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Sample records for electrical safety code

  1. Electrical safety code manual a plan language guide to national electrical code, OSHA and NFPA 70E

    CERN Document Server

    Keller, Kimberley

    2010-01-01

    Safety in any workplace is extremely important. In the case of the electrical industry, safety is critical and the codes and regulations which determine safe practices are both diverse and complicated. Employers, electricians, electrical system designers, inspectors, engineers and architects must comply with safety standards listed in the National Electrical Code, OSHA and NFPA 70E. Unfortunately, the publications which list these safety requirements are written in very technically advanced terms and the average person has an extremely difficult time understanding exactly what they need to

  2. Section 60 revisited (The Ontario Electrical Safety Code)

    Energy Technology Data Exchange (ETDEWEB)

    Olechna, T.

    2003-06-01

    Recent changes to the Ontario Electrical Safety Code (OESC), specifically the deletion of Section 60, Electrical Communication Systems, are discussed in an effort to explain the history behind the decision and the time frame of the changes. Communication systems include telephone, telegraph, data communications, intercommunications, wired music and paging systems. In brief, the deletion of Section 60 occurred in 1983, and resulted from the fact that communication-type wiring was historically the property of the communications utility and under federal jurisdiction. Since such equipment was under federal jurisdiction, they were not inspected in Ontario, hence the deletion of Section 60 from the Ontario Code. It should be noted that although Section 60 is deleted, a number of rules applicable to communications circuits are spread throughout various sections of the Code, notably in Rule 1-032 dealing with damage and interference, Rule 4-022 involving harmonics issues, Rule 12-904(2) regulates the use of conductors that are of different sources of voltage, and Rule 10-708 which specifies the spacing and bonding requirements for communications systems. The end result is that even though Section 60 was deleted, there are these and other rules in the OESC that have direct impact on communications circuits and in effect help to protect the integrity of the system.

  3. DOE handbook electrical safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    Electrical Safety Handbook presents the Department of Energy (DOE) safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety guidance and information for DOE installations to effect a reduction or elimination of risks associated with the use of electrical energy. The objectives of this handbook are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.

  4. Electrical safety guidelines

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-01

    The Electrical Safety Guidelines prescribes the DOE safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety standards and guidance for DOE installations in order to affect a reduction or elimination of risks associated with the use of electrical energy. The objectives of these guidelines are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.

  5. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

    Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

  6. Principles of electrical safety

    CERN Document Server

    Sutherland, Peter E

    2015-01-01

    Principles of Electrical Safety discusses current issues in electrical safety, which are accompanied by series' of practical applications that can be used by practicing professionals, graduate students, and researchers. .  Provides extensive introductions to important topics in electrical safety Comprehensive overview of inductance, resistance, and capacitance as applied to the human body Serves as a preparatory guide for today's practicing engineers

  7. Managing electrical safety

    CERN Document Server

    Wiggins, James H, Jr

    2001-01-01

    Managing Electrical Safety provides an overview of electric basics, hazards, and established standards that enables you to understand the hazards you are likely to encounter in your workplace. Focusing on typical industrial environments-which utilize voltages much higher than household or office circuits-the author identifies the eight key components of an electrical safety program and examines each using a model safety management process. You'll learn how to identify electrical hazards, how to prescribe necessary electrical Personal Protective Equipment, how to ensure that equipment is de-ene

  8. Code of safety for nuclear merchant ships

    International Nuclear Information System (INIS)

    1982-01-01

    The Code is in chapters, entitled: general (including general safety principles and principles of risk acceptance); design criteria and conditions; ship design, construction and equipment; nuclear steam supply system; machinery and electrical installations; radiation safety (including radiological protection design; protection of persons; dosimetry; radioactive waste management); operation (including emergency operation procedures); surveys. Appendices cover: sinking velocity calculations; seaway loads depending on service periods; safety assessment; limiting dose-equivalent rates for different areas and spaces; quality assurance programme; application of single failure criterion. Initial application of the Code is restricted to conventional types of ships propelled by nuclear propulsion plants with pressurized light water type reactors. (U.K.)

  9. Verification of reactor safety codes

    International Nuclear Information System (INIS)

    Murley, T.E.

    1978-01-01

    The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

  10. Nuclear safety code study

    Energy Technology Data Exchange (ETDEWEB)

    Hu, H.H.; Ford, D.; Le, H.; Park, S.; Cooke, K.L.; Bleakney, T.; Spanier, J.; Wilburn, N.P.; O' Reilly, B.; Carmichael, B.

    1981-01-01

    The objective is to analyze an overpower accident in an LMFBR. A simplified model of the primary coolant loop was developed in order to understand the instabilities encountered with the MELT III and SAS codes. The computer programs were translated for switching to the IBM 4331. Numerical methods were investigated for solving the neutron kinetics equations; the Adams and Gear methods were compared. (DLC)

  11. Casebook on electric safety accidents

    International Nuclear Information System (INIS)

    1987-09-01

    This book gives concentration on electric safety accidents in domestic and abroad, which introduces general electrical safety with property of electricity, safe equipment and maintenance and protection of electric shock. It lists the cases of accident caused of electricity in domestic like accident in power substation, utilization equipment, load system and another accident by electricity like death in electric shock another by electricity like death in electric shock in new building construction, the cases caused of electricity in abroad like damage in electric shock by high voltage electric transformer, electric shock in summer and earth fault accident by fault cooling tower.

  12. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  13. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  14. Code of Ethics for Electrical Engineers

    Science.gov (United States)

    Matsuki, Junya

    The Institute of Electrical Engineers of Japan (IEEJ) has established the rules of practice for its members recently, based on its code of ethics enacted in 1998. In this paper, first, the characteristics of the IEEJ 1998 ethical code are explained in detail compared to the other ethical codes for other fields of engineering. Secondly, the contents which shall be included in the modern code of ethics for electrical engineers are discussed. Thirdly, the newly-established rules of practice and the modified code of ethics are presented. Finally, results of questionnaires on the new ethical code and rules which were answered on May 23, 2007, by 51 electrical and electronic students of the University of Fukui are shown.

  15. Home Electrical Safety Checklist

    Science.gov (United States)

    ... Interrupter Protection for Pools, Spas and Hot Tubs Metal Ladders and Electricity Don’t Mix Electrocution Hazard with Do-It-Yourself Repairs of Microwave Ovens Preventing Home Fires: Arc Fault Circuit Interrupters (AFCIs) Power up with ...

  16. Electrical, instrumentation, and control codes and standards

    International Nuclear Information System (INIS)

    Kranning, A.N.

    1978-01-01

    During recent years numerous documents in the form of codes and standards have been developed and published to provide design, fabrication and construction rules and criteria applicable to instrumentation, control and power distribution facilities for nuclear power plants. The contents of this LTR were prepared by NUS Corporation under Subcontract K5108 and provide a consolidated index and listing of the documents selected for their application to procurement of materials and design of modifications and new construction at the LOFT facility. These codes and standards should be applied together with the National Electrical Code, the ID Engineering Standards and LOFT Specifications to all LOFT instrument and electrical design activities

  17. Certification plan for safety and PRA codes

    International Nuclear Information System (INIS)

    Toffer, H.; Crowe, R.D.; Ades, M.J.

    1990-05-01

    A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA ampersand PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations

  18. Electrical safety in flammable gas/vapor laden atmospheres

    CERN Document Server

    Korver, WOE

    1992-01-01

    This book provides comprehensive coverage of electrical system installation within areas where flammable gases and liquids are handled and processed. The accurate hazard evaluation of flammability risks associated with chemical and petrochemical locations is critical in determining the point at which the costs of electrical equipment and installation are balanced with explosion safety requirements. The book offers the most current code requirements along with tables and illustrations as analytic tools. Environmental characteristics are covered in Section 1 along with recommended electrical ins

  19. Pulse coded safety logic for PFBR

    International Nuclear Information System (INIS)

    Anwer, Md. Najam; Satheesh, N.; Nagaraj, C.P.; Krishnakumar, B.

    2002-01-01

    Full text: Reactor safety logic is designed to initiate safety action against design basis events. The reactor is shutdown by de-energizing electromagnets and dropping the absorber rods under gravity. In prototype fast breeder reactor (PFBR), shutdown is affected by two independent shutdown systems, viz., control and safety rod drive mechanism (CSRDM) and diverse safety rod drive mechanism (DSRDM). Two separate safety logics are proposed for CSRDM and DSRDM, i.e. solid state logic with on-line fine impulse test (FIT) for CSRDM and pulse coded safety logic (PCSL) for DSRDM. The PCSL primarily utilizes the fact that the vast majority of faults in the logic circuitry result in static conditions at the output. It is arranged such that the presence of pulses are required to hold the shutdown actuators and any DC logic state, either logic 0 or logic 1 releases them. It is a dynamic, self-testing logic and used in a number of reactors. This paper describes the principle of operation of PCSL, its advantages, the concept of guard line logic (GLL), detection of stuck at 0 and stuck at 1 faults, fail safe and diversity features. The implementation of PCSL using Altera Max+Plus II software for PFBR trip signals and the results of simulation are discussed. This paper also describes a test jig using 80186 based system for testing PCSL for various input parameter's combinations and monitoring the outputs

  20. Code on the safety of nuclear power plants: Siting

    International Nuclear Information System (INIS)

    1988-01-01

    This Code provides criteria and procedures that are recommended for safety in nuclear power plant siting. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  1. 38 CFR 61.20 - Life Safety Code capital grants.

    Science.gov (United States)

    2010-07-01

    ... (CONTINUED) VA HOMELESS PROVIDERS GRANT AND PER DIEM PROGRAM § 61.20 Life Safety Code capital grants. (a) This section sets forth provisions for obtaining a Life Safety Code capital grant under 38 U.S.C. 2012... 38 Pensions, Bonuses, and Veterans' Relief 2 2010-07-01 2010-07-01 false Life Safety Code capital...

  2. MOSEG code for safety oriented maintenance management Safety of management of maintenance oriented by MOSEG code

    International Nuclear Information System (INIS)

    Torres Valle, Antonio

    2005-01-01

    Full text: One of the main reasons that makes maintenance contribute highly when facing safety problems and facilities availability is the lack of maintenance management systems to solve these fields in a balanced way. Their main setbacks are shown in this paper. It briefly describes the development of an integrating algorithm for a safety and availability-oriented maintenance management by virtue of the MOSEG Win 1.0 code. (author)

  3. IAEA code and safety guides on quality assurance

    International Nuclear Information System (INIS)

    Raisic, N.

    1980-01-01

    In the framework of its programme in safety standards development, the IAEA has recently published a Code of Practice on Quality Assurance for Safety in Nuclear Power Plants. The Code establishes minimum requirements for quality assurance which Member States should use in the context of their own nuclear safety requirements. A series of 10 Safety Guides which describe acceptable methods of implementing the requirements of specific sections of the Code are in preparation. (orig.)

  4. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  5. SAFETY IN THE DESIGN OF SCIENCE LABORATORIES AND BUILDING CODES.

    Science.gov (United States)

    HOROWITZ, HAROLD

    THE DESIGN OF COLLEGE AND UNIVERSITY BUILDINGS USED FOR SCIENTIFIC RESEARCH AND EDUCATION IS DISCUSSED IN TERMS OF LABORATORY SAFETY AND BUILDING CODES AND REGULATIONS. MAJOR TOPIC AREAS ARE--(1) SAFETY RELATED DESIGN FEATURES OF SCIENCE LABORATORIES, (2) LABORATORY SAFETY AND BUILDING CODES, AND (3) EVIDENCE OF UNSAFE DESIGN. EXAMPLES EMPHASIZE…

  6. Home electrical system safety in Italy

    Energy Technology Data Exchange (ETDEWEB)

    Auditor,

    1990-06-01

    Italy, amongst the industrialized countries, has the highest mortality rate due to accidents associated with the improper use or maintenance of home electrical systems. The increasing use of domestic electrical appliances has raised the risk of accidents, especially in homes equipped with out-dated, low-capacity electrical plants and worn wiring. Within this context, this paper reports on the results of survey to establish the worthiness and type of electrical systems in use in a sample of 1,000 residential buildings. The paper then assesses the efficacy of recent normatives designed to increase the safety and efficiency of home electrical installations.

  7. Safety policy in the production of electricity

    International Nuclear Information System (INIS)

    Siddall, E.

    1982-01-01

    When safety is properly understood, defined and quantified, it can be seen that the development of our present industrial civilization has resulted in a progressive improvement in human safety. Increased safety has come with increased wealth in such close association that a high degree of cause-and-effect relationship must be considered. The quantitative relationship between wealth production and safety improvement is derived from different sources of evidence. When this is applied to the wealth production from electricity generation in a standard module of population in an advanced society, a safety benefit is indicated which exceeds the assessed direct risk associated with the electricity generation by orders of magnitude. It appears that a goal or policy intended to confer the greatest safety benefit to the population would result in attitudes and actions diametrically opposite to those which are conventional at the moment

  8. Safety policy in the production of electricity

    International Nuclear Information System (INIS)

    Siddall, E.

    1983-01-01

    When safety is properly understood, defined and quantified, it can be seen that the development of our present industrial civilization has resulted in a progressive and great improvement in human safety which is still continuing. Increased safety has come with increased wealth in such close association that a high degree of cause-and-effect relationship must be considered. The quantitative relationship between wealth production and safety improvement is derived from different sources of evidence. When this is applied to the wealth production from electricity generation in a standard module of population in an advanced society, a safety benefit is indicated which exceeds the assessed direct risk associated with the electricity generation by orders of magnitude. It appears that a goal or policy intended to confer the greatest safety benefit to the population would result in attitudes and actions diametrically opposite to those which are conventional at the moment

  9. New conducted electrical weapons: Electrical safety relative to relevant standards.

    Science.gov (United States)

    Panescu, Dorin; Nerheim, Max; Kroll, Mark W; Brave, Michael A

    2017-07-01

    We have previously published about TASER ® conducted electrical weapons (CEW) compliance with international standards. CEWs deliver electrical pulses that can inhibit a person's neuromuscular control or temporarily incapacitate. An eXperimental Rotating-Field (XRF) waveform CEW and the X2 CEW are new 2-shot electrical weapon models designed to target a precise amount of delivered charge per pulse. They both can deploy 1 or 2 dart pairs, delivered by 2 separate cartridges. Additionally, the XRF controls delivery of incapacitating pulses over 4 field vectors, in a rotating sequence. As in our previous study, we were motivated by the need to understand the cardiac safety profile of these new CEWs. The goal of this paper is to analyze the nominal electrical outputs of TASER XRF and X2 CEWs in reference to provisions of all relevant international standards that specify safety requirements for electrical medical devices and electrical fences. Although these standards do not specifically mention CEWs, they are the closest electrical safety standards and hence give very relevant guidance. The outputs of several TASER XRF and X2 CEWs were measured under normal operating conditions. The measurements were compared against manufacturer specifications. CEWs electrical output parameters were reviewed against relevant safety requirements of UL 69, IEC 60335-2-76 Ed 2.1, IEC 60479-1, IEC 60479-2, AS/NZS 60479.1, AS/NZS 60479.2, IEC 60601-1 and BS EN 60601-1. Our study confirmed that the nominal electrical outputs of TASER XRF and X2 CEWs lie within safety bounds specified by relevant standards.

  10. Code on the safety of nuclear power plants: Governmental organization

    International Nuclear Information System (INIS)

    1988-01-01

    This Code recommends requirements for a regulatory body responsible for regulating the siting, design, construction, commissioning, operation and decommissioning of nuclear power plants for safety. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  11. Validation and verification plan for safety and PRA codes

    International Nuclear Information System (INIS)

    Ades, M.J.; Crowe, R.D.; Toffer, H.

    1991-04-01

    This report discusses a verification and validation (V ampersand V) plan for computer codes used for safety analysis and probabilistic risk assessment calculations. The present plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River Office (DOE-SRO) to bring the essential safety analysis and probabilistic risk assessment codes in compliance with verification and validation requirements

  12. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  13. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  14. Electrical Safety and Arc Flash Protections

    Energy Technology Data Exchange (ETDEWEB)

    R. Camp

    2008-03-04

    Over the past four years, the Electrical Safety Program at PPPL has evolved in addressing changing regulatory requirements and lessons learned from accident events, particularly in regards to arc flash hazards and implementing NFPA 70E requirements. This presentation will discuss PPPL's approaches to the areas of electrical hazards evaluation, both shock and arc flash; engineered solutions for hazards mitigation such as remote racking of medium voltage breakers, operational changes for hazards avoidance, targeted personnel training and hazard appropriate personal protective equipment. Practical solutions for nominal voltage identification and zero voltage checks for lockout/tagout will also be covered. Finally, we will review the value of a comprehensive electrical drawing program, employee attitudes expressed as a personal safety work ethic, integrated safety management, and sustained management support for continuous safety improvement.

  15. Electrical Safety and Arc Flash Protections

    International Nuclear Information System (INIS)

    Camp, R.

    2008-01-01

    Over the past four years, the Electrical Safety Program at PPPL has evolved in addressing changing regulatory requirements and lessons learned from accident events, particularly in regards to arc flash hazards and implementing NFPA 70E requirements. This presentation will discuss PPPL's approaches to the areas of electrical hazards evaluation, both shock and arc flash; engineered solutions for hazards mitigation such as remote racking of medium voltage breakers, operational changes for hazards avoidance, targeted personnel training and hazard appropriate personal protective equipment. Practical solutions for nominal voltage identification and zero voltage checks for lockout/tagout will also be covered. Finally, we will review the value of a comprehensive electrical drawing program, employee attitudes expressed as a personal safety work ethic, integrated safety management, and sustained management support for continuous safety improvement.

  16. Reactor safety computer code development at INEL

    International Nuclear Information System (INIS)

    Johnsen, G.W.

    1985-01-01

    This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

  17. Code on the safety of civilian nuclear fuel cycle installations

    International Nuclear Information System (INIS)

    1996-01-01

    The 'Code' was promulgated by the National Nuclear Safety Administration (NSSA) on June 17, 1993, which is applicable to civilian nuclear fuel fabrication, processing, storage and reprocessing installations, not including the safety requirements for the use of nuclear fuel in reactors. The contents of the 'Code' involve siting, design, construction, commissioning, operation and decommissioning of fuel cycle installation. The NNSA shall be responsible for the interpretation of this 'Code'

  18. Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

    CERN Document Server

    Marshall, T; Topilski, L; Merrill, B

    2002-01-01

    For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

  19. Electrical Safety During a Hurricane

    Centers for Disease Control (CDC) Podcasts

    2006-08-10

    Power outages and flooding can cause electrical hazards. Never touch a downed power line or anything in contact with one.  Created: 8/10/2006 by Emergency Communications System.   Date Released: 10/22/2007.

  20. Method of accounting for code safety valve setpoint drift in safety analyses

    International Nuclear Information System (INIS)

    Rousseau, K.R.; Bergeron, P.A.

    1989-01-01

    In performing the safety analyses for transients that result in a challenge to the reactor coolant system (RCS) pressure boundary, the general acceptance criterion is that the peak RCS pressure not exceed the American Society of Mechanical Engineers limit of 110% of the design pressure. Without crediting non-safety-grade pressure mitigating systems, protection from this limit is mainly provided by the primary and secondary code safety valves. In theory, the combination of relief capacity and setpoints for these valves is designed to provide this protection. Generally, banks of valves are set at varying setpoints staggered by 15- to 20-psid increments to minimize the number of valves that would open by an overpressure challenge. In practice, however, when these valves are removed and tested (typically during a refueling outage), setpoints are sometimes found to have drifted by >50 psid. This drift should be accounted for during the performance of the safety analysis. This paper describes analyses performed by Yankee Atomic Electric Company (YAEC) to account for setpoint drift in safety valves from testing. The results of these analyses are used to define safety valve operability or acceptance criteria

  1. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  2. Safety culture development in nuclear electric plc

    International Nuclear Information System (INIS)

    Gibson, G.P.; Low, M.B.J.

    1995-01-01

    Nuclear Electric plc (NE) has always given the highest priority to safety. However, past emphasis has been directed towards ensuring safety thorough engineering design and hazard control procedures. Whilst the company did achieve high safety standards, particularly with respect to accidents, it was recognized that further improvements could be obtained. Analysis of the safety performance across a wide range of industries showed that the key to improving safety performance lay in developing a strong safety culture within the company. Over the last five years, NE has made great strides to improve its safety culture. This has resulted in a considerable improvement in its measured safety performance indicators, such as the number of incidents at international nuclear event scale (INES) rating 1, the number of lost time accidents and the collective radiation dose. However, despite this success, the company is committed to further improvement and a means by which this process becomes self-sustaining. In this way the company will achieve its prime goal, to ''ensure the safety of people, plant and the environment''. The paper provides an overview of the development of safety culture in NE since its formation in November 1989. It describes the research and international developments that have influenced the company's understanding of safety culture, the key initiatives that the company has undertaken to enhance its safety culture and the future initiatives being considered to ensure continual improvement. (author). 5 refs, 2 figs, 2 tabs

  3. Uniform emergency codes: will they improve safety?

    Science.gov (United States)

    2005-01-01

    There are pros and cons to uniform code systems, according to emergency medicine experts. Uniformity can be a benefit when ED nurses and other staff work at several facilities. It's critical that your staff understand not only what the codes stand for, but what they must do when codes are called. If your state institutes a new system, be sure to hold regular drills to familiarize your ED staff.

  4. Application of coupled codes for safety analysis and licensing issues

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    2006-01-01

    An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

  5. Consequences of electricity deregulation on nuclear safety

    International Nuclear Information System (INIS)

    Podjavorsek, M.

    2007-01-01

    The evolution of deregulation of electricity market started a couple of years ago and has not been finished yet. Deregulation causes increased pressure to reduce the costs of electricity generation. This presents a new challenge to regulatory bodies. They have to assess the impact of these changes on the safety of nuclear power plants. Accordingly, it is important to identify the risks to the nuclear power industry resulting from the deregulation. Today's trend is that the number of electricity generating power companies will be reduced in Europe and also in Slovenia due to tough competition in the electricity market. The electricity price has decreased after the introduction of the deregulated market in most countries. This has been also the main reason for less investment to new generating capacities since the price has been lower than the generation costs. Investment problems are also present for the existing units, because of danger of inappropriate maintenance and reduction of the number of staff and their qualifications below the desired level that leads to loss of institutional memory. It is expected that only the biggest companies can stand the consequences of competition in electricity prices and consequential pressure to reduce the cost. In order to review the impact of deregulation of the electricity market some relevant points are discussed in this paper such as the need to cut costs of companies by reducing the number of their activities and increasing the efficiency in the remaining activities and /or outsourcing of activities, power station operating regime, safety culture, grid reliability, reliability and safety of operation, increased number of transients, ageing of components, outage duration, extended cycle and response of nuclear regulators. From a regulatory point of view the impact of deregulation on nuclear safety is an important issue. This paper also discusses analyses and evaluations of this impact and proposes some measures how to

  6. SAFETY ALERT: Electrical insulation defect on safety helmets

    CERN Multimedia

    HSE Unit

    2013-01-01

    Contrarily to the information provided until 31 May 2013, some “Euro Protection” safety helmets do not respect any of the requirements for electrical insulation.   This alert concerns the safety helmets identified under the following SCEM numbers: 50.43.30.050.4 white 50.43.30.060.2 yellow 50.43.30.070.0 blue This amounts up to several hundreds of helmets on the CERN site. People who need to wear an electrically insulated safety helmet for their activities, must from now on acquire a duly insulated item to be found on the CERN store under the following SCEM numbers: 50.43.30.210.6: Petzl Vertex ST Helmet (without vent) 50.43.30.300.1: IDRA Helmet with a visor for electrical work As for the people who do not need to wear an electrically insulated helmet for their activities, they can continue working with the aforementioned helmets. For your information, please take note of the maximum use limit of each helmet: “Euro Protection” Safety Helme...

  7. Fire safety requirements for electrical cables towards nuclear reactor safety

    International Nuclear Information System (INIS)

    Raju, M.R.

    2002-01-01

    Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

  8. Heat Transfer treatment in computer codes for safety analysis

    International Nuclear Information System (INIS)

    Jerele, A.; Gregoric, M.

    1984-01-01

    Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)

  9. Electrical safety in health care area

    International Nuclear Information System (INIS)

    Amer, G.M.

    2011-01-01

    An electrical safety in health care area is necessary to protect patients and staff from potential electrical hazards.Functional, accurate and safe clinical equipment is an essential requirement in the provision of health services. Well-maintained equipment will give clinicians greater confidence in the reliability of its performance and contribute to a high standard of client care. Clinical equipment, like all health services, requires annual or periodic servicing of medical equipment. In addition to planned servicing and preventative maintenance, there may be the unexpected failure of medical (and other) equipment, necessitating repair. In general, clinical equipment that has an electrical power source and has direct contact with the client must be serviced as a first priority. In this presentation, a review of the main concepts related to the electrical safety in health area,theinternational standard, the distribution of electric power in hospital and protection against shockwill be introduced. Protection system in hospital will be presented in its two ways: inpower distribution in hospitaland inbiomedical equipment design,finally the optimum maintenance technology and safety tests in health care areawill presented also.

  10. Computer codes for level 1 probabilistic safety assessment

    International Nuclear Information System (INIS)

    1990-06-01

    Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs

  11. Code of practice for safety in laboratory - non ionising radiation

    International Nuclear Information System (INIS)

    Ramli Jaya; Mohd Yusof Mohd Ali; Khoo Boo Huat; Khatijah Hashim

    1995-01-01

    The code identifies the non-ionizing radiation encountered in laboratories and the associated hazards. The code is intended as a laboratory standard reference document for general information on safety requirements relating to the usage of non-ionizing radiations in laboratories. The nonionizing radiations cover in this code, namely, are ultraviolet radiation, visible light, radio-frequency radiation, lasers, sound waves and ultrasonic radiation. (author)

  12. Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

    International Nuclear Information System (INIS)

    D'Ardenne, W.H.; Sherwood, G.G.

    1985-01-01

    The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)

  13. Code of Conduct on the Safety of Research Reactors

    International Nuclear Information System (INIS)

    2006-09-01

    The Board of Governors of the International Atomic Energy Agency (IAEA) adopted the Code of Conduct on the Safety of Research Reactors on 8 March 2004. The Board's action was the culmination of several years of work to develop the Code and obtain a consensus on its provisions. The process leading to the Code began in 1998, when the International Nuclear Safety Advisory Group (INSAG) informed the Director General of concerns about the safety of research reactors. In 2000, INSAG recommended that the Secretariat begin developing an international protocol or a similar legal instrument to address those concerns. In September 2000, in resolution GC(44)/RES/14, the General Conference requested the Secretariat ''within its available resources, to continue work on exploring options to strengthen the international nuclear safety arrangements for civil research reactors, taking due account of input from INSAG and the views of other relevant bodies''. A working group convened by the Secretariat pursuant to that request recommended that ''the Agency consider establishing an international action plan for research reactors'' and that the action plan include preparation of a Code of Conduct ''that would clearly establish the desirable attributes for management of research reactor safety''. In September 2001, the Board requested that the Secretariat develop and implement, in conjunction with Member States, an international research reactor safety enhancement plan which included preparation of a Code of Conduct on the Safety of Research Reactors. Subsequently, in resolution GC(45)/RES/10.A, the General Conference endorsed the Board's request. Pursuant to that request, a Code of Conduct on the Safety of Research Reactors was drafted at two meetings of an Open-ended Working Group of Legal and Technical Experts. This draft Code of Conduct was circulated to all Member States for comment. On the basis of the responses received, a revised draft of the Code was prepared by the Secretariat

  14. Lightning protection design of nuclear power plants. KTA safety code, version 6/99

    International Nuclear Information System (INIS)

    1999-06-01

    This KTA safety code does not cover calculation methods for determination of lightning-induced voltage inputs to control room systems within the reactor building, as the literature presents a variety of applicable methods, which however cannot be directly applied to any power plant, due to the great diversity of geometries of the electrical systems and control room systems in nuclear power plants. Compliance with the design requirements of this safety code for shielding of buildings, and installation and shielding of cables, can be considered to offer the appropriate protection. (orig./CB) [de

  15. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

    2006-08-01

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

  16. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  17. Italian electricity supply contracts optimization: ECO computer code

    International Nuclear Information System (INIS)

    Napoli, G.; Savelli, D.

    1993-01-01

    The ECO (Electrical Contract Optimization) code written in the Microsoft WINDOWS 3.1 language can be handled with a 286 PC and a minimum of RAM. It consists of four modules, one for the calculation of ENEL (Italian National Electricity Board) tariffs, one for contractual time-of-use tariffs optimization, a table of tariff coefficients, and a module for monthly power consumption calculations based on annual load diagrams. The optimization code was developed by ENEA (Italian Agency for New Technology, Energy and the Environment) to help Italian industrial firms comply with new and complex national electricity supply contractual regulations and tariffs. In addition to helping industrial firms determine optimum contractual arrangements, the code also assists them in optimizing their choice of equipment and production cycles

  18. Development and improvement of safety analysis code for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  19. A finite element code for electric motor design

    Science.gov (United States)

    Campbell, C. Warren

    1994-01-01

    FEMOT is a finite element program for solving the nonlinear magnetostatic problem. This version uses nonlinear, Newton first order elements. The code can be used for electric motor design and analysis. FEMOT can be embedded within an optimization code that will vary nodal coordinates to optimize the motor design. The output from FEMOT can be used to determine motor back EMF, torque, cogging, and magnet saturation. It will run on a PC and will be available to anyone who wants to use it.

  20. RCC-E a Design Code for I and C and Electrical Systems

    International Nuclear Information System (INIS)

    Haure, J.M.

    2015-01-01

    The paper deals with the stakes and strength of the RCC-E code applicable to Electrical and Instrumentation and control systems and components as regards dealing with safety class functions. The document is interlacing specifications between Owners, safety authorities, designers, and suppliers IAEA safety guides and IEC standards. The code is periodically updated and published by French Society for Design and Construction rules for Nuclear Island Components (AFCEN). The code is compliant with third generation PWR nuclear islands and aims to suit with national regulations as needed in a companion document. The Feedback experience of Fukushima and the licensing of UKEPR in the framework of Generic Design Assessment are lessons learnt that should be considered in the upgrading of the code. The code gathers a set of requirements and relevant good practices of several PWR design and construction practices related to the electrical and I and C systems and components, and electrical engineering documents dealing with systems, equipment and layout designs. Comprehensive statement including some recent developments will be provided about: - Offsite and onsite sources requirements including sources dealing the total loss of off sites and main onsite sources. - Highlights of a relevant protection level against high frequencies disturbances emitted by lightning strokes, Interfaces data used by any supplier or designer such as site data, rooms temperature, equipment maximum design temperature, alternative current and direct current electrical network voltages and frequency variation ranges, environmental conditions decoupling data, - Environmental Qualification process including normal, mild (earthquake resistant), harsh and severe accident ambient conditions. A suit made approach based on families, which are defined as a combination of mission time, duration and abnormal conditions (pressure, temperature, radiation), enables to better cope with Environmental Qualifications

  1. Safety, codes and standards for hydrogen installations. Metrics development and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Harris, Aaron P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dedrick, Daniel E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); San Marchi, Christopher W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-04-01

    Automakers and fuel providers have made public commitments to commercialize light duty fuel cell electric vehicles and fueling infrastructure in select US regions beginning in 2014. The development, implementation, and advancement of meaningful codes and standards is critical to enable the effective deployment of clean and efficient fuel cell and hydrogen solutions in the energy technology marketplace. Metrics pertaining to the development and implementation of safety knowledge, codes, and standards are important to communicate progress and inform future R&D investments. This document describes the development and benchmarking of metrics specific to the development of hydrogen specific codes relevant for hydrogen refueling stations. These metrics will be most useful as the hydrogen fuel market transitions from pre-commercial to early-commercial phases. The target regions in California will serve as benchmarking case studies to quantify the success of past investments in research and development supporting safety codes and standards R&D.

  2. Independent peer review of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Boyack, B.E.; Jenks, R.P.

    1993-01-01

    A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC

  3. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

  4. Code conversion for system design and safety analysis of NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)

  5. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  6. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  7. Method for quantitative assessment of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Dearien, J.A.; Davis, C.B.; Matthews, L.J.

    1979-01-01

    A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

  8. IAEA codes and guides for safety of nuclear power plants

    International Nuclear Information System (INIS)

    Raisic, N.

    1980-01-01

    The objectives and scope of the Agency's programme of nuclear safety standards are described and the role of these documents in regulation of nuclear power im Member States is discussed. For each of the five areas of safety standards development, i.e. siting, design, operation, quality assurance and governmental organization, a set of principles underlying requirements and recommendations contained in the Code of Practice and Safety Guides will be presented. Safety Guides in each of the five areas will be reviewed in respect of the scope and content. A consideration will be given to the future development of the safety standards and to the revision and updating of the published documents. (orig./RW)

  9. Photovoltaic power systems and the National Electrical Code: Suggested practices

    Energy Technology Data Exchange (ETDEWEB)

    Wiles, J. [New Mexico State Univ., Las Cruces, NM (United States). Southwest Technology Development Inst.

    1996-12-01

    This guide provides information on how the National Electrical Code (NEC) applies to photovoltaic systems. The guide is not intended to supplant or replace the NEC; it paraphrases the NEC where it pertains to photovoltaic systems and should be used with the full text of the NEC. Users of this guide should be thoroughly familiar with the NEC and know the engineering principles and hazards associated with electrical and photovoltaic power systems. The information in this guide is the best available at the time of publication and is believed to be technically accurate; it will be updated frequently. Application of this information and results obtained are the responsibility of the user.

  10. Photovoltaic Power Systems and the National Electrical Code: Suggested Practices

    Energy Technology Data Exchange (ETDEWEB)

    None

    2002-02-01

    This guide provides information on how the National Electrical Code (NEC) applies to photovoltaic systems. The guide is not intended to supplant or replace the NEC; it paraphrases the NEC where it pertains to photovoltaic systems and should be used with the full text of the NEC. Users of this guide should be thoroughly familiar with the NEC and know the engineering principles and hazards associated with electrical and photovoltaic power systems. The information in this guide is the best available at the time of publication and is believed to be technically accurate; it will be updated frequently.

  11. ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES

    Energy Technology Data Exchange (ETDEWEB)

    Poole, B R; Nelson, S D; Langdon, S

    2005-05-05

    The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes.

  12. ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES

    International Nuclear Information System (INIS)

    Poole, B R; Nelson, S D; Langdon, S

    2005-01-01

    The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes

  13. Development and assessment of best estimate integrated safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

    2007-03-15

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

  14. Development and assessment of best estimate integrated safety analysis code

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu

    2007-03-01

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published

  15. Vectorization of the KENO V.a criticality safety code

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Dodds, H.L.; Petrie, L.M.

    1991-01-01

    The development of the vector processor, which is used in the current generation of supercomputers and is beginning to be used in workstations, provides the potential for dramatic speed-up for codes that are able to process data as vectors. Unfortunately, the stochastic nature of Monte Carlo codes prevents the old scalar version of these codes from taking advantage of the vector processors. New Monte Carlo algorithms that process all the histories undergoing the same event as a batch are required. Recently, new vectorized Monte Carlo codes have been developed that show significant speed-ups when compared to the scalar version of themselves or equivalent codes. This paper discusses the vectorization of an already existing and widely used criticality safety code, KENO V.a All the changes made to KENO V.a are transparent to the user making it possible to upgrade from the standard scalar version of KENO V.a to the vectorized version without learning a new code

  16. Environmental codes of practice for steam electric power generation

    International Nuclear Information System (INIS)

    1985-03-01

    The Design Phase Code is one of a series of documents being developed for the steam electric power generation industry. This industry includes fossil-fuelled stations (gas, oil and coal-fired boilers), and nuclear-powered stations (CANDU heavy water reactors). In this document, environmental concerns associated with water-related and solid waste activities of steam electric plants are discussed. Design recommendations are presented that will minimize the detrimental environmental effects of once-through cooling water systems, of wastewaters discharged to surface waters and groundwaters, and of solid waste disposal sites. Recommendations are also presented for the design of water-related monitoring systems and programs. Cost estimates associated with the implementation of these recommendations are included. These technical guides for new or modified steam electric stations are the result to consultation with a federal-provincial-industry task force

  17. Safety in nuclear power plant siting. A code of practice

    International Nuclear Information System (INIS)

    1978-01-01

    This publication is brought out within the framework of establishing Codes of Practice and Safety Guides for nuclear power plants: NUSS programme. The scope of the document encompasses site and site-plant interaction factors related to operational states and accident conditions. The purpose of the Code is to give criteria and procedures to be applied as appropriate to operational states and accident conditions, including those which could lead to emergency situations. This Code is mainly concerned with severe events of low probability which relate to the siting of nuclear power plants and have to be considered in designing a particular nuclear power plant. Annex: Examples of natural and man-made events relevant for design basis evaluation

  18. Application of software to development of reactor-safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1980-09-01

    Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis

  19. Advanced Electric and Magnetic Material Models for FDTD Electromagnetic Codes

    CERN Document Server

    Poole, Brian R; Nelson, Scott D

    2005-01-01

    The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which requires nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes an...

  20. Implementation of probabilistic safety concepts in international codes

    International Nuclear Information System (INIS)

    Borges, J.F.

    1977-01-01

    Recent progress in the implementation of safety concepts in international structure codes is briefly presented. Special attention is paid to the work of the Joint-Committee on Structural Safety. The discussion is centered on some problems such as: safety differentiation, definition and combination of actions, spaces for checking safety and non-linear structural behaviour. When discussing safety differentiation it should be considered that the total probability of failure derives from a theoretical probability of failure and a probability of failure due to error and gross negligence. Optimization of design criteria should take into account both causes of failure. The quantification of reliability implies a probabilistic idealization of all basic variables. Steps taken to obtain an improved definition of different types of actions and rules for their combination are described. Safety checking can be carried out in terms of basic variables, action-effects, or any other suitable variable. However, the advantages and disadvantages of the different types of formulation should be discussed, particularly in the case of non-linear structural behaviour. (orig.) [de

  1. 77 FR 24560 - National Highway Traffic Safety Administration Electric Vehicle Safety Technical Symposium

    Science.gov (United States)

    2012-04-24

    ... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Electric Vehicle... discuss safety considerations for electric vehicles powered by lithium-ion (Li-ion) batteries. The... vehicles. Electric vehicles show great promise as an innovative and fuel- efficient option for American...

  2. Partial Safety Factors and Target Reliability Level in Danish Structural Codes

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard; Hansen, J. O.; Nielsen, T. A.

    2001-01-01

    The partial safety factors in the newly revised Danish structural codes have been derived using a reliability-based calibration. The calibrated partial safety factors result in the same average reliability level as in the previous codes, but a much more uniform reliability level has been obtained....... The paper describes the code format, the stochastic models and the resulting optimised partial safety factors....

