Nuclear criticality safety handbook. Version 2

International Nuclear Information System (INIS)

1999-03-01

The Nuclear Criticality Safety Handbook, Version 2 essentially includes the description of the Supplement Report to the Nuclear Criticality Safety Handbook, released in 1995, into the first version of Nuclear Criticality Safety Handbook, published in 1988. The following two points are new: (1) exemplifying safety margins related to modelled dissolution and extraction processes, (2) describing evaluation methods and alarm system for criticality accidents. Revision is made based on previous studies for the chapter that treats modelling the fuel system: e.g., the fuel grain size that the system can be regarded as homogeneous, non-uniformity effect of fuel solution, and burnup credit. This revision solves the inconsistencies found in the first version between the evaluation of errors found in JACS code system and criticality condition data that were calculated based on the evaluation. (author)

Nuclear criticality safety parameter evaluation for uranium metallic alloy

Energy Technology Data Exchange (ETDEWEB)

Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

2013-07-01

Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

SIMCRI: a simple computer code for calculating nuclear criticality parameters

International Nuclear Information System (INIS)

Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.

1986-03-01

This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)

A PC version of the Monte Carlo criticality code OMEGA

International Nuclear Information System (INIS)

Seifert, E.

1996-05-01

A description of the PC version of the Monte Carlo criticality code OMEGA is given. The report contains a general description of the code together with a detailed input description. Furthermore, some examples are given illustrating the generation of an input file. The main field of application is the calculation of the criticality of arrangements of fissionable material. Geometrically complicated arrangements that often appear inside and outside a reactor, e.g. in a fuel storage or transport container, can be considered essentially without geometrical approximations. For example, the real geometry of assemblies containing hexagonal or square lattice structures can be described in full detail. Moreover, the code can be used for special investigations in the field of reactor physics and neutron transport. Many years of practical experience and comparison with reference cases have shown that the code together with the built-in data libraries gives reliable results. OMEGA is completely independent on other widely used criticality codes (KENO, MCNP, etc.), concerning programming and the data base. It is a good practice to run difficult criticality safety problems by different independent codes in order to mutually verify the results. In this way, OMEGA can be used as a redundant code within the family of criticality codes. An advantage of OMEGA is the short calculation time: A typical criticality safety application takes only a few minutes on a Pentium PC. Therefore, the influence of parameter variations can simply be investigated by running many variants of a problem. (orig.)

Criticality safety engineer training at WSRC

International Nuclear Information System (INIS)

Williamson, T.G.; Mincey, J.F.

1993-01-01

Two programs designed to prepare engineers for certification as criticality safety engineers are offered at Westinghouse Savannah River Company (WSRC). One program, Student On Loan Criticality Engineer Training (SOLCET), is an intensive 2-yr course involving lectures, rigorous problem assignments, and mentoring. The other program, In-Field Criticality Engineer Training (IN-FIELD), is a less intensive series of lectures and problem assignments. Both courses are conducted by members of the Applied Physics Group (APG) of the Savannah River Technical Center, the organization at WSRC responsible for the operation and maintenance of criticality codes and for training of code users

The International Criticality Safety Benchmark Evaluation Project

International Nuclear Information System (INIS)

Briggs, B. J.; Dean, V. F.; Pesic, M. P.

2001-01-01

experimenters or individuals who are familiar with the experimenters or the experimental facility; (3) compile the data into a standardized format; (4) perform calculations of each experiment with standard criticality safety codes, and (5) formally document the work into a single source of verified benchmark critical data. The work of the ICSBEP is documented as an OECD handbook, in 7 volumes, entitled, 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. This handbook is available on CD-ROM or on the Internet (http://icsbep.inel.gov/icsbep). Over 150 scientists from around the world have combined their efforts to produce this Handbook. The 2000 publication of the handbook will span over 19,000 pages and contain benchmark specifications for approximately 284 evaluations containing 2352 critical configurations. The handbook is currently in use in 45 different countries by criticality safety analysts to perform necessary validation of their calculation techniques and it is expected to be a valuable tool for decades to come. As a result of the efforts of the ICSBEP: (1) a large portion of the tedious, redundant, and very costly research and processing of criticality safety experimental data has been eliminated; (2) the necessary step in criticality safety analyses of validating computer codes with benchmark data is greatly streamlined; (3) gaps in data are being highlighted; (4) lost data are being retrieved; (5) deficiencies and errors in cross section processing codes and neutronic codes are being identified, and (6) over a half-century of valuable criticality safety data are being preserved. (author)

Criticality safety studies at VTT Energy

International Nuclear Information System (INIS)

Roine, T.; Anttila, M.

1995-01-01

At VTT Energy a compact reactor physics calculation system is applied in many kind of problems. Generation of group constants for static and dynamic core calculations, flux and dose rate calculations as well as criticality safety studies are performed basically with the same codes. In the presentation a short overview of the wide variety of criticality safety problems analyzed at VTT Energy is given. The calculation system with some illustrative examples is also described. (12 refs., 1 tab.)

Maintaining scale as a realiable computational system for criticality safety analysis

International Nuclear Information System (INIS)

Bowmann, S.M.; Parks, C.V.; Martin, S.K.

1995-01-01

Accurate and reliable computational methods are essential for nuclear criticality safety analyses. The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer code system was originally developed at Oak Ridge National Laboratory (ORNL) to enable users to easily set up and perform criticality safety analyses, as well as shielding, depletion, and heat transfer analyses. Over the fifteen-year life of SCALE, the mainstay of the system has been the criticality safety analysis sequences that have featured the KENO-IV and KENO-V.A Monte Carlo codes and the XSDRNPM one-dimensional discrete-ordinates code. The criticality safety analysis sequences provide automated material and problem-dependent resonance processing for each criticality calculation. This report details configuration management which is essential because SCALE consists of more than 25 computer codes (referred to as modules) that share libraries of commonly used subroutines. Changes to a single subroutine in some cases affect almost every module in SCALE exclamation point Controlled access to program source and executables and accurate documentation of modifications are essential to maintaining SCALE as a reliable code system. The modules and subroutine libraries in SCALE are programmed by a staff of approximately ten Code Managers. The SCALE Software Coordinator maintains the SCALE system and is the only person who modifies the production source, executables, and data libraries. All modifications must be authorized by the SCALE Project Leader prior to implementation

New enhancements to SCALE for criticality safety analysis

International Nuclear Information System (INIS)

Hollenbach, D.F.; Bowman, S.M.; Petrie, L.M.; Parks, C.V.

1995-01-01

As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portability of the SCALE code package, and (9) other minor enhancements, modifications, and corrections to SCALE. Each of these additions and enhancements to the criticality safety analysis capabilities of the SCALE code system are discussed below

Tank farms criticality safety manual

International Nuclear Information System (INIS)

FORT, L.A.

2003-01-01

This document defines the Tank Farms Contractor (TFC) criticality safety program, as required by Title 10 Code of Federal Regulations (CFR-), Subpart 830.204(b)(6), ''Documented Safety Analysis'' (10 CFR- 830.204 (b)(6)), and US Department of Energy (DOE) 0 420.1A, Facility Safety, Section 4.3, ''Criticality Safety.'' In addition, this document contains certain best management practices, adopted by TFC management based on successful Hanford Site facility practices. Requirements in this manual are based on the contractor requirements document (CRD) found in Attachment 2 of DOE 0 420.1A, Section 4.3, ''Nuclear Criticality Safety,'' and the cited revisions of applicable standards published jointly by the American National Standards Institute (ANSI) and the American Nuclear Society (ANS) as listed in Appendix A. As an informational device, requirements directly imposed by the CRD or ANSI/ANS Standards are shown in boldface. Requirements developed as best management practices through experience and maintained consistent with Hanford Site practice are shown in italics. Recommendations and explanatory material are provided in plain type

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)

Computational methods for criticality safety analysis within the scale system

International Nuclear Information System (INIS)

Parks, C.V.; Petrie, L.M.; Landers, N.F.; Bucholz, J.A.

1986-01-01

The criticality safety analysis capabilities within the SCALE system are centered around the Monte Carlo codes KENO IV and KENO V.a, which are both included in SCALE as functional modules. The XSDRNPM-S module is also an important tool within SCALE for obtaining multiplication factors for one-dimensional system models. This paper reviews the features and modeling capabilities of these codes along with their implementation within the Criticality Safety Analysis Sequences (CSAS) of SCALE. The CSAS modules provide automated cross-section processing and user-friendly input that allow criticality safety analyses to be done in an efficient and accurate manner. 14 refs., 2 figs., 3 tabs

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

Accomplishment of 10-year research in NUCEF and future development. Criticality safety research

International Nuclear Information System (INIS)

Miyoshi, Yoshinori

2005-01-01

Since 1995, static and transient critical experiments on low enriched uranyl nitrate solution have been performed using two solution type criticality facilities, STACY and TRACY constructed in NUCEF. The obtained fundamental and systematic data on aqueous solution were used to validate the criticality safety calculation codes and to develop the transient analyses codes for criticality accident evaluation. This paper describes the outline of the criticality safety research conducted in NUCEF. (author)

Development of the criticality accident analysis code, AGNES

International Nuclear Information System (INIS)

Nakajima, Ken

1989-01-01

In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

SCALE 5: Powerful new criticality safety analysis tools

International Nuclear Information System (INIS)

Bowman, Stephen M.; Hollenbach, Daniel F.; Dehart, Mark D.; Rearden, Bradley T.; Gauld, Ian C.; Goluoglu, Sedat

2003-01-01

Version 5 of the SCALE computer software system developed at Oak Ridge National Laboratory, scheduled for release in December 2003, contains several significant new modules and sequences for criticality safety analysis and marks the most important update to SCALE in more than a decade. This paper highlights the capabilities of these new modules and sequences, including continuous energy flux spectra for processing multigroup problem-dependent cross sections; one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations; two-dimensional flexible mesh discrete ordinates code; automated burnup-credit analysis sequence; and one-dimensional material distribution optimization for criticality safety. (author)

Nuclear Criticality Safety Handbook, Version 2. English translation

International Nuclear Information System (INIS)

2001-08-01

The Nuclear Criticality Safety Handbook, Version 2 essentially includes the description of the Supplement Report to the Nuclear Criticality Safety Handbook, released in 1995, into the first version of the Nuclear Criticality Safety Handbook, published in 1988. The following two points are new: (1) exemplifying safety margins related to modeled dissolution and extraction processes, (2) describing evaluation methods and alarm system for criticality accidents. Revision has been made based on previous studies for the chapter that treats modeling the fuel system: e.g., the fuel grain size that the system can be regarded as homogeneous, non-uniformity effect of fuel solution, an burnup credit. This revision has solved the inconsistencies found in the first version between the evaluation of errors found in JACS code system and the criticality condition data that were calculated based on the evaluation. This report is an English translation of the Nuclear Criticality Safety Handbook, Version 2, originally published in Japanese as JAERI 1340 in 1999. (author)

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

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

Development of INCTAC code for analyzing criticality accident phenomena

International Nuclear Information System (INIS)

Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

2003-01-01

Aiming at understanding nuclear transients and thermal- and hydraulic-phenomena of the criticality accident, a code named INCTAC has been newly developed at the Institute of Nuclear Safety. The code is applicable to the analysis of criticality accident transients of aqueous homogenous fuel solution system. Neutronic transient model is composed of equations for the kinetics and for the spatial distributions, which are deduced from the time dependent multi-group transport equations with the quasi steady state assumption. Thermal-hydraulic transient model is composed of a complete set of the mass, momentum and energy equations together with the two-phase flow assumptions. Validation tests of INCTAC were made using the data obtained at TRACY, a transient experiment criticality facility of JAERI. The calculated results with INCTAC showed a very good agreement with the experiment data, except a slight discrepancy of the time when the peak of reactor power was attained. But, the discrepancy was resolved with the use of an adequate model for movement and transfer of the void in the fuel solution mostly generated by radiolysis. With a simulation model for the transport of radioactive materials through ventilation systems to the environment, INCTAC will be used as an overall safety evaluation code of the criticality accident. (author)

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

Qualification of safety-critical software for digital reactor safety system in nuclear power plants

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo

2013-01-01

This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)

Regulatory considerations for computational requirements for nuclear criticality safety

International Nuclear Information System (INIS)

Bidinger, G.H.

1995-01-01

As part of its safety mission, the U.S. Nuclear Regulatory Commission (NRC) approves the use of computational methods as part of the demonstration of nuclear criticality safety. While each NRC office has different criteria for accepting computational methods for nuclear criticality safety results, the Office of Nuclear Materials Safety and Safeguards (NMSS) approves the use of specific computational methods and methodologies for nuclear criticality safety analyses by specific companies (licensees or consultants). By contrast, the Office of Nuclear Reactor Regulation approves codes for general use. Historically, computational methods progressed from empirical methods to one-dimensional diffusion and discrete ordinates transport calculations and then to three-dimensional Monte Carlo transport calculations. With the advent of faster computational ability, three-dimensional diffusion and discrete ordinates transport calculations are gaining favor. With the proper user controls, NMSS has accepted any and all of these methods for demonstrations of nuclear criticality safety

Proceedings of the first annual Nuclear Criticality Safety Technology Project

International Nuclear Information System (INIS)

Rutherford, D.A.

1994-09-01

This document represents the published proceedings of the first annual Nuclear Criticality Safety Technology Project (NCSTP) Workshop, which took place May 12--14, 1992, in Gaithersburg, Md. The conference consisted of four sessions, each dealing with a specific aspect of nuclear criticality safety issues. The session titles were ''Criticality Code Development, Usage, and Validation,'' ''Experimental Needs, Facilities, and Measurements,'' ''Regulation, Compliance, and Their Effects on Nuclear Criticality Technology and Safety,'' and ''The Nuclear Criticality Community Response to the USDOE Regulations and Compliance Directives.'' The conference also sponsored a Working Group session, a report of the NCSTP Working Group is also presented. Individual papers have been cataloged separately

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

International Nuclear Information System (INIS)

Mitake, Susumu

2003-01-01

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions

Request from nuclear fuel cycle and criticality safety design

International Nuclear Information System (INIS)

Hamasaki, Manabu; Sakashita, Kiichiro; Natsume, Toshihiro

2005-01-01

The quality and reliability of criticality safety design of nuclear fuel cycle systems such as fuel fabrication facilities, fuel reprocessing facilities, storage systems of various forms of nuclear materials or transportation casks have been largely dependent on the quality of criticality safety analyses using qualified criticality calculation code systems and reliable nuclear data sets. In this report, we summarize the characteristics of the nuclear fuel cycle systems and the perspective of the requirements for the nuclear data, with brief comments on the recent issue about spent fuel disposal. (author)

Uncertainties associated with the use of the KENO Monte Carlo criticality codes

International Nuclear Information System (INIS)

Landers, N.F.; Petrie, L.M.

1989-01-01

The KENO multi-group Monte Carlo criticality codes have earned the reputation of being efficient, user friendly tools especially suited for the analysis of situations commonly encountered in the storage and transportation of fissile materials. Throughout their twenty years of service, a continuing effort has been made to maintain and improve these codes to meet the needs of the nuclear criticality safety community. Foremost among these needs is the knowledge of how to utilize the results safely and effectively. Therefore it is important that code users be aware of uncertainties that may affect their results. These uncertainties originate from approximations in the problem data, methods used to process cross sections, and assumptions, limitations and approximations within the criticality computer code itself. 6 refs., 8 figs., 1 tab

Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed

International Nuclear Information System (INIS)

1984-01-01

The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.

Criticality safety validation of MCNP5 using continuous energy libraries

International Nuclear Information System (INIS)

Salome, Jean A.D.; Pereira, Claubia; Assuncao, Jonathan B.A.; Veloso, Maria Auxiliadora F.; Costa, Antonella L.; Silva, Clarysson A.M. da

2013-01-01

The study of subcritical systems is very important in the design, installation and operation of various devices, mainly nuclear reactors and power plants. The information generated by these systems guide the decisions to be taken in the executive project, the economic viability and the safety measures to be employed in a nuclear facility. Simulating some experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments, the code MCNP5 was validated to nuclear criticality analysis. Its continuous libraries were used. The average values and standard deviation (SD) were evaluated. The results obtained with the code are very similar to the values obtained by the benchmark experiments. (author)

Criticality safety analysis of Hanford Waste Tank 241-101-SY

International Nuclear Information System (INIS)

Perry, R.T.; Sapir, J.L.; Krohn, B.J.

1993-01-01

As part of a safety assessment for proposed pump mixing operations to mitigate episodic gas releases in Tank 241-101-SY at the Hanford Site, Richland, Washington, a criticality safety analysis was made using the Sn transport code ONEDANT. The tank contains approximately one million gallons of waste and an estimated 910 G of plutonium. the criticality analysis considers reconfiguration and underestimation of plutonium content. The results indicate that Tank SY-101 does not present a criticality hazard. These methods are also used in criticality analyses of other Hanford tanks

USNRC licensing process as related to nuclear criticality safety

International Nuclear Information System (INIS)

Ketzlach, N.

1987-01-01

The U.S. Code of Federal Regulations establishes procedures and criteria for the issuance of licenses to receive title to, own, acquire, deliver, receive, possess, use, and initially transfer special nuclear material; and establishes and provides for the terms and conditions upon which the Nuclear Regulatory Commission (NRC) will issue such licenses. Section 70.22 of the regulations, ''Contents of Applications'', requires that applications for licenses contain proposed procedures to avoid accidental conditions of criticality. These procedures are elements of a nuclear criticality safety program for operations with fissionable materials at fuels and materials facilities (i.e., fuel cycle facilities other than nuclear reactors) in which there exists a potential for criticality accidents. To assist the applicant in providing specific information needed for a nuclear criticality safety program in a license application, the NRC has issued regulatory guides. The NRC requirements for nuclear criticality safety include organizational, administrative, and technical requirements. For purely technical matters on nuclear criticality safety these guides endorse national standards. Others provide guidance on the standard format and content of license applications, guidance on evaluating radiological consequences of criticality accidents, or guidance for dealing with other radiation safety issues. (author)

Design aspects of safety critical instrumentation of nuclear installations

Energy Technology Data Exchange (ETDEWEB)

Swaminathan, P. [Electronics Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)]. E-mail: swamy@igcar.ernet.in

2005-07-01

Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

Design aspects of safety critical instrumentation of nuclear installations

International Nuclear Information System (INIS)

Swaminathan, P.

2005-01-01

Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

The International Criticality Safety Benchmark Evaluation Project (ICSBEP)

International Nuclear Information System (INIS)

Briggs, J.B.

2003-01-01

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organisation for Economic Cooperation and Development (OECD) - Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Israel, Spain, and Brazil are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled 'International Handbook of Evaluated Criticality Safety Benchmark Experiments.' The 2003 Edition of the Handbook contains benchmark model specifications for 3070 critical or subcritical configurations that are intended for validating computer codes that calculate effective neutron multiplication and for testing basic nuclear data. (author)

New developments enhancing MCNP for criticality safety

International Nuclear Information System (INIS)

Hendricks, J.S.; McKinney, G.W.; Forster, R.A.

1993-01-01

Since the early 80's MCNP has had three estimates of k eff : collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime. These are calculated for every cycle and are averaged over the cycles as simple averages and covariance weighted averages. Correlation coefficients between estimators are also calculated. These criticality estimators are all in addition to the extensive summary information and tally edits used in shielding and other problems. A number of significant new developments have been made to enhance the MCNP Monte Carlo radiation transport code for criticality safety applications. These are available in the newly released MCNP4A version of the code

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)

Criticality safety assessment of FBTR fuel sub-assemblies using WIMS cross section set

International Nuclear Information System (INIS)

Gupta, H.C.; Chakraborty, B.

2002-01-01

Full text: FBTR's irradiated fuel sub-assemblies (FSAs) are sent to RML at Indira Gandhi Centre for Atomic Research for post irradiation examination. The FSAs are cut open and the fuel pins are separated for examination in the hot cells. It was required to evaluate the criticality safety in handling the FSAs in the hot cells. Criticality safety studies for handling two as well as three irradiated FSAs in the hot cells under dry conditions were carried out by the Safety Group at IGCAR, Kalpakkam. Monte Carlo code KENO (Version Va) which uses 16-group Hansen-Roach cross-section set was used for the calculations. Subsequently, during the safety review of the proposition by the Safety Review Committee (SARCOP) of AERB, it was stipulated to carry out the criticality safety studies under flooded condition also. We carried out the criticality safety studies for these fuel sub assemblies in different configurations under dry (buried in concrete) as well as wet condition (flooded with light water) using Monte Carlo codes MONALI (developed at BARC) and KENO4 using WlMS-69 group cross section set. Results of our analyses under various conditions are presented in this paper

Criticality safety analysis for plutonium dissolver using silver mediated electrolytic oxidation method

International Nuclear Information System (INIS)

Umeda, Miki; Sugikawa, Susumu; Nakamura, Kazuhito; Egashira, Tetsurou

1998-08-01

Design and construction of a plutonium dissolver using silver mediated electrolytic oxidation method are promoted in NUCEF. Criticality safety analysis for the plutonium dissolver is described in this report. The electrolytic plutonium dissolver consists of connection pipes and three pots for MOX powder supply, circulation and electrolysis. The criticality control for the dissolver is made by geometrically safe shape with mass limitation. Monte Carlo code KENO-IV using MGCL-137 library based on ENDF/B-IV was used for the criticality safety analysis for the plutonium dissolver. Considering the required size for construction and criticality safety, diameter of pot and distance between two pots were determined. On this condition, the criticality safety analysis for the plutonium dissolver with connection pipes was carried out. As the result of the criticality safety analysis, an effective neutron multiplication factor keff of 0.91 was obtained and the criticality safety of the plutonium dissolver was confirmed on the basis of criteria of ≤0.95. (author)

ACRR fuel storage racks criticality safety analysis

International Nuclear Information System (INIS)

Bodette, D.E.; Naegeli, R.E.

1997-10-01

This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-11

Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C_k's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.

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

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

International Nuclear Information System (INIS)

Kang, H-S; Jang, M-S; Kim, S-R; Park, J-M; Kim, K-N

2015-01-01

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

Energy Technology Data Exchange (ETDEWEB)

Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.

Verification of criticality Safety for ETRR-2 Fuel Manufacturing pilot Plant (FMPP) at Inshas

International Nuclear Information System (INIS)

Aziz, M.; Gadalla, A.A.; Orabi, G.

2006-01-01

The criticality safety of the fuel manufacturing pilot plant (FMPP) at inshas is studied and analyzed during normal and abnormal operation conditions. the multiplication factor during all stages of the manufacturing processes is determined. several accident scenarios were simulated and the criticality of these accidents were investigated. two codes are used in the analysis : MCNP 4 B code, based on monte Carlo method, and CITATION code , based on diffusion theory. the results are compared with the designer calculations and satisfactory agreement were found. the results of the study indicated that the safety of the fuel manufacturing pilot plant is confirmed

ICSBEP-2007, International Criticality Safety Benchmark Experiment Handbook

International Nuclear Information System (INIS)

Blair Briggs, J.

2007-01-01

1 - Description: The Critically Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United Sates Department of Energy. The project quickly became an international effort as scientist from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization of Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA). This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material. The example calculations presented do not constitute a validation of the codes or cross section data. The work of the ICSBEP is documented as an International Handbook of Evaluated Criticality Safety Benchmark Experiments. Currently, the handbook spans over 42,000 pages and contains 464 evaluations representing 4,092 critical, near-critical, or subcritical configurations and 21 criticality alarm placement/shielding configurations with multiple dose points for each and 46 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. The handbook is intended for use by criticality safety analysts to perform necessary validations of their calculational techniques and is expected to be a valuable tool for decades to come. The ICSBEP Handbook is available on DVD. You may request a DVD by completing the DVD Request Form on the internet. Access to the Handbook on the Internet requires a password. You may request a password by completing the Password Request Form. The Web address is: http://icsbep.inel.gov/handbook.shtml 2 - Method of solution: Experiments that are found

A software engineering process for safety-critical software application

International Nuclear Information System (INIS)

Kang, Byung Heon; Kim, Hang Bae; Chang, Hoon Seon; Jeon, Jong Sun

1995-01-01

Application of computer software to safety-critical systems in on the increase. To be successful, the software must be designed and constructed to meet the functional and performance requirements of the system. For safety reason, the software must be demonstrated not only to meet these requirements, but also to operate safely as a component within the system. For longer-term cost consideration, the software must be designed and structured to ease future maintenance and modifications. This paper presents a software engineering process for the production of safety-critical software for a nuclear power plant. The presentation is expository in nature of a viable high quality safety-critical software development. It is based on the ideas of a rational design process and on the experience of the adaptation of such process in the production of the safety-critical software for the shutdown system number two of Wolsung 2, 3 and 4 nuclear power generation plants. This process is significantly different from a conventional process in terms of rigorous software development phases and software design techniques, The process covers documentation, design, verification and testing using mathematically precise notations and highly reviewable tabular format to specify software requirements and software requirements and software requirements and code against software design using static analysis. The software engineering process described in this paper applies the principle of information-hiding decomposition in software design using a modular design technique so that when a change is required or an error is detected, the affected scope can be readily and confidently located. it also facilitates a sense of high degree of confidence in the 'correctness' of the software production, and provides a relatively simple and straightforward code implementation effort. 1 figs., 10 refs. (Author)

Machine-Checked Sequencer for Critical Embedded Code Generator

Science.gov (United States)

Izerrouken, Nassima; Pantel, Marc; Thirioux, Xavier

This paper presents the development of a correct-by-construction block sequencer for GeneAuto a qualifiable (according to DO178B/ED12B recommendation) automatic code generator. It transforms Simulink models to MISRA C code for safety critical systems. Our approach which combines classical development process and formal specification and verification using proof-assistants, led to preliminary fruitful exchanges with certification authorities. We present parts of the classical user and tools requirements and derived formal specifications, implementation and verification for the correctness and termination of the block sequencer. This sequencer has been successfully applied to real-size industrial use cases from various transportation domain partners and led to requirement errors detection and a correct-by-construction implementation.

Lecture notes for criticality safety

International Nuclear Information System (INIS)

Fullwood, R.

1992-03-01

These lecture notes for criticality safety are prepared for the training of Department of Energy supervisory, project management, and administrative staff. Technical training and basic mathematics are assumed. The notes are designed for a two-day course, taught by two lecturers. Video tapes may be used at the options of the instructors. The notes provide all the materials that are necessary but outside reading will assist in the fullest understanding. The course begins with a nuclear physics overview. The reader is led from the macroscopic world into the microscopic world of atoms and the elementary particles that constitute atoms. The particles, their masses and sizes and properties associated with radioactive decay and fission are introduced along with Einstein's mass-energy equivalence. Radioactive decay, nuclear reactions, radiation penetration, shielding and health-effects are discussed to understand protection in case of a criticality accident. Fission, the fission products, particles and energy released are presented to appreciate the dangers of criticality. Nuclear cross sections are introduced to understand the effectiveness of slow neutrons to produce fission. Chain reactors are presented as an economy; effective use of the neutrons from fission leads to more fission resulting in a power reactor or a criticality excursion. The six-factor formula is presented for managing the neutron budget. This leads to concepts of material and geometric buckling which are used in simple calculations to assure safety from criticality. Experimental measurements and computer code calculations of criticality are discussed. To emphasize the reality, historical criticality accidents are presented in a table with major ones discussed to provide lessons-learned. Finally, standards, NRC guides and regulations, and DOE orders relating to criticality protection are presented

Sensitivity and uncertainty analyses applied to criticality safety validation. Volume 2

International Nuclear Information System (INIS)

Broadhead, B.L.; Hopper, C.M.; Parks, C.V.

1999-01-01

This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies developed in Volume 1 to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the existing S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently in use by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The methods for application of S/U and generalized linear-least-square methodology (GLLSM) tools to the criticality safety validation procedures were described in Volume 1 of this report. Volume 2 of this report presents the application of these procedures to the validation of criticality safety analyses supporting uranium operations where enrichments are greater than 5 wt %. Specifically, the traditional k eff trending analyses are compared with newly developed k eff trending procedures, utilizing the D and c k coefficients described in Volume 1. These newly developed procedures are applied to a family of postulated systems involving U(11)O 2 fuel, with H/X values ranging from 0--1,000. These analyses produced a series of guidance and recommendations for the general usage of these various techniques. Recommendations for future work are also detailed

Module Testing Techniques for Nuclear Safety Critical Software Using LDRA Testing Tool

International Nuclear Information System (INIS)

Moon, Kwon-Ki; Kim, Do-Yeon; Chang, Hoon-Seon; Chang, Young-Woo; Yun, Jae-Hee; Park, Jee-Duck; Kim, Jae-Hack

2006-01-01

The safety critical software in the I and C systems of nuclear power plants requires high functional integrity and reliability. To achieve those requirement goals, the safety critical software should be verified and tested according to related codes and standards through verification and validation (V and V) activities. The safety critical software testing is performed at various stages during the development of the software, and is generally classified as three major activities: module testing, system integration testing, and system validation testing. Module testing involves the evaluation of module level functions of hardware and software. System integration testing investigates the characteristics of a collection of modules and aims at establishing their correct interactions. System validation testing demonstrates that the complete system satisfies its functional requirements. In order to generate reliable software and reduce high maintenance cost, it is important that software testing is carried out at module level. Module testing for the nuclear safety critical software has rarely been performed by formal and proven testing tools because of its various constraints. LDRA testing tool is a widely used and proven tool set that provides powerful source code testing and analysis facilities for the V and V of general purpose software and safety critical software. Use of the tool set is indispensable where software is required to be reliable and as error-free as possible, and its use brings in substantial time and cost savings, and efficiency

The automatic programming for safety-critical software in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Eom, Heung Seop; Choi, You Rark

1998-06-01

We defined the Korean unique safety-critical software development methodology by modifying Dr. Harel`s statechart-based on formal methods in order to digitalized the reactor protection system. It is suggested software requirement specification guideline to specify design specification which is basis for requirement specification and automatic programming by the caused by shutdown parameter logic of the steam generator water level for Wolsung 2/3/4 unit SDS no.1 and simulated it by binding the Graphic User Interface (GUI). We generated the K and R C code automatically by utilizing the Statemate MAGNUM Sharpshooter/C code generator. Auto-generated K and R C code is machine independent code and has high productivity, quality and provability. The following are the summaries of major research and development. - Set up the Korean unique safety-critical software development methodology - Developed software requirement specification guidelines - Developed software design specification guidelines - Reactor trip modeling for steam generator waster level Wolsung 2/3/4 SDS no. 1 shutdown parameter logic - Graphic panel binding with GUI. (author). 20 refs., 12 tabs., 15 figs

The automatic programming for safety-critical software in nuclear power plants

International Nuclear Information System (INIS)

Kim, Jang Yeol; Eom, Heung Seop; Choi, You Rark

1998-06-01

We defined the Korean unique safety-critical software development methodology by modifying Dr. Harel's statechart-based on formal methods in order to digitalized the reactor protection system. It is suggested software requirement specification guideline to specify design specification which is basis for requirement specification and automatic programming by the caused by shutdown parameter logic of the steam generator water level for Wolsung 2/3/4 unit SDS no.1 and simulated it by binding the Graphic User Interface (GUI). We generated the K and R C code automatically by utilizing the Statemate MAGNUM Sharpshooter/C code generator. Auto-generated K and R C code is machine independent code and has high productivity, quality and provability. The following are the summaries of major research and development. - Set up the Korean unique safety-critical software development methodology - Developed software requirement specification guidelines - Developed software design specification guidelines - Reactor trip modeling for steam generator waster level Wolsung 2/3/4 SDS no. 1 shutdown parameter logic - Graphic panel binding with GUI. (author). 20 refs., 12 tabs., 15 figs

Preparation for the second edition of nuclear criticality safety handbook

International Nuclear Information System (INIS)

Okuno, Hiroshi; Nomura, Yasushi

1997-01-01

The making of the second edition of Nuclear Criticality Safety Handbook entered the final stage of investigation by the working group. In the second edition, the newest results of the researches in Japan were taken. In this report, among the subjects which were examined continuously from the first edition published in 1988, the size of fuel particles which can be regarded as homogeneous even in a heterogeneous system, the reactivity effect when fuel concentration distribution became not uniform in a homogeneous fuel system, the method of evaluating criticality safety in which submersion is not assumed, and the criticality data when fuel burning is considered are explained. Further, about the matters related to the criticality in chemical processes and the matters related to criticality accident, the outlines are introduced. Finally, the state of preparation for aiming at the third edition is mentioned. Criticality safety control is important for overall nuclear fuel cycle including the transportation and storage of fuel. The course of the publication of this Handbook is outlined. The matters which have been successively examined from the first edition, the results of criticality safety analysis for the dissolving tanks of fuel reprocessing, and the analysis code and the simplified evaluation method for criticality accident are reported. (K.I.)

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

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.

International handbook of evaluated criticality safety benchmark experiments

International Nuclear Information System (INIS)

2010-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span over 55,000 pages and contain 516 evaluations with benchmark specifications for 4,405 critical, near critical, or subcritical configurations, 24 criticality alarm placement / shielding configurations with multiple dose points for each, and 200 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these evaluations; however, benchmark specifications are not derived for such experiments (in some cases models are provided in an appendix). Approximately 770 experimental configurations are categorized as unacceptable for use as criticality safety benchmark experiments. Additional evaluations are in progress and will be

Criticality Safety Problems Related to Storage of Highly Active Liquid Waste

International Nuclear Information System (INIS)

Amin, E.

1999-01-01

The geometries of liquid waste storage tanks are not generally safe against criticality. Normally, this does not cause problems as fissile materials exist in nitric acid solution only as depleted uranium or in insignificant concentration of the originally reprocessed inventory of plutonium. However, if sedimentation of solid particles would occur, the deposited material would cause criticality safety problems. Particularly, non-horizontal installation of the storage tanks would increase the Eigen value. The effect of the storage tank inclination and the presence of transplutonium elements on the criticality safety are investigated using the NCNSRC code packages. The results are compared well with a similar German published results

Criticality safety and shielding analysis of WWER-440 fuel configurations

International Nuclear Information System (INIS)

Christoskov, I.

2008-01-01

An overview is made of some studies performed on the criticality safety and radiation shielding analysis of irradiated WWER-440 fuel storage and handling configurations. The analytical tools are based on the SCALE 4.4a code system, in combination with the TORT discrete ordinates transport code and the BUGLE-96 cross-sections library. The accuracy of some important results is assessed through comparison with independent evaluations and with measurement data. (author)

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)

Overview of the activities of the OECD/NEA/NSC working party on nuclear criticality safety

International Nuclear Information System (INIS)

Nouri, A.; Blomquist, R.; Bradyraap, M.; Briggs, B.; Cousinou, P.; Nomura, Y.; Weber, W.

2003-01-01

The OECD Nuclear Energy Agency (NEA) started dealing with criticality-safety related subjects back in the seventies. In the mid-nineties, several activities related to criticality-safety were grouped together into the Working Party on Nuclear Criticality Safety. This working party has since been operating and reporting to the Nuclear Science Committee. Six expert groups co-ordinate various activities ranging from experimental evaluations to code and data inter-comparisons for the study of static and transient criticality behaviours. The paper describes current activities performed in this framework and the achievements of the various expert groups. (author)

Nuclear criticality safety guide

International Nuclear Information System (INIS)

Pruvost, N.L.; Paxton, H.C.

1996-09-01

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators

Nuclear criticality safety guide

Energy Technology Data Exchange (ETDEWEB)

Pruvost, N.L.; Paxton, H.C. [eds.

1996-09-01

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

Sensitivity and uncertainty analyses applied to criticality safety validation, methods development. Volume 1

International Nuclear Information System (INIS)

Broadhead, B.L.; Hopper, C.M.; Childs, R.L.; Parks, C.V.

1999-01-01

This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the available S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently used by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The S/U methods that are presented in this volume are designed to provide a formal means of establishing the range (or area) of applicability for criticality safety data validation studies. The development of parameters that are analogous to the standard trending parameters forms the key to the technique. These parameters are the D parameters, which represent the differences by group of sensitivity profiles, and the ck parameters, which are the correlation coefficients for the calculational uncertainties between systems; each set of parameters gives information relative to the similarity between pairs of selected systems, e.g., a critical experiment and a specific real-world system (the application)

International Criticality Safety Benchmark Evaluation Project (ICSBEP) - ICSBEP 2015 Handbook

International Nuclear Information System (INIS)

Bess, John D.

2015-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy (DOE). The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirements and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross-section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span approximately 69000 pages and contain 567 evaluations with benchmark specifications for 4874 critical, near-critical or subcritical configurations, 31 criticality alarm placement/shielding configurations with multiple dose points for each, and 207 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the handbook are benchmark specifications for neutron activation foil and thermoluminescent dosimeter measurements performed at the SILENE critical assembly in Valduc, France as part of a joint venture in 2010 between the US DOE and the French Alternative Energies and Atomic Energy Commission (CEA). A photograph of this experiment is shown on the front cover. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these

International Handbook of Evaluated Criticality Safety Benchmark Experiments - ICSBEP (DVD), Version 2013

International Nuclear Information System (INIS)

2013-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical experiment facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span nearly 66,000 pages and contain 558 evaluations with benchmark specifications for 4,798 critical, near critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the Handbook are benchmark specifications for Critical, Bare, HEU(93.2)- Metal Sphere experiments referred to as ORSphere that were performed by a team of experimenters at Oak Ridge National Laboratory in the early 1970's. A photograph of this assembly is shown on the front cover

Nuclear Data Activities in Support of the DOE Nuclear Criticality Safety Program

International Nuclear Information System (INIS)

Westfall, R.M.; McKnight, R.D.

2005-01-01

The DOE Nuclear Criticality Safety Program (NCSP) provides the technical infrastructure maintenance for those technologies applied in the evaluation and performance of safe fissionable-material operations in the DOE complex. These technologies include an Analytical Methods element for neutron transport as well as the development of sensitivity/uncertainty methods, the performance of Critical Experiments, evaluation and qualification of experiments as Benchmarks, and a comprehensive Nuclear Data program coordinated by the NCSP Nuclear Data Advisory Group (NDAG).The NDAG gathers and evaluates differential and integral nuclear data, identifies deficiencies, and recommends priorities on meeting DOE criticality safety needs to the NCSP Criticality Safety Support Group (CSSG). Then the NDAG identifies the required resources and unique capabilities for meeting these needs, not only for performing measurements but also for data evaluation with nuclear model codes as well as for data processing for criticality safety applications. The NDAG coordinates effort with the leadership of the National Nuclear Data Center, the Cross Section Evaluation Working Group (CSEWG), and the Working Party on International Evaluation Cooperation (WPEC) of the OECD/NEA Nuclear Science Committee. The overall objective is to expedite the issuance of new data and methods to the DOE criticality safety user. This paper describes these activities in detail, with examples based upon special studies being performed in support of criticality safety for a variety of DOE operations

Review of criticality safety and shielding analysis issues for transportation packages

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.

1995-01-01

The staff of the Nuclear Engineering Applications Section (NEAS) at Oak Ridge National Laboratory (ORNL) have been involved for over 25 years with the development and application of computational tools for use in analyzing the criticality safety and shielding features of transportation packages carrying radioactive material (RAM). The majority of the computational tools developed by ORNL/NEAS have been included within the SCALE modular code system (SCALE 1995). This code system has been used throughout the world for the evaluation of nuclear facility and package designs. With this development and application experience as a basis, this paper highlights a number of criticality safety and shielding analysis issues that confront the designer and reviewer of a new RAM package. Changes in the types and quantities of material that need to be shipped will keep these issues before the technical community and provide challenges to future package design and certification

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.)

Preparation of data for criticality safety evaluation of nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Yoshiyama, Hiroshi; Tonoike, Kotaro; Miyoshi, Yoshinori

2005-01-01

Nuclear Criticality Safety Handbook/Data Collection, Version 2 was submitted to the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan as a contract report. In this presentation paper, its outline and related recent works are presented. After an introduction in Chapter 1, useful information to obtain the atomic number densities was collected in Chapter 2. The nuclear characteristic parameters for 11 nuclear fuels were provided in Chapter 3, and subcriticality judgment graphs were given in Chapter 4. The estimated critical and estimated lower-limit critical values were supplied for the 11 nuclear fuels as results of calculations by using the Japanese Evaluated Nuclear Data Library, JENDL-3.2, and the continuous energy Monte Carlo neutron transport code MVP in Chapter 5. The results of benchmark calculations based on the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook were summarized into six fuel categories in Chapter 6. As for recent works, subcriticality judgment graphs for U-SiO 2 and Pu-SiO 2 were obtained. Benchmark calculations were made with the combination of the latest version of the library JENDL-3.3 and MVP code for a series of STACY experiments and the estimated critical and estimated lower-limit critical values of 10 wt%-enriched uranium nitrate solutions were calculated. (author)

Validation of KENO V.a for criticality safety calculations involving WR-1 fast-neutron fuel arrangements

Energy Technology Data Exchange (ETDEWEB)

Gauld, I. C.

1991-07-15

The KENO V.a criticality safety code, used with the SCALE 27-energy-group ENDF/B-IV-based cross-section library, has been validated for low-enriched uranium carbide (UC) WR-1 fast-neutron (FN) fuel arrangements. Because of a lack of relevant experimental data for UC fuel in the published literature, the validation is based primarily on calculational comparisons with critical experiments for fuel types with a range of enrichments and densities that cover those of the FN UC fuel. The ability of KENO V.a to handle the unique annular pin arrangement of the WR-1 FN fuel bundle was established using a comparison with the MCNP3B code used with a continuous-energy ENDF/B-V-based cross-section library. This report is part of the AECL--10146 report series documenting the validation of the KENO V.a criticality safety code.

Migration of nuclear criticality safety software from a mainframe to a workstation environment

International Nuclear Information System (INIS)

Bowie, L.J.; Robinson, R.C.; Cain, V.R.

1993-01-01

The Nuclear Criticality Safety Department (NCSD), Oak Ridge Y-12 Plant has undergone the transition of executing the Martin Marietta Energy Systems Nuclear Criticality Safety Software (NCSS) on IBM mainframes to a Hewlett-Packard (HP) 9000/730 workstation (NCSSHP). NCSSHP contains the following configuration controlled modules and cross-section libraries: BONAMI, CSAS, GEOMCHY, ICE, KENO IV, KENO Va, MODIIFY, NITAWL SCALE, SLTBLIB, XSDRN, UNIXLIB, albedos library, weights library, 16-Group HANSEN-ROACH master library, 27-Group ENDF/B-IV master library, and standard composition library. This paper will discuss the method used to choose the workstation, the hardware setup of the chosen workstation, an overview of Y-12 software quality assurance and configuration control methodology, code validation, difficulties encountered in migrating the codes, and advantages to migrating to a workstation environment

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...

Nuclear criticality safety guide

International Nuclear Information System (INIS)

Ro, Seong Ki; Shin, Hee Seong; Park, Seong Won; Shin, Young Joon.

1997-06-01

Nuclear criticality safety guide was described for handling, transportation and storage of nuclear fissile materials in this report. The major part of the report was excerpted frp, TID-7016(revision 2) and nuclear criticality safety written by Knief. (author). 16 tabs., 44 figs., 5 refs

French safety and criticality testing programmes

International Nuclear Information System (INIS)

Barbry, F.; Leclerc, J.; Manaranche, J.C.; Maubert, L.

1982-01-01

This article underlines the need to include experimental safety-criticality programmes in the French nuclear effort. The means and methods used at the Section of Experimental Nuclear Safety and Criticality Research, attached to the CEA Valduc Centre, are described. Three experimental programmes are presented: safety-criticality of the PWR fuel cycle, neutron poisoning of plutonium solutions by gadolinium and safety-criticality of slightly enriched and slightly moderated uranium oxide. Criticality accidents studies in solution are then described [fr

ALARP considerations in criticality safety assessments

International Nuclear Information System (INIS)

Bowden, Russell L.; Barnes, Andrew; Thorne, Peter R.; Venner, Jack

2003-01-01

Demonstrating that the risk to the public and workers is As Low As Reasonably Practicable (ALARP) is a fundamental requirement of safety cases for nuclear facilities in the United Kingdom. This is embodied in the Safety Assessment Principles (SAPs) published by the Regulator, the essence of which is incorporated within the safety assessment processes of the various nuclear site licensees. The concept of ALARP within criticality safety assessments has taken some time to establish in the United Kingdom. In principle, the licensee is obliged to search for a deterministic criticality safety solution, such as safe geometry vessels and passive control features, rather than placing reliance on active measurement devices and plant administrative controls. This paper presents a consideration of some ALARP issues in relation to the development of criticality safety cases. The paper utilises some idealised examples covering a range of issues facing the criticality safety assessor, including new plant design, operational plant and decommissioning activities. These examples are used to outline the elements of the criticality safety cases and present a discussion of ALARP in the context of criticality safety assessments. (author)

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

2011 Annual Criticality Safety Program Performance Summary

Energy Technology Data Exchange (ETDEWEB)

Andrea Hoffman

2011-12-01

The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The

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...

Role of computers in quality assurance in the LLL Criticality Safety Program

International Nuclear Information System (INIS)

Koponen, B.L.

1978-01-01

Some of the aspects of computational criticality safety quality assurance that have been emphasized in recent years at LLL are summarized. In particular, computer code changes that have been made that help the criticality analyst reduce the number of errors that he makes and to locate those that he does make; and how a computerized ''benchmark'' data base aids him in the validation of his computational methods are discussed

Nuclear criticality safety: 2-day training course

International Nuclear Information System (INIS)

Schlesser, J.A.

1997-02-01

This compilation of notes is presented as a source reference for the criticality safety course. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used as Los Alamos; be able to identify examples of circumstances present during criticality accidents; have participated in conducting two critical experiments; be asked to complete a critique of the nuclear criticality safety training course

Nuclear criticality safety: 2-day training course

Energy Technology Data Exchange (ETDEWEB)

Schlesser, J.A. [ed.] [comp.

1997-02-01

This compilation of notes is presented as a source reference for the criticality safety course. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used as Los Alamos; be able to identify examples of circumstances present during criticality accidents; have participated in conducting two critical experiments; be asked to complete a critique of the nuclear criticality safety training course.

Safety-critical Java for embedded systems

DEFF Research Database (Denmark)

Schoeberl, Martin; Dalsgaard, Andreas Engelbredt; Hansen, René Rydhof

2016-01-01

This paper presents the motivation for and outcomes of an engineering research project on certifiable Javafor embedded systems. The project supports the upcoming standard for safety-critical Java, which defines asubset of Java and libraries aiming for development of high criticality systems....... The outcome of this projectinclude prototype safety-critical Java implementations, a time-predictable Java processor, analysis tools formemory safety, and example applications to explore the usability of safety-critical Java for this applicationarea. The text summarizes developments and key contributions...

Outline of criticality safety research project

International Nuclear Information System (INIS)

Kobayashi, Iwao; Tachimori, Shoichi; Suzaki, Takenori; Takeshita, Isao; Miyoshi, Yoshinori; Nakajima, Ken; Sakurai, Satoshi; Yanagisawa, Hiroshi

1987-01-01

As the power generation capacity of LWRs in Japan increased, the establishment and development of nuclear fuel cycle have become the important subject. Conforming to the safety research project of the nation, the Japan Atomic Energy Research Institute has advanced the project of constructing a new research facility, that is, Nuclear Fuel Cycle Engineering Research Facility (NUCEF). In this facility, it is planned to carry out the research on criticality safety, upgraded reprocessing techniques, and the treatment and disposal of transuranium element wastes. In this paper, the subjects of criticality safety research and the research carried out with a criticality safety experiment facility which is expected to be installed in the NUCEF are briefly reported. The experimental data obtained from the criticality safety handbooks and published literatures in foreign countries are short of the data on the mixture of low enriched uranium and plutonium which is treated in the reprocessing of spent fuel from LWRs. The acquisition of the criticality data for various forms of fuel, the elucidation of the scenario of criticality accidents, and the soundness of the confinement system for gaseous fission products and plutonium are the main subjects. The Static Criticality Safety Facility, Transient Criticality Safety Facility and pulse column system are the main facilities. (Kako, I.)

Safety, danger and catastrophe inevitability in operation of safety-critical software algorithms: a possible new look at software safety analysis

International Nuclear Information System (INIS)

Povyakalo, A.A.

2000-01-01

The paper provides basic definitions and describes the basic procedure of the Formal Qualitative Safety Analysis (FQSA) of critical software algorithms. The procedure is described by C-based pseudo-code. It uses the notion of weakest precondition and representation of a given critical algorithm by a Gurevich's Abstract State Mashine (GASM). For a given GASM and a given Catastrophe Condition the procedure results in a Catastrophe Inevitability Condition (it means that every sequence of algorithm steps lead to a catastrophe early or late), Danger Condition (it means that next step may lead to a catastrophe or make a catastrophe to be inevitable, but a catastrophe may be prevented yet), Safety Condition (it means that a next step can not lead to a catastrophe or make a catastrophe to be inevitable). The using of proposed procedure is illustrated by a simplest test example of algorithm. The FQSA provides a logical basis for PSA of critical algorithm. (author)

A critical overview of safety-related and technological criteria for nuclear fuel

International Nuclear Information System (INIS)

Lahodova, M.; Valach, M.

2000-10-01

A detailed overview of the safety criteria, methods of analysis and computer codes used in OECD countries is presented. A critical analysis of the validity of criteria in the high burnup domain was performed, and recommendations for testing their validity based on available experimental data are put forth. (author)

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.

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.

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

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

Criticality analysis of the storage tubes for irradiated fuel elements from the IEA-R1 with the MCNP code

International Nuclear Information System (INIS)

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

1992-01-01

A criticality safety analysis has been carried out for the storage tubes for irradiated fuel elements from the IEA-R1 research reactor. The analysis utilized the MCNP computer code which allows exact simulations of complex geometries. Aiming reducing the amount of input data, the fuel element cross-sections have been spatially smeared out. The earth material interstice between fuel elements has been approximated conservatively as concrete because its composition was unknown. The storage tubes have been found subcritical for the most adverse conditions (water flooding and un-irradiated fuel elements). A similar analysis with the KENO-IV computer code overestimated the KEF result but still confirmed the criticality safety of the storage tubes. (author)

Criticality safety analysis for mockup facility

International Nuclear Information System (INIS)

Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

2000-03-01

Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

Nuclear criticality safety: 2-day training course

International Nuclear Information System (INIS)

Schlesser, J.A.

1992-11-01

This compilation of notes is presented as a source reference for the criticality safety course. At the completion of this training course, the attendee will: (1) be able to define terms commonly used in nuclear criticality safety; (2) be able to appreciate the fundamentals of nuclear criticality safety; (3) be able to identify factors which affect nuclear criticality safety; (4) be able to identify examples of criticality controls as used at Los Alamos; (5) be able to identify examples of circumstances present during criticality accidents; (6) have participated in conducting two critical experiments

Critical lengths of error events in convolutional codes

DEFF Research Database (Denmark)

Justesen, Jørn

1994-01-01

If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes......If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes...

Critical Lengths of Error Events in Convolutional Codes

DEFF Research Database (Denmark)

Justesen, Jørn; Andersen, Jakob Dahl

1998-01-01

If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes......If the calculation of the critical length is based on the expurgated exponent, the length becomes nonzero for low error probabilities. This result applies to typical long codes, but it may also be useful for modeling error events in specific codes...

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

Applicability of object-oriented design methods and C++ to safety-critical systems

International Nuclear Information System (INIS)

Cuthill, B.B.

1994-01-01

This paper reports on a study identifying risks and benefits of using a software development methodology containing object-oriented design (OOD) techniques and using C++ as a programming language relative to selected features of safety-critical systems development. These features are modularity, functional diversity, removing ambiguous code, traceability, and real-time performance

Elements of a nuclear criticality safety program

International Nuclear Information System (INIS)

Hopper, C.M.

1995-01-01

Nuclear criticality safety programs throughout the United States are quite successful, as compared with other safety disciplines, at protecting life and property, especially when regarded as a developing safety function with no historical perspective for the cause and effect of process nuclear criticality accidents before 1943. The programs evolved through self-imposed and regulatory-imposed incentives. They are the products of conscientious individuals, supportive corporations, obliged regulators, and intervenors (political, public, and private). The maturing of nuclear criticality safety programs throughout the United States has been spasmodic, with stability provided by the volunteer standards efforts within the American Nuclear Society. This presentation provides the status, relative to current needs, for nuclear criticality safety program elements that address organization of and assignments for nuclear criticality safety program responsibilities; personnel qualifications; and analytical capabilities for the technical definition of critical, subcritical, safety and operating limits, and program quality assurance

Traceability of Software Safety Requirements in Legacy Safety Critical Systems

Science.gov (United States)

Hill, Janice L.

2007-01-01

How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

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

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

Criticality safety evaluation in Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Shirai, Nobutoshi; Nakajima, Masayoshi; Takaya, Akikazu; Ohnuma, Hideyuki; Shirouzu, Hidetomo; Hayashi, Shinichiro; Yoshikawa, Koji; Suto, Toshiyuki

2000-04-01

Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 'Criticality safety of single unit' in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units. (author)

Development and validation of a criticality calculation scheme based on French deterministic transport codes

International Nuclear Information System (INIS)

Santamarina, A.

1991-01-01

A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

Engineering design guidelines for nuclear criticality safety

International Nuclear Information System (INIS)

Waltz, W.R.

1988-08-01

This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

Requirement analysis of the safety-critical software implementation for the nuclear power plant

International Nuclear Information System (INIS)

Chang, Hoon Seon; Jung, Jae Cheon; Kim, Jae Hack; Nam, Sang Ku; Kim, Hang Bae

2005-01-01

The safety critical software shall be implemented under the strict regulation and standards along with hardware qualification. In general, the safety critical software has been implemented using functional block language (FBL) and structured language like C in the real project. Software design shall comply with such characteristics as; modularity, simplicity, minimizing the use of sub-routine, and excluding the interrupt logic. To meet these prerequisites, we used the computer-aided software engineering (CASE) tool to substantiate the requirements traceability matrix that were manually developed using Word processors or Spreadsheets. And the coding standard and manual have been developed to confirm the quality of software development process, such as; readability, consistency, and maintainability in compliance with NUREG/CR-6463. System level preliminary hazard analysis (PHA) is performed by analyzing preliminary safety analysis report (PSAR) and FMEA document. The modularity concept is effectively implemented for the overall module configurations and functions using RTP software development tool. The response time imposed on the basis of the deterministic structure of the safety-critical software was measured

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.)

KEOPS and other VENUS experiments dedicated to the criticality safety of a MOX fuel fabrication facility

International Nuclear Information System (INIS)

Lance, Benoit; Van Den Hende, Paul; Marloye, Daniel; Basselier, Jacques; Libon, Henri; De Vleeschhauwer, Marc; Moerenhout, Jeremie; Baeten, Peter

2005-01-01

The qualification scheme of criticality computer codes for Pu bearing powders lies upon databases which suffer from a lack of recent experimental results. As a MOX manufacture, BELGONUCLEAIRE is especially concerned by criticality safety and would like to address such an issue by launching with SCK-CEN an International Programme called KEOPS. (author)

Criticality safety basics, a study guide

Energy Technology Data Exchange (ETDEWEB)

V. L. Putman

1999-09-01

This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates.

Criticality safety basics, a study guide

International Nuclear Information System (INIS)

Putman, V.L.

1999-01-01

This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates

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

The SWAN-SCALE code for the optimization of critical systems

International Nuclear Information System (INIS)

Greenspan, E.; Karni, Y.; Regev, D.; Petrie, L.M.

1999-01-01

The SWAN optimization code was recently developed to identify the maximum value of k eff for a given mass of fissile material when in combination with other specified materials. The optimization process is iterative; in each iteration SWAN varies the zone-dependent concentration of the system constituents. This change is guided by the equal volume replacement effectiveness functions (EVREF) that SWAN generates using first-order perturbation theory. Previously, SWAN did not have provisions to account for the effect of the composition changes on neutron cross-section resonance self-shielding; it used the cross sections corresponding to the initial system composition. In support of the US Department of Energy Nuclear Criticality Safety Program, the authors recently removed the limitation on resonance self-shielding by coupling SWAN with the SCALE code package. The purpose of this paper is to briefly describe the resulting SWAN-SCALE code and to illustrate the effect that neutron cross-section self-shielding could have on the maximum k eff and on the corresponding system composition

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.)

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

Nuclear criticality safety department training implementation

International Nuclear Information System (INIS)

Carroll, K.J.; Taylor, R.G.; Worley, C.A.

1996-01-01

The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. The NCSD Qualification Program is described in Y/DD-694, Qualification Program, Nuclear Criticality Safety Department This document provides a listing of the roles and responsibilities of NCSD personnel with respect to training and details of the Training Management System (TMS) programs, Mentoring Checklists and Checksheets, as well as other documentation utilized to implement the program. This document supersedes Y/DD-696, Revision 2, dated 3/27/96, Training Implementation, Nuclear Criticality Safety Department. There are no backfit requirements associated with revisions to this document

Critical experiments facility and criticality safety programs at JAERI

International Nuclear Information System (INIS)

Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Miyoshi, Yoshinori; Nomura, Yasushi

1985-10-01

The nuclear criticality safety is becoming a key point in Japan in the safety considerations for nuclear installations outside reactors such as spent fuel reprocessing facilities, plutonium fuel fabrication facilities, large scale hot alboratories, and so on. Especially a large scale spent fuel reprocessing facility is being designed and would be constructed in near future, therefore extensive experimental studies are needed for compilation of our own technical standards and also for verification of safety in a potential criticality accident to obtain public acceptance. Japan Atomic Energy Research Institute is proceeding a construction program of a new criticality safety experimental facility where criticality data can be obtained for such solution fuels as mainly handled in a reprocessing facility and also chemical process experiments can be performed to investigate abnormal phenomena, e.g. plutonium behavior in solvent extraction process by using pulsed colums. In FY 1985 detail design of the facility will be completed and licensing review by the government would start in FY 1986. Experiments would start in FY 1990. Research subjects and main specifications of the facility are described. (author)

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.

Autoclave nuclear criticality safety analysis

Energy Technology Data Exchange (ETDEWEB)

D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

1991-12-31

Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

A Criticality Safety Study on Storing Unirradiated Cintichem-Type Targets at Sandia National Laboratories

International Nuclear Information System (INIS)

Romero, D.J.; Parma, E.J.; Busch, R.D.

1999-01-01

This criticality safety analysis is performed to determine the effective multiplication factor (k eff ) for a storage cabinet filled with unirradiated Cintichem-type targets. These targets will be used to produce 99 Mo at Sandia National Laboratories and will be stored on-site prior to irradiation in the Annular Core Research Reactor. The analysis consisted of using the Monte Carlo code MCNP (Version 4A) to model and predict the k eff for the proposed dry storage configuration under credible loss of geometry and moderator control. Effects of target pitch, non-uniform loading, and target internal/external flooding are evaluated. Further studies were done with deterministic methods to verify the results obtained from MCNP and to obtain a clearer understanding of the parameters affecting system criticality. The diffusion accelerated neutral particle transport code ONEDANT was used to model the target in a one-dimensional, infinite half-slab geometry and determine the critical slab thickness. Hand calculations were also completed to determine the critical slab thickness with modified one-group, and one-group, two region approximations. Results obtained from ONEDANT and the hand calculations were compared to applicable cases in a commonly used criticality safety analysis handbook. Overall, the critical slab thicknesses obtained in the deterministic analysis were much larger than the dimensions of the cabinet and further support the predictions by MCNP that a critical system cannot be attained for the base case or in conditions where loss of geometry and moderation control occur

Anatomy of safety-critical computing problems

International Nuclear Information System (INIS)

Swu Yih; Fan Chinfeng; Shirazi, Behrooz

1995-01-01

This paper analyzes the obstacles faced by current safety-critical computing applications. The major problem lies in the difficulty to provide complete and convincing safety evidence to prove that the software is safe. We explain this problem from a fundamental perspective by analyzing the essence of safety analysis against that of software developed by current practice. Our basic belief is that in order to perform a successful safety analysis, the state space structure of the analyzed system must have some properties as prerequisites. We propose the concept of safety analyzability, and derive its necessary and sufficient conditions; namely, definability, finiteness, commensurability, and tractability. We then examine software state space structures against these conditions, and affirm that the safety analyzability of safety-critical software developed by current practice is severely restricted by its state space structure and by the problem of exponential growth cost. Thus, except for small and simple systems, the safety evidence may not be complete and convincing. Our concepts and arguments successfully explain the current problematic situation faced by the safety-critical computing domain. The implications are also discussed

The octopus burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

1996-09-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

The octopus burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

1996-01-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

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…

Software reliability for safety-critical applications

International Nuclear Information System (INIS)

Everett, B.; Musa, J.

1994-01-01

In this talk, the authors address the question open-quotes Can Software Reliability Engineering measurement and modeling techniques be applied to safety-critical applications?close quotes Quantitative techniques have long been applied in engineering hardware components of safety-critical applications. The authors have seen a growing acceptance and use of quantitative techniques in engineering software systems but a continuing reluctance in using such techniques in safety-critical applications. The general case posed against using quantitative techniques for software components runs along the following lines: safety-critical applications should be engineered such that catastrophic failures occur less frequently than one in a billion hours of operation; current software measurement/modeling techniques rely on using failure history data collected during testing; one would have to accumulate over a billion operational hours to verify failure rate objectives of about one per billion hours

Nuclear Criticality Safety Department Qualification Program

International Nuclear Information System (INIS)

Carroll, K.J.; Taylor, R.G.; Worley, C.A.

1996-01-01

The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of highly qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document defines the Qualification Program to address the NCSD technical and managerial qualification as required by the Y-1 2 Training Implementation Matrix (TIM). This Qualification Program is in compliance with DOE Order 5480.20A and applicable Lockheed Martin Energy Systems, Inc. (LMES) and Y-1 2 Plant procedures. It is implemented through a combination of WES plant-wide training courses and professional nuclear criticality safety training provided within the department. This document supersedes Y/DD-694, Revision 2, 2/27/96, Qualification Program, Nuclear Criticality Safety Department There are no backfit requirements associated with revisions to this document

Reusable libraries for safety-critical Java

DEFF Research Database (Denmark)

Rios Rivas, Juan Ricardo; Schoeberl, Martin

2014-01-01

The large collection of Java class libraries is a main factor of the success of Java. However, these libraries assume that a garbage-collected heap is used. Safety-critical Java uses scope-based memory areas instead of a garbage-collected heap. Therefore, the Java class libraries are problematic...... to use in safety-critical Java. We have identified common programming patterns in the Java class libraries that make them unsuitable for safety-critical Java. We propose ways to improve the libraries to avoid the impact of the identified problematic patterns. We illustrate these changes by implementing...

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

Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5

International Nuclear Information System (INIS)

Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.

2011-01-01

The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)

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....

A desktop 3D printer in safety-critical Java

DEFF Research Database (Denmark)

Strøm, Tórur Biskopstø; Schoeberl, Martin

2012-01-01

there exist several safety-critical Java framework implementations, there is a lack of safety-critical use cases implemented according to the specification. In this paper we present a 3D printer and its safety-critical Java level 1 implementation as a use case. With basis in the implementation we evaluate......It is desirable to bring Java technology to safety-critical systems. To this end The Open Group has created the safety-critical Java specification, which will allow Java applications, written according to the specification, to be certifiable in accordance with safety-critical standards. Although...

Evaluation and validation of criticality codes for fuel dissolver calculations

International Nuclear Information System (INIS)

Santamarina, A.; Smith, H.J.; Whitesides, G.E.

1991-01-01

During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

Criticality Safety Evaluation for the TACS at DAF

Energy Technology Data Exchange (ETDEWEB)

Percher, C. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2011-06-10

Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, Guidance for Nuclear Criticality Safety Engineer Training and Qualification. This document is a criticality safety evaluation of the training activities and operations associated with HS-3201-P, Nuclear Criticality 4-Day Training Course (Practical). This course was designed to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program1. The hands-on, or laboratory, portion of the course will utilize the Training Assembly for Criticality Safety (TACS) and will be conducted in the Device Assembly Facility (DAF) at the Nevada Nuclear Security Site (NNSS). The training activities will be conducted by Lawrence Livermore National Laboratory following the requirements of an Integrated Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of an LLNL Certified Fissile Material Handler.

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)

The Department of Energy nuclear criticality safety program

International Nuclear Information System (INIS)

Felty, J.R.

2004-01-01

This paper broadly covers key events and activities from which the Department of Energy Nuclear Criticality Safety Program (NCSP) evolved. The NCSP maintains fundamental infrastructure that supports operational criticality safety programs. This infrastructure includes continued development and maintenance of key calculational tools, differential and integral data measurements, benchmark compilation, development of training resources, hands-on training, and web-based systems to enhance information preservation and dissemination. The NCSP was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 97-2, Criticality Safety, and evolved from a predecessor program, the Nuclear Criticality Predictability Program, that was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 93-2, The Need for Critical Experiment Capability. This paper also discusses the role Dr. Sol Pearlstein played in helping the Department of Energy lay the foundation for a robust and enduring criticality safety infrastructure.

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

Overview of DOE/ONS criticality safety projects

International Nuclear Information System (INIS)

Barber, R.W.; Brown, B.P.; Hopper, C.M.

1985-01-01

The evolution of Federal involvement with nuclear criticality safety has traversed through the 1940's and early 1950's with the Manhattan Engineering District, the 1950's and 1960's with the Atomic Energy Commission, the early 1970's with the Energy Research and Development Administration, and the late 1970's to date with the US Department of Energy. The importance of nuclear criticality safety has been maintained throughout these periods; however, criticality safety has received shifting emphases in research/applications, promulgations of regulations/standards, origins of fiscal support and organization. In June 1981 the Office of Nuclear Safety was established in response to a Department of Energy study of the impact of the March 1979 Three Mile Island accident. The organizational structure of the ONS, its program for establishing and maintaining a progressive nuclear criticality safety program, and associated projects, and current history of ONS's fiscal support of program projects is presented. With the establishment of the ONS came concomitant missions to develop and maintain nuclear safety policy and requirements, to provide independent assurance that nuclear operations are performed safely, to provide resources and management for DOE responses to nuclear accidents, and to provide technical support. In the past four years, ONS has developed and initiated a continuing Department Nuclear Criticality Safety Program in such areas as communications and information, physics of criticality, knowledge of factors affecting criticality, and computational capability

Criticality benchmarks for COG: A new point-wise Monte Carlo code

International Nuclear Information System (INIS)

Alesso, H.P.; Pearson, J.; Choi, J.S.

1989-01-01

COG is a new point-wise Monte Carlo code being developed and tested at LLNL for the Cray computer. It solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) charged particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems. However, its point-wise cross-sections also make it effective for a wide variety of criticality problems. COG has some similarities to a number of other computer codes used in the shielding and criticality community. These include the Lawrence Livermore National Laboratory (LLNL) codes TART and ALICE, the Los Alamos National Laboratory code MCNP, the Oak Ridge National Laboratory codes 05R, 06R, KENO, and MORSE, the SACLAY code TRIPOLI, and the MAGI code SAM. Each code is a little different in its geometry input and its random-walk modification options. Validating COG consists in part of running benchmark calculations against critical experiments as well as other codes. The objective of this paper is to present calculational results of a variety of critical benchmark experiments using COG, and to present the resulting code bias. Numerous benchmark calculations have been completed for a wide variety of critical experiments which generally involve both simple and complex physical problems. The COG results, which they report in this paper, have been excellent

Evaluation and validation of criticality codes for fuel dissolver calculations

International Nuclear Information System (INIS)

Santamarina, A.; Smith, H.J.; Whitesides, G.E.

1991-01-01

During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

Nuclear criticality predictability

International Nuclear Information System (INIS)

Briggs, J.B.

1999-01-01

As a result of lots of efforts, a large portion of the tedious and redundant research and processing of critical experiment data has been eliminated. The necessary step in criticality safety analyses of validating computer codes with benchmark critical data is greatly streamlined, and valuable criticality safety experimental data is preserved. Criticality safety personnel in 31 different countries are now using the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Much has been accomplished by the work of the ICSBEP. However, evaluation and documentation represents only one element of a successful Nuclear Criticality Safety Predictability Program and this element only exists as a separate entity, because this work was not completed in conjunction with the experimentation process. I believe; however, that the work of the ICSBEP has also served to unify the other elements of nuclear criticality predictability. All elements are interrelated, but for a time it seemed that communications between these elements was not adequate. The ICSBEP has highlighted gaps in data, has retrieved lost data, has helped to identify errors in cross section processing codes, and has helped bring the international criticality safety community together in a common cause as true friends and colleagues. It has been a privilege to associate with those who work so diligently to make the project a success. (J.P.N.)

A Profile for Safety Critical Java

DEFF Research Database (Denmark)

Schoeberl, Martin; Søndergaard, Hans; Thomsen, Bent

2007-01-01

We propose a new, minimal specification for real-time Java for safety critical applications. The intention is to provide a profile that supports programming of applications that can be validated against safety critical standards such as DO-178B [15]. The proposed profile is in line with the Java...... specification request JSR-302: Safety Critical Java Technology, which is still under discussion. In contrast to the current direction of the expert group for the JSR-302 we do not subset the rather complex Real-Time Specification for Java (RTSJ). Nevertheless, our profile can be implemented on top of an RTSJ...

Nuclear criticality safety in Canada

International Nuclear Information System (INIS)

Shultz, K.R.

1980-04-01

The approach taken to nuclear criticality safety in Canada has been influenced by the historical development of participants. The roles played by governmental agencies and private industry since the Atomic Energy Control Act was passed into Canadian Law in 1946 are outlined to set the scene for the current situation and directions that may be taken in the future. Nuclear criticality safety puts emphasis on the control of materials called special fissionable material in Canada. A brief account is given of the historical development and philosophy underlying the existing regulations governing special fissionable material. Subsequent events have led to a change in emphasis in the regulatory process that has not yet been fully integrated into Canadian legislation and regulations. Current efforts towards further development of regulations governing the practice of nuclear criticality safety are described. (auth)

Criticality safety calculations for the nuclear waste disposal canisters

International Nuclear Information System (INIS)

Anttila, M.

1996-12-01

The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

Impact of axial burnup profile on criticality safety of ANPP spent fuel cask

International Nuclear Information System (INIS)

Bznuni, S.

2006-01-01

Criticality safety assessment for WWER-440 NUHOMS cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in such way that the neutron multiplication factor k eff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. NRSC ANRA in order to improve future fuel storage efficiency initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon was not taken into account in the Safety Analysis Report of NUHOMS spent fuel storage constructed on the site of ANPP. Although ANRA does not yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burnup profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality calculations of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn't taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The actinides and actinides + fission products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the k eff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect

Nuclear criticality safety: 3-day training course

International Nuclear Information System (INIS)

Schlesser, J.A.

1993-06-01

The open-quotes 3-Day Training Courseclose quotes is an intensive course in criticality safety consisting of lectures and laboratory sessions, including active student participation in actual critical experiments, a visit to a plutonium processing facility, and in-depth discussions on safety philosophy. The program is directed toward personnel who currently have criticality safety responsibilities in the capacity of supervisory staff and/or line management. This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. It should be noted that when chapters were extracted, an attempt was made to maintain footnotes and references as originally written. Photographs and illustrations are numbered sequentially

Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

International Nuclear Information System (INIS)

Pesic, M. P.

1997-01-01

Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

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

Effects of neutron data libraries and criticality codes on IAEA criticality benchmark problems

International Nuclear Information System (INIS)

Sarker, Md.M.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

1993-10-01

In order to compare the effects of neutron data libraries and criticality codes to thermal reactors (LWR), the IAEA criticality benchmark calculations have been performed. The experiments selected in this study include TRX-1 and TRX-2 with a simple geometric configuration. Reactor lattice calculation codes WIMS-D/4, MCNP-4, JACS (MGCL, KENO), and SRAC were used in the present calculations. The TRX cores were analyzed by WIMS-D/4 using WIMS original library and also by MCNP-4, JACS (MGCL, KENO), and SRAC using the libraries generated from JENDL-3 and ENDF/B-IV nuclear data files. An intercomparison work for the above mentioned code systems and cross section libraries was performed by analyzing the LWR benchmark experiments TRX-1 and TRX-2. The TRX cores were also analyzed for supercritical and subcritical conditions and these results were compared. In the case of critical condition, the results were in good agreement. But for the supercritical and subcritical conditions, the difference of the results obtained by using the different cross section libraries become larger than for the critical condition. (author)

Criticality safety evaluation for TWR-S fuel assembly transportation using TK-S16 containers

International Nuclear Information System (INIS)

Pesic, M.P.; Steljic, M.M.; Antic, D.P.

2002-01-01

Criticality safety issues, concerning transportation of fresh high-enriched uranium fuel elements (TWR-S fuel assembly type) with Russian containers TK-S16, are objects of study in this paper. Three-dimensional (3D) models of fuel element and container were made, based upon their well-known geometry and material structure. The way to pack fuel elements in a bundle inside of the container is proposed. Calculations were done by MCNP4B2 computer code. This Monte Carlo criticality code determined the effective multiplication factor from the cross-section data and specific geometry data. This evaluation demonstrated the subcriticality of a single package and an array of packages during normal conditions of transport and various hypothetical accident conditions. (author)

Status of criticality safety research at NUCEF

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Two critical facilities, named STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility), at the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) started their hot operations in 1995. Since then, basic experimental data for criticality safety research have been accumulated using STACY, and supercritical experiments for the study of criticality accident in a reprocessing plant have been performed using TRACY. In this paper, the outline of those critical facilities and the main results of TRACY experiments are presented. (author)

Development of Safety-Critical Software for Nuclear Power Plant using a CASE Tool

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Ho; Oh, Do Young; Kim, Koh Eun; Choi, Woong Seock; Sohn, Se Do; Kim, Jae Hack; Kim, Hang Bae [KEPCO E and C, Daejeon (Korea, Republic of)

2011-08-15

The Integrated SOftware Development Environment (ISODE) is developed to provide the major S/W life cycle processes that are composed of development process, V/V process, requirements traceability process, and automated document generation process and target importing process to Programmable Logic Controller (PLC) platform. This provides critical safety software developers with a certified, domain optimized, model-based development environment, and the associated services to reduce time and efforts to develop software such as debugging, simulation, code generation and document generation. This also provides critical safety software verifiers with integrated V/V features of each phase of the software life cycle using appropriate tools such as model test coverage, formal verification, and automated report generation. In addition to development and verification, the ISODE gives a complete traceability solution from the SW design phase to the testing phase. Using this information, the coverage and impact analysis can be done easily whenever software modification is necessary. The final source codes of ISODE are imported into the newly developed PLC environment, as a module based after automatically converted into the format required by PLC. Additional tests for module and unit level are performed on the target platform.

Development of Safety-Critical Software for Nuclear Power Plant using a CASE Tool

International Nuclear Information System (INIS)

Kim, Chang Ho; Oh, Do Young; Kim, Koh Eun; Choi, Woong Seock; Sohn, Se Do; Kim, Jae Hack; Kim, Hang Bae

2011-01-01

The Integrated SOftware Development Environment (ISODE) is developed to provide the major S/W life cycle processes that are composed of development process, V/V process, requirements traceability process, and automated document generation process and target importing process to Programmable Logic Controller (PLC) platform. This provides critical safety software developers with a certified, domain optimized, model-based development environment, and the associated services to reduce time and efforts to develop software such as debugging, simulation, code generation and document generation. This also provides critical safety software verifiers with integrated V/V features of each phase of the software life cycle using appropriate tools such as model test coverage, formal verification, and automated report generation. In addition to development and verification, the ISODE gives a complete traceability solution from the SW design phase to the testing phase. Using this information, the coverage and impact analysis can be done easily whenever software modification is necessary. The final source codes of ISODE are imported into the newly developed PLC environment, as a module based after automatically converted into the format required by PLC. Additional tests for module and unit level are performed on the target platform

Proceedings of the Nuclear Criticality Technology Safety Workshop

Energy Technology Data Exchange (ETDEWEB)

Rene G. Sanchez

1998-04-01

This document contains summaries of most of the papers presented at the 1995 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 16 and 17 at San Diego, Ca. The meeting was broken up into seven sessions, which covered the following topics: (1) Criticality Safety of Project Sapphire; (2) Relevant Experiments For Criticality Safety; (3) Interactions with the Former Soviet Union; (4) Misapplications and Limitations of Monte Carlo Methods Directed Toward Criticality Safety Analyses; (5) Monte Carlo Vulnerabilities of Execution and Interpretation; (6) Monte Carlo Vulnerabilities of Representation; and (7) Benchmark Comparisons.

Program of nuclear criticality safety experiment at JAERI

International Nuclear Information System (INIS)

Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Ohnishi, Nobuaki

1983-11-01

JAERI is promoting the nuclear criticality safety research program, in which a new facility for criticality safety experiments (Criticality Safety Experimental Facility : CSEF) is to be built for the experiments with solution fuel. One of the experimental researches is to measure, collect and evaluate the experimental data needed for evaluation of criticality safety of the nuclear fuel cycle facilities. Another research area is a study of the phenomena themselves which are incidental to postulated critical accidents. Investigation of the scale and characteristics of the influences caused by the accident is also included in this research. The result of the conceptual design of CSEF is summarized in this report. (author)

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'

Criticality safety

International Nuclear Information System (INIS)

Walker, G.

1983-01-01

When a sufficient quantity of fissile material is brought together a self-sustaining neutron chain reaction will be started in it and will continue until some change occurs in the fissile material to stop the chain reaction. The quantity of fissile material required is the 'Critical Mass'. This is not a fixed quantity even for a given type of fissile material but varies between quite wide limits depending on a number of factors. In a nuclear reactor the critical mass of fissile material is assembled under well-defined condition to produce a controllable chain reaction. The same materials have to be handled outside the reactor in all stages of fuel element manufacture, storage, transport and irradiated fuel reprocessing. At any stage it is possible (at least in principle) to assemble a critical mass and thus initiate an accidental and uncontrollable chain reaction. Avoiding this is what criticality safety is all about. A system is just critical when the rate of production of neutrons balances the rate of loss either by escape or by absorption. The factors affecting criticality are, therefore, those which effect neutron production and loss. The principal ones are:- type of nuclide and enrichment (or isotopic composition), moderation, reflection, concentration (density), shape and interaction. Each factor is considered in detail. (author)

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

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.

FELIX - a computer code for simulation of criticality excursions in liquid fissile solutions

International Nuclear Information System (INIS)

Gmal, B.; Weber, J.

1989-01-01

Knowledge of characteristic parameters like evolved power fission yield during an accidental excursion is of essential importance to estimate possible radiological consequences and resulting safety hazards. The computer code 'FELIX' simulates excursion characteristics of aqueous critical assemblies: Starting out from given initial conditions the space-dependent neutron kinetic equations are solved in one-dimensional geometry. Power, fission yield, reactivity and temperature are calculated as a function of time. Reactivity-feedback includes density effects and radiolytic gas voids. Results from calculations are compared with CRAC-experiments. (orig.)

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)

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

Architecture Level Safety Analyses for Safety-Critical Systems

Directory of Open Access Journals (Sweden)

K. S. Kushal

2017-01-01

Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

Plant safety review from mass criticality accident

International Nuclear Information System (INIS)

Susanto, B.G.

2000-01-01

The review has been done to understand the resent status of the plant in facing postulated mass criticality accident. From the design concept of the plant all the components in the system including functional groups have been designed based on favorable mass/geometry safety principle. The criticality safety for each component is guaranteed because all the dimensions relevant to criticality of the components are smaller than dimensions of 'favorable mass/geometry'. The procedures covering all aspects affecting quality including the safety related are developed and adhered to at all times. Staff are indoctrinated periodically in short training session to warn the important of the safety in process of production. The plant is fully equipped with 6 (six) criticality detectors in strategic places to alert employees whenever the postulated mass criticality accident occur. In the event of Nuclear Emergency Preparedness, PT BATAN TEKNOLOGI has also proposed the organization structure how promptly to report the crisis to Nuclear Energy Control Board (BAPETEN) Indonesia. (author)

Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

2001-02-01

This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

International Nuclear Information System (INIS)

Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

2001-02-01

This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

The SCALE Web site: Resources for the worldwide nuclear criticality safety community

International Nuclear Information System (INIS)

Bowman, S.M.

2000-01-01

The Standardized Computer Analyses for Licensing Evaluations (SCALE) computer software system developed at Oak Ridge National Laboratory (ORNL) is widely used and accepted around the world for criticality safety analyses. SCALE includes the well-known KENO V.a and KENO VI three-dimensional Monte Carlo criticality computer codes. For several years, the SCALE staff at ORNL has maintained a Web site to provide information and support to sponsors and users in the worldwide criticality safety community. The SCALE WEB site is located at www.cped.ornl.gov/scale and provides information in the following areas: 1. important notices to users; 2. SCALE Users Electronic Notebook; 3. current and past issues of the SCALE Newsletter; 4. verification and validation (V and V) and benchmark reports; 5. download updates, utilities, and V and V input files; 6. SCALE training course information; 7. SCALE Manual on-line; 8. overview of SCALE system; 9. how to install and run SCALE; 10. SCALE quality assurance documents; and 11. nuclear resources on the Internet

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

Realism in nuclear criticality safety

International Nuclear Information System (INIS)

McLaughlin, T. P.

2009-01-01

Commercial nuclear power plant operation and regulation have made remarkable progress since the Three Mile Island Accident. This is attributed largely to a heavy dose of introspection and self-regulation by the industry and to a significant infusion of risk-informed and performance-based regulation by the Nuclear Regulatory Commission. This truly represents reality in action both by the plant operators and the regulators. On the other hand, the implementation of nuclear criticality safety in ex-reactor operations involving significant quantities of fissile material has not progressed, but, tragically, it has regressed. Not only is the practice of the discipline in excess of a factor of ten more expensive than decades ago; the trend continues. This unfortunate reality is attributed to a lack of coordination within the industry (as contrasted to what occurred in the reactor operations sector), and to a lack of implementation of risk-informed and performance-based regulation by the NRC While the criticality safety discipline is orders of magnitude smaller than the reactor safety discipline, both operators and regulators must learn from the progress made in reactor safety and apply it to the former to reduce the waste, inefficiency and potentially increased accident risks associated with current practices. Only when these changes are made will there be progress made toward putting realism back into nuclear criticality safety. (authors)

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)

Introduction to 'International Handbook of Criticality Safety Benchmark Experiments'

International Nuclear Information System (INIS)

Komuro, Yuichi

1998-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated year by year by the working group of the project. This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used. The author briefly introduces the informative handbook and would like to encourage Japanese engineers who are in charge of nuclear criticality safety to use the handbook. (author)

Assessment CANDU physics codes using experimental data - part 1: criticality measurement

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok; Jeong, Chang Joon

2001-08-01

In order to assess the applicability of MCNP-4B code to the heavy water moderated, light water cooled and pressure-tube type reactor, the MCNP-4B physics calculations has been carried out for the Deuterium Critical Assembly (DCA), and the results were compared with those of the experimental data. In this study, the key safety parameters like as the multiplication factor, void coefficient, local power peaking factor and bundle power distribution in the scattered core are simulated. In order to use the cross section data consistently for the fuels to be analyzed in the future, new MCNP libraries have been generated from ENDF/B-VI release 3. Generally, the MCNP-4B calculation results show a good agreement with experimental data of DCA core. After benchmarking MCNP-4B against available experimental data, it will be used as the reference tool to benchmark design and analysis codes for the advanced CANDU fuels

Minimum qualifications for nuclear criticality safety professionals

International Nuclear Information System (INIS)

Ketzlach, N.

1990-01-01

A Nuclear Criticality Technology and Safety Training Committee has been established within the U.S. Department of Energy (DOE) Nuclear Criticality Safety and Technology Project to review and, if necessary, develop standards for the training of personnel involved in nuclear criticality safety (NCS). The committee is exploring the need for developing a standard or other mechanism for establishing minimum qualifications for NCS professionals. The development of standards and regulatory guides for nuclear power plant personnel may serve as a guide in developing the minimum qualifications for NCS professionals

SRTC criticality safety technical review: Nuclear Criticality Safety Evaluation 93-04 enriched uranium receipt

International Nuclear Information System (INIS)

Rathbun, R.

1993-01-01

Review of NMP-NCS-930087, open-quotes Nuclear Criticality Safety Evaluation 93-04 Enriched Uranium Receipt (U), July 30, 1993, close quotes was requested of SRTC (Savannah River Technology Center) Applied Physics Group. The NCSE is a criticality assessment to determine the mass limit for Engineered Low Level Trench (ELLT) waste uranium burial. The intent is to bury uranium in pits that would be separated by a specified amount of undisturbed soil. The scope of the technical review, documented in this report, consisted of (1) an independent check of the methods and models employed, (2) independent HRXN/KENO-V.a calculations of alternate configurations, (3) application of ANSI/ANS 8.1, and (4) verification of WSRC Nuclear Criticality Safety Manual procedures. The NCSE under review concludes that a 500 gram limit per burial position is acceptable to ensure the burial site remains in a critically safe configuration for all normal and single credible abnormal conditions. This reviewer agrees with that conclusion

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)

Nuclear Criticality Safety Data Book

Energy Technology Data Exchange (ETDEWEB)

Hollenbach, D. F. [Y-12 National Security Complex, Oak Ridge, TN (United States)

2016-11-14

The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

Nuclear Criticality Safety Data Book

International Nuclear Information System (INIS)

Hollenbach, D. F.

2016-01-01

The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

Criticality safety study of dry spent fuel cask loaded with increased enrichment fuel

International Nuclear Information System (INIS)

Bznuni, S.; Baghdasaryan, N.; Amirjanyan, A.

2013-01-01

Existing Dry Spent Fuel Casks (DSC) for transporting and storing of Armenian NPP fuel was licensed for WWER-440 fuel assemblies with 3.6% enrichment. Having in mind that ANPP introduced new fuel assemblies with increased enrichment (3.82 %) re-assessment of criticality safety analysis for DSC is required. Criticality safety analysis of DSC was performed by KENO-VI program using 238-GROUP ENDF/B-VII.0 LIBRARY (V7-238). Results of analysis showed that additional 8 borated racks for fuel assemblies should be included in the design of DSC. In addition feasibility study was performed to find out level of burnup-credit approach implementation to keep current design of DSC unchanged. Burnup-credit analysis was performed by STARBUCS program using axial burnup profiles from Armenian NPP neutronics analysis carried out by BIPR code. (authors)

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)

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.

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

Criticality Safety Evaluation of Hanford Tank Farms Facility

Energy Technology Data Exchange (ETDEWEB)

WEISS, E.V.

2000-12-15

Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

Criticality Safety Evaluation of Hanford Tank Farms Facility

International Nuclear Information System (INIS)

WEISS, E.V.

2000-01-01

Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste

DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

International Nuclear Information System (INIS)

C.E. Sanders

2005-01-01

This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

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

heavily on experience and engineering judgement, consistent with the ALARA philosophy. Special care is taken to ensure that the best estimate dose rates are used to the extent possible when applying ALARA. Provisions for safeguards equipment are made throughout the fuel-handling route in CANDU and ACR reactors. For example, the fuel bundle counters rely on the decay gammas from the fission products in spent-fuel bundles to record the number of fuel movements. The International Atomic Energy Agency (IAEA) Safeguards system for CANDU and ACR reactors is based on item (fuel bundle) accounting. It involves a combination of IAEA inspection with containment and surveillance, and continuous unattended monitoring. The spent fuel bundle counter monitors spent fuel bundles as they are transferred from the fuelling machine to the spent fuel bay. The shielding and dose-rate analysis need to be carried out so that the bundle counter functions properly. This paper includes two codes used in criticality safety analyses. Criticality safety is a unique phenomenon and codes that address criticality issues will demand specific validations. However, it is recognised that some of the codes used in radiation physics will also be used in criticality safety assessments. (authors)

Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

International Nuclear Information System (INIS)

Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

1993-01-01

Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking

Automated Translation of Safety Critical Application Software Specifications into PLC Ladder Logic

Science.gov (United States)

Leucht, Kurt W.; Semmel, Glenn S.

2008-01-01

The numerous benefits of automatic application code generation are widely accepted within the software engineering community. A few of these benefits include raising the 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 the NASA Kennedy Space Center (KSC) recognized the need for PLC code generation while developing their new ground checkout and launch processing system. They developed a process and a prototype software tool that automatically translates a high-level representation or specification of safety critical application software into ladder logic that executes on a PLC. This process and tool are expected to increase the reliability of the PLC code over that which is written manually, and may even lower life-cycle costs and shorten the development schedule of the new control system at KSC. This paper examines the problem domain and discusses the process and software tool that were prototyped by the KSC software engineers.

A Web-Based Nuclear Criticality Safety Bibliographic Database

International Nuclear Information System (INIS)

Koponen, B L; Huang, S

2007-01-01

A bibliographic criticality safety database of over 13,000 records is available on the Internet as part of the U.S. Department of Energy's (DOE) Nuclear Criticality Safety Program (NCSP) website. This database is easy to access via the Internet and gets substantial daily usage. This database and other criticality safety resources are available at ncsp.llnl.gov. The web database has evolved from more than thirty years of effort at Lawrence Livermore National Laboratory (LLNL), beginning with compilations of critical experiment reports and American Nuclear Society Transactions

K-effective as a measure of criticality safety

International Nuclear Information System (INIS)

Venner, J.; Haley, R.M.; Bowden, R.L.

2003-01-01

This paper considers the relation between the neutron multiplication of a system, k-effective, and critical parameters. It aims to investigate whether k-effective is always the most appropriate measure of safety. For simple systems handbook data can be effectively utilized, applying a safety factor to critical masses. In such situations, the criticality safety margin is readily apparent. However, more complex systems may use the calculated value of neutron multiplication to assess the criticality safety of the system under investigation. A problem arises because there is no exact consistency between k-effective and the physical margin of subcriticality, in terms of parameters such as mass. In the UK, commonly accepted safety criteria are applied to limit the k-effective of the system being assessed. These margins of subcriticality have no definitive justification to support the values chosen and might be considered rather arbitrary in nature. This paper aims to answer this question of suitability by investigating the relation between k-effective and the physical critical parameters for a wide range of systems. It concludes that the safety criteria currently applied in the UK are valid, but some difference exists between safety factors applied to the mass of fissile material present and the corresponding value of k-effective. (author)

Proceedings of the nuclear criticality technology safety project

Energy Technology Data Exchange (ETDEWEB)

Sanchez, R.G. [comp.

1997-06-01

This document contains summaries of the most of the papers presented at the 1994 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 10 and 11 at Williamsburg, Va. The meeting was broken up into seven sessions, which covered the following topics: (1) Validation and Application of Calculations; (2) Relevant Experiments for Criticality Safety; (3) Experimental Facilities and Capabilities; (4) Rad-Waste and Weapons Disassembly; (5) Criticality Safety Software and Development; (6) Criticality Safety Studies at Universities; and (7) Training. The minutes and list of participants of the Critical Experiment Needs Identification Workgroup meeting, which was held on May 9 at the same venue, has been included as an appendix. A second appendix contains the names and addresses of all NCTSP meeting participants. Separate abstracts have been indexed to the database for contributions to this proceedings.

Proceedings of the nuclear criticality technology safety project

International Nuclear Information System (INIS)

Sanchez, R.G.

1997-06-01

This document contains summaries of the most of the papers presented at the 1994 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 10 and 11 at Williamsburg, Va. The meeting was broken up into seven sessions, which covered the following topics: (1) Validation and Application of Calculations; (2) Relevant Experiments for Criticality Safety; (3) Experimental Facilities and Capabilities; (4) Rad-Waste and Weapons Disassembly; (5) Criticality Safety Software and Development; (6) Criticality Safety Studies at Universities; and (7) Training. The minutes and list of participants of the Critical Experiment Needs Identification Workgroup meeting, which was held on May 9 at the same venue, has been included as an appendix. A second appendix contains the names and addresses of all NCTSP meeting participants. Separate abstracts have been indexed to the database for contributions to this proceedings

Trustworthy Variant Derivation with Translation Validation for Safety Critical Product Lines

DEFF Research Database (Denmark)

Iosif-Lazăr, Alexandru Florin; Wasowski, Andrzej

2016-01-01

Software product line (SPL) engineering facilitates development of entire families of software products with systematic reuse. Model driven SPLs use models in the design and development process. In the safety critical domain, validation of models and testing of code increases the quality...... of the products altogether. However, to maintain this trustworthiness it is necessary to know that the SPL tools, which manipulate models and code to derive concrete product variants, do not introduce errors in the process. We propose a general technique of checking correctness of product derivation tools through...... translation validation. We demonstrate it using Featherweight VML—a core language for separate variability modeling relying on a single kind of variation point to define transformations of artifacts seen as object models. We use Featherweight VML with its semantics as a correctness specification...

HSE's safety assessment principles for criticality safety

International Nuclear Information System (INIS)

Simister, D N; Finnerty, M D; Warburton, S J; Thomas, E A; Macphail, M R

2008-01-01

The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf). (memorandum)

MCNPTM criticality primer and training experiences

International Nuclear Information System (INIS)

Briesmeister, J.; Forster, R.A.; Busch, R.

1995-01-01

With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, the analyst may have little experience with the specific codes available at his or her facility. Usually, the codes are quite complex, black boxes capable of analyzing numerous problems with a myriad of input options. Documentation for these codes is designed to cover all the possible configurations and types of analyses but does not give much detail on any particular type of analysis. For criticality calculations, the user of a code is primarily interested in the value of the effective multiplication factor for a system (k eff ). Most codes will provide this, and truckloads of other information that may be less pertinent to criticality calculations. Based on discussions with code users in the nuclear criticality safety community, it was decided that a simple document discussing the ins and outs of criticality calculations with specific codes would be quite useful. The Transport Methods Group, XTM, at Los Alamos National Laboratory (LANL) decided to develop a primer for criticality calculations with their Monte Carlo code, MCNP. This was a joint task between LANL with a knowledge and understanding of the nuances and capabilities of MCNP and the University of New Mexico with a knowledge and understanding of nuclear criticality safety calculations and educating first time users of neutronics calculations. The initial problem was that the MCNP manual just contained too much information. Almost everything one needs to know about MCNP can be found in the manual; the problem is that there is more information than a user requires to do a simple k eff calculation. The basic concept of the primer was to distill the manual to create a document whose only focus was criticality calculations using MCNP

New Improved Nuclear Data for Nuclear Criticality and Safety

International Nuclear Information System (INIS)

Guber, Klaus H.; Leal, Luiz C.; Lampoudis, C.; Kopecky, S.; Schillebeeckx, P.; Emiliani, F.; Wynants, R.; Siegler, P.

2011-01-01

The Geel Electron Linear Accelerator (GELINA) was used to measure neutron total and capture cross sections of 182,183,184,186 W and 63,65 Cu in the energy range from 100 eV to ∼200 keV using the time-of-flight method. GELINA is the only high-power white neutron source with excellent timing resolution and ideally suited for these experiments. Concerns about the use of existing cross-section data in nuclear criticality calculations using Monte Carlo codes and benchmarks were a prime motivator for the new cross-section measurements. To support the Nuclear Criticality Safety Program, neutron cross-section measurements were initiated using GELINA at the EC-JRC-IRMM. Concerns about data deficiencies in some existing cross-section evaluations from libraries such as ENDF/B, JEFF, or JENDL for nuclear criticality calculations were the prime motivator for new cross-section measurements. Over the past years many troubles with existing nuclear data have emerged, such as problems related to proper normalization, neutron sensitivity backgrounds, poorly characterized samples, and use of improper pulse-height weighting functions. These deficiencies may occur in the resolved- and unresolved-resonance region and may lead to erroneous nuclear criticality calculations. An example is the use of the evaluated neutron cross-section data for tungsten in nuclear criticality safety calculations, which exhibit discrepancies in benchmark calculations and show the need for reliable covariance data. We measured the neutron total and capture cross sections of 182,183,184,186 W and 63,65 Cu in the neutron energy range from 100 eV to several hundred keV. This will help to improve the representation of the cross sections since most of the available evaluated data rely only on old measurements. Usually these measurements were done with poor experimental resolution or only over a very limited energy range, which is insufficient for the current application.

Methods and tools used at the IPSN for the safety assessment of critical software

International Nuclear Information System (INIS)

Regnier, P.; Henry, J.Y.

1998-01-01

A significant feature of EDF's latest 1400MWe ''N4'' generation of pressurized water reactor (PWR) is the extensive use of computerized instrumentation and control, including a fully digital system for the reactor protection function. For the safety assessment of the software driving the operation of this digital reactor protection called SPIN, IPSN has developed and implemented a set of methods and tools. Using the lessons learned from this experience, IPSN has worked at improving those methods and tools, mainly trying to make them more automatic to use, and has participated in an international assessment exercise to test some other methods and tools, either new products on the market or self-developed products. As a result of these works, this paper presents an up to date overview of the IPSN methods and tools used for the assessment of safety critical software. This assessment, which consists of an analysis of all the documentation associated with the technical specifications and of a representative set of functions, is usually carried out in five steps: (1) critical examination of the documents, (2) evaluation of the quality of the code, (3) determination of the critical software components, (4) development of test cases and choice of testing strategy, (5) dynamic analysis (consistency and robustness). This paper also presents methods and tools developed or implemented by IPSN in order to: evaluate the completeness and consistency of specification and design documents written in natural language; build a model and simulate specification or design items; evaluate the quality of the source code; carry out FMEA analysis; run the binary code and perform tests (CLAIRE); perform random or mutational tests. (author)

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.

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

Software Safety Risk in Legacy Safety-Critical Computer Systems

Science.gov (United States)

Hill, Janice L.; Baggs, Rhoda

2007-01-01

Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

Use of a Web Site to Enhance Criticality Safety Training

International Nuclear Information System (INIS)

Huang, S T; Morman, J

2003-01-01

Currently, a website dedicated to enhancing communication and dissemination of criticality safety information is sponsored by the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This website was developed as part of the DOE response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. The website is the focal point for DOE nuclear criticality safety (NCS) activities, resources and references, including hyperlinks to other sites actively involved in the collection and dissemination of criticality safety information. The website is maintained by the Lawrence Livermore National Laboratory (LLNL) under auspices of the NCSP management. One area of the website contains a series of Nuclear Criticality Safety Engineer Training (NCSET) modules. During the past few years, many users worldwide have accessed the NCSET section of the NCSP website and have downloaded the training modules as an aid for their training programs. This trend was remarkable in that it points out a continuing need of the criticality safety community across the globe. It has long been recognized that training of criticality safety professionals is a continuing process involving both knowledge-based training and experience-based operations floor training. As more of the experienced criticality safety professionals reach retirement age, the opportunities for mentoring programs are reduced. It is essential that some method be provided to assist the training of young criticality safety professionals to replenish this limited human expert resource to support on-going and future nuclear operations. The main objective of this paper is to present the features of the NCSP website, including its mission, contents, and most importantly its use for the dissemination of training modules to the criticality safety community. We will discuss lessons learned and several ideas

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Jang-Soo; Jee, Eunkyoung

2016-01-01

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee-Choon; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jee, Eunkyoung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents.

Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

2001-08-15

Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

Criticality Safety Basics for INL FMHs and CSOs

Energy Technology Data Exchange (ETDEWEB)

V. L. Putman

2012-04-01

Nuclear power is a valuable and efficient energy alternative in our energy-intensive society. However, material that can generate nuclear power has properties that require this material be handled with caution. If improperly handled, a criticality accident could result, which could severely harm workers. This document is a modular self-study guide about Criticality Safety Principles. This guide's purpose it to help you work safely in areas where fissionable nuclear materials may be present, avoiding the severe radiological and programmatic impacts of a criticality accident. It is designed to stress the fundamental physical concepts behind criticality controls and the importance of criticality safety when handling fissionable materials outside nuclear reactors. This study guide was developed for fissionable-material-handler and criticality-safety-officer candidates to use with related web-based course 00INL189, BEA Criticality Safety Principles, and to help prepare for the course exams. These individuals must understand basic information presented here. This guide may also be useful to other Idaho National Laboratory personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. This guide also includes additional information that will not be included in 00INL189 tests. The additional information is in appendices and paragraphs with headings that begin with 'Did you know,' or with, 'Been there Done that'. Fissionable-material-handler and criticality-safety-officer candidates may review additional information at their own discretion. This guide is revised as needed to reflect program changes, user requests, and better information. Issued in 2006, Revision 0 established the basic text and integrated various programs from former contractors. Revision 1 incorporates operation and program changes implemented since 2006. It also incorporates suggestions, clarifications

Nuclear criticality safety program at the Fuel Cycle Facility

International Nuclear Information System (INIS)

Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

1994-01-01

The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

ICNC2003: Proceedings of the seventh international conference on nuclear criticality safety. Challenges in the pursuit of global nuclear criticality safety

International Nuclear Information System (INIS)

2003-10-01

This proceedings contain (technical, oral and poster papers) presented papers at the Seventh International Conference on Nuclear Criticality Safety ICNC2003 held on 20-24 October 2003, in Tokai, Ibaraki, Japan, following ICNC'99 in Versailles, France. The theme of this conference is 'Challenges in the Pursuit of Global Nuclear Criticality Safety'. This proceedings represent the current status of nuclear criticality safety research throughout the world. The 81 of the presented papers are indexed individually. (J.P.N.)

ICNC2003: Proceedings of the seventh international conference on nuclear criticality safety. Challenges in the pursuit of global nuclear criticality safety

International Nuclear Information System (INIS)

2003-10-01

This proceedings contain (technical, oral and poster papers) presented papers at the Seventh International Conference on Nuclear Criticality Safety ICNC2003 held on 20-24 October 2003, in Tokai, Ibaraki, Japan, following ICNC'99 in Versailles, France. The theme of this conference is 'Challenges in the Pursuit of Global Nuclear Criticality Safety'. This proceedings represent the current status of nuclear criticality safety research throughout the world. The 79 of the presented papers are indexed individually. (J.P.N.)

CRITICALITY SAFETY LIMIT EVALUATION PROGRAM (CSLEP's) AND QUICK SCREENS: ANSWERS TO EXPEDITED PROCESSING LEGACY CRITICALITY SAFETY LIMITS AND EVALUATIONS

International Nuclear Information System (INIS)

TOFFER, H.

2006-01-01

Since the end of the cold war, the need for operating weapons production facilities has faded. Criticality Safety Limits and controls supporting production modes in these facilities became outdated and furthermore lacked the procedure based rigor dictated by present day requirements. In the past, in many instances, the formalism of present day criticality safety evaluations was not applied. Some of the safety evaluations amounted to a paragraph in a notebook with no safety basis and questionable arguments with respect to double contingency criteria. When material stabilization, clean out, and deactivation activities commenced, large numbers of these older criticality safety evaluations were uncovered with limits and controls backed up by tenuous arguments. A dilemma developed: on the one hand, cleanup activities were placed on very aggressive schedules; on the other hand, a highly structured approach to limits development was required and applied to the cleanup operations. Some creative approaches were needed to cope with the limits development process

Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

J. D. Bess; J. B. Briggs; A. S. Garcia

2011-09-01

One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

International Nuclear Information System (INIS)

Bess, J.D.; Briggs, J.B.; Garcia, A.S.

2011-01-01

One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

Statistical methods for accurately determining criticality code bias

International Nuclear Information System (INIS)

Trumble, E.F.; Kimball, K.D.

1997-01-01

A system of statistically treating validation calculations for the purpose of determining computer code bias is provided in this paper. The following statistical treatments are described: weighted regression analysis, lower tolerance limit, lower tolerance band, and lower confidence band. These methods meet the criticality code validation requirements of ANS 8.1. 8 refs., 5 figs., 4 tabs

Criticality calculations for safety analysis

International Nuclear Information System (INIS)

Vellozo, S.O.

1981-01-01

Criticality studies in uranium nitrate and plutonium nitrate aqueous solutions were done. For uranium compound three basic computer codes are used: GAMTEC-II, DTF-IV, KENO-IV. Water was used as refletor and the results obtained with the different computer codes were analyzed and compared with the 'Handbuck zur Kriticalitat'. The cross sections and the cylindrical geometry were generated by Gamtec-II computer code. In the second compound the thickness of the recipient with plutonium nitrate are used with rectangular geometry and concret reflector. The effective multiplication constant was calculated with the Gamtec-II and Keno-IV library. The results show many differences. (E.G) [pt

Prerequisites of ideal safety-critical organizations

International Nuclear Information System (INIS)

Takeuchi, Michiru; Hikono, Masaru; Matsui, Yuko; Goto, Manabu; Sakuda, Hiroshi

2013-01-01

This study explores the prerequisites of ideal safety-critical organizations, marshalling arguments of 4 areas of organizational research on safety, each of which has overlap: a safety culture, high reliability organizations (HROs), organizational resilience, and leadership especially in safety-critical organizations. The approach taken in this study was to retrieve questionnaire items or items on checklists of the 4 research areas and use them as materials of abduction (as referred to in the KJ method). The results showed that the prerequisites of ideal safety-oriented organizations consist of 9 factors as follows: (1) The organization provides resources and infrastructure to ensure safety. (2) The organization has a sharable vision. (3) Management attaches importance to safety. (4) Employees openly communicate issues and share wide-ranging information with each other. (5) Adjustments and improvements are made as the organization's situation changes. (6) Learning activities from mistakes and failures are performed. (7) Management creates a positive work environment and promotes good relations in the workplace. (8) Workers have good relations in the workplace. (9) Employees have all the necessary requirements to undertake their own functions, and act conservatively. (author)

The Health and Safety Executive's regulatory framework for control of nuclear criticality safety

International Nuclear Information System (INIS)

Smith, K.; Simister, D.N.

1991-01-01

In the United Kingdom the Health and Safety at Work Act, 1974 is the main legal instrument under which risks to people from work activities are controlled. Certain sections of the Nuclear Installations Act, 1965 which deal with the licensing of nuclear sites and the regulatory control of risks arising from them, including the risk from accidental criticality, are relevant statutory provisions of the Health and Safety at Work Act. The responsibility for safety rests with the operator who has to make and implement arrangements to prevent accidental criticality. The adequacy of these arrangements must be demonstrated in a safety case to the regulatory authorities. Operators are encouraged to treat each plant on its own merits and develop the safety case accordingly. The Nuclear Installations Inspectorate (NII), for its part, assesses the adequacy of the operator's safety case against the industry's own standards and criteria, but more particularly against the NII's safety assessment principles and guides, and international standards. Risks should be made as low as reasonably practicable. Generally, the NII seeks improvements in safety using an enforcement policy which operates at a number of levels, ranging from persuasion through discussion to the ultimate deterrent of withdrawal of a site licence. This paper describes the role of the NII, which includes a specialist criticality expertise, within the Health and Safety Executive, in regulating the nuclear sites from the criticality safety viewpoint. (Author)

OFFSCALE: A PC input processor for the SCALE code system. The CSASIN processor for the criticality sequences

International Nuclear Information System (INIS)

Bowman, S.M.

1994-11-01

OFFSCALE is a suite of personal computer input processor programs developed at Oak Ridge National Laboratory to provide an easy-to-use interface for modules in the SCALE-4 code system. CSASIN (formerly known as OFFSCALE) is a program in the OFFSCALE suite that serves as a user-friendly interface for the Criticality Safety Analysis Sequences (CSAS) available in SCALE-4. It is designed to assist a SCALE-4 user in preparing an input file for execution of criticality safety problems. Output from CSASIN generates an input file that may be used to execute the CSAS control module in SCALE-4. CSASIN features a pulldown menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up a CSAS input file and perform data checking. This capability increases productivity and decreases the chance of user error

Assessment of critical flow models of RELAP5-MOD2 and CATHARE codes

International Nuclear Information System (INIS)

Hao Laomi; Zhu Zhanchuan

1992-01-01

The critical flow tests for the long and short nozzles conducted on the SUPER MOBY-DICK facility were analyzed using the RELAP5-MOD2 and CATHARE 1.3 codes to assess the critical flow models of two codes. The critical mass flux calculated for two nozzles are given. The CATHARE code has used the thermodynamic nonequilibrium sound velocity of the two-phase fluid as the critical flow criterion, and has the better interphase transfer models and calculates the critical flow velocities with the completely implicit solution. Therefore, it can well calculate the critical flowrate and can describe the effect of the geometry L/D on the critical flowrate

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)

Primer for criticality calculations with DANTSYS

International Nuclear Information System (INIS)

Busch, R.D.

1996-01-01

With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his or her facility. Typically, two types of codes are available: deterministic codes such as ANISN or DANTSYS that solve an approximate model exactly and Monte Carlo Codes such as KENO or MCNP that solve an exact model approximately. Often, the analyst feels that the deterministic codes are too simple and will not provide the necessary information, so most modeling uses Monte Carlo methods. This sometimes means that hours of effort are expended to produce results available in minutes from deterministic codes. A substantial amount of reliable information on nuclear systems can be obtained using deterministic methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico in cooperation with the Radiation Transport Group at Los Alamos National Laboratory has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. (DANTSYS is the name of a suite of codes that users more commonly know as ONEDANT, TWODANT, TWOHEX, and THREEDANT.) It assumes a college education in a technical field, but there is no assumption of familiarity with neutronics codes in general or with DANTSYS in particular. The primer is designed to teach by example, with each example illustrating two or three DANTSYS features useful in criticality analyses

Framatome-ANP France UO2 fuel fabrication. Criticality safety analysis in the light of the JCO accident

International Nuclear Information System (INIS)

Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.

2003-01-01

In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The FRAMATOME-ANP production of its French low enriched (5 w/o) UO2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach FRAMATOME-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (author)

On the Evaluation of Pebble Bead Reactor Critical Experiments Using the Pebbed Code

International Nuclear Information System (INIS)

Gougar, Hans D.; Sen, R. Sonat

2014-01-01

Critical experiments pose a particular but necessary challenge to validating pebble bed reactor design codes. Fuel and core heterogeneities, impurities in graphite, variable packing of pebbles, and moderately strong neutronic coupling are among the factors that inject uncertainty into the results obtained with lower fidelity core physics models. Some of these are addressed in this study. The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling

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

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

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

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

A Methodological Framework for Software Safety in Safety Critical Computer Systems

OpenAIRE

P. V. Srinivas Acharyulu; P. Seetharamaiah

2012-01-01

Software safety must deal with the principles of safety management, safety engineering and software engineering for developing safety-critical computer systems, with the target of making the system safe, risk-free and fail-safe in addition to provide a clarified differentaition for assessing and evaluating the risk, with the principles of software risk management. Problem statement: Prevailing software quality models, standards were not subsisting in adequately addressing the software safety ...

Criticality safety validation: Simple geometry, single unit 233U systems

International Nuclear Information System (INIS)

Putman, V.L.

1997-06-01

Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL 233 U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in 233 U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed 233 U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k eff calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va

Proceedings of KURRI symposium on criticality safety

International Nuclear Information System (INIS)

Nishina, Kojiro; Kanda, Keiji

1984-01-01

On August 8, 1984, at the Reactor Application Center of the Research Reactor Institute, Kyoto University, the symposium on criticality safety was held, and 81 participants from various fields of reactor physics, nuclear fuel cycle engineering, reactor chemistry, nuclear chemistry, health physics and so on discussed the problem. The gists of the presentation are collected in this report. The contents are the techniques of evaluating criticality safety in respective fuel facilities, the system of control and its concept, the course and plan of the research on criticality safety in Japan and foreign countries, the techniques of determining multiplication factor and so on, and the review of present status, the pointing-out of problems and the report of new techniques were made. The measures coping with criticality safety have been mostly to meet urgent demand, but its fundamental examination and long term research should be carried out. This symposium was planned as the preparation for such research project, and favorable comment was given by the participants. In the next symposium, it is considered better to limit the themes and to allot more time to respective lectures. (Kako, I.)

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)

SCALE Graphical Developments for Improved Criticality Safety Analyses

International Nuclear Information System (INIS)

Barnett, D.L.; Bowman, S.M.; Horwedel, J.E.; Petrie, L.M.

1999-01-01

New computer graphic developments at Oak Ridge National Ridge National Laboratory (ORNL) are being used to provide visualization of criticality safety models and calculational results as well as tools for criticality safety analysis input preparation. The purpose of this paper is to present the status of current development efforts to continue to enhance the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system. Applications for criticality safety analysis in the areas of 3-D model visualization, input preparation and execution via a graphical user interface (GUI), and two-dimensional (2-D) plotting of results are discussed

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)

Criticality Safety in the Handling of Fissile Material. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-05-15

This Safety Guide provides guidance and recommendations on how to meet the relevant requirements for ensuring subcriticality when dealing with fissile material and for planning the response to criticality accidents. The guidance and recommendations are applicable to both regulatory bodies and operating organizations. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences of this if it were to occur. The Safety Guide makes recommendations on how to ensure subcriticality in systems involving fissile materials during normal operation, anticipated operational occurrences, and, in the case of accident conditions, within design basis accidents, from initial design through commissioning, operation, and decommissioning and disposal.

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

New SCALE graphical interface for criticality safety

International Nuclear Information System (INIS)

Bowman, Stephen M.; Horwedel, James E.

2003-01-01

The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer software system developed at Oak Ridge National Laboratory is widely used and accepted around the world for criticality safety analyses. SCALE includes the well-known KENO V.a and KENO-VI three-dimensional (3-D) Monte Carlo criticality computer codes. One of the current development efforts aimed at making SCALE easier to use is the SCALE Graphically Enhanced Editing Wizard (GeeWiz). GeeWiz is compatible with SCALE 5 and runs on Windows personal computers. GeeWiz provides input menus and context-sensitive help to guide users through the setup of their input. It includes a direct link to KENO3D to allow the user to view the components of their geometry model as it is constructed. Once the input is complete, the user can click a button to run SCALE and another button to view the output. KENO3D has also been upgraded for compatibility with SCALE 5 and interfaces directly with GeeWiz. GeeWiz and KENO3D for SCALE 5 are planned for release in late 2003. The presentation of this paper is designed as a live demonstration of GeeWiz and KENO3D for SCALE 5. (author)

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.

Criticality Safety Information Resource Center Web portal: www.csirc.net

International Nuclear Information System (INIS)

Harmon, C.D. II; Jones, T.

2000-01-01

The Nuclear Criticality Safety Group (ESH-6) at Los Alamos National Laboratory (LANL) is in the process of collecting and archiving historical and technical information related to nuclear criticality safety from LANL and other facilities. In an ongoing effort, this information is being made available via the Criticality Safety Information Resource Center (CSIRC) web site, which is hosted and maintained by ESH-6 staff. Recently, the CSIRC Web site was recreated as a Web portal that provides the criticality safety community with much more than just archived data

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+.

Application of SN and Monte Carlo codes to the SHEBA critical assemblies

International Nuclear Information System (INIS)

O'Dell, R.D.

1993-01-01

The Solution High-Energy Burst Assembly (SHEBA) at Los Alamos is a low-enriched (4.95 wt. %) aqueous uranyl fluoride solution critical assembly. There are two SHEBA configurations, both consisting of right circular cylinders with a central control rod. The first configuration, hereafter called the old SHEBA, had a fuel solution diameter of 54.6 cm and a measured critical solution height of 36.5 cm. An improved modification, hereafter called the new SHEBA, has a fuel solution diameter of 48.9 cm but since it is not yet operational, the critical solution height has not yet been measured. In this presentation the application of the discrete-ordinates (S N ) code TWODANT using Hansen-Roach cross sections and the MCNP Monte Carlo code using continuous-energy cross sections for calculating the critical solution heights for both the old and new SHEBA assemblies is described. The code's predictions are compared and it is shown that a single calculation with a standard computer code may yield misleading results, especially when using a Monte Carlo code

Safe operation of critical assemblies and research reactors. Code of practice and Technical appendix. 1971 ed

International Nuclear Information System (INIS)

Cox, J.

1971-01-01

This book is in two parts. The first is a Code of Practice for the Safe Operation of Critical Assemblies and Research Reactors, prepared as a result of a meeting of experts which took place in Vienna on 20-24 May 1968. The Code has been prepared by the International Atomic Energy Agency in co-operation with the World Health Organization, and its publication is sponsored by both organizations. In addition, the Code was approved by the Board of Governors of the International Atomic Energy Agency on 16 December 1968 as part of the Agency's safety standards, which are applied to operations undertaken by Member States with the assistance of the Agency. The Board, in approving the publication of the present book, also recommended Member States to take the Code into account in the formulation of national regulations and recommendations. The second part of the book is a Technical Appendix to give information and illustrative samples that would be helpful in implementing the Code of Practice. This second part, although published under the same cover, is not part of the Code. An extensive Bibliography, amplifying the Technical Appendix, is included at the end.

Applications of PRA in nuclear criticality safety

International Nuclear Information System (INIS)

McLaughlin, T.P.

1992-01-01

Traditionally, criticality accident prevention at Los Alamos has been based on a thorough review and understanding of proposed operations of changes to operations, involving both process supervision and criticality safety staff. The outcome of this communication was usually an agreement, based on professional judgement, that certain accident sequences were credible and had to be reduced in likelihood either by administrative controls or by equipment design and others were not credible, and thus did not warrant expenditures to further reduce their likelihood. The extent of analysis and documentation was generally in proportion to the complexity of the operation but did not include quantified risk assessments. During the last three years nuclear criticality safety related Probabilistic Risk Assessments (PRAs) have been preformed on operations in two Los Alamos facilities. Both of these were conducted in order to better understand the cost/benefit aspects of PRA's as they apply to largely ''hands-on'' operations with fissile material for which human errors or equipment failures significant to criticality safety are both rare and unique. Based on these two applications and an appreciation of the historical criticality accident record (frequency and consequences) it is apparent that quantified risk assessments should be performed very selectively

A study of software safety analysis system for safety-critical software

International Nuclear Information System (INIS)

Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

2004-01-01

The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

Writing robust C++ code for critical applications

CERN Multimedia

CERN. Geneva

2015-01-01

**C++** is one of the most **complex**, expressive and powerful languages out there. However, its complexity makes it hard to write **robust** code. When using C++ to code **critical** applications, ensuring **reliability** is one of the key topics. Testing, debugging and profiling are all a major part of this kind of work. In the BE department we use C++ to write a big part of the controls system for beam operation, which implies putting a big focus on system stability and ensuring smooth operation. This talk will try to: - Highlight potential problems when writing C++ code, giving guidelines on writing defensive code that could have avoided such issues - Explain how to avoid common pitfalls (both in writing C++ code and at the debugging & profiling phase) - Showcase some tools and tricks useful to C++ development The attendees' proficiency in C++ should not be a concern. Anyone is free to join, even people that do not know C++, if only to learn the pitfalls a language may have. This may benefit f...

Nuclear criticality safety training: guidelines for DOE contractors

International Nuclear Information System (INIS)

Crowell, M.R.

1983-09-01

The DOE Order 5480.1A, Chapter V, Safety of Nuclear Facilities, establishes safety procedures and requirements for DOE nuclear facilities. This guide has been developed as an aid to implementing the Chapter V requirements pertaining to nuclear criticality safety training. The guide outlines relevant conceptual knowledge and demonstrated good practices in job performance. It addresses training program operations requirements in the areas of employee evaluations, employee training records, training program evaluations, and training program records. It also suggests appropriate feedback mechanisms for criticality safety training program improvement. The emphasis is on academic rather than hands-on training. This allows a decoupling of these guidelines from specific facilities. It would be unrealistic to dictate a universal program of training because of the wide variation of operations, levels of experience, and work environments among DOE contractors and facilities. Hence, these guidelines do not address the actual implementation of a nuclear criticality safety training program, but rather they outline the general characteristics that should be included

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

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

Martin Marietta Energy Systems Nuclear Criticality Safety Improvement Program

International Nuclear Information System (INIS)

Speas, I.G.

1987-01-01

This report addresses questions raised by criticality safety violation at several DOE plants. Two charts are included that define the severity and reporting requirements for the six levels of accidents. A summary is given of all reported criticality incident at the DOE plants involved. The report concludes with Martin Marietta's Nuclear Criticality Safety Policy Statement

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)

Phenomenological modeling of critical heat flux: The GRAMP code and its validation

International Nuclear Information System (INIS)

Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.

2013-01-01

Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate

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

Explicit Precedence Constraints in Safety-Critical Java

DEFF Research Database (Denmark)

Puffitsch, Wolfgang; Noulard, Eric; Pagetti, Claire

2013-01-01

Safety-critical Java (SCJ) aims at making the amenities of Java available for the development of safety-critical applications. The multi-rate synchronous language Prelude facilitates the specification of the communication and timing requirements of complex real-time systems. This paper combines...... to provide explicit support for precedence constraints. We present the considerations behind the design of this extension and discuss our experiences with a first prototype implementation based on the SCJ implementation of the Java Optimized Processor....

Continuous-energy version of KENO V.a for criticality safety applications

International Nuclear Information System (INIS)

Dunn, Michael E.; Greene, N. Maurice; Petrie, Lester M.

2003-01-01

KENO V.a is a multigroup Monte Carlo code that solves the Boltzmann transport equation and is used extensively in the criticality safety community to calculate the effective multiplication factor of systems with fissionable material. In this work, a continuous-energy or pointwise version of KENO V.a has been developed by first designing a new continuous-energy cross-section format and then by developing the appropriate Monte Carlo transport procedures to sample the new cross-section format. In order to generate pointwise cross sections for a test library, a series of cross-section processing modules were developed and used to process 50 ENDF/B-6 Release 7 nuclides for the test library. Once the cross-section processing procedures were in place, a continuous-energy version of KENO V.a was developed and tested by calculating 21 critical benchmark experiments. The point KENO-calculated results for the 21 benchmarks are in agreement with calculated results obtained with the multigroup version of KENO V.a using the 238-group ENDF/B-5 and 199-group ENDF/B-6 Release 3 libraries. Based on the calculated results with the prototypic cross-section library, a continuous-energy version of the KENO V.a code has been successfully developed and demonstrated for modeling systems with fissionable material. (author)

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

Historically, new entrants to the practice of nuclear criticality safety have learned their job primarily by on-the-job training (OJT) often by association with an experienced nuclear criticality safety engineer who probably also learned their job by OJT. Typically, the new entrant learned what he/she needed to know to solve a particular problem and accumulated experience as more problems were solved. It is likely that more formalism will be required in the future. Current US Department of Energy requirements for those positions which have to demonstrate qualification indicate that it should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis i's incompletely developed in some areas. Details of this analysis are provided in this report

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)

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

For non-reactor nuclear facilities, the U.S. Department of Energy (DOE) does not require that nuclear criticality safety engineers demonstrate qualification for their job. It is likely, however, that more formalism will be required in the future. Current DOE requirements for those positions which do have to demonstrate qualification indicate that qualification should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis is incompletely developed in some areas

Criticality safety (prospect of study in NUCEF)

International Nuclear Information System (INIS)

Itagaki, Masafumi

1996-01-01

Experimental studies of criticality safety are under way using STACY and TRACY in NUCEF. Collection of fundamental data on criticality in a solution system is undergoing with STACY to confirm that the likelihood of criticality safety in the system constructed on the assumption of apparatuses in a reprocessing plant is enough large. Whereas some experiments simulating criticality accidents in a reprocessing plant using TRACY were designed to investigate the behaviors of fuel solution and radioactive matters in order to clarify whether it is possible to safely shut them in the facility even if a critical accident occurs. Both STACY and TRACY reached the criticality in 1995. Up to now a series of criticality experiments have been done using STACY with a core tank φ60 cm and the first periodical examination is now under way. On the other hand, we have a plan using TRACY to investigate the behaviors of nuclear heat solution at a criticality accident, and the releasing, transfer and deposition of radioactive materials. After reaching the criticality for the first, the performance verification test has been conducted. The full-scale study using TRACY is planned to begin in the second half of 1996. (M.N.)

Monte Carlo code criticality benchmark comparisons for waste packaging

International Nuclear Information System (INIS)

Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.

1992-07-01

COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented

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

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)

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

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.

Use of a web site to enhance criticality safety training

International Nuclear Information System (INIS)

Huang, Song T.; Morman, James A.

2003-01-01

Establishment of the NCSP (Nuclear Criticality Safety Program) website represents one attempt by the NCS (Nuclear Criticality Safety) community to meet the need to enhance communication and disseminate NCS information to a wider audience. With the aging work force in this important technical field, there is a common recognition of the need to capture the corporate knowledge of these people and provide an easily accessible, web-based training opportunity to those people just entering the field of criticality safety. A multimedia-based site can provide a wide range of possibilities for criticality safety training. Training modules could range from simple text-based material, similar to the NCSET (Nuclear Criticality Safety Engineer Training) modules, to interactive web-based training classes, to video lecture series. For example, the Los Alamos National Laboratory video series of interviews with pioneers of criticality safety could easily be incorporated into training modules. Obviously, the development of such a program depends largely upon the need and participation of experts who share the same vision and enthusiasm of training the next generation of criticality safety engineers. The NCSP website is just one example of the potential benefits that web-based training can offer. You are encouraged to browse the NCSP website at http://ncsp.llnl.gov. We solicit your ideas in the training of future NCS engineers and welcome your participation with us in developing future multimedia training modules. (author)

The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project

International Nuclear Information System (INIS)

Hopper, Calvin Mitchell

2011-01-01

In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT and SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT and SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National

Present status of Japanese Criticality Safety Handbook

International Nuclear Information System (INIS)

Okuno, Hiroshi

1999-01-01

A draft of the second edition of Nuclear Criticality Safety Handbook has been finalized, and it is under examination by reviewing committee for JAERI Report. Working Group designated for revising the Japanese Criticality Safety Handbook, which is chaired by Prof. Yamane, is now preparing for 'Guide on Burnup Credit for Storage and Transport of Spent Nuclear Fuel' and second edition of 'Data Collection' part of Handbook. Activities related to revising the Handbook might give a hint for a future experiment at STACY. (author)

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)

Supplement report to the Nuclear Criticality Safety Handbook of Japan

International Nuclear Information System (INIS)

Okuno, Hiroshi; Komuro, Yuichi; Nakajima, Ken

1995-10-01

Supplementing works to 'The Nuclear Criticality Safety Handbook' of Japan have been continued since 1988, the year the handbook edited by the Science and Technology Agency first appeared. This report publishes the fruits obtained in the supplementing works. Substantial improvements are made in the chapters of 'Modelling the evaluation object' and 'Methodology for analytical safety assessment', and newly added are chapters of 'Criticality safety of chemical processes', 'Criticality accidents and their evaluation methods' and 'Basic principles on design and installation of criticality alarm system'. (author)

Criticality safety and facility design considerations

International Nuclear Information System (INIS)

Waltz, W.R.

1991-06-01

Operations with fissile material introduce the risk of a criticality accident that may be lethal to nearby personnel. In addition, concerns over criticality safety can result in substantial delays and shutdown of facility operations. For these reasons, it is clear that the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The emphasis of this report will be placed on engineering design considerations in the prevention of criticality. The discussion will not include other important aspects, such as the physics of calculating limits nor criticality alarm systems

Development and experimental qualification of the new safety-criticality CRISTAL package

International Nuclear Information System (INIS)

Mattera, Ch.

1998-11-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA 93) and the newest accurate codes (APOLLO 2, MORET 4, TRIPOLI 4). The CRISTAL system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route. To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OECD working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

Development and experimental testing of the new safety-criticality Cristal package

International Nuclear Information System (INIS)

Mattera, Ch.

1998-01-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA93) and the newest accurate codes (APOLLO2, MORET4, TRIPOLI4). The cristal system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route.) To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OCDE working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

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.

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

Fission, critical mass and safety-a historical review

International Nuclear Information System (INIS)

Meggitt, Geoff

2006-01-01

Since the discovery of fission, the notion of a chain reaction in a critical mass releasing massive amounts of energy has haunted physicists. The possibility of a bomb or a reactor prompted much of the early work on determining a critical mass, but the need to avoid an accidental critical excursion during processing or transport of fissile material drove much that took place subsequently. Because of the variety of possible situations that might arise, it took some time to develop adequate theoretical tools for criticality safety and the early assessments were based on direct experiment. Some extension of these experiments to closely similar situations proved possible, but it was not until the 1960s that theoretical methods (and computers to run them) developed enough for them to become reliable assessment tools. Validating such theoretical methods remained a concern, but by the end of the century they formed the backbone of criticality safety assessment. This paper traces the evolution of these methods, principally in the UK and USA, and summarises some related work concerned with the nature of criticality accidents and their radiological consequences. It also indicates how the results have been communicated and used in ensuring nuclear safety. (review)

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

USAEC Controls for Nuclear Criticality Safety

Energy Technology Data Exchange (ETDEWEB)

McCluggage, W. C. [Division of Operational Safety, United States Atomic Energy Commission Washington, DC (United States)

1966-05-15

This is a paper written to provide a broad general view of the United States Atomic Energy Commission's controls for nuclear criticality safety within its own facilities. Included also is a brief' discussion of the USAEC's methods of obtaining assurance that the controls are being applied. The body of the document contains three sections. The first two describe the functions of the USAEC; the third deals with the contractors. The provisions of the Atomic Energy Act applicable to health and safety are discussed in relation to nuclear criticality safety. The use of United States Atomic Energy Commission manual chapters and Federal regulations is described. The functions of the USAEC Headquarters' offices and the operations offices are briefly outlined. Comments regarding the USAEC's inspection, auditing and appraisal programmes are included. Also briefly mentioned are the basic qualifications which must be met to become a contractor to possess and process or use fissionable materials. On the plant, factory or facility level the duties and responsibilities of industrial management are briefly outlined. The fundamental standards and their origin, together with the principal documents and guides are mentioned. The chief methods of control used by contractors operating large USAEC facilities and plants are described and compared. These include diagrams of how a typical nuclear criticality safety problem is handled from inception, design, construction and finally plant operation. Also included is a brief discussion of the contractors' methods of assuring strict employee compliance with the operating rules and limits. (author)

MORET: Version 4.B. A multigroup Monte Carlo criticality code

International Nuclear Information System (INIS)

Jacquet, Olivier; Miss, Joachim; Courtois, Gerard

2003-01-01

MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)

Criticality safety benchmarking of PASC-3 and ECNJEF1.1

International Nuclear Information System (INIS)

Li, J.

1992-09-01

To validate the code system PASC-3 and the multigroup cross section library ECNJEF1.1 on various applications many benchmarks are required. This report presents the results of critically safety benchmarking for five calculational and four experimental benchmarks. These benchmