolution and behaviour of fuel compounds during normal, off-normal and accident conditions. The final objective of the second axis is the determination of predictive tools applicable to systems and conditions different from those from which they were derived. State of the art experimental facilities, extensive networks of partnerships and collaboration with other organizations worldwide, and a developing programme for training and education are essential in this approach. This strategy has been implemented through various programs and projects. The SUPERFACT programme constitutes the main body of existing knowledge on the behavior in-pile of MOX fuel containing minor actinides. It encompassed all steps of a closed fuel cycle. Another international project investigating the safety of a closed cycle is METAPHIX. In this case a U-Pu19-Zr10 metal alloy containing Np, Am and Cm constitutes the fuel. 9 test pins have been prepared and irradiated. In addition to the PIE (Post Irradiation Examination), pyrometallurgical separation of the irradiated fuel has been performed, to demonstrate all the steps of a multiple recycling closed cycle and characterize their safety relevant aspects. Basic studies like thermodynamic fuel properties, fuel-cladding-coolant interactions have also been carried out at JRC-ITU.

  20. A new principle for low-cost hydrogen sensors for fuel cell technology safety

    NASA Astrophysics Data System (ADS)

    Liess, Martin

    2014-03-01

    Hydrogen sensors are of paramount importance for the safety of hydrogen fuel cell technology as result of the high pressure necessary in fuel tanks and its low explosion limit. I present a novel sensor principle based on thermal conduction that is very sensitive to hydrogen, highly specific and can operate on low temperatures. As opposed to other thermal sensors it can be operated with low cost and low power driving electronics. On top of this, as sensor element a modified standard of-the shelf MEMS thermopile IR-sensor can be used. The sensor principle presented is thus suited for the future mass markets of hydrogen fuel cell technology.S

  1. A new principle for low-cost hydrogen sensors for fuel cell technology safety

    SciTech Connect

    Liess, Martin

    2014-03-24

    Hydrogen sensors are of paramount importance for the safety of hydrogen fuel cell technology as result of the high pressure necessary in fuel tanks and its low explosion limit. I present a novel sensor principle based on thermal conduction that is very sensitive to hydrogen, highly specific and can operate on low temperatures. As opposed to other thermal sensors it can be operated with low cost and low power driving electronics. On top of this, as sensor element a modified standard of-the shelf MEMS thermopile IR-sensor can be used. The sensor principle presented is thus suited for the future mass markets of hydrogen fuel cell technology.S.

  2. The importance of safety in achieving the widespread use of hydrogen as a fuel

    SciTech Connect

    Edeskuty, F.J.

    1997-09-01

    The advantages of hydrogen fuel have been adequately demonstrated on numerous occasions. However, two major disadvantages have prevented any significant amount of corresponding development. These disadvantages have been in the economics of producing sufficient quantities of hydrogen and in the safety (both real and perceived) of its use. To date work has mostly been properly centered on solving the economic problems. However, a greater effort on the safety of new hydrogen systems now being proposed also deserves consideration. To achieve the greatest safety in the expansion of the use of hydrogen into its wide-spread use as a fuel, attention must be given to four considerations. These are, obtaining knowledge of all the physical principles involved in the new uses, having in place the regulations that allow the safe interfacing of the new systems, designing and constructing the new systems with safety in mind, and the training of the large number of people that will become the handlers of the hydrogen. Existing organizations that produce, transport, or use hydrogen on a large scale have an excellent safety record. This safety record comes as a consequence of dedicated attention to the above-mentioned principles. However, where these principles were not closely followed, accidents have resulted. Some examples can be cited. As the use of hydrogen becomes more widespread, there must be a mechanism for assuring the universal application of these principles. Larger and more numerous fleet operations with hydrogen fuel may be the best way to begin the indoctrination of the general public to the more general use of hydrogen fuel. Demonstrated safe operation with hydrogen is vital to its final acceptance as the fuel of choice.

  3. Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

    SciTech Connect

    Stevens, P.F.

    1994-10-13

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment.

  4. 75 FR 12123 - Federal Motor Vehicle Safety Standards; Side Impact Protection; Fuel System Integrity; Electric...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-15

    ... Standards; Side Impact Protection; Fuel System Integrity; Electric-Powered Vehicles: Electrolyte Spillage and Electrical Shock Protection AGENCY: National Highway Traffic Safety Administration (NHTSA... Standard (FMVSS) No. 214, ``Side impact protection,'' (72 FR 51908, Docket No. NHTSA-29134).\\1\\ Until...

  5. Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

    SciTech Connect

    Taylor, L.L.; Shikashio, R.

    1993-09-01

    The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms.

  6. Secondary Protection for 70 MPa Fueling - A White Paper from the Hydrogen Safety Panel

    SciTech Connect

    Weiner, Steven C.; Kallman, Richard A.

    2009-07-01

    In developing a 70 megapascal (MPa) fueling infrastructure, it is critical to ensure that a vehicle equipped with a lower service pressure fuel tank is never filled from a 70 MPa fueling source. Filling of a lower service pressure vehicle at a 70 MPa fueling source is likely to result in a catastrophic event with severe injuries or fatalities. The Hydrogen Safety Panel recommends that DOE undertake a two-step process to address this issue: 1. Perform an independent risk analysis of a 70MPa dispenser filling a lower pressure vehicle tank and develop different approaches for prevention and mitigation to meet an acceptable level of safety. Cost effectiveness, reliability, advantages and disadvantages are among the factors that should be evaluated for each approach considered. 2. Until such time as this analysis is complete and any recommended actions implemented, communicate the potential risk to responsible parties and strongly encourage those parties to add a secondary layer of protection to the existing system of mechanically non-interchangeable nozzles/receptacles. This will reduce the probability of a pressure mismatch during this developmental phase for hydrogen fuel cell vehicles and infrastructure. This step can be reassessed after further analysis is completed and the need and effectiveness of secondary protection methods are evaluated. This paper provides background discussion of the problem, current safety systems and strategy and examples of potential future solutions to support the above recommendations.

  7. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    SciTech Connect

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J.

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool

  8. Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions

    SciTech Connect

    Glumac, B; Ravnik, M.; Logar, M.

    1997-02-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10{sup {minus}6} per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool.

  9. CSER 01-011 Criticality Safety Evaluation for Light Water Reactor Fuel in NAC-1 Casks

    SciTech Connect

    ERICKSON, D.G.

    2001-11-01

    This analysis references a previous analysis (Larson, 1999) for a qualitative acceptability argument, and an appropriate ANS standard (ANS, 1998), and a reference (Clark 1966) for safe cylinder diameters as a function of {sup 235}U enrichment in UO{sub 2} fuel rods (pins) in water, for a quantitative acceptability argument. The previous analysis established the criticality safety of PWR assemblies without broken pins, pin segments or powdered fuel. This addendum extends the previous range of applicability in accordance with the controlling procedure (FH 2001b). The operation, when conducted according to the established limits stated in this document, complies with the incredibility principle. The evaluations demonstrated criticality safety for PWR assemblies with broken pins and pin segments of UO{sub 2} fuel.

  10. Integrated indicator to evaluate vehicle performance across: Safety, fuel efficiency and green domains.

    PubMed

    Torrao, G; Fontes, T; Coelho, M; Rouphail, N

    2016-07-01

    In general, car manufacturers face trade-offs between safety, efficiency and environmental performance when choosing between mass, length, engine power, and fuel efficiency. Moreover, the information available to the consumers makes difficult to assess all these components at once, especially when aiming to compare vehicles across different categories and/or to compare vehicles in the same category but across different model years. The main objective of this research was to develop an integrated tool able to assess vehicle's performance simultaneously for safety and environmental domains, leading to the research output of a Safety, Fuel Efficiency and Green Emissions (SEG) indicator able to evaluate and rank vehicle's performance across those three domains. For this purpose, crash data was gathered in Porto (Portugal) for the period 2006-2010 (N=1374). The crash database was analyzed and crash severity prediction models were developed using advanced logistic regression models. Following, the methodology for the SEG indicator was established combining the vehicle's safety and the environmental evaluation into an integrated analysis. The obtained results for the SEG indicator do not show any trade-off between vehicle's safety, fuel consumption and emissions. The best performance was achieved for newer gasoline passenger vehicles (<5year) with a smaller engine size (<1400cm(3)). According to the SEG indicator, a vehicle with these characteristics can be recommended for a safety-conscious profile user, as well as for a user more interested in fuel economy and/or in green performance. On the other hand, for larger engine size vehicles (>2000cm(3)) the combined score for safety user profile was in average more satisfactory than for vehicles in the smaller engine size group (<1400cm(3)), which suggests that in general, larger vehicles may offer extra protection. The achieved results demonstrate that the developed SEG integrated methodology can be a helpful tool for

  11. Optimization of a Dry, Mixed Nuclear Fuel Storage Array for Nuclear Criticality Safety

    NASA Astrophysics Data System (ADS)

    Baranko, Benjamin T.

    A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different research and test reactor fuel types in up to 636 fuel storage canisters. New analysis demonstrates that the current arrangement of the different fuel-type canisters does not minimize the system neutron multiplication factor (keff), and that the entire facility storage capacity cannot be utilized without exceeding the subcritical limit (ksafe) for ensuring nuclear criticality safety. This work determines a more optimal arrangement of the stored fuels with a goal to minimize the system keff, but with a minimum of potential fuel canister relocation movements. The solution to this multiple-objective optimization problem will allow for both an improvement in the facility utilization while also offering an enhancement in the safety margin. The solution method applies stochastic approximation and a Tabu search metaheuristic to an empirical model developed from supporting MCNP calculations. The results establish an optimal relocation of between four to sixty canisters, which will allow the current thirty-one empty canisters to be used for storage while reducing the array keff by up to 0.018 +/- 0.003 relative to the current arrangement.

  12. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    SciTech Connect

    Wagner, J.C.; Parks, C.V.

    2000-09-01

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of

  13. Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

    SciTech Connect

    Not Available

    1980-10-01

    The Assistant Secretary for Environment has responsibility for identifying, characterizing, and ameliorating the environmental, health, and safety issues and public concerns associated with commercial operation of specific energy systems. The need for developing a safety and environmental control assessment for liquefied gaseous fuels was identified by the Environmental and Safety Engineering Division as a result of discussions with various governmental, industry, and academic persons having expertise with respect to the particular materials involved: liquefied natural gas, liquefied petroleum gas, hydrogen, and anhydrous ammonia. This document is arranged in three volumes and reports on progress in the Liquefied Gaseous Fuels (LGF) Safety and Environmental Control Assessment Program made in Fiscal Year (FY)-1979 and early FY-1980. Volume 1 (Executive Summary) describes the background, purpose and organization of the LGF Program and contains summaries of the 25 reports presented in Volumes 2 and 3. Annotated bibliographies on Liquefied Natural Gas (LNG) Safety and Environmental Control Research and on Fire Safety and Hazards of Liquefied Petroleum Gas (LPG) are included in Volume 1.

  14. 77 FR 70193 - Shaw Areva MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-23

    ... COMMISSION Shaw Areva MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and Licensing Board Reconstitution Pursuant to 10 CFR 2.313(c) and 2.321(b), the Atomic Safety and Licensing... Administrative Judge, Atomic Safety and Licensing Board Panel. BILLING CODE 7590-01-P...

  15. Spent nuclear fuel project cold vacuum drying facility safety equipment list

    SciTech Connect

    IRWIN, J.J.

    1999-02-24

    This document provides the safety equipment list (SEL) for the Cold Vacuum Drying Facility (CVDF). The SEL was prepared in accordance with the procedure for safety structures, systems, and components (SSCs) in HNF-PRO-516, ''Safety Structures, Systems, and Components,'' Revision 0 and HNF-PRO-097, Engineering Design and Evaluation, Revision 0. The SEL was developed in conjunction with HNF-SO-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998). The SEL identifies the SSCs and their safety functions, the design basis accidents for which they are required to perform, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. This SEL has been developed for the CVDF Phase 2 Safety Analysis Report (SAR) and shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR until the CVDF final SAR is approved.

  16. Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

    SciTech Connect

    Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

    1980-05-01

    The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

  17. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    SciTech Connect

    Richard, R.F.

    1995-05-11

    It has been postulated that a degradation phenomenon, referred to as ``hot cell rot``, may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ``Hot cell rot`` refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ``hot cell rot`` phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical.

  18. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    SciTech Connect

    DeHart, M.D.

    1999-08-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  19. Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies

    SciTech Connect

    Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

    1980-05-01

    Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

  20. Development of Criteria for Flameholding Tendencies within Premixer Passages for High Hydrogen Content Fuels

    SciTech Connect

    Lewis, Elliot; McDonell, Vincent

    2015-03-31

    Due to increasingly stringent air quality requirements stationary power gas turbines have moved to lean-premixed operation, which reduces pollutant emissions but can result in flashback. Flashback can cause serious damage to the premixer hardware. Curtailing flashback can be difficult with hydrocarbon fuels and becomes even more challenging when hydrogen is used as the fuel. The two main approaches for coping with flashback are either to design a combustor that is resistant to flashback, or to design a premixer that will not anchor a flame if flashback occurs. Even with a well-designed combustor flashback can occur under certain circumstances, thus it is necessary to determine how to avoid flameholding within the premixer passageways of a gas turbine. To this end, an experiment was designed that would determine the flameholding propensities at elevated pressures and temperatures of three different classes of geometric features commonly found in gas turbine premixers, with both natural gas and hydrogen fuel. Experiments to find the equivalence ratio at blow off were conducted within an optically accessible test apparatus with four flameholders: 0.25 and 0.50 inch diameter cylinders, a reverse facing step with a height of 0.25 inches, and a symmetric airfoil with a thickness of 0.25 inches and a chord length of one inch. Tests were carried out at temperatures between 300 K and 750 K, at pressures up to 9 atmospheres. Typical bulk velocities were between 40 and 100 m/s. The effect of airfoil’s angle of rotation was also investigated. Blow off for hydrogen flames was found to occur at much lower adiabatic flame temperatures than natural gas flames. Additionally it was observed that at high pressures and high turbulence intensities, reactant velocity does not have a noticeable effect on the point of blow off due in large part to corresponding increases in turbulent flame speed. Finally a semi empirical correlation was developed that predicts flame extinction for both

  1. Failure Criteria for Evaluating Accidental Drops of Fuel Containers at INTEC

    SciTech Connect

    Miller, G. K.

    1998-10-01

    This report presents a failure criterion that has been developed for use in evaluating fuel containers at the Idaho Nuclear Technology and Engineering Center (INTEC) for accidental drop events. The criterion would typically be used in dynamic finite element analyses using the ABA-QUS/Explicit program. The failure criterion used in the past is generally considered to substantially underestimate the strength and ductility of the materials involved. The new criterion is intended to be more realistic, allowing for more accurate impact analyses. The criterion is based on the distortion energy theory, which is considered to be appropriate for the ductile materials typically used in fuel containers. Also addressed in development of the criterion were the effects of strain rate and hydrostatic stress. The importance of these factors, however, is highly dependent on the material used. Three materials specifically addressed in this study were stainless steel, aluminum, and lead. The criterion is presented in the form of guidelines and recommendations that are based on material data obtained from the literature. The most significant difference between these and the previous criterion is that ductile materials are allowed to strain to much higher levels before they are considered to fail.

  2. Environmental, health, and safety issues of fuel cells in transportation. Volume 1: Phosphoric acid fuel-cell buses

    SciTech Connect

    Ring, S.

    1994-12-01

    The U.S. Department of Energy (DOE) chartered the Phosphoric Acid Fuel-Cell (PAFC) Bus Program to demonstrate the feasibility of fuel cells in heavy-duty transportation systems. As part of this program, PAFC- powered buses are being built to meet transit industry design and performance standards. Test-bed bus-1 (TBB-1) was designed in 1993 and integrated in March 1994. TBB-2 and TBB-3 are under construction and should be integrated in early 1995. In 1987 Phase I of the program began with the development and testing of two conceptual system designs- liquid- and air-cooled systems. The liquid-cooled PAFC system was chosen to continue, through a competitive award, into Phase H, beginning in 1991. Three hybrid buses, which combine fuel-cell and battery technologies, were designed during Phase III. After completing Phase II, DOE plans a comprehensive performance testing program (Phase HI) to verify that the buses meet stringent transit industry requirements. The Phase III study will evaluate the PAFC bus and compare it to a conventional diesel bus. This NREL study assesses the environmental, health, and safety (EH&S) issues that may affect the commercialization of the PAFC bus. Because safety is a critical factor for consumer acceptance of new transportation-based technologies the study focuses on these issues. The study examines health and safety together because they are integrally related. In addition, this report briefly discusses two environmental issues that are of concern to the Environmental Protection Agency (EPA). The first issue involves a surge battery used by the PAFC bus that contains hazardous constituents. The second issue concerns the regulated air emissions produced during operation of the PAFC bus.

  3. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    SciTech Connect

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  4. Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

    SciTech Connect

    Not Available

    1991-10-18

    The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173.

  5. Quality and safety of medication use in primary care: consensus validation of a new set of explicit medication assessment criteria and prioritisation of topics for improvement

    PubMed Central

    2012-01-01

    Background Addressing the problem of preventable drug related morbidity (PDRM) in primary care is a challenge for health care systems internationally. The increasing implementation of clinical information systems in the UK and internationally provide new opportunities to systematically identify patients at risk of PDRM for targeted medication review. The objectives of this study were (1) to develop a set of explicit medication assessment criteria to identify patients with sub-optimally effective or high-risk medication use from electronic medical records and (2) to identify medication use topics that are perceived by UK primary care clinicians to be priorities for quality and safety improvement initiatives. Methods For objective (1), a 2-round consensus process based on the RAND/UCLA Appropriateness Method (RAM) was conducted, in which candidate criteria were identified from the literature and scored by a panel of 10 experts for 'appropriateness' and 'necessity'. A set of final criteria was generated from candidates accepted at each level. For objective (2), thematically related final criteria were clustered into 'topics', from which a panel of 26 UK primary care clinicians identified priorities for quality improvement in a 2-round Delphi exercise. Results (1) The RAM process yielded a final set of 176 medication assessment criteria organised under the domains 'quality' and 'safety', each classified as targeting 'appropriate/necessary to do' (quality) or 'inappropriate/necessary to avoid' (safety) medication use. Fifty-two final 'quality' assessment criteria target patients with unmet indications, sub-optimal selection or intensity of beneficial drug treatments. A total of 124 'safety' assessment criteria target patients with unmet needs for risk-mitigating agents, high-risk drug selection, excessive dose or duration, inconsistent monitoring or dosing instructions. (2) The UK Delphi panel identified 11 (23%) of 47 scored topics as 'high priority' for quality

  6. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance. [LMFBR

    SciTech Connect

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment.

  7. Optimum design of a fuel-cell powertrain based on multiple design criteria

    NASA Astrophysics Data System (ADS)

    Sarioglu, Ismail Levent; Czapnik, Bartosch; Bostanci, Emine; Klein, Olaf P.; Schröder, Hendrik; Küçükay, Ferit

    2014-11-01

    As the number of fuel-cell vehicles on the roads increase, the vehicle designs are gaining more importance. Clearly, one major topic in this field is the optimization of powertrain designs. In this design process, the aim of the car manufacturers is to meet the expectations of the potential customer best, while creating a sustainable product. However, due to several trade-offs in the design, it would be non-realistic to expect a single solution that fulfills all design objectives. Therefore, a systematical approach, which includes a trade-off analysis and evaluation methods for this multiobjective design problem, is required. In this paper, a suitable methodology is presented and applied in a case study, where an optimum powertrain design for a typical European long-range passenger car is sought. Simulation-aided powertrain models and scalable component models are used to increase the accuracy of the design process. Furthermore, various visual and quantitative evaluation techniques are applied in order to support the decision making process.

  8. High Energy Density Additives for Hybrid Fuel Rockets to Improve Performance and Enhance Safety

    NASA Technical Reports Server (NTRS)

    Jaffe, Richard L.

    2014-01-01

    We propose a conceptual study of prototype strained hydrocarbon molecules as high energy density additives for hybrid rocket fuels to boost the performance of these rockets without compromising safety and reliability. Use of these additives could extend the range of applications for which hybrid rockets become an attractive alternative to conventional solid or liquid fuel rockets. The objectives of the study were to confirm and quantify the high enthalpy of these strained molecules and to assess improvement in rocket performance that would be expected if these additives were blended with conventional fuels. We confirmed the chemical properties (including enthalpy) of these additives. However, the predicted improvement in rocket performance was too small to make this a useful strategy for boosting hybrid rocket performance.

  9. Fabrication of UO/sub 2/-stainless steel fuel for a fast-reactor safety test

    SciTech Connect

    Rhude, H.V.; O'Keefe, G.B.; Noland, R.A.

    1983-01-01

    As part of its fast breeder reactor (LMFBR) safety research program, Argonne National Laboratory is conducting a series of tests in the TREAT Reactor to clarify the boilup and freezing behavior of UO/sub 2/ fuel and stainless steel in the transition phase of an accident sequence. Although much work has been done on UO/sub 2/ dispersed in a stainless steel matrix, no references could be found on work relating to the dispersal of stainless steel in a UO/sub 2/ matrix. Consequently a development program was started to produce the necessary fuel. The major difficulty encountered is attributable to the melting range of 316 stainless steel (1391 to 1399/sup 0/C) which is well below the normal sintering temperature of UO/sub 2/ (1750/sup 0/C). Attempts to produce the required fuel pellets by standard sintering techniques did not work. Alternative approaches to sintering were investigated to determine the density achievable without sintering.

  10. Advanced Fuel Cycles for Fusion Reactors: Passive Safety and Zero-Waste Options

    NASA Astrophysics Data System (ADS)

    Zucchetti, Massimo; Sugiyama, Linda E.

    2006-05-01

    Nuclear fusion is seen as a much ''cleaner'' energy source than fission. Most of the studies and experiments on nuclear fusion are currently devoted to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition. The recent stress on safety by the world's community has stimulated the research on other fuel cycles than the DT one, based on 'advanced' reactions, such as the Deuterium-Helium-3 (DHe) one. These reactions pose problems, such as the availability of 3He and the attainment of the higher plasma parameters that are required for burning. However, they have many advantages, like for instance the very low neutron activation, while it is unnecessary to breed and fuel tritium. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) has been studied. Results show that Candor does reach the passive safety and zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate response to the environmental requirements for future nuclear power plants.

  11. Environmental safety aspects of the new spent nuclear fuel management and storage system at Ignalina NPP

    SciTech Connect

    Poskas, P.; Ragaisis, V.; Adomaitis, J. E.

    2007-07-01

    In the framework of the preparation for the decommissioning of the Ignalina Nuclear Power Plant (INPP) a new Interim Spent Nuclear Fuel Storage Facility (ISFSF) will be built in the existing sanitary protection zone (SPZ) of INPP. In addition to the ISFSF, the new spent nuclear fuel management activity will include all necessary spent nuclear fuel retrieval and packaging operations at the Reactor Units, transfer of storage casks to the ISFSF, and other activities appropriate to the chosen design solution and required for the safe removal of the existing spent nuclear fuel from storage pools and insertion into the new ISFSF. The Republic of Lithuania regulations require that the average annual dose to the critical group members of population due to operation of nuclear facility shall not exceed dose constraint. If several nuclear facilities are located in the same SPZ, the same dose constraint shall envelope radiological impacts from all operating and planned nuclear facilities. The paper discusses radiological safety assessment aspects as relevant for the new nuclear activity to be implemented in the SPZ of INPP considering specificity of Lithuanian regulatory requirements. The safety assessment methodology aspects, results and conclusions as concern public exposure are outlined and discussed. (authors)

  12. Liquefied gaseous fuels safety and environmental control assessment program: third status report

    SciTech Connect

    Not Available

    1982-03-01

    This Status Report contains contributions from all contractors currently participating in the DOE Liquefied Gaseous Fuels (LG) Safety and Environmental Control Assessment Program and is presented in two principal sections. Section I is an Executive Summary of work done by all program participants. Section II is a presentation of fourteen individual reports (A through N) on specific LGF Program activities. The emphasis of Section II is on research conducted by Lawrence Livermore National Laboratory (Reports A through M). Report N, an annotated bibliography of literature related to LNG safety and environmental control, was prepared by Pacific Northwest Laboratory (PNL) as part of its LGF Safety Studies Project. Other organizations who contributed to this Status Report are Aerojet Energy Conversion Company; Applied Technology Corporation; Arthur D. Little, Incorporated; C/sub v/ International, Incorporated; Institute of Gas Technology; and Massachusetts Institute of Technology. Separate abstracts have been prepared for Reports A through N for inclusion in the Energy Data Base.

  13. Design Criteria for Future Fuels and Related Power Systems Addressing the Impacts of Non-CO2 Pollutants on Human Health and Climate Change.

    PubMed

    Schauer, James Jay

    2015-01-01

    Concerns over the economics, supply chain, and emissions of greenhouse gases associated with the wide use of fossil fuels have led to increasing interest in developing alternative and renewable fuels for stationary power generation and transportation systems. Although there is considerable uncertainty regarding the economic and environmental impacts of alternative and renewable fuels, there is a great need for assessment of potential and emerging fuels to guide research priorities and infrastructure investment. Likewise, there is a great need to identify potential unintended adverse impacts of new fuels and related power systems before they are widely adopted. Historically, the environmental impacts of emerging fuels and power systems have largely focused on carbon dioxide emissions, often called the carbon footprint, which is used to assess impacts on climate change. Such assessments largely ignore the large impacts of emissions of other air pollutants. Given the potential changes in emissions of air pollutants associated with the large-scale use of new and emerging fuels and power systems, there is a great need to better guide efforts to develop new fuels and power systems that can avoid unexpected adverse impacts on the environment and human health. This review covers the nature of emissions, including the key components and impacts from the use of fuels, and the design criteria for future fuels and associated power systems to assure that the non-CO2 adverse impacts of stationary power generation and transportation are minimized. PMID:26134739

  14. Use of Burnup Credit as a Safety Factor in Handling of NIST Fuel Assemblies in the L Basin of SRS

    SciTech Connect

    Eghbali, DA

    2004-01-07

    Burnup credit was recently used for the first time in criticality safety analysis to support the handling of the National Institute of Standards and Technology spent fuel assemblies in the L Basin of Savannah River Site. Previous criticality safety analyses were based on the fissile content of fresh, unirradiated fuel assemblies, resulting in handling of a group of 10 or less fuel assemblies at a time. Using burnup credit, it was demonstrated that an isolated configuration of up to 14 NITS fuel assemblies, the maximum number of fuel assemblies in a full basket, submerged in a concrete-lined, water-filled pool is subcritical, resulting in several administrative controls being modified or eliminated without compromising safety.

  15. [Experience of justification of hygienic standards of food safety with the use of criteria for the risk population health].

    PubMed

    Zaytseva, N V; Tutelyan, V A; Shur, P Z; Khotimchenko, S A; Sheveleva, S A

    2014-01-01

    In the article there is presented the experience of justification of hygienic standards of food safety with the use of criteria for the risk for population health. Health risk assessment under the impact of tetracyclines with food showed that the content of residual amounts of these antibiotics at the level of 10 mg/kg (permissible residual tetracycline accepted in Customs Union Member Countries (CUMC) will not increase the risk to public health, including the most sensitive groups of the population. The assessment ofthe health risk associated with the receipt of ractopamine with food, showed that eating foods containing ractopamine at ADI level (0-1 mg/kg body weight), and even at the limit of quantification levels in meat products, is inadmissible because of unacceptable risk of functional disorders and diseases of the cardiovascular system. The results of the substantiation of the permissible levels of nitrates content in crop production showed that at the level of exposure according to hygienic standards established in the CUMC as at the recommended and actual consumption levels of products ofplant origin, the health risk as carcinogenic and non-carcinogenic, does not exceed acceptable levels. The results of the assessment of the risk associated with the permissible levels of L. monocytogenes in certain food groups showed that an exposure level of hygienic standards established in the CUMC, standards of Codex Alimentarius Commission and EU documents (before release to the market by the manufacturer) the health risk does not exceed the maximum permissible level of the appearance of serious diseases. Adoption of standards of Codex Alimentarius Commission and the EU (for handling products in the market) is not acceptable because it can lead to an unacceptable risk of listeriosis for the population of the Russian Federation as a whole, and for the most sensitive groups. PMID:25831934

  16. Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

    SciTech Connect

    Turner, John A; Clarno, Kevin T; Hansen, Glen A

    2009-09-01

    -development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach.

  17. Unreviewed safety question evaluation of 100K East and 100K West in-basin fuel characterization program activities

    SciTech Connect

    Alwardt, L.D.

    1995-01-12

    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.

  18. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  19. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

  20. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    SciTech Connect

    Klein, Andrew; Matthews, Topher; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  1. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  2. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  3. Fuel efficiency and automobile safety: Single-vehicle highway fatalities for passenger cars

    SciTech Connect

    Khazzoom, J.D.

    1994-12-31

    This paper reports the results of an effort to shed some light on the relationship that might exist between enhanced standards and single-vehicle passenger car highway fatalities. Quantification of this relationship is not an easy task Not surprisingly, the literature on modeling the relationship between fuel economy and highway fatalities is very scant. Our analytic framework consists of two submodels: a corporate average fuel economy (CAFE) submodel and a single-vehicle highway fatalities submodel. Some of the variables that enter the CAFE relationship affect single-vehicle fatalities, as well. The results of this study are not unequivocal in every respect. However, they indicate that enhanced standards and automobile safety need not be at odds with each other. A main message that emerges from this study is the need not to confuse car downsizing with down weighting. Quantatative studies of highway fatalities have mostly treated weight and size interchangeably, and have used only the weight variable in the fatalities equation to avoid dealing with multicollinearity. Such references as {open_quote}size/weight{close_quote} which lump size and weight together as if they were the same variable are not uncommon in the safety literature. Our study indicates that weight and size are not a proxy to each other, and that in single vehicle crashes they are likely to have opposite effects on safety. Men researchers choose to drop the size variable and include only the weight variable in the fatalities equation, the weight estimate may end up with a negative sign, not necessarily because weight has a beneficial effect on safety, but because the omitted size variable has a dominant beneficial effect on safety, which is picked up by the weight variable that appears in the equation. 65 refs., 7 tabs.

  4. Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

    SciTech Connect

    A.E. Craft; R. C. O'Brien; S. D. Howe; J. C. King

    2014-07-01

    Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact, fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.

  5. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    SciTech Connect

    Chiang, R. T.

    2013-07-01

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  6. First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace

    SciTech Connect

    Demkowicz, Paul A.; Reber, Edward L.; Scates, Dawn M.; Scott, Les; Collin, Blaise P.

    2015-09-01

    Three TRISO fuel compacts from the AGR-1 irradiation experiment were subjected to safety tests at 1600 and 1800 °C for approximately 300 h to evaluate the fission product retention characteristics. Silver behavior was dominated by rapid release of an appreciable fraction of the compact inventory (3–34%) at the beginning of the tests, believed to be from inventory residing in the compact matrix and outer pyrocarbon (OPyC) prior to the safety test. Measurable release of silver from intact particles appears to become apparent only after ~60 h at 1800 °C. The release rate for europium and strontium was nearly constant for 300 h at 1600 °C (reaching maximum values of approximately 2×10⁻³ and 8×10⁻⁴ respectively), and at this temperature the release may be mostly limited to inventory in the compact matrix and OPyC prior to the safety test. The release rate for both elements increased after approximately 120 h at 1800 °C, possibly indicating additional measurable release through the intact particle coatings. Cesium fractional release from particles with intact coatings was <10⁻⁶ after 300 h at 1600 °C or 100 h at 1800 °C, but release from the rare particles that experienced SiC failure during the test could be significant. However, Kr release was still very low for 300 h 1600 °C (<2 × 10⁻⁶). At 1800 °C, krypton release increased noticeably after SiC failure, reflecting transport through the intact outer pyrocarbon layer. Nonetheless, the krypton and cesium release fractions remained less than approximately 10⁻³ after 277 h at 1800 °C.

  7. First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace

    NASA Astrophysics Data System (ADS)

    Demkowicz, Paul A.; Reber, Edward L.; Scates, Dawn M.; Scott, Les; Collin, Blaise P.

    2015-09-01

    Three TRISO fuel compacts from the AGR-1 irradiation experiment were subjected to safety tests at 1600 and 1800 °C for approximately 300 h to evaluate the fission product retention characteristics. Silver behavior was dominated by rapid release of an appreciable fraction of the compact inventory (3-34%) at the beginning of the tests, believed to be from inventory residing in the compact matrix and outer pyrocarbon (OPyC) prior to the safety test. Measurable release of silver from intact particles appears to become apparent only after ∼60 h at 1800 °C. The release rate for europium and strontium was nearly constant for 300 h at 1600 °C (reaching maximum values of approximately 2 × 10-3 and 8 × 10-4 respectively), and at this temperature the release may be mostly limited to inventory in the compact matrix and OPyC prior to the safety test. The release rate for both elements increased after approximately 120 h at 1800 °C, possibly indicating additional measurable release through the intact particle coatings. Cesium fractional release from particles with intact coatings was <10-6 after 300 h at 1600 °C or 100 h at 1800 °C, but release from the rare particles that experienced SiC failure during the test could be significant. However, Kr release was still very low for 300 h 1600 °C (<2 × 10-6). At 1800 °C, krypton release increased noticeably after SiC failure, reflecting transport through the intact outer pyrocarbon layer. Nonetheless, the krypton and cesium release fractions remained less than approximately 10-3 after 277 h at 1800 °C.

  8. General-purpose heat source project and space nuclear safety fuels program. Progress report, February 1980

    SciTech Connect

    Maraman, W.J.

    1980-05-01

    This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space Nuclear Safety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work.

  9. Probabilistic Modeling Approach for Evaluating the Compliance of Ready-To-Eat Foods with New European Union Safety Criteria for Listeria monocytogenes▿

    PubMed Central

    Koutsoumanis, Konstantinos; Angelidis, Apostolos S.

    2007-01-01

    Among the new microbiological criteria that have been incorporated in EU Regulation 2073/2005, of particular interest are those concerning Listeria monocytogenes in ready-to eat (RTE) foods, because for certain food categories, they no longer require zero tolerance but rather specify a maximum allowable concentration of 100 CFU/g or ml. This study presents a probabilistic modeling approach for evaluating the compliance of RTE sliced meat products with the new safety criteria for L. monocytogenes. The approach was based on the combined use of (i) growth/no growth boundary models, (ii) kinetic growth models, (iii) product characteristics data (pH, aw, shelf life) collected from 160 meat products from the Hellenic retail market, and (iv) storage temperature data recorded from 50 retail stores in Greece. This study shows that probabilistic analysis of the above components using Monte Carlo simulation, which takes into account the variability of factors affecting microbial growth, can lead to a realistic estimation of the behavior of L. monocytogenes throughout the food supply chain, and the quantitative output generated can be further used by food managers as a decision-making tool regarding the design or modification of a product's formulation or its “use-by” date in order to ensure its compliance with the new safety criteria. The study also argues that compliance of RTE foods with the new safety criteria should not be considered a parameter with a discrete and binary outcome because it depends on factors such as product characteristics, storage temperature, and initial contamination level, which display considerable variability even among different packages of the same RTE product. Rather, compliance should be expressed and therefore regulated in a more probabilistic fashion. PMID:17557858

  10. Probabilistic modeling approach for evaluating the compliance of ready-to-eat foods with new European Union safety criteria for Listeria monocytogenes.

    PubMed

    Koutsoumanis, Konstantinos; Angelidis, Apostolos S

    2007-08-01

    Among the new microbiological criteria that have been incorporated in EU Regulation 2073/2005, of particular interest are those concerning Listeria monocytogenes in ready-to eat (RTE) foods, because for certain food categories, they no longer require zero tolerance but rather specify a maximum allowable concentration of 100 CFU/g or ml. This study presents a probabilistic modeling approach for evaluating the compliance of RTE sliced meat products with the new safety criteria for L. monocytogenes. The approach was based on the combined use of (i) growth/no growth boundary models, (ii) kinetic growth models, (iii) product characteristics data (pH, a(w), shelf life) collected from 160 meat products from the Hellenic retail market, and (iv) storage temperature data recorded from 50 retail stores in Greece. This study shows that probabilistic analysis of the above components using Monte Carlo simulation, which takes into account the variability of factors affecting microbial growth, can lead to a realistic estimation of the behavior of L. monocytogenes throughout the food supply chain, and the quantitative output generated can be further used by food managers as a decision-making tool regarding the design or modification of a product's formulation or its "use-by" date in order to ensure its compliance with the new safety criteria. The study also argues that compliance of RTE foods with the new safety criteria should not be considered a parameter with a discrete and binary outcome because it depends on factors such as product characteristics, storage temperature, and initial contamination level, which display considerable variability even among different packages of the same RTE product. Rather, compliance should be expressed and therefore regulated in a more probabilistic fashion. PMID:17557858

  11. International Workshop on Characterization and PIE Needs for Fundamental Understanding of Fuels Performance and Safety

    SciTech Connect

    Not Listed

    2011-12-01

    The International Workshop on Characterization and PIE Needs to Support Science-Based Development of Innovative Fuels was held June 16-17, 2011, in Paris, France. The Organization for Economic Co-operation and Development (OECD), Nuclear Energy Agency (NEA) Working Party on the Fuel Cycle (WPFC) sponsored the workshop to identify gaps in global capabilities that need to be filled to meet projected needs in the 21st century. First and foremost, the workshop brought nine countries and associated international organizations, together in support of common needs for nuclear fuels and materials testing, characterization, PIE, and modeling capabilities. Finland, France, Germany, Republic of Korea, Russian Federation, Sweden, Switzerland, United Kingdom, United States of America, IAEA, and ITU (on behalf of European Union Joint Research Centers) discussed issues and opportunities for future technical advancements and collaborations. Second, the presentations provided a base level of understanding of current international capabilities. Three main categories were covered: (1) status of facilities and near term plans, (2) PIE needs from fuels engineering and material science perspectives, and (3) novel PIE techniques being developed to meet the needs. The International presentations provided valuable data consistent with the outcome of the National Workshop held in March 2011. Finally, the panel discussion on 21st century PIE capabilities, created a unified approach for future collaborations. In conclusion, (1) existing capabilities are not sufficient to meet the needs of a science-based approach, (2) safety issues and fuels behavior during abnormal conditions will receive more focus post-Fukushima; therefore we need to adopt our techniques to those issues, and (3) International collaboration is needed in the areas of codes and standards development for the new techniques.

  12. Study for the optimization of a transport aircraft wing for maximum fuel efficiency. Volume 1: Methodology, criteria, aeroelastic model definition and results

    NASA Technical Reports Server (NTRS)

    Radovcich, N. A.; Dreim, D.; Okeefe, D. A.; Linner, L.; Pathak, S. K.; Reaser, J. S.; Richardson, D.; Sweers, J.; Conner, F.

    1985-01-01

    Work performed in the design of a transport aircraft wing for maximum fuel efficiency is documented with emphasis on design criteria, design methodology, and three design configurations. The design database includes complete finite element model description, sizing data, geometry data, loads data, and inertial data. A design process which satisfies the economics and practical aspects of a real design is illustrated. The cooperative study relationship between the contractor and NASA during the course of the contract is also discussed.

  13. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    SciTech Connect

    Maraman, W.J.

    1980-02-01

    Studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of LASL are presented. The three programs involved are: general-purpose heat source development; space nuclear safety; and fuels program. Three impact tests were conducted to evaluate the effects of a high temperature reentry pulse and the use of CBCF on impact performance. Additionally, two /sup 238/PuO/sub 2/ pellets were encapsulated in Ir-0.3% W for impact testing. Results of the clad development test and vent testing are noted. Results of the environmental tests are summarized. Progress on the Stirling isotope power systems test and the status of the improved MHW tests are indicated. The examination of the impact failure of the iridium shell of MHFT-65 at a fuel pass-through continued. A test plan was written for vibration testing of the assembled light-weight radioisotopic heater unit. Progress on fuel processing is reported.

  14. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... of 10 CFR 52.157 to include in its final safety analysis report a description of the quality... construction permit is required by the provisions of § 50.34 to include in its preliminary safety analysis... operating license is required to include, in its final safety analysis report, information pertaining to...

  15. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  16. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    SciTech Connect

    Miko, David K.; Desimone, David J.

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  17. A FRAMEWORK TO DEVELOP FLAW ACCEPTANCE CRITERIA FOR STRUCTURAL INTEGRITY ASSESSMENT OF MULTIPURPOSE CANISTERS FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

    SciTech Connect

    Lam, P.; Sindelar, R.; Duncan, A.; Adams, T.

    2014-04-07

    A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

  18. Safety of interim storage solutions of used nuclear fuel during extended term

    SciTech Connect

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M.

    2013-07-01

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

  19. Health and safety impacts related to the management of spent nuclear fuels

    SciTech Connect

    Jilek, D.C.

    1996-06-01

    Under the Nuclear Waste Policy Act of 1982, as amended, the U.S. Department of Energy is responsible for managing the disposal of spent nuclear fuel from civilian nuclear power plants. Deployment of a multipurpose canister (MPC) system for dry storage of commercial spent nuclear fuel at reactor sites was determined to be an option for managing spent nuclear fuel until either a permanent repository or interim central storage facility (commonly called a Monitored Retrievable Storage Facility, or MRS) becomes available. Routine health and safety impacts to workers from handling and storage operations at nuclear facilities for four separate scenarios were evaluated for the MPC system: an on-time repository with an MRS; an on-time repository with no MRS; a delayed repository with an MRS; and a delayed repository with no MRS. In addition to evaluating the MPC system, five alternatives were analyzed. These included the No Action Alternative (NAA), Current Technology (CTr), the Transposable Storage Cask (TSC), the Dual-Purpose Canister (DPC), and the Small MPC (SmMPC). Health effects are expressed as collective doses in person- rem per year and risks as latent cancer fatalities per year for incident-free operations for each alternative and scenario. Results show that both dose and risks to workers vary as much as 68{percent} among scenarios and alternatives. Although dose estimates and risks fall below limits for radiation dose to workers as specified in Title 10, Part 20, of the Code of Federal Regulations, additional measures could be applied to reduce potential doses and resultant health risk. 5 refs., 2 tabs.

  20. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    SciTech Connect

    William J. O’Donnell; Donald S. Griffin

    2007-05-07

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  1. Comparison of safety parameters and transient behavior of a generic 10 MW reactor with HEU and LEU fuels

    SciTech Connect

    Matos, J.E.; Freese, K.E.; Woodruff, W.L.

    1983-01-01

    Key safety parameters are compared for equilibrium cores of the IAEA generic 10 MW reactor with HEU and LEU fuels. These parameters include kinetics parameters, reactivity feedback coefficients, control rod worths, power peaking factors, and shutdown margins. Reactivity insertion and loss-of-flow transients are compared. Results indicate that HEU and LEU cores will behave in a very similar manner.

  2. Landscape modeling for dose calculations in the safety assessment of a repository for spent nuclear fuel

    SciTech Connect

    Lindborg, Tobias; Kautsky, Ulrik; Brydsten, Lars

    2007-07-01

    The Swedish Nuclear Fuel and Waste Management Co.,(SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site. Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the

  3. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    NASA Astrophysics Data System (ADS)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  4. Safety assessment of plutonium mixed oxide fuel irradiated up to 37.7 GW day tonne-1

    NASA Astrophysics Data System (ADS)

    Somers, J.; Papaioannou, D.; McGinley, J.; Sommer, D.

    2013-06-01

    In this irradiation test, the safety performance of (Th,Pu)O2 fuel was evaluated. The fuel pellets were synthesised from powders prepared using a sol gel method to give a product exhibiting an atomically homogeneous distribution of the elements. The fuel pellets, of conventional pressurised water reactor (PWR) dimensions, were encapsulated in zircaloy cladding, and irradiated during four reactor cycles, reaching a burnup of 37.7 GW day tonne-1 in the KWO pressurised water reactor at Obrigheim, Germany. The irradiation test was performed under representative conditions. Intermediate inspection of the fuel pin during reactor outages revealed a cladding creep down within the bounds observed for UO2 fuels under similar conditions. Hydriding of the cladding was found predominantly on the outer liner of the duplex cladding. Fission gas analysis revealed a release of about 0.5%, which is somewhat lower than U-MOX fuels at the same burnup, but the latter were operated at higher linear heating rate. The Xe/Kr ratio of 11 is much lower than (U,Pu)O2 fuel (typically 16), indicating significant 233U generation and fissioning thereof during the irradiation experiment. Examination of the microstructure indicates that the pellet - cladding gap is almost closed. The grain size remained similar to the fresh fuel (4 μm) and no intragranular porosity was observed.

  5. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... approval or design certification under part 52 of this chapter is required by the provisions of 10 CFR 52... of 10 CFR 52.157 to include in its final safety analysis report a description of the...

  6. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... approval or design certification under part 52 of this chapter is required by the provisions of 10 CFR 52... of 10 CFR 52.157 to include in its final safety analysis report a description of the...

  7. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... approval or design certification under part 52 of this chapter is required by the provisions of 10 CFR 52... of 10 CFR 52.157 to include in its final safety analysis report a description of the...

  8. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  9. Performance Criteria and Evaluation System

    Energy Science and Technology Software Center (ESTSC)

    1992-06-18

    The Performance Criteria and Evaluation System (PCES) was developed in order to make a data base of criteria accessible to radiation safety staff. The criteria included in the package are applicable to occupational radiation safety at DOE reactor and nonreactor nuclear facilities, but any data base of criteria may be created using the Criterion Data Base Utiliity (CDU). PCES assists personnel in carrying out oversight, line, and support activities.

  10. Evaluating the road safety effects of a fuel cost increase measure by means of zonal crash prediction modeling.

    PubMed

    Pirdavani, Ali; Brijs, Tom; Bellemans, Tom; Kochan, Bruno; Wets, Geert

    2013-01-01

    Travel demand management (TDM) consists of a variety of policy measures that affect the transportation system's effectiveness by changing travel behavior. The primary objective to implement such TDM strategies is not to improve traffic safety, although their impact on traffic safety should not be neglected. The main purpose of this study is to evaluate the traffic safety impact of conducting a fuel-cost increase scenario (i.e. increasing the fuel price by 20%) in Flanders, Belgium. Since TDM strategies are usually conducted at an aggregate level, crash prediction models (CPMs) should also be developed at a geographically aggregated level. Therefore zonal crash prediction models (ZCPMs) are considered to present the association between observed crashes in each zone and a set of predictor variables. To this end, an activity-based transportation model framework is applied to produce exposure metrics which will be used in prediction models. This allows us to conduct a more detailed and reliable assessment while TDM strategies are inherently modeled in the activity-based models unlike traditional models in which the impact of TDM strategies are assumed. The crash data used in this study consist of fatal and injury crashes observed between 2004 and 2007. The network and socio-demographic variables are also collected from other sources. In this study, different ZCPMs are developed to predict the number of injury crashes (NOCs) (disaggregated by different severity levels and crash types) for both the null and the fuel-cost increase scenario. The results show a considerable traffic safety benefit of conducting the fuel-cost increase scenario apart from its impact on the reduction of the total vehicle kilometers traveled (VKT). A 20% increase in fuel price is predicted to reduce the annual VKT by 5.02 billion (11.57% of the total annual VKT in Flanders), which causes the total NOCs to decline by 2.83%. PMID:23200453

  11. Key Performance Criteria Affecting the Most the Safety of a Nuclear Waste Long Term Storage : A Case Study Commissioned by CEA

    SciTech Connect

    Marvy, A.; Lioure, A; Heriard-Dubreuil, G.; Gadbois, S.; Schneider, T.; Schieber, C.

    2003-02-24

    As part of the work scope set in the French law on high level long lived waste R&D passed in 1991, CEA is conducting a research program to establish the scientific basis and assess the feasibility of long term storage as an option for the safe management of nuclear waste for periods as long as centuries. This goal is a significant departure from the current industrial practice where storage facilities are usually built to last only a few decades. From a technical viewpoint such an extension in time seems feasible provided care and maintenance is exercised. Considering such long periods of time, the risk for Society of loosing oversight and control of such a facility is real, which triggers the question of whether and how long term storage safety can be actually achieved. Therefore CEA commissioned a study (1) in which MUTADIS Consultants (2) and CEPN (3) were both involved. The case study looks into several past and actual human enterprises conducted over significant periods o f time, one of them dating back to the end of the 18th century, and all identified out of the nuclear field. Then-prevailing societal behavior and organizational structures are screened out to show how they were or are still able to cope with similar oversight and control goals. As a result, the study group formulated a set of performance criteria relating to issues like responsibility, securing funds, legal and legislative implications, economic sustainable development, all being areas which are not traditionally considered as far as technical studies are concerned. These criteria can be most useful from the design stage onward, first in an attempt to define the facility construction and operating guiding principles, and thereafter to substantiate the safety case for long term storage and get geared to the public dialogue on that undertaking should it become a reality.

  12. Application of the MERIT survey in the multi-criteria quality assessment of occupational health and safety management

    PubMed Central

    Korban, Zygmunt

    2015-01-01

    Occupational health and safety management systems apply audit examinations as an integral element of these systems. The examinations are used to verify whether the undertaken actions are in compliance with the accepted regulations, whether they are implemented in a suitable way and whether they are effective. One of the earliest solutions of that type applied in the mining industry in Poland involved the application of audit research based on the MERIT survey (Management Evaluation Regarding Itemized Tendencies). A mathematical model applied in the survey facilitates the determination of assessment indexes WOPi for each of the assessed problem areas, which, among other things, can be used to set up problem area rankings and to determine an aggregate (synthetic) assessment. In the paper presented here, the assessment indexes WOPi were used to calculate a development measure, and the calculation process itself was supplemented with sensitivity analysis. PMID:26414772

  13. Application of the MERIT survey in the multi-criteria quality assessment of occupational health and safety management.

    PubMed

    Korban, Zygmunt

    2015-01-01

    Occupational health and safety management systems apply audit examinations as an integral element of these systems. The examinations are used to verify whether the undertaken actions are in compliance with the accepted regulations, whether they are implemented in a suitable way and whether they are effective. One of the earliest solutions of that type applied in the mining industry in Poland involved the application of audit research based on the MERIT survey (Management Evaluation Regarding Itemized Tendencies). A mathematical model applied in the survey facilitates the determination of assessment indexes WOPi for each of the assessed problem areas, which, among other things, can be used to set up problem area rankings and to determine an aggregate (synthetic) assessment. In the paper presented here, the assessment indexes WOPi were used to calculate a development measure, and the calculation process itself was supplemented with sensitivity analysis. PMID:26414772

  14. Analytical criteria for performance characteristics of IgE binding methods for evaluating safety of biotech food products.

    PubMed

    Holzhauser, Thomas; Ree, Ronald van; Poulsen, Lars K; Bannon, Gary A

    2008-10-01

    There is detailed guidance on how to perform bioinformatic analyses and enzymatic degradation studies for genetically modified crops under consideration for approval by regulatory agencies; however, there is no consensus in the scientific community on the details of how to perform IgE serum studies. IgE serum studies are an important safety component to acceptance of genetically modified crops when the introduced protein is novel, the introduced protein is similar to known allergens, or the crop is allergenic. In this manuscript, we describe the characteristics of the reagents, validation of assay performance, and data analysis necessary to optimize the information obtained from serum testing of novel proteins and genetically modified (GM) crops and to make results more accurate and comparable between different investigations. PMID:18727951

  15. An advanced deterministic method for spent-fuel criticality safety analysis

    SciTech Connect

    DeHart, M.D.

    1998-09-01

    Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte Carlo methods for criticality calculations. Monte Carlo has become popular because of its capability to model complex, nonorthogonal configurations or fissile materials, typical of real-world problems. In the last few years, however, interest in determinist transport methods has been revived, due to shortcomings in the stochastic nature of Monte Carlo approaches for certain types of analyses. Specifically, deterministic methods are superior to stochastic methods for calculations requiring accurate neutron density distributions or differential fluxes. Although Monte Carlo methods are well suited for eigenvalue calculations, they lack the localized detail necessary to assess uncertainties and sensitivities important in determining a range of applicability. Monte Carlo methods are also inefficient as a transport solution for multiple-pin depletion methods. Discrete ordinates methods have long been recognized as one of the most rigorous and accurate approximations used to solve the transport equation. However, until recently, geometric constrains in finite differencing schemes have made discrete ordinates methods impractical for nonorthogonal configurations such as reactor fuel assemblies. The development of an extended step characteristic (ESC) technique removes the grid structure limitation of traditional discrete ordinates methods. The NEWT computer code, a discrete ordinates code built on the ESC formalism, is being developed as part of the SCALE code system. This paper demonstrates the power, versatility, and applicability of NEWT as a state-of-the-art solution for current computational needs.

  16. Possibility of experimental validation of criticality safety methodology in support of underground fuel storage efforts

    SciTech Connect

    Nikolaev, M.N.; Briggs, J.B.

    1997-10-01

    Critical systems which might be formed in geologic repositories as a result of long-term degradation of the storage media, leaching of plutonium from the storage media, and the redistribution of low concentrations of plutonium into underground sand layers or lenses can be characterized by positive reactivity feedback. Formation of such a systems can not be excluded when considering the burial of high enriched uranium or plutonium contaminated wastes or spent nuclear fuels. Although the probability of formation of a critical systems under such conditions is very low, the reliable prediction of neutron multiplication properties appears to be of great interest from a criticality safety view point. At the present time, all estimations of criticality are based only on evaluated neutron data because critical experiments are not available for large systems containing small quantities plutonium distributed throughout a typically encountered matrix material such as silicon dioxide. The possibility of providing such an experiment using the large Russian critical assemblies, BFS-1 or BFS-2, is considered. It is shown that critical systems containing small amounts of hydrogenous material (polyethylene) with positive reactivity feedback can by modeled in the BFS Facility.

  17. Bounding criticality safety analyses for shipments of unconfigured spent nuclear fuel

    SciTech Connect

    Lichtenwalter, J.J.; Parks, C.V.

    1998-06-01

    In November 1996, a request was made to the US Department of Energy for a waiver for three shipments of spent nuclear fuel (SNF) from Oak Ridge National Laboratory (ORNL) to the Savannah River Site (SRS) in the US NRC certified BMI-1 cask (CoC 5957). Although the post-irradiation fissile mass (based on chemical assays) in each shipment was less than 800 g, a criticality safety analysis was needed because the pre-irradiation mass exceeded 800 g, the fissile material limit in the CoC. The analyses were performed on SNF consisting of aluminum-clad U{sub 3}O{sub 8}, UAl{sub x}, and U{sub 3}Si{sub 2} plates, fragments and pieces that had been irradiated at ORNL during the Reduced Enrichment Research and Test Reactor Program of the 1980s. The highlights of the approach used to analyze this unique SNF and the benefits of the waiver are presented in this paper.

  18. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... approval or design certification under part 52 of this chapter is required by the provisions of 10 CFR 52... of 10 CFR 52.157 to include in its final safety analysis report a description of the quality...: reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of...

  19. Recommendations from the workshop on Comparative Approaches to Safety Assessment of GM Plant Materials: A road toward harmonized criteria?

    PubMed

    Bartholomaeus, Andrew; Batista, Juan Carlos; Burachik, Moisés; Parrott, Wayne

    2015-01-01

    An international meeting of genetically modified (GM) food safety assessors from the main importing and exporting countries from Asia and the Americas was held in Buenos Aires, Argentina, between June 26(th) and 28(th), 2013. Participants shared their evaluation approaches, identified similarities and challenges, and used their experience to propose areas for future work. Recommendations for improving risk assessment procedures and avenues for future collaboration were also discussed. The deliberations of the meeting were also supported by a survey of participants which canvassed risk assessment approaches across the regions from which participants came. This project was initiated by Argentine Agri-Food Health and Quality National Service (SENASA, Ministry of Agriculture, Argentina), with the support of the International Life Sciences Institute (ILSI) and other partner institutions. The importance of making all possible efforts toward more integrated and harmonized regulatory oversight for GM organisms (GMOs) was strongly emphasized. This exercise showed that such harmonization is a feasible goal that would contribute to sustain a fluid trade of commodities and ultimately enhance food security. Before this can be achieved, key issues identified in this meeting will have to be addressed in the near future to enable regulatory collaboration or joint work. The authors propose that the recommendations coming out of the meeting should be used as a basis for continuing work, follow up discussions and concrete actions. PMID:25706477

  20. A case study on the influence of THM coupling on the near field safety of a spent fuel repository in sparsely fractured granite

    SciTech Connect

    Nguyen, T.S.; Borgesson, L.; Chijimatsu, M.; Hernelind, J.; Jing, L.; Kobayashi, A.; Rutqvist, J.

    2009-03-01

    In order to demonstrate the feasibility of geological disposal of spent CANDU fuel in Canada, a safety assessment was performed for a hypothetical repository in the Canadian Shield. The assessment shows that such repository would meet international criteria for dose rate; however, uncertainties in the assumed evolution of the repository were identified. Such uncertainties could be resolved by the consideration of coupled Thermal-Hydro-Mechanical-Chemical (THMC) processes. In Task A of the DECOVALEX-THMC project, THM models were developed within the framework of the theory of poroelasticity. Such model development was performed in an iterative manner, using experimental data from laboratory and field tests. The models were used to perform near-field simulations of the evolution of the repository in order to address the above uncertainties. This paper presents the definition and rationale of task A and the results of the simulations. From a repository safety point of view, the simulations predict that the maximum temperature would be well below the design target of 100 C, however the load on the container can marginally exceed the design value of 15 MPa. However, the most important finding from the simulations is that a rock damage zone could form around the emplacement borehole. Such damage zone can extend a few metres from the walls of the emplacement holes, with permeability values that are orders of magnitude higher than the initial values. The damage zone has the potential to increase the radionuclide transport flux from the geosphere; the effect of such an increase should be taken into account in the safety assessment and mitigated if necessary by the provision of sealing systems.

  1. A case study on the influence of THM coupling on the near field safety of a spent fuel repository in sparsely fractured granite

    NASA Astrophysics Data System (ADS)

    Nguyen, Thanh Son; Börgesson, Lennart; Chijimatsu, Masakazu; Hernelind, Jan; Jing, Lanru; Kobayashi, Akira; Rutqvist, Jonny

    2009-05-01

    In order to demonstrate the feasibility of geological disposal of spent CANDU fuel in Canada, a safety assessment was performed for a hypothetical repository in the Canadian Shield. The assessment shows that the maximum long term radionuclide release from such repository would meet international criteria for dose rate; however, uncertainties in the assumed evolution of the repository were identified. Such uncertainties could be resolved by the consideration of coupled Thermal-Hydro-Mechanical-Chemical (THMC) processes. In Task A of the DECOVALEX-THMC project, THM models were developed within the framework of the theory of poroelasticity. Such model development was performed in an iterative manner, using experimental data from laboratory and field tests. The models were used to perform near-field simulations of the evolution of the repository in order to address the above-mentioned uncertainties. This paper presents the definition and rationale of task A and the results of the simulations. From a repository safety point of view, the simulations predict that the maximum temperature would be well below the design target of 100°C; however, the stress on the container can marginally exceed the design value of 15 MPa. However, the most important finding from the simulations is that a rock damage zone could form around the emplacement borehole. Such damage zone can extend a few metres from the walls of the emplacement holes, with permeability values that are orders of magnitude higher than the initial values. The damage zone has the potential to increase the radionuclide transport flux from the geosphere; the effect of such an increase should be taken into account in the safety assessment and mitigated if necessary by the provision of sealing systems.

  2. An Integrated Marine Propulsion System Utilising TRIGA{sup TM} Fuel

    SciTech Connect

    Manach, G.; Monnez, J-P.; Freeman, M.J.; Newell, A.; Brushwood, J.M.; Thompson, A.; Collins, C.; Scholes, N.; Hamilton, P.J.; Beeley, P.A.

    2004-07-01

    This paper describes the reactor physics, shielding, thermal hydraulics, reactor dynamics and safety studies conducted to develop a proposed Integrated Marine Propulsion System (IMPS) utilising TRIGA{sup TM} type uranium zirconium hydride fuel. The study has demonstrated that the IMPS plant is feasible and meets the design safety principles and safety criteria imposed on the study. (authors)

  3. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.

  4. Spent Nuclear Fuel (SNF) project Integrated Safety Management System phase I and II Verification Review Plan

    SciTech Connect

    CARTER, R.P.

    1999-11-19

    The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78). Integrated Safety Management (ISM) requires contractors to integrate safety into management and work practices at all levels so that missions are achieved while protecting the public, the worker, and the environment. The contractor is required to describe the Integrated Safety Management System (ISMS) to be used to implement the safety performance objective.

  5. Safety.

    ERIC Educational Resources Information Center

    Education in Science, 1996

    1996-01-01

    Discusses safety issues in science, including: allergic reactions to peanuts used in experiments; explosions in lead/acid batteries; and inspection of pressure vessels, such as pressure cookers or model steam engines. (MKR)

  6. Preliminary safety evaluation for the spent nuclear fuel project`s cold vacuum drying system

    SciTech Connect

    Garvin, L.J., Westinghouse Hanford

    1996-07-01

    This preliminary safety evaluation (PSE) considers only the Cold Vacuum Drying System (CVDS) facility and its mission as it relates to the integrated process strategy (WHC 1995). The purpose of the PSE is to identify those CBDS design functions that may require safety- class and safety-significant accident prevention and mitigation features.

  7. Safety considerations in testing a fuel-rich aeropropulsion gas generator

    NASA Technical Reports Server (NTRS)

    Rollbuhler, R. James; Hulligan, David D.

    1991-01-01

    A catalyst containing reactor is being tested using a fuel-rich mixture of Jet A fuel and hot input air. The reactor product is a gaseous fuel that can be utilized in aeropropulsion gas turbine engines. Because the catalyst material is susceptible to damage from high temperature conditions, fuel-rich operating conditions are attained by introducing the fuel first into an inert gas stream in the reactor and then displacing the inert gas with reaction air. Once a desired fuel-to-air ratio is attained, only limited time is allowed for a catalyst induced reaction to occur; otherwise the inert gas is substituted for the air and the fuel flow is terminated. Because there presently is not a gas turbine combustor in which to burn the reactor product gas, the gas is combusted at the outlet of the test facility flare stack. This technique in operations has worked successfully in over 200 tests.

  8. 24 CFR 51.203 - Safety standards.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 24 Housing and Urban Development 1 2013-04-01 2013-04-01 false Safety standards. 51.203 Section 51.203 Housing and Urban Development Office of the Secretary, Department of Housing and Urban Development ENVIRONMENTAL CRITERIA AND STANDARDS Siting of HUD-Assisted Projects Near Hazardous Operations Handling Conventional Fuels or Chemicals of...

  9. Criticality Safety Scoping Study for the Transport of Weapons-Grade Mixed-Oxide Fuel Using the MO-1 Shipping Package

    SciTech Connect

    Dunn, M.E.; Fox, P.B.

    1999-05-01

    This report provides the criticality safety information needed for obtaining certification of the shipment of mixed-oxide (MOX) fuel using the MO-1 [USA/9069/B()F] shipping package. Specifically, this report addresses the shipment of non-weapons-grade MOX fuel as certified under Certificate of Compliance 9069, Revision 10. The report further addresses the shipment of weapons-grade MOX fuel using a possible Westinghouse fuel design. Criticality safety analysis information is provided to demonstrate that the requirements of 10 CFR S 71.55 and 71.59 are satisfied for the MO-1 package. Using NUREG/CR-5661 as a guide, a transport index (TI) for criticality control is determined for the shipment of non-weapons-grade MOX fuel as specified in Certificate of Compliance 9069, Revision 10. A TI for criticality control is also determined for the shipment of weapons-grade MOX fuel. Since the possible weapons-grade fuel design is preliminary in nature, this report is considered to be a scoping evaluation and is not intended as a substitute for the final criticality safety analysis of the MO-1 shipping package. However, the criticality safety evaluation information that is presented in this report does demonstrate the feasibility of obtaining certification for the transport of weapons-grade MOX lead test fuel using the MO-1 shipping package.

  10. Criteria for Developing Criteria Sets.

    ERIC Educational Resources Information Center

    Martin, James L.

    Criteria sets are a necessary step in the systematic development of evaluation in education. Evaluation results from the combination of criteria and evidence. There is a need to develop explicit tools for evaluating criteria, similar to those used in evaluating evidence. The formulation of such criteria depends on distinguishing between terms…

  11. TITLE: Environmental, health, and safety issues offuel cells in transportation. Volume 1: Phosphoricacid fuel-cell buses

    NASA Astrophysics Data System (ADS)

    Ring, Shan

    1994-12-01

    The U.S. Department of Energy (DOE) chartered the Phosphoric Acid Fuel-Cell (PAFC) Bus Program to demonstrate the feasibility of fuel cells in heavy-duty transportation systems. As part of this program, PAFC- powered buses are being built to meet transit industry design and performance standards. Test-bed bus-1 (TBB-1) was designed in 1993 and integrated in March 1994. TBB-2 and TBB-3 are under construction and should be integrated in early 1995. In 1987 Phase 1 of the program began with the development and testing of two conceptual system designs- liquid- and air-cooled systems. The liquid-cooled PAFC system was chosen to continue, through a competitive award, into Phase H, beginning in 1991. Three hybrid buses, which combine fuel-cell and battery technologies, were designed during Phase 3. After completing Phase 2, DOE plans a comprehensive performance testing program (Phase H1) to verify that the buses meet stringent transit industry requirements. The Phase 3 study will evaluate the PAFC bus and compare it to a conventional diesel bus. This NREL study assesses the environmental, health, and safety (EH&S) issues that may affect the commercialization of the PAFC bus. Because safety is a critical factor for consumer acceptance of new transportation-based technologies the study focuses on these issues. The study examines health and safety together because they are integrally related. In addition, this report briefly discusses two environmental issues that are of concern to the Environmental Protection Agency (EPA). The first issue involves a surge battery used by the PAFC bus that contains hazardous constituents. The second issue concerns the regulated air emissions produced during operation of the PAFC bus.

  12. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    SciTech Connect

    Pesic, Milan P

    2003-10-15

    The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

  13. WNA's worldwide overview on front-end nuclear fuel cycle growth and health, safety and environmental issues.

    PubMed

    Saint-Pierre, Sylvain; Kidd, Steve

    2011-01-01

    This paper presents the WNA's worldwide nuclear industry overview on the anticipated growth of the front-end nuclear fuel cycle from uranium mining to conversion and enrichment, and on the related key health, safety, and environmental (HSE) issues and challenges. It also puts an emphasis on uranium mining in new producing countries with insufficiently developed regulatory regimes that pose greater HSE concerns. It introduces the new WNA policy on uranium mining: Sustaining Global Best Practices in Uranium Mining and Processing-Principles for Managing Radiation, Health and Safety and the Environment, which is an outgrowth of an International Atomic Energy Agency (IAEA) cooperation project that closely involved industry and governmental experts in uranium mining from around the world. PMID:21399410

  14. General-purpose heat source project and space nuclear safety and fuels program. Progress reportt, January 1980

    SciTech Connect

    Maraman, W.J.

    1980-04-01

    This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are the general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work.

  15. The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

    SciTech Connect

    Logar, Marjan; Jeraj, Robert; Glumac, Bogdan

    2003-02-15

    It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool k{sub eff} is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor k{sub eff}. In the interval in which the pitch was changed, k{sub eff} decreased for up to {approx}0.4 for standard and {approx}0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, k{sub eff} values are smaller for {approx}0.2 ({approx}0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is {approx}0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the k{sub eff} of spent-fuel pool and can be neglected in spent-fuel pool design.

  16. Fire protection design criteria

    SciTech Connect

    1997-03-01

    This Standard provides supplemental fire protection guidance applicable to the design and construction of DOE facilities and site features (such as water distribution systems) that are also provided for fire protection. It is intended to be used in conjunction with the applicable building code, national Fire Protection Association Codes and Standards, and any other applicable DOE construction criteria. This Standard, along with other delineated criteria, constitutes the basic criteria for satisfying DOE fire and life safety objectives for the design and construction or renovation of DOE facilities.

  17. Plutonium storage criteria

    SciTech Connect

    Chung, D.; Ascanio, X.

    1996-05-01

    The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less than 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.

  18. 30 CFR 75.1904 - Underground diesel fuel tanks and safety cans.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... have liquid tight welded seams; (4) Not leak; and (5) For stationary tanks in permanent underground... for stationary tanks in permanent underground diesel fuel storage facilities and self-closing caps...

  19. 30 CFR 75.1904 - Underground diesel fuel tanks and safety cans.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... have liquid tight welded seams; (4) Not leak; and (5) For stationary tanks in permanent underground... for stationary tanks in permanent underground diesel fuel storage facilities and self-closing caps...

  20. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    SciTech Connect

    Wiesenack, W.

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  1. Severe accidents in spent fuel pools in support of generic safety, Issue 82

    SciTech Connect

    Sailor, V.L.; Perkins, K.R.; Weeks, J.R.; Connell, H.R.

    1987-07-01

    This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel storage pool designs are considered based on a preliminary screening study which tried to identify vulnerabilities. Internal and external events and accidents are assessed. Conditions which could lead to failure of the spent fuel Zircaloy cladding as a result of cladding rupture or as a result of a self-sustaining oxidation reaction are presented. Propagation of a cladding fire to older stored fuel assemblies is evaluated. Spent fuel pool fission product inventory is estimated and the releases and consequences for the various cladding scenarios are provided. Possible preventive or mitigative measures are qualitatively evaluated. The uncertainties in the risk estimate are large, and areas where additional evaluations are needed to reduce uncertainty are identified.

  2. The importance of criticality in the safety analysis of the spent-fuel waste container

    SciTech Connect

    Culbreth, W.G.; Zielinski, P.

    1993-12-31

    The storage of high-level spent reactor fuel in a proposed national geologic repository will require the construction of containers to be placed in boreholes drilled into the host rock. Federal regulations require that the fuel be maintained subcritical under normal or accident conditions. This is determined through the calculation of a neutron multiplication factor, k{sub eff}, that must remain below 0.95. Criticality will play an important role in the container design, the internal configuration of the fuel, and the selection of neutron poisons. An analysis of k{sub eff} should be a normal step in the conceptualization of new waste container designs. Unlike thermal effects in a proposed repository, criticality will remain a problem long after the 10,000 year lifetime of the facility. In this study, nuclear criticality has been determined for the proposed spent fuel container in various situations that include varying fuel enrichment and partial air gap flooding. Results will be presented to demonstrate the impact of these variables on the design of a safe spent fuel container.

  3. Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

    SciTech Connect

    O'Donnell, F.R.; Hoy, H.C.

    1981-10-01

    Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/sup 12/ Btu of energy output, and per other appropriate units of output.

  4. Criticality safety requirements for transporting EBR-II fuel bottles stored at INTEC

    SciTech Connect

    Lell, R. M.; Pope, C. L.

    2000-03-14

    Two carrier/shipping cask options are being developed to transport bottles of EBR-II fuel elements stored at INTEC. Some fuel bottles are intact, but some have developed leaks. Reactivity control requirements to maintain subcriticality during the hypothetical transport accident have been examined for both transport options for intact and leaking bottles. Poison rods, poison sleeves, and dummy filler bottles were considered; several possible poison materials and several possible dummy filler materials were studied. The minimum number of poison rods or dummy filler bottles has been determined for each carrier for transport of intact and leaking bottles.

  5. 30 CFR 75.1903 - Underground diesel fuel storage facilities and areas; construction and safety precautions.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... percent of the maximum capacity of the fuel storage system; and (7) Provided with a competent concrete...) When it is necessary to weld, cut, or solder pipelines, tanks, or other containers that may have..., tanks, or containers that have been welded, soldered, brazed, or cut until the metal has cooled...

  6. 30 CFR 75.1903 - Underground diesel fuel storage facilities and areas; construction and safety precautions.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... percent of the maximum capacity of the fuel storage system; and (7) Provided with a competent concrete...) When it is necessary to weld, cut, or solder pipelines, tanks, or other containers that may have..., tanks, or containers that have been welded, soldered, brazed, or cut until the metal has cooled...

  7. Standard review plan for reviewing safety analysis reports for dry metallic spent fuel storage casks

    SciTech Connect

    Not Available

    1988-01-01

    The Cask Standard Review Plan (CSRP) has been prepared as guidance to be used in the review of Cask Safety Analysis Reports (CSARs) for storage packages. The principal purpose of the CSRP is to assure the quality and uniformity of storage cask reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The CSRP also sets forth solutions and approaches determined to be acceptable in the past by the NRC staff in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a CSAR does not have to follow the solutions or approaches presented in the CSRP. However, applicants should recognize that the NRC staff has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches.

  8. Data quality objective to support resolution of the organic fuel rich tank safety issue

    SciTech Connect

    Buckley, L.L.

    1995-04-28

    During years of Hanford process history, large quantities of complexants used in waste management operations as well as an unknown quantity of degradation products of the solvents used in fuel reprocessing and metal recovery were added to man of the 149 single-shell tanks. These waste tanks also contain a presumed stoichiometric excess of sodium nitrate/nitrite oxidizers, sufficient to exothermically oxidize the organic compounds if suitably initiated. This DQO identifies the questions that must be answered to appropriately disposition organic watchlist tanks, identifies a strategy to deal with false positive or negative judgements associated with analytical uncertainty, and list the analytes of concern to support dealing with organic watchlist concerns. Uncertainties associated with both assay limitations and matrix effects complicate selection of analytes. This results in requiring at least two independent measures of potential fuel reactivity.

  9. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    DOE PAGESBeta

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-06-15

    Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  10. Full-length high-temperature severe fuel damage test No. 2. Final safety analysis

    SciTech Connect

    Hesson, G.M.; Lombardo, N.J.; Pilger, J.P.; Rausch, W.N.; King, L.L.; Hurley, D.E.; Parchen, L.J.; Panisko, F.E.

    1993-09-01

    Hazardous conditions associated with performing the Full-Length High- Temperature (FLHT). Severe Fuel Damage Test No. 2 experiment have been analyzed. Major hazards that could cause harm or damage are (1) radioactive fission products, (2) radiation fields, (3) reactivity changes, (4) hydrogen generation, (5) materials at high temperature, (6) steam explosion, and (7) steam pressure pulse. As a result of this analysis, it is concluded that with proper precautions the FLHT- 2 test can be safely conducted.

  11. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    SciTech Connect

    Williams, M.L. Wiarda, D.; Ilas, G.; Marshall, W.J.; Rearden, B.T.

    2015-01-15

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  12. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2015-01-01

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  13. Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

    SciTech Connect

    1980-10-01

    Volume 2 consists of 19 reports describing technical effort performed by Government Contractors in the area of LNG Safety and Environmental Control. Report topics are: simulation of LNG vapor spread and dispersion by finite element methods; modeling of negatively buoyant vapor cloud dispersion; effect of humidity on the energy budget of a liquefied natural gas (LNG) vapor cloud; LNG fire and explosion phenomena research evaluation; modeling of laminar flames in mixtures of vaporized liquefied natural gas (LNG) and air; chemical kinetics in LNG detonations; effects of cellular structure on the behavior of gaseous detonation waves under transient conditions; computer simulation of combustion and fluid dynamics in two and three dimensions; LNG release prevention and control; the feasibility of methods and systems for reducing LNG tanker fire hazards; safety assessment of gelled LNG; and a four band differential radiometer for monitoring LNG vapors.

  14. The long term storage of radioactive waste and spent fuel: safety and policy considerations

    SciTech Connect

    Rowat, J.; Metcalf, P.

    2007-07-01

    Storage is a necessary step in the overall management of radioactive waste. In recent years, due to the unavailability of disposal facilities, storage facilities intended originally as temporary, have had their lifetimes extended and consideration has been given, in some countries, to the use of long term storage (LTS) as a management option. In 2003, the IAEA published a position paper titled 'The Long Term Storage of Radioactive Waste: Safety and Sustainability'. The position paper, which written for a non-specialist audience, focused on seven key factors for safety and sustainability of LTS, namely: safety, maintenance/institutional control, retrieval, security, costs, community attitudes and retention of information. The Agency is preparing a follow-up report to the position paper that elaborates in a more technical manner upon the issues raised in the position paper and issues important for implementation of LTS. It also provides some discussion of the reasons for implementing a LTS option and contrasts LTS with aspects of other management options. The present paper provides an overview of the draft follow-up report. (authors)

  15. Technology, safety and costs of decommissioning reference nuclear fuel cycle facilities

    SciTech Connect

    Elder, H.K.

    1986-05-01

    The radioactive wastes expected to result from decommissioning nuclear fuel cycle facilities are reviewed and classified in accordance with 10 CFR 61. Most of the wastes from the MOX plant (exclusive of the lagoon wastes) will require interim storage (11% Class A 49 m/sup 3/; 89% interim storage, 383 m/sup 3/). The MOX plant lagoon wastes are Class A waste (2930 m/sup 3/). All of the wastes from the U-Fab and UF/sub 6/ plants are designated as Class A waste (U-Fab 1090 m/sup 3/, UF/sub 6/ 1259 m/sup 3/).

  16. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    SciTech Connect

    Carvo, Alan E.

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  17. Bibliography on aircraft fire hazards and safety. Volume 2: Safety. Part 1: Key numbers 1 to 524

    NASA Technical Reports Server (NTRS)

    Pelouch, J. J., Jr. (Compiler); Hacker, P. T. (Compiler)

    1974-01-01

    Bibliographic citations are presented to describe and define aircraft safety methods, equipment, and criteria. Some of the subjects discussed are: (1) fire and explosion suppression using whiffle balls, (2) ultraviolet flame detecting sensors, (3) evaluation of flame arrestor materials for aircraft fuel systems, (4) crash fire prevention system for supersonic commercial aircraft, and (5) fire suppression for aerospace vehicles.

  18. Well-to-Wheels Analysis of Advanced Fuel/Vehicle Systems: A North American Study of Energy Use, Greenhouse Gas Emissions, and Criteria Pollutant Emissions

    SciTech Connect

    Brinkman, Norman; Wang, Michael; Weber, Trudy; Darlington, Thomas

    2005-05-01

    An accurate assessment of future fuel/propulsion system options requires a complete vehicle fuel-cycle analysis, commonly called a well-to-wheels (WTW) analysis. This WTW study analyzes energy use and emissions associated with fuel production (or well-to-tank [WTT]) activities and energy use and emissions associated with vehicle operation (or tank-to-wheels [TTW]) activities.

  19. Development Trends in Nuclear Technology and Related Safety Aspects

    SciTech Connect

    Kuczera, B.; Juhn, P.E.; Fukuda, K.

    2002-07-01

    The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO sub-task 'User Requirements and Nuclear Energy Development Criteria in the Area of Safety' have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R and D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle. (authors)

  20. Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors

    SciTech Connect

    Biswas, D; Mennerdahl, D

    2008-06-23

    The ANSI/ANS 8.12 standard was first approved in July 1978. At that time, this edition was applicable to operations with plutonium-uranium oxide (MOX) fuel mixtures outside reactors and was limited to subcritical limits for homogeneous systems. The next major revision, ANSI/ANS-8.12-1987, included the addition of subcritical limits for heterogeneous systems. The standard was subsequently reaffirmed in February 1993. During late 1990s, substantial work was done by the ANS 8.12 Standard Working Group to re-examine the technical data presented in the standard using the latest codes and cross section sets. Calculations performed showed good agreement with the values published in the standard. This effort resulted in the reaffirmation of the standard in March 2002. The standard is currently in a maintenance mode. After 2002, activities included discussions to determine the future direction of the standard and to follow the MOX standard development by the International Standard Organization (ISO). In 2007, the Working Group decided to revise the standard to extend the areas of applicability by providing a wider range of subcritical data. The intent is to cover a wider domain of MOX fuel fabrication and operations. It was also decided to follow the ISO MOX standard specifications (related to MOX density and isotopics) and develop a new set of subcritical limits for homogeneous systems. This has resulted in the submittal (and subsequent approval) of the project initiation notification system form (PINS) in 2007.

  1. GNS-12 Packaging design criteria

    SciTech Connect

    Clements, E.P., Westinghouse Hanford

    1996-07-24

    The purpose of this Packaging Design Criteria (PDC) is to provide criteria for the Safety Analysis Report for Packaging (SARP)(Onsite). The SARP provides the evaluation to demonstrate that the onsite transportation safety criteria are met for the transport and storage of the 324 Building vitrified encapsulated material in the GNS-12 cask. In this application, the approved PDC provides a formal set of standards for the payload requirements, and guidance for the current cask transport configuration and a revised storage seal and primary lid modification design.

  2. Oxygen and Fuel Jet Diffusion Flame Studies in Microgravity Motivated by Spacecraft Oxygen Storage Fire Safety

    NASA Technical Reports Server (NTRS)

    Sunderland, P. B.; Yuan, Z.-G.; Krishnan, S. S.; Abshire, J. M.; Gore, J. P.

    2003-01-01

    Owing to the absence of past work involving flames similar to the Mir fire namely oxygen-enhanced, inverse gas-jet diffusion flames in microgravity the objectives of this work are as follows: 1. Observe the effects of enhanced oxygen conditions on laminar jet diffusion flames with ethane fuel. 2. Consider both earth gravity and microgravity. 3. Examine both normal and inverse flames. 4. Compare the measured flame lengths and widths with calibrated predictions of several flame shape models. This study expands on the work of Hwang and Gore which emphasized radiative emissions from oxygen-enhanced inverse flames in earth gravity, and Sunderland et al. which emphasized the shapes of normal and inverse oxygen-enhanced gas-jet diffusion flames in microgravity.

  3. AREVA NP next generation fresh UO{sub 2} fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    SciTech Connect

    Doucet, M.; Landrieu, M.; Montgomery, R.; O' Donnell, B.

    2007-07-01

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO{sub 2} shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - Boral{sup TM} as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The

  4. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    SciTech Connect

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  5. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect

    Not Available

    1980-09-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  6. 49 CFR Appendix to Subpart H of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Non-North America-Domiciled...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... experienced during the past 12 months. Recordable accident, as defined in 49 CFR 390.5, means an accident... Evaluation Criteria for Non-North America-Domiciled Motor Carriers Appendix to Subpart H of Part 385... Special Rules for New Entrant Non-North America-Domiciled Carriers Pt. 385, Subpt. H, App. Appendix...

  7. 49 CFR Appendix to Subpart H of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Non-North America-Domiciled...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... experienced during the past 12 months. Recordable accident, as defined in 49 CFR 390.5, means an accident... Evaluation Criteria for Non-North America-Domiciled Motor Carriers Appendix to Subpart H of Part 385... Special Rules for New Entrant Non-North America-Domiciled Carriers Pt. 385, Subpt. H, App. Appendix...

  8. 49 CFR Appendix A to Subpart E of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Mexico-Domiciled Motor...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... months. Recordable accident, as defined in 49 CFR 390.5, means an accident involving a commercial motor... Evaluation Criteria for Mexico-Domiciled Motor Carriers A Appendix A to Subpart E of Part 365 Transportation... OPERATING AUTHORITY Special Rules for Certain Mexico-domiciled Carriers Pt. 365, Subpt. E, App. A Appendix...

  9. Unreviewed Safety Question Determination for TOPAZ II uranium fuel pellet production at the Plutonium Handling Facility (PF-4), Technical Area 55, Los Alamos National Laboratory

    SciTech Connect

    Gordon, D.J.P.

    1993-09-29

    Enriched uranium oxide, nitride, and carbide fuel pellets have been produced at PF-4 since the facility became operational in the late 1970s. The TOPAZ II reactors require fuel enriched to 97% uranium-235. Approximately 75 kilograms (kgs) of uranium will be processed per year in support of this program. The amount of fuel processed per year at PF-4 will not be increased for these programs, but the batch size will be increased to approximately 3 kgs of uranium. The current DOE-approved Final Safety Analysis Report (FSAR) calls for batches containing 45 grams (gms) of plutonium-239 and 172 gms of uranium-235. The impact of increasing the uranium batch size on the facility authorization basis is analyzed in the attached Safety Evaluation Worksheet. In addition, the structural modification for the transformer and vacuum pump installation, required to support the operation, is evaluated. Based on the attached Safety Evaluation, it has been determined that the change in uranium batch size does not constitute an Unreviewed Safety Question (USQ), the increase in uranium batch size does not increase the probability or consequences of any accidents previously analyzed and does not create the possibility for a new type of accident or reduce the margin of safety in the Operational Safety Requirements (OSRs). Similarly, the structural modifications required for the transformer and vacuum pump installation do not increase the probability or consequence of any accident previously analyzed and do not create the possibility for a new type of accident or reduce any margin of safety in the OSRS.

  10. FHR Generic Design Criteria

    SciTech Connect

    Flanagan, G.F.; Holcomb, D.E.; Cetiner, S.M.

    2012-06-15

    The purpose of this document is to provide an initial, focused reference to the safety characteristics of and a licensing approach for Fluoride-Salt-Cooled High-Temperature Reactors (FHRs). The document does not contain details of particular reactor designs nor does it attempt to identify or classify either design basis or beyond design basis accidents. Further, this document is an initial attempt by a small set of subject matter experts to document the safety and licensing characteristics of FHRs for a larger audience. The document is intended to help in setting the safety and licensing research, development, and demonstration path forward. Input from a wider audience, further technical developments, and additional study will be required to develop a consensus position on the safety and licensing characteristics of FHRs. This document begins with a brief overview of the attributes of FHRs and then a general description of their anticipated safety performance. Following this, an overview of the US nuclear power plant approval process is provided that includes both test and power reactors, as well as the role of safety standards in the approval process. The document next describes a General Design Criteria (GDC)–based approach to licensing an FHR and provides an initial draft set of FHR GDCs. The document concludes with a description of a path forward toward developing an FHR safety standard that can support both a test and power reactor licensing process.

  11. FHR Generic Design Criteria

    SciTech Connect

    Flanagan, George F; Holcomb, David Eugene; Cetiner, Sacit M

    2012-06-01

    The purpose of this document is to provide an initial, focused reference to the safety characteristics of and a licensing approach for Fluoride-Salt-Cooled High-Temperature Reactors (FHRs). The document does not contain details of particular reactor designs nor does it attempt to identify or classify either design basis or beyond design basis accidents. Further, this document is an initial attempt by a small set of subject matter experts to document the safety and licensing characteristics of FHRs for a larger audience. The document is intended to help in setting the safety and licensing research, development, and demonstration path forward. Input from a wider audience, further technical developments, and additional study will be required to develop a consensus position on the safety and licensing characteristics of FHRs. This document begins with a brief overview of the attributes of FHRs and then a general description of their anticipated safety performance. Following this, an overview of the US nuclear power plant approval process is provided that includes both test and power reactors, as well as the role of safety standards in the approval process. The document next describes a General Design Criteria (GDC) - based approach to licensing an FHR and provides an initial draft set of FHR GDCs. The document concludes with a description of a path forward toward developing an FHR safety standard that can support both a test and power reactor licensing process.

  12. Strain-Based Acceptance Criteria for Energy-Limited Events

    SciTech Connect

    Spencer D. Snow; Dana K. Morton

    2009-07-01

    The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code was primarily written with stress-based acceptance criteria. These criteria are applicable to force, displacement, and energy-controlled loadings and ensure a factor of safety against failure. However, stress-based acceptance criteria are often quite conservative for one time energy-limited events such as accidental drops and impacts. For several years, the ASME Working Group on Design of Division 3 Containments has been developing the Design Articles for Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High-Level Radioactive Material and Waste,” and has wanted to establish strain-based acceptance criteria for accidental drops of containments. This Division 3 working group asked the Working Group on Design Methodology (WGDM) to assist in developing these strain-based acceptance criteria. This paper discusses the current proposed strain-based acceptance criteria, associated limitations of use, its background development, and the current status.

  13. Spent nuclear fuel storage. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1997-07-01

    The bibliography contains citations concerning spent nuclear fuel storage technologies, facilities, sites, and assessment. References review wet and dry storage, spent fuel casks and pools, underground storage, monitored and retrievable storage systems, and aluminum-clad spent fuels. Environmental impact, siting criteria, regulations, and risk assessment are also discussed. Computer codes and models for storage safety are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  14. Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel

    SciTech Connect

    Gauld, I. C.; Ryman, J. C.

    2000-12-11

    This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance. The importance is investigated as a function of increasing burnup to assist in identifying the key changes in spent fuel characteristics between conventional- and extended-burnup regimes. Studies involving both pressurized water-reactor (PWR) fuel assemblies and boiling-water-reactor (BWR) assemblies are included. This study is seen to be a necessary first step in identifying the high-burnup spent fuel characteristics that may adversely affect the accuracy of current computational methods and data, assess the potential impact on previous guidance on isotopic source terms and decay-heat values, and thus help identify areas for methods and data improvement. Finally, several recommendations on the direction of possible future code validation efforts for high-burnup spent fuel predictions are presented.

  15. 40 CFR 600.115-08 - Criteria for determining the fuel economy label calculation method for 2011 and later model year...

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON... 5-cycle method for determining fuel economy label values, as specified in § 600.210-08 (a)(2) or (b... the vehicle-specific 5-cycle method specified in § 600.210-08(a)(1) or (b)(1), as applicable....

  16. Key Differences in the Fabrication, Irradiation, and Safety Testing of U.S. and German TRISO-coated Particle Fuel and Their Implications on Fuel Performance

    SciTech Connect

    Petti, David Andrew; Maki, John Thomas; Buongiorno, Jacopo; Hobbins, Richard Redfield

    2002-06-01

    High temperature gas reactor technology is achieving a renaissance around the world. This technology relies on high quality production and performance of coated particle fuel. Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the United States. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than U.S. fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the U.S. and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the U.S. fuel has not faired as well, and what process/ production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer U.S. irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, and degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

  17. 29 CFR 1904.4 - Recording criteria.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 29 Labor 5 2013-07-01 2013-07-01 false Recording criteria. 1904.4 Section 1904.4 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RECORDING AND REPORTING OCCUPATIONAL INJURIES AND ILLNESSES Recordkeeping Forms and Recording Criteria § 1904.4 Recording criteria. (a)...

  18. Review of Overall Safety Manual for space nuclear systems. An evaluation of a nuclear safety analysis methodology for plutonium-fueled space nuclear systems

    SciTech Connect

    Coleman, J.; Inhaber, H.

    1984-02-01

    As part of its duties in connection with space missions involving nuclear power sources, the Office of Nuclear Safety (ONS) of the Office of Assistant Secretary for Environmental Protection, Safety, and Emergency Preparedness has been assigned the task of reviewing the Overall Safety Manual (OSM) (memo from B.J. Rock to J.R. Maher, December 1, 1982). The OSM, dated July 1981 and in four volumes, was prepared by NUS Corporation, Rockville, Maryland, for the US Department of Energy. The OSM provides many of the technical models and much of the data which are used by (1) space launch contractors in safety analysis reports and (2) the broader Interagency Nuclear Safety Review Panel (INSRP) safety evaluation reports. If fhs interaction between the OSM, contractors, and INSRP is to work effectively, the OSM must be accurate, comprehensive, understandable, and usable.

  19. The use of individual and societal risk criteria within the Dutch flood safety policy--nationwide estimates of societal risk and policy applications.

    PubMed

    Jonkman, Sebastiaan N; Jongejan, Ruben; Maaskant, Bob

    2011-02-01

    The Dutch government is in the process of revising its flood safety policy. The current safety standards for flood defenses in the Netherlands are largely based on the outcomes of cost-benefit analyses. Loss of life has not been considered separately in the choice for current standards. This article presents the results of a research project that evaluated the potential roles of two risk metrics, individual and societal risk, to support decision making about new flood safety standards. These risk metrics are already used in the Dutch major hazards policy for the evaluation of risks to the public. Individual risk concerns the annual probability of death of a person. Societal risk concerns the probability of an event with many fatalities. Technical aspects of the use of individual and societal risk metrics in flood risk assessments as well as policy implications are discussed. Preliminary estimates of nationwide levels of societal risk are presented. Societal risk levels appear relatively high in the southwestern part of the country where densely populated dike rings are threatened by a combination of river and coastal floods. It was found that cumulation, the simultaneous flooding of multiple dike rings during a single flood event, has significant impact on the national level of societal risk. Options for the application of the individual and societal risk in the new flood safety policy are presented and discussed. PMID:20883529

  20. HTR fuel design, qualification and analyses at PBMR

    SciTech Connect

    Van Der Merwe, J. J.; Venter, J. H.

    2006-07-01

    This paper presents an overview of the safety and design requirements of PBMR fuel, design and performance analyses performed, analyses models and software being developed, and the current program to qualify PBMR fuel for use in the demonstration power plant. PBMR fuel design is based on the German reference fuel design, and will be utilised inside the operating envelope of the original German fuel qualification program. Fuel design, safety functions of the fuel, phenomena that influence fuel performance and fission product release and the design criteria derived from these functions and phenomena are described. Fuel qualification and validation of analyses methods are achieved by evaluations of previous experimental irradiation data and a fuel qualification programme for PBMR type fuel. The performed and planned validation and qualification efforts are presented with some results and issues discussed. The fuel performance analyses methods and legacy software products inherited from the German fuel program are being further developed at PBMR. New models and software are being developed as new requirements such as Monte Carlo design analyses become necessary. (authors)

  1. Overview of the Safety Issues Associated with the Compressed Natural Gas Fuel System and Electric Drive System in a Heavy Hybrid Electric Vehicle

    SciTech Connect

    Nelson, S.C.

    2002-11-14

    This report evaluates the hazards that are unique to a compressed-natural-gas (CNG)-fueled heavy hybrid electric vehicle (HEV) design compared with a conventional heavy vehicle. The unique design features of the heavy HEV are the CNG fuel system for the internal-combustion engine (ICE) and the electric drive system. This report addresses safety issues with the CNG fuel system and the electric drive system. Vehicles on U. S. highways have been propelled by ICEs for several decades. Heavy-duty vehicles have typically been fueled by diesel fuel, and light-duty vehicles have been fueled by gasoline. The hazards and risks posed by ICE vehicles are well understood and have been generally accepted by the public. The economy, durability, and safety of ICE vehicles have established a standard for other types of vehicles. Heavy-duty (i.e., heavy) HEVs have recently been introduced to U. S. roadways, and the hazards posed by these heavy HEVs can be compared with the hazards posed by ICE vehicles. The benefits of heavy HEV technology are based on their potential for reduced fuel consumption and lower exhaust emissions, while the disadvantages are the higher acquisition cost and the expected higher maintenance costs (i.e., battery packs). The heavy HEV is more suited for an urban drive cycle with stop-and-go driving conditions than for steady expressway speeds. With increasing highway congestion and the resulting increased idle time, the fuel consumption advantage for heavy HEVs (compared with conventional heavy vehicles) is enhanced by the HEVs' ability to shut down. Any increase in fuel cost obviously improves the economics of a heavy HEV. The propulsion system for a heavy HEV is more complex than the propulsion system for a conventional heavy vehicle. The heavy HEV evaluated in this study has in effect two propulsion systems: an ICE fueled by CNG and an electric drive system with additional complexity and failure modes. This additional equipment will result in a less

  2. Criteria for onsite transfers of radioactive material

    SciTech Connect

    Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

    1992-12-31

    A general description of the requirements for making onsite transfers of radioactive material is provided in Chapter 2, along with the required sequencey of activities. Various criteria for package use are identified in Chapters 3-13. These criteria provide protection against undue radiation exposure. Package shielding, containment, and surface contamination requirements are established. Criteria for providing criticality safety are enumerated in Chapter 6. Criteria for providing hazards information are established in Chapter 13. A glossary is provided.

  3. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  4. Fuel-Coolant-Interaction modeling and analysis work for the High Flux Isotope Reactor Safety Analysis Report

    SciTech Connect

    Taleyarkhan, R.P.; Georgevich, V.; Nestor, C.W.; Chang, S.J.; Freels, J.; Gat, U.; Lepard, B.L.; Gwaltney, R.C.; Luttrell, C.; Kirkpatrick, J.

    1993-07-01

    A brief historical background and a description of short- and long-term task plan development for effective closure of this important safety issue for the HFIR are given. Short-term aspects deal with Fuel-Coolant-Interaction (FCI) issues experimentation, modeling, and analysis for the flow-blockage-induced steam explosion events in direct support of the SAR. Long-term aspects deal with addressing FCI issues resulting from other accidents in conjunction with issues dealing with aluminum ignition, which can result in an order of magnitude increase in overall energetics. Problem formulation, modeling, and computer code simulation for the various phases of steam explosions are described. The evaluation of core melt initiation propagation, and melt superheat are described. Core melt initiation and propagation have been studied using simple conservative models as well as from modeling and analysis using RELAP5. Core debris coolability, heatup, and melting/freezing aspects have been studied by use of the two-dimensional melting/freezing analysis code 2DKO, which was also benchmarked with MELCOR code predictions. Descriptions are provided for the HM, BH, FCIMOD, and CTH computer codes that have been implemented for studying steam explosion energetics from the standpoint of evaluating bounding loads by thermodynamic models or best-estimate loads from one- and two-dimensional simulations of steam explosion energetics. Vessel failure modeling and analysis was conducted using the principles of probabilistic fracture mechanics in conjunction with ADINA code calculations. Top head bolts failure modeling has also been conducted where the failure criterion was based upon stresses in the bolts exceeding the material yield stress for a given time duration. Missile transport modeling and analysis was conducted by setting up a one-dimensional mathematical model that accounts for viscous dissipation, virtual mass effects, and material inertia.

  5. 30 CFR 104.3 - Pattern criteria.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... recurrence of violations of mandatory safety standards or health standards which significantly and substantially contribute to the cause and effect of mine safety or health hazards. These criteria are— (1) A... Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR PATTERN OF VIOLATIONS PATTERN...

  6. Highly distributed multi-point, temperature and pressure compensated, fiber optic oxygen sensors (FOxSense) for aircraft fuel tank environment and safety monitoring

    NASA Astrophysics Data System (ADS)

    Mendoza, Edgar A.; Kempen, Cornelia; Sun, Sunjian; Esterkin, Yan

    2014-09-01

    This paper describes recent progress towards the development and qualification of a highly distributed, multi-point, all optical pressure and temperature compensated, fiber optic oxygen sensor (FOxSense™) system for closed-loop monitoring and safety of the oxygen ullage environment inside fuel tanks of military and commercial aircraft. The alloptical FOxSense™ system uses a passive, multi-parameter (O2/T&P) fiber optic sensor probe with no electrical connections leading to the sensors install within the fuel tanks of an aircraft. The all optical sensor consists of an integrated multi-parameter fiber optic sensor probe that integrates a fuel insensitive fluorescence based optical oxygen optrode with built-in temperature and pressure optical optrodes for compensation of temperature and pressure variants induced in the fluorescence response of the oxygen optrode. The distributed (O2/T&P) fiber optic sensors installed in the fuel tanks of the aircraft are connected to the FOxSense optoelectronic system via a fiber optic cable conduit reaching to each fuel tank in the aircraft. A multichannel frequency-domain fiber optic sensor read-out (FOxSense™) system is used to interrogate the optical signal of all three sensors in real-time and to display the fuel tank oxygen environment suitable for aircraft status and alarm applications. Preliminary testing of the all optical fiber optic oxygen sensor have demonstrated the ability to monitor the oxygen environment inside a simulated fuel tank in the range of 0% O2 to 40% O2 concentrations, temperatures from (-) 40°C to (+) 60°C, and altitudes from 0-ft to 40,000-ft.

  7. Packaging design criteria for the MCO cask

    SciTech Connect

    Edwards, W.S.

    1996-09-11

    Approximately 2,100 metric tons of unprocessed, irradiated nuclear fuel elements are presently stored in the K Basins. To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the K Basins to a Canister Storage Building in the 200 East Area. The purpose of this packaging design criteria is to provide criteria for the design,fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multiple Canister Overpacks.

  8. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    SciTech Connect

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  9. High radon levels in subterranean environments: monitoring and technical criteria to ensure human safety (case of Castañar cave, Spain).

    PubMed

    Alvarez-Gallego, Miriam; Garcia-Anton, Elena; Fernandez-Cortes, Angel; Cuezva, Soledad; Sanchez-Moral, Sergio

    2015-07-01

    Castañar cave contains the highest radon gas ((222)Rn) concentration in Spain with an annual average of 31.9 kBq m(-)(3). Seasonal variations with summer minimums and maximum values in fall were recorded. The reduction of air-filled porosity of soil and rock by condensation or rainfalls hides the radon exchange by gas diffusion, determining this seasonal stair-step pattern of the radon activity concentration in underground air. The effective total dose and the maximum hours permitted have been evaluated for the guides and public safety with a highly detailed radon measurement along 2011 and 2012. A network of 12 passive detectors (kodalphas) has been installed, as well as, two radon continuous monitoring in the most interesting geological sites of the subterranean environment. A follow up of the recommended time (max. 50 min) inside the underground environment has been analysed since the reopen to public visitors for not surpassing the legal maximum effective dose for tourists and guides. Results shown that public visitors would receive in fall a 12.1% of the total effective dose permitted per visit, whereas in summer it is reduced to 8.6%, while the cave guide received a total effective dose of 6.41 mSv in four months. The spatial radon maps allow defining the most suitable touristic paths according to the radon concentration distribution and therefore, appropriate fall and summer touristic paths are recommended. PMID:25863322

  10. White paper on the proposed design, development, and implementation of a monitored retrievable storage module and the siting criteria for spent nuclear fuel

    SciTech Connect

    Villarreal, B.; Knobeloch, D.

    1996-01-01

    Congress enacted the Nuclear Waste Policy (NWP) Act in 1982 as comprehensive legislation for the DOE to locate, build, and operate repositories to permanently dispose of spent nuclear fuel and other high-level wastes. In 1987, Congress amended the NWP Act and authorized the DOE to site, construct, and operate one Monitored Retrievable Storage (MRS) facility. The MRS facility was planned as a means to enhance the flexibility and reliability of the overall waste management system. This white paper presents a broad prospectus of the scientific and regulatory capabilities at Los Alamos National Laboratory and outlines the methodology to design and implement an MRS test module. This proposed module will incorporate the flexibility to store all types of spent nuclear fuel above or below ground level and will be fully monitored for the residence time of the spent fuel in the MRS module. The purpose of this test module is to define the parameters necessary to build a simple and economical MRS system. Demonstration of the proposed MRS test module will be important because it will form the basis for an integrated MRS site model.

  11. Criticality Safety Evaluation Report CSER-96-019 for Spent Nuclear Fuel (SNF) Processing and Storage Facilities Multi Canister Overpack (MCO)

    SciTech Connect

    KESSLER, S.F.

    1999-10-20

    This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially built baskets containing a maximum of either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the cask-MCO, operation at the Cold Vacuum Drying Facility,a nd storage in the Canister Storage Building. Many conservatisms have been built into this analysis, the primary one being the selection of the K{sub eff} = 0.95 criticality safety limit. This revision incorporates the analyses for the sampling/weld station in the Canister Storage Building and additional analysis of the MCO during the draining at CVDF. Additional discussion of the scrap basket model was added to show why the addition of copper divider plates was not included in the models.

  12. Safety Assurance for ATR Irradiations

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL) is the world’s premiere test reactor for performing high fluence, large volume, irradiation test programs. The ATR has many capabilities and a wide variety of tests are performed in this truly one of a kind reactor, including isotope production, simple self-contained static capsule experiments, instrumented/controlled experiments, and loop testing under pressurized water conditions. Along with the five pressurized water loops, ATR may also have gas (temperature controlled) lead experiments, fuel boosted fast flux experiments, and static sealed capsules all in the core at the same time. In addition, any or all of these tests may contain fuel or moderating materials that can affect reactivity levels in the ATR core. Therefore the safety analyses required to ensure safe operation of each experiment as well as the reactor itself are complex. Each test has to be evaluated against stringent reactor control safety criteria, as well as the effects it could have on adjacent tests and the reactor as well as the consequences of those effects. The safety analyses of each experiment are summarized in a document entitled the Experiment Safety Assurance Package (ESAP). The ESAP references and employs the results of the reactor physics, thermal, hydraulic, stress, seismic, vibration, and all other analyses necessary to ensure the experiment can be irradiated safely in the ATR. The requirements for reactivity worth, chemistry compatibilities, pressure limitations, material issues, etc. are all specified in the Technical Safety Requirements and the Upgraded Final Safety Analysis Report (UFSAR) for the ATR. This paper discusses the ESAP process, types of analyses, types of safety requirements and the approvals necessary to ensure an experiment can be safely irradiated in the ATR.

  13. Industrial Fuel Gas Demonstration Plant Program. Task III, Demonstration plant safety, industrial hygiene, and major disaster plan (Deliverable No. 35)

    SciTech Connect

    1980-03-01

    This Health and Safety Plan has been adopted by the IFG Demonstration Plant managed by Memphis Light, Gas and Water at Memphis, Tennessee. The plan encompasses the following areas of concern: Safety Plan Administration, Industrial Health, Industrial Safety, First Aid, Fire Protection (including fire prevention and control), and Control of Safety Related Losses. The primary objective of this plan is to achieve adequate control of all potentially hazardous activities to assure the health and safety of all employees and eliminate lost work time to both the employees and the company. The second objective is to achieve compliance with all Federal, state and local laws, regulations and codes. Some thirty specific safe practice instruction items are included.

  14. Application of Neutron-Absorbing Structural-Amorphous Metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Control

    SciTech Connect

    Choi, J

    2007-01-12

    This report describes the analysis and modeling approaches used in the evaluation for criticality-control applications of the neutron-absorbing structural-amorphous metal (SAM) coatings. The applications of boron-containing high-performance corrosion-resistant material (HPCRM)--amorphous metal as the neutron-absorbing coatings to the metallic support structure can enhance criticality safety controls for spent nuclear fuel in baskets inside storage containers, transportation casks, and disposal containers. The use of these advanced iron-based, corrosion-resistant materials to prevent nuclear criticality in transportation, aging, and disposal containers would be extremely beneficial to the nuclear waste management programs.

  15. 10 CFR 830.122 - Quality assurance criteria.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Quality Assurance Requirements § 830.122 Quality assurance criteria. The QAP must address the following management, performance, and assessment criteria: (a) Criterion 1—Management/Program. (1) Establish an organizational structure, functional...

  16. 10 CFR 830.122 - Quality assurance criteria.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Quality Assurance Requirements § 830.122 Quality assurance criteria. The QAP must address the following management, performance, and assessment criteria: (a) Criterion 1—Management/Program. (1) Establish an organizational structure, functional...

  17. Crew Transportation Technical Standards and Design Evaluation Criteria

    NASA Technical Reports Server (NTRS)

    Lueders, Kathryn L.; Thomas, Rayelle E. (Compiler)

    2015-01-01

    Crew Transportation Technical Standards and Design Evaluation Criteria contains descriptions of technical, safety, and crew health medical processes and specifications, and the criteria which will be used to evaluate the acceptability of the Commercial Providers' proposed processes and specifications.

  18. Launch Commit Criteria Monitoring Agent

    NASA Technical Reports Server (NTRS)

    Semmel, Glenn S.; Davis, Steven R.; Leucht, Kurt W.; Rowe, Dan A.; Kelly, Andrew O.; Boeloeni, Ladislau

    2005-01-01

    The Spaceport Processing Systems Branch at NASA Kennedy Space Center has developed and deployed a software agent to monitor the Space Shuttle's ground processing telemetry stream. The application, the Launch Commit Criteria Monitoring Agent, increases situational awareness for system and hardware engineers during Shuttle launch countdown. The agent provides autonomous monitoring of the telemetry stream, automatically alerts system engineers when predefined criteria have been met, identifies limit warnings and violations of launch commit criteria, aids Shuttle engineers through troubleshooting procedures, and provides additional insight to verify appropriate troubleshooting of problems by contractors. The agent has successfully detected launch commit criteria warnings and violations on a simulated playback data stream. Efficiency and safety are improved through increased automation.

  19. Disposing of Canada's used fuel

    SciTech Connect

    Torgerson, D.F.

    1990-01-01

    The Canadian Nuclear Fuel Waste Management Program is assessing the permanent disposal of used nuclear fuel in a waste vault located 500 to 1,000 m deep in the Precambrian granitic rock of the Canadian Shield. The specific objectives of the program are to develop and demonstrate the technology to site, design, build, and operate a disposal facility in a way that creates no, or negligible, burden on future generations. In addition, the program must develop a methodology to evaluate the performance of the disposal system against safety criteria and demonstrate that sites are likely to exist in the Canadian Shield that satisfy regulatory criteria. These criteria are very stringent. As in other national high-level waste management programs, the Canadian concept for the permanent disposal of nuclear fuel wastes employs a multiple barrier system for isolating contaminants from the environment. The current phase of the work is generic in nature and is not site specific. Research and development (R and D) has advanced to the point where the generic concept will be evaluated under the Canadian environmental assessment review process, which involves public hearings and independent scientific review.

  20. Environmental Impacts, Health and Safety Impacts, and Financial Costs of the Front End of the Nuclear Fuel Cycle

    SciTech Connect

    Brett W Carlsen; Urairisa Phathanapirom; Eric Schneider; John S. Collins; Roderick G. Eggert; Brett Jordan; Bethany L. Smith; Timothy M. Ault; Alan G. Croff; Steven L. Krahn; William G. Halsey; Mark Sutton; Clay E. Easterly; Ryan P. Manger; C. Wilson McGinn; Stephen E. Fisher; Brent W. Dixon; Latif Yacout

    2013-07-01

    FEFC processes, unlike many of the proposed fuel cycles and technologies under consideration, involve mature operational processes presently in use at a number of facilities worldwide. This report identifies significant impacts resulting from these current FEFC processes and activities. Impacts considered to be significant are those that may be helpful in differentiating between fuel cycle performance and for which the FEFC impact is not negligible relative to those from the remainder of the full fuel cycle. This report: • Defines ‘representative’ processes that typify impacts associated with each step of the FEFC, • Establishes a framework and architecture for rolling up impacts into normalized measures that can be scaled to quantify their contribution to the total impacts associated with various fuel cycles, and • Develops and documents the bases for estimates of the impacts and costs associated with each of the representative FEFC processes.

  1. 47 CFR 90.713 - Entry criteria.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false Entry criteria. 90.713 Section 90.713 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES PRIVATE LAND MOBILE RADIO SERVICES Regulations Governing Licensing and Use of Frequencies in the 220-222 MHz Band § 90.713 Entry criteria. (a) As set forth in...

  2. Design criteria for SW-205 capillary system

    SciTech Connect

    Lyons, W.J.

    1989-04-01

    This design criteria covers the converting of the SW-250 Capillary System from fumehood manual operation to sealed glovebox automated operation. The design criteria contains general guidelines and includes drawings reflecting a similar installation at another site. Topics include purpose and physical description, architectural-engineering requirements, reference document, electrical, fire protection, occupational safety and health, quality assurance, and security.

  3. 46 CFR 154.471 - Design criteria.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 5 2012-10-01 2012-10-01 false Design criteria. 154.471 Section 154.471 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR SELF-PROPELLED VESSELS CARRYING BULK LIQUEFIED GASES Design, Construction and Equipment Support System § 154.471 Design criteria. (a) The...

  4. Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. 2. Reentrainment and discharge of radioactive materials

    SciTech Connect

    Davis, W Jr

    1981-07-01

    This report describes results of a parametric study of quantities of radioactive materials that might be discharged by a tornado-generated depressurization on contaminated process cells within the presently inoperative Nuclear Fuel Services' (NFS) fuel reprocessing facility near West Valley, New York. The study involved the following tasks: determining approximate quantities of radioactive materials in the cells and characterizing particle-size distribution; estimating the degree of mass reentrainment from particle-size distribution and from air speed data presented in Part 1; and estimating the quantities of radioactive material (source term) released from the cells to the atmosphere. The study has shown that improperly sealed manipulator ports in the Process Mechanical Cell (PMC) present the most likely pathway for release of substantial quantities of radioactive material in the atmosphere under tornado accident conditions at the facility.

  5. Spent nuclear fuel project product specification

    SciTech Connect

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  6. 78 FR 4830 - National Advisory Committee on Microbiological Criteria for Foods; Reestablishment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-23

    ... Food Safety and Inspection Service National Advisory Committee on Microbiological Criteria for Foods; Reestablishment AGENCY: Food Safety and Inspection Service, USDA. ACTION: Notice of reestablishment of Committee... reestablishment of the National Advisory Committee on Microbiological Criteria for Foods (NACMCF). The...

  7. Answering Key Fuel Cycle Questions

    SciTech Connect

    Piet, S.J.; Dixon, B.W.; Bennett, R.G.; Smith, J.D.; Hill, R.N.

    2004-10-03

    Given the range of fuel cycle goals and criteria, and the wide range of fuel cycle options, how can the set of options eventually be narrowed in a transparent and justifiable fashion? It is impractical to develop all options. We suggest an approach that starts by considering a range of goals for the Advanced Fuel Cycle Initiative (AFCI) and then posits seven questions, such as whether Cs and Sr isotopes should be separated from spent fuel and, if so, what should be done with them. For each question, we consider which of the goals may be relevant to eventually providing answers. The AFCI program has both ''outcome'' and ''process'' goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geologic repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are rea diness to proceed and adaptability and robustness in the face of uncertainties.

  8. 16 CFR 1301.5 - Banning criteria.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Banning criteria. 1301.5 Section 1301.5 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT REGULATIONS BAN OF... being distributed in commerce on or after the effective date of this ban and which has an...

  9. Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. Volume I. Tornado effects on head-end cell airflow

    SciTech Connect

    Holloway, L.J.; Andrae, R.W.

    1981-09-01

    This report describes results of a parametric study of the impacts of a tornado-generated depressurization on airflow in the contaminated process cells within the presently inoperative Nuclear Fuel Services fuel reprocessing facility near West Valley, NY. The study involved the following tasks: (1) mathematical modeling of installed ventilation and abnormal exhaust pathways from the cells and prediction of tornado-induced airflows in these pathways; (2) mathematical modeling of individual cell flow characteristics and prediction of in-cell velocities induced by flows from step 1; and (3) evaluation of the results of steps 1 and 2 to determine whether any of the pathways investigated have the potential for releasing quantities of radioactively contaminated air from the main process cells. The study has concluded that in the event of a tornado strike, certain pathways from the cells have the potential to release radioactive materials of the atmosphere. Determination of the quantities of radioactive material released from the cells through pathways identified in step 3 is presented in Part II of this report.

  10. Packaging design criteria for the K east basin sludge transportation system

    SciTech Connect

    Tomaszewski, T.A., Westinghouse Hanford

    1996-07-11

    This packaging design criteria (PDC) establishes the onsite transportation safety criteria for a reusable packaging and transport system to transport K East Basin sludge and water.This PDC provides the basis for the development of a safety analysis report for packaging; establishes the packaging contents and safety class of the package; and provides design criteria for the package, packaging, and transport systems.

  11. Safety review of the design, operation, and radiation sections of the General Electric Morris Operation Consolidated Safety Analysis Report

    SciTech Connect

    McBride, J.P.

    1981-01-30

    A safety review was made of Sections 4 through 9 of the Consolidated Safety Analysis Report (CSAR) for the GE Morris Operation spent-fuel storage facility. The sections reviewed include Design Criteria and Compliance, Facility Design and Description, Radiation Protection, Accident Analysis, and Conduct of Operations. The safety review was performed in accordance with the Code of Federal Regulations, Title 10, Part 72, ''Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation'' and contains independent estimations of source terms and dose-commitments from postulated accidents in the storage facility and a structural analysis of the Morris Operation cranes as an appendix. The review confirms that the features of the facility as described in Sections 4 through 9 of the CSAR fulfilled the safety requirements of 10 CFR 72, and it is concluded that spent-fuel handling and storage at the Morris Operation do not present significant risks to public health and safety. 15 refs., 3 tabs.

  12. State of Washington Department of Ecology criteria pollutants and toxic air pollutants phase 1 notice of construction for the Hanford site spent nuclear fuel project - hot conditioning system annex, project W-484

    SciTech Connect

    Turnbaugh, J.E.

    1996-08-15

    This notice of construction (NOC) provides information regarding the source and the air toxic and criteria pollutants resulting from operation of the Hot: Conditioning System Annex (HCSA). Additional details on emissions generated by the operation of the HCSA will be, discussed again in the Phase 11 NOC. This Phase I NOC is defined as, constructing the substructure, including but not limited to pouring the concrete for the floor; construction of the process pits and `exterior walls; making necessary interface connections to the Canister Storage Building (CSB) ventilation and utility systems for personnel comfort; and extending the multi-canister overpack (MCO) handling machine rails into the HCSA. A Phase 11 NOC, will be submitted for approval prior to installing and is defined as the completion of the HCSA, which will consist of installation of the Hot Conditioning System Equipment (HCSE), air emissions control equipment and emissions monitoring equipment. About 80 percent of the !U.S. Department of Energy`s spent nuclear fuel (SNF) inventory is stored under water in the Hanford Site K Basins; spent nuclear fuel in the K West Basin is contained in closed canisters, while the SNF in the K East Basin is contained in open canisters, which allow free release of corrosion products to the K Basin water. Storage in the K Basins was `originally intended to be on an as-needed basis to sustain operation of the N Reactor while the Plutonium-Uranium Extraction (PUREX) Plant was refurbished and restarted. The decision in December 1992 to deactivate the PUREX Plant left approximately 2,300 MT (2,530 tons) of N Reactor SNF in the K Basins with no means for near-term removal and processing.

  13. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    SciTech Connect

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical.

  14. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory. Part 2, Chemical constituents

    SciTech Connect

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical.

  15. Packaging Design Criteria for the MCO Cask

    SciTech Connect

    FLANAGAN, B.D.

    2000-08-01

    Approximately 2,100 metric tons of unprocessed, irradiated, nuclear fuel elements are presently stored in the K Basins (including approximately 700 additional elements from the Plutonium-Uranium Extraction Plant, N Reactor, and 327 Laboratory). To permit cleanup of the K Basins and fuel conditioning, the fuel will be transported from the 100 K Area to a Canister Storage Building (CSB) in the 200 East Area. The purpose of this packaging design criteria is to provide criteria for the design, fabrication, and use of a packaging system to transport the large quantities of irradiated nuclear fuel elements positioned within Multi-canister Overpacks. Concurrent with the K Basin cleanup, 72 Shippingport Pressurized Water Reactor Core 2 fuel assemblies will be transported from T Plant to the CSB to provide space at T Plant for K Basin sludge canisters.

  16. Estimating Source Terms for Diverse Spent Nuclear Fuel Types

    SciTech Connect

    Brett Carlsen; Layne Pincock

    2004-11-01

    The U.S. Department of Energy (DOE) National Spent Nuclear Fuel Program is responsible for developing a defensible methodology for determining the radionuclide inventory for the DOE spent nuclear fuel (SNF) to be dispositioned at the proposed Monitored Geologic Repository at the Yucca Mountain Site. SNF owned by DOE includes diverse fuels from various experimental, research, and production reactors. These fuels currently reside at several DOE sites, universities, and foreign research reactor sites. Safe storage, transportation, and ultimate disposal of these fuels will require radiological source terms as inputs to safety analyses that support design and licensing of the necessary equipment and facilities. This paper summarizes the methodology developed for estimating radionuclide inventories associated with DOE-owned SNF. The results will support development of design and administrative controls to manage radiological risks and may later be used to demonstrate conformance with repository acceptance criteria.

  17. Packaging design criteria for the Hanford Ecorok Packaging

    SciTech Connect

    Mercado, M.S.

    1996-01-19

    The Hanford Ecorok Packaging (HEP) will be used to ship contaminated water purification filters from K Basins to the Central Waste Complex. This packaging design criteria documents the design of the HEP, its intended use, and the transportation safety criteria it is required to meet. This information will serve as a basis for the safety analysis report for packaging.

  18. 29 CFR 1904.7 - General recording criteria.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 5 2010-07-01 2010-07-01 false General recording criteria. 1904.7 Section 1904.7 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RECORDING AND REPORTING OCCUPATIONAL INJURIES AND ILLNESSES Recordkeeping Forms and Recording Criteria § 1904.7 General recording criteria....

  19. 29 CFR 1904.7 - General recording criteria.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 29 Labor 5 2012-07-01 2012-07-01 false General recording criteria. 1904.7 Section 1904.7 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RECORDING AND REPORTING OCCUPATIONAL INJURIES AND ILLNESSES Recordkeeping Forms and Recording Criteria § 1904.7 General recording criteria....

  20. 29 CFR 1904.7 - General recording criteria.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 29 Labor 5 2011-07-01 2011-07-01 false General recording criteria. 1904.7 Section 1904.7 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RECORDING AND REPORTING OCCUPATIONAL INJURIES AND ILLNESSES Recordkeeping Forms and Recording Criteria § 1904.7 General recording criteria....

  1. 49 CFR 538.8 - Gallon Equivalents for Gaseous Fuels.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION MANUFACTURING INCENTIVES FOR ALTERNATIVE FUEL... Measurements for Gaseous Fuels per 100 Standard Cubic Feet Fuel Gallon equivalent measurement...

  2. 49 CFR 538.8 - Gallon Equivalents for Gaseous Fuels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION MANUFACTURING INCENTIVES FOR ALTERNATIVE FUEL... Measurements for Gaseous Fuels per 100 Standard Cubic Feet Fuel Gallon equivalent measurement...

  3. 49 CFR 538.8 - Gallon Equivalents for Gaseous Fuels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION MANUFACTURING INCENTIVES FOR ALTERNATIVE FUEL... Measurements for Gaseous Fuels per 100 Standard Cubic Feet Fuel Gallon equivalent measurement...

  4. 49 CFR 538.8 - Gallon Equivalents for Gaseous Fuels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION MANUFACTURING INCENTIVES FOR ALTERNATIVE FUEL... Measurements for Gaseous Fuels per 100 Standard Cubic Feet Fuel Gallon equivalent measurement...

  5. 49 CFR 538.8 - Gallon Equivalents for Gaseous Fuels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION MANUFACTURING INCENTIVES FOR ALTERNATIVE FUEL... Measurements for Gaseous Fuels per 100 Standard Cubic Feet Fuel Gallon equivalent measurement...

  6. 49 CFR 238.423 - Fuel tanks.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 4 2012-10-01 2012-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at...

  7. 49 CFR 238.423 - Fuel tanks.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at...

  8. 49 CFR 238.423 - Fuel tanks.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 4 2011-10-01 2011-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at...

  9. 49 CFR 238.423 - Fuel tanks.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 4 2013-10-01 2013-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at...

  10. 49 CFR 238.423 - Fuel tanks.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 4 2014-10-01 2014-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at...

  11. Hanford Site Solid Waste Acceptance Criteria

    SciTech Connect

    MCDOWELL, A.K.; TRINER, G.C.

    2002-03-28

    DOE Order 435.1 requires that each treatment, storage, and/or disposal facility (referred to in this document as TSD unit) that manages low-level or transuranic (TRU) waste (including mixed waste and TSCA PCB waste) maintain waste acceptance criteria. These criteria must address the various requirements to operate the TSD unit in compliance with applicable safety and environmental requirements. This document sets forth the baseline criteria for acceptance of waste at TSD units operated by WMP. The criteria for each TSD unit have been established to ensure that waste accepted can be managed in a manner that is within the operating requirements of the unit, including environmental regulations, DOE Orders, permits, technical safety requirements, waste analysis plans, performance assessments, and other applicable requirements. Revisions to the acceptance criteria document require an Unreviewed Safety Question review to document that the changes are consistent with current applicable safety analysis. Acceptance criteria apply to the following TSD units: the Low-Level Burial Grounds (LLBG) including both the nonregulated portions of the LLBG and trenches 31 and 34 of the 218-W-5 Burial Ground for mixed waste disposal; Central Waste Complex (CWC); Waste Receiving and Processing (WRAP) Facility; and T-Plant facility. Waste from all generators, both from the Hanford Site and from offsite facilities, must comply with these criteria. Exceptions can be granted as provided in Section 1.6. Specific waste streams could have additional requirements based on the identified TSD pathway. These requirements are communicated in the waste specification records (WSRds) and/or waste stream profile sheet approvals. The Hanford Site manages nonradioactive waste through direct shipments to offsite contractors. The waste acceptance requirements of the offsite TSD facility must be met for these nonradioactive wastes. This document does not address the acceptance requirements of these offsite

  12. Strength criteria for composite materials (a literature survey)

    NASA Technical Reports Server (NTRS)

    Roode, F.

    1982-01-01

    Literature concerning strength (failure) criteria for composite materials is reviewed with emphasis on phenomenological failure criteria. These criteria are primarily intended to give a good estimation of the safety margin with respect to failure for arbitrary multiaxial stress states. The failure criteria do not indicate the types of fracture that will occur in the material. The collection of failure criteria is evaluated for applicability for the glass reinforced plastics used in mine detectors. Material tests necessary to determine the parameters in the failure criteria are discussed.

  13. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  14. Hand Safety

    MedlinePlus

    ... en gatillo See More... Hand Anatomy Hand Safety Fireworks Safety Lawnmower Safety Snowblower safety Pumpkin Carving Gardening ... en gatillo See More... Hand Anatomy Hand Safety Fireworks Safety Lawnmower Safety Snowblower safety Pumpkin Carving Gardening ...

  15. Hand Safety

    MedlinePlus

    ... Fireworks Safety Lawnmower Safety Snowblower safety Pumpkin Carving Gardening Safety Turkey Carving Removing a Ring Español Artritis ... Fireworks Safety Lawnmower Safety Snowblower safety Pumpkin Carving Gardening Safety Turkey Carving Removing a Ring Español Artritis ...

  16. Safety and Performance Advantages of Nitrous Oxide Fuel Blends (NOFBX) Propellants for Manned and Unmanned Spaceflight Applications

    NASA Astrophysics Data System (ADS)

    Taylor, R.

    2012-01-01

    Hydrazine, N2H4, is the current workhorse monopropellant in the spacecraft industry. Although widely used since the 1960's, hydrazine is highly toxic and its specific impulse (ISP) performance of ~230s is far lower than bipropellants and solid motors. NOFBX™ monopropellants were originally developed under NASA's Mars Advanced Technology program (2004-2007) for deep space Mars missions. This work focused on characterizing various Nitrous Oxide Fuel Blend (NOFB) monopropellants which exhibited many favorable attributes to include: (1) Mono-propulsion, (2) Isp > 320s, (3) Non-toxic constituents, (4) Non-toxic effluents, (5) Low Cost, (6) High Density Specific Impulse, (7) Non-cryogenic, (8) Wide Storable Temperature Range, (9) Deeply throttlable [between 5 - 100lbs], (10) Self Pressurizing, (11) Wide Range of materials compatibility, along with many, many other benefits. All rocket propellants carry with them a history or stigma associated with either the development or implementation of that propellant and NOFBX™ is no exception. This paper examines the benefits of NOFBX™ propellants while addressing or dispelling a number of critiques N2O based propellants acquired through the decades of rocket propellant testing.

  17. Standard format and content for the safety analysis report for an independent spent fuel storage installation or monitored retrievable storage installation (dry storage): Revision 1, Task CE 406-4

    SciTech Connect

    Not Available

    1989-08-01

    Part 72, ''Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,'' of Title 10 of the Code of Federal Regulations specifies the information to be supplied in applications for licenses to store spent fuel in an independent spent fuel storage installation (ISFSI) or to store spent fuel and high-level radioactive waste in a monitored retrievable storage (MRS) installation. However, Part 72 does not specify the format for presentation of the safety analysis report (SAR). Guidance on the content of the SAR will vary, depending on the type of installation that is planned. This guide represents a Standard Format that is acceptable to the NRC staff for the SAR required for the license application. Conformance with this Standard Format, however, is not mandatory. License applications with differing SAF formats will be acceptable to the staff if they provide an adequate basis for the findings required for the issuance of a license.

  18. 30 CFR 75.1403-7 - Criteria-Mantrips.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Criteria-Mantrips. 75.1403-7 Section 75.1403-7 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Hoisting and Mantrips § 75.1403-7 Criteria—Mantrips. (a) Mantrips should be...

  19. 30 CFR 75.1403-4 - Criteria-Automatic elevators.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Criteria-Automatic elevators. 75.1403-4 Section 75.1403-4 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Hoisting and Mantrips § 75.1403-4 Criteria—Automatic elevators. (a) The doors...

  20. 30 CFR 75.1403-9 - Criteria-Shelter holes.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Criteria-Shelter holes. 75.1403-9 Section 75.1403-9 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Hoisting and Mantrips § 75.1403-9 Criteria—Shelter holes. (a) Shelter holes should...

  1. 30 CFR 75.1403-9 - Criteria-Shelter holes.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Criteria-Shelter holes. 75.1403-9 Section 75.1403-9 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Hoisting and Mantrips § 75.1403-9 Criteria—Shelter holes. (a) Shelter holes should...

  2. 30 CFR 75.1403-9 - Criteria-Shelter holes.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Criteria-Shelter holes. 75.1403-9 Section 75.1403-9 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Hoisting and Mantrips § 75.1403-9 Criteria—Shelter holes. (a) Shelter holes should...

  3. Nutrient Criteria Research

    EPA Science Inventory

    EPA has developed methodologies for deriving nutrient criteria, default criteria for the variety of waters and eco-regions found in the U.S., and a strategy for implementing the criteria including guidance on the use and development of biocriteria. Whereas preliminary research ha...

  4. A statistical approach to nuclear fuel design and performance

    NASA Astrophysics Data System (ADS)

    Cunning, Travis Andrew

    As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

  5. K Basin safety analysis

    SciTech Connect

    Porten, D.R.; Crowe, R.D.

    1994-12-16

    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.

  6. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    SciTech Connect

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report describes health and safety concerns associated with the Mixed and Low-level Waste Treatment Facility at the Idaho National Engineering Laboratory. Various hazards are described such as fire, electrical, explosions, reactivity, temperature, and radiation hazards, as well as the potential for accidental spills, exposure to toxic materials, and other general safety concerns.

  7. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory. Part 1, Waste streams and treatment technologies

    SciTech Connect

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report describes health and safety concerns associated with the Mixed and Low-level Waste Treatment Facility at the Idaho National Engineering Laboratory. Various hazards are described such as fire, electrical, explosions, reactivity, temperature, and radiation hazards, as well as the potential for accidental spills, exposure to toxic materials, and other general safety concerns.

  8. DIPS Space Exploration Initiative safety

    NASA Astrophysics Data System (ADS)

    Dix, Terry E.

    The Dynamic Isotope Power Subsystem has been identified for potential applications for the Space Exploration Initiative. A qualitative safety assessment has been performed to demonstrate the overall safety adequacy of the Dynamic Isotope Power Subsystem for these applications. Mission profiles were defined for reference lunar and Martian flights. Accident scenarios were qualitatively defined for all mission phases. Safety issue were then identified. The safety issues included radiation exposure, fuel containment, criticality, diversion, toxic materials, heat flux to the extravehicular mobility unit, and disposal. The design was reviewed for areas where safety might be further improved. Safety would be improved by launching the fuel separate from the rest of the subsystem on expendable launch vehicles, using a fuel handling tool during unloading of the hot fuel canister, and constructing a cage-like structure around the reversible heat removal system lithium heat pipes. The results of the safety assessment indicate that the DIPS design with minor modifications will produce a low risk concept.

  9. Analysis of the Relationship Between Vehicle Weight/Size and Safety, and Implications for Federal Fuel Economy Regulation

    SciTech Connect

    Wenzel, Thomas P.

    2010-03-02

    This report analyzes the relationship between vehicle weight, size (wheelbase, track width, and their product, footprint), and safety, for individual vehicle makes and models. Vehicle weight and footprint are correlated with a correlation coefficient (R{sup 2}) of about 0.62. The relationship is stronger for cars (0.69) than for light trucks (0.42); light trucks include minivans, fullsize vans, truck-based SUVs, crossover SUVs, and pickup trucks. The correlation between wheelbase and track width, the components of footprint, is about 0.61 for all light vehicles, 0.62 for cars and 0.48 for light trucks. However, the footprint data used in this analysis does not vary for different versions of the same vehicle model, as curb weight does; the analysis could be improved with more precise data on footprint for different versions of the same vehicle model. Although US fatality risk to drivers (driver fatalities per million registered vehicles) decreases as vehicle footprint increases, there is very little correlation either for all light vehicles (0.01), or cars (0.07) or trucks (0.11). The correlation between footprint and fatality risks cars impose on drivers of other vehicles is also very low (0.01); for trucks the correlation is higher (0.30), with risk to others increasing as truck footprint increases. Fatality risks reported here do not account for differences in annual miles driven, driver age or gender, or crash location by vehicle type or model. It is difficult to account for these factors using data on national fatal crashes because the number of vehicles registered to, for instance, young males in urban areas is not readily available by vehicle type or model. State data on all police-reported crashes can be used to estimate casualty risks that account for miles driven, driver age and gender, and crash location. The number of vehicles involved in a crash can act as a proxy of the number of miles a given vehicle type, or model, is driven per year, and is a

  10. Design criteria document, electrical system, K-Basin essential systems recovery, Project W-405

    SciTech Connect

    Hoyle, J.R.

    1994-12-12

    This Design Criteria Document provides the criteria for design and construction of electrical system modifications for 100K Area that are essential to protect the safe operation and storage of spent nuclear fuel in the K-Basin facilities.

  11. 50 CFR 665.965 - Fishing permit procedures and criteria.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Rose Atoll Marine National Monument § 665.965 Fishing permit procedures and criteria. (a) Rose Atoll... expenses, including but not limited to ice, bait, fuel, or food. (b) Rose Atoll Monument...

  12. 50 CFR 665.965 - Fishing permit procedures and criteria.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Rose Atoll Marine National Monument § 665.965 Fishing permit procedures and criteria. (a) Rose Atoll... expenses, including but not limited to ice, bait, fuel, or food. (b) Rose Atoll Monument...

  13. Revision of testing criteria for air cleaning unit of renovated APR-1000 and APR-1400 NPPS.

    PubMed

    Jeong, Seung-Young

    2011-07-01

    Designing Air Cleaning Units (ACU) of an Engineered Safety Feature and normal atmosphere clean-up system at the renovated APR-1000 and APR-1400 NPP, and fuel cycle facilities in Korea, is required to meet the standards of ASME AG-1 (1997), ASME N509/N510 (1989) and KEPIC-MH (2001) to enhance the removal efficiency of aerosols and particulates from the effluents. The revised ACU testing criteria are allowed to use alternative challenge agents of the dioctyl phthalate and Refrigerant-11 for in situ testing of high efficiency particulate air filters and adsorption banks. The operability testing time of engineered safety feature (ESF) trains was changed from 10 h to 15 min. The activated carbon in adsorption banks should undergo laboratory tests at a temperature of 30 °C and relative humidity 95 %. The removal criteria of methyl iodide should be over 99.5 % for ESF and 99 % for normal systems. This paper provides the background of the changed criteria for designing and testing of the ACU system in nuclear facilities. PMID:21502294

  14. Costs and benefits of automotive fuel economy improvement: A partial analysis

    SciTech Connect

    Greene, D.L.; Duleep, K.G.

    1992-03-01

    This paper is an exercise in estimating the costs and benefits of technology-based fuel economy improvements for automobiles and light trucks. Benefits quantified include vehicle cots, fuel savings, consumer`s surplus effects, the effect of reduced weight on vehicle safety, impacts on emissions of CO{sub 2} and criteria pollutants, world oil market and energy security benefits, and the transfer of wealth from US consumes to oil producers. A vehicle stock model is used to capture sales, scrappage, and vehicle use effects under three fuel price scenarios. Three alternative fuel economy levels for 2001 are considered, ranging from 32.9 to 36.5 MPG for cars and 24.2 to 27.5 MPG for light trucks. Fuel economy improvements of this size are probably cost-effective. The size of the benefit, and whether there is a benefit, strongly depends on the financial costs of fuel economy improvement and judgments about the values of energy security, emissions, safety, etc. Three sets of values for eight parameters are used to define the sensitivity of costs and benefits to key assumptions. The net present social value (1989$) of costs and benefits ranges from a cost of $11 billion to a benefit of $286 billion. The critical parameters being the discount rate (10% vs. 3%) and the values attached to externalities. The two largest components are always the direct vehicle costs and fuel savings, but these tend to counterbalance each other for the fuel economy levels examined here. Other components are the wealth transfer, oil cost savings, CO{sub 2} emissions reductions, and energy security benefits. Safety impacts, emissions of criteria pollutants, and consumer`s surplus effects are relatively minor components. The critical issues for automotive fuel economy are therefore: (1) the value of present versus future costs and benefits, (2) the values of external costs and benefits, and (3) the financially cost-effective level of MPG achievable by available technology. 53 refs.

  15. Costs and benefits of automotive fuel economy improvement: A partial analysis

    SciTech Connect

    Greene, D.L. ); Duleep, K.G. )

    1992-03-01

    This paper is an exercise in estimating the costs and benefits of technology-based fuel economy improvements for automobiles and light trucks. Benefits quantified include vehicle cots, fuel savings, consumer's surplus effects, the effect of reduced weight on vehicle safety, impacts on emissions of CO{sub 2} and criteria pollutants, world oil market and energy security benefits, and the transfer of wealth from US consumes to oil producers. A vehicle stock model is used to capture sales, scrappage, and vehicle use effects under three fuel price scenarios. Three alternative fuel economy levels for 2001 are considered, ranging from 32.9 to 36.5 MPG for cars and 24.2 to 27.5 MPG for light trucks. Fuel economy improvements of this size are probably cost-effective. The size of the benefit, and whether there is a benefit, strongly depends on the financial costs of fuel economy improvement and judgments about the values of energy security, emissions, safety, etc. Three sets of values for eight parameters are used to define the sensitivity of costs and benefits to key assumptions. The net present social value (1989$) of costs and benefits ranges from a cost of $11 billion to a benefit of $286 billion. The critical parameters being the discount rate (10% vs. 3%) and the values attached to externalities. The two largest components are always the direct vehicle costs and fuel savings, but these tend to counterbalance each other for the fuel economy levels examined here. Other components are the wealth transfer, oil cost savings, CO{sub 2} emissions reductions, and energy security benefits. Safety impacts, emissions of criteria pollutants, and consumer's surplus effects are relatively minor components. The critical issues for automotive fuel economy are therefore: (1) the value of present versus future costs and benefits, (2) the values of external costs and benefits, and (3) the financially cost-effective level of MPG achievable by available technology. 53 refs.

  16. 27 CFR 19.911 - Criteria for issuance of permit.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... Permits § 19.911 Criteria for issuance of permit. In general, an alcohol fuel producer's permit will be... applications are set forth in 27 CFR Part 71. (Sec. 201, Pub. L. 85-859, 72 Stat. 1370, as amended (26 U.S.C... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Criteria for issuance...

  17. Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

    SciTech Connect

    Dr. Michael A. Lehto; MAL

    2007-05-01

    A recent accomplishment of the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Nuclear Safety analysis group was to obtain DOE-ID approval for the inter-facility transfer of greater-than-Hazard-Category-3 quantity radioactive/fissionable waste in Department of Transportation (DOT) Type A drums at MFC. This accomplishment supported excellence in operations through safety analysis by better integrating nuclear safety requirements with waste requirements in the Transportation Safety Document (TSD); reducing container and transport costs; and making facility operations more efficient. The MFC TSD governs and controls the inter-facility transfer of greater-than-Hazard-Category-3 radioactive and/or fissionable materials in non-DOT approved containers. Previously, the TSD did not include the capability to transfer payloads of greater-than-Hazard-Category-3 radioactive and/or fissionable materials using DOT Type A drums. Previous practice was to package the waste materials to less-than-Hazard-Category-3 quantities when loading DOT Type A drums for transfer out of facilities to reduce facility waste accumulations. This practice allowed operations to proceed, but resulted in drums being loaded to less than the Waste Isolation Pilot Plant (WIPP) waste acceptance criteria (WAC) waste limits, which was not cost effective or operations friendly. An improved and revised safety analysis was used to gain DOE-ID approval for adding this container configuration to the MFC TSD safety basis. In the process of obtaining approval of the revised safety basis, safety analysis practices were used effectively to directly support excellence in operations. Several factors contributed to the success of MFC’s effort to obtain approval for the use of DOT Type A drums, including two practices that could help in future safety basis changes at other facilities. 1) The process of incorporating the DOT Type A drums into the TSD at MFC helped to better integrate nuclear safety

  18. PCES 2.0. Performance Criteria and Evaluation System v. 2.0

    SciTech Connect

    Jackson, L.L.

    1992-04-01

    The Performance Criteria and Evaluation System (PCES) was developed in order to make a data base of criteria accessible to radiation safety staff. The criteria included in the package are applicable to occupational radiation safety at DOE reactor and nonreactor nuclear facilities, but any data base of criteria may be created using the Criterion Data Base Utiliity (CDU). PCES assists personnel in carrying out oversight, line, and support activities.

  19. Optimization strategies for sustainable fuel cycle of the BR2 Reactor

    SciTech Connect

    Kalcheva, S.; Van Den Branden, G.; Koonen, E.

    2013-07-01

    The objective of the present study is to achieve a sustainable fuel cycle in a long term of reactor operation applying advanced in-core loading strategies. The optimization criteria concern mainly enhancement of nuclear safety by means of reactivity margins and minimization of the operational fuel cycle cost at a given (constant) power level and same or longer cycle length. An important goal is also to maintain the same or to improve the experimental performances. Current developments are focused on optimization of control rods localization; optimization of fresh and burnt fuel assemblies in-core distribution; optimization of azimuth and axial fuel burn up strategies, including fuel assembly rotating and flipping upside down. (authors)

  20. Conceptual design and selection of a biodiesel fuel processor for a vehicle fuel cell auxiliary power unit

    NASA Astrophysics Data System (ADS)

    Specchia, S.; Tillemans, F. W. A.; van den Oosterkamp, P. F.; Saracco, G.

    Within the European project BIOFEAT (biodiesel fuel processor for a fuel cell auxiliary power unit for a vehicle), a complete modular 10 kW e biodiesel fuel processor capable of feeding a PEMFC will be developed, built and tested to generate electricity for a vehicle auxiliary power unit (APU). Tail pipe emissions reduction, increased use of renewable fuels, increase of hydrogen-fuel economy and efficient supply of present and future APU for road vehicles are the main project goals. Biodiesel is the chosen feedstock because it is a completely natural and thus renewable fuel. Three fuel processing options were taken into account at a conceptual design level and compared for hydrogen production: (i) autothermal reformer (ATR) with high and low temperature shift (HTS/LTS) reactors; (ii) autothermal reformer (ATR) with a single medium temperature shift (MTS) reactor; (iii) thermal cracker (TC) with high and low temperature shift (HTS/LTS) reactors. Based on a number of simulations (with the AspenPlus® software), the best operating conditions were determined (steam-to-carbon and O 2/C ratios, operating temperatures and pressures) for each process alternative. The selection of the preferential fuel processing option was consequently carried out, based on a number of criteria (efficiency, complexity, compactness, safety, controllability, emissions, etc.); the ATR with both HTS and LTS reactors shows the most promising results, with a net electrical efficiency of 29% (LHV).

  1. WATER QUALITY CRITERIA DOCUMENTS

    EPA Science Inventory

    Background

    Water quality standards and criteria are the foundation for a wide range of programs under the Clean Water Act. Specifically, under section 304(a)(1) of the Clean Water Act it requires EPA to develop criteria for water quality that accurately re...

  2. 49 CFR Appendix B to Part 236 - Risk Assessment Criteria

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 4 2011-10-01 2011-10-01 false Risk Assessment Criteria B Appendix B to Part 236... B to Part 236—Risk Assessment Criteria The safety-critical performance of each product for which risk assessment is required under this part must be assessed in accordance with the following...

  3. 16 CFR 1610.7 - Test sequence and classification criteria.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Test sequence and classification criteria. 1610.7 Section 1610.7 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION FLAMMABLE FABRICS ACT REGULATIONS STANDARD FOR THE FLAMMABILITY OF CLOTHING TEXTILES The Standard § 1610.7 Test sequence and classification criteria. (a) Preliminary...

  4. 16 CFR 1031.13 - Criteria for employee involvement.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Criteria for employee involvement. 1031.13 Section 1031.13 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION GENERAL COMMISSION PARTICIPATION AND COMMISSION EMPLOYEE INVOLVEMENT IN VOLUNTARY STANDARDS ACTIVITIES Employee Involvement § 1031.13 Criteria for employee involvement....

  5. 29 CFR 1926.95 - Criteria for personal protective equipment.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 29 Labor 8 2014-07-01 2014-07-01 false Criteria for personal protective equipment. 1926.95 Section..., DEPARTMENT OF LABOR (CONTINUED) SAFETY AND HEALTH REGULATIONS FOR CONSTRUCTION Personal Protective and Life Saving Equipment § 1926.95 Criteria for personal protective equipment. (a) Application....

  6. 29 CFR 1926.95 - Criteria for personal protective equipment.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 8 2010-07-01 2010-07-01 false Criteria for personal protective equipment. 1926.95 Section..., DEPARTMENT OF LABOR (CONTINUED) SAFETY AND HEALTH REGULATIONS FOR CONSTRUCTION Personal Protective and Life Saving Equipment § 1926.95 Criteria for personal protective equipment. (a) Application....

  7. Packaging Design Criteria for the Steel Waste Package

    SciTech Connect

    BOEHNKE, W.M.

    2000-10-19

    This packaging design criteria provides the criteria for the design, fabrication, safety evaluation, and use of the steel waste package (SWP) to transport remote-handled waste and special-case waste from the 324 facility to Central Waste Complex (CWC) for interim storage.

  8. Fast Flux Test Facility final safety analysis report. Amendment 73

    SciTech Connect

    Gantt, D.A.

    1993-08-01

    This report provides Final Safety Analysis Report (FSAR) Amendment 73 for incorporation into the Fast Flux Test Facility (FFTR) FSAR set. This page change incorporates Engineering Change Notices (ECNs) issued subsequent to Amendment 72 and approved for incorparoration before May 6, 1993. These changes include: Chapter 3, design criteria structures, equipment, and systems; chapter 5B, reactor coolant system; chapter 7, instrumentation and control systems; chapter 9, auxiliary systems; chapter 11, reactor refueling system; chapter 12, radiation protection and waste management; chapter 13, conduct of operations; chapter 17, technical specifications; chapter 20, FFTF criticality specifications; appendix C, local fuel failure events; and appendix Fl, operation at 680{degrees}F inlet temperature.

  9. Nuclear safety

    NASA Technical Reports Server (NTRS)

    Buden, D.

    1991-01-01

    Topics dealing with nuclear safety are addressed which include the following: general safety requirements; safety design requirements; terrestrial safety; SP-100 Flight System key safety requirements; potential mission accidents and hazards; key safety features; ground operations; launch operations; flight operations; disposal; safety concerns; licensing; the nuclear engine for rocket vehicle application (NERVA) design philosophy; the NERVA flight safety program; and the NERVA safety plan.

  10. Fuel flexible fuel injector

    SciTech Connect

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  11. 27 CFR 19.678 - Criteria for issuance of permit.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... BUREAU, DEPARTMENT OF THE TREASURY ALCOHOL DISTILLED SPIRITS PLANTS Distilled Spirits for Fuel Use... will issue an alcohol fuel plant permit to any person who completes the required application for a... 27 Alcohol, Tobacco Products and Firearms 1 2014-04-01 2014-04-01 false Criteria for issuance...

  12. 27 CFR 19.678 - Criteria for issuance of permit.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... BUREAU, DEPARTMENT OF THE TREASURY ALCOHOL DISTILLED SPIRITS PLANTS Distilled Spirits for Fuel Use... will issue an alcohol fuel plant permit to any person who completes the required application for a... 27 Alcohol, Tobacco Products and Firearms 1 2013-04-01 2013-04-01 false Criteria for issuance...

  13. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect

    C.E. Sanders

    2004-09-10

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging

  14. HTGR Fuel performance basis

    SciTech Connect

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600/sup 0/C, and complete fuel failure occurs at 2660/sup 0/C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents.

  15. Investigations and Recommendations on the Use of Existing Experiments in Criticality Safety Analysis of Nuclear Fuel Cycle Facilities for Weapons-Grade Plutonium

    SciTech Connect

    Rearden, B.T.

    2002-05-29

    report is given in Sect. 2. This report pertains to two of the five AOAs identified by the licensee [Duke, Cogema, Stone and Webster (DCS)] for the validation of criticality codes in the design of the Mixed-Oxide Fuel Fabrication Facility (MFFF). The five AOAs are as follows: (1) Pu-nitrate aqueous solutions (homogeneous systems), (2) Mixed-oxide (MOX) pellets, fuel rods and fuel assemblies (heterogeneous systems), (3) PuO{sub 2} powders, (4) MOX powders, and (5) Aqueous solutions of Pu compounds (Pu-oxalate solutions). This report addresses a S/U analysis pertaining to AOA 3, PuO{sub 2} powders, and AOA 4, MOX powders. AOA 3 and AOA 4 are the subject of this report since the other AOAs (solutions and heterogeneous systems) appear to be well represented in the documented benchmark experiments used in the criticality safety community. Prior to this work, DCS used traditional criticality validation techniques to identify numerous experimental benchmarks that are applicable to AOAs 3 and 4. Traditional techniques for selection of applicable benchmark experiments essentially consist of evaluating the area of applicability for important design parameters (e.g., Pu content or average neutron energy) and ensuring experiments have similar characteristics that bound or nearly bound the range of conditions requiring design analysis. DCS provided ORNL with compositions and dimensions for critical systems used to establish preliminary mass limits for facility powder and fuel pellet handling areas corresponding to AOAs 3 and 4. ORNL has reviewed existing critical experiments to identify those, which, in addition to those provided by DCS, may be applicable to the criticality code validation for AOAs 3 and 4. A S/U analysis was then performed to calculate the integral parameters used to determine the similarity of each critical experiment to each design system provided by DCS. This report contains a review of the S/U theory, a description of the design systems, a brief description of

  16. Ice sheet sensitivity experiments as part of an assessment of long-term safety for a planned repository for spent nuclear fuel in Sweden

    NASA Astrophysics Data System (ADS)

    Wekerle, Claudia; Colleoni, Florence; Masina, Simona; Näslund, Jens-Ove; Brandefelt, Jenny

    2014-05-01

    An application to build a deep geological repository for spent nuclear fuel in Forsmark in south-central Sweden is currently under consideration by Swedish authorities. As part of the safety assessment, the response of the repository to an extensive glaciation over time scales of several hundred thousand years, in terms of ice thickness, bedrock depression and hydrostatic pressure, has to be evaluated. The most extensive glaciation over Eurasia recorded in geological proxies occurred during the MIS 6, at around 140 kyrs BP (Late Saalian glaciation). At this time, the few existing numerical ice-sheet reconstructions suggest that the Eurasian ice volume reached more than 70 m SLE, which is at least three times larger than during the Last Glacial Maximum (21 kyrs BP). The reconstruction of this ice sheet is complicated by the fact that the timing of the maximum ice volume may not have been coeval with the maximum eastern and southern extent of the Saalian ice sheet. In the present study, the maximum geographical extension of the Late Saalian glaciation serves as an extreme test case to assess the impact of ice thickness over the Forsmark repository site. We use the 3D-thermodynamical ice sheet-ice shelves and ice stream model GRISLI (Ritz et al. 2001) to simulate the Northern Hemisphere ice sheet topography of the Late Saalian glaciation. The model is forced by steady-state climatic fields (surface air temperature and precipitation) computed using the coupled atmosphere-ocean Community Earth System Model (CESM, NCAR) at ~1°x1° resolution, with boundary and forcing conditions representative for the MIS6 glacial maximum. Ice sheet simulations are run on a 20 km regular rectangular grid over the northern high latitudes and allow for floating ice. First, as part of the model validation, we show a numerical reconstruction of the MIS 6 Eurasian ice sheet using standard parameters for lapse rate, PDD coefficients and basal hydrology. Second, sensitivity experiments are

  17. 29 CFR 1926.95 - Criteria for personal protective equipment.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 29 Labor 8 2011-07-01 2011-07-01 false Criteria for personal protective equipment. 1926.95 Section 1926.95 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) SAFETY AND HEALTH REGULATIONS FOR CONSTRUCTION Personal Protective and Life Saving Equipment § 1926.95...

  18. 78 FR 58567 - Criteria to Certify Coal Mine Rescue Teams

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-24

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF LABOR Mine Safety and Health Administration Criteria to Certify Coal Mine Rescue Teams AGENCY: Mine Safety and Health Administration, Labor. ACTION: Notice of availability; request for comments. SUMMARY: The...

  19. 78 FR 79010 - Criteria to Certify Coal Mine Rescue Teams

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-27

    ... training. MSHA published a notice in the Federal Register (78 FR 58567) announcing the availability of the... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF LABOR Mine Safety and Health Administration Criteria to Certify Coal Mine Rescue Teams AGENCY: Mine Safety...

  20. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents

    SciTech Connect

    Not Available

    1993-11-01

    The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

  1. Design Criteria for Microbiological Facilities at Fort Detrick. Volume II: Design Criteria

    ERIC Educational Resources Information Center

    Army Biological Labs., Fort Detrick, MD. Industrial Health and Safety Div.

    Volume II of a two-volume manual of design criteria, based primarily on biological safety considerations. It is prepared for the use of architect-engineers in designing new or modified microbiological facilities for Fort Detrick, Maryland. Volume II is divided into the following sections: (1) architectural, (2) heating, ventilating, and air…

  2. 30 CFR 75.1906 - Transport of diesel fuel.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Transport of diesel fuel. 75.1906 Section 75... diesel fuel. (a) Diesel fuel shall be transported only by diesel fuel transportation units or in safety... fuel storage facilities. (c) Safety cans that leak must be promptly removed from the mine. (d)...

  3. 40 CFR 80.584 - What are the precision and accuracy criteria for approval of test methods for determining the...

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... criteria for approval of test methods for determining the sulfur content of motor vehicle diesel fuel, NRLM diesel fuel, and ECA marine fuel? 80.584 Section 80.584 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor...

  4. 40 CFR 80.584 - What are the precision and accuracy criteria for approval of test methods for determining the...

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... criteria for approval of test methods for determining the sulfur content of motor vehicle diesel fuel, NRLM diesel fuel, and ECA marine fuel? 80.584 Section 80.584 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor...

  5. NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

    SciTech Connect

    1982-07-01

    This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules.

  6. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  7. Extending dry storage of spent LWR fuel for 100 years.

    SciTech Connect

    Einziger, R. E.

    1998-12-16

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

  8. NASA Lewis Wind Tunnel Model Systems Criteria

    NASA Technical Reports Server (NTRS)

    Soeder, Ronald H.; Haller, Henry C.

    1994-01-01

    This report describes criteria for the design, analysis, quality assurance, and documentation of models or test articles that are to be tested in the aeropropulsion facilities at the NASA Lewis Research Center. The report presents three methods for computing model allowable stresses on the basis of the yield stress or ultimate stress, and it gives quality assurance criteria for models tested in Lewis' aeropropulsion facilities. Both customer-furnished model systems and in-house model systems are discussed. The functions of the facility manager, project engineer, operations engineer, research engineer, and facility electrical engineer are defined. The format for pretest meetings, prerun safety meetings, and the model criteria review are outlined Then, the format for the model systems report (a requirement for each model that is to be tested at NASA Lewis) is described, the engineers that are responsible for developing the model systems report are listed, and the time table for its delivery to the facility manager is given.

  9. A Silent Safety Program

    NASA Technical Reports Server (NTRS)

    Goodin, James Ronald

    2006-01-01

    NASA's Columbia Accident Investigation Board (CAIB) referred 8 times to the NASA "Silent Safety Program." This term, "Silent Safety Program" was not an original observation but first appeared in the Rogers Commission's Investigation of the Challenger Mishap. The CAIB on page 183 of its report in the paragraph titled 'Encouraging Minority Opinion,' stated "The Naval Reactor Program encourages minority opinions and "bad news." Leaders continually emphasize that when no minority opinions are present, the responsibility for a thorough and critical examination falls to management. . . Board interviews revealed that it is difficult for minority and dissenting opinions to percolate up through the agency's hierarchy. . ." The first question and perhaps the only question is - what is a silent safety program? Well, a silent safety program may be the same as the dog that didn't bark in Sherlock Holmes' "Adventure of the Silver Blaze" because system safety should behave as a devil's advocate for the program barking on every occasion to insure a critical review inclusion. This paper evaluates the NASA safety program and provides suggestions to prevent the recurrence of the silent safety program alluded to in the Challenger Mishap Investigation. Specifically targeted in the CAM report, "The checks and balances the safety system was meant to provide were not working." A silent system safety program is not unique to NASA but could emerge in any and every organization. Principles developed by Irving Janis in his book, Groupthink, listed criteria used to evaluate an organization's cultural attributes that allows a silent safety program to evolve. If evidence validates Jams's criteria, then Jams's recommendations for preventing groupthink can also be used to improve a critical evaluation and thus prevent the development of a silent safety program.

  10. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    SciTech Connect

    Fung Poon, Cindy; Poston, David I.

    2006-01-20

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel)

  11. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Horning, W.A.; Lanning, D.D.; Donahue, D.J.

    1959-10-01

    A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

  12. Hanford Site solid waste acceptance criteria

    SciTech Connect

    Ellefson, M.D.

    1998-07-01

    Order 5820.2A requires that each treatment, storage, and/or disposal facility (referred to in this document as TSD unit) that manages low-level or transuranic waste (including mixed waste and TSCA PCB waste) maintain waste acceptance criteria. These criteria must address the various requirements to operate the TSD unit in compliance with applicable safety and environmental requirements. This document sets forth the baseline criteria for acceptance of radioactive waste at TSD units operated by WMH. The criteria for each TSD unit have been established to ensure that waste accepted can be managed in a manner that is within the operating requirements of the unit, including environmental regulations, DOE Orders, permits, technical safety requirements, waste analysis plans, performance assessments, and other applicable requirements. Acceptance criteria apply to the following TSD units: the Low-Level Burial Grounds (LLBG) including both the nonregulated portions of the LLBG and trenches 31 and 34 of the 218-W-5 Burial Ground for mixed waste disposal; Central Waste Complex (CWC); Waste Receiving and Processing Facility (WRAP); and T Plant Complex. Waste from all generators, both from the Hanford Site and from offsite facilities, must comply with these criteria. Exceptions can be granted as provided in Section 1.6. Specific waste streams could have additional requirements based on the 1901 identified TSD pathway. These requirements are communicated in the Waste Specification Records (WSRds). The Hanford Site manages nonradioactive waste through direct shipments to offsite contractors. The waste acceptance requirements of the offsite TSD facility must be met for these nonradioactive wastes. This document does not address the acceptance requirements of these offsite facilities.

  13. Alternative aircraft fuels technology

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1976-01-01

    NASA is studying the characteristics of future aircraft fuels produced from either petroleum or nonpetroleum sources such as oil shale or coal. These future hydrocarbon based fuels may have chemical and physical properties that are different from present aviation turbine fuels. This research is aimed at determining what those characteristics may be, how present aircraft and engine components and materials would be affected by fuel specification changes, and what changes in both aircraft and engine design would be required to utilize these future fuels without sacrificing performance, reliability, or safety. This fuels technology program was organized to include both in-house and contract research on the synthesis and characterization of fuels, component evaluations of combustors, turbines, and fuel systems, and, eventually, full-scale engine demonstrations. A review of the various elements of the program and significant results obtained so far are presented.

  14. Liquid fuel cells

    PubMed Central

    2014-01-01

    Summary The advantages of liquid fuel cells (LFCs) over conventional hydrogen–oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented. PMID:25247123

  15. Lung donor selection criteria

    PubMed Central

    Chaney, John; Suzuki, Yoshikazu; Cantu, Edward

    2014-01-01

    The criteria that define acceptable physiologic and social parameters for lung donation have remained constant since their empiric determination in the 1980s. These criteria include a donor age between 25-40, a arterial partial pressure of oxygen (PaO2)/FiO2 ratio greater than 350, no smoking history, a clear chest X-ray, clean bronchoscopy, and a minimal ischemic time. Due to the paucity of organ donors, and the increasing number of patients requiring lung transplant, finding a donor that meets all of these criteria is quite rare. As such, many transplants have been performed where the donor does not meet these stringent criteria. Over the last decade, numerous reports have been published examining the effects of individual acceptance criteria on lung transplant survival and graft function. These studies suggest that there is little impact of the historical criteria on either short or long term outcomes. For age, donors should be within 18 to 64 years old. Gender may relay benefit to all female recipients especially in male to female transplants, although results are mixed in these studies. Race matched donor/recipients have improved outcomes and African American donors convey worse prognosis. Smoking donors may decrease recipient survival post transplant, but provide a life saving opportunity for recipients that may otherwise remain on the transplant waiting list. No specific gram stain or bronchoscopic findings are reflected in recipient outcomes. Chest radiographs are a poor indicator of lung donor function and should not adversely affect organ usage aside for concerns over malignancy. Ischemic time greater than six hours has no documented adverse effects on recipient mortality and should not limit donor retrieval distances. Brain dead donors and deceased donors have equivalent prognosis. Initial PaO2/FiO2 ratios less than 300 should not dissuade donor organ usage, although recruitment techniques should be implemented with intent to transplant. PMID:25132970

  16. Lung donor selection criteria.

    PubMed

    Chaney, John; Suzuki, Yoshikazu; Cantu, Edward; van Berkel, Victor

    2014-08-01

    The criteria that define acceptable physiologic and social parameters for lung donation have remained constant since their empiric determination in the 1980s. These criteria include a donor age between 25-40, a arterial partial pressure of oxygen (PaO2)/FiO2 ratio greater than 350, no smoking history, a clear chest X-ray, clean bronchoscopy, and a minimal ischemic time. Due to the paucity of organ donors, and the increasing number of patients requiring lung transplant, finding a donor that meets all of these criteria is quite rare. As such, many transplants have been performed where the donor does not meet these stringent criteria. Over the last decade, numerous reports have been published examining the effects of individual acceptance criteria on lung transplant survival and graft function. These studies suggest that there is little impact of the historical criteria on either short or long term outcomes. For age, donors should be within 18 to 64 years old. Gender may relay benefit to all female recipients especially in male to female transplants, although results are mixed in these studies. Race matched donor/recipients have improved outcomes and African American donors convey worse prognosis. Smoking donors may decrease recipient survival post transplant, but provide a life saving opportunity for recipients that may otherwise remain on the transplant waiting list. No specific gram stain or bronchoscopic findings are reflected in recipient outcomes. Chest radiographs are a poor indicator of lung donor function and should not adversely affect organ usage aside for concerns over malignancy. Ischemic time greater than six hours has no documented adverse effects on recipient mortality and should not limit donor retrieval distances. Brain dead donors and deceased donors have equivalent prognosis. Initial PaO2/FiO2 ratios less than 300 should not dissuade donor organ usage, although recruitment techniques should be implemented with intent to transplant. PMID:25132970

  17. Safety management of complex research operators

    NASA Technical Reports Server (NTRS)

    Brown, W. J.

    1981-01-01

    Complex research and technology operations present varied potential hazards which are addressed in a disciplined, independent safety review and approval process. Potential hazards vary from high energy fuels to hydrocarbon fuels, high pressure systems to high voltage systems, toxic chemicals to radioactive materials and high speed rotating machinery to high powered lasers. A Safety Permit System presently covers about 600 potentially hazardous operations. The Safety Management Program described is believed to be a major factor in maintaining an excellent safety record.

  18. 75 FR 63433 - National Advisory Committee on Microbiological Criteria for Foods; Re-Establishment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-15

    ...; ] DEPARTMENT OF AGRICULTURE Food Safety and Inspection Service National Advisory Committee on Microbiological Criteria for Foods; Re-Establishment AGENCY: Food Safety and Inspection Service, USDA. ACTION: Notice of re... announcing the re-chartering of the National Advisory Committee on Microbiological Criteria for Foods...

  19. Fuel cell generator energy dissipator

    SciTech Connect

    Veyo, S.E.; Dederer, J.T.; Gordon, J.T.; Shockling, L.A.

    2000-02-15

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel inventory in the generator. The invention provides a safety function in eliminating the fuel energy, and also provides protection to the fuel cell stack by eliminating overheating.

  20. DOE Standard: Fire protection design criteria

    SciTech Connect

    Not Available

    1999-07-01

    The development of this Standard reflects the fact that national consensus standards and other design criteria do not comprehensively or, in some cases, adequately address fire protection issues at DOE facilities. This Standard provides supplemental fire protection guidance applicable to the design and construction of DOE facilities and site features (such as water distribution systems) that are also provided for fire protection. It is intended to be used in conjunction with the applicable building code, National Fire Protection Association (NFPA) Codes and Standards, and any other applicable DOE construction criteria. This Standard replaces certain mandatory fire protection requirements that were formerly in DOE 5480.7A, ``Fire Protection``, and DOE 6430.1A, ``General Design Criteria``. It also contains the fire protection guidelines from two (now canceled) draft standards: ``Glove Box Fire Protection`` and ``Filter Plenum Fire Protection``. (Note: This Standard does not supersede the requirements of DOE 5480.7A and DOE 6430.1A where these DOE Orders are currently applicable under existing contracts.) This Standard, along with the criteria delineated in Section 3, constitutes the basic criteria for satisfying DOE fire and life safety objectives for the design and construction or renovation of DOE facilities.

  1. Revisiting Bioaccumulation Criteria

    EPA Science Inventory

    The objective of workgroup 5 was to revisit the B(ioaccumulation) criteria that are currently being used to identify POPs under the Stockholm Convention and PBTs under CEPA, TSCA, REACh and other programs. Despite the lack of a recognized definition for a B substance, we defined ...

  2. Laboratory Equipment Criteria.

    ERIC Educational Resources Information Center

    State Univ. Construction Fund, Albany, NY.

    Requirements for planning, designing, constructing and installing laboratory furniture are given in conjunction with establishing facility criteria for housing laboratory equipment. Furniture and equipment described include--(1) center tables, (2) reagent racks, (3) laboratory benches and their mechanical fixtures, (4) sink and work counters, (5)…

  3. Graphite criteria peer review

    SciTech Connect

    1986-09-01

    This report documents a review of the stress criteria proposed for the graphite components of the modular high temperature gas-cooled reactor (MHTGR) core. The review was conducted by a panel of six independent consultants, chosen for their expertise over a range of relevant disciplines.

  4. CRITERIA FOR COUNSELOR PERFORMANCE.

    ERIC Educational Resources Information Center

    MILLER, LEONARD A.; MUTHARD, JOHN E.

    THIS RESEARCH CONCERNS THE RELATIONSHIPS AMONG REHABILITATION COUNSELOR PERFORMANCE CRITERIA CURRENTLY BEING USED OR READILY AVAILABLE TO STATE VOCATIONAL REHABILITATION AGENCIES. THE 143 COUNSELORS STUDIED CAME FROM MIDDLE-SIZED AGENCIES IN SIX STATES AND, SINCE COWORKER RATINGS WERE REQUIRED, THE SAMPLE WAS LIMITED TO COUNSELORS WORKING WITH TWO…

  5. Safety Tips: Hazardous Chemical Storage.

    ERIC Educational Resources Information Center

    Williamson, J. R.

    1983-01-01

    Discusses storage of hazardous chemicals and provides a list of eight basic safety rules to use in developing a safe storage system. Suggestions include not storing materials alphabetically, storing nonreactive chemicals together, and not storing oxidizers and fuels together. (JN)

  6. An underwater neutron coincidence counter for measurement of spent fuels

    SciTech Connect

    Staples, P.; Halbig, J.; Lestone, J.; Sprinkle, J.

    1999-07-01

    An underwater neutron coincident counter has been designed and constructed for the measurement of spent nuclear fuel--the spent-fuel coincident counter (SFCC). The SFCC is a medium-detection-efficiency design that incorporates an ionization chamber (IC) for gamma-ray dose evaluation from the spent nuclear fuel. The absolute neutron detection efficiency is 14.5% for {sup 252}Cf sources. The SFCC is hermetically sealed, as it is installed {approximately}5 m below water level in a spent-fuel storage pond. There are 20 {sup 3}He tubes arranged within a polyethylene ring in a single band. There is an inner ring of 6.8 cm of lead to provide shielding from the fission product gamma rays. A single IC is primarily used to determine the dose impinging upon the {sup 3}He tubes and to determine the appropriate operational parameters to avoid gamma-ray pile effects in the {sup 3}He tubes. To further minimize gamma-ray pileup effects, each {sup 3}He tube is connected to a PDT110A preamplifier. The single and double neutron count rates, in addition to the IC measurement information from the SFCC, are used to determine the Pu mass of the spent-fuel assemblies and the decay heat and for classification of the assembly type. This information is required such that safety criteria are met for the safe packaging of the spent-fuel assemblies.

  7. Application of Neutron-Absorbing Structural-Amorphous Metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls

    SciTech Connect

    Choi, Jor-Shan; Lee, Chuck; Farmer, Joseph; Day, Dan; Wall, Mark; Saw, Cheng; Boussoufi, Moe; Liu, Ben; Egbert, Harold; Branagan, Dan; D'Amato, Andy

    2007-07-01

    Spent nuclear fuel contains fissionable materials ({sup 235}U, {sup 239}Pu, {sup 241}Pu, etc.). To prevent nuclear criticality in spent fuel storage, transportation, and during disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral{sup TM}, Metamic{sup TM}, Ni-Gd, and others) would have to be applied. The success in demonstrating that the High-Performance Corrosion- Resistant Material (HPCRM){sup [1]} can be thermally applied as coating onto base metal to provide for corrosion resistance for many naval applications raises the interest in applying the HPCRM to USDOE/OCRWM spent fuel management program. The fact that the HPCRM relies on the high content of boron to make the material amorphous - an essential property for corrosion resistance - and that the boron has to be homogeneously distributed in the HPCRM qualify the material to be a neutron poison. (authors)

  8. Safety evaluation report related to the evaluation of low-enriched uranium silicide-aluminum dispersion fuel for use in non-power reactors

    SciTech Connect

    Not Available

    1988-07-01

    Low-enriched uranium silicide-aluminum dispersion plate-type fuels have been extensively researched and developed under the international program, Reduced Enrichment in Research and Test Reactors. The international effort was led by Argonne National Laboratory (ANL) in the United States. This evaluation is based primarily on reports issued by ANL that discuss and summarize the developmental tests and experiments, including postirradiation examinations, of both miniature and full-sized plates of prototypical fuel compositions. This evaluation concludes that plate-type fuels suitable and acceptable for use in research and test reactors can be fabricated with U/sub 3/Si/sub 2/-Al dispersion compacts with uranium densities up to 4.8 g/cm/sup 3/. 4 refs., 1 fig.

  9. Safety system status monitoring

    SciTech Connect

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.

  10. 33 CFR 183.538 - Metallic fuel line materials.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Metallic fuel line materials. 183... (CONTINUED) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.538 Metallic fuel line materials. Each metallic fuel line connecting the fuel tank with the fuel inlet connection...

  11. 33 CFR 183.528 - Fuel stop valves.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel stop valves. 183.528 Section...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.528 Fuel stop valves. (a) Each electrically operated fuel stop valve in a fuel line between the fuel tank and the...

  12. 33 CFR 183.528 - Fuel stop valves.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 2 2011-07-01 2011-07-01 false Fuel stop valves. 183.528 Section...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.528 Fuel stop valves. (a) Each electrically operated fuel stop valve in a fuel line between the fuel tank and the...

  13. 33 CFR 183.528 - Fuel stop valves.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Fuel stop valves. 183.528 Section...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.528 Fuel stop valves. (a) Each electrically operated fuel stop valve in a fuel line between the fuel tank and the...

  14. 33 CFR 183.528 - Fuel stop valves.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 33 Navigation and Navigable Waters 2 2013-07-01 2013-07-01 false Fuel stop valves. 183.528 Section...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.528 Fuel stop valves. (a) Each electrically operated fuel stop valve in a fuel line between the fuel tank and the...

  15. 33 CFR 183.528 - Fuel stop valves.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Fuel stop valves. 183.528 Section...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.528 Fuel stop valves. (a) Each electrically operated fuel stop valve in a fuel line between the fuel tank and the...

  16. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    SciTech Connect

    Cook, David Howard

    2009-01-01

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

  17. A comparison of low-pressure and supercharged operation of polymer electrolyte membrane fuel cell systems for aircraft applications

    NASA Astrophysics Data System (ADS)

    Werner, C.; Preiß, G.; Gores, F.; Griebenow, M.; Heitmann, S.

    2016-08-01

    Multifunctional fuel cell systems are competitive solutions aboard future generations of civil aircraft concerning energy consumption, environmental issues, and safety reasons. The present study compares low-pressure and supercharged operation of polymer electrolyte membrane fuel cells with respect to performance and efficiency criteria. This is motivated by the challenge of pressure-dependent fuel cell operation aboard aircraft with cabin pressure varying with operating altitude. Experimental investigations of low-pressure fuel cell operation use model-based design of experiments and are complemented by numerical investigations concerning supercharged fuel cell operation. It is demonstrated that a low-pressure operation is feasible with the fuel cell device under test, but that its range of stable operation changes between both operating modes. Including an external compressor, it can be shown that the power demand for supercharging the fuel cell is about the same as the loss in power output of the fuel cell due to low-pressure operation. Furthermore, the supercharged fuel cell operation appears to be more sensitive with respect to variations in the considered independent operating parameters load requirement, cathode stoichiometric ratio, and cooling temperature. The results indicate that a pressure-dependent self-humidification control might be able to exploit the potential of low-pressure fuel cell operation for aircraft applications to the best advantage.

  18. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  19. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    SciTech Connect

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  20. Unified nonclassicality criteria

    NASA Astrophysics Data System (ADS)

    Ryl, S.; Sperling, J.; Agudelo, E.; Mraz, M.; Köhnke, S.; Hage, B.; Vogel, W.

    2015-07-01

    In this work we generalize the Bochner criterion addressing the characteristic function, i.e., the Fourier transform, of the Glauber-Sudarshan phase-space function. For this purpose we extend the Bochner theorem by including derivatives of the characteristic function. The resulting necessary and sufficient nonclassicality criteria unify previously known moment-based criteria with those based on the characteristic function. For applications of the generalized nonclassicality probes, we provide direct sampling formulas for balanced homodyne detection. A squeezed vacuum state is experimentally realized and characterized with our method. This complete framework—theoretical unification, sampling approach, and experimental implementation—presents an efficient toolbox to characterize quantum states of light for applications in quantum technology.

  1. Application of Neutron-Absorbing Structural-Amorphous Metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls

    SciTech Connect

    Choi, J; Lee, C; Day, D; Wall, M; Saw, C; MoberlyChan, W; Farmer, J; Boussoufl, M; Liu, B; Egbert, H; Branagan, D; D'Amato, A

    2006-11-13

    Spent nuclear fuel contains fissionable materials ({sup 235}U, {sup 239}Pu, {sup 241}Pu, etc.). Neutron multiplication and the potential for criticality are enhanced by the presence of a moderator during cask loading in water, water incursion in accidents conditions during spent fuel storage or transport. To prevent nuclear criticality in spent fuel storage, transportation, and during disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral{trademark}, Metamic{trademark}, Ni-Gd, and others) would have to be applied. The success in demonstrating that the High-Performance Corrosion-Resistant material (HPCRM) can be thermally applied as coating onto base metal to provide for corrosion resistance for many naval applications raises the interest in applying the HPCRM to USDOE/OCRWM spent fuel management program. The fact that the HPCRM relies on the high content of boron to make the material amorphous--an essential property for corrosion resistance--and that the boron has to be homogeneously distributed in the HPCRM qualify the material to be a neutron poison.

  2. PML diagnostic criteria

    PubMed Central

    Aksamit, Allen J.; Clifford, David B.; Davis, Larry; Koralnik, Igor J.; Sejvar, James J.; Bartt, Russell; Major, Eugene O.; Nath, Avindra

    2013-01-01

    Objective: To establish criteria for the diagnosis of progressive multifocal leukoencephalopathy (PML). Methods: We reviewed available literature to identify various diagnostic criteria employed. Several search strategies employing the terms “progressive multifocal leukoencephalopathy” with or without “JC virus” were performed with PubMed, SCOPUS, and EMBASE search engines. The articles were reviewed by a committee of individuals with expertise in the disorder in order to determine the most useful applicable criteria. Results: A consensus statement was developed employing clinical, imaging, pathologic, and virologic evidence in support of the diagnosis of PML. Two separate pathways, histopathologic and clinical, for PML diagnosis are proposed. Diagnostic classification includes certain, probable, possible, and not PML. Conclusion: Definitive diagnosis of PML requires neuropathologic demonstration of the typical histopathologic triad (demyelination, bizarre astrocytes, and enlarged oligodendroglial nuclei) coupled with the techniques to show the presence of JC virus. The presence of clinical and imaging manifestations consistent with the diagnosis and not better explained by other disorders coupled with the demonstration of JC virus by PCR in CSF is also considered diagnostic. Algorithms for establishing the diagnosis have been recommended. PMID:23568998

  3. Preliminary criteria for the fracture control of space shuttle structures

    NASA Technical Reports Server (NTRS)

    1971-01-01

    The complex and multidisciplinary factors are presented which relate to the prevention of structural failure due to the initiation or propagation of cracks or crack-like defects. The fracture control criteria are applicable to space shuttle components which are: (1) susceptible to cracking or fracture on the basis of anticipated loads and environment, and (2) critical to either crew safety or system performance. The criteria define the design, fabrication, environmental control, inspection, maintenance, repair, and verification procedures required for adequate fracture control.

  4. State of Washington Department of Ecology criteria pollutants and toxic air polluntants phase II notice of construction for the Hanford Site spent nuclear fuel project--cold vacuum dryingfacility, Project W-441

    SciTech Connect

    Jansky, M.T., Westinghouse Hanford

    1997-01-24

    This Phase 11 notice of construction (NOC) provides the additional information committed to in the Phase I NOC submittal (DOE/RL-96- 55) regarding the air toxic and criteria pollutants that could potentially be emitted during operation of the Cold Vacuum Drying Facility (CVDF). This Phase 11 NOC is being submitted to ensure the CVDF is in full compliance with Washington Administrative Code (WAC) 173-460-040(8), `Commencement of Construction`. The Phase I NOC (approved September 30, 1996) was defined as constructing the substructure, including but not limited to, pouring the concrete for the floor, and construction of the exterior. This Phase 11 NOC is being submitted for approval before installation and operation of the process equipment that will generate any potential air emissions at the CVDF, and installation and operation of the emissions control equipment.

  5. Answering Key Fuel Cycle Questions

    SciTech Connect

    Steven J. Piet; Brent W. Dixon; J. Stephen Herring; David E. Shropshire; Mary Lou Dunzik-Gougar

    2003-10-01

    The Advanced Fuel Cycle Initiative (AFCI) program has both “outcome” and “process” goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are readiness to proceed and adaptability and robustness in the face of uncertainties. A classic decision-making approach to such a multi-attribute problem would be to weight individual quantified criteria and calculate an overall figure of merit. This is inappropriate for several reasons. First, the goals are not independent. Second, the importance of different goals varies among stakeholders. Third, the importance of different goals is likely to vary with time, especially the “energy future.” Fourth, some key considerations are not easily or meaningfully quantifiable at present. Instead, at this point, we have developed 16 questions the AFCI program should answer and suggest an approach of determining for each whether relevant options improve meeting each of the program goals. We find that it is not always clear which option is best for a specific question and specific goal; this helps identify key issues for future work. In general, we suggest attempting to create as many win-win decisions (options that are attractive or neutral to most goals) as possible. Thus, to help clarify why the program is exploring the options it is, and to set the stage for future narrowing of options, we have developed 16 questions, as follows: · What are the AFCI program goals? · Which potential waste disposition approaches do we plan for? · What are the major separations, transmutation, and fuel options? · How do we address proliferation resistance? · Which potential energy futures do we plan for? · What potential external triggers do we

  6. 33 CFR 183.542 - Fuel systems.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 33 Navigation and Navigable Waters 2 2013-07-01 2013-07-01 false Fuel systems. 183.542 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.542 Fuel systems. (a) Each fuel system in a boat must have been tested by the boat manufacturer and not leak when subjected to...

  7. 33 CFR 183.510 - Fuel tanks.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Fuel tanks. 183.510 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.510 Fuel tanks. (a) Each fuel tank in a boat must have been tested by its manufacturer under § 183.580 and not leak...

  8. 33 CFR 183.542 - Fuel systems.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Fuel systems. 183.542 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.542 Fuel systems. (a) Each fuel system in a boat must have been tested by the boat manufacturer and not leak when subjected to...

  9. 33 CFR 183.524 - Fuel pumps.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Fuel pumps. 183.524 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.524 Fuel pumps. (a) Each diaphragm pump must not leak fuel from the pump if the primary diaphragm fails. (b) Each...

  10. 33 CFR 183.524 - Fuel pumps.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Fuel pumps. 183.524 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.524 Fuel pumps. (a) Each diaphragm pump must not leak fuel from the pump if the primary diaphragm fails. (b) Each...

  11. 33 CFR 183.510 - Fuel tanks.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel tanks. 183.510 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.510 Fuel tanks. (a) Each fuel tank in a boat must have been tested by its manufacturer under § 183.580 and not leak...

  12. 33 CFR 183.510 - Fuel tanks.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 2 2011-07-01 2011-07-01 false Fuel tanks. 183.510 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.510 Fuel tanks. (a) Each fuel tank in a boat must have been tested by its manufacturer under § 183.580 and not leak...

  13. 33 CFR 183.510 - Fuel tanks.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Fuel tanks. 183.510 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.510 Fuel tanks. (a) Each fuel tank in a boat must have been tested by its manufacturer under § 183.580 and not leak...

  14. 33 CFR 183.510 - Fuel tanks.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 33 Navigation and Navigable Waters 2 2013-07-01 2013-07-01 false Fuel tanks. 183.510 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.510 Fuel tanks. (a) Each fuel tank in a boat must have been tested by its manufacturer under § 183.580 and not leak...

  15. 33 CFR 183.542 - Fuel systems.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 2 2011-07-01 2011-07-01 false Fuel systems. 183.542 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.542 Fuel systems. (a) Each fuel system in a boat must have been tested by the boat manufacturer and not leak when subjected to...

  16. 33 CFR 183.524 - Fuel pumps.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel pumps. 183.524 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.524 Fuel pumps. (a) Each diaphragm pump must not leak fuel from the pump if the primary diaphragm fails. (b) Each...

  17. 33 CFR 183.542 - Fuel systems.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel systems. 183.542 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.542 Fuel systems. (a) Each fuel system in a boat must have been tested by the boat manufacturer and not leak when subjected to...

  18. 33 CFR 183.524 - Fuel pumps.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 33 Navigation and Navigable Waters 2 2013-07-01 2013-07-01 false Fuel pumps. 183.524 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.524 Fuel pumps. (a) Each diaphragm pump must not leak fuel from the pump if the primary diaphragm fails. (b) Each...

  19. 33 CFR 183.542 - Fuel systems.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Fuel systems. 183.542 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.542 Fuel systems. (a) Each fuel system in a boat must have been tested by the boat manufacturer and not leak when subjected to...

  20. 33 CFR 183.524 - Fuel pumps.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 2 2011-07-01 2011-07-01 false Fuel pumps. 183.524 Section 183... SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Equipment Standards § 183.524 Fuel pumps. (a) Each diaphragm pump must not leak fuel from the pump if the primary diaphragm fails. (b) Each...

  1. Criteria for Assessing Naturalistic Inquiries as Reports.

    ERIC Educational Resources Information Center

    Lincoln, Yvonna S.; Guba, Egon G.

    Research on the assessment of naturalistic inquiries is reviewed, and criteria for assessment are outlined. Criteria reviewed include early foundational and non-foundational criteria, trustworthiness criteria, axiomatic criteria, rhetorical criteria, action criteria, and application/transferability criteria. Case studies that are reports of…

  2. Low conversion ratio fuel studies.

    SciTech Connect

    Smith, M. A.

    2006-02-28

    Recent studies on TRU disposition in fast reactors indicated viable reactor performance for a sodium cooled low conversion ratio reactor design. Additional studies have been initiated to refine the earlier work and consider the feasibility of alternate fuel forms such as nitride and oxide fuel (rather than metal fuel). These alternate fuel forms may have significant impacts upon the burner design and the safety behavior. The work performed thus far has focused on compiling the necessary fuel form property information and refinement of the physics models. For this limited project, the burner design and performance using nitride fuel will be assessed.

  3. HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

    SciTech Connect

    DUNCAN, D.R.

    1999-10-07

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

  4. 30 CFR 75.222 - Roof control plan-approval criteria.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Roof control plan-approval criteria. 75.222 Section 75.222 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Roof Support § 75.222 Roof...

  5. 30 CFR 75.222 - Roof control plan-approval criteria.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Roof control plan-approval criteria. 75.222 Section 75.222 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Roof Support § 75.222 Roof...

  6. 30 CFR 75.222 - Roof control plan-approval criteria.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Roof control plan-approval criteria. 75.222 Section 75.222 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Roof Support § 75.222 Roof...

  7. 30 CFR 75.222 - Roof control plan-approval criteria.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Roof control plan-approval criteria. 75.222 Section 75.222 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Roof Support § 75.222 Roof...

  8. 30 CFR 75.222 - Roof control plan-approval criteria.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Roof control plan-approval criteria. 75.222 Section 75.222 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Roof Support § 75.222 Roof...

  9. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports

    SciTech Connect

    Not Available

    1993-11-01

    A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

  10. Magnetic criteria of aromaticity.

    PubMed

    Gershoni-Poranne, Renana; Stanger, Amnon

    2015-09-21

    This review describes the current state of magnetic criteria of aromaticity. The introduction contains the fundamentals of ring currents in aromatic and antiaromatic systems, followed by a brief description of experimental and computational tools: NMR, diamagnetic susceptibility exaltation, current density analyses (CDA) and nucleus independent chemical shifts (NICS). This is followed by more comprehensive chapters: NMR - focusing on the work of R. Mitchell - NICS and CDA - describing the progress and development of the methods to their current state and presenting some examples of representative work. PMID:26035305

  11. Criteria for software modularization

    NASA Technical Reports Server (NTRS)

    Card, David N.; Page, Gerald T.; Mcgarry, Frank E.

    1985-01-01

    A central issue in programming practice involves determining the appropriate size and information content of a software module. This study attempted to determine the effectiveness of two widely used criteria for software modularization, strength and size, in reducing fault rate and development cost. Data from 453 FORTRAN modules developed by professional programmers were analyzed. The results indicated that module strength is a good criterion with respect to fault rate, whereas arbitrary module size limitations inhibit programmer productivity. This analysis is a first step toward defining empirically based standards for software modularization.

  12. 49 CFR 229.93 - Safety cut-off device.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 4 2013-10-01 2013-10-01 false Safety cut-off device. 229.93 Section 229.93..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Internal Combustion Equipment § 229.93 Safety cut-off device. The fuel line shall have a safety cut-off device that— (a)...

  13. Performance-based waste acceptance criteria preliminary baseline assumptions

    SciTech Connect

    Not Available

    1994-10-24

    The Department of Energy`s (DOE`s) strategy for the management of transuranic (TRU) and TRU mixed wastes has focused on the development of the Waste Isolation Pilot Plant (WIPP). The WIPP repository is designated to receive DOE defense wastes that meet the established criteria for acceptance. As a national strategy [DOE, 1993], DOE does not intend to treat candidate wastes unless treatment or processing are necessary to meet the safety, health, and regulatory criteria for transport and disposal at WIPP. The WIPP WAC has evolved over the past 10 years to include criteria and requirements in support of the Waste Characterization program and other related compliance programs. In aggregate, the final health, safety and regulatory criteria for the waste will be documented in the Disposal WAC. This document serves two purposes. First, it familiarizes regulators and stakeholders with the concept of performance based waste acceptance criteria as an augmentation within a final Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria. Second, the document preliminarily identifies certain waste characteristics that appear important to the performance assessment process for WIPP; therefore, these could become component characteristics in the Performance Based Waste Acceptance Criteria (PBWAC). Identification of the final PBWAC will be accomplished through iterative runs of the System Prioritization Method (SPM). These iterations will serve to more clearly isolate and identify those waste characteristics that directly and predominately impact on the performance assessment.

  14. Gout Classification Criteria: Update and Implications

    PubMed Central

    Vargas-Santos, Ana Beatriz; Taylor, William J.

    2016-01-01

    Gout is the most common inflammatory arthritis, with a rising prevalence and incidence worldwide. There has been a resurgence in gout research, fueled, in part, by a number of advances in pharmacologic therapy for gout. The conduct of clinical trials and other observational research in gout requires a standardized and validated means of assembling well-defined groups of patients with gout for such research purposes. Recently, an international collaborative effort that involved a data-driven process with state-of-the art methodology supported by the American College of Rheumatology and the European League Against Rheumatism led to publication of new gout classification criteria. PMID:27342957

  15. Development of quantitative risk acceptance criteria

    SciTech Connect

    Griesmeyer, J. M.; Okrent, D.

    1981-01-01

    Some of the major considerations for effective management of risk are discussed, with particular emphasis on risks due to nuclear power plant operations. Although there are impacts associated with the rest of the fuel cycle, they are not addressed here. Several previously published proposals for quantitative risk criteria are reviewed. They range from a simple acceptance criterion on individual risk of death to a quantitative risk management framework. The final section discussed some of the problems in the establishment of a framework for the quantitative management of risk.

  16. Gout Classification Criteria: Update and Implications.

    PubMed

    Vargas-Santos, Ana Beatriz; Taylor, William J; Neogi, Tuhina

    2016-07-01

    Gout is the most common inflammatory arthritis, with a rising prevalence and incidence worldwide. There has been a resurgence in gout research, fueled, in part, by a number of advances in pharmacologic therapy for gout. The conduct of clinical trials and other observational research in gout requires a standardized and validated means of assembling well-defined groups of patients with gout for such research purposes. Recently, an international collaborative effort that involved a data-driven process with state-of-the art methodology supported by the American College of Rheumatology and the European League Against Rheumatism led to publication of new gout classification criteria. PMID:27342957

  17. Handling glacially induced faults in the assessment of the long term safety of a repository for spent nuclear fuel at Forsmark, Sweden

    NASA Astrophysics Data System (ADS)

    Munier, R.

    2011-12-01

    Located deep into the Baltic shield, far from active plate boundaries and volcanism, Swedish bedrock is characterised by a low frequency of earthquakes of small magnitudes. Yet, faults, predominantly in the Lapland region, offsetting the quarternary regolith ten meters or more, reveal that Swedish bedrock suffered from substantial earthquake activity in connection to the retreat of the latest continental glacier, Weichsel. Storage of nuclear wastes, hazardous for hundreds of thousand years, requires, firstly, isolation of radionuclides and, secondly, retardation of the nuclides should the barriers fail. Swedish regulations require that safety is demonstrated for a period of a million years. Consequently, the repository must be designed to resist the impact of several continental glaciers. Large, glacially induced, earthquakes near the repository have the potential of triggering slip along fractures across the canisters containing the nuclear wastes, thereby simultaneously jeopardising isolation, retardation and, hence, long term safety. It has therefore been crucial to assess the impact of such intraplate earthquake upon the primary functions of the repository. We conclude that, by appropriate design of the repository, the negative impact of earthquakes on long term safety can be considerably lessened. We were, additionally, able to demonstrate compliance with Swedish regulations in our safety assessment, SR-Site, submitted to the authorities earlier this year. However, the assessment required a number of critical assumptions, e.g. concerning the strain rate and the fracture properties of the rock, many of which are subject of current research in the geoscientific community. By a conservative approach, though, we judge to have adequately propagated critical uncertainties through the assessment and bound the uncertainty space.

  18. Quartic Rotation Criteria and Algorithms.

    ERIC Educational Resources Information Center

    Clarkson, Douglas B.; Jennrich, Robert I.

    1988-01-01

    Most of the current analytic rotation criteria for simple structure in factor analysis are summarized and identified as members of a general symmetric family of quartic criteria. A unified development of algorithms for orthogonal and direct oblique rotation using arbitrary criteria from this family is presented. (Author/TJH)

  19. Summertime Safety

    MedlinePlus

    ... Violence & Safety Life Stages & Populations Travelers' Health Workplace Safety & Health Features Media Sign up for Features Get Email Updates ... Submit What's this? Submit Button Past Emails CDC Features Summertime Safety Recommend on Facebook Tweet Share Compartir The feature ...

  20. Drug Safety

    MedlinePlus

    ... over-the-counter drug. The FDA evaluates the safety of a drug by looking at Side effects ... clinical trials The FDA also monitors a drug's safety after approval. For you, drug safety means buying ...