  3. The history of 20 years of electrical safety

    International Nuclear Information System (INIS)

    1994-10-01

    This book describes the establishment of Korea electrical safety corporation including the background and procession of the safety corporation, based on the building to make infrastructure and solve the difficulty, the time of growth in stability and development. Also it deals with personnel management and financial of the company. Next, it describes the operation of the each department, that is, regulant check, management agency of electrical safety, speciation task, and check for the request. Last, it deals with the voluntary service and safe management and technical development of the electrical safety check and study.

  4. Nuclear power plants. Electrical equipment of the safety system. Qualification

    International Nuclear Information System (INIS)

    2001-01-01

    This International Standard applies to electrical parts of safety systems employed at nuclear power plants, including components and equipment of any interface whose failure could affect unfavourably properties of the safety system. The standard also applies to non-electrical safety-related interfaces. Furthermore, the standard describes the generic process of qualification certification procedures and methods of qualification testing and related documentation. (P.A.)

  5. Changing An Electrical Safety Culture - The Importance of Understanding Why.

    Energy Technology Data Exchange (ETDEWEB)

    Waters, Richard Thomas [Idaho National Laboratory

    2015-12-01

    Abstract – Electrical workers, regardless of experience, are faced with a major barrier when first introduced to NFPA 70E, “The Standard for Electrical Safety in the Workplace,” and an erroneous electrical safety culture pre-exists. This paper describes, from the author’s point of view, the barrier that he and other electrical workers have experienced and his insight into overcoming the barrier. The author in conclusion will present a series of techniques that can be used to assist other electrical workers in overcoming the barrier.

  6. Application of software engineering to development of reactor safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1981-01-01

    Software Engineering, which is a systematic methodology by which a large scale software development project is partitioned into manageable pieces, has been applied to the development of LMFBR safety codes. The techniques have been applied extensively in the business and aerospace communities and have provided an answer to the drastically increasing cost of developing and maintaining software. The five phases of software engineering (Survey, Analysis, Design, Implementation, and Testing) were applied in turn to development of these codes, along with Walkthroughs (peer review) at each stage. The application of these techniques has resulted in SUPERIOR SOFTWARE which is well documented, thoroughly tested, easy to modify, easier to use and maintain. The development projects have resulted in lower overall cost. (orig.) [de

  7. Fire-safety engineering and performance-based codes

    DEFF Research Database (Denmark)

    Sørensen, Lars Schiøtt

    project administrators, etc. The book deals with the following topics: • Historical presentation on the subject of fire • Legislation and building project administration • European fire standardization • Passive and active fire protection • Performance-based Codes • Fire-safety Engineering • Fundamental......Fire-safety Engineering is written as a textbook for Engineering students at universities and other institutions of higher education that teach in the area of fire. The book can also be used as a work of reference for consulting engineers, Building product manufacturers, contractors, building...... thermodynamics • Heat exchange during the fire process • Skin burns • Burning rate, energy release rate and design fires • Proposal to Risk-based design fires • Proposal to a Fire scale • Material ignition and flame spread • Fire dynamics in buildings • Combustion products and toxic gases • Smoke inhalation...

  8. Change to CERN Safety Rules: Abolition of Safety Code A7

    CERN Multimedia

    2016-01-01

    As from 3 June 2016 Safety Code A7 “Road traffic at CERN” is abolished.   CERN's current practice to follow French or Swiss road traffic regulations on the corresponding parts of the CERN site will continue to apply. HSE Unit

  9. Coding conspecific identity and motion in the electric sense.

    Directory of Open Access Journals (Sweden)

    Na Yu

    Full Text Available Interactions among animals can result in complex sensory signals containing a variety of socially relevant information, including the number, identity, and relative motion of conspecifics. How the spatiotemporal properties of such evolving naturalistic signals are encoded is a key question in sensory neuroscience. Here, we present results from experiments and modeling that address this issue in the context of the electric sense, which combines the spatial aspects of vision and touch, with the temporal aspects of audition. Wave-type electric fish, such as the brown ghost knifefish, Apteronotus leptorhynchus, used in this study, are uniquely identified by the frequency of their electric organ discharge (EOD. Multiple beat frequencies arise from the superposition of the EODs of each fish. We record the natural electrical signals near the skin of a "receiving" fish that are produced by stationary and freely swimming conspecifics. Using spectral analysis, we find that the primary beats, and the secondary beats between them ("beats of beats", can be greatly influenced by fish swimming; the resulting motion produces low-frequency envelopes that broaden all the beat peaks and reshape the "noise floor". We assess the consequences of this motion on sensory coding using a model electroreceptor. We show that the primary and secondary beats are encoded in the afferent spike train, but that motion acts to degrade this encoding. We also simulate the response of a realistic population of receptors, and find that it can encode the motion envelope well, primarily due to the receptors with lower firing rates. We discuss the implications of our results for the identification of conspecifics through specific beat frequencies and its possible hindrance by active swimming.

  10. Electrical Safety for Non-Electricians

    Science.gov (United States)

    ... handled cement finishing floats • Metal ladders • Raised dump truck beds • Scaffolding But electrical hazards are also at ... must be grounded. Your employer must check all electric systems, including wiring and switches, to be sure ...

  11. Safety in nuclear power plant operation, including commissioning and decommissioning. A code of practice

    International Nuclear Information System (INIS)

    1978-01-01

    Safe operation of a nuclear power plant postulates satisfactory siting, design, construction and commissioning, together with proper management and operation of the plant. This Code of Practice deals with the safety aspects of management, commissioning, operation and decommissioning of the plant. It forms part of the Agency's programme, referred to as the NUSS programme, for establishing Codes of Practice and Safety Guides relating to land-based stationary thermal neutron power plants. It has been prepared for the use of those responsible for the operation of stationary nuclear power plants, the main function of which is the generation of electrical and/or thermal power, and for the use of those responsible for regulating the operation of such plants. It is not intended for application to reactors used solely for experimental or research purposes. The provisions in the Code are designed to provide assurance that operational activities are carried out without undue radiological hazard to the general public and to persons on the site. It should be understood that the provisions in the Code set forth minimum requirements which shall be met in order to achieve safe operation of a nuclear power plant

  12. WSRC approach to validation of criticality safety computer codes

    International Nuclear Information System (INIS)

    Finch, D.R.; Mincey, J.F.

    1991-01-01

    Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K eff ) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope 236 U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed

  13. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  14. Electrical Safety Program: Nonelectrical Crafts at LANL, Live #12175

    Energy Technology Data Exchange (ETDEWEB)

    Glass, George [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-12-22

    Los Alamos National Laboratory (LANL) and the federal government require those working with or near electrical equipment to be trained on electrical hazards and how to avoid them. Although you might not be trained to work on electrical systems, your understanding of electricity, how it can hurt you, and what precautions to take when working near electricity could save you or others from injury or death. This course, Electrical Safety Program: Nonelectrical Crafts at LANL (12175), provides knowledge of basic electrical concepts, such as current, voltage, and resistance, and their relationship to each other. You will learn how to apply these concepts to safe work practices while learning about the dangers of electricity—and associated hazards—that you may encounter on the job. The course also discusses what you can do to prevent electrical accidents and what you should do in the event of an electrical emergency. The LANL Electrical Safety Program is defined by LANL Procedure (P) 101-13. An electrical safety officer (ESO) is well versed in this document and should be consulted regarding electrical questions. Appointed by the responsible line manager (RLM), ESOs can tell you if a piece of equipment or an operation is safe or how to make it safe.

  15. Interface design of VSOP'94 computer code for safety analysis

    International Nuclear Information System (INIS)

    Natsir, Khairina; Andiwijayakusuma, D.; Wahanani, Nursinta Adi; Yazid, Putranto Ilham

    2014-01-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects

  16. Interface design of VSOP'94 computer code for safety analysis

    Science.gov (United States)

    Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

    2014-09-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

  17. Safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1976-01-01

    The Codes of Practice and Safety Guides that are being developed by the International Atomic Energy Agency are divided in five topical areas: Governmental Organization, Siting, Design, Operation and Quality Assurance. In each area, a scientific secretary is responsible for developing the documents and five Technical Review Committees composed of 10 to 12 experts from various Members Countries revise the drafts at different stages. A Senior Advisory Group supervises the entire programme and revises the document. A scientific co-ordinator is responsible for the co-ordination within the programme with other sections of the IAEA, and with other international organizations. In preparing a document, information on the practice adopted by Member States is collected, a group of experts is convened for preparing a preliminary draft on the basis of this material and the draft is then reviewed by the appropriate Technical Review Committee. The document is translated into various languages, reviewed by the Senior Advisory Group and sent to Member States for comments. After the comments of Member States have been received, the Technical Review Committee and then the Senior Advisory Group are convened again for the final revision of the document. Some 25 drafts, are in different stages of development. The preparation of a document in its final form takes about two years. The programme started in 1975 and to date most of the safety codes and a few safety guides have been sent to Member States for comments. These documents will have gone through the entire development procedure by early 1977. The Senior Advisory Groups and the Technical Review Committees meet on the average four times a year for a week at a time. Until now these meetings have been mainly concerned with the development of new documents or with that part of the procedure which precedes the transmission of the draft to Member States for comments. The next series of meetings will deal with the revisions needed to

  18. ELECTRICAL SAFETY IMPROVEMENT PROJECT A COMPLEX WIDE TEAMING INITIATIVE

    Energy Technology Data Exchange (ETDEWEB)

    GRAY BJ

    2007-11-26

    This paper describes the results of a year-long project, sponsored by the Energy Facility Contractors Group (EFCOG) and designed to improve overall electrical safety performance throughout Department of Energy (DOE)-owned sites and laboratories. As evidenced by focused metrics, the Project was successful primarily due to the joint commitment of contractor and DOE electrical safety experts, as well as significant support from DOE and contractor senior management. The effort was managed by an assigned project manager, using classical project-management principles that included execution of key deliverables and regular status reports to the Project sponsor. At the conclusion of the Project, the DOE not only realized measurable improvement in the safety of their workers, but also had access to valuable resources that will enable them to do the following: evaluate and improve electrical safety programs; analyze and trend electrical safety events; increase electrical safety awareness for both electrical and non-electrical workers; and participate in ongoing processes dedicated to continued improvement.

  19. Validation of thermal hydraulic codes for fusion reactors safety

    International Nuclear Information System (INIS)

    Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

    2006-01-01

    A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

  20. Improving radiopharmaceutical supply chain safety by implementing bar code technology.

    Science.gov (United States)

    Matanza, David; Hallouard, François; Rioufol, Catherine; Fessi, Hatem; Fraysse, Marc

    2014-11-01

    The aim of this study was to describe and evaluate an approach for improving radiopharmaceutical supply chain safety by implementing bar code technology. We first evaluated the current situation of our radiopharmaceutical supply chain and, by means of the ALARM protocol, analysed two dispensing errors that occurred in our department. Thereafter, we implemented a bar code system to secure selected key stages of the radiopharmaceutical supply chain. Finally, we evaluated the cost of this implementation, from overtime, to overheads, to additional radiation exposure to workers. An analysis of the events that occurred revealed a lack of identification of prepared or dispensed drugs. Moreover, the evaluation of the current radiopharmaceutical supply chain showed that the dispensation and injection steps needed to be further secured. The bar code system was used to reinforce product identification at three selected key stages: at usable stock entry; at preparation-dispensation; and during administration, allowing to check conformity between the labelling of the delivered product (identity and activity) and the prescription. The extra time needed for all these steps had no impact on the number and successful conduct of examinations. The investment cost was reduced (2600 euros for new material and 30 euros a year for additional supplies) because of pre-existing computing equipment. With regard to the radiation exposure to workers there was an insignificant overexposure for hands with this new organization because of the labelling and scanning processes of radiolabelled preparation vials. Implementation of bar code technology is now an essential part of a global securing approach towards optimum patient management.

  1. Investigations of safety risks in converted electric vehicles

    NARCIS (Netherlands)

    Bolech, M.; Foster, D.L.; Lange, R. de; Rodarius, C.

    2010-01-01

    Within the departments Environmentally Sustainable Transport and Automotive of TNO (Netherlands organisation for applied scientific research) several projects investigating safety aspects of electric vehicles have been conducted, including one in cooperation with KEMA and RDW of the Netherlands.

  2. 76 FR 77549 - Colorado River Indian Tribes-Amendment to Health & Safety Code, Article 2. Liquor

    Science.gov (United States)

    2011-12-13

    ... Health & Safety Code, Article 2. Liquor AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice. SUMMARY: This notice publishes the amendment to the Colorado River Tribal Health and Safety Code, Article... Code, Article 2, Liquor by Ordinance No. 10-03 on December 13, 2010. This notice is published in...

  3. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  4. Assessment of Electrical Safety Beliefs and Practices: A Case Study

    Directory of Open Access Journals (Sweden)

    S. Boubaker

    2017-12-01

    Full Text Available In this paper, the electrical safety beliefs and practices in Hail region, Saudi Arabia, have been assessed. Based on legislative recommendations and rules applied in Saudi Arabia, on official statistics regarding the electricity-caused accidents and on the analysis of more than 200 photos captured in Hail (related to electrical safety, a questionnaire composed of 36 questions (10 for the respondents information, 16 for the home safety culture and 10 for the electrical devices purchasing culture has been devised and distributed to residents. 228 responses have been collected and analyzed. Using a scale similar to the one adopted for a university student GPA calculation, the electrical safety level (ESL in Hail region has been found to be 0.76 (in a scale of 4 points which is a very low score and indicates a poor electrical safety culture. Several recommendations involving different competent authorities have been proposed. Future work will concern the assessment of safety in industrial companies in Hail region.

  5. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  6. A computer code SPHINCS for sodium fire safety evaluation

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    2000-01-01

    A computer code SPHINCS solves coupled phenomena of thermal-hydraulics and sodium fire based on a multi-zone model. It deals with arbitrary number of rooms each of which is connected mutually by doorway and penetrations. With regard to the combustion phenomena, flame sheet model and liquid droplet combustion model are used for pool and spray fire, respectively, with the chemical equilibrium model using Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermal-hydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The author analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS has been validated with regard to the pool combustion phenomena. Further research needs are identified for the pool spreading modeling considering thermal deformation of liner and measurement of pool fluidity property of a mixture of liquid sodium and reaction products. SPHINCS code is to be used mainly in the safety evaluation of the consequence of sodium fire accident of liquid metal cooled fast reactor. (author)

  7. Legal bases of safety regulations in electrical engineering

    Energy Technology Data Exchange (ETDEWEB)

    Jeiter, W

    1981-12-01

    Apart from the governmental regulations the rule for the prevention of accidents 'Electric plants and equipment' must be observed in order to protect the insurants. Actually, all these regulations do not contain any independent instructions. They rather utilize the VDE regulations and refer to them. The laws of electrical safety engineering are strongly influenced by harmonization efforts particularly within the European Communitties.

  8. Kayenta Township Building & Safety Department, Tribal Green Building Code Summit Presentation

    Science.gov (United States)

    Tribal Green Building Code Summit Presentation by Kayenta Township Building & Safety Department showing how they established the building department, developed a code adoption and enforcement process, and hired staff to carry out the work.

  9. Job safety and awareness analysis of safety implementation among electrical workers in airport service company

    Directory of Open Access Journals (Sweden)

    Putra Perdana Suteja

    2018-01-01

    Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

  10. Compendium of computer codes for the safety analysis of fast breeder reactors

    International Nuclear Information System (INIS)

    1977-10-01

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available

  11. Safety analysis and code development for nuclear fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    We are estimating that the debris containing fuel are piled in the containment and the pressure vessel bottoms of Fukushima-Daiichi NPPs. A radioactive Xe concentration discharged in recriticality is being monitored by utilizing the gas management system set up in NPPs unit 1-3. For this reason, we can confirm the recriticality might not be broken out. However, the debris conditions distributed in the containment vessel and the pressure vessel bottoms are not clear. The internal and external surrounding changes will make recriticality become possible. According to TEPCO's roadmap, TEPCO will launch extracting task within 10 years. Even in the case that the fuel condition changes due to debris relocation and mixture, subcriticality must be secured. Criticality safety analysis with non-uniform effect should therefore be essential for the molten debris. For above reasons, we studies the optimum distributions with some parameters that have a large reactivity change were assessed with OPT-DANT code. Finally, the boron concentration was estimated in order to keep subcriticality. (author)

  12. Probabilist methods applied to electric source problems in nuclear safety

    International Nuclear Information System (INIS)

    Carnino, A.; Llory, M.

    1979-01-01

    Nuclear Safety has frequently been asked to quantify safety margins and evaluate the hazard. In order to do so, the probabilist methods have proved to be the most promising. Without completely replacing determinist safety, they are now commonly used at the reliability or availability stages of systems as well as for determining the likely accidental sequences. In this paper an application linked to the problem of electric sources is described, whilst at the same time indicating the methods used. This is the calculation of the probable loss of all the electric sources of a pressurized water nuclear power station, the evaluation of the reliability of diesels by event trees of failures and the determination of accidental sequences which could be brought about by the 'total electric source loss' initiator and affect the installation or the environment [fr

  13. Radiation and electrical safety systems for PEP

    International Nuclear Information System (INIS)

    Smith, H.; Constant, T.; Crook, K.; Fitch, J.; Taylor, T.

    1981-02-01

    At SLAC, the Personnel Protection System (PPS) protects people from radiation hazards. For PEP, the system has been expanded to include protection against electrical and RF hazards. This paper describes the overall system design, giving particular attention to the novel features not found in similar systems in other areas of SLAC. These include the Restricted Access Mode to allow limited occupancy in the ring while high voltage or RF may be present, the automatic badge reader system for improving the efficiency of entry logging and control, and the solid state lighting control system for switching large lighting loads with minimum electro-magetic interference

  14. Software coding for reliable data communication in a reactor safety system

    International Nuclear Information System (INIS)

    Maghsoodi, R.

    1978-01-01

    A software coding method is proposed to improve the communication reliability of a microprocessor based fast-reactor safety system. This method which replaces the conventional coding circuitry, applies a program to code the data which is communicated between the processors via their data memories. The system requirements are studied and the suitable codes are suggested. The problems associated with hardware coders, and the advantages of software coding methods are discussed. The product code which proves a faster coding time over the cyclic code is chosen as the final code. Then the improvement of the communication reliability is derived for a processor and its data memory. The result is used to calculate the reliability improvement of the processing channel as the basic unit for the safety system. (author)

  15. Nuclear electric power safety, operation, and control aspects

    CERN Document Server

    Knowles, J Brian

    2013-01-01

    Assesses the engineering of renewable sources for commercial power generation and discusses the safety, operation, and control aspects of nuclear electric power From an expert who advised the European Commission and UK government in the aftermath of Three Mile Island and Chernobyl comes a book that contains experienced engineering assessments of the options for replacing the existing, aged, fossil-fired power stations with renewable, gas-fired, or nuclear plants. From geothermal, solar, and wind to tidal and hydro generation, Nuclear Electric Power: Safety, Operation, and Control Aspects ass

  16. Three computer codes for safety and stability of large superconducting magnets

    International Nuclear Information System (INIS)

    Turner, L.R.

    1985-01-01

    For analyzing the safety and stability of large superconducting magnets, three computer codes TASS, SHORTURN, and SSICC have been developed, applicable to bath-cooled magnets, bath-cooled magnets with shorted turns, and magnets with internally cooled conductors respectively. The TASS code is described, and the use of the three codes is reviewed

  17. Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

  18. DEVELOPMENT OF SALES APPLICATION OF PREPAID ELECTRICITY VOUCHER BASED ON ANFROID PLATFORM USING QUICK RESPONSE CODE (QR CODE

    Directory of Open Access Journals (Sweden)

    Ricky Akbar

    2017-09-01

    Full Text Available Perusahaan Listrik Negara (PLN has implemented a smart electricity system or prepaid electricity. The customers pay the electricity voucher first before use the electricity. The token contained in electricity voucher that has been purchased by the customer is inserted into the Meter Prabayar (MPB installed in the location of customers. When a customer purchases a voucher, it will get a receipt that contains all of the customer's identity and the 20-digit of voucher code (token to be entered into MPB as a substitute for electrical energy credit. Receipts obtained by the customer is certainly vulnerable to loss, or hijacked by unresponsible parties. In this study, authors designed and develop an android based application by utilizing QR code technology as a replacement for the receipt of prepaid electricity credit which contains the identity of the customer and the 20-digit voucher code. The application is developed by implemented waterfall methodology. The implementation process of the waterfall methods used, are (1 analysis of functional requirement of the system by conducting a preliminary study and data collection based on field studies and literature, (2 system design by using UML diagrams and Business Process Model Notation (BPMN and Entity Relationship diagram (ERD, (3 design implementation by using OOP (Object Oriented programming technique. Web application is developed by using laravel PHP framework and database MySQL while mobile application is developed by using B4A (4 developed system is tested by using blackbox method testing. Final result of this research is a Web and mobile applications for the sale of electricityvoucher by QR Code technology.

  19. Controlling Electrical Hazards

    National Research Council Canada - National Science Library

    1997-01-01

    ...). In general, OSHA's electrical standards are based on the National Fire Protection Associations Standard NFPA 70E, Electrical Safety Requirements for Employee Workplaces, and in turn, from the National Electrical Code (NEC...

  20. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

  1. Cross-index to DOE-prescribed occupational safety codes and standards

    International Nuclear Information System (INIS)

    1981-01-01

    This Cross-Index volume is the 1981 compilation of detailed information from more than three hundred and fifty DOE prescribed or OSHA referenced industrial safety codes and standards and is revised yearly to provide information from current codes. Condensed data from individual code portions are listed according to reference code, section, paragraph and page. Each code is given a two-digit reference code number or letter in the Contents section (pages C to L) of this volume. This reference code provides ready identification of any code listed in the Cross-Index. The computerized information listings are on the left-hand portion of Cross-Index page; in order to the right of the listing are the reference code letters or numbers, the section, paragraph and page of the referenced code containing expanded information on the individual listing

  2. Twenty sixth DAE safety and occupational health professionals meet: cryogenic safety, electrical safety and ergonomics at work place

    International Nuclear Information System (INIS)

    2009-01-01

    This conference gathered knowledge in safety and occupational health hazards in various fields of nuclear science and technology like radiations, high voltages, ultra-low temperature, ultra-high magnetic fields, electrical breakdown, fire, ergonomics and cryogenic safety at work place. Papers relevant to INIS database are indexed separately

  3. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  4. Cross-index to DOE-prescribed occupational safety codes and standards

    International Nuclear Information System (INIS)

    1982-01-01

    A compilation of detailed information from more than three hundred and fifty DOE-prescribed or OSHA-referenced industrial safety codes and standards is presented. Condensed data from individual code portions are listed according to reference code, section, paragraph and page. A glossary of letter initials/abbreviations for the organizations or documents whose codes or standards are contained in this Cross-Index, is listed

  5. Code of Practice containing definitions for Safety Codes of Practice for nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    This Code provides definitions of the technical terms used in the licensing applications to be submitted to the Turkish Atomic Energy Commission (TAEC), in accordance with national licensing regulations. The Code is based mainly on the International Atomic Energy Agency's Code of Practice on the subject. (NEA) [fr

  6. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  7. Is the international safety management code an organisational tool ...

    African Journals Online (AJOL)

    The birth of the International Management Code for the Safe Operation of Ships and for Pollution Prevention (hereinafter ISM Code) is said to a reaction to the sinking of the Herald of free Enterprise on 6th March 1987.The human element is said to be a generic term used to describe what makes humans behave the way ...

  8. ITER safety studies: starting the quality plan code AINA

    International Nuclear Information System (INIS)

    Dies, J.; Rivas, J. C.; Bargallo, E.

    2010-01-01

    This contribution discusses the implementation of the Quality Plan AINA Code 2.0. The work has affected areas such as the life cycle model, the software reliability plan, the design of the equipment or software tools to use, and has finally produced a new code.

  9. Validation of computer codes used in the safety analysis of Canadian research reactors

    International Nuclear Information System (INIS)

    Bishop, W.E.; Lee, A.G.

    1998-01-01

    AECL has embarked on a validation program for the suite of computer codes that it uses in performing the safety analyses for its research reactors. Current focus is on codes used for the analysis of the two MAPLE reactors under construction at Chalk River but the program will be extended to include additional codes that will be used for the Irradiation Research Facility. The program structure is similar to that used for the validation of codes used in the safety analyses for CANDU power reactors. (author)

  10. Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

    International Nuclear Information System (INIS)

    Lajtman, S.; Feretic, D.; Debrecin, N.

    1989-01-01

    This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

  11. Development of safety analysis codes for light water reactor

    International Nuclear Information System (INIS)

    Akimoto, Masayuki

    1985-01-01

    An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

  12. MOSEG code for safety oriented maintenance management Safety of management of maintenance oriented by MOSEG code; Codigo MOSEG para la gestion de mantenimiento orientada a la seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Torres Valle, Antonio [Instituto Superior de Tecnologias y Ciencias Aplicadas, La Habana (Cuba). Dept. Ingenieria Nuclear]. E-mail: atorres@fctn.isctn.edu.cu; Rivero Oliva, Jose de Jesus [Centro de Gestion de la Informacion y Desarrollo de la Energia (CUBAENERGIA) (Cuba)]. E-mail: jose@cubaenergia.cu

    2005-07-01

    Full text: One of the main reasons that makes maintenance contribute highly when facing safety problems and facilities availability is the lack of maintenance management systems to solve these fields in a balanced way. Their main setbacks are shown in this paper. It briefly describes the development of an integrating algorithm for a safety and availability-oriented maintenance management by virtue of the MOSEG Win 1.0 code. (author)

  13. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  14. Demonstration study on shielding safety analysis code (8)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

    2001-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

  15. An electric mandate. The EU procedure for harmonising cross-border network codes for electricity

    Energy Technology Data Exchange (ETDEWEB)

    Jevnaker, Torbjoerg

    2012-07-01

    The research question addressed in this report is why the EU procedure for developing network codes for electricity was enacted in its particular form. Passed by the EU in 2009, European organisations partly outside of the formal EU structure were given a mandate to make rules that would apply across the EU. This was puzzling given the observed resistance on part of the member states to let go of national control over energy issues. Drawing on institutionalist perspectives, the analysis shows that the procedure would not have been passed without support from and compromise among the Commission, European Parliament and the member states in Council; that parts of the procedure imitated existing practices within related policy areas; that horizontal and vertical specialization within the nation-states along with a Commission actively promoting transnational cooperation changed the feedback mechanisms, which changed the direction of European energy market regulation; and finally, that the new actors played an active role vis-a-vis EU bodies as the latter were legislating on the procedure. (Author)

  16. Self assessment of safety culture in HANARO using the code of conduct on the safety of research reactor by IAEA

    International Nuclear Information System (INIS)

    Lim, I.C.; Hwang, S.Y.; Woo, J.S.; Lee, M.; Jun, B.J.

    2003-01-01

    Full text: The safety culture in HANARO was self-assessed in accordance with the Code of Conduct on the Safety of Research Reactor drafted by IAEA. From 2002, IAEA has worked on the development of the Code of Conduct to achieve and maintain high level of nuclear safety in research reactors worldwide through the enhancement of national measures and international co-operation including, where appropriate, safety related technical cooperation. It defines the role of the state, the role of the regulatory body, the role of the operating organization and the role of the IAEA. As for the role of operating organization, the code specifies general requirements in assessment and verification of safety, financial and human resources, quality assurance, human factors, radiation protection and emergency preparedness. It also defines the role of operating organization for safety of research reactor in siting, design, operation, maintenance, modification and utilization as well. All of these items are the subjects for safety culture implementation, which means the Code could be a guideline for an operating organization to assess its safety culture. The self-assessment of safety culture in HANARO was made by using the sections of the Code describing the role of the operating organization for safety of research reactor. The major assessment items and the practices in HANARO for each items are as follow: The SAR of HANARO was reviewed by the regulatory body before the construction and the fuel loading of HANARO. Major design modifications and new installation of utilization facility needs the approval from regulatory body and safety assessment is a requirement for the approval. The Tech. Spec. for HANARO Operation specifies the analysis, surveillance, testing and inspection for HANARO operation. The reactor operation is mainly supported by the government and partly by nuclear R and D fund. The education and training of operation staff are one of major tasks of operating organization

  17. IAEA program for the preparation of safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    1975-01-01

    On the 13th of September, 1974, the IAEA Governors' Council has given its consent to the programme for the establishment of safety codes and guides (annex VII to IAEA document G.C. (XVIII/526)). The programme envisages the establishment of one code of practice for each of the issues governmental organization, siting, design, operation and quality assurance and also of about 50 safety guides between 1975 and 1980. These codes will contain the minimum requirements for the safety of the nuclear power stations, their systems and components. The guides will recommend methods to achieve the aims stated in the codes. It is the purpose of these IAEA activities to provide recommendations and guiding rules which may serve as standards for the assessment of the safety of nuclear power stations for all nations which may become participants in the peaceful use of nuclear energy within the next few years. (orig./AK) [de

  18. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  19. Safety code 19: recommended safety procedures for the selection, installation and use of x-ray diffraction equipment

    International Nuclear Information System (INIS)

    1984-01-01

    This document is one of a series of Safety Codes prepared by the Radiation Protection Bureau to set out requirements for the safe use of radiation emitting devices. The equipment and installation guidelines and safety procedures detailed in this Code are primarily for the instruction and guidance of persons employed in Federal Public Service Departments and Agencies, as well as those coming under the jurisdiction of the Canada Labour Code. This Safety Code is also intended to assist other users of X-ray diffraction equipment to select safe equipment and to install and use it so that the radiation hazard to the operator and other persons in its vicinity is negligible. It should be noted that facilities under provincial jurisdiction may be subject to requirements specified under provincial statutes. This Code supersedes Safety Code RPD-SC-7, entitled 'Requirements For Non-Medical X-Ray Equipment, Use and Installation', insofar as X-ray diffraction equipment is concerned, and it is intended to complement X-ray equipment design, construction and performance standards promulgated under the Radiation Emitting Devices Act

  20. The history of 10 years of electrical safety

    International Nuclear Information System (INIS)

    1984-12-01

    This book describes the foundation course of Korea electrical power corporation, including the process of the establishment with difficulty in the early period, growth through the rough passage, maintain of stability with voluntary service for public benefit. Next it deals with the management of the organization, and personnel management, financial affairs the management of business, examination for safety. The last part is an appendix for the administration law of each deportment.

  1. The safety relief valve handbook design and use of process safety valves to ASME and International codes and standards

    CERN Document Server

    Hellemans, Marc

    2009-01-01

    The Safety Valve Handbook is a professional reference for design, process, instrumentation, plant and maintenance engineers who work with fluid flow and transportation systems in the process industries, which covers the chemical, oil and gas, water, paper and pulp, food and bio products and energy sectors. It meets the need of engineers who have responsibilities for specifying, installing, inspecting or maintaining safety valves and flow control systems. It will also be an important reference for process safety and loss prevention engineers, environmental engineers, and plant and process designers who need to understand the operation of safety valves in a wider equipment or plant design context. . No other publication is dedicated to safety valves or to the extensive codes and standards that govern their installation and use. A single source means users save time in searching for specific information about safety valves. . The Safety Valve Handbook contains all of the vital technical and standards informat...

  2. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  3. Utilization of the MCNP-3A code for criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.; Moreira, J.M.L.

    1996-01-01

    In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

  4. Cross-Index to DOE-prescribed industrial safety codes and standards

    International Nuclear Information System (INIS)

    1980-01-01

    This Cross-Index volume is the 1980 compilation of detailed information from more than two hundred and ninety Department of Energy (DOE) prescribed or Occupational Health and Safety Administration (OSHA) referenced industrial safety codes and standards. The compilation of this material was conceived and initiated in 1973, and is revised yearly to provide information from current codes. Condensed data from individual code portions are listed according to reference code, section, paragraph, and page. Each code is given a two-digit reference code number or letter in the Contents section. This reference code provides ready identification of any code listed in the Cross-Index. The computerized information listings are on the left-hand portion of Cross-Index page; in order to the right of the listing are the reference code letters or numbers, the section, paragraph, and page of the referenced code containing expanded information on the individual listing. Simplified How to Use directions are listed. A glossary of letter initials/abbreviations for the organizations or documents, whose codes or standards are contained in this Cross-Index, is included

  5. Safety analyses of the electrical systems on VVER NPP

    International Nuclear Information System (INIS)

    Andel, J.

    2004-01-01

    Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

  6. The qualification of electrical components and instrumentations relevant to safety

    CERN Document Server

    Zambardi, F

    1989-01-01

    Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and ...

  7. Code on the safety of nuclear power plants: Quality assurance

    International Nuclear Information System (INIS)

    1988-01-01

    This revised Code provides the principles and objectives for the establishment and implementation of quality assurance programmes applied to both the overall and each of the constituent activities associated with a nuclear power plant project. The quality assurance principles enumerated in the present Code can be usefully applied to nuclear facilities other than nuclear power plants. The quality assurance programme encompasses: (1) the activities that are necessary to achieve the appropriate quality of the respective item or service; and (2) the activities that are necessary for verifying that the required quality is achieved and that objective evidence is produced to that effect. Quality assurance is an essential aspect of good management and the quality assurance programme is the main management tool for a disciplined approach to all activities affecting quality, including, where appropriate, verification that each task has been satisfactorily performed and that necessary corrective actions have been implemented. The principles and objectives provided by the Code are applicable by all those responsible for the nuclear power plant, by plant designers, suppliers, architect-engineers, plant constructors, plant operators and other organizations participating in activities affecting quality. The Code is a revision of the previous Code of Practice (1978) on the same subject of interest to regulatory bodies and experts in quality assurance for design, siting and operation of nuclear power plants. Contents: Definitions; 1. Introduction; 2. Quality assurance programmes; 3. Organization; 4. Document control; 5. Design control; 6. Procurement control; 7. Control of items; 8. Process control; 9. Inspection and test control; 10. Non-conformance control; 11. Corrective actions; 12, Records; 13. Audits

  8. R&D for Safety Codes and Standards: Materials and Components Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Somerday, Brian P. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); LaFleur, Chris [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Marchi, Chris San [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2015-08-01

    This project addresses the following technical barriers from the Safety, Codes and Standards section of the 2012 Fuel Cell Technologies Office Multi-Year Research, Development and Demonstration Plan (section 3.8): (A) Safety data and information: limited access and availability (F) Enabling national and international markets requires consistent RCS (G) Insufficient technical data to revise standards.

  9. Code on the safety of nuclear research reactors: Design

    International Nuclear Information System (INIS)

    1992-01-01

    The main objective of this publication is to provide a safety basis for the design of a research reactor and for the assessment of the design. Another objective is to cover certain aspects related to regulatory supervision, siting and quality assurance, as far as these are related to activities for the design of a research reactor. These objectives are expressed in terms of requirements and recommendations for the design of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop specific regulations and safety criteria for its research reactor programme.

  10. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  11. Code on the safety of nuclear research reactors: Operation

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this publication is to provide the essential requirements and recommendations for the safe operation of research reactors, with emphasis on the supervisory and managerial aspects. However, the publication also provides some guidance and information on topics concerning all the organizations involved in operation. These objectives are expressed in terms of requirements and recommendations for the safe operation of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on the ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop regulations and safety criteria for its research reactor programme.

  12. System code improvements for modelling passive safety systems and their validation

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

  13. Development of NPP Safety Requirements into Kenya's Grid Codes

    Energy Technology Data Exchange (ETDEWEB)

    Ndirangu, Nguni James; Koo, Chang Choong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand.

  14. Development of NPP Safety Requirements into Kenya's Grid Codes

    International Nuclear Information System (INIS)

    Ndirangu, Nguni James; Koo, Chang Choong

    2015-01-01

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand

  15. Demonstration study on shielding safety analysis code. 7

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    2000-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) To improve the detection sensitivity of pulse neutron measurement, two neutron detectors and some electronic circuits are added to the system constructed last year. (2) To estimate the neutron dose at the distant point from the facility instead of the commercialized rem-counter, a {sup 3}He detector with paraffin moderator is equipped to the system. (3) Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility was measured in the distance up to 300 m. The results show that the time structure of pulsed neutrons almost disappears at the further points than 150 m. (4) In the distance from 90 m to 300 m ordinal total counting method without gate pulse are applied to detect the neutrons. (5) The experimental results of space dependency up to 300 m is fitted fairly well by the Gui's response function. (author)

  16. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup

    2011-04-01

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4∼'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  17. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup [KAERI, Daejeon (Korea, Republic of)

    2011-04-15

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4{approx}'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  18. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Y. M.; Park, H. S.; Kim, T. H.

    2015-01-01

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software

  19. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. M.; Park, H. S. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

    2015-10-15

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software.

  20. MKENO-DAR: a direct angular representation Monte Carlo code for criticality safety analysis

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Komuro, Yuichi; Tsunoo, Yukiyasu; Nakayama, Mitsuo.

    1984-03-01

    Improving the Monte Carlo code MULTI-KENO, the MKENO-DAR (Direct Angular Representation) code has been developed for criticality safety analysis in detail. A function was added to MULTI-KENO for representing anisotropic scattering strictly. With this function, the scattering angle of neutron is determined not by the average scattering angle μ-bar of the Pl Legendre polynomial but by the random work operation using probability distribution function produced with the higher order Legendre polynomials. This code is avilable for the FACOM-M380 computer. This report is a computer code manual for MKENO-DAR. (author)

  1. LISA. A code for safety assessment in nuclear waste disposals program description and user guide

    International Nuclear Information System (INIS)

    Saltelli, A.; Bertozzi, G.; Stanners, D.A.

    1984-01-01

    The code LISA (Long term Isolation Safety Assessment), developed at the Joint Research Centre, Ispra is a useful tool in the analysis of the hazard due to the disposal of nuclear waste in geological formations. The risk linked to preestablished release scenarios is assessed by the code in terms of dose rate to a maximum exposed individual. The various submodels in the code simulate the system of barriers -both natural and man made- which are interposed between the contaminants and man. After a description of the code features a guide for the user is supplied and then a test case is presented

  2. Safety requirements for a nuclear power plant electric power system

    Energy Technology Data Exchange (ETDEWEB)

    Fouad, L F; Shinaishin, M A

    1988-06-15

    This work aims at identifying the safety requirements for the electric power system in a typical nuclear power plant, in view of the UNSRC and the IAEA. Description of a typical system is provided, followed by a presentation of the scope of the information required for safety evaluation of the system design and performance. The acceptance and design criteria that must be met as being specified by both regulatory systems, are compared. Means of implementation of such criteria as being described in the USNRC regulatory guides and branch technical positions on one hand and in the IAEA safety guides on the other hand are investigated. It is concluded that the IAEA regulations address the problems that may be faced with in countries having varying grid sizes ranging from large stable to small potentially unstable ones; and that they put emphasis on the onsite standby power supply. Also, in this respect the Americans identify the grid as the preferred power supply to the plant auxiliaries, while the IAEA leaves the possibility that the preferred power supply could be either the grid or the unit main generator depending on the reliability of each. Therefore, it is found that it is particularly necessary in this area of electric power supplies to deal with the IAEA and the American sets of regulations as if each complements and not supplements the other. (author)

  3. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  4. Computer codes in nuclear safety, radiation transport and dosimetry

    International Nuclear Information System (INIS)

    Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.

    2006-01-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations

  5. Audit of data and code use in the SR-Can safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Hicks, T.W.; Baldwin, T.D. [Galson Sciences Ltd, 5 Grosvenor House, Melton R oad, Oakham, Rutland LE15 6AX (United Kingdom)

    2008-03-15

    Building on the findings of previous studies on data and code quality assurance (QA) in safety assessments, this report provides a review of data and code QA in the SR-Can safety assessment. The data quality audit aimed to check that the selection and use of data in the SR-Can safety assessment was appropriate, focusing on the data that underpin representations of and assumptions about canister, insert, buffer, and backfill behaviour. The SR-Can Data Report provided the initial focus for examining the traceability and reliability of data used in the safety assessment; the Data Report is one of the series of SR-Can safety assessment reports and, in this review, it was anticipated that it would provide the primary source of data on the canister, insert, buffer, and backfill. However, other safety assessment reports (the SR-Can Main Report, the Initial State Report, the Fuel and Canister Process Report, and the Buffer and Backfill Process Report) were found to provide key information on data used in the safety assessment. The quality audit of codes aimed to check that code use in the SR-Can safety assessment has been justified through a transparent and traceable process of code development and selection. The Model Summary Report provided the focus for reviewing the QA status of the codes used in the safety assessment. As well as highlighting a number of concerns regarding QA aspects of specific data sets, parameter values, and codes used in the SR-Can safety assessment (which are presented in the report), the review has led to several general observations on data and code QA that should be considered by SKB in the development and implementation of a QA system for the SR-Site safety assessment: - The SR-Site safety assessment and associated QA records should include information that demonstrates that a full QA system has been implemented in order to build confidence in the validity of the assessment. - The data and parameter values used directly in the safety

  6. Risk limitation, safety and environmental compatibility in electricity generation

    International Nuclear Information System (INIS)

    Angelini, A.M.

    1981-01-01

    The purpose of this paper is to present the problem of meeting future electricity needs while at the same time reducing to a minimum the risks, the pollution of air and water and the environmental effects of power stations. The first resource to exploit is the ''virtual source'' represented by energy saving pursued to the limit of the possible. The second, in order of priority, is that of renewable resources as yet unused and under development. Unfortunately, in most countries these latter resources are far from sufficient: it is then necessary to choose between the use of conventional fossil fuels and nuclear fuels. In this paper it is shown that, of all the possible fossil fuels, only coal can be considered for electricity production. As a result, in meeting new electricity needs, the choice will have to be made between coal and nuclear power. Attention is directed to factors having a significant influence on this choice, particularly the risks and safety problems in the widest sense, with a view to making a global evaluation comprising not just generating stations but the entire production cycle, from the search for the primary source to the supplying of electricity to the user. The most important problems that arise in this connection are briefly analysed in the paper, which concludes with an appeal for more objectivity in providing information on energy, such information being at present very ''polluted'' and exerting a major influence on the views of experts. (author)

  7. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

  8. Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

    International Nuclear Information System (INIS)

    Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2013-01-01

    SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

  9. Electric capacitance tomography and two-phase flow for the nuclear reactor safety analysis

    International Nuclear Information System (INIS)

    Lee, Jae Young

    2008-01-01

    Recently electric capacitance tomography has been developed to be used in the analysis of two-phase flow. Although its electric field is not focused as the hard ray tomography such as the X-ray or gamma ray, its convenience of easy access to the system and easy maintenance due to no requirement of radiation shielding benefits us in its application in the two-phase flow study, one of important area in the nuclear safety analysis. In the present paper, the practical technologies in the electric capacitance tomography are represented in both parts of hardware and software. In the software part, both forward problem and inverse problem are discussed and the method of regularization. In the hardware part, the brief discussion of the electronics circuits is made which provides femto farad resolution with a reasonable speed (150 frame/sec for 16 electrodes). Some representative ideal cases are studied to demonstrate its potential capability for the two-phase flow analysis. Also, some variations of the tomography such as axial tomography, and three dimensional tomography are discussed. It was found that the present ECT is expected to become a useful tool to understand the complicated three dimensional two-phase flow which may be an important feature to be equipped by the safety analysis codes. (author)

  10. Development and application of best-estimate LWR safety analysis codes

    International Nuclear Information System (INIS)

    Reocreux, M.

    1997-01-01

    This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

  11. International assessment of application of the Code of Conduct on the Safety of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor

    2015-11-15

    The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; notably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

  12. International assessment of application of the Code of Conduct on the Safety of Research Reactors

    International Nuclear Information System (INIS)

    Shokr, A.M.

    2015-01-01

    The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; notably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

  13. Code of Sustainable Practice in Occupational and Environmental Health and Safety for Corporations.

    Science.gov (United States)

    Castleman, Barry; Allen, Barbara; Barca, Stefania; Bohme, Susanna Rankin; Henry, Emmanuel; Kaur, Amarjit; Massard-Guilbaud, Genvieve; Melling, Joseph; Menendez-Navarro, Alfredo; Renfrew, Daniel; Santiago, Myrna; Sellers, Christopher; Tweedale, Geoffrey; Zalik, Anna; Zavestoski, Stephen

    2008-01-01

    At a conference held at Stony Brook University in December 2007, "Dangerous Trade: Histories of Industrial Hazard across a Globalizing World," participants endorsed a Code of Sustainable Practice in Occupational and Environmental Health and Safety for Corporations. The Code outlines practices that would ensure corporations enact the highest health and environmentally protective measures in all the locations in which they operate. Corporations should observe international guidelines on occupational exposure to air contaminants, plant safety, air and water pollutant releases, hazardous waste disposal practices, remediation of polluted sites, public disclosure of toxic releases, product hazard labeling, sale of products for specific uses, storage and transport of toxic intermediates and products, corporate safety and health auditing, and corporate environmental auditing. Protective measures in all locations should be consonant with the most protective measures applied anywhere in the world, and should apply to the corporations' subsidiaries, contractors, suppliers, distributors, and licensees of technology. Key words: corporations, sustainability, environmental protection, occupational health, code of practice.

  14. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  15. Basic recognition on safety of nuclear electric power generation

    International Nuclear Information System (INIS)

    Miyazaki, Keiji

    1995-01-01

    The safety of nuclear electric power generation is not to inflict radiation damage on public. Natural radiation is about 1 mSv every year. As far as the core melting on large scale does not occur, there is not the possibility of exerting serious radiation effect to public. The way of thinking on ensuring the safety is defense in depth. The first protection is the prevention of abnormality, the second protection is the prevention of accidents, and the third protection is the relaxation of effect. As design base accidents, the loss of coolant accident due to the breakdown of inlet pipings of reactors and the breaking of fine tubes in steam generators are included. The suitability of location is evaluated. As the large scale accidents of nuclear power stations in the past, Chernobyl accident and Three Mile Island accident are explained. The features of the countermeasures to the accident in Mihama No. 2 plant are described. The countermeasures to severe accidents, namely accident management and general preventive maintenance are explained. The background of the nonconfidence feeling to nuclear electric power generation and the importance of opening information to public are shown. (K.I.)

  16. Internal Arc: People safety in the electrical wiring

    International Nuclear Information System (INIS)

    Inchausti, J. M.

    2009-01-01

    The aim of this article is to describe the internal arc phenomenon, an extremely fast, almost explosive and unattended process of transformation form an initial electric power to the generation of a pressure and heat wave inside the medium its produced its consequences for safety, current methods of limiting them and current regulations in general for equipment used in medium-voltage electrical distribution networks. Taking into account that this type of equipment is found thought the distribution network in both public buildings and unrestricted access areas, safety (of operators and the general public) must be taken into account in the design of equipment and installations to minimize the risk of internal arcs occurring. This is the gist of, for example, ITC 16 of the Spanish Regulation on Power Plants and transformer substations. In addition to the construction aspects specific to each device in which the manufacturer has to takes steps to minimize the risks of an internal arcs occurring. This is the gist of, for example, ITC 16 of the Spanish Regulation on Power Plants and transformer substations. In addition to the construction aspects specific to each device in which an internal arc occurring, it is understood to be vitally important that users, installers and designers of Medium Voltage installations are familiar with the installation conditions stated by the manufacturer and thus avoid risks. (Author) 14 refs

  17. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  18. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  19. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  20. Code of conduct on the safety and security of radioactive sources

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    The objective of this Code is to achieve and maintain a high level of safety and security of radioactive sources through the development, harmonization and enforcement of national policies, laws and regulations, and through tile fostering of international co-operation. In particular, this Code addresses the establishment of an adequate system of regulatory control from the production of radioactive sources to their final disposal, and a system for the restoration of such control if it has been lost.

  1. Code of conduct on the safety and security of radioactive sources

    International Nuclear Information System (INIS)

    2001-03-01

    The objective of this Code is to achieve and maintain a high level of safety and security of radioactive sources through the development, harmonization and enforcement of national policies, laws and regulations, and through tile fostering of international co-operation. In particular, this Code addresses the establishment of an adequate system of regulatory control from the production of radioactive sources to their final disposal, and a system for the restoration of such control if it has been lost

  2. Preliminary study for unified management of CANDU safety codes and construction of database system

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, Hyoung Tae

    2003-03-01

    It is needed to develop the Graphical User Interface(GUI) for the unified management of CANDU safety codes and to construct database system for the validation of safety codes, for which the preliminary study is done in the first stage of the present work. The input and output structures and data flow of CATHENA and PRESCON2 are investigated and the interaction of the variables between CATHENA and PRESCON2 are identified. Furthermore, PC versions of CATHENA and PRESCON2 codes are developed for the interaction of these codes and GUI(Graphic User Interface). The PC versions are assessed by comparing the calculation results with those by HP workstation or from FSAR(Final Safety Analysis Report). Preliminary study on the GUI for the safety codes in the unified management system are done. The sample of GUI programming is demonstrated preliminarily. Visual C++ is selected as the programming language for the development of GUI system. The data for Wolsong plants, reactor core, and thermal-hydraulic experiments executed in the inside and outside of the country, are collected and classified following the structure of the database system, of which two types are considered for the final web-based database system. The preliminary GUI programming for database system is demonstrated, which is updated in the future work

  3. Safety standards, legislation and codes of practice for fuel cell manufacture and operation

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, C.P.

    1999-07-01

    This report examines safety standards, legislation and codes of practice for fuel cell manufacture and operation in the UK, Europe and internationally. Management of health and safety in the UK is discussed, and the characteristics of phosphoric acid (PAFC), proton exchange membrane (PEM), molten carbonate (MCFC), solid oxide (SOFC) fuel cells are described. Fuel cell power plant standards and manufacture in the UK, design and operational considerations, end of life disposal, automotive fuel cell system, and fuelling and vehicular concerns are explored, and standards, legislation and codes of practice are explained in the appendix.

  4. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1995-01-01

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  5. Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

    1994-01-01

    Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

  6. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  7. Britain's Nuclear Electric pursues a major programme to enhance safety

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    The UK state-owned nuclear utility, Nuclear Electric, is actively pursuing a major initiative inherited from its predecessor the Central Electricity Generating Board, known as the Station Enhancement Programme. The Enhancement Programme is a multi-million pound project and covers all Nuclear Electric's stations (including relevant aspects of the de-commissioning site at Berkeley). Its fundamental objective is to enhance the arrangements, practices and attitudes on the power stations to create an overall environment likely to promote ever safer and more reliable operation, and to take full account of the best international practices. The four year programme started in the spring of 1989, and is expected to be completed in 1993. Key elements of the programme include quality assurance based management systems; improvements to site and plant material conditions; a more formal approach to training; additional resources and controls to aid the management of maintenance defects; higher profile for the feedback of operational experience; management targets; a station evaluation programme; and formal feedback of lessons learned from the IAEA operational and safety review team visit to the Oldbury-on-Severn station in 1989. (author)

  8. The qualification of electrical components and instrumentations relevant to safety

    International Nuclear Information System (INIS)

    Zambardi, F.

    1989-03-01

    Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and peculiar aspects are discussed and the qualification by testing is taken into special account. Technical and organizational aspects related to a plant qualification program are also focused. The report ends with a look to the most significant research and development activities. (author)

  9. Driver perceptions of the safety implications of quiet electric vehicles.

    Science.gov (United States)

    Cocron, Peter; Krems, Josef F

    2013-09-01

    Previous research on the safety implications of quiet electric vehicles (EVs) has mostly focused on pedestrians' acoustic perception of EVs, and suggests that EVs are more difficult for pedestrians to hear and, therefore, compromise traffic safety. The two German field studies presented here examine the experiences of 70 drivers with low noise emissions of EVs and the drivers' long-term evaluation of the issue. Participants were surveyed via interviews and questionnaires before driving an EV for the first time, after 3 months of driving, and in the first study, again after 6 months. Based on participants' reports, a catalogue of safety-relevant incidents was composed in Study 1. The catalogue revealed that low noise-related critical incidents only rarely occur, and mostly take place in low-speed environments. The degree of hazard related to these incidents was rated as low to medium. In Study 1, driver concern for vulnerable road users as a result of low noise diminished with increasing driving experience, while perceived comfort due to this feature increased. These results were replicated in Study 2. In the second study, it was additionally examined, if drivers adjust their perceived risk of harming other road users over time. Results show that the affective assessment of risk also decreased with increased driving experience. Based on individual experience, drivers adjust their evaluation of noise-related hazards, suggesting that dangers associated with low noise emissions might be less significant than previously expected. Copyright © 2013 Elsevier Ltd. All rights reserved.

  10. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  11. Code of practice on quality assurance for safety in nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    The code sets forth the management principles and objectives to be met during the implementation of activities in different phases of the nuclear power plants (NPPs) for assuring safety. It is intended for use by organisations and individuals responsible for safety related functions in design, manufacturing, construction, commissioning, operation and decommissioning of NPPs. It covers the functions of management, performance, verification and corrective action. It also deals with the quality assurance records. (M.G.B.)

  12. Report on nuclear industry quality assurance procedures for safety analysis computer code development and use

    International Nuclear Information System (INIS)

    Sheron, B.W.; Rosztoczy, Z.R.

    1980-08-01

    As a result of a request from Commissioner V. Gilinsky to investigate in detail the causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March, 1978, the staff undertook an extensive investigation of the vendor quality assurance practices applied to safety analysis computer code development and use. This investigation included inspections of code development and use practices of the four major Light Water Reactor Nuclear Steam Supply System vendors and a major reload fuel supplier. The conclusion reached by the staff as a result of the investigation is that vendor practices for code development and use are basically sound. A number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality assurance (QA) review and inspection process for computer codes and identified areas for improvement

  13. Recent development and application of a new safety analysis code for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: Brad.Merrill@inl.gov; Humrickhouse, Paul W.; Shimada, Masashi

    2016-11-01

    Highlights: • This paper presents recent code development activities for the MELCOR for fusion and Tritium Migration Analysis Program computer codes at the Idaho National Engineering Laboratory. • The capabilities of these computer codes are being merged into a single safety analysis tool for fusion reactor accidents. • The result of benchmarking these codes against previous code versions is presented by the authors of this paper. • This new capability is applied to study the tritium inventory and permeation rate for a water cold tungsten divertor that has neutron damage at 0.3 dpa. - Abstract: This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this

  14. Recent development and application of a new safety analysis code for fusion reactors

    International Nuclear Information System (INIS)

    Merrill, Brad J.; Humrickhouse, Paul W.; Shimada, Masashi

    2016-01-01

    Highlights: • This paper presents recent code development activities for the MELCOR for fusion and Tritium Migration Analysis Program computer codes at the Idaho National Engineering Laboratory. • The capabilities of these computer codes are being merged into a single safety analysis tool for fusion reactor accidents. • The result of benchmarking these codes against previous code versions is presented by the authors of this paper. • This new capability is applied to study the tritium inventory and permeation rate for a water cold tungsten divertor that has neutron damage at 0.3 dpa. - Abstract: This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this

  15. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  16. IAEA activities to prepare safety codes and guides for thermal neutron nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1977-01-01

    In accordance with the programme presented to, and endorsed by, the eighteenth General Conference in September 1974, the IAEA is now developing a complete set of safety codes and guides that will represent recommendations for the safety of thermal neutron power plants. The safety codes outline the minimum requirements for achieving this safety, and the safety guides set forth the criteria, procedures and methods to implement the safety codes. The whole programme is directed towards the five areas of Governmental Organization, Siting, Design, Operation, and Quality Assurance. One Scientific Secretary from the Agency Secretariat is responsible for each of these areas and a Co-ordinator takes care of common problems. For the development of each of these documents a working group of a few world experts is first convened which prepare a preliminary draft. This draft is then reviewed by a larger, international Technical Review Committee (one for each of the five areas) and a subsequent review by the Senior Advisory Group - with representatives from 20 states - ensures that the document is well coordinated within the programme. At this stage, it is sent to Member States for comments. The Technical Review Committee concerned is reconvened to integrate these comments into the document, and, after a final review by the Senior Advisory Group, the document is ready for transmission to the Director General of the Agency for endorsement and publication. A preliminary to this procedure is the collation by the Secretariat of large amounts of information submitted by Member States so that the first draft is really based on a very complete knowledge of what is done in each area all over the world. This collation frequently reveals differences in approach which are not random but due, rather, to the local conditions and the types of reactors. These differences must be harmonized in the documents produced without detracting from the effectiveness of the code or guide. The whole

  17. SACRD: a data base for fast reactor safety computer codes, general description

    International Nuclear Information System (INIS)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Simpson, D.B.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics, and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases. A general description of the SACRD system is presented in the report

  18. On application of CFD codes to problems of nuclear reactor safety

    International Nuclear Information System (INIS)

    Muehlbauer, Petr

    2005-01-01

    The 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in May 2002 at Aix-en-Province, France, recommended formation of writing groups to report the need of guidelines for use and assessment of CFD in single-phase nuclear reactor safety problems, and on recommended extensions to CFD codes to meet the needs of two-phase problems in nuclear reactor safety. This recommendations was supported also by Working Group on the Analysis and Management of Accidents and led to formation oaf three Writing Groups. The first writing Group prepared a summary of existing best practice guidelines for single phase CFD analysis and made a recommendation on the need for nuclear reactor safety specific guidelines. The second Writing Group selected those nuclear reactor safety applications for which understanding requires or is significantly enhanced by single-phase CFD analysis, and proposed a methodology for establishing assesment matrices relevant to nuclear reactor safety applications. The third writing group performed a classification of nuclear reactor safety problems where extension of CFD to two-phase flow may bring real benefit, a classification of different modeling approaches, and specification and analysis of needs in terms of physical and numerical assessments. This presentation provides a review of these activities with the most important conclusions and recommendations (Authors)

  19. Nuclear electric propulsion operational reliability and crew safety study

    International Nuclear Information System (INIS)

    Karns, J.J.; Fragola, J.R.; Kahan, L.; Pelaccio, D.

    1993-01-01

    The central purpose of this analysis is to assess the ''achievability'' of a nuclear electric propulsion (NEP) system in a given mission. ''Achievability'' is a concept introduced to indicate the extent to which a system that meets or achieves its design goals might be implemented using the existing technology base. In the context of this analysis, the objective is to assess the achievability of an NEP system for a manned Mars mission as it pertains to operational reliability and crew safety goals. By varying design parameters, then examining the resulting system achievability, the design and mission risk drivers can be identified. Additionally, conceptual changes in design approach or mission strategy which are likely to improve overall achievability of the NEP system can be examined

  20. Status of electrical safety in Indira Gandhi Centre for Atomic Research

    International Nuclear Information System (INIS)

    Kandasamy, S.; Karthikeyan, S.V.; Senthilkumar, B.; Shunmugam, U.; Kannan, S.E.

    1999-01-01

    Electrical Systems in Indira Gandhi Centre for Atomic Research (IGCAR), comprising of facilities such as Fast Breeder Test Reactor (FBTR) and various Research and Development laboratories form a complex network providing practically uninterrupted power supply to all the facilities meeting their energy demands. The safety record of the Centre for the past 25 years has been very good and the number of electrical accidents in the Centre has been less. This paper brings out the status of the electrical safety in IGCAR and indicates the steps to be taken to improve the safety culture. It also utilizes the measure to be provided for improving the electrical safety in the works executed by contractors

  1. Application of Code Of Conduct on the Safety of Research Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Ahmad Nabil Abd Rahim; Zarina Masood

    2014-01-01

    The implementation and the practices of the effective safety system at research reactors are important to ensure that the worker, public and environment do not receive any abnormal causes. Many international safety related support agencies for research reactor such as International Atomic Energy Agency (IAEA) providing guidelines that can be applied to enhance and strengthen the enforcement of safety namely Code of Conduct on the Safety of Research Reactor (IAEA/CODEOC/RR/2006). The excellent safety management, reliability, and maintainability of RTP reactor structures, coupled with personnel numerous lessons and experiences learned, Reactor TRIGA PUSPATI research reactor providing Nuclear Malaysia personnel and visitor the very safe working and visiting environment. This paper will discuss the status, practices and improvement strategies over the past few years. (author)

  2. Implementation of reactor safety analysis code CATHARE and its use on FACOM M-380

    International Nuclear Information System (INIS)

    Ishiguro, Misako; Shinozawa, Naohisa; Tomiyama, Mineyoshi; Fujisaki, Masahide

    1986-05-01

    CATHARE is an advanced safety analysis code developed at the Nuclear Research Center of Grenoble in France. The code simulates thermohydraulic phenomena involved in loss of coolant accidents in pressurized water reactors. The code has been introduced into JAERI as a part of the technical exchange between the JAERI ROSA-IV Program and the French BETHSY-CATHARE Program. The code was delivered in the form of 23 files containing 115,000 statements in total. A large part of CATHARE code has been written in an extended Fortran language 'Esope' which is mainly used for managing dynamic memory allocation. The JAERI version is created from the IBM version which has been used on Amdhal computer at ISPRA. Some modifications are required in order to implement the CATHARE code at JAERI because of difference in softwares. In this report, the overview of the code structure, the JAERI usage, the implementation method, the error correction method, the problems special to install the code in JAERI, and the distribution of computing time are described. (author)

  3. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    International Nuclear Information System (INIS)

    Touati, Said; Chennai, Salim; Souli, Aissa

    2015-01-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  4. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    Energy Technology Data Exchange (ETDEWEB)

    Touati, Said; Chennai, Salim; Souli, Aissa [Nuclear Research Centre of Birine, Ain Oussera, Djelfa Province (Algeria)

    2015-07-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  5. Applying Hamming Code to Memory System of Safety Grade PLC (POSAFE-Q) Processor Module

    International Nuclear Information System (INIS)

    Kim, Taehee; Hwang, Sungjae; Park, Gangmin

    2013-01-01

    If some errors such as inverted bits occur in the memory, instructions and data will be corrupted. As a result, the PLC may execute the wrong instructions or refer to the wrong data. Hamming Code can be considered as the solution for mitigating this mis operation. In this paper, we apply hamming Code, then, we inspect whether hamming code is suitable for to the memory system of the processor module. In this paper, we applied hamming code to existing safety grade PLC (POSAFE-Q). Inspection data are collected and they will be referred for improving the PLC in terms of the soundness. In our future work, we will try to improve time delay caused by hamming calculation. It will include CPLD optimization and memory architecture or parts alteration. In addition to these hamming code-based works, we will explore any methodologies such as mirroring for the soundness of safety grade PLC. Hamming code-based works can correct bit errors, but they have limitation in multi bits errors

  6. Development and Verification of the Computer Codes for the Fast Reactors Nuclear Safety Justification

    International Nuclear Information System (INIS)

    Kisselev, A.E.; Mosunova, N.A.; Strizhov, V.F.

    2015-01-01

    The information on the status of the work on development of the system of the nuclear safety codes for fast liquid metal reactors is presented in paper. The purpose of the work is to create an instrument for NPP neutronic, thermohydraulic and strength justification including human and environment radiation safety. The main task that is to be solved by the system of codes developed is the analysis of the broad spectrum of phenomena taking place on the NPP (including reactor itself, NPP components, containment rooms, industrial site and surrounding area) and analysis of the impact of the regular and accidental releases on the environment. The code system is oriented on the ability of fully integrated modeling of the NPP behavior in the coupled definition accounting for the wide range of significant phenomena taking place on the NPP under normal and accident conditions. It is based on the models that meet the state-of-the-art knowledge level. The codes incorporate advanced numerical methods and modern programming technologies oriented on the high-performance computing systems. The information on the status of the work on verification of the separate codes of the system of codes is also presented. (author)

  7. Quantitative software-reliability analysis of computer codes relevant to nuclear safety

    International Nuclear Information System (INIS)

    Mueller, C.J.

    1981-12-01

    This report presents the results of the first year of an ongoing research program to determine the probability of failure characteristics of computer codes relevant to nuclear safety. An introduction to both qualitative and quantitative aspects of nuclear software is given. A mathematical framework is presented which will enable the a priori prediction of the probability of failure characteristics of a code given the proper specification of its properties. The framework consists of four parts: (1) a classification system for software errors and code failures; (2) probabilistic modeling for selected reliability characteristics; (3) multivariate regression analyses to establish predictive relationships among reliability characteristics and generic code property and development parameters; and (4) the associated information base. Preliminary data of the type needed to support the modeling and the predictions of this program are described. Illustrations of the use of the modeling are given but the results so obtained, as well as all results of code failure probabilities presented herein, are based on data which at this point are preliminary, incomplete, and possibly non-representative of codes relevant to nuclear safety

  8. Applying Hamming Code to Memory System of Safety Grade PLC (POSAFE-Q) Processor Module

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taehee; Hwang, Sungjae; Park, Gangmin [POSCO Nuclear Technology, Seoul (Korea, Republic of)

    2013-05-15

    If some errors such as inverted bits occur in the memory, instructions and data will be corrupted. As a result, the PLC may execute the wrong instructions or refer to the wrong data. Hamming Code can be considered as the solution for mitigating this mis operation. In this paper, we apply hamming Code, then, we inspect whether hamming code is suitable for to the memory system of the processor module. In this paper, we applied hamming code to existing safety grade PLC (POSAFE-Q). Inspection data are collected and they will be referred for improving the PLC in terms of the soundness. In our future work, we will try to improve time delay caused by hamming calculation. It will include CPLD optimization and memory architecture or parts alteration. In addition to these hamming code-based works, we will explore any methodologies such as mirroring for the soundness of safety grade PLC. Hamming code-based works can correct bit errors, but they have limitation in multi bits errors.

  9. Demonstration of Emulator-Based Bayesian Calibration of Safety Analysis Codes: Theory and Formulation

    Directory of Open Access Journals (Sweden)

    Joseph P. Yurko

    2015-01-01

    Full Text Available System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC sampling feasible. This work uses Gaussian Process (GP based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.

  10. Quality assurance for safety in nuclear power plants. A code of practice

    International Nuclear Information System (INIS)

    1978-01-01

    The Code of Practice is a part of the International Atomic Energy Agency's programme, referred to as the NUSS programme (Nuclear Safety Standards), for establishing Codes of Practice and Safety Guides relating to land-based stationary thermal neutron power plants. The documents are based on documentation and experience from various national systems and practices. The present document provides the recommended principles and objectives for the establishment and implementation of a quality assurance programme during design, manufacture, construction, commissioning and operation of structures, system and components important to safety. They are applicable by all those responsible for the power plant, by plant designers, suppliers, architect-engineers, plant constructors, plant operators and other organizations participating in activities affecting quality. The Lists of relevant definition and the Provisional List of NUSS Programme Titles are given

  11. Development of the Monju core safety analysis numerical models by super-COPD code

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Minami, Masaki

    2010-12-01

    Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model. The numerical model for core safety analysis has been built based on the best estimate model validated by the actually measured plant behavior up to 40% rated power conditions, taking over safety analysis models of conventionally applied COPD and HARHO-IN codes, to be capable of overall calculations of the entire plant with the safety protection and control systems. Among the constituents of the analytical model, neutronic-thermal model, heat transfer and hydraulic models of PHTS, SHTS, and water/steam system are individually verified by comparisons with the conventional calculations. Comparisons are also made with the actually measured plant behavior up to 40% rated power conditions to confirm the calculation adequacy and conservativeness of the input data. The unified analytical model was applied to analyses of in total 8 anomaly events; reactivity insertion, abnormal power distribution, decrease and increase of coolant flow rate in PHTS, SHTS and water/steam systems. The resulting maximum values and temporal variations of the key parameters in safety evaluation; temperatures of fuel, cladding, in core sodium coolant and RV inlet and outlet coolant have negligible discrepancies against the existing analysis result in the annex to the construction permit application, verifying the unified analytical model. These works have enabled analytical evaluation of Monju core heat removal capability by Super-COPD utilizing the

  12. Intercomparison and validation of computer codes for thermalhydraulic safety analysis of heavy water reactors

    International Nuclear Information System (INIS)

    2004-08-01

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and co-operative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled: Intercomparison and validation of computer codes for thermalhydraulics safety analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. The RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommendations made

  13. From chain liability to chain responsibility: MNE approaches to implement safety and health codes in international supply chains

    NARCIS (Netherlands)

    van Tulder, R.; van Wijk, J.; Kolk, A.

    2009-01-01

    This article examines whether the involvement of stakeholders in the design of corporate codes of conduct leads to a higher implementation likelihood of the code. The empirical focus is on Occupational Safety and Health (OSH). The article compares the inclusion of OSH issues in the codes of conduct

  14. Code of conduct on the safety and security of radioactive sources

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    The objective of the code of conduct is to achieve and maintain a high level of safety and security of radioactive sources through the development, harmonization and enforcement of national policies, laws and regulations, and through the fostering of international co-operation. In particular, this code addresses the establishment of an adequate system of regulatory control from the production of radioactive sources to their final disposal, and a system for the restoration of such control if it has been lost. (N.C.)

  15. Radiation protection in dentistry. Recommended safety procedures for the use of dental x-ray equipment. Safety code 30

    International Nuclear Information System (INIS)

    1994-01-01

    The Radiation Protection Bureau has prepared a series of documents on safety codes to set out requirements for the safe use of radiation-emitting equipment. This Safety Code has been prepared to provide specific guidance to the dentist, dental hygienist, dental assistant and other support personnel concerned with safety procedures and equipment performance. Dental radiography is one of the most valuable tools used in modern dental health care. It makes possible the diagnosis of physical conditions that would otherwise be difficult to identify. The use of dental radiological procedures must be carefully managed, because x-radiation has the potential for damaging healthy cells and tissues. Although no known occurrence of cancer or genetic damage has been observed from radiation doses delivered in modern dentistry, and until more evidence is available, one should practice radiation hygiene with the same care as would be dictated if a hazard were known to exist. The aim of radiation protection in dentistry is to obtain the desired clinical information with minimal radiation exposure to patients, dental personnel and the public. 15 tabs

  16. Radiation protection in dentistry. Recommended safety procedures for the use of dental x-ray equipment. Safety code 30

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The Radiation Protection Bureau has prepared a series of documents on safety codes to set out requirements for the safe use of radiation-emitting equipment. This Safety Code has been prepared to provide specific guidance to the dentist, dental hygienist, dental assistant and other support personnel concerned with safety procedures and equipment performance. Dental radiography is one of the most valuable tools used in modern dental health care. It makes possible the diagnosis of physical conditions that would otherwise be difficult to identify. The use of dental radiological procedures must be carefully managed, because x-radiation has the potential for damaging healthy cells and tissues. Although no known occurrence of cancer or genetic damage has been observed from radiation doses delivered in modern dentistry, and until more evidence is available, one should practice radiation hygiene with the same care as would be dictated if a hazard were known to exist. The aim of radiation protection in dentistry is to obtain the desired clinical information with minimal radiation exposure to patients, dental personnel and the public. 15 tabs.

  17. Temporal code-driven stimulation: definition and application to electric fish signaling

    Directory of Open Access Journals (Sweden)

    Angel Lareo

    2016-10-01

    Full Text Available Closed-loop activity-dependent stimulation is a powerful methodology to assess information processing in biological systems. In this context, the development of novel protocols, their implementation in bioinformatics toolboxes and their application to different description levels open up a wide range of possibilities in the study of biological systems. We developed a methodology for studying biological signals representing them as temporal sequences of binary events. A specific sequence of these events (code is chosen to deliver a predefined stimulation in a closed-loop manner. The response to this code-driven stimulation can be used to characterize the system. This methodology was implemented in a real time toolbox and tested in the context of electric fish signaling. We show that while there are codes that evoke a response that cannot be distinguished from a control recording without stimulation, other codes evoke a characteristic distinct response. We also compare the code-driven response to open-loop stimulation. The discussed experiments validate the proposed methodology and the software toolbox.

  18. ELECTROMAGNETIC SAFETY OF ELECTRIC TRANSPORT SYSTEMS: MAIN SOURCES AND PARAMETERS OF MAGNETIC FIELDS

    Directory of Open Access Journals (Sweden)

    N. G. Ptitsyna

    2013-03-01

    Full Text Available Magnetic fields produced by electric drive vehicles may break electromagnetic safety. For electromagnetic safety and electromagnetic compatibility knowledge about characteristics and sources of magnetic fields in the electric transport is necessary. The article deals with analysis of available data about magnetic fields in electric cars and comparison with results of our measurements carried out in the other types of electrified transport systems.

  19. IAEA programme to support development and validation of advanced design and safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J., E-mail: J.H.Choi@iaea.org [International Atomic Energy Agency, Vienna (Austria)

    2013-07-01

    The International Atomic Energy Agency (IAEA) has been organized many international collaboration programs to support the development and validation of design and safety analysis computer codes for nuclear power plants. These programs are normally implemented with a frame of Coordinated Research Project (CRP) or International Collaborative Standard Problem (ICSP). This paper introduces CRPs and ICSPs currently being organized or recently completed by IAEA for this purpose. (author)

  20. Application of software engineering to development of reactor-safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1980-11-01

    As a result of the drastically increasing cost of software and the lack of an engineering approach, the technology of Software Engineering is being developed. Software Engineering provides an answer to the increasing cost of developing and maintaining software. It has been applied extensively in the business and aerospace communities and is just now being applied to the development of scientific software and, in particular, to the development of reactor safety codes at HEDL

  1. TOOKUIL: A case study in user interface development for safety code application

    International Nuclear Information System (INIS)

    Gray, D.L.; Harkins, C.K.; Hoole, J.G.

    1997-01-01

    Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today's safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL

  2. TOOKUIL: A case study in user interface development for safety code application

    International Nuclear Information System (INIS)

    Gray, D.L.; Harkins, C.K.; Hoole, J.G.; Peebles, R.C.; Smith, R.J.

    1996-11-01

    Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today's safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL

  3. TOOKUIL: A case study in user interface development for safety code application

    Energy Technology Data Exchange (ETDEWEB)

    Gray, D.L.; Harkins, C.K.; Hoole, J.G. [and others

    1997-07-01

    Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today`s safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL.

  4. Code of conduct on the safety and security of radioactive sources

    International Nuclear Information System (INIS)

    2004-01-01

    The objectives of the Code of Conduct are, through the development, harmonization and implementation of national policies, laws and regulations, and through the fostering of international co-operation, to: (i) achieve and maintain a high level of safety and security of radioactive sources; (ii) prevent unauthorized access or damage to, and loss, theft or unauthorized transfer of, radioactive sources, so as to reduce the likelihood of accidental harmful exposure to such sources or the malicious use of such sources to cause harm to individuals, society or the environment; and (iii) mitigate or minimize the radiological consequences of any accident or malicious act involving a radioactive source. These objectives should be achieved through the establishment of an adequate system of regulatory control of radioactive sources, applicable from the stage of initial production to their final disposal, and a system for the restoration of such control if it has been lost. This Code relies on existing international standards relating to nuclear, radiation, radioactive waste and transport safety and to the control of radioactive sources. It is intended to complement existing international standards in these areas. The Code of Conduct serves as guidance in general issues, legislation and regulations, regulatory bodies as well as import and export of radioactive sources. A list of radioactive sources covered by the code is provided which includes activities corresponding to thresholds of categories

  5. Code of conduct on the safety and security of radioactive sources

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-01-01

    The objectives of the Code of Conduct are, through the development, harmonization and implementation of national policies, laws and regulations, and through the fostering of international co-operation, to: (i) achieve and maintain a high level of safety and security of radioactive sources; (ii) prevent unauthorized access or damage to, and loss, theft or unauthorized transfer of, radioactive sources, so as to reduce the likelihood of accidental harmful exposure to such sources or the malicious use of such sources to cause harm to individuals, society or the environment; and (iii) mitigate or minimize the radiological consequences of any accident or malicious act involving a radioactive source. These objectives should be achieved through the establishment of an adequate system of regulatory control of radioactive sources, applicable from the stage of initial production to their final disposal, and a system for the restoration of such control if it has been lost. This Code relies on existing international standards relating to nuclear, radiation, radioactive waste and transport safety and to the control of radioactive sources. It is intended to complement existing international standards in these areas. The Code of Conduct serves as guidance in general issues, legislation and regulations, regulatory bodies as well as import and export of radioactive sources. A list of radioactive sources covered by the code is provided which includes activities corresponding to thresholds of categories.

  6. Interactive computer codes for education and training on nuclear safety and radioprotection

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2008-01-01

    Two interactive computer codes for education and training on nuclear safety and radioprotection developed at RA6 Reactor Division-Bariloche Atomic Center-CNEA are presented on this paper. The first code named SIMREACT has been developed in order to simulate the control of a research nuclear reactor in real time with a simple but accurate approach. The code solves the equations of neutron punctual kinetics with time variable reactivity. Utilizing the timer of the computer and the controls of a PC keyboard, with an adequate graphic interface, a simulation in real time of the temporal behavior of a research reactor is obtained. The reactivity can be changed by means of the extraction or insertion of control rods. It was implemented also the simulation of automatic pilot and scram. The use of this code is focalized on practices of nuclear reactor control like start-up from the subcritical state with external source up to power to a desired level, change of power level, calibration of a control rod with different methods, and approach to critical condition by interpolation of the answer in function of reactivity. The second code named LICEN has been developed in order to help the studies of all the topics included in examination programs for obtaining licenses for research reactor operators and radioprotection officials. Using the PC mouse, with an adequate graphic interface, the student can gradually learn the topics related with general and special licenses. The general option includes nuclear reactor engineering, radioprotection, nuclear safety, documentation and normative. The specific option includes mandatory documentation, description of the installation and task on normal and emergency situations. For each of these topics there are sub-items with all the relevant information. The objective of this code is to joint in one electronic place a large amount of information which usually it is disseminated on difficult to find separated papers. (author)

  7. Revised IAEA Code of Conduct on the Safety and Security of Radioactive Sources

    International Nuclear Information System (INIS)

    Wheatley, J. S.

    2004-01-01

    The revised Code of Conduct on the Safety and Security of Radioactive Sources is aimed primarily at Governments, with the objective of achieving and maintaining a high level of safety and security of radioactive sources through the development, harmonization and enforcement of national policies, laws and regulations; and through the fostering of international co-operation. It focuses on sealed radioactive sources and provides guidance on legislation, regulations and the regulatory body, and import/export controls. Nuclear materials (except for sources containing 239Pu), as defined in the Convention on the Physical Protection of Nuclear Materials, are not covered by the revised Code, nor are radioactive sources within military or defence programmes. An earlier version of the Code was published by IAEA in 2001. At that time, agreement was not reached on a number of issues, notably those relating to the creation of comprehensive national registries for radioactive sources, obligations of States exporting radioactive sources, and the possibility of unilateral declarations of support. The need to further consider these and other issues was highlighted by the events of 11th September 2001. Since then, the IAEA's Secretariat has been working closely with Member States and relevant International Organizations to achieve consensus. The text of the revised Code was finalized at a meeting of technical and legal experts in August 2003, and it was submitted to IAEA's Board of Governors for approval in September 2003, with a recommendation that the IAEA General Conference adopt it and encourage its wide implementation. The IAEA General Conference, in September 2003, endorsed the revised Code and urged States to work towards following the guidance contained within it. This paper summarizes the history behind the revised Code, its content and the outcome of the discussions within the IAEA Board of Governors and General Conference. (Author) 8 refs

  8. Stochastic methods for uncertainty treatment of functional variables in computer codes: application to safety studies

    International Nuclear Information System (INIS)

    Nanty, Simon

    2015-01-01

    This work relates to the framework of uncertainty quantification for numerical simulators, and more precisely studies two industrial applications linked to the safety studies of nuclear plants. These two applications have several common features. The first one is that the computer code inputs are functional and scalar variables, functional ones being dependent. The second feature is that the probability distribution of functional variables is known only through a sample of their realizations. The third feature, relative to only one of the two applications, is the high computational cost of the code, which limits the number of possible simulations. The main objective of this work was to propose a complete methodology for the uncertainty analysis of numerical simulators for the two considered cases. First, we have proposed a methodology to quantify the uncertainties of dependent functional random variables from a sample of their realizations. This methodology enables to both model the dependency between variables and their link to another variable, called co-variate, which could be, for instance, the output of the considered code. Then, we have developed an adaptation of a visualization tool for functional data, which enables to simultaneously visualize the uncertainties and features of dependent functional variables. Second, a method to perform the global sensitivity analysis of the codes used in the two studied cases has been proposed. In the case of a computationally demanding code, the direct use of quantitative global sensitivity analysis methods is intractable. To overcome this issue, the retained solution consists in building a surrogate model or meta model, a fast-running model approximating the computationally expensive code. An optimized uniform sampling strategy for scalar and functional variables has been developed to build a learning basis for the meta model. Finally, a new approximation approach for expensive codes with functional outputs has been

  9. The Use of Coupled Code Technique for Best Estimate Safety Analysis of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Bousbia Salah, A.; D'Auria, F.

    2006-01-01

    Issues connected with the thermal-hydraulics and neutronics of nuclear plants still challenge the design, safety and the operation of Light Water nuclear Reactors (LWR). The lack of full understanding of complex mechanisms related to the interaction between these issues imposed the adoption of conservative safety limits. Those safety margins put restrictions on the optimal exploitation of the plants and consequently reduced economic profit of the plant. In the light of the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. In fact, during the last decades Best Estimate (BE) neutronic and thermal-hydraulic calculations were so far carried out following rather parallel paths with only few interactions between them. Nowadays, it becomes possible to switch to new generation of computational tools, namely, Coupled Code technique. The application of such method is mandatory for the analysis of accident conditions where strong coupling between the core neutronics and the primary circuit thermal-hydraulics, and more especially when asymmetrical processes take place in the core leading to local space-dependent power generation. Through the current study, a demonstration of the maturity level achieved in the calculation of 3-D core performance during complex accident scenarios in NPPs is emphasized. Typical applications are outlined and discussed showing the main features and limitations of this technique. (author)

  10. The use of case tools in OPG safety analysis code qualification

    International Nuclear Information System (INIS)

    Pascoe, J.; Cheung, A.; Westbye, C.

    2001-01-01

    Ontario Power Generation (OPG) is currently qualifying its critical safety analysis software. The software quality assurance (SQA) framework is described. Given the legacy nature of much of the safety analysis software the reverse engineering methodology has been adopted. The safety analysis suite of codes was developed over a period of many years to differing standards of quality and had sparse or incomplete documentation. Key elements of the reverse engineering process require recovery of design information from existing coding. This recovery, if performed manually, could represent an enormous effort. Driven by a need to maximize productivity and enhance the repeatability and objectivity of software qualification activities the decision was made to acquire or develop and implement Computer Aided Software Engineering (CASE) tools. This paper presents relevant background information on CASE tools and discusses how the OPG SQA requirements were used to assess the suitability of available CASE tools. Key findings from the application of CASE tools to the qualification of the OPG safety analysis software are discussed. (author)

  11. Current status and applications of intergrated safety assessment and simulation code system for ISA

    Energy Technology Data Exchange (ETDEWEB)

    Izquierdo, J. M.; Hortal, J.; Perea, M. Sanchez; Melendez, E. [Modeling and Simulation Area (MOSI), Nuclear Safety Council (CSN), Madrid (Spain); Queral, E.; Rivas-Lewicky, J. [Energy and Fuels Department, Technical University of Madrid (UPM), Madrid (Spain)

    2017-03-15

    This paper reviews current status of the unified approach known as integrated safety assessment (ISA), as well as the associated SCAIS (simulation codes system for ISA) computer platform. These constitute a proposal, which is the result of collaborative action among the Nuclear Safety Council (CSN), University of Madrid (UPM), and NFQ Solutions S.L, aiming to allow independent regulatory verification of industry quantitative risk assessments. The content elaborates on discussions of the classical treatment of time in conventional probabilistic safety assessment (PSA) sequences and states important conclusions that can be used to avoid systematic and unacceptable underestimation of the failure exceedance frequencies. The unified ISA method meets this challenge by coupling deterministic and probabilistic mutual influences. The feasibility of the approach is illustrated with some examples of its application to a real size plant.

  12. Application of best estimate thermalhydraulic codes for the safety analysis of research reactors

    International Nuclear Information System (INIS)

    Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

    2006-01-01

    An established international expertise in relation to computational tools, procedures for their application including Best Estimate (BE) methods supported by uncertainty evaluation, and comprehensive experimental database exists within the safety technology of Nuclear Power Plant (NPP). The importance of transferring NPP safety technology tools and methods to RR safety technology has been noted in recent IAEA activities. However, the ranges of parameters of interest to RR are different from those for NPP: this is namely true for fuel composition, system pressure, adopted materials and overall system geometric configuration. The large variety of research reactors prevented so far the achievement of systematic and detailed lists of initiating events based upon qualified Probabilistic Safety Assessment (PSA) studies with results endorsed by the international community. However, bounding and generalized lists of events are available from IAEA documents and can be considered for deeper studies in the area. In the area of acceptance criteria, established standards accepted by the international community are available. Therefore no major effort is needed, but an effort appears worthwhile to check that those standards are adopted and that the related thresholds are fulfilled. The importance of suitable experimental assessment is recognized. A large amount of data exists as the kinetic dynamic core behaviour form SPERT reactors tests. However, not all data are accessible to all institutions and the relationship between the range of parameters of experiments and the range of parameters relevant to RR technology is not always established. However, code-assessment through relevant set of experimental data are recorded and properly stored. An established technology exists for development, qualification and application of system thermal-hydraulics codes suitable to be adopted for accident analysis in research reactors. This derives from NPP technology. The applicability of

  13. Staff report on the environmental qualification of safety-related electrical equipment

    International Nuclear Information System (INIS)

    1977-12-01

    The current NRC safety review process for nuclear power plants includes criteria related to the qualification of certain electrical equipment. These criteria require that electrical equipment important to safety must be qualified to function in the environment that might result from various accident conditions. Although such criteria have been applied since the early days of commercial nuclear power, the details of these criteria have been changed over the years. The evolution of environmental qualification of safety-related electrical equipment is described in Appendix A

  14. Safety studies of plasma-wall events with AINA code for Japanese DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Rivas, J.C., E-mail: jose.carlos.rivas@upc.edu [International Fusion Energy Research Centre (IFERC) (Japan); Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia-BarcelonaTech (Spain); Nakamura, M.; Someya, Y.; Hoshino, K.; Asakura, N. [Japan Atomic Energy Agency (JAEA) (Japan); Takase, H. [International Fusion Energy Research Centre (IFERC) (Japan); Miyoshi, Y.; Utoh, H.; Tobita, K. [Japan Atomic Energy Agency (JAEA) (Japan); Dies, J.; Blas, A. de; Riego, A.; Fabbri, M. [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia-BarcelonaTech (Spain)

    2016-11-01

    Highlights: • Work done in AINA code during 2014 and 2015 at IFERC to develop a version for safety studies of a Japanese DEMO design. • A thermal model for a WCPB breeding blanket has been developed based in parametric input data from neutronics calculations. • A breakthrough for the safety studies of plasma-divertor transients: An integrated SOL-pedestal model + using melting time as objective variable + using optimization algorithm. • The results for the case of divertor show that both loss of plasma control (LOPC) transients and ex-vessel LOCA transient can induce severe melting. The difference is that while in the first case melting happens at PFC surface, in the second case it happens at copper heat sink. • Conclusions suggest that, because the minimum melting times are same order of magnitude than the energy confinement time, recovery time for plasma control system should be lower order. - Abstract: In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times. The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case

  15. Safety studies of plasma-wall events with AINA code for Japanese DEMO

    International Nuclear Information System (INIS)

    Rivas, J.C.; Nakamura, M.; Someya, Y.; Hoshino, K.; Asakura, N.; Takase, H.; Miyoshi, Y.; Utoh, H.; Tobita, K.; Dies, J.; Blas, A. de; Riego, A.; Fabbri, M.

    2016-01-01

    Highlights: • Work done in AINA code during 2014 and 2015 at IFERC to develop a version for safety studies of a Japanese DEMO design. • A thermal model for a WCPB breeding blanket has been developed based in parametric input data from neutronics calculations. • A breakthrough for the safety studies of plasma-divertor transients: An integrated SOL-pedestal model + using melting time as objective variable + using optimization algorithm. • The results for the case of divertor show that both loss of plasma control (LOPC) transients and ex-vessel LOCA transient can induce severe melting. The difference is that while in the first case melting happens at PFC surface, in the second case it happens at copper heat sink. • Conclusions suggest that, because the minimum melting times are same order of magnitude than the energy confinement time, recovery time for plasma control system should be lower order. - Abstract: In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times. The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case

  16. Evaluation of temporary non-code repairs in safety class 3 piping systems

    International Nuclear Information System (INIS)

    Godha, P.C.; Kupinski, M.; Azevedo, N.F.

    1996-01-01

    Temporary non-ASME Code repairs in safety class 3 pipe and piping components are permissible during plant operation in accordance with Nuclear Regulatory Commission Generic Letter 90-05. However, regulatory acceptance of such repairs requires the licensee to undertake several timely actions. Consistent with the requirements of GL 90-05, this paper presents an overview of the detailed evaluation and relief request process. The technical criteria encompasses both ductile and brittle piping materials. It also lists appropriate evaluation methods that a utility engineer can select to perform a structural integrity assessment for design basis loading conditions to support the use of temporary non-Code repair for degraded piping components. Most use of temporary non-code repairs at a nuclear generating station is in the service water system which is an essential safety related system providing the ultimate heat sink for various plant systems. Depending on the plant siting, the service water system may use fresh water or salt water as the cooling medium. Various degradation mechanisms including general corrosion, erosion/corrosion, pitting, microbiological corrosion, galvanic corrosion, under-deposit corrosion or a combination thereof continually challenge the pressure boundary structural integrity. A good source for description of corrosion degradation in cooling water systems is provided in a cited reference

  17. Validation study of computer code SPHINCS for sodium fire safety evaluation of fast reactor

    International Nuclear Information System (INIS)

    Yamaguchi, Akira; Tajima, Yuji

    2003-01-01

    A computer code SPHINCS solves coupled phenomena of thermal hydraulics and sodium fire based on a multi-zone model. It deals with an arbitrary number of rooms, each of which is connected mutually by doorways and penetrations. With regard to the combustion phenomena, a flame sheet model and a liquid droplet combustion model are used for pool and spray fires, respectively, with the chemical equilibrium model based on the Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermalhydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The authors analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS code has been validated with regard to pool combustion phenomena. Further research needs are identified for pool spreading modeling considering thermal deformation of steel liner and measurement of pool fluidity property as a mixture of liquid sodium and reaction products. The SPHINCS code is to be used mainly in the safety evaluation of the consequence of a sodium fire accident in a liquid metal cooled fast reactor as well as fire safety analysis in general

  18. FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-30

    The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variable lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

  19. Current algorithms used in reactor safety codes and the impact of future computer development on these algorithms

    International Nuclear Information System (INIS)

    Mahaffy, J.H.; Liles, D.R.; Woodruff, S.B.

    1985-01-01

    Computational methods and solution procedures used in the US Nuclear Regulatory Commission's reactor safety systems codes, Transient Reactor Analysis Code (TRAC) and Reactor Leak and Power Safety Excursion Code (RELAP), are reviewed. Methods used in TRAC-PF1/MOD1, including the stability-enhancing two-step (SETS) technique, which permits fast computations by allowing time steps larger than the material Courant stability limit, are described in detail, and the differences from RELAP5/MOD2 are noted. Developments in computing, including parallel and vector processing, and their applicability to nuclear reactor safety codes are described. These developments, coupled with appropriate numerical methods, make detailed faster-than-real-time reactor safety analysis a realistic near-term possibility

  20. Authorization request for potential non-compliance with the American Standard Safety Code for Elevators Dumbwaiters and Escalators

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, J.E.

    1964-09-28

    A Third Party inspection of the reactor work platforms was conducted by representatives of the Travelers Insurance Company in 1958. An inspection report submitted by these representatives described hazardous conditions noted and presented a series of recommendations to improve the operational safety of the systems. Project CGI-960, ``C`` & ``D`` Work Platform Safety Improvements -- All Reactors, vas initiated to modify the platforms in compliance with the Third Party recommendations. The American Standard Safety Code for Elevators Dumbwaiters and Escalators (A-17.1) is used as a guide by the Third Party in formulating their recommendations. This code is used because there is no other applicable code for this type of equipment. While the work platforms do not and in some cases can not comply with this code because of operational use, every effort is made to comply with the intent of the code.

  1. Data and code for the exploratory data analysis of the electrical energy demand in the time domain in Greece.

    Science.gov (United States)

    Tyralis, Hristos; Karakatsanis, Georgios; Tzouka, Katerina; Mamassis, Nikos

    2017-08-01

    We present data and code for visualizing the electrical energy data and weather-, climate-related and socioeconomic variables in the time domain in Greece. The electrical energy data include hourly demand, weekly-ahead forecasted values of the demand provided by the Greek Independent Power Transmission Operator and pricing values in Greece. We also present the daily temperature in Athens and the Gross Domestic Product of Greece. The code combines the data to a single report, which includes all visualizations with combinations of all variables in multiple time scales. The data and code were used in Tyralis et al. (2017) [1].

  2. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  3. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  4. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  5. Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-

    International Nuclear Information System (INIS)

    Park, Chang Kyu; Kim, Tae Woon; Ha, Jae Joo; Han, Sang Hoon; Cho, Yeong Kyun; Jeong, Won Dae; Jang, Seung Cheol; Choi, Young; Seong, Tae Yong; Kang, Dae Il; Hwang, Mi Jeong; Choi, Seon Yeong; An, Kwang Il

    1994-07-01

    This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The Improvement of Level-1 PSA Computer Codes' is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author)

  6. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1996-03-01

    During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

  7. Application of an integrated PC-based neutronics code system to criticality safety

    International Nuclear Information System (INIS)

    Briggs, J.B.; Nigg, D.W.

    1991-01-01

    An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)

  8. Status of SACRD: a data base for fast reactor safety computer codes

    International Nuclear Information System (INIS)

    Greene, N.M.; Flanagan, G.F.; Alter, H.

    1982-01-01

    In 1975 work was initiated to provide a central computerized data collection of evaluated data for use in fast reactor safety computer codes. This data base is called SACRD and is intended to encompass handbook and other nonproblem-dependent data related to LMFBR's, especially at extreme conditions where little or no experimental data are available. Version 1 of the data base was released in the latter part of 1978 and remained the standard version until Version 81, which was released in October 1981

  9. Code development incorporating environmental, safety and economic aspects of fusion reactors

    International Nuclear Information System (INIS)

    Fowler, T.K.; Greenspan, E.; Holdren, J.P.

    1993-01-01

    This document is a proposal to continue the authors work on the Environmental, Safety and Economic (ESE) aspects of fusion reactors under DOE contract DE-FR03-89ER52514. The grant objectives continue those from the previous grant: (1) completion of first-generation Environmental, Safety and Economic (ESE) computer modules suitable as integral components of tokamak systems codes. (2) continuation of work on special topics, in support of the above and in response to OFE requests. The proposal also highlights progress on the contract in the twelve months since April, 1992. This has included work with the ARIES and ITER design teams, work on tritium management, studies on materials activation, and calculation of radioactive inventories in fusion reactors

  10. Testing the new stochastic neutronic code ANET in simulating safety important parameters

    International Nuclear Information System (INIS)

    Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.

    2017-01-01

    Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement

  11. Japanese standard method for safety evaluation using best estimate code based on uncertainty and scaling analyses with statistical approach

    International Nuclear Information System (INIS)

    Mizokami, Shinya; Hotta, Akitoshi; Kudo, Yoshiro; Yonehara, Tadashi; Watada, Masayuki; Sakaba, Hiroshi

    2009-01-01

    Current licensing practice in Japan consists of using conservative boundary and initial conditions(BIC), assumptions and analytical codes. The safety analyses for licensing purpose are inherently deterministic. Therefore, conservative BIC and assumptions, such as single failure, must be employed for the analyses. However, using conservative analytical codes are not considered essential. The standard committee of Atomic Energy Society of Japan(AESJ) has drawn up the standard for using best estimate codes for safety analyses in 2008 after three-years of discussions reflecting domestic and international recent findings. (author)

  12. Acoustic characteristics of hybrid electric vehicles and the safety of pedestrians who are blind

    Science.gov (United States)

    2010-08-01

    Quieter cars such as electric vehicles (EVs) and hybrid electric vehicles (HEVs) may reduce auditory cues used by pedestrians to assess the state of nearby traffic and, as a result, their use may have an adverse impact on pedestrian safety. In order ...

  13. 78 FR 2797 - Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and Electric Vehicles

    Science.gov (United States)

    2013-01-14

    ... Sound Requirements for Hybrid and Electric Vehicles; Draft Environmental Assessment for Rulemaking To Establish Minimum Sound Requirements for Hybrid and Electric Vehicles; Proposed Rules #0;#0;Federal Register...-0148] RIN 2127-AK93 Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and...

  14. Fundamental challenging problems for developing new nuclear safety standard computer codes

    International Nuclear Information System (INIS)

    Wong, P.K.; Wong, A.E.; Wong, A.

    2005-01-01

    Based on the claims of the US Basic patents number 5,084,232; 5,848,377 and 6,430,516 that can be obtained from typing the Patent Numbers into the Box of the Web site http://164.195.100.11/netahtml/srchnum.htm and their associated published technical papers having been presented and published at International Conferences in the last three years and that all these had been sent into US-NRC by E-mail on March 26, 2003 at 2:46 PM., three fundamental challenging problems for developing new nuclear safety standard computer codes had been presented at the US-NRC RIC2003 Session W4. 2:15-3:15 PM. at the Washington D.C. Capital Hilton Hotel, Presidential Ballroom on April 16, 2003 in front of more than 800 nuclear professionals from many countries worldwide. The objective and scope of this paper is to invite all nuclear professionals to examine and evaluate all the current computer codes being used in their own countries by means of comparison of numerical data from these three specific openly challenging fundamental problems in order to set up a global safety standard for all nuclear power plants in the world. (authors)

  15. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, Peter Andrew

    2011-12-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  16. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    International Nuclear Information System (INIS)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on 'The improvement of level-1 PSA computer codes' is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author)

  17. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC)

    International Nuclear Information System (INIS)

    Schultz, Peter Andrew

    2011-01-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  18. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on `The improvement of level-1 PSA computer codes` is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author).

  19. The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

    International Nuclear Information System (INIS)

    Hall, P.; Hutt, P.

    1994-01-01

    This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

  20. Interface design of VSOP'94 computer code for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Natsir, Khairina, E-mail: yenny@batan.go.id; Andiwijayakusuma, D.; Wahanani, Nursinta Adi [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia); Yazid, Putranto Ilham [Center for Nuclear Technology, Material and Radiometry- National Nuclear Energy Agency, Jl. Tamansari No.71, Bandung 40132 (Indonesia)

    2014-09-30

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

  1. Code coverage measurement methodology for MMI software of safety-class I and C system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eun Hyung; Jung, Beom Young; Choi, Seok Joo [Suresofttech, Seoul (Korea, Republic of)

    2016-10-15

    MMI (Man-Machine Interface) software of the safety instrumentation and control system used in nuclear power plants carry out an important functions, such as displaying and transmitting the commend to another system, and change setpoints the safety-related information. Yet, this has been recognized reliability of the MMI software plays an important role in enhancing nuclear power plants are operating, regulatory standards have been strengthened with it. Strengthening of regulatory standards has affected even perform software testing soon, and accordingly, the current regulatory require the measurement of code coverage with legal standard. In this paper, it poses a problem of the conventional method used for measuring the above-mentioned code coverage, presents a new coverage measuring method for solving the exposed problems. In this paper, we checked the problems such as limit and the low efficiency of the existing test coverage measuring method on the MMI software using in nuclear power instrumentation and control systems, and it proposed a new test coverage measuring method as a solution for this. If you apply a new method of Top-Down approach, can mitigate all of the problems of existing test coverage measurement methods and possible coverage achievement of the desired objectives. Of course, it is still necessary to secure more cases, and the methodology should be systematization based on the cases. Thus, if later the efficient and reliable are ensured through the application in many cases, as well as nuclear power instrumentation and control, may be used to ensure code coverage of software of the many areas where the GUI is utilized.

  2. Cellular and circuit properties supporting different sensory coding strategies in electric fish and other systems.

    Science.gov (United States)

    Marsat, Gary; Longtin, André; Maler, Leonard

    2012-08-01

    Neural codes often seem tailored to the type of information they must carry. Here we contrast the encoding strategies for two different communication signals in electric fish and describe the underlying cellular and network properties that implement them. We compare an aggressive signal that needs to be quickly detected, to a courtship signal whose quality needs to be evaluated. The aggressive signal is encoded by synchronized bursts and a predictive feedback input is crucial in separating background noise from the communication signal. The courtship signal is accurately encoded through a heterogenous population response allowing the discrimination of signal differences. Most importantly we show that the same strategies are used in other systems arguing that they evolved similar solutions because they faced similar tasks. Copyright © 2012 Elsevier Ltd. All rights reserved.

  3. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

  4. Safety leadership in the teaching laboratories of electrical and electronic engineering departments at Taiwanese Universities.

    Science.gov (United States)

    Wu, Tsung-Chih

    2008-01-01

    Safety has always been one of the principal goals in teaching laboratories. Laboratories cannot serve their educational purpose when accidents occur. The leadership of department heads has a major impact on laboratory safety, so this study discusses the factors affecting safety leadership in teaching laboratories. This study uses a mail survey to explore the perceived safety leadership in electrical and electronic engineering departments at Taiwanese universities. An exploratory factor analysis shows that there are three main components of safety leadership, as measured on a safety leadership scale: safety controlling, safety coaching, and safety caring. The descriptive statistics also reveals that among faculty, the perception of department heads' safety leadership is in general positive. A two-way MANOVA shows that there are interaction effects on safety leadership between university size and instructor age; there are also interaction effects between presence of a safety committee and faculty gender and faculty age. It is therefore necessary to assess organizational factors when determining whether individual factors are the cause of differing perceptions among faculty members. The author also presents advice on improving safety leadership for department heads at small universities and at universities without safety committees.

  5. Utility regulation-The scope and structure of electrical safety regulation

    International Nuclear Information System (INIS)

    Abbott, Malcolm; Cohen, Bruce

    2011-01-01

    As a consequence of policies in Australia and New Zealand to increase competition in the utilities sector, regulatory agencies have been created in each state to provide independent and authorative advice on matters such as electricity pricing, access to infrastructure, service quality and security of supply. In addition arrangements have been established to maintain safety standards in the industry. The purpose of this paper is to discuss the major issues that have arisen in the creation of regulatory agencies responsible for electrical safety standards in Australia and New Zealand, and how they have impacted on liberalised electricity markets. - Highlights: → Policies in Australia and New Zealand to increase competition have led to the creation of electrical safety agencies. → These agencies have been created in response to perceived market failures. → There is a variance in agencies in terms of their independence and industry coverage. → These agencies have been created at a time of falling fatalities.

  6. Safety assessment of emergency electric power systems for nuclear power plants

    International Nuclear Information System (INIS)

    1986-09-01

    This paper is intended to assist the safety assessor within a regulatory body, or one working as a consultant, in assessing a given design of the Emergency Electrical Power System. Those non-electric power systems which may be used in a plant design to serve as emergency energy sources are addressed only in their general safety aspects. The paper thus relates closely to Safety Series 50-SG-D7 ''Emergency Power Systems at Nuclear Power Plants'' (1982), as far as it addresses emergency electric power systems. Several aspects are dealt with: the information the assessor may expect from the applicant to fulfill his task of safety review; the main questions the reviewer has to answer in order to determine the compliance with requirements of the NUSS documents; the national or international standards which give further guidance on a certain system or piece of equipment; comments and suggestions which may help to judge a variety of possible solutions

  7. The impact of changing computing technology on EPRI [Electric Power Research Institute] nuclear analysis codes

    International Nuclear Information System (INIS)

    Breen, R.J.

    1988-01-01

    The Nuclear Reload Management Program of the Nuclear Power Division (NPD) of the Electric Power Research Institute (EPRI) has the responsibility for initiating and managing applied research in selected nuclear engineering analysis functions for nuclear utilities. The computer systems that result from the research projects consist of large FORTRAN programs containing elaborate computational algorithms used to access such areas as core physics, fuel performance, thermal hydraulics, and transient analysis. This paper summarizes a study of computing technology trends sponsored by the NPD. The approach taken was to interview hardware and software vendors, industry observers, and utility personnel focusing on expected changes that will occur in the computing industry over the next 3 to 5 yr. Particular emphasis was placed on how these changes will impact engineering/scientific computer code development, maintenance, and use. In addition to the interviews, a workshop was held with attendees from EPRI, Power Computing Company, industry, and utilities. The workshop provided a forum for discussing issues and providing input into EPRI's long-term computer code planning process

  8. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  9. Constitutive model development needs for reactor safety thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Kelly, J.M.

    1998-01-01

    This paper discusses the constitutive model development needs for our current and future generation of reactor safety thermal-hydraulic analysis codes. Rather than provide a simple 'shopping list' of models to be improved, a detailed description is given of how a constitutive model works within the computational framework of a current reactor safety code employing the two-fluid model of two-phase flow. The intent is to promote a better understanding of both the types of experiments and the instrumentation needs that will be required in the USNRCs code improvement program. First, a summary is given of the modeling considerations that need to be taken into account when developing constitutive models for use in reactor safety thermal-hydraulic codes. Specifically, the two-phase flow model should be applicable to a control volume formulation employing computational volumes with dimensions on the order of meters but containing embedded structure with a dimension on the order of a centimeter. The closure relations are then required to be suitable when averaged over such large volumes containing millions or even tens of millions of discrete fluid particles (bubbles/drops). This implies a space and time averaging procedure that neglects the intermittency observed in slug and chum turbulent two-phase flows. Furthermore, the geometries encountered in reactor systems are complex, the constitutive relations should therefore be component specific (e.g., interfacial shear in a tube does not represent that in a rod bundle nor in the downcomer). When practicable, future modeling efforts should be directed towards resolving the spatial evolution of two-phase flow patterns through the introduction of interfacial area transport equations and by modeling the individual physical processes responsible for the creation or destruction of interfacial area. Then the example of the implementation and assessment of a subcooled boiling model in a two-fluid code is given. The primary parameter

  10. Electrical drives of the safety system in nuclear power plants

    International Nuclear Information System (INIS)

    1990-09-01

    Actuating drives, control magnets for ventilators, machine drives and control member drives are part of this rule. The rule deals with the security and technical requirements for design, construction, calculation, fabrication, assembling, testing and operation. Furthermore, it places significant demands, with regard to planning and arrangement of electrical drives, on the accompanying technical systems. Furthermore, demands are placed on the aggregate protection for electrical drives of the security systems. The signals given to these systems do not, however, have precedence over the protection signals of the reactor. The rule is identical with KTA-3504, version 9/1988. (orig./HP) [de

  11. Development of SAGE, A computer code for safety assessment analyses for Korean Low-Level Radioactive Waste Disposal

    International Nuclear Information System (INIS)

    Zhou, W.; Kozak, Matthew W.; Park, Joowan; Kim, Changlak; Kang, Chulhyung

    2002-01-01

    This paper describes a computer code, called SAGE (Safety Assessment Groundwater Evaluation) to be used for evaluation of the concept for low-level waste disposal in the Republic of Korea (ROK). The conceptual model in the code is focused on releases from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. Doses can be calculated for several biosphere systems including drinking contaminated groundwater, and subsequent contamination of foods, rivers, lakes, or the ocean by that groundwater. The flexibility of the code will permit both generic analyses in support of design and site development activities, and straightforward modification to permit site-specific and design-specific safety assessments of a real facility as progress is made toward implementation of a disposal site. In addition, the code has been written to easily interface with more detailed codes for specific parts of the safety assessment. In this way, the code's capabilities can be significantly expanded as needed. The code has the capability to treat input parameters either deterministic ally or probabilistic ally. Parameter input is achieved through a user-friendly Graphical User Interface.

  12. Outlines and verifications of the codes used in the safety analysis of High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Kunitomi, Kazuhiko; Maruyama, Soh; Fujita, Shigeki; Nakagawa, Shigeaki; Iyoku, Tatsuo; Shindoh, Masami; Sudo, Yukio; Hirano, Masashi.

    1990-03-01

    This paper presents brief description of the computer codes used in the safety analysis of High Temperature Engineering Test Reactor. The list of the codes is: 1. BLOOST-J2 2. THYDE-HTGR 3. TAC-NC 4. RATSAM6 5. COMPARE-MOD1 6. GRACE 7. OXIDE-3F 8. FLOWNET/TRUMP. Of described above, 1, 3, 4, 5, 6 and 7 were developed for the multi-hole type gas cooled reactor and improved for HTTR and 2 was originated by THYDE-codes which were developed to treat the transient thermo-hydraulics during LOCA of LWR. Each code adopted the models and properties which yield conservative analytical results. Adequacy of each code was verified by the comparison with the experimental results and/or the analytical results obtained from the other codes which were already proven. (author)

  13. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    International Nuclear Information System (INIS)

    Chang Oh

    2006-01-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 900 C and operational fuel temperatures above 1250 C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR's higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  14. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  15. Report of the Task Group on Electrical Safety of Department of Energy facilities

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-01-01

    The Task Group on Electrical Safety at DOE Facilities (Task Group), which was formally established on October 27, 1992. The Task Group reviewed the electrical safety-related occurrence history of, and conducted field visits to, seven DOE sites chosen to represent a cross section of the Department`s electrical safety activities. The purpose of the field visits was to review, firsthand, electrical safety programs and practices and to gain greater insight to the root causes and corrective actions taken for recently reported incidents. The electrical safety environment of the DOE complex is extremely varied, ranging from common office and industrial electrical systems to large high-voltage power distribution systems (commercial transmission line systems). It includes high-voltage/high-power systems associated with research programs such as linear accelerators and experimental fusion confinement systems. Age, condition, and magnitude of the facilities also varies, with facilities dating from the Manhattan Project, during World War II, to the most modem complexes. The complex is populated by Federal (DOE and other agencies) and contractor employees engaged in a wide variety of occupations and activities in office, research and development, and industrial settings. The sites visited included all of these variations and are considered by the Task Group to offer a valid representation of the Department`s electrical safety issues. The sites visited were Oak Ridge National Laboratory (ORNL), Stanford Linear Accelerator Center (SLAC), Idaho National Engineering Laboratory (INEL), Nevada Test Site (NTS), Savannah River Site (SRS), Hanford Reservation (Hanford), and the Uranium Mill Tailings Remedial Action Project (UMTRA) located at Grand Junction, Colorado.

  16. A Novel Series Connected Batteries State of High Voltage Safety Monitor System for Electric Vehicle Application

    Directory of Open Access Journals (Sweden)

    Qiang Jiaxi

    2013-01-01

    Full Text Available Batteries, as the main or assistant power source of EV (Electric Vehicle, are usually connected in series with high voltage to improve the drivability and energy efficiency. Today, more and more batteries are connected in series with high voltage, if there is any fault in high voltage system (HVS, the consequence is serious and dangerous. Therefore, it is necessary to monitor the electric parameters of HVS to ensure the high voltage safety and protect personal safety. In this study, a high voltage safety monitor system is developed to solve this critical issue. Four key electric parameters including precharge, contact resistance, insulation resistance, and remaining capacity are monitored and analyzed based on the equivalent models presented in this study. The high voltage safety controller which integrates the equivalent models and control strategy is developed. By the help of hardware-in-loop system, the equivalent models integrated in the high voltage safety controller are validated, and the online electric parameters monitor strategy is analyzed and discussed. The test results indicate that the high voltage safety monitor system designed in this paper is suitable for EV application.

  17. A novel series connected batteries state of high voltage safety monitor system for electric vehicle application.

    Science.gov (United States)

    Jiaxi, Qiang; Lin, Yang; Jianhui, He; Qisheng, Zhou

    2013-01-01

    Batteries, as the main or assistant power source of EV (Electric Vehicle), are usually connected in series with high voltage to improve the drivability and energy efficiency. Today, more and more batteries are connected in series with high voltage, if there is any fault in high voltage system (HVS), the consequence is serious and dangerous. Therefore, it is necessary to monitor the electric parameters of HVS to ensure the high voltage safety and protect personal safety. In this study, a high voltage safety monitor system is developed to solve this critical issue. Four key electric parameters including precharge, contact resistance, insulation resistance, and remaining capacity are monitored and analyzed based on the equivalent models presented in this study. The high voltage safety controller which integrates the equivalent models and control strategy is developed. By the help of hardware-in-loop system, the equivalent models integrated in the high voltage safety controller are validated, and the online electric parameters monitor strategy is analyzed and discussed. The test results indicate that the high voltage safety monitor system designed in this paper is suitable for EV application.

  18. Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

  19. Utilisation of best estimate system codes and best estimate methods in safety analyses of VVER reactors in the Czech Republic

    International Nuclear Information System (INIS)

    Macek, Jiri; Kral, Pavel

    2010-01-01

    The content of the presentation was as follows: Conservative versus best estimate approach, Brief description and selection of methodology, Description of uncertainty methods, Examples of the BE methodology. It is concluded that where BE computer codes are used, uncertainty and sensitivity analyses should be included; if best estimate codes + uncertainty are used, the safety margins increase; and BE + BSA is the next step in licensing analyses. (P.A.)

  20. Inventory of Safety-related Codes and Standards for Energy Storage Systems with some Experiences related to Approval and Acceptance

    Energy Technology Data Exchange (ETDEWEB)

    Conover, David R.

    2014-09-11

    The purpose of this document is to identify laws, rules, model codes, codes, standards, regulations, specifications (CSR) related to safety that could apply to stationary energy storage systems (ESS) and experiences to date securing approval of ESS in relation to CSR. This information is intended to assist in securing approval of ESS under current CSR and to identification of new CRS or revisions to existing CRS and necessary supporting research and documentation that can foster the deployment of safe ESS.

  1. Design and implementation of safety traceability system for candied fruits based on two-dimension code technology

    Directory of Open Access Journals (Sweden)

    ZHAO Kun

    2014-12-01

    Full Text Available Traceability is the basic principle of food safety.A food safety traceability system based on QR code and cloud computing technology was introduced in this paper.First of all we introduced the QR code technology and the concept of traceability.And then through the field investigation,we analyzed the traceability process.At the same time,we designed the system and database were found,and the consumer experiencing technology is studied.Finally we expounded the traceability information collection,transmission and final presentation style and expected the future development of traceability system.

  2. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

    International Nuclear Information System (INIS)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-01-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  3. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  4. Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

    International Nuclear Information System (INIS)

    2014-12-01

    In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

  5. Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-12-15

    In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

  6. Overview of Development and Deployment of Codes, Standards and Regulations Affecting Energy Storage System Safety in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Conover, David R.

    2014-08-22

    This report acquaints stakeholders and interested parties involved in the development and/or deployment of energy storage systems (ESS) with the subject of safety-related codes, standards and regulations (CSRs). It is hoped that users of this document gain a more in depth and uniform understanding of safety-related CSR development and deployment that can foster improved communications among all ESS stakeholders and the collaboration needed to realize more timely acceptance and approval of safe ESS technology through appropriate CSR.

  7. Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

    Directory of Open Access Journals (Sweden)

    Enrico Zio

    2008-01-01

    Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

  8. Low-frequency electrical dosimetry: research agenda of the IEEE International Committee on Electromagnetic Safety.

    Science.gov (United States)

    Reilly, J Patrick; Hirata, Akimasa

    2016-06-21

    This article treats unsettled issues in the use of numerical models of electrical dosimetry as applied to international limits on human exposure to low-frequency (typically  IEEE-ICES (International Committee on Electromagnetic Safety) Technical Committee 95. The paper discusses 25 issues needing attention, fitting into three general categories: induction models; electrostimulation models; and human exposure limits. Of these, 9 were voted as 'high priority' by members of Subcommittee 6. The list is presented as a research agenda for refinements in numerical modeling with applications to human exposure limits. It is likely that such issues are also important in medical and electrical product safety design applications.

  9. A Wii-controlled safety device for electric chainsaws

    Directory of Open Access Journals (Sweden)

    R. Gubiani

    2013-09-01

    Full Text Available Forestry continues to represent one of the most hazardous economic sectors of human activity, and historically, the operation of chainsaws has mainly been restricted to professional lumberjacks. In recent years, because of low cost, chainsaws have become popular among unprofessionals, e.g. for cutting firewood and trimming trees. Serious or lethal lesions due to the use of chainsaws or electric chainsaws are often observed by traumatologists or forensic pathologists. Such serious accidents often occur during occupational activities and are essentially due to kickback or uncorrected use of the tool, or when the operator falls down losing the control of the implement. A new device in order to stop a cutting chain was developed and adapted to an electric chainsaw. The device is based on a Wiimote controller (Nintendo™, including two accelerometers and two gyroscopes for detecting rotation and inclination. A Bluetooth wireless technology is used to transfer data to a portable computer. The data collected about linear and angular acceleration are filtered by an algorithm, based on the Euclid norm, capable to distinguishing between normal movements and dangerous chainsaw movements. The result show a good answer to device and when happen a dangerous situation an alarm signal is sent back to the implement in order to stop the cutting chain. The device show a correct behavior in tested dangerous situations and is envisaged to extend to combustion engine chainsaws, as well as to other portable equipment used in agriculture and forestry operations and for this objectives were patented.

  10. Can envelope codes reduce electricity and CO2 emissions in different types of buildings in the hot climate of Bahrain?

    International Nuclear Information System (INIS)

    Radhi, H.

    2009-01-01

    The depletion of non-renewable resources and the environmental impact of energy consumption, particularly energy use in buildings, have awakened considerable interest in energy efficiency. Building energy codes have recently become effective techniques to achieve efficiency targets. The Electricity and Water Authority in Bahrain has set a target of 40% reduction of building electricity consumption and CO 2 emissions to be achieved by using envelope thermal insulation codes. This paper investigates the ability of the current codes to achieve such a benchmark and evaluates their impact on building energy consumption. The results of a simulation study are employed to investigate the impact of the Bahraini codes on the energy and environmental performance of buildings. The study focuses on air-conditioned commercial buildings and concludes that envelope codes, at best, are likely to reduce the energy use of the commercial sector by 25% if the building envelope is well-insulated and efficient glazing is used. Bahraini net CO 2 emissions could drop to around 7.1%. The simulation results show that the current energy codes alone are not sufficient to achieve a 40% reduction benchmark, and therefore, more effort should be spent on moving towards a more comprehensive approach

  11. The safety and reliability of the electricity transportation network, and the opening of markets

    International Nuclear Information System (INIS)

    Merlin, A.

    2003-01-01

    Following a decade in which the electricity sector was concerned by the opening of markets the world over, the year 2003 highlighted the challenges of secure supply and of the safety of electrical systems. Major incidents are nothing new throughout history, and occurred well before the trend towards opening the markets dot underway. However, it is necessary to verify that the rules for the organisation, clarification of responsibilities and regulation make it possible to control the growing complexity of interconnected systems related to opening the market up to competition. Faced with such situations, changes are necessary in Europe. Firstly, it is important to have the second European directive on the internal electricity market introduced everywhere, this directive providing common rules to be met by all key players to ensure greater safety within the electricity system, and strengthening the role of the electricity transport network manager, who is separate from the other players in the market. It is also necessary to draft a third directive, focusing on both security of supply and the operational safety of the European electricity system. (authors)

  12. Assessment of CFD Codes for Nuclear Reactor Safety Problems - Revision 2

    International Nuclear Information System (INIS)

    Smith, B.L.; Andreani, M.; Bieder, U.; Ducros, F.; Bestion, D.; Graffard, E.; Heitsch, M.; Scheuerer, M.; Henriksson, M.; Hoehne, T.; Houkema, M.; Komen, E.; Mahaffy, J.; Menter, F.; Moretti, F.; Morii, T.; Muehlbauer, P.; Rohde, U.; Krepper, E.; Song, C.H.; Watanabe, T.; Zigh, G.; Boyd, C.F.; Archambeau, F.; Bellet, S.; Munoz-Cobo, J.M.; Simoneau, J.P.

    2015-01-01

    Following recommendations made at an 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety (NRS) Problems', held in Aix-en-Provence, France, 15-16 May, 2002, and a follow-up meeting 'Use of Computational Fluid Dynamics (CFD) Codes for Safety Analysis of Reactor Systems including Containment', which took place in Pisa on 11-14 Nov., 2002, a CSNI action plan was drawn up which resulted in the creation of three Writing Groups, with mandates to perform the following tasks: (1) Provide a set of guidelines for the application of CFD to NRS problems; (2) Evaluate the existing CFD assessment bases, and identify gaps that need to be filled; (3) Summarise the extensions needed to CFD codes for application to two-phase NRS problems. Work began early in 2003. In the case of Writing Group 2 (WG2), a preliminary report was submitted to Working Group on the Analysis and Management of Accidents (WGAMA) in September 2004 that scoped the work needed to be carried out to fulfil its mandate, and made recommendations on how to achieve the objective. A similar procedure was followed by the other two groups, and in January 2005 all three groups were reformed to carry out their respective tasks. In the case of WG2, this resulted in the issue of a CSNI report (NEA/CSNI/R(2007)13), issued in January 2008, describing the work undertaken. The writing group met on average twice per year during the period March 2005 to May 2007, and coordinated activities strongly with the sister groups WG1 (Best Practice Guidelines) and WG3 (Multiphase Extensions). The resulting document prepared at the end of this time still represents the core of the present revised version, though updates have been made as new material has become available. After some introductory remarks, Chapter 3 lists twenty-three (23) NRS issues for which it is considered that the application of CFD would bring real benefits

  13. Performance Assessment and analysis of national building codes with fire safety in all wards of a hospital

    Directory of Open Access Journals (Sweden)

    M. Mahdinia

    2009-04-01

    Full Text Available Background and aimsAIDS as a re-emergent disease and Viral hepatitis (B and C as one of thBackground and objective: Fire safety is an important problem in hospitals. Movement less, lack of awareness and special situation of residents are the reasons of this subject. In more countries such as Iran, fire protection regulations have compiled within the framework of national building codes. Current building codes don't create sufficient safety for patient in the hospitals in different situations and more of the advanced countries in the world effort to establish building code, base  on performance. This study to be accomplished with this goal that determination of fire risk level in the wards of a hospital and survey the efficiency of the national building codes. Methodsfire risk assesses is done, using "engineering fire risk assessment method" with the checklist for Data gathering. In this manner, risk calculate in all compartments and in the next  stage for survey the effect of building codes, with this supposition that all compartment is  conforming to building code requirement, risk level calculate in two situation.Resultsthe results of present study reveals that, risk level in all wards is more than one and even though risk less than one is acceptable, consequently minimum of safely situations didn't  produce in most wards. The results show the national building code in the different conditions  don't have appropriate efficient for creation of suitable safety. Conclusionin order to access to a fire safety design with sufficient efficiency, suitable selection is use of risk assessment based on, design methods.

  14. Safety and injury profile of conducted electrical weapons used by law enforcement officers against criminal suspects.

    Science.gov (United States)

    Bozeman, William P; Hauda, William E; Heck, Joseph J; Graham, Derrel D; Martin, Brian P; Winslow, James E

    2009-04-01

    Conducted electrical weapons such as the Taser are commonly used by law enforcement agencies. The safety of these weapons has been the subject of scrutiny and controversy; previous controlled studies in animals and healthy humans may not accurately reflect the risks of conducted electrical weapons used in actual conditions. We seek to determine the safety and injury profile of conducted electrical weapons used against criminal suspects in a field setting. This prospective, multicenter, observational trial tracked a consecutive case series of all conducted electrical weapon uses against criminal suspects at 6 US law enforcement agencies. Mandatory review of each conducted electrical weapon use incorporated physician review of police and medical records. Injuries were classified as mild, moderate, or severe according to a priori definitions. The primary outcome was a composite of moderate and severe injuries, termed significant injuries. Conducted electrical weapons were used against 1,201 subjects during 36 months. One thousand one hundred twenty-five subjects (94%) were men; the median age was 30 years (range 13 to 80 years). Mild or no injuries were observed after conducted electrical weapon use in 1,198 subjects (99.75%; 95% confidence interval 99.3% to 99.9%). Of mild injuries, 83% were superficial puncture wounds from conducted electrical weapon probes. Significant injuries occurred in 3 subjects (0.25%; 95% confidence interval 0.07% to 0.7%), including 2 intracranial injuries from falls and 1 case of rhabdomyolysis. Two subjects died in police custody; medical examiners did not find conducted electrical weapon use to be causal or contributory in either case. To our knowledge, these findings represent the first large, independent, multicenter study of conducted electrical weapon injury epidemiology and suggest that more than 99% of subjects do not experience significant injuries after conducted electrical weapon use.

  15. The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

    International Nuclear Information System (INIS)

    Caroli, Cataldo; Bleyer, Alexandre; Bentaib, Ahmed; Chatelard, Patrick; Cranga, Michel; Van Dorsselaere, Jean-Pierre

    2006-01-01

    IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE...). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and

  16. Application of probabilistic safety assessment for Macedonian electric power system

    International Nuclear Information System (INIS)

    Kancev, D.; Causevski, A.; Cepin, M.; Volkanovski, A.

    2007-01-01

    Due to the complex and integrated nature of a power system, failures in any part of the system can cause interruptions, which range from inconveniencing a small number of local residents to a major and widespread catastrophic disruption of supply known as blackout. The objective of the paper is to show that the methods and tools of probabilistic safety assessment are applicable for assessment and improvement of real power systems. The method used in this paper is developed based on the fault tree analysis and is adapted for the power system reliability analysis. A particular power system i.e. the Macedonian power system is the object of the analysis. The results show that the method is suitable for application of real systems. The reliability of Macedonian power system assumed as the static system is assessed. The components, which can significantly impact the power system are identified and analysed in more details. (author)

  17. 1E Qualification of Electrical Equipment - Requirement for Safety Nuclear Power Plants

    International Nuclear Information System (INIS)

    Geambasu, C.; Segarceanu, D.; Albu, J.

    2002-01-01

    The paper presents the qualification methods of the safety related equipment according to the safety class 1E. There are presented the qualification principles, procedure and documents, emphasis being laid on the qualification approach by type tests. This approach assumes the equipment test under both normal and accident conditions (design basis events) simulating the operational conditions and covers the largest part of electrical equipment from a nuclear power plant.The safety related equipment is to be qualified is subjected to a sequential test that will be detailed in the paper. (author)

  18. Computational Fluid Dynamics (CFD) in Nuclear Reactor Safety (NRS) - Proceedings of the workshop on Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS)

    International Nuclear Information System (INIS)

    2008-01-01

    Computational Fluid Dynamics (CFD) is to an increasing extent being adopted in nuclear reactor safety analyses as a tool that enables specific safety relevant phenomena occurring in the reactor coolant system to be better described. The Committee on the Safety of Nuclear Installations (CSNI), which is responsible for the activities of the OECD Nuclear Energy Agency that support advancing the technical base of the safety of nuclear installations, has in recent years conducted an important activity in the CFD area. This activity has been carried out within the scope of the CSNI working group on the analysis and management of accidents (GAMA), and has mainly focused on the formulation of user guidelines and on the assessment and verification of CFD codes. It is in this GAMA framework that a first workshop CFD4NRS was organized and held in Garching, Germany in 2006. Following the CFD4NRS workshop, this XCFD4NRS Workshop was intended to extend the forum created for numerical analysts and experimentalists to exchange information in the field of Nuclear Reactor Safety (NRS) related activities relevant to Computational Fluid Dynamics (CFD) validation, but this time with more emphasis placed on new experimental techniques and two-phase CFD applications. The purpose of the workshop was to provide a forum for numerical analysts and experimentalists to exchange information in the field of NRS-related activities relevant to CFD validation, with the objective of providing input to GAMA CFD experts to create a practical, state-of-the-art, web-based assessment matrix on the use of CFD for NRS applications. The scope of XCFD4NRS includes: - Single-phase and two-phase CFD simulations with an emphasis on validation in areas such as: boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These applications should relate to NRS-relevant issues such as: pressurized thermal shocks, critical heat flux, pool heat exchangers, boron dilution, hydrogen

  19. Environmental, health, and safety issues of sodium-sulfur batteries for electric and hybrid vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Ohi, J.M.

    1992-09-01

    This report is the first of four volumes that identify and assess the environmental, health, and safety issues involved in using sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles that may affect the commercialization of Na/S batteries. This and the other reports on recycling, shipping, and vehicle safety are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD D) program for Na/S battery technology. The reports review the status of Na/S battery RD D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers cell design and engineering as the basis of safety for Na/S batteries and describes and assesses the potential chemical, electrical, and thermal hazards and risks of Na/S cells and batteries as well as the RD D performed, under way, or to address these hazards and risks. The report is based on a review of the literature and on discussions with experts at DOE, national laboratories and agencies, universities, and private industry. Subsequent volumes will address environmental, health, and safety issues involved in shipping cells and batteries, using batteries to propel electric vehicles, and recycling and disposing of spent batteries. The remainder of this volume is divided into two major sections on safety at the cell and battery levels. The section on Na/S cells describes major component and potential failure modes, design, life testing and failure testing, thermal cycling, and the safety status of Na/S cells. The section on batteries describes battery design, testing, and safety status. Additional EH S information on Na/S batteries is provided in the appendices.

  20. 42 CFR 137.368 - Is the Secretary responsible for oversight and compliance of health and safety codes during...

    Science.gov (United States)

    2010-10-01

    ... compliance of health and safety codes during construction projects being performed by a Self-Governance Tribe... SERVICES TRIBAL SELF-GOVERNANCE Construction Roles of the Secretary in Establishing and Implementing Construction Project Agreements § 137.368 Is the Secretary responsible for oversight and compliance of health...

  1. Development of a computer code for low-and intermediate-level radioactive waste disposal safety assessment

    International Nuclear Information System (INIS)

    Park, J. W.; Kim, C. L.; Lee, E. Y.; Lee, Y. M.; Kang, C. H.; Zhou, W.; Kozak, M. W.

    2002-01-01

    A safety assessment code, called SAGE (Safety Assessment Groundwater Evaluation), has been developed to describe post-closure radionuclide releases and potential radiological doses for low- and intermediate-level radioactive waste (LILW) disposal in an engineered vault facility in Korea. The conceptual model implemented in the code is focused on the release of radionuclide from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. The radionuclide transport equations are solved by spatially discretizing the disposal system into a series of compartments. Mass transfer between compartments is by diffusion/dispersion and advection. In all compartments, radionuclides are decayed either as a single-member chain or as multi-member chains. The biosphere is represented as a set of steady-state, radionuclide-specific pathway dose conversion factors that are multiplied by the appropriate release rate from the far field for each pathway. The code has the capability to treat input parameters either deterministically or probabilistically. Parameter input is achieved through a user-friendly Graphical User Interface. An application is presented, which is compared against safety assessment results from the other computer codes, to benchmark the reliability of system-level conceptual modeling of the code

  2. Auditory localisation of conventional and electric cars : laboratory results and implications for cycling safety.

    NARCIS (Netherlands)

    Stelling-Konczak, A. Hagenzieker, M.P. Commandeur, J.J.F. Agterberg, M.J.H. & Wee, B. van

    2016-01-01

    When driven at low speeds, cars operating in electric mode have been found to be quieter than conventional cars. As a result, the auditory cues which pedestrians and cyclists use to assess the presence, proximity and location oncoming traffic may be reduced, posing a safety hazard. This laboratory

  3. Auditory localisation of conventional and electric cars: laboratory results and implications for cycling safety

    NARCIS (Netherlands)

    Stelling-Konczak, A.; Hagenzieker, M.P.; Commandeur, J.J.F.; Agterberg, M.J.H.; van Wee, B.

    2016-01-01

    When driven at low speeds, cars operating in electric mode have been found to be quieter than conventional cars. As a result, the auditory cues which pedestrians and cyclists use to assess the presence, proximity and location oncoming traffic may be reduced, posing a safety hazard. This laboratory

  4. On safety 1E qualification of electrical equipment for nuclear power stations

    International Nuclear Information System (INIS)

    Segarceanu, D.; Geambasu, C.; Avramescu, M.

    1995-01-01

    Electrical equipment and systems for the emergency shutdown of a nuclear reactor are qualified according to safety class 1E. Methods of qualification meeting the requirements should be used, either individually or in combination include, type-test qualification, qualification by operating experience, qualification by analysis, combined qualification. These methods qualification principles, procedures and documents are discussed. (N.T.). 1 fig

  5. Does employee safety influence customer satisfaction? Evidence from the electric utility industry.

    Science.gov (United States)

    Willis, P Geoffrey; Brown, Karen A; Prussia, Gregory E

    2012-12-01

    Research on workplace safety has not examined implications for business performance outcomes such as customer satisfaction. In a U.S. electric utility company, we surveyed 821 employees in 20 work groups, and also had access to archival safety data and the results of a customer satisfaction survey (n=341). In geographically-based work units where there were more employee injuries (based on archival records), customers were less satisfied with the service they received. Safety climate, mediated by safety citizenship behaviors (SCBs), added to the predictive power of the group-level model, but these two constructs exerted their influence independently from actual injuries. In combination, two safety-related predictor paths (injuries and climate/SCB) explained 53% of the variance in customer satisfaction. Results offer preliminary evidence that workplace safety influences customer satisfaction, suggesting that there are likely spillover effects between the safety environment and the service environment. Additional research will be needed to assess the specific mechanisms that convert employee injuries into palpable results for customers. Better safety climate and reductions in employee injuries have the potential to offer payoffs in terms of what customers experience. Copyright © 2012 National Safety Council and Elsevier Ltd. All rights reserved.

  6. Aspects of the design of the automated system for code generation of electrical items of technological equipment

    Directory of Open Access Journals (Sweden)

    Erokhin V.V.

    2017-09-01

    Full Text Available The article presents the aspects of designing an automated system for generating codes for electrical elements of process equipment using CASE-means. We propose our own technology of iterative development of such systems. The proposed methodology uses the tool to develop the ERwin Data Modeler databases of Computer Associates and the author's tool for the automatic generation of ERwin Class Builder code. The implemented design tool is a superstructure over the ERwin Data Modeler from Computer Associates, which extends its functionality. ERwin Data Modeler works with logical and physical data models and allows you to generate a description of the database and ddl-scripts.

  7. Turbine model for the Laguna Verde nucleo electric central based in the RELAP code

    International Nuclear Information System (INIS)

    Cortes M, F.S.; Ramos P, J.C.; Salazar S, E.; Chavez M, C.

    2003-01-01

    The Nuclear Power stations as Laguna Verde occupy at the present time a place every time but important as non pollutant alternative, economic and trusty to generate electricity. It is for it that the Group of Nuclear Engineering of the Engineering Faculty (GrlNFI) of the National Autonomous University of Mexico (UNAM) it develops investigation projects applied to Nuclear Centrals. One of the projects in process is the development of a Classroom simulator, which it can configures to consent to diverse models of nuclear systems with training purposes in normal operation, or, to consent to specialized nuclear codes for the analysis of transitory events and have a severe accident. This work describes the development, implementation and it proves of a simplified model of the Main turbine to be integrated to the group of models of the Classroom simulator. It is part of the current effort of GrlNFI guided to obtain a representation of all the dynamic models necessary to simulate the Plant Balance of the Laguna Verde Central. It is included the development of the unfolding graphic which represent the modeling of the Main turbine, and of the control interface that allows the user to manipulate in simple way, direct and interactive this device during the training or the analysis. With this work it is wanted to contribute to the training of new technicians and to support the operation personnel of the Centrals. Also, the developed infrastructure is guided to contribute in the design and analysis of new Nuclear Power stations with the contribution of new computational tools for the visualization, simulation and process control. (Author)

  8. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J M; Kodeli, I; Menard, St; Bouchet, J L; Renard, F; Martin, E; Blazy, L; Voros, S; Bochud, F; Laedermann, J P; Beaugelin, K; Makovicka, L; Quiot, A; Vermeersch, F; Roche, H; Perrin, M C; Laye, F; Bardies, M; Struelens, L; Vanhavere, F; Gschwind, R; Fernandez, F; Quesne, B; Fritsch, P; Lamart, St; Crovisier, Ph; Leservot, A; Antoni, R; Huet, Ch; Thiam, Ch; Donadille, L; Monfort, M; Diop, Ch; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

  9. Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 3

    International Nuclear Information System (INIS)

    Bestion, D.; Anglart, H.; Mahaffy, J.; Lucas, D.; Song, C.H.; Scheuerer, M.; Zigh, G.; Andreani, M.; Kasahara, F.; Heitsch, M.; Komen, E.; Moretti, F.; Morii, T.; Muehlbauer, P.; Smith, B.L.; Watanabe, T.

    2014-11-01

    The Writing Group 3 on the extension of CFD to two-phase flow safety problems was formed following recommendations made at the 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in Aix-en-Provence, in May 2002. Extension of CFD codes to two-phase flow is significant potentiality for the improvement of safety investigations, by giving some access to smaller scale flow processes which were not explicitly described by present tools. Using such tools as part of a safety demonstration may bring a better understanding of physical situations, more confidence in the results, and an estimation of safety margins. The increasing computer performance allows a more extensive use of 3D modelling of two-phase Thermal hydraulics with finer nodalization. However, models are not as mature as in single phase flow and a lot of work has still to be done on the physical modelling and numerical schemes in such two-phase CFD tools. The Writing Group listed and classified the NRS problems where extension of CFD to two-phase flow may bring real benefit, and classified different modelling approaches in a first report (Bestion et al., 2006). First ideas were reported about the specification and analysis of needs in terms of validation and verification. It was then suggested to focus further activity on a limited number of NRS issues with a high priority and a reasonable chance to be successful in a reasonable period of time. The WG3-step 2 was decided with the following objectives: - selection of a limited number of NRS issues having a high priority and for which two-phase CFD has a reasonable chance to be successful in a reasonable period of time; - identification of the remaining gaps in the existing approaches using two-phase CFD for each selected NRS issue; - review of the existing data base for validation of two-phase CFD application to the selected NRS problems

  10. A simple in-surge pressure analysis using the SPACE code

    International Nuclear Information System (INIS)

    Youn, Bum Soo; Kim, Yo Han; Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2010-01-01

    Currently, nuclear safety analysis codes used in Korea are developed by all the overseas. These codes are paying huge fee and permission must be obtained for use in the country. In addition, orders for nuclear power plants must ensure the safety analysis code for independent domestic technology. Therefore, Korea Electric Power Research Institute(KEPRI) is developing the domestic nuclear power safety analysis, SPACE(Safety and Performance Analysis Code for nuclear power plants). To determine the computational power of pressurizer model in development SPACE code, it was compared with existing commercial nuclear power safety analysis code, RETRAN

  11. Analysis of SCARABEE BE+3 experiment with ASTEC-Na and comparison with other SFR safety analysis codes

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Ederli, Stefano; Perez-Martin, Sara; Pfrang, Werner; Girault, Nathalie; Cloarec, Laure

    2017-01-01

    The ASTEC-Na code was further developed and assessed in the frame of JASMIN project of the 7th EU Framework Program to extend the original capability of ASTEC, dealing with severe accident analysis in LWR to Sodium-cooled Fast Reactors (SFR). The in-pile BE+3 experiment from the SCARABEE-N program has been simulated with ASTEC-Na for thermal-hydraulic models validation purpose. The adequacy of ASTEC-Na thermal-hydraulic models has been also investigated through the comparison with other safety analysis codes. The analysis of SCARABEE BE+3 test confirms the good performance of ASTEC-Na code in the calculation of single-phase conditions and boiling onset, while larger deviations are encountered in the analysis of the two-phase conditions, mainly regarding the propagation of the boiling front. Furthermore, reasonable agreement was found with other code results. (author)

  12. A criticality safety analysis code using a vectorized Monte Carlo method on the HITAC S-810 supercomputer

    International Nuclear Information System (INIS)

    Morimoto, Y.; Maruyama, H.

    1987-01-01

    A vectorized Monte Carlo criticality safety analysis code has been developed on the vector supercomputer HITAC S-810. In this code, a multi-particle tracking algorithm was adopted for effective utilization of the vector processor. A flight analysis with pseudo-scattering was developed to reduce the computational time needed for flight analysis, which represents the bulk of computational time. This new algorithm realized a speed-up of factor 1.5 over the conventional flight analysis. The code also adopted the multigroup cross section constants library of the Bodarenko type with 190 groups, with 132 groups being for fast and epithermal regions and 58 groups being for the thermal region. Evaluation work showed that this code reproduce the experimental results to an accuracy of about 1 % for the effective neutron multiplication factor. (author)

  13. The Development of Computer Code for Safety Injection Tank (SIT) with Fluidic Device(FD) Blowdown Test

    International Nuclear Information System (INIS)

    Lee, Joo Hee; Kim, Tae Han; Choi, Hae Yun; Lee, Kwang Won; Chung, Chang Kyu

    2007-01-01

    Safety Injection Tanks (SITs) with the Fluidic Device (FD) of APR1400 provides a means of rapid reflooding of the core following a large break Loss Of Coolant Accident (LOCA), and keeping it covered until flow from the Safety Injection Pump (SIP) becomes available. A passive FD can provide two operation stages of a safety water injection into the RCS and allow more effective use of borated water in case of LOCA. Once a large break LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced to a lower flow rate. The conventional computer code 'TURTLE' used to simulate the blowdown of OPR1000 SIT can not be directly applied to simulate a blowdown process of the SIT with FD. A new computer code is needed to be developed for the blowdown test evaluation of the APR1400 SIT with FD. Korea Power Engineering Company (KOPEC) has developed a new computer code to analyze the characteristics of the SIT with FD and validated the code through the comparison of the calculation results with the test results obtained by Ulchin 5 and 6 units pre-operational test and VAlve Performance Evaluation Rig (VAPER) tests performed by The Korea Atomic Energy Research Institute (KAERI)

  14. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V.; Farnaam, M.R.K.

    1977-01-01

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  15. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  16. Reliability evaluation of the power supply of an electrical power net for safety-relevant applications

    International Nuclear Information System (INIS)

    Dominguez-Garcia, Alejandro D.; Kassakian, John G.; Schindall, Joel E.

    2006-01-01

    In this paper, we introduce a methodology for the dependability analysis of new automotive safety-relevant systems. With the introduction of safety-relevant electronic systems in cars, it is necessary to carry out a thorough dependability analysis of those systems to fully understand and quantify the failure mechanisms in order to improve the design. Several system level FMEAs are used to identify the different failure modes of the system and, a Markov model is constructed to quantify their probability of occurrence. A new power net architecture with application to new safety-relevant automotive systems, such as Steer-by-Wire or Brake-by-Wire, is used as a case study. For these safety-relevant loads, loss of electric power supply means loss of control of the vehicle. It is, therefore, necessary and critical to develop a highly dependable power net to ensure power to these loads under all circumstances

  17. Efforts to improve safety and reliability of nuclear power plants in Kyushu Electric Power

    International Nuclear Information System (INIS)

    Yamamoto, Satoshi

    2014-01-01

    After the Fukushima accident, Kyushu Electric Power Co. took emergency safety measures requested by government to ensure power supply, coolant supply pumps and cooling water so as to keep cooling fuels in the reactor and spent fuel storage pool in case of losses of ordinary cooling capability caused by earthquake and tsunami. In order to improve safety and reliability of nuclear power plants, further efforts based on lessons learned from the Fukushima accident had been made to diversify corresponding equipment of safety measures in terms of prevention of core damage, prevention of containment failure, mitigation of radioactive materials release, cooling of spent fuel pit and ensurance of power supply, and to enhance emergency response capability so as to make operational management more complete. Additional safety measures applicable to new regulatory requirements against severe accidents were in progress. This article introduced details of such activities. (T. Tanaka)

  18. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 92--94). Final report

    International Nuclear Information System (INIS)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1994-01-01

    This is the Final Report for a three-year (FY 92--94) study of the Environmental, Safety, and Economic (ESE) aspects of fusion energy systems, emphasizing development of computerized approaches suitable for incorporation as modules in fusion system design codes. First, as is reported in Section 2, the authors now have operating a simplified but complete environment and safety evaluation code, BESAFE. The first tests of BESAFE as a module of the SUPERCODE, a design optimization systems code at LLNL, are reported in Section 3. Secondly, as reported in Section 4, the authors have maintained a strong effort in developing fast calculational schemes for activation inventory evaluation. In addition to these major accomplishments, considerable progress has been made on research on specific topics as follows. A tritium modeling code TRIDYN was developed in collaboration with the TSTA group at LANL and the Fusion Nuclear Technology group at UCLA. A simplified algorithm has been derived to calculate the transient temperature profiles in the blanket during accidents. The scheme solves iteratively a system of non-linear ordinary differential equations describing about 10 regions of the blanket by preserving energy balance. The authors have studied the physics and engineering aspects of divertor modeling for safety applications. Several modifications in the automation and characterization of environmental and safety indices have been made. They have applied this work to the environmental and safety comparisons of stainless steel with alternative structural materials for fusion reactors. A methodology in decision analysis utilizing influence and decision diagrams has been developed to model fusion reactor design problems. Most of the work during this funding period has been reported in 26 publications including theses, journal publications, conference papers, and technical reports, as listed in Section 11

  19. The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

  20. Safety evaluation of liquid radioactive effluents treatment system in a BWR reactor, through the LIQM03 code

    International Nuclear Information System (INIS)

    Zorrilla R, S.H.

    1978-01-01

    In this work we made a safety evaluation of the liquid radioactive effluents system in a plant using a BWR similar to that now installed in Laguna Verde. For that purpose, the computation program ORIGENwas modified, in order to keep up to date and adapt it to the PDP 10 computer, which is operating at the Computation Department of the Nuclear Center of Mexico, the code LIQM03 was the result of this modification. As usual in this work we dealt with problems which were solved opportunely, now we have at our disposal the code LIQM03 which will be in the future a very useful tool for this kind of evaluations. (author)

  1. Calculation of fluid-structure interaction for reactor safety with the Cassiopee code

    International Nuclear Information System (INIS)

    Graveleau, J.L.; Louvet, P.D.

    1979-01-01

    The cassiopee code is an eulerian-lagrangian coupled code for computations where the hydrodynamic is coupled with structural domains. It is completely explicit. The fluid zones may be computed either in lagrangian or in eulerian coordinates; thin shells can be computed wih their flexural behaviour; elastic plastic zones must be calculated in a lagrangian way. This code is under development in Cadarache. Its purpose is to compute the hypothetical core disruptive accident of a LMFBR when lagrangian codes are not sufficient. This paper contains a description of the code and two examples of computations, one of which has been compared with experimental results

  2. Realistic integration of sorption processes in transport codes for long-term safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Noseck, Ulrich; Fluegge, Judith; Britz, Susan; Schneider, Anke [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Brendler, Vinzenz; Stockmann, Madlen; Schikora, Johannes [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Lampe, Michael [Frankfurt Univ. (Germany). Goethe Center for Scientific Computing

    2012-09-15

    One important aspect in long-term safety assessment is related to radionuclide transport in geologic formations. In order to assess its consequences over assessment periods of one million years numerical models describing flow and transport are applied. Sorption on mineral surfaces is the most relevant process retarding radionuclide transport. On the one hand an increased transport time might cause a decrease in radionuclide concentration by radioactive decay. On the other hand it might increase concentrations of dose-relevant daughter nuclides in decay chains. In order to treat the radionuclide sorption processes in natural systems close to reality the so-called smart K{sub d}-concept is implemented into the transport program r{sup 3}t, which is applied to large model areas and very long time scales in long-term safety assessment. In the first stage this approach is developed for a typical sedimentary system covering rock salt and clay formations in Northern Germany. The smart K{sub d}-values are based on mechanistic surface complexation models (SCM), varying in time and space and de-pending on the actual geochemical conditions, which might change in the future e. g. due to the impact of climate changes. The concept developed and introduced here is based on a feasible treatment of the most relevant geochemical parameters in the transport code as well as on a matrix of smart K{sub d}-values calculated in dependence on these parameters. The implementation of the concept comprises the selection of relevant elements and minerals to be considered, an experimental program to fill data gaps of the thermody-namic sorption database, an uncertainty and sensitivity analysis to identify the most important environmental parameters influencing sorption of long-term relevant radionu-clides, the creation of a matrix with K{sub d}-values dependent on the selected environmental parameters, and the development and realisation of the conceptual model for treatment of temporal and

  3. The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-01

    The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

  4. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

    Science.gov (United States)

    Mosunova, N. A.

    2018-05-01

    The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

  5. Regulatory inspection practices for industrial safety (electrical, mechanical, material handling and conventional aspects)

    International Nuclear Information System (INIS)

    Agarwal, K.

    2017-01-01

    Regulatory Inspection (RI) of BARC facilities and projects are carried out under the guidance of BARC Safety Council (BSC) Secretariat. Basically facilities and projects have been divided into two board categories viz. radiological facilities and non-radiological facilities. The Rls of radiological facilities should be carried out under OPSRC and of non-radiological facilities under CFSRC. Periodicity of inspection shall be at least once in a year. The RI of projects is carried out under concerned DSRC. RI practices with industrial safety which includes electrical, mechanical, material handling and conventional aspect for these facilities starts with check lists. The inspection areas are prepared in the form of checklists which includes availability of approved documents, compliance status of previous RIT and various safety committee's recommendations, radiological status of facilities, prompt reporting of safety related unusual occurrences, major incident, site visit for verification of actual status of system/plant. The practices for inspection in the area of electrical safety shall include checking of maintenance procedure for all critical class IV system equipment's such as HT panel, LT panel, transformer and motors. Load testing of Class III system such as D.G. set etc. shall be carried out as technical specification surveillance schedule. Status of aviation lights, number of qualified staff, availability of qualified staff etc. shall be form of inspection

  6. Overview of the U.S. DOE Hydrogen Safety, Codes and Standards Program. Part 4: Hydrogen Sensors; Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Buttner, William J.; Rivkin, Carl; Burgess, Robert; Brosha, Eric; Mukundan, Rangachary; James, C. Will; Keller, Jay

    2016-12-01

    Hydrogen sensors are recognized as a critical element in the safety design for any hydrogen system. In this role, sensors can perform several important functions including indication of unintended hydrogen releases, activation of mitigation strategies to preclude the development of dangerous situations, activation of alarm systems and communication to first responders, and to initiate system shutdown. The functionality of hydrogen sensors in this capacity is decoupled from the system being monitored, thereby providing an independent safety component that is not affected by the system itself. The importance of hydrogen sensors has been recognized by DOE and by the Fuel Cell Technologies Office's Safety and Codes Standards (SCS) program in particular, which has for several years supported hydrogen safety sensor research and development. The SCS hydrogen sensor programs are currently led by the National Renewable Energy Laboratory, Los Alamos National Laboratory, and Lawrence Livermore National Laboratory. The current SCS sensor program encompasses the full range of issues related to safety sensors, including development of advance sensor platforms with exemplary performance, development of sensor-related code and standards, outreach to stakeholders on the role sensors play in facilitating deployment, technology evaluation, and support on the proper selection and use of sensors.

  7. Nuclear Electric flask design and safety case development during the last ten years

    International Nuclear Information System (INIS)

    Dougall, I.; Jones, D.K.

    1994-01-01

    The scope of Nuclear Electric's (NE) requirements for irradiated fuel transport and the relevant safety and regulatory compliance standards are outlined. To illustrate NE's approach to demonstrating package design compliance the basis of the safety case for unbottled Magnox fuel transport in the Mk M2 Magnox flask is described. The considerations which lay behind the development of the Mk A2 AGR Flask are indicated and a description is given of the main design features of this flask. Finally there is a summary of progress in obtaining Type B(M) approvals, based on the 1985 IAEA Regulations, for the requisite range of flask contents. (author)

  8. Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

  9. Safety Design for Smart Electric Vehicle Charging with Current and Multiplexing Control

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ching-Yen; Youn, Edward; Chynoweth, Joshua; Qiu, Charlie; Chu, Chi-Cheng; Gadh, Rajit

    2013-10-21

    As Electric Vehicles (EVs) increase, charging infrastructure becomes more important. When during the day there is a power shortage, the charging infrastructure should have the options to either shut off the power to the charging stations or to lower the power to the EVs in order to satisfy the needs of the grid. This paper proposes a design for a smart charging infrastructure capable of providing power to several EVs from one circuit by multiplexing power and providing charge control and safety systems to prevent electric shock. The safety design is implemented in different levels that include both the server and the smart charging stations. With this smart charging infrastructure, the shortage of energy in a local grid could be solved by our EV charging management system.

  10. Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

    2012-03-15

    JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

  11. Simulation codes of chemical separation process of spent fuel reprocessing. Tool for process development and safety research

    International Nuclear Information System (INIS)

    Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji

    2005-01-01

    This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount. (author)

  12. SACRD: a data base for fast reactor safety computer codes, contents and glossary of Version 1 of the system

    International Nuclear Information System (INIS)

    Greene, N.M.; Forsberg, V.M.; Raiford, G.B.; Arwood, J.W.; Flanagan, G.F.

    1979-01-01

    SACRD is a data base of material properties and other handbook data needed in computer codes used for fast reactor safety studies. This document lists the contents of Version 1 and also serves as a glossary of terminology used in the data base. Data are available in the thermodynamics, heat transfer, fluid mechanics, structural mechanics, aerosol transport, meteorology, neutronics and dosimetry areas. Tabular, graphical and parameterized data are provided in many cases

  13. Compliance Behavior Analysis of the Ship Crew to the International Safety Management (Ism) Code in Indonesia???

    OpenAIRE

    Desi Albert Mamahit; Heny K Daryanto; Ujang Sumarwan; Eva Zhoriva Yusuf

    2013-01-01

    The purpose of this code is to provide international standards for the management and safe operation of ships and pollution prevention Furthermore, this study has the objective to identify the role of the ISM Code on maritime activities in Indonesia, knowing the perceptions and attitudes regarding the conduct of the crew boat ISM Code. Location research is conducted on the crew that was in the Port of Tanjung Priok in Jakarta. Data collection and processing is done for 3 months. The study was...

  14. TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1982-01-01

    The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

  15. Modification of the SAS4A Safety Analysis Code for Integration with the ADAPT Discrete Dynamic Event Tree Framework.

    Energy Technology Data Exchange (ETDEWEB)

    Jankovsky, Zachary Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    It is difficult to assess the consequences of a transient in a sodium-cooled fast reactor (SFR) using traditional probabilistic risk assessment (PRA) methods, as numerous safety-related sys- tems have passive characteristics. Often there is significant dependence on the value of con- tinuous stochastic parameters rather than binary success/failure determinations. One form of dynamic PRA uses a system simulator to represent the progression of a transient, tracking events through time in a discrete dynamic event tree (DDET). In order to function in a DDET environment, a simulator must have characteristics that make it amenable to changing physical parameters midway through the analysis. The SAS4A SFR system analysis code did not have these characteristics as received. This report describes the code modifications made to allow dynamic operation as well as the linking to a Sandia DDET driver code. A test case is briefly described to demonstrate the utility of the changes.

  16. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  17. Evaluation of Electrical Characteristics of Protective Equipment - a Prerequisite for Ensuring Safety and Health of Workers at Work

    Science.gov (United States)

    Buică, G.; Beiu, C.; Antonov, A.; Dobra, R.; Păsculescu, D.

    2017-06-01

    The protecting electrical equipment in use are subject to various factors generated by the use, maintenance, storage and working environment, which may change the characteristics of protection against electric shock. The study presents the results of research on the behaviour over time of protective characteristics of insulating covers of material of work equipment in use, in order to determine the type and periodicity of safety tests. There were tested and evaluated safety equipment with plastic and insulating rubber covers used in operations of verifying functionality, safety and maintenance of machinery used in manufacturing industries and specific services from electric, energy and food sector.

  18. Impact Safety Control Strategy for the Battery System of an Example Electric Bus

    Directory of Open Access Journals (Sweden)

    Zhen-po Wang

    2015-01-01

    Full Text Available This paper proposes a side impact safety control strategy for the battery system, aiming at defusing the hazards of unacceptable behaviors of the battery system such as high-voltage hazards. Based on some collision identification metrics, a side impact discrimination algorithm and a side impact severity algorithm are developed for electric buses. Based on the study on the time to break for power battery, the side impact discrimination algorithm response time is about 20 ms posing a great challenge to the side impact discrimination algorithm. At the same time, the reliability of the impact safety control strategy developed in this paper is evaluated for other plausible side impact signals generated by finite element analysis. The results verify that the impact safety control strategy exhibits robust performance and is able to trigger a breaking signal for power battery system promptly and accurately.

  19. Validation and verification of the ORNL Monte Carlo codes for nuclear safety analysis

    International Nuclear Information System (INIS)

    Emmett, M.B.

    1993-01-01

    The process of ensuring the quality of computer codes can be very time consuming and expensive. The Oak Ridge National Laboratory (ORNL) Monte Carlo codes all predate the existence of quality assurance (QA) standards and configuration control. The number of person-years and the amount of money spent on code development make it impossible to adhere strictly to all the current requirements. At ORNL, the Nuclear Engineering Applications Section of the Computing Applications Division is responsible for the development, maintenance, and application of the Monte Carlo codes MORSE and KENO. The KENO code is used for doing criticality analyses; the MORSE code, which has two official versions, CGA and SGC, is used for radiation transport analyses. Because KENO and MORSE were very thoroughly checked out over the many years of extensive use both in the United States and in the international community, the existing codes were open-quotes baselined.close quotes This means that the versions existing at the time the original configuration plan is written are considered to be validated and verified code systems based on the established experience with them

  20. Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C.

    2001-01-01

    The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical model in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future

  1. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    Energy Technology Data Exchange (ETDEWEB)

    Staedtke, H. [Commission of the European Union, Ispra (Italy)

    1997-07-01

    The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.

  2. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    International Nuclear Information System (INIS)

    Staedtke, H.

    1997-01-01

    The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved

  3. Self-assessment of application of the Code of Conduct on the safety of research reactors - Mexico

    International Nuclear Information System (INIS)

    Mamani-Alegria, Y.R.; Salgado-Gonzalez, J.R.; Miranda-Aldaco, J.

    2009-01-01

    In Mexico, the nuclear regulatory body is the National Commission on Nuclear Safety and Safeguards (CNSNS), and there is one research reactor, a TRIGA MARK III, operated by the National Institute for Nuclear Research (ININ). The main aspects of the Self-assessment of application of The Code of Conduct on the Safety of Research Reactor are given for the case of the TRIGA reactor. Furthermore, in this paper we give a brief description of the legal framework of the licensing process, for nuclear activities in a research reactor, there are also highlights of the major reactor features, the uses of the reactor for isotope production, the management and verification of safety, the radiation protection management program, the emergency planning and the training and qualification of the operation personnel. (author)

  4. Current state of the auto-evaluation process of the behaviour code in the safety of research reactors in Mexico

    International Nuclear Information System (INIS)

    Mamani A, Y. R.; Salgado G, J. R.

    2011-11-01

    In Mexico, the regulator organism in nuclear matter is the National Commission of Nuclear Safety and Safeguards, and a nuclear research reactor exists, the TRIGA Mark III, operated by the National Institute of Nuclear Research. In this work the main aspects of the current state and the future challenges are presented with relationship to the installation of the auto-evaluation process of the behaviour code in the safety of research reactors for the TRIGA reactor case. Additionally, the legal mark of the licensing process for the nuclear activities in a research reactor is described in a brief way, and the main characteristics of the reactor, the uses for the isotopes production, the administration and the verification of the safety, the administration program of the radiological protection, the emergency plan and the operation personnel qualification are emphasized. (Author)

  5. Longitudinal safety evaluation of electric vehicles with the partial wireless charging lane on freeways.

    Science.gov (United States)

    Li, Ye; Wang, Wei; Xing, Lu; Fan, Qi; Wang, Hao

    2018-02-01

    As an environment friendly transportation mode, the electric vehicle (EV) has drawn an increasing amount of attention from governments, vehicle manufactories and researchers recently. One of the biggest issue impeding EV's popularization associates with the charging process. The wireless charging lane (WCL) has been proposed as a convenient charging facility for EVs. Due to the high costs, the application of WCL on the entire freeways is impractical in the near future, while the partial WCL (PWCL) may be a feasible solution. This study aims to evaluate longitudinal safety of EVs with PWCL on freeways based on simulations. The simulation experiments are firstly designed, including deployment of PWCL on freeways and distribution of state of charge (SOC) of EVs. Then, a vehicle behavior model for EVs is proposed based on the intelligent driver model (IDM). Two surrogate safety measures, derived from time-to-collision (TTC), are utilized as indicators for safety evaluations. Sensitivity analysis is also conducted for related factors. Results show that the distribution of EVs' SOC significantly affect longitudinal safety when the PWCL is utilized. The low SOC in traffic consisting of EVs has the negative effect on longitudinal safety. The randomness and incompliance of EV drivers worsens the safety performance. The sensitivity analysis indicates that the larger maximum deceleration rate results in the higher longitudinal crash risks of EVs, while the length of PWCL has no monotonous effect. Different TTC thresholds also show no impact on results. A case study shows the consistent results. Based on the findings, several suggestions are discussed for EVs' safety improvement. Results of this study provide useful information for freeway safety when EVs are applied in the future. Copyright © 2017 Elsevier Ltd. All rights reserved.

  6. RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

    International Nuclear Information System (INIS)

    Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

    1989-03-01

    A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

  7. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  8. Modeling of a confinement bypass accident with CONSEN, a fast-running code for safety analyses in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it [Sapienza University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Roma (Italy); Giannetti, Fabio [Sapienza University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Roma (Italy); Porfiri, Maria Teresa [ENEA FUS C.R. Frascati, Via Enrico Fermi, 45, 00044 Frascati, Roma (Italy)

    2013-12-15

    Highlights: • The CONSEN code for thermal-hydraulic transients in fusion plants is introduced. • A magnet induced confinement bypass accident in ITER has been simulated. • A comparison with previous MELCOR results for the accident is presented. -- Abstract: The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.

  9. An Agent-Based Model for Zip-Code Level Diffusion of Electric Vehicles and Electricity Consumption in New York City

    Directory of Open Access Journals (Sweden)

    Azadeh Ahkamiraad

    2018-03-01

    Full Text Available Current power grids in many countries are not fully prepared for high electric vehicle (EV penetration, and there is evidence that the construction of additional grid capacity is constantly outpaced by EV diffusion. If this situation continues, then it will compromise grid reliability and cause problems such as system overload, voltage and frequency fluctuations, and power losses. This is especially true for densely populated areas where the grid capacity is already strained with existing old infrastructure. The objective of this research is to identify the zip-code level electricity consumption that is associated with large-scale EV adoption in New York City, one of the most densely populated areas in the United States (U.S.. We fuse the Fisher and Pry diffusion model and Rogers model within the agent-based simulation to forecast zip-code level EV diffusion and the required energy capacity to satisfy the charging demand. The research outcomes will assist policy makers and grid operators in making better planning decisions on the locations and timing of investments during the transition to smarter grids and greener transportation.

  10. Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

    International Nuclear Information System (INIS)

    Brus, N.A.; Ioussoupov, O.E.

    2000-01-01

    This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

  11. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

    International Nuclear Information System (INIS)

    Podlazov, L. N.

    1998-01-01

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

  12. Safety aspects of electric energy production and supplies in conditions of the Slovenske elektrarne, a.s

    International Nuclear Information System (INIS)

    Sip, M.; Danilak, M. et al.

    2005-01-01

    In this presentation author deals with safety aspects of electric energy production and supplies in conditions of the Slovenske elektrarne, a.s. Some terrorist attack and accidents are presented. Four video-sequences are included

  13. Revision of AESJ standard 'the code of implemnetation of periodic safety review of nuclear power plants'

    International Nuclear Information System (INIS)

    Hirano, Masashi; Narumiya, Yoshiyuki

    2010-01-01

    The Periodic Safety Review (PSR) was launched in June 1992, when the Agency for Natural Resources and Energy issued a notification that required licensees to conduct comprehensive review on the safety of each existing nuclear power plant (NPP) once approximately every ten years based on the latest technical findings for the purpose of improving the safety of the NPP. In 2006, the Standard Committee of the Atomic Energy Society of Japan established the first version of 'The Standard of Implementation for Periodic Safety Review of Nuclear Power Plants: 2006'. Taking into account developments in safety regulation of PSR after the issuance of the first version, the Standard Committee has revised the Standard. This paper summarizes background on PSR, such developments are major contents of the Standard as well as the focal points of the revision. (author)

  14. Objective Oriented Design of Architecture for TH System Safety Analysis Code and Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong

    2008-03-15

    In this work, objective oriented design of generic system analysis code has been tried based on the previous works in KAERI for two phase three field Pilot code. It has been performed to implement of input and output design, TH solver, component model, special TH models, heat structure solver, general table, trip and control, and on-line graphics. All essential features for system analysis has been designed and implemented in the final product SYSTF code. The computer language C was used for implementation in the Visual studio 2008 IDE (Integrated Development Environment) since it has easier and lighter than C++ feature. The code has simple and essential features of models and correlation, special component, special TH model and heat structure model. However the input features is able to simulate the various scenarios, such as steady state, non LOCA transient and LOCA accident. The structure validity has been tested through the various verification tests and it has been shown that the developed code can treat the non LOCA and LOCA simulation. However more detailed design and implementation of models are required to get the physical validity of SYSTF code simulation.

  15. Objective Oriented Design of Architecture for TH System Safety Analysis Code and Verification

    International Nuclear Information System (INIS)

    Chung, Bub Dong

    2008-03-01

    In this work, objective oriented design of generic system analysis code has been tried based on the previous works in KAERI for two phase three field Pilot code. It has been performed to implement of input and output design, TH solver, component model, special TH models, heat structure solver, general table, trip and control, and on-line graphics. All essential features for system analysis has been designed and implemented in the final product SYSTF code. The computer language C was used for implementation in the Visual studio 2008 IDE (Integrated Development Environment) since it has easier and lighter than C++ feature. The code has simple and essential features of models and correlation, special component, special TH model and heat structure model. However the input features is able to simulate the various scenarios, such as steady state, non LOCA transient and LOCA accident. The structure validity has been tested through the various verification tests and it has been shown that the developed code can treat the non LOCA and LOCA simulation. However more detailed design and implementation of models are required to get the physical validity of SYSTF code simulation

  16. MONK 6.3 code new version for microcomputers and its impact on the safety-criticality analysis

    International Nuclear Information System (INIS)

    Albornoz, A.F.; Jatuff, F.E.

    1994-01-01

    The Monte Carlo method is utilized in neutronics in different fields, such as critical experiment analysis, shielding and criticality-safety analysis. This broad use of is due mainly to its great ability in geometrical representation. In this work it is presented in detail the modifications made on version 6.3 of english code MONK, in order to install it in a system based in microcomputers. The description of some enhancements made to the code are also described. The modifications basically comprised to rewrite code with non-standard use of FORTRAN, and to modify the multidimensional arrays arithmetic in order to save another non-standard use of language. The random number generator was also tested, and it was recognized that it presented a layer structure like all congruent-linear generators. For this reason, it was changed by another with better randomness characteristics. The economic motivations that promote down sizing work, the difficulties found in reorganizing group work, and the impact on the quality of safety analysis of facilities with fissile material are also discussed. (author)

  17. The IAEA's activities in safeguarding nuclear materials and in developing internationally acceptable safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    Rometsch, Rudolf; Specter, Herschel

    1977-01-01

    Promoting the peaceful use of nuclear energy and aiming at the international sharing of its benefits are objectives that guide the activities of the Agency. But this promotional work is carried out on condition that security and safety are provided for. All Agency assistance involving nuclear facilities will be subjected to standards of safety or other standards, which are proposed by a State the Agency finds essentially equivalent. Safeguards are always applied on the basis of agreement. States party to NPT are obligated to negotiate and conclude with the Agency agreements which cover all their peaceful nuclear activities. Safeguards agreements concluded outside NPT are applied to specific supplies of facilities, equipment and material. To assist countries in laying down their nuclear safety regulations the Agency's program for the developing of codesof practice and safety guides for nuclear power plants draws up guidelines for governmental organizations, siting, design, operation and quality assurance. Codes are the fundamental documents laying down the objectives of each field of nuclear safety

  18. Analysis of the impact of correlated benchmark experiments on the validation of codes for criticality safety analysis

    International Nuclear Information System (INIS)

    Bock, M.; Stuke, M.; Behler, M.

    2013-01-01

    The validation of a code for criticality safety analysis requires the recalculation of benchmark experiments. The selected benchmark experiments are chosen such that they have properties similar to the application case that has to be assessed. A common source of benchmark experiments is the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) compiled by the 'International Criticality Safety Benchmark Evaluation Project' (ICSBEP). In order to take full advantage of the information provided by the individual benchmark descriptions for the application case, the recommended procedure is to perform an uncertainty analysis. The latter is based on the uncertainties of experimental results included in most of the benchmark descriptions. They can be performed by means of the Monte Carlo sampling technique. The consideration of uncertainties is also being introduced in the supplementary sheet of DIN 25478 'Application of computer codes in the assessment of criticality safety'. However, for a correct treatment of uncertainties taking into account the individual uncertainties of the benchmark experiments is insufficient. In addition, correlations between benchmark experiments have to be handled correctly. For example, these correlations can arise due to different cases of a benchmark experiment sharing the same components like fuel pins or fissile solutions. Thus, manufacturing tolerances of these components (e.g. diameter of the fuel pellets) have to be considered in a consistent manner in all cases of the benchmark experiment. At the 2012 meeting of the Expert Group on 'Uncertainty Analysis for Criticality Safety Assessment' (UACSA) of the OECD/NEA a benchmark proposal was outlined that aimed for the determination of the impact on benchmark correlations on the estimation of the computational bias of the neutron multiplication factor (k eff ). The analysis presented here is based on this proposal. (orig.)

  19. Research on Heat Dissipation of Electric Vehicle Based on Safety Architecture Optimization

    Science.gov (United States)

    Zhou, Chao; Guo, Yajuan; Huang, Wei; Jiang, Haitao; Wu, Liwei

    2017-10-01

    In order to solve the problem of excessive temperature in the discharge process of lithium-ion battery and the temperature difference between batteries, a heat dissipation of electric vehicle based on safety architecture optimization is designed. The simulation is used to optimize the temperature field of the heat dissipation of the battery. A reasonable heat dissipation control scheme is formulated to achieve heat dissipation requirements. The results show that the ideal working temperature range of the lithium ion battery is 20?∼45?, and the temperature difference between the batteries should be controlled within 5?. A cooling fan is arranged at the original air outlet of the battery model, and the two cooling fans work in turn to realize the reciprocating flow. The temperature difference is controlled within 5? to ensure the good temperature uniformity between the batteries of the electric vehicle. Based on the above finding, it is concluded that the heat dissipation design for electric vehicle batteries is safe and effective, which is the most effective methods to ensure battery life and vehicle safety.

  20. On the electrical safety of dielectric elastomer actuators in proximity to the human body

    Science.gov (United States)

    Pourazadi, S.; Shagerdmootaab, A.; Chan, H.; Moallem, M.; Menon, C.

    2017-11-01

    Novel devices based on the use of dielectric elastomer actuators (DEA) have been proposed for a large variety of different applications. In many of these applications, DEAs are envisioned to be in direct or close proximity to the human body. Since DEAs usually require high voltage for their actuation, the safety of individuals operating or using these devices should be ensured. In this paper, safety standards based on safe limits for electrical discharge are investigated. Flat and cylindrical DEA configurations, which are generally considered as the building blocks for the design of DEA-based systems, are investigated in detail. Relevant elements and factors that affect the electrical discharge of DEA devices are analyzed and guidelines to design DEA-based devices that are not of harm for humans are provided. The performed analyses are experimentally validated using flat DEA samples. The safety requirements that should be considered when wrapping DEAs around the body (specifically the legs) are also briefly investigated to provide a practical example of interest for the biomedical community.

  1. Safety studies: Review of loss of plasma control transients in ITER with AINA 3.0 code

    International Nuclear Information System (INIS)

    Rivas, J.C.; Dies, J.

    2013-01-01

    Highlights: ► We have examined the methodology employed in the analysis of the “Loss of plasma transients in ITER” safety reference events. ► We have developed a new methodology based on the study of the plasma operating window. ► We have concluded that the combined effect of different perturbations should be studied also to determine the most severe transients. -- Abstract: The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. In the past, conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer showed that a hypothetical scenario of first wall coolant tubes melting and subsequent entering of water in the vacuum vessel could not be totally excluded. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering Laboratory (FEEL) in Barcelona. It uses a 0D–1D architecture, similar to that used for previous analyses of ITER loss of plasma control events. The results of this study show the simultaneous effect of two perturbations (overfuelling and overheating) over a plasma transient, and compare it with the isolated effects of each perturbation. It is shown that the combined effect can be more severe, and a method is outlined to locate the most dangerous transients over a nT diagram

  2. Roadmap to an Engineering-Scale Nuclear Fuel Performance and Safety Code

    International Nuclear Information System (INIS)

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

    2009-01-01

    -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.

  3. [QR-Code based patient tracking: a cost-effective option to improve patient safety].

    Science.gov (United States)

    Fischer, M; Rybitskiy, D; Strauß, G; Dietz, A; Dressler, C R

    2013-03-01

    Hospitals are implementing a risk management system to avoid patient or surgery mix-ups. The trend is to use preoperative checklists. This work deals specifically with a type of patient identification, which is realized by storing patient data on a patient-fixed medium. In 127 ENT surgeries data relevant for patient identification were encrypted in a 2D-QR-Code. The code, as a separate document coming with the patient chart or as a patient wristband, has been decrypted in the OR and the patient data were presented visible for all persons. The decoding time, the compliance of the patient data, as well as the duration of the patient identification was compared with the traditional patient identification by inspection of the patient chart. A total of 125 QR codes were read. The time for the decrypting of QR-Code was 5.6 s, the time for the screen view for patient identification was 7.9 s, and for a comparison group of 75 operations traditional patient identification was 27.3 s. Overall, there were 6 relevant information errors in the two parts of the experiment. This represents a ratio of 0.6% for 8 relevant classes per each encrypted QR code. This work allows a cost effective way to technically support patient identification based on electronic patient data. It was shown that the use in the clinical routine is possible. The disadvantage is a potential misinformation from incorrect or missing information in the HIS, or due to changes of the data after the code was created. The QR-code-based patient tracking is seen as a useful complement to the already widely used identification wristband. © Georg Thieme Verlag KG Stuttgart · New York.

  4. Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, including containment. Summary report of a technical meeting

    International Nuclear Information System (INIS)

    2003-11-01

    Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and

  5. Traffic sounds and cycling safety : the use of electronic devices by cyclists and the quietness of hybrid and electric cars.

    NARCIS (Netherlands)

    Stelling-Konczak, A. Hagenzieker, M.P. & Wee, B. van

    2015-01-01

    The growing popularity of electric devices and the increasing number of hybrid and electric cars have recently raised concerns about the use of auditory signals by vulnerable road users. This paper consolidates current knowledge about the two trends in relation to cycling safety. Both a literature

  6. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C

    2006-10-15

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions.

  7. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    International Nuclear Information System (INIS)

    Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C.

    2006-10-01

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions

  8. Safety related investigations of the VVER-1000 reactor type by the coupled code system TRACE/PARCS

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Lischke, Wolfgang; Sanchez Espinoza, Victor Hugo

    2007-01-01

    This study was performed at the Institute of Reactor Safety at the Research Center Karlsruhe. It is embedded in the ongoing investigations of the international code application and maintenance program (CAMP) for qualification and validation of system codes like TRACE [1] and PARCS [2]. The predestinated reactor type for the validation of these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2 [3] includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The posttest-investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement to the measured data. The coolant mixing pattern especially in the downcomer has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provides good results compared to reference values and the ones of other participants of the benchmark. It can be pointed out that the developed three-dimensional nodalisation of the reactor pressure vessel (RPV) is appropriate for the description of transients where the thermal-hydraulics and the neutronics are strongly linked. (author)

  9. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    International Nuclear Information System (INIS)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications

  10. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.; Boss, C. R.

    2006-01-01

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  11. The Nuremberg Code subverts human health and safety by requiring animal modeling

    Directory of Open Access Journals (Sweden)

    Greek Ray

    2012-07-01

    Full Text Available Abstract Background The requirement that animals be used in research and testing in order to protect humans was formalized in the Nuremberg Code and subsequent national and international laws, codes, and declarations. Discussion We review the history of these requirements and contrast what was known via science about animal models then with what is known now. We further analyze the predictive value of animal models when used as test subjects for human response to drugs and disease. We explore the use of animals for models in toxicity testing as an example of the problem with using animal models. Summary We conclude that the requirements for animal testing found in the Nuremberg Code were based on scientifically outdated principles, compromised by people with a vested interest in animal experimentation, serve no useful function, increase the cost of drug development, and prevent otherwise safe and efficacious drugs and therapies from being implemented.

  12. Insights provided by Probabilistic Safety Assessment Relating to the Loss of Electrical Sources

    International Nuclear Information System (INIS)

    Lanore, Jeanne-Marie

    2015-01-01

    The loss of electrical sources is generally an important contributor to the risk related to nuclear plants. In particular the external hazards initiating events lead generally to a loss of electrical sources. This importance was underscored by the Fukushima accident. A strength of PSA is to provide insights not only into the causes of the event but also into the potential consequences (core damage prevention, large release prevention, and mitigation) with the corresponding risk impact. PSA could provide a measure of Defence-in-Depth in case of loss of a safety function. The task intends to illustrate the PSA capabilities with outstanding practical examples. The task will rely on a survey of existing PSAs. It will provide a complementary view for ROBELSYS task. The content and status of the task are summarized in 2 slides

  13. Interim staff position on environmental qualification of safety-related electrical equipment: including staff responses to public comments. Regulatory report

    International Nuclear Information System (INIS)

    Szukiewicz, A.J.

    1981-07-01

    This document provides the NRC staff positions regarding selected areas of environmental qualification of safety-related electrical equipment, in the resolution of Unresolved Safety Issue A-24, 'Qualification of Class IE Safety-Related Equipment.' The positions herein are applicable to plants that are or will be in the construction permit (CP) or operating license (OL) review process and that are required to satisfy the requirements set forth in either the 1971 or the 1974 version of IEEE-323 standard

  14. Direct and indirect health and safety impacts of electrical generation options

    International Nuclear Information System (INIS)

    Habegger, L.J.; Gasper, J.R.; Brown, C.D.

    1982-01-01

    This report is an analysis of the health and safety risks of seven electrical generation systems, all of which have the potential for commercial availability after the year 2000. The systems are compared on the basis of expected public and occupational deaths and lost workdays associated with average unit generation of 1000 MW(e) per year. The risks and associated uncertainties are estimated for all phases of the energy production cycle, including fuel extraction and processing, on-site construction and system operation and maintenance. Also included are the risks of direct and indirect component manufacture, materials production and energy inputs, all of which are major contributors to the risks of the more capital-intensive solar technologies. The potential significance of major health and safety issues that remain largely unquantifiable are also considered. (author)

  15. Safety Assessment for Electrical Motor Drive System Based on SOM Neural Network

    Directory of Open Access Journals (Sweden)

    Linghui Meng

    2016-01-01

    Full Text Available With the development of the urban rail train, safety and reliability have become more and more important. In this paper, the fault degree and health degree of the system are put forward based on the analysis of electric motor drive system’s control principle. With the self-organizing neural network’s advantage of competitive learning and unsupervised clustering, the system’s health clustering and safety identification are worked out. With the switch devices’ faults data obtained from the dSPACE simulation platform, the health assessment algorithm is verified. And the results show that the algorithm can achieve the system’s fault diagnosis and health assessment, which has a point in the health assessment and maintenance for the train.

  16. Comparison of safety assessment codes for near-surface disposal of LILW with the compartment model: SAGE and VR-KHNP

    International Nuclear Information System (INIS)

    Kim, H. J.; Park, J. W.; Park, J. B.; Kim, C. L.

    2004-01-01

    Safety Assessment Groundwater Evaluation and Virtual Repository for KHNP codes for performance assessment of the LILW disposal repository were developed by joint collaboration between KHNP and foreign consulting organizations. In both code, the disposal facility consists of a series of compartments that represent the waste form, the engineered barrier system, the unsaturated and saturated zone

  17. Application of the integrated safety assessment methodology to the protection of electric systems

    International Nuclear Information System (INIS)

    Hortal, Javier; Izquierdo, Jose M.

    1996-01-01

    The generalization of classical techniques for risk assessment incorporating dynamic effects is the main objective of the Integrated Safety Assessment Methodology, as practical implementation of Protection Theory. Transient stability, contingency analysis and protection setpoint verification in electric power systems are particularly appropriate domains of application, since the coupling of reliability and dynamic analysis in the protection assessment process is being increasingly demanded. Suitable techniques for dynamic simulation of sequences of switching events in power systems are derived from the use of quasi-linear equation solution algorithms. The application of the methodology, step by step, is illustrated in a simple but representative example

  18. Validation studies of thermal-hydraulic code for safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Haapalehto, T.

    1995-01-01

    The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.)

  19. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  20. Development and validation of safety climate scales for mobile remote workers using utility/electrical workers as exemplar.

    Science.gov (United States)

    Huang, Yueng-Hsiang; Zohar, Dov; Robertson, Michelle M; Garabet, Angela; Murphy, Lauren A; Lee, Jin

    2013-10-01

    The objective of this study was to develop and test the reliability and validity of a new scale designed for measuring safety climate among mobile remote workers, using utility/electrical workers as exemplar. The new scale employs perceived safety priority as the metric of safety climate and a multi-level framework, separating the measurement of organization- and group-level safety climate items into two sub-scales. The question of the emergence of shared perceptions among remote workers was also examined. For the initial survey development, several items were adopted from a generic safety climate scale and new industry-specific items were generated based on an extensive literature review, expert judgment, 15-day field observations, and 38 in-depth individual interviews with subject matter experts (i.e., utility industry electrical workers, trainers and supervisors of electrical workers). The items were revised after 45 cognitive interviews and a pre-test with 139 additional utility/electrical workers. The revised scale was subsequently implemented with a total of 2421 workers at two large US electric utility companies (1560 participants for the pilot company and 861 for the second company). Both exploratory (EFA) and confirmatory factor analyses (CFA) were adopted to finalize the items and to ensure construct validity. Reliability of the scale was tested based on Cronbach's α. Homogeneity tests examined whether utility/electrical workers' safety climate perceptions were shared within the same supervisor group. This was followed by an analysis of the criterion-related validity, which linked the safety climate scores to self-reports of safety behavior and injury outcomes (i.e., recordable incidents, missing days due to work-related injuries, vehicle accidents, and near misses). Six dimensions (Safety pro-activity, General training, Trucks and equipment, Field orientation, Financial Investment, and Schedule flexibility) with 29 items were extracted from the EFA to

  1. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  2. Neutron kinetics developments of the SIMMER-III safety code for APS application

    International Nuclear Information System (INIS)

    Rineiski, A.; Kiefhaber, E.; Merk, B.; Maschek, W.

    2000-01-01

    Recent developments extending the capabilities of the SIMMER-III code for dealing with transients and accidents in an ADS are presented. The impact of weighting functions on the point-kinetics parameters at steady-state is shown. Some preliminary results of using a space-time kinetics model for beam-trip related transients are highlighted. (orig.)

  3. Research on the improvement of nuclear safety -Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Kim, Dong Ha; Park, Won Seok; Hwang, Mi Jeong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated. These results will give a physical insight in the development of a new dispersion model. A wind tunnel experiment with bell shaped hill model was made in order to develop a new dispersion model. And an improved dispersion model was developed based on the concentration distribution data obtained from the wind tunnel experiment. This model will be added as an option to the atmospheric dispersion code. A stand-alone atmospheric code using MS Visual Basic programming language which runs at the Windows environment of a PC was developed. A user can easily select a necessary data file and type input data by clicking menus, and can select calculation options such building wake, plume rise etc., if necessary. And a user can easily understand the meaning of concentration distribution on the map around the plant site as well as output files. Also the methodologies for the estimation of radiation exposure and for the calculation of risks was established. These methodologies will be used for the development of modules for the radiation exposure and risks respectively. These modules will be developed independently and finally will be combined to the atmospheric dispersion code in order to develop a level 3 PSA code. 30 tabs., 56 figs., refs. (Author).

  4. Applicability of Coupled Thermalhydraulic Codes for Safety Analysis of Nuclear Reactors

    International Nuclear Information System (INIS)

    Gairola, A.; Bhowmik, P. K.; Shamim, J. A.; Suh, K. Y.

    2014-01-01

    To this end computational codes like RELAP and TRACE are used to model thermal-hydraulic response of nuclear power plant during an accident. By careful modeling and significant user experience these system codes are able to simulate the behavior of primary system and the containment to a reasonable extent. Comparatively decoupled simulation is simple but might not produce reality and the physics involved in an accurate manner. Thus simulation using two different system codes is interesting as the whole system is coupled through the pressure in the containment and flow through the break. Using this methodology it might be possible to get new insight about the primary and containment behavior by the precise simulation of the accident both in the current reactors and future Gen-III/III+ reactors. Couple thermalhydraulic code methodology is still new and require further investigations. Applicability of such methodology to the GEN-II plants have met with limited success, however a number of situations in which this methodology could be applied are still unexplored and thus provides a room for improvement and modifications

  5. Applicability of Coupled Thermalhydraulic Codes for Safety Analysis of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gairola, A.; Bhowmik, P. K.; Shamim, J. A.; Suh, K. Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    To this end computational codes like RELAP and TRACE are used to model thermal-hydraulic response of nuclear power plant during an accident. By careful modeling and significant user experience these system codes are able to simulate the behavior of primary system and the containment to a reasonable extent. Comparatively decoupled simulation is simple but might not produce reality and the physics involved in an accurate manner. Thus simulation using two different system codes is interesting as the whole system is coupled through the pressure in the containment and flow through the break. Using this methodology it might be possible to get new insight about the primary and containment behavior by the precise simulation of the accident both in the current reactors and future Gen-III/III+ reactors. Couple thermalhydraulic code methodology is still new and require further investigations. Applicability of such methodology to the GEN-II plants have met with limited success, however a number of situations in which this methodology could be applied are still unexplored and thus provides a room for improvement and modifications.

  6. Research on the improvement of nuclear safety -Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Kim, Dong Ha; Park, Won Seok; Hwang, Mi Jeong

    1995-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated. These results will give a physical insight in the development of a new dispersion model. A wind tunnel experiment with bell shaped hill model was made in order to develop a new dispersion model. And an improved dispersion model was developed based on the concentration distribution data obtained from the wind tunnel experiment. This model will be added as an option to the atmospheric dispersion code. A stand-alone atmospheric code using MS Visual Basic programming language which runs at the Windows environment of a PC was developed. A user can easily select a necessary data file and type input data by clicking menus, and can select calculation options such building wake, plume rise etc., if necessary. And a user can easily understand the meaning of concentration distribution on the map around the plant site as well as output files. Also the methodologies for the estimation of radiation exposure and for the calculation of risks was established. These methodologies will be used for the development of modules for the radiation exposure and risks respectively. These modules will be developed independently and finally will be combined to the atmospheric dispersion code in order to develop a level 3 PSA code. 30 tabs., 56 figs., refs. (Author)

  7. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    International Nuclear Information System (INIS)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H 2 /air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author)

  8. A cloud medication safety support system using QR code and Web services for elderly outpatients.

    Science.gov (United States)

    Tseng, Ming-Hseng; Wu, Hui-Ching

    2014-01-01

    Drug is an important part of disease treatment, but medication errors happen frequently and have significant clinical and financial consequences. The prevalence of prescription medication use among the ambulatory adult population increases with advancing age. Because of the global aging society, outpatients need to improve medication safety more than inpatients. The elderly with multiple chronic conditions face the complex task of medication management. To reduce the medication errors for the elder outpatients with chronic diseases, a cloud medication safety supporting system is designed, demonstrated and evaluated. The proposed system is composed of a three-tier architecture: the front-end tier, the mobile tier and the cloud tier. The mobile tier will host the personalized medication safety supporting application on Android platforms that provides some primary functions including reminders for medication, assistance with pill-dispensing, recording of medications, position of medications and notices of forgotten medications for elderly outpatients. Finally, the hybrid technology acceptance model is employed to understand the intention and satisfaction level of the potential users to use this mobile medication safety support application system. The result of the system acceptance testing indicates that this developed system, implementing patient-centered services, is highly accepted by the elderly. This proposed M-health system could assist elderly outpatients' homecare in preventing medication errors and improving their medication safety.

  9. About the use of the CATHARE code for best estimate Large Break LOCA calculations: benefits for Safety and constraints

    International Nuclear Information System (INIS)

    Vacher, J.L.

    1994-01-01

    Since 1979, EDF has participated to the development of CATHARE, a best estimate accidental thermalhydraulic code, in collaboration with CEA and FRAMATOME. EDF is now investigating the use of this code for licensing studies and particularly for Large Break LOCA calculations. Until now, the work done at EDF, in relation with FRAMATOME and CEA, has mainly focused on the physical analysis of the transient and on the identification of the key phenomena. This task is a necessary step before uncertainty evaluation. To illustrate this point of view, a peculiar example of calculated Large Break transients for a 900 MW three loop plant is presented. In one of these calculations, a high value of Peak Cladding Temperature was obtained. This peculiar scenario was initiated by a large entrainment of water to the steam generators at the very beginning of the reflooding stage, followed by a strong pressurization which led to a lasting draining of the reactor vessel. The physical phenomena which determine the existence and amplitude of this scenario were identified and their influence was explained: condensation at the accumulator injection, heat exchange in the core, entrainment process to the steam generators. It appeared obvious that the large observed uncertainty was associated to only a few parameters. Although this peculiar system behaviour was obtained for only a particular combination of parameters and a narrow range of thermalhydraulic conditions, the capability of the code to simulate these phenomena was investigated in regard to experimental data. It was concluded that this scenario was definitely unrealistic on a reactor. Nevertheless, this peculiar example tends to demonstrate, firstly, that the use of a best-estimate code improves Safety as it makes possible to point out physical phenomena that could not be considered when using non mechanistic codes, secondly, that the uncertainty evaluation must be guided by a pertinent physical analysis of the transient, focusing

  10. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  11. Comparative health and safety assessment of the satellite power system and other electrical generation alternatives

    International Nuclear Information System (INIS)

    1980-12-01

    The work reported here is an analysis of existing data on the health and safety risks of a satellite power system and six electrical generation systems: a combined-cycle coal power system with a low-Btu gasifier and open-cycle gas turbine; a light water fission power system without fuel reprocessing; a liquid-metal, fast-breeder fission reactor; a centralized and decentralized, terrestrial, solar-photovoltaic power system; and a first-generation design for a fusion power system. The systems are compared on the basis of expected deaths and person-days lost per year associated with 1000 MW of average electricity generation. Risks are estimated and uncertainties indicated for all phases of the energy production cycle, including fuel and raw material extraction and processing, direct and indirect component manufacture, on-site construction, and system operation and maintenance. Also discussed is the potential significance of related major health and safety issues that remain largely unquantifiable. The appendices provide more detailed information on risks, uncertainties, additional research needed, and references for the identified impacts of each system

  12. Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III

    International Nuclear Information System (INIS)

    Suzuki, Tohru; Tobita, Yoshiharu; Kondo, Satoru; Saito, Yasushi; Mishima, Kaichiro

    2003-01-01

    SIMMER-III, a safety analysis code for liquid-metal fast reactors (LMFRs), includes a momentum exchange model based on conventional correlations for ordinary gas-liquid flows, such as an air-water system. From the viewpoint of safety evaluation of core disruptive accidents (CDAs) in LMFRs, we need to confirm that the code can predict the two-phase flow behaviors with high liquid-to-gas density ratios formed during a CDA. In the present study, the momentum exchange model of SIMMER-III was assessed and improved using experimental data of two-phase flows containing liquid metal, on which fundamental information, such as bubble shapes, void fractions and velocity fields, has been lacking. It was found that the original SIMMER-III can suitably represent high liquid-to-gas density ratio flows including ellipsoidal bubbles as seen in lower gas fluxes. In addition, the employment of Kataoka-Ishii's correlation has improved the accuracy of SIMMER-III for gas-liquid metal flows with cap-shape bubbles as identified in higher gas fluxes. Moreover, a new procedure, in which an appropriate drag coefficient can be automatically selected according to bubble shape, was developed. Through this work, the reliability and the precision of SIMMER-III have been much raised with regard to bubbly flows for various liquid-to-gas density ratios

  13. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  14. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  15. Patient safety with blood products administration using wireless and bar-code technology.

    Science.gov (United States)

    Porcella, Aleta; Walker, Kristy

    2005-01-01

    Supported by a grant from the Agency for Healthcare Research and Quality, a University of Iowa Hospitals and Clinics interdisciplinary research team created an online data-capture-response tool utilizing wireless mobile devices and bar code technology to track and improve blood products administration process. The tool captures 1) sample collection, 2) sample arrival in the blood bank, 3) blood product dispense from blood bank, and 4) administration. At each step, the scanned patient wristband ID bar code is automatically compared to scanned identification barcode on requisition, sample, and/or product, and the system presents either a confirmation or an error message to the user. Following an eight-month, 5 unit, staged pilot, a 'big bang,' hospital-wide implementation occurred on February 7, 2005. Preliminary results from pilot data indicate that the new barcode process captures errors 3 to 10 times better than the old manual process.

  16. Generating Safety-Critical PLC Code From a High-Level Application Software Specification

    Science.gov (United States)

    2008-01-01

    The benefits of automatic-application code generation are widely accepted within the software engineering community. These benefits include raised abstraction level of application programming, shorter product development time, lower maintenance costs, and increased code quality and consistency. Surprisingly, code generation concepts have not yet found wide acceptance and use in the field of programmable logic controller (PLC) software development. Software engineers at Kennedy Space Center recognized the need for PLC code generation while developing the new ground checkout and launch processing system, called the Launch Control System (LCS). Engineers developed a process and a prototype software tool that automatically translates a high-level representation or specification of application software into ladder logic that executes on a PLC. All the computer hardware in the LCS is planned to be commercial off the shelf (COTS), including industrial controllers or PLCs that are connected to the sensors and end items out in the field. Most of the software in LCS is also planned to be COTS, with only small adapter software modules that must be developed in order to interface between the various COTS software products. A domain-specific language (DSL) is a programming language designed to perform tasks and to solve problems in a particular domain, such as ground processing of launch vehicles. The LCS engineers created a DSL for developing test sequences of ground checkout and launch operations of future launch vehicle and spacecraft elements, and they are developing a tabular specification format that uses the DSL keywords and functions familiar to the ground and flight system users. The tabular specification format, or tabular spec, allows most ground and flight system users to document how the application software is intended to function and requires little or no software programming knowledge or experience. A small sample from a prototype tabular spec application is

  17. Safety analysis code input automation using the Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Kopp, H.; Leung, J.; Tajbakhsh, A.; Viles, F.

    1985-01-01

    The Nuclear Plant Data Bank (NPDB) is a computer-based system that organizes a nuclear power plant's technical data, providing mechanisms for data storage, retrieval, and computer-aided engineering analysis. It has the specific objective to describe thermohydraulic systems in order to support: rapid information retrieval and display, and thermohydraulic analysis modeling. The Nuclear Plant Data Bank (NPBD) system fully automates the storage and analysis based on this data. The system combines the benefits of a structured data base system and computer-aided modeling with links to large scale codes for engineering analysis. Emphasis on a friendly and very graphically oriented user interface facilitates both initial use and longer term efficiency. Specific features are: organization and storage of thermohydraulic data items, ease in locating specific data items, graphical and tabular display capabilities, interactive model construction, organization and display of model input parameters, input deck construction for TRAC and RELAP analysis programs, and traceability of plant data, user model assumptions, and codes used in the input deck construction process. The major accomplishments of this past year were the development of a RELAP model generation capability and the development of a CRAY version of the code

  18. Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2017-11-15

    The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

  19. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

    Energy Technology Data Exchange (ETDEWEB)

    Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

    2011-02-01

    This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

  20. Chronic electrical stimulation with a suprachoroidal retinal prosthesis: a preclinical safety and efficacy study.

    Directory of Open Access Journals (Sweden)

    David A X Nayagam

    Full Text Available To assess the safety and efficacy of chronic electrical stimulation of the retina with a suprachoroidal visual prosthesis.Seven normally-sighted feline subjects were implanted for 96-143 days with a suprachoroidal electrode array and six were chronically stimulated for 70-105 days at levels that activated the visual cortex. Charge balanced, biphasic, current pulses were delivered to platinum electrodes in a monopolar stimulation mode. Retinal integrity/function and the mechanical stability of the implant were assessed monthly using electroretinography (ERG, optical coherence tomography (OCT and fundus photography. Electrode impedances were measured weekly and electrically-evoked visual cortex potentials (eEVCPs were measured monthly to verify that chronic stimuli were suprathreshold. At the end of the chronic stimulation period, thresholds were confirmed with multi-unit recordings from the visual cortex. Randomized, blinded histological assessments were performed by two pathologists to compare the stimulated and non-stimulated retina and adjacent tissue.All subjects tolerated the surgical and stimulation procedure with no evidence of discomfort or unexpected adverse outcomes. After an initial post-operative settling period, electrode arrays were mechanically stable. Mean electrode impedances were stable between 11-15 kΩ during the implantation period. Visually-evoked ERGs & OCT were normal, and mean eEVCP thresholds did not substantially differ over time. In 81 of 84 electrode-adjacent tissue samples examined, there were no discernible histopathological differences between stimulated and unstimulated tissue. In the remaining three tissue samples there were minor focal fibroblastic and acute inflammatory responses.Chronic suprathreshold electrical stimulation of the retina using a suprachoroidal electrode array evoked a minimal tissue response and no adverse clinical or histological findings. Moreover, thresholds and electrode impedance remained

  1. Electrical safety of commercial Li-ion cells based on NMC and NCA technology compared to LFP technology

    OpenAIRE

    Brand, Martin; Gläser, Simon; Geder, Jan; Menacher, Stefan; Obpacher, Sebastian; Jossen, Andreas; Quinger, Daniel

    2013-01-01

    Since a laptop caught fire in 2006 at the latest, Li-ion cells were considered as more dangerous than other accumulators [1]. Recent incidents, such as the one involving a BYD e6 electric taxi [2] or the Boeing Dreamliner [3], give rise to questions concerning the safety of L#i-ion cells. This is a crucial point, since Li-ion cells are increasingly integrated in all kinds of (electric) vehicles. Therefore the economic success of hybrid electric vehicles (HEV) and battery electric vehicles (BE...

  2. Safety assessment for a disposal option of TENORM wastes coming from the electric generation in Cuba

    International Nuclear Information System (INIS)

    Leyva, Dennys; Gil, Reinaldo; Peralta, Jose L.; Odalys Ramos

    2008-01-01

    The aim of the present paper was the safety assessment for a disposal option of ashes wastes coming from the electric generation in Cuba. The ashes are planned to be disposed as subsurface layer, covered with soil under controlled conditions. The composition of theses wastes are TENORM ( 226 Ra and 224 Ra) and heavy metals (vanadium, chromium, zinc), therefore, their disposal should accomplish the national and international defined regulations. The adopted safety assessment methodology, allowed the identification and selection of the main scenarios to evaluate, the mathematical models to apply and the comparison against the assessment criteria. According to the assessment context and the site characteristics, the atmospheric and groundwater scenarios were evaluated. During the modelling stage were included the identification of the main exposure pathways and the most relevant assessment processes were modelled (transport of contaminants, radioactive decay, etc.). For atmospheric dispersion, the SCREEN3 model was adopted, including the radioactive decay and other radiological properties. The DRAF model was used for the groundwater scenario. The doses for inhalation, external irradiation and foodstuff ingestion were obtained using several dosimetric models. The results showed that the 226 Ra concentration values were higher than the 228 Ra in the evaluation points, for atmospheric and groundwater scenarios. This behaviour is influenced by the small radioactive inventory, the shorter half life of the 228 Ra and the distance between the disposal site and the evaluation points. The obtained external doses were always below the dose limits for the members of the public and for all scenarios, including the more conservatives. The lower dose (by ingestion) values were associated to the scenarios of radionuclides transport through the geosphere. According the safety assessment and the established scenarios, the evaluated disposal practice does not represent a relevant

  3. Status of non-electric nuclear heat applications: Technology and safety

    International Nuclear Information System (INIS)

    2000-11-01

    Nuclear energy plays an important role in electricity generation, producing 16% of the world's electricity at the beginning of 1999. It has proven to be safe, reliable, economical and has only a minimal impact on the environment. Most of the world's energy consumption, however, is in the form of heat. The market potential for nuclear heat was recognized early. Some of the first reactors were used for heat supply, e.g. Calder Hall (United Kingdom), Obninsk (Russian Federation), and Agesta (Sweden). Now, over 60 reactors are supplying heat for district heating, industrial processes and seawater desalination. But the nuclear option could be better deployed if it would provide a larger share of the heat market. In particular, seawater desalination using nuclear heat is of increasing interest to some IAEA Member States. In consideration of the growing experience being accumulated, the IAEA periodically reviews the progress and new developments in the field of nuclear heat applications. This publication summarizes the recent activities among Member States presented at a Technical Committee meeting in April 1999. The purpose of the meeting was to provide a forum for the exchange of up to date information on the prospect, design, safety and licensing aspects, and development of non-electrical applications of nuclear heat for industrial use. This mainly included seawater desalination and hydrogen production

  4. Current status of environmental, health, and safety issues of nickel metal-hydride batteries for electric vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Corbus, D; Hammel, C J; Mark, J

    1993-08-01

    This report identifies important environment, health, and safety issues associated with nickel metal-hydride (Ni-MH) batteries and assesses the need for further testing and analysis. Among the issues discussed are cell and battery safety, workplace health and safety, shipping requirements, and in-vehicle safety. The manufacture and recycling of Ni-MH batteries are also examined. This report also overviews the ``FH&S`` issues associated with other nickel-based electric vehicle batteries; it examines venting characteristics, toxicity of battery materials, and the status of spent batteries as a hazardous waste.

  5. Current status of environmental, health, and safety issues of nickel metal-hydride batteries for electric vehicles

    International Nuclear Information System (INIS)

    Corbus, D.; Hammel, C.J.; Mark, J.

    1993-08-01

    This report identifies important environment, health, and safety issues associated with nickel metal-hydride (Ni-MH) batteries and assesses the need for further testing and analysis. Among the issues discussed are cell and battery safety, workplace health and safety, shipping requirements, and in-vehicle safety. The manufacture and recycling of Ni-MH batteries are also examined. This report also overviews the ''FH ampersand S'' issues associated with other nickel-based electric vehicle batteries; it examines venting characteristics, toxicity of battery materials, and the status of spent batteries as a hazardous waste

  6. Environmental, health, and safety issues of sodium-sulfur batteries for electric and hybrid vehicles. Volume 1, Cell and battery safety

    Energy Technology Data Exchange (ETDEWEB)

    Ohi, J M

    1992-09-01

    This report is the first of four volumes that identify and assess the environmental, health, and safety issues involved in using sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles that may affect the commercialization of Na/S batteries. This and the other reports on recycling, shipping, and vehicle safety are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD&D) program for Na/S battery technology. The reports review the status of Na/S battery RD&D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers cell design and engineering as the basis of safety for Na/S batteries and describes and assesses the potential chemical, electrical, and thermal hazards and risks of Na/S cells and batteries as well as the RD&D performed, under way, or to address these hazards and risks. The report is based on a review of the literature and on discussions with experts at DOE, national laboratories and agencies, universities, and private industry. Subsequent volumes will address environmental, health, and safety issues involved in shipping cells and batteries, using batteries to propel electric vehicles, and recycling and disposing of spent batteries. The remainder of this volume is divided into two major sections on safety at the cell and battery levels. The section on Na/S cells describes major component and potential failure modes, design, life testing and failure testing, thermal cycling, and the safety status of Na/S cells. The section on batteries describes battery design, testing, and safety status. Additional EH&S information on Na/S batteries is provided in the appendices.

  7. R&D for Safety Codes and Standards: Materials and Components Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    San Marchi, Christopher W. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States)

    2014-08-01

    A principal challenge to the widespread adoption of hydrogen infrastructure is the lack of quantifiable data on its safety envelope and concerns about additional risk from hydrogen. To convince regulatory officials, local fire marshals, fuel suppliers, and the public at large that hydrogen refueling is safe for consumer use, the risk to personnel and bystanders must be quantified and minimized to an acceptable level. Such a task requires strong confidence in the safety performance of high pressure hydrogen systems. Developing meaningful materials characterization and qualification methodologies in addition to enhancing understanding of performance of materials is critical to eliminating barriers to the development of safe, low-cost, high-performance high-pressure hydrogen systems for the consumer environment.

  8. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  9. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    International Nuclear Information System (INIS)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

  10. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    International Nuclear Information System (INIS)

    Hwnag, M.

    2001-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented

  11. Order of the 29. of January 2010 approving the decision n. 2009-DC-0146 made on the 16. of July 2010 by the Nuclear Safety Authority (ASN), modified by the decision n. 2009-DC-0162 of the 20. of October 2009, defining the list electric equipment generating X rays which are hold or used for purposes of biomedical research or medical, dental, forensics or veterinary diagnosis and submitted to a declaration regime on the account of the first paragraph of the article R. 1333-19 of the Public Health code and abrogating the 23 April 1969 order related to equipment and installations using ionizing radiations on medical purposes

    International Nuclear Information System (INIS)

    2010-01-01

    This legal document indicates the list of equipment which must now be declared to the ASN (the French Nuclear Safety Authority), specifies that any use of these equipment in conditions other than normal requires an authorization on the account of an article of the Public Health Code. A list of equipment is given which are used in medical and forensics radio-diagnosis, in biomedical research, in dental and veterinary radio-diagnosis

  12. Order of the 29. of January 2010 approving the decision n. 2009-DC-0146 made on the 16. of July 2010 by the Nuclear Safety Authority (ASN), modified by the decision n. 2009-DC-0162 of the 20. of October 2009, defining the list electric equipment generating X rays which are hold or used for purposes of biomedical research or medical, dental, forensics or veterinary diagnosis and submitted to a declaration regime on the account of the first paragraph of the article R. 1333-19 of the Public Health code and abrogating the 23 April 1969 order related to equipment and installations using ionizing radiations on medical purposes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    This legal document indicates the list of equipment which must now be declared to the ASN (the French Nuclear Safety Authority), specifies that any use of these equipment in conditions other than normal requires an authorization on the account of an article of the Public Health Code. A list of equipment is given which are used in medical and forensics radio-diagnosis, in biomedical research, in dental and veterinary radio-diagnosis

  13. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  14. Calibration of brick masonry partial safety factors for the South African Code

    CSIR Research Space (South Africa)

    Mahachi, J

    2007-09-01

    Full Text Available VR of the resistance of the member. The safety index β is determined as follows: )( fp1−−= φβ (4) Where φ-1() is the inverse of the cumulative normal distribution and pf is the probability of failure at the ultimate limit state... carefully controlled. However, data on the effect of inspection on Rn and VR, and on the variability in construction quality control in South Africa is not available. Current partial resistance factors based on the modified British stochastic models...

  15. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  16. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  17. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2004-01-01

    Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

  18. Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Honda, T.; Maki, K.; Okazaki, T.

    1994-01-01

    Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

  19. Proceedings of the workshop on Benchmarking of CFD Codes for Application to Nuclear Reactor Safety (CFD4NRS)

    International Nuclear Information System (INIS)

    2007-01-01

    Computational Fluid Dynamics (CFD) is to an increasing extent being adopted in nuclear reactor safety analyses as a tool that enables specific safety relevant phenomena occurring in the reactor coolant system to be better described. The Committee on the Safety of Nuclear Installations (CSNI), which is responsible for the activities of the Nuclear Energy Agency that support advancing the technical base of the safety of nuclear installations, has in recent years conducted an important activity in the CFD area. This activity has been carried out within the scope of the CSNI working group on the analysis and management of accidents (GAMA), and has mainly focused on the formulation of user guidelines and on the assessment and verification of CFD codes. It is in this GAMA framework that the present workshop was organized and carried out. The purpose of the workshop was to provide a forum for numerical analysts and experimentalists to exchange information in the field of NRS-related activities relevant to CFD validation, with the objective of providing input to GAMA CFD experts to create a practical, state-of-the-art, web-based assessment matrix on the use of CFD for NRS applications. Numerical simulations with a strong emphasis on validation were welcomed in such areas as heat transfer, buoyancy, stratification, natural circulation, free-surface modelling, turbulent mixing and multi-phase flow. These would relate to such NRS-relevant issues as: pressurized thermal shocks, boron dilution, hydrogen distribution, induced breaks, thermal striping, etc. The use of systematic error quantification and Best Practice Guidelines was encouraged. Papers reporting experiments providing high-quality data suitable for CFD validation, specifically in the area of NRS, were given high priority. Here, emphasis was placed on the availability of local measurements, especially multi-dimensional velocity measurements obtained using such techniques as laser-doppler velocimetry, hot

  20. Challenges for the nuclear safety of the deregulation of electricity markets

    International Nuclear Information System (INIS)

    2001-01-01

    Eurosafe 2000 was organised around two round tables on the first day and four seminars on the second day. The first round table dealt with general aspects of deregulation including the economic constraints and the special challenges arising during transition from regulated to deregulated structures. The second round table focussed on technical and organisational safety issues which are directly or indirectly related to the changes introduced by deregulation. The four seminars hold in order to provide opportunities for comparing experiences and learning about recent activities of IPSN, GRS and their partners in the European Union and Eastern Europe: Seminar 1 (Nuclear installation safety, assessment and analysis): assessment of the flooding incident at the Blayais nuclear power plant; PSA data base, comparison of the French and German approach; assessment of the Balakovo fire probabilistic study and elaboration of a guide for reviewing fire PSA; comprehensive technical assessment of an advanced German PWR by PSA - objectives and main results; PSA approach for the safety assessment of low-power and shutdown states; correlation of initiating events with the PSA level-2 results; safety assessment for fission products tests in the Phebus reactor; use of NPP simulators for applied human factor studies; assessment of the 'deterministic realistic method' applied to large LOCA analysis; assessment of the feasibility of an improvement programme enabling operation of units 3 and 4 of Kozloduy nuclear power plant. Seminar 2 (nuclear installation safety, research): PHEBUS 2K project on severe accidents; current status of the COCOSYS development; fission product modeling in ASTEC; Euratom Framework Programme (FP) research in reactor safety: main achievements of FP- 4 (1994-1998), some preliminary results of FP-5 (1998-2002) and prospects for beyond 2002; development of coupled systems of 3D neutronics and fluid-dynamic system codes and their application for safety analyses

  1. Assessment of the safety of aquatic animal commodities for international trade: the OIE Aquatic Animal Health code.

    Science.gov (United States)

    Oidtmann, B; Johnston, C; Klotins, K; Mylrea, G; Van, P T; Cabot, S; Martin, P Rosado; Ababouch, L; Berthe, F

    2013-02-01

    Trading of aquatic animals and aquatic animal products has become increasingly globalized during the last couple of decades. This commodity trade has increased the risk for the spread of aquatic animal pathogens. The World Organisation for Animal Health (OIE) is recognized as the international standard-setting organization for measures relating to international trade in animals and animal products. In this role, OIE has developed the Aquatic Animal Health Code, which provides health measures to be used by competent authorities of importing and exporting countries to avoid the transfer of agents pathogenic for animals or humans, whilst avoiding unjustified sanitary barriers. An OIE ad hoc group developed criteria for assessing the safety of aquatic animals or aquatic animal products for any purpose from a country, zone or compartment not declared free from a given disease 'X'. The criteria were based on the absence of the pathogenic agent in the traded commodity or inactivation of the pathogenic agent by the commercial processing used to produce the commodity. The group also developed criteria to assess the safety of aquatic animals or aquatic animal products for retail trade for human consumption from potentially infected areas. Such commodities were assessed considering the form and presentation of the product, the expected volume of waste tissues generated by the consumer and the likely presence of viable pathogenic agent in the waste. The ad hoc group applied the criteria to commodities listed in the individual disease chapters of the Aquatic Animal Health Code (2008 edition). Revised lists of commodities for which no additional measures should be required by the importing countries regardless of the status for disease X of the exporting country were developed and adopted by the OIE World Assembly of Delegates in May 2011. The rationale of the criteria and their application will be explained and demonstrated using examples. © 2012 Crown Copyright. Reproduced

  2. RDS; A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

    2013-01-01

    Full-text: Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBIMOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges. (author)

  3. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    International Nuclear Information System (INIS)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat

    2014-01-01

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges

  4. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  5. On the complex analysis of the reliability, safety, and economic efficiency of atomic electric power stations

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Klemin, A.I.; Polyakov, E.F.

    1977-01-01

    The problem is posed of effectively increasing the engineering performance of nuclear electric power stations (APS). The principal components of the engineering performance of modern large APS are considered: economic efficiency, radiation safety, reliability, and their interrelationship. A nomenclature is proposed for the quantitative indices which most completely characterize the enumerated properties and are convenient for the analysis of the engineering performance. The urgent problem of developing a methodology for the complex analysis and optimization of the principal performance components is considered; this methodology is designed to increase the efficiency of the work on high-performance competitive APS. The principle of complex optimization of the reliability, safety, and economic-efficiency indices is formulated; specific recommendations are made for the practical realization of this principle. The structure of the complex quantiative analysis of the enumerated performance components is given. The urgency and promise of the complex approach to solving the problem of APS optimization is demonstrated, i.e., the solution of the problem of creating optimally reliable, fairly safe, and maximally economically efficient stations

  6. Comparative health and safety assessment of alternative future electrical-generation systems

    International Nuclear Information System (INIS)

    Habegger, L.J.; Gasper, J.R.; Brown, C.D.

    1980-01-01

    The report is an analysis of health and safety risks of seven alternative electrical generation systems, all of which have potential for commercial availability in the post-2000 timeframe. The systems are compared on the basis of expected public and occupational deaths and lost workdays per year associated with 1000 MWe average unit generation. Risks and their uncertainties are estimated for all phases of the energy production cycle, including fuel and raw material extraction and processing, direct and indirect component manufacture, on-site construction, and system operation and maintenance. Also discussed is the potential significance of related major health and safety issues that remain largely unquantifiable. The technologies include: the SPS; a low-Btu coal gasification system with an open-cycle gas turbine combined with a steam topping cycle (CG/CC); a light water fission reactor system without fuel reprocessing (LWR); a liquid metal fast breeder fission reactor system (LMFBR); a central station terrestrial photovoltaic system (CTPV); and a first generation fusion system with magnetic confinement. For comparison with the baseload technologies, risk from a decentralized roof-top photovoltaic system with 6 kWe peak capacity and battery storage (DTPV) was also evaluated

  7. Lessons learned while implementing a safety parameter display system at the Comanche Peak steam electric station

    International Nuclear Information System (INIS)

    Hagar, B.

    1987-01-01

    With the completion of site Verification and Validation tests, the Safety Parameter Display System (SPDS) will be fully operational at the Comanche Peak Steam Electric Station. Implementation of the SPDS, which began in 1982, included: modifying generic Safety Assessment System Software; developing site-specific displays and features; installing and integrating system equipment into the plant; modifying station heating, ventilation, and air conditioning systems to provide necessary cooling; installing an additional uninterruptible power supply system to provide necessary power; and training station personnel in the operation and use of the system. Lessons learned during this project can be discussed in terms of an ideal SPDS implementation project. Such a project would design and implement an SPDS for a plant that is already under construction or operating, and would progress through a sequence of activities that includes: (1) developing and documenting the system design bases, and including all major design influences; (2) developing a database description and system functional specifications to clarify specific system requirements; (3) developing detailed system hardware and software design specifications to fully describe the system, and to enable identification of necessary site design changes early in the project; (4) implementing the system design; (5) configuring and extensively testing the system prior to routine system operation; and (6) tuning the system after the completion of system installation. The ideal project would include future system users in design development and system testing, and would use Verification and Validation techniques throughout the project to ensure that each sequential step is appropriate and correct

  8. Quantifying reactor safety margins: Part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology

    International Nuclear Information System (INIS)

    Boyack, B.E.; Duffey, R.B.; Griffith, P.

    1988-01-01

    In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of emergency core cooling systems (ECCS) entitled ''Emergency Core Cooling System; Revisions to Acceptance Criteria.'' The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and included that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. It is an auditable, traceable, and practical method for combining quantitative analyses and expert opinions to arrive at computed values of uncertainty. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 /times/ 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3. 23 refs., 5 figs., 1 tab

  9. Current status of environmental, health, and safety issues of lithium ion electric vehicle batteries

    Energy Technology Data Exchange (ETDEWEB)

    Vimmerstedt, L.J.; Ring, S.; Hammel, C.J.

    1995-09-01

    The lithium ion system considered in this report uses lithium intercalation compounds as both positive and negative electrodes and has an organic liquid electrolyte. Oxides of nickel, cobalt, and manganese are used in the positive electrode, and carbon is used in the negative electrode. This report presents health and safety issues, environmental issues, and shipping requirements for lithium ion electric vehicle (EV) batteries. A lithium-based electrochemical system can, in theory, achieve higher energy density than systems using other elements. The lithium ion system is less reactive and more reliable than present lithium metal systems and has possible performance advantages over some lithium solid polymer electrolyte batteries. However, the possibility of electrolyte spills could be a disadvantage of a liquid electrolyte system compared to a solid electrolyte. The lithium ion system is a developing technology, so there is some uncertainty regarding which materials will be used in an EV-sized battery. This report reviews the materials presented in the open literature within the context of health and safety issues, considering intrinsic material hazards, mitigation of material hazards, and safety testing. Some possible lithium ion battery materials are toxic, carcinogenic, or could undergo chemical reactions that produce hazardous heat or gases. Toxic materials include lithium compounds, nickel compounds, arsenic compounds, and dimethoxyethane. Carcinogenic materials include nickel compounds, arsenic compounds, and (possibly) cobalt compounds, copper, and polypropylene. Lithiated negative electrode materials could be reactive. However, because information about the exact compounds that will be used in future batteries is proprietary, ongoing research will determine which specific hazards will apply.

  10. Development of a graphical user interface for the TRAC plant/safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, A.E.; Harkins, C.K.; Smith, R.J.

    1995-09-01

    A graphical user interface (GUI) for the Transient Reactor Analysis Code (TRAC) has been developed at Knolls Atomic Power Laboratory. This X Window based GUI supports the design and analysis process, acting as a preprocessor, runtime editor, help system and post processor to TRAC-PF1/MOD2. TRAC was developed at the Los Alamos National Laboratory (LANL). The preprocessor is an icon-based interface which allows the user to create a TRAC model. When the model is complete, the runtime editor provides the capability to execute and monitor TRAC runs on the workstation or supercomputer. After runs are made, the output processor allows the user to extract and format data from the TRAC graphics file. The TRAC GUI is currently compatible with TRAC-PF1/MOD2 V5.3 and is available with documentation from George Niederauer, Section Leader of the Software Development Section, Group TSA-8, at LANL. Users may become functional in creating, running, and interpreting results from TRAC without having to know Unix commands and the detailed format of any of the data files. This reduces model development and debug time and increases quality control. Integration with post-processing and visualization tools increases engineering effectiveness.

  11. Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

    International Nuclear Information System (INIS)

    Todorova, N.K.; Ivanov, K.N.

    2003-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement

  12. Related research with thermo hydraulics safety by means of Trace code

    International Nuclear Information System (INIS)

    Chaparro V, F. J.; Del Valle G, E.; Rodriguez H, A.; Gomez T, A. M.; Sanchez E, V. H.; Jager, W.

    2014-10-01

    In this article the results of the design of a pressure vessel of a BWR/5 similar to the type of Laguna Verde NPP are presented, using the Trace code. A thermo hydraulics Vessel component capable of simulating the behavior of fluids and heat transfer that occurs within the reactor vessel was created. The Vessel component consists of a three-dimensional cylinder divided into 19 axial sections, 4 azimuthal sections and two concentric radial rings. The inner ring is used to contain the core and the central part of the reactor, while the outer ring is used as a down comer. Axial an azimuthal divisions were made with the intention that the dimensions of the internal components, heights and orientation of the external connections match the reference values of a reactor BWR/5 type. In the model internal components as, fuel assemblies, steam separators, jet pumps, guide tubes, etc. are included and main external connections as, steam lines, feed-water or penetrations of the recirculation system. The model presents significant simplifications because the object is to keep symmetry between each azimuthal section of the vessel. In most internal components lack a detailed description of the geometry and initial values of temperature, pressure, fluid velocity, etc. given that it only considered the most representative data, however with these simulations are obtained acceptable results in important parameters such as the total flow through the core, the pressure in the vessel, percentage of vacuums fraction, pressure drop in the core and the steam separators. (Author)

  13. Development of a graphical user interface for the TRAC plant/safety analysis code

    International Nuclear Information System (INIS)

    Kelly, A.E.; Harkins, C.K.; Smith, R.J.

    1995-01-01

    A graphical user interface (GUI) for the Transient Reactor Analysis Code (TRAC) has been developed at Knolls Atomic Power Laboratory. This X Window based GUI supports the design and analysis process, acting as a preprocessor, runtime editor, help system and post processor to TRAC-PF1/MOD2. TRAC was developed at the Los Alamos National Laboratory (LANL). The preprocessor is an icon-based interface which allows the user to create a TRAC model. When the model is complete, the runtime editor provides the capability to execute and monitor TRAC runs on the workstation or supercomputer. After runs are made, the output processor allows the user to extract and format data from the TRAC graphics file. The TRAC GUI is currently compatible with TRAC-PF1/MOD2 V5.3 and is available with documentation from George Niederauer, Section Leader of the Software Development Section, Group TSA-8, at LANL. Users may become functional in creating, running, and interpreting results from TRAC without having to know Unix commands and the detailed format of any of the data files. This reduces model development and debug time and increases quality control. Integration with post-processing and visualization tools increases engineering effectiveness

  14. Environmental, health, and safety issues of sodium-sulfur batteries for electric and hybrid vehicles. Volume 4, In-vehicle safety

    Energy Technology Data Exchange (ETDEWEB)

    Mark, J.

    1992-11-01

    This report is the last of four volumes that identify and assess the environmental, health, and safety issues that may affect the commercial-scale use of sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles. The reports are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD&D) program for Na/S battery technology. The reports review the status of Na/S battery RD&D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers the in-vehicle safety issues of electric vehicles powered by Na/S batteries. The report is based on a review of the literature and on discussions with experts at DOE, national laboratories and agencies, and private industry. It has three major goals: (1) to identify the unique hazards associated with electric vehicle (EV) use; (2) to describe the existing standards, regulations, and guidelines that are or could be applicable to these hazards; and (3) to discuss the adequacy of the existing requirements in addressing the safety concerns of EVs.

  15. Development of a Code for the Long Term Radiological Safety Assessment of Radioactive Wastes from Advanced Nuclear Fuel Cycle Facilities in Republic of Korea

    International Nuclear Information System (INIS)

    Hwang, Yong Soo

    2010-01-01

    For the purpose of evaluating annual individual doses from a potential repository disposing of radioactive wastes from the operation of the prospective advanced nuclear fuel cycle facilities in Korea, the new safety assessment code based on the Goldsim has been developed. It was designed to compare the environmental impacts from many fuel cycle options such as direct disposal, wet and dry recycling. The code based on the compartment theory can be applied to assess both normal and what if scenarios

  16. Quantifying reactor safety margins: Application of CSAU [Code Scalability, Applicability and Uncertainty] methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

    International Nuclear Information System (INIS)

    Wulff, W.; Boyack, B.E.; Duffey, R.B.

    1988-01-01

    Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs

  17. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    Science.gov (United States)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  18. Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3 (Docket No. 50-382)

    International Nuclear Information System (INIS)

    1985-03-01

    Supplement 10 to the Safety Evaluation Report for the application filed by Louisiana Power and Light Company for a license to operate the Waterford Steam Electric Station, Unit 3 (Docket No. 50-382), located in St. Charles Parish, Louisiana, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of information submitted by the licensee since the Safety Evaluation Report and its nine previous supplements were issued

  19. PSAPACK 4.2. A code for probabilistic safety assessment level 1. User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    Only limited use has been made until now of the large amount of information contained in probabilistic safety assessments (PSAs). This is mainly due to the complexity of the PSA reports and the difficulties in obtaining intermediate results and in performing updates and recalculations. Moreover, PSA software was developed for mainframe computers, and the files of information such as fault trees and accident sequences were intended for the use of the analysts carrying out PSA studies or other skilled PSA practitioners. The increasing power and availability of personal computers (PCs) and developments in recent years in both hardware and software have made it possible to develop PSA software for use in PCs. Furthermore, the operational characteristics of PCs make them attractive not only for performing PSAs but also for updating the results and in using them in day-to-day applications. The IAEA has therefore developed in co-operation with its Member States, a software package (PSAPACK) for PCs for use in performing a Level 1 PSA and for easy interrogation of the results. Figs.

  20. PSAPACK 4.2. A code for probabilistic safety assessment level 1. User's manual

    International Nuclear Information System (INIS)

    1995-01-01

    Only limited use has been made until now of the large amount of information contained in probabilistic safety assessments (PSAs). This is mainly due to the complexity of the PSA reports and the difficulties in obtaining intermediate results and in performing updates and recalculations. Moreover, PSA software was developed for mainframe computers, and the files of information such as fault trees and accident sequences were intended for the use of the analysts carrying out PSA studies or other skilled PSA practitioners. The increasing power and availability of personal computers (PCs) and developments in recent years in both hardware and software have made it possible to develop PSA software for use in PCs. Furthermore, the operational characteristics of PCs make them attractive not only for performing PSAs but also for updating the results and in using them in day-to-day applications. The IAEA has therefore developed in co-operation with its Member States, a software package (PSAPACK) for PCs for use in performing a Level 1 PSA and for easy interrogation of the results. Figs

  1. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  2. Requirements to be taken into account in the design, qualification startup and operation of electrical equipment for safety-related electrical systems

    International Nuclear Information System (INIS)

    1985-07-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to provide the rules to be respected in order that safety-related electrical systems can perform its function under plausible operating conditions

  3. Overlapping Nuclear Safety Control Provisions of the Atomic Energy Act and Electric Utility Act

    International Nuclear Information System (INIS)

    Chang, Gun-Hyun; Kim, Sang-Won; Koh, Jae-Dong; Ahn, Hyung-Joon; Kim, Chang-Bum

    2007-01-01

    Before May 17, 2005, Korea's nuclear power plant (hereinafter referred to as 'NNP') regulation system was two-pronged. Every NPP system consists of primary or secondary system, and each type was respectively regulated by the Atomic Energy Act(hereinafter referred to as 'AEA') and the Electric Utility Act(hereinafter referred to as 'EUA'). This unusual regulatory regime gave rise to a number of problems with respect to operation and safety. For this reason, the Enforcement Regulation of AEA and applicable Notice were revised on May 17, 2005 to the effect that all regulation on NPPs subject to EUA was brought under the purview of AEA, except regulation on business license for nuclear power generation under Article 7 of EUA and approval of plan of works for setting up electric installations (hereinafter referred to as 'construction plan') (including approval of any changes; the same shall apply hereinafter) under Article 61 thereof. From the point of view of the Ministry of Science and Technology, the regulation of NPPs by a single law has enhanced their safety. However, the Ministry of Commerce, Industry and Energy retains regulatory authority regarding NPPs. It reviews and approves construction plans for secondary system pursuant to Article 61 of EUA and Article 28 of the Enforcement Regulation thereof. This situation arose because Article 28 of the Enforcement Regulation of EUA continues to provide for matters related with nuclear power. Therefore, continued control of NPPs under EUA ignores the relationship and respective nature of AEA and EUA. There is also possibility of violation of a superseding law. Even if said provision is not in violation of a superseding law, Article 28 of the Enforcement Regulation of EUA poses the possibility of overlapping regulation, which may violate the principle of prohibiting excessive regulation, one of the principles of the Korean Constitution. Assessment of the dual regulatory system for review of secondary system requires (i

  4. CNE (Embalse nuclear power plant): probabilistic safety study. Electric power supply. Events sequence

    International Nuclear Information System (INIS)

    Figueroa, N.

    1987-01-01

    The plant response to the occurrence of the starting event 'total loss of electric power supply to class IV and class III' is analyzed. This involves the study of automatical actions of safety and process systems as well as the operator actions. The probabilistic evaluation of starting event frequency is performed through fault-tree techniques. The frequency of occurrence 'loss of electric power supply to class IV (λIV = 0.56/year) and the probability of failure to demand of 'reserve' generating groups (Pd III 6.79 x 10 -3 ) contribute to the mentioned frequency. As soon as the starting event occurs, the reactor power must be reduced to 0%, the fuel must be cooled through the thermo siphon and decay heat has to be removed. The events sequence analysis leads to the conclusion that the non shutting down of the reactor with any of the shutdown systems is 'incredible' (10 -6 /year). In all cases the fuel is cooled by building the thermo siphon except when a substantial inventory loss exist due to a closure failure of some valve of pressure and inventory control system. The order of magnitude of the failure of decay heat removal through the steam generators is 4 x 10 -4 . This removal would be assured by the emergency water system. Therefore, the frequency of the sequence of possible core meltdown, when the reactor does not shut down is: λ = 5 x 10 -9 /year and for the failure of heat removal: λ = 2 x 10 -6 /year. (Author)

  5. Current status of environmental, health, and safety issues of lithium polymer electric vehicle batteries

    Energy Technology Data Exchange (ETDEWEB)

    Corbus, D; Hammel, C J

    1995-02-01

    Lithium solid polymer electrolyte (SPE) batteries are being investigated by researchers worldwide as a possible energy source for future electric vehicles (EVs). One of the main reasons for interest in lithium SPE battery systems is the potential safety features they offer as compared to lithium battery systems using inorganic and organic liquid electrolytes. However, the development of lithium SPE batteries is still in its infancy, and the technology is not envisioned to be ready for commercialization for several years. Because the research and development (R&D) of lithium SPE battery technology is of a highly competitive nature, with many companies both in the United States and abroad pursuing R&D efforts, much of the information concerning specific developments of lithium SPE battery technology is proprietary. This report is based on information available only through the open literature (i.e., information available through library searches). Furthermore, whereas R&D activities for lithium SPE cells have focused on a number of different chemistries, for both electrodes and electrolytes, this report examines the general environmental, health, and safety (EH&S) issues common to many lithium SPE chemistries. However, EH&S issues for specific lithium SPE cell chemistries are discussed when sufficient information exists. Although lithium batteries that do not have a SPE are also being considered for EV applications, this report focuses only on those lithium battery technologies that utilize the SPE technology. The lithium SPE battery technologies considered in this report may contain metallic lithium or nonmetallic lithium compounds (e.g., lithium intercalated carbons) in the negative electrode.

  6. State of the art of computer codes and experimental investigations on safety of nuclear power plants with reactors of WWER type

    International Nuclear Information System (INIS)

    Asmolov, V.G.; Volkov, G.A.; Elikin, I.V.; Mysenkov, A.I.

    1987-01-01

    As is well-known investigations on mathematical models of accidental situations from the point of view of nuclear safety as well as their experimental investigation are of great importance in the design of reactor units for nuclear power plants. This paper gives a review of the state of the art of thermodynamic models and computer codes used for safety analysis of WWER reactors in the USSR, the experimental basis and experimental investigations of the appropriate thermal processes. The actual and future trends of theoretical and experimental investigation on safety problems of WWER type nuclear power plants are briefly described. (author)

  7. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  8. Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2016-05-15

    Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

  9. Development of thermal-hydraulic safety codes for HTGRs with gas-turbine and hydrogen process cycles

    International Nuclear Information System (INIS)

    No, Hee Cheon; Yoon, Ho Joon; Lee, Byung Jin; Kim, Yong Soo; Jin, Hyeng Gon; Kim, Ji Hwan; Kim, Hyeun Min; Lim, Hong Sik

    2008-01-01

    We present three nuclear/hydrogen-related R and D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA in which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed SANA code to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found out that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The GAMMA-SANA coupled code was assessed by comparing its results with the steady-state of the GTHTR300, and the load reduction transient was simulated for the 100% to 70% power operation. The calculation results confirm that two-dimensional throughflow modeling can be successfully used to describe the gas turbine behavior. The dynamic equations for the distillation column of the HI process in the I-S cycle are described with 4 material components involved in the HI process: H2O, HI, I2, and H2. For the VLE prediction in the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. Relative to the experimental data, the improved Neumann model shows deviations of 8.6% in maximum and 2.5% in average for the total pressure, and 9.5% in maximum for the liquid-liquid separation composition. Through a parametric analysis using the published experimental data related to the Bunsen reaction and liquid-liquid separation, an optimized operating condition for the

  10. Validation and application of the system code TRACE for safety related investigations of innovative nuclear energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim

    2011-12-19

    of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

  11. The nuclear safety regulation in Japan and the response to changes of circumstances surrounding the nuclear electricity generation

    International Nuclear Information System (INIS)

    Hombu, K.; Hirota, M.; Taniguchi, T.; Tanaka, N.; Akimoto, S.

    2001-01-01

    The influences of external factors on nuclear safety are discussed in this paper, based on the views on the circumstances of nuclear electricity generation. The following external factors, which might have some potential impacts on nuclear safety, are selected for discussion: (1) The deregulation in the electricity generation industry; (2) The modification of approval/certification system in the regulation of electricity generation; (3) The influences on social atmosphere due to the occurrence of a series of troubles; (4) The government reform and the structural adjustment of industry and (5) Others. Our further discussion seems to focus on the following 2 issues: (a) Whether nuclear power and the other electrical sources should compete with each other for short term economical cost, or whether factors of cost stability and competitiveness as well as longer term energy supply security and global environmental issues ranging over several decades should be considered; (b) How to realize the appropriate regulation from the perspective of public acceptance and confidence (when a series of troubles occur) without imposing unnecessary burdens on industry and without jeopardizing safety. These issues may be common among many countries and can be widely discussed. (author)

  12. Do code of conduct audits improve chemical safety in garment factories? Lessons on corporate social responsibility in the supply chain from Fair Wear Foundation.

    Science.gov (United States)

    Lindholm, Henrik; Egels-Zandén, Niklas; Rudén, Christina

    2016-10-01

    In managing chemical risks to the environment and human health in supply chains, voluntary corporate social responsibility (CSR) measures, such as auditing code of conduct compliance, play an important role. To examine how well suppliers' chemical health and safety performance complies with buyers' CSR policies and whether audited factories improve their performance. CSR audits (n = 288) of garment factories conducted by Fair Wear Foundation (FWF), an independent non-profit organization, were analyzed using descriptive statistics and statistical modeling. Forty-three per cent of factories did not comply with the FWF code of conduct, i.e. received remarks on chemical safety. Only among factories audited 10 or more times was there a significant increase in the number of factories receiving no remarks. Compliance with chemical safety requirements in garment supply chains is low and auditing is statistically correlated with improvements only at factories that have undergone numerous audits.

  13. Do code of conduct audits improve chemical safety in garment factories? Lessons on corporate social responsibility in the supply chain from Fair Wear Foundation

    Science.gov (United States)

    2016-01-01

    Background In managing chemical risks to the environment and human health in supply chains, voluntary corporate social responsibility (CSR) measures, such as auditing code of conduct compliance, play an important role. Objectives To examine how well suppliers’ chemical health and safety performance complies with buyers’ CSR policies and whether audited factories improve their performance. Methods CSR audits (n = 288) of garment factories conducted by Fair Wear Foundation (FWF), an independent non-profit organization, were analyzed using descriptive statistics and statistical modeling. Results Forty-three per cent of factories did not comply with the FWF code of conduct, i.e. received remarks on chemical safety. Only among factories audited 10 or more times was there a significant increase in the number of factories receiving no remarks. Conclusions Compliance with chemical safety requirements in garment supply chains is low and auditing is statistically correlated with improvements only at factories that have undergone numerous audits. PMID:27611103

  14. Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

    International Nuclear Information System (INIS)

    Kasselmann, S.; Druska, C.; Lauer, A.

    2010-01-01

    The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

  15. Simulation of the thermal behaviour of electric industrial resistance furnaces using the I-DEAS/TMG code; Simulation du comportement thermique des fours electriques industriels a resistances a l`aide du code I-DEAS/TMG

    Energy Technology Data Exchange (ETDEWEB)

    Plard, Ch; Branchu, K; Le Cloirec, B [Electricite de France, 77 - Moret sur Loing (France). Direction des Etudes et Recherches

    1997-12-31

    In order to answer the modeling needs of manufacturers and users of electrical resistance furnaces, Electricite de France (EdF) has been appealed to search for a numerical simulation tool for the modeling of thermal phenomena and to carry out its qualification. The TNG software has been retained according to its modeling characteristics of radiant heat transfers and to the coupling with thermal conduction. After a description of the main characteristics of this code, two examples of application to electrical furnaces are presented. The first example illustrates how it can be possible to accurately reproduce the behaviour of a big industrial furnace. The second example is an illustration of numerical simulation possibilities for the optimization of processes performed with an electric furnace. (J.S.)

  16. Simulation of the thermal behaviour of electric industrial resistance furnaces using the I-DEAS/TMG code; Simulation du comportement thermique des fours electriques industriels a resistances a l`aide du code I-DEAS/TMG

    Energy Technology Data Exchange (ETDEWEB)

    Plard, Ch.; Branchu, K.; Le Cloirec, B. [Electricite de France, 77 - Moret sur Loing (France). Direction des Etudes et Recherches

    1996-12-31

    In order to answer the modeling needs of manufacturers and users of electrical resistance furnaces, Electricite de France (EdF) has been appealed to search for a numerical simulation tool for the modeling of thermal phenomena and to carry out its qualification. The TNG software has been retained according to its modeling characteristics of radiant heat transfers and to the coupling with thermal conduction. After a description of the main characteristics of this code, two examples of application to electrical furnaces are presented. The first example illustrates how it can be possible to accurately reproduce the behaviour of a big industrial furnace. The second example is an illustration of numerical simulation possibilities for the optimization of processes performed with an electric furnace. (J.S.)

  17. Application of low bitrate image coding to surveillance of electric power facilities. Part 1. Proposal of low bitrate coding for surveillance of electric power facilities and examination of facilities region extraction method; Denryoku setsubi kanshi eno tei rate fugoka hoshiki no tekiyo. 1. Setsubi kanshiyo fugoka hoshiki no teian to setsubi ryoiki chushutsuho no kento

    Energy Technology Data Exchange (ETDEWEB)

    Murata, H.; Ishino, R. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    1996-03-01

    Current status of low bitrate image coding has been investigated, and a low bitrate coding suitable for the surveillance of electric power facilities has been proposed, to extract its problems to be solved. For the conventional image coding, the waveform coding has been used by which the images are processed as signals. While, for the MPEG-4, a coding method with considering the image information has been proposed. For these coding methods, however, image information lacks details primarily, when lowering the bitrate. Accordingly, these methods can not be applied when the details in the images are important, such as in the case of surveillance of facilities. Then, the coding method has been proposed by expanding the partially detailed coding, and by separating constituent images of facilities, such as power cables and steel towers, designated by operators. It is the special feature of this method that the method can easily respond to the low bitrate and the detailed information can be conserved by using the structure extraction coding for the designated partial image which is generally processed by the low bitrate waveform coding. 29 refs., 17 figs., 1 tab.

  18. Discussion on Integration of Welding Coordinator in Welding Quality System of KEPIC(Korea Electric Power Industry code)

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyun-Jae; Sohn, Myoung-Sung; Cho, Kyoung-Youn; Kim, Jong-Hae [Korea Electric Association, Seoul (Korea, Republic of)

    2015-05-15

    The welding quality system of KEPIC-MQW 'Welding Qualification' referencing ASME BPVC Sec.IX, Part QW requires welding procedures and performance qualification of welder or welding operator excluding welding coordinator. It means that there is potential possibility of any problem in process of welding on nuclear power plants or shop in absence of an welding coordinator who can resolve welding troubles. Therefore, the integration of welding coordinators in the welding quality system of KEPIC can improve welding quality and enhance safety of construction and management of power plants. The introduction of welding coordinator requirement would put economic problems on manufactures for new employment and subsequent management works (eg. training) and field problems making authorized nuclear inspectors to be confused on inspection work scopes. Those predictable problems are expected to be minimized or eliminated through public hearings and/or seminars with regulatory body, the owner, and manufacturers and, most significantly, cooperation with related KEPIC committees. The revision draft was reviewed and discussed with personnel in nuclear industry by holding three workshop and public hearings from 2011 to 2012 and by having a presentation in 2014 KEPIC-Week. Industrial consensus on need for integration of welding coordinators in welding quality system of KEPIC was performed by reasons that it would improve welding quality, guarantee welding reliability, advance expertise, and help export to abroad. However, economic problems on manufacturers for new employment and subsequent management works, for example training, are predicted. Therefore, introduction in stages for minimizing industrial impact regarding manufacturer's scale and permission of utilizing external welding coordinator for small scale manufacturers are required. A new draft version of KEPIC-MQW (if possible, appendices of MQW) including requirements and directives for solving these economic

  19. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  20. Electric shock and electrical fire specialty

    International Nuclear Information System (INIS)

    2011-02-01

    This book deals with electric shock and electrical fire, which is made up seven chapters. It describes of special measurement for electric shock and electrical fire. It mentions concretely about electrical fire analysis and precautionary measurement, electrical shock analysis cases, occurrence of static electricity and measurement, gas accident, analysis of equipment accident and precautionary measurement. The book is published to educate the measurement on electric shock and electrical fire by electrical safety technology education center in Korea Electrical Safety Corporation.

  1. Electricity

    CERN Document Server

    Basford, Leslie

    2013-01-01

    Electricity Made Simple covers the fundamental principles underlying every aspect of electricity. The book discusses current; resistance including its measurement, Kirchhoff's laws, and resistors; electroheat, electromagnetics and electrochemistry; and the motor and generator effects of electromagnetic forces. The text also describes alternating current, circuits and inductors, alternating current circuits, and a.c. generators and motors. Other methods of generating electromagnetic forces are also considered. The book is useful for electrical engineering students.

  2. Health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California. Volume 1. Health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1977-01-01

    This report presents an overview of a project on the health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California. In addition to presenting an executive summary of the project, it sets forth the main results of the four tasks of the project: to review the health impacts (and related standards) of these forms of power generation, to review the status of standards related to plant safety (with an emphasis on nuclear power), to consider the role of the California Energy Resources Conservation and Development Commission in selection of standards, and to set forth methodologies whereby that Commission may review the health and safety aspects of proposed sites and facilities

  3. Nuclear Criticality Safety Assessment Using the SCALE Computer Code Package. A demonstration based on an independent review of a real application

    International Nuclear Information System (INIS)

    Mennerdahl, Dennis

    1998-06-01

    The purpose of this project was to instruct a young scientist from the Lithuanian Energy Institute (LEI) on how to carry out an independent review of a safety report. In particular, emphasis, was to be put on how to use the personal computer version of the calculation system SCALE 4.3 in this process. Nuclear criticality safety together with radiation shielding from gamma and neutron sources were areas of interest. This report concentrates on nuclear criticality safety aspects while a separate report covers radiation shielding. The application was a proposed storage cask for irradiated fuel assemblies from the Ignalina RBMK reactors in Lithuania. The safety report contained various documents involving many design and safety considerations. A few other documents describing the Ignalina reactors and their operation were available. The time for the project was limited to approximately one month, starting 'clean' with a SCALE 4.3 CD-ROM, a thick safety report and a fast personal computer. The results should be of general interest to Swedish authorities, in particular related to shielding where experience in using advanced computer codes like those available in SCALE is limited. It has been known for many years that criticality safety is very complicated, and that independent reviews are absolutely necessary to reduce the risk from quite common errors in the safety assessments. Several important results were obtained during the project. Concerning use of SCALE 4.3, it was confirmed that a young scientist, without extensive previous experience in the code system, can learn to use essentially all options. During the project, it was obvious that familiarity with personal computers, operating systems (including network system) and office software (word processing, spreadsheet and Internet browser software) saved a lot of time. Some of the Monte Carlo calculations took several hours. Experience is valuable in quickly picking out input or source document errors. Understanding

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

    International Nuclear Information System (INIS)

    Boyack, B.; Duffey, R.; Wilson, G.; Griffith, P.; Lellouche, G.; Levy, S.; Rohatgi, U.; Wulff, W.; Zuber, N.

    1989-12-01

    The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

  6. Analysis of Electrical Safety Conditions Taking into Account Soil Conductivity Determined on the Basis of Fuzzy Logic

    OpenAIRE

    Manusov, V.Z.; Zaytseva, N.M.

    2017-01-01

    The goal of this work is to prove a possibility of determining soil parameters that influence its conductivity being the basis of grounding, step voltage and touch voltage calculation. This in its turn increases the safety level of electric equipment operation. The article is devoted to development of new, no conventional models of soil conductivity using the theory of fuzzy sets and fuzzy logic. The description of the solution includes the following sections: fuzzy models of specific electri...

  7. 77 FR 43167 - Safety Zone; Electric Zoo Fireworks, East River, Randall's Island, NY

    Science.gov (United States)

    2012-07-24

    ... establishes a temporary safety zone on the waters of the East River in the vicinity of Randall's Island, NY.... This temporary safety zone will restrict vessels from a portion of the East River around the location... area is a temporary safety zone: all navigable waters of the East River within a 164-yard radius of the...

  8. Safety and operational aspects in in-situ electrical baking of large vacuum systems of Indus accelerators

    International Nuclear Information System (INIS)

    Bhatnagar, Prateek; Bhange, Nilesh; Joshi, Sujata; Sridhar, R.

    2016-01-01

    In order to achieve pressures in UHV (Ultra High Vacuum) range, the vacuum chambers and associated vacuum components, necessarily made of UHV compatible materials, should be baked sufficiently long enough so as to reduce outgassing rates. The baking period usually ranges from 48 hours to 72 hours for an electrical load of more than 30 kW per sector, in which baking temperatures, a characteristic parameter and specific to material, range from 150°C-180°C for Aluminium alloy and 250°C-300°C for SS (stainless steel). Indus accelerators vacuum installations include more than 300 m long vacuum chamber which are subjected to rigorous, standardized and labour intensive electrical baking procedure involving an intelligent ON-OFF distributed temperature control system with in-built defence of electrical safety and expandability as per the needs. The paper discusses various in built electrical safely features, operational aspects, work practices and challenges involved in accomplishing an interrupt free continuous electrical baking for distributed electrical load of more than 30 kW for a typical vacuum segment of Indus accelerator. (author)

  9. Towards advanced code simulators

    International Nuclear Information System (INIS)

    Scriven, A.H.

    1990-01-01

    The Central Electricity Generating Board (CEGB) uses advanced thermohydraulic codes extensively to support PWR safety analyses. A system has been developed to allow fully interactive execution of any code with graphical simulation of the operator desk and mimic display. The system operates in a virtual machine environment, with the thermohydraulic code executing in one virtual machine, communicating via interrupts with any number of other virtual machines each running other programs and graphics drivers. The driver code itself does not have to be modified from its normal batch form. Shortly following the release of RELAP5 MOD1 in IBM compatible form in 1983, this code was used as the driver for this system. When RELAP5 MOD2 became available, it was adopted with no changes needed in the basic system. Overall the system has been used for some 5 years for the analysis of LOBI tests, full scale plant studies and for simple what-if studies. For gaining rapid understanding of system dependencies it has proved invaluable. The graphical mimic system, being independent of the driver code, has also been used with other codes to study core rewetting, to replay results obtained from batch jobs on a CRAY2 computer system and to display suitably processed experimental results from the LOBI facility to aid interpretation. For the above work real-time execution was not necessary. Current work now centers on implementing the RELAP 5 code on a true parallel architecture machine. Marconi Simulation have been contracted to investigate the feasibility of using upwards of 100 processors, each capable of a peak of 30 MIPS to run a highly detailed RELAP5 model in real time, complete with specially written 3D core neutronics and balance of plant models. This paper describes the experience of using RELAP5 as an analyzer/simulator, and outlines the proposed methods and problems associated with parallel execution of RELAP5

  10. Dungeness 'A' Nuclear Power Station. The findings of NII's assessment of Nuclear Electric's long term safety review

    International Nuclear Information System (INIS)

    1994-01-01

    The assessment is reported of Nuclear Electrics' Long Term Safety Reviews (LTSR) of the Dungeness A magnox reactors. The assessment was undertaken by the Health and Safety Executive's Nuclear Installations Inspectorate (NII) which is responsible for regulating the safety of nuclear installations in the United Kingdom. This was one of a programme of LTSRs for all the UK magnox reactors. The LTSR for each plant was proceeded by a Generic Issues programme. The results of both the LTSR and the Generic Issues programme have been used by NII in forming the conclusions of this assessment. Overall the safety case for Dungeness A is satisfactory for continued operation. A programme of additional modifications and inspections has been put in place which further enhances the safety justification. Reactor operations will continue to be monitored and regulated in accordance with the inspections required under the licensing arrangements. Provided these requirements and the agreed further analysis, improvements and inspections give satisfactory results it is expected that the station will be able to operate safely till each reactor is at least 30 years old. Beyond this point a further Periodic Safety Review will be required. (UK)

  11. Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

    2010-09-01

    This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

  12. Challenge problem and milestones for: Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC)

    International Nuclear Information System (INIS)

    Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe Jr.

    2010-01-01

    This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

  13. The effects of electric power industry restructuring on the safety of nuclear power plants in the United States

    Science.gov (United States)

    Butler, Thomas S.

    Throughout the United States the electric utility industry is restructuring in response to federal legislation mandating deregulation. The electric utility industry has embarked upon an extraordinary experiment by restructuring in response to deregulation that has been advocated on the premise of improving economic efficiency by encouraging competition in as many sectors of the industry as possible. However, unlike the telephone, trucking, and airline industries, the potential effects of electric deregulation reach far beyond simple energy economics. This dissertation presents the potential safety risks involved with the deregulation of the electric power industry in the United States and abroad. The pressures of a competitive environment on utilities with nuclear power plants in their portfolio to lower operation and maintenance costs could squeeze them to resort to some risky cost-cutting measures. These include deferring maintenance, reducing training, downsizing staff, excessive reductions in refueling down time, and increasing the use of on-line maintenance. The results of this study indicate statistically significant differences at the .01 level between the safety of pressurized water reactor nuclear power plants and boiling water reactor nuclear power plants. Boiling water reactors exhibited significantly more problems than did pressurized water reactors.

  14. Results of evaluation of periodic safety review for No. 1 plant in Mihama Power Station, Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1994-01-01

    No. 1 plant in Mihama Power Station started the commercial power generation in November, 1970, and has continued the operation for more than 23 years. During this period, the counter measures to troubles, periodic inspections and the maintenance by the electric power company have been carried out. These states of No. 1 plant in Mihama Power Station for more than 23 years are to be recollected from the view-points of the comprehensive evaluation of operation experiences and the reflection of latest technological knowledge, and the safety and reliability are to be further improved in the periodic safety review. Agency of Natural Resources and Energy evaluated the report of the periodic safety review for No. 1 plant in Mihama Power Station made by Kansai Electric Power Co., and summarized the results. The course of the evaluation of the report is shown. The facility utilization factor is 43.3% on the average of about 23 years, but in the last 10 years, it was improved to 69.4%. In the last five years, the rate of occurrence of unexpected shutoff was 0.6 times/year. These are the results of preventive maintenance and the improvement of the facilities and operation management. Operation management, maintenance management, fuel management, radiation control, and radioactive waste management have been carried out properly. The work plan for preventing disasters was established, and the experience of troubles and the latest technological knowledge were well reflected to improve the safety. (K.I.)

  15. The IAEA code of conduct on the safety of radiation sources and the security of radioactive materials. A step forwards or backwards?

    International Nuclear Information System (INIS)

    Boustany, K.

    2001-01-01

    About the finalization of the Code of Conduct on the Safety and Security of radioactive Sources, it appeared that two distinct but interrelated subject areas have been identified: the prevention of accidents involving radiation sources and the prevention of theft or any other unauthorized use of radioactive materials. What analysis reveals is rather that there are gaps in both the content of the Code and the processes relating to it. Nevertheless, new standards have been introduced as a result of this exercise and have thus, as an enactment of what constitutes appropriate behaviour in the field of the safety and security of radioactive sources, emerged into the arena of international relations. (N.C.)

  16. First analysis of AGS0, LT2 and E9 CABRI tests with the new SFR safety code ASTEC-Na

    International Nuclear Information System (INIS)

    Perez-Martin, Sara; Bandini, Giacomino; Matuzas, Vaidas; Buck, Michael; Girault, Nathalie

    2015-01-01

    Within the framework of the European JASMIN project, the ASTEC-Na code is being developed for safety analysis of severe accidents in SFR. In the first phase of validation of the ASTEC-Na fuel thermo-mechanical models three in-pile tests conducted in the CABRI experimental reactor have been selected to be analysed. We present here the preliminary results of the simulation of two Transient Over Power tests and one power ramp test (AGS0, LT2 and E9, respectively) where no pin failure occurred during the transient. We present the comparison of ASTEC-Na results against experimental data and other safety code results for the initial steady state conditions prior to the transient onset as well as for the fuel pin behaviour during the transients. (author)

  17. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  18. Safety Evaluation Report related to the operation of Wm. H. Zimmer Nuclear Power Station, Unit No. 1. Docket No. 50-358. Cincinnati Gas and Electric Company

    International Nuclear Information System (INIS)

    1982-08-01

    Information is presented concerning site characteristics; design criteria for structures, systems, and components; reactor; reactor coolant system and connected systems; engineered safety features; instrumentation and controls; electric power; auxiliary systems; conduct of operations; and TMI-2 requirements

  19. Electromagnetic Field Modeling of Transcranial Electric and Magnetic Stimulation: Targeting, Individualization, and Safety of Convulsive and Subconvulsive Applications

    Science.gov (United States)

    Deng, Zhi-De

    neural stimulation strength and focality of ECT and MST. Across and within ECT studies, there is marked unexplained variability in seizure threshold and clinical outcomes. It is not known to what extent the age and sex effects on seizure threshold are mediated by interindividual variation in neural excitability and/or anatomy of the head. Addressing this question, we examine the effect on ECT and MST induced field characteristics of the variability in head diameter, scalp and skull thicknesses and conductivities, as well as brain volume, in a range of values that are representative of the patient population. Variations in the local tissue properties such as scalp and skull thickness and conductivity affect the existing ECT configurations more than MST. On the other hand, the existing MST coil configurations show greater sensitivity to head diameter variation compared to ECT. Due to the high focality of MST compared to ECT, the stimulated brain volume in MST is more sensitive to variation in tissue layer thicknesses. We further demonstrate how individualization of the stimulus pulse current amplitude, which is not presently done in ECT or MST, can be used as a means of compensating for interindividual anatomical variability, which could lead to better and more consistent clinical outcomes. Part III of the dissertation aims to systemically investigate, both computationally and experimentally, the safety of TMS and ECT in patients with a deep-brain stimulation system, and propose safety guidelines for the dual-device therapy. We showed that the induction of significant voltages in the subcutaneous leads in the scalp during TMS could result in unintended and potentially dangerous levels of electrical currents in the DBS electrode contacts. When applying ECT in patients with intracranial implants, we showed that there is an increase in the electric field strength in the brain due to conduction through the burr holes, especially when the burr holes are not fitted with

  20. 75 FR 33515 - Federal Motor Vehicle Safety Standards; Electric-Powered Vehicles; Electrolyte Spillage and...

    Science.gov (United States)

    2010-06-14

    ... instrument used for measuring electrical resistance which consists of two main elements: (1) A DC generator... physiological impacts of direct current (DC) are less than those of alternating current (AC), this rule specifies lower electrical isolation requirements for certain DC components than for AC components. The...

  1. Computational Fluid Dynamics (CFD) for Nuclear Reactor Safety Applications - Workshop Proceedings, CFD4NRS-3 - Experimental Validation and Application of CFD and CMFD Codes to Nuclear Reactor Safety Issues

    International Nuclear Information System (INIS)

    2012-01-01

    related to nuclear reactor safety issues. The conference consisted of 14 technical sessions. Among the topics included were containment, advanced reactors, multiphase flows, flow in a rod bundle, fire analysis, flows in dry casks, thermal analysis, mixing flows and pressurized thermal shock (PTS). About 1/3 of the papers were concerned with two-phase flow issues and the rest were devoted to single-phase CFD validation. South Korea is a candidate to host a follow-up meeting scheduled in 2012, organized by KAERI. KAERI also volunteered to sponsor and organize the second OECD/NEA CFD benchmark exercise. In the closure meeting after the panel session discussion, the representative from the Paul Scherrer Institut (PSI) proposed to host a future workshop scheduled for 2014, and to organize and sponsor the third OECD/NEA benchmark exercise based on a stratification experiment in the PANDA facility at PSI. The great majority of participants were interested in attending a follow-up workshop within two years. Comments were made during the panel session on the content of CFD4NRS-3. Two of the comments are that experiments can provide insight into the physics, and that CFD is now an accepted analysis tool, though it is very important to follow BPGs. There was a consensus on the need to maintain the high quality of the papers. The promotion of international benchmarking exercises for CFD was strongly encouraged. Another comment suggested that such workshops should be a forum to discuss novel approaches, but that one must also keep in mind that the end users are people from the nuclear safety community. The CFD4NRS, XCFD4NRS and CFD4NRS-3 workshops have proved to be very valuable means to assess the status of CFD code capabilities and validation, to exchange experiences in CFD code applications, and to monitor future progress

  2. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  3. Quality assurance for safety in nuclear power plants and other nuclear installations. Code and safety guides Q1-Q14. A publication within the NUSS programme

    International Nuclear Information System (INIS)

    1996-01-01

    The code provides the basic requirements for establishing and implementing quality assurance programmes for the stages of siting, design, construction, commissioning, operation and decommissioning of nuclear power plants. These basic requirements apply to all individuals and organizations, including designers, suppliers, constructors, manufacturers and operators. The basic quality assurance requirements presented in this Code also apply, with appropriate modifications, to nuclear installations other than nuclear power plants

  4. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  5. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

    International Nuclear Information System (INIS)

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  6. Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

  7. Code of Conduct on the Safety and Security of Radioactive Sources and the Supplementary Guidance on the Import and Export of Radioactive Sources

    International Nuclear Information System (INIS)

    2005-01-01

    In operative paragraph 4 of its resolution GC(47)/RES/7.B, the General Conference, having welcomed the approval by the Board of Governors of the revised IAEA Code of Conduct on the Safety and Security of Radioactive Sources (GC(47)/9), and while recognizing that the Code is not a legally binding instrument, urged each State to write to the Director General that it fully supports and endorses the IAEA's efforts to enhance the safety and security of radioactive sources and is working toward following the guidance contained in the IAEA Code of Conduct. In operative paragraph 5, the Director General was requested to compile, maintain and publish a list of States that have made such a political commitment. The General Conference, in operative paragraph 6, recognized that this procedure 'is an exceptional one, having no legal force and only intended for information, and therefore does not constitute a precedent applicable to other Codes of Conduct of the Agency or of other bodies belonging to the United Nations system'. In operative paragraph 7 of resolution GC(48)/RES/10.D, the General Conference welcomed the fact that more than 60 States had made political commitments with respect to the Code in line with resolution GC(47)/RES/7.B and encouraged other States to do so. In operative paragraph 8 of resolution GC(48)/RES/10.D, the General Conference further welcomed the approval by the Board of Governors of the Supplementary Guidance on the Import and Export of Radioactive Sources (GC(48)/13), endorsed this Guidance while recognizing that it is not legally binding, noted that more than 30 countries had made clear their intention to work towards effective import and export controls by 31 December 2005, and encouraged States to act in accordance with the Guidance on a harmonized basis and to notify the Director General of their intention to do so as supplementary information to the Code of Conduct, recalling operative paragraph 6 of resolution GC(47)/RES/7.B. 4. The

  8. Practical experience with software tools to assess and improve the quality of existing nuclear analysis and safety codes

    International Nuclear Information System (INIS)

    Marshall, N.H.; Marwil, E.S.; Matthews, S.D.; Stacey, B.J.

    1990-01-01

    Within the constraints of schedule and budget, software tools and techniques were applied to existing FORTRAN codes determining software quality metrics and improving the code quality. Specifically discussed are INEL experiences in applying pretty printers, cross-reference analyzers, and computer aided software engineering (CASE) tools and techniques. These have provided management with measures of the risk potential for individual program modules so that rational decisions can be made on resource allocation. Selected program modules have been modified to reduce the complexity, achieve higher functional independence, and improve the code vectorization. (orig.)

  9. Literature review of environmental qualification of safety-related electric cables: Summary of past work. Volume 1

    International Nuclear Information System (INIS)

    Subudhi, M.

    1996-04-01

    This report summarizes the findings from a review of published documents dealing with research on the environmental qualification of safety-related electric cables used in nuclear power plants. Simulations of accelerated aging and accident conditions are important considerations in qualifying the cables. Significant research in these two areas has been performed in the US and abroad. The results from studies in France, Germany, and Japan are described in this report. In recent years, the development of methods to monitor the condition of cables has received special attention. Tests involving chemical and physical examination of cable's insulation and jacket materials, and electrical measurements of the insulation properties of cables are discussed. Although there have been significant advances in many areas, there is no single method which can provide the necessary information about the condition of a cable currently in service. However, it is possible that further research may identify a combination of several methods that can adequately characterize the cable's condition

  10. The Development of the Safety Related Valve Class 1E Electrical Motor, the Target and the Results

    International Nuclear Information System (INIS)

    Saban, I.; Grgic, D.; Fancev, T.; Flegar, Lj.; Novosel, N.

    1996-01-01

    The development of the safety related valves class 1E electric motor is described. The design implemented in order to satisfy the 1E requirements, and a way in which related 1E standards are addressed, are shown. The development was realized in three stages. In the first stage eight motorettes were made and the insulation system was tested. In the second stage the motor was produced in accordance with producer's prototype QA program. In the third stage part of the testing of the produced motor was made. The results of the testing, finished until now, show that produced motor, as well as similarly produced motors, is able to perform its safety function in the design bases accident conditions as requested by class 1E requirements. The rest of the testing (LOCA test) can be made on the same or similar motor in the future. (author)

  11. Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Unit 2 (Docket No. 50-446)

    International Nuclear Information System (INIS)

    1992-09-01

    This document supplement 25 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Unit 2 (NUREG-0797), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement reports the status of certain issues that had not been resolved when the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, 12, 21, 22, 23, and 24 to that report were published. This supplement deals primarily with Unit 2 issues; however, it also references evaluations for several Unit 1 licensing items resolved since Supplement 24 was issued

  12. Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Unit 2 (Docket No. 50-446)

    International Nuclear Information System (INIS)

    1993-02-01

    Supplement 26 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Unit 2, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement reports the status of certain issues that had not been resolved when the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, 12, 21, 22, 23, 24, and 25 to that report were published. This supplement deals primarily with Unit 2 issues; however, it also references evaluations for several licensing issues that relate to Unit 1, which have been resolved since Supplement 25 was issued

  13. Safety verification for the ECCS driven by the electrically 4 trains during LBLOCA reflood phase using ATLAS

    International Nuclear Information System (INIS)

    Park, Yusun; Park, Hyun-sik; Kang, Kyoung-ho; Choi, Nam-hyun; Min, Kyoung-ho; Choi, Ki-yong

    2014-01-01

    Highlights: • Safety improvement by adopting 4 train emergency core cooling system was validated experimentally. • General thermal hydraulic behaviors of the system during LBLOCA reflood phase were successfully demonstrated. • Key parameters such as the liquid levels, the PCTs, the quenching time, and the ECC bypass ratios were investigated. • Asymmetric effects of the different combination of safety injection were negligible during the reflood period. - Abstract: The APR1400 is equipped with four safety injection pumps driven by two emergency diesel generators. However, the design has been changed so that the four safety injection pumps are driven by 4 emergency diesel generators during the design certification process from the U.S. NRC. Thus, 4 safety injection pumps (SIPs) are completely independent electrically and mechanically and three safety injection pumps are available in a single failure condition. This design change could have a certain effects on the thermal-hydraulic phenomenon occurring in the downcomer region during the late reflood phase of a large break loss of coolant accident (LBLOCA). Thus, in this study, a verification experiment for the reflood phase of a LBLOCA was performed to evaluate the core cooling performance of the 4 train emergency core cooling system (ECCS) with an assumption of a single failure. And the different combinations of three SIPs positions were tested to investigate the asymmetric effects on the reactor core cooling performance. The overall experimental results revealed the typical thermal–hydraulic trends expected to occur during the reflood phase of a large-break LOCA scenario for the APR1400. Experiment with the injection of three SIPs showed a faster core quenching time and lower bypass ratio than that of the case in which two SIPs were injected. The RPV wall temperature distributions showed the similar trend in spite of the different SIP combinations

  14. Safety

    International Nuclear Information System (INIS)

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  15. Safety Evaluation Report related to the operation of Vogtle Electric Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 50-425)

    International Nuclear Information System (INIS)

    1987-03-01

    In June 1985, the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-1137) regarding the application of Georgia Power Company, Municipal Electric Authority of Georgia, Oglethorpe Power Corporation, and the City of Dalton, Georgia, for licenses to operate the Vogtle Electric Generating Plant, Units 1 and 2 (Docket Nos. 50-424 and 50-425). This sixth supplement of NUREG-1137 provides recent information regarding resolution of some of the open and confirmatory items that remained unresolved at the time the Safety Evaluation Report was issued. These areas are performance testing, reactor cooling hydraulics, loose parts monitoring, and electric power systems

  16. Development and verification of a leningrad NPP unit 1 living PSA model in the INL SAPHIRE code format for prompt operational safety level monitoring

    International Nuclear Information System (INIS)

    Bronislav, Vinnikov

    2007-01-01

    The first part of the paper presents results of the work, that was carried out in complete conformity with the Technical Assignment, which was developed by the Leningrad Nuclear Power Plant. The initial scientific and technical information, contained into the In-Depth Safety Assessment Reports, was given to the author of the work. This information included graphical Fault Trees of Safety Systems and Auxiliary Technical Systems, Event Trees for the necessary number of Initial Events, and also information about failure probabilities of basic components of the nuclear unit. On the basis of this information and fueling it to the Usa Idaho National Laboratory (INL) SAPHIRE code, we have developed an electronic version of the Data Base for failure probabilities of the components of technical systems. Then, we have developed both the electronic versions of the necessary Fault Trees, and an electronic versions of the necessary Event Trees. And at last, we have carried out the linkage of the Event Trees. This work has resulted in the Living PSA (LPSA - Living Probabilistic Safety Assessment) Model of the Leningrad NPP Unit 1. The LPSA-model is completely adapted to be consistent with the USA INL SAPHIRE Risk Monitor. The second part of the paper results in analysis of fire consequences in various places of Leningrad NPP Unit 1. The computations were carried out with the help of the LPSA-model, developed in SAPHIRE code format. On the basis of the computations the order of priority of implementation of fire prevention measures was established. (author)

  17. Implementation of an enlarged model of the safety valves and relief in the plant integral model for the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Amador G, R.; Ortiz V, J.; Castillo D, R.; Hernandez L, E. J.; Galeana R, J. C.; Gutierrez, V. H.

    2013-10-01

    The present work refers to the implementation of a new model on the logic of the safety valves and relief in the integral model of the Nuclear Power Plant of Laguna Verde of the thermal-hydraulic compute code RELAP/SCDAPSIM Mod. 3.4. The new model was developed with the compute package SIMULINK-MATLAB and contemplates all the operation options of the safety valves and relief, besides including the availability options of the valves in all the operation ways and of blockage in the ways of relief and low-low. The implementation means the elimination of the old model of the safety valves and to analyze the group of logical variables, of discharge and available control systems to associate them to the model of package SIMULINK-MATLAB. The implementation has been practically transparent and 27 cases corresponding to a turbine discharge were analyzed with the code RELAP/SCDAPSIM Mod. 3.4. The results were satisfactory. (Author)

  18. Electricity

    International Nuclear Information System (INIS)

    Tombs, F.

    1983-01-01

    The subject is discussed, with particular reference to the electricity industry in the United Kingdom, under the headings; importance and scope of the industry's work; future fuel supplies (estimated indigenous fossil fuels reserves); outlook for UK energy supplies; problems of future generating capacity and fuel mix (energy policy; construction programme; economics and pricing; contribution of nuclear power - thermal and fast reactors; problems of conversion of oil-burning to coal-burning plant). (U.K.)

  19. Investigating the burning characteristics of electric cables used in the nuclear power plant by way of 3-D transient FDS code

    Energy Technology Data Exchange (ETDEWEB)

    Ferng, Y.M., E-mail: ymferng@ess.nthu.edu.t [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Liu, C.H. [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)

    2011-01-15

    Burning characteristics of electrical cables are one of the key parameters for the fire hazard assessment of nuclear power plants (NPPs) since the cables are the essential sources of fire in the plants. A three-dimensional (3-D) transient computational fluid dynamics (CFD) code{sub F}DS is adopted in this paper to simulate these characteristics related to the cable burning. Being one of the NRC licensing fire codes, the FDS includes the thermal-hydraulic equations, the turbulence model and the chemical combustion model, etc. In order to assess the CFD fire models used in this code, a burning test using the control cable with the outer jacket of polyvinylchloride (PVC) and the inner insulation of cross-linked polyethylene (XLPE) is conducted. The measured parameters associated with the burning characteristics include the heat release rate (HRR), O{sub 2} depletion, and CO and CO{sub 2} production, etc. Except the amount of O{sub 2} consumption, the predicted transient behaviors of other parameters can reproduce the measured data. Based on the chemical combustion model in the FDS code, this discrepancy may be essentially resulted from the default value of hydrogen fraction (H{sub frac}) contained in the soot since the soot yield for the burning of PVC material is high enough that the uncertainty in the H{sub frac} value has a prominent effect on the amount of O{sub 2} consumption. This explanation can be confirmed by a benchmark calculation for simulating a burning test with the polymethylmethacrylate (PMMA) fuel of low-soot yield. The present simulation works can provide the useful information for the plant staff or the researcher as they would perform the fire hazard analysis in the NPPs using the FDS code.

  20. Evaluation of practicality of ASME code, Section XI

    International Nuclear Information System (INIS)

    Mattu, R.K.; Lauderdale, J.R.; Liu, S.N.; Lance, J.J.

    2004-01-01

    Many nuclear power plants have found that it is impractical or unduly burdensome to comply with some ASME Boiler and Pressure Code provisions and have sought relief from those provisions from the Nuclear Regulatory Commission. An Electric Power Research Institute (EPRI) project is evaluating such Code provisions and alternatives to them that will meet the safety intent of the Code with less burden on utilities. The methodology is to extract data from an on-line data base of relief requests since 1980, analyse the data to identify burdensome provisions for which there are satisfactory alternatives, and recommend changes in the Code to the ASME. (author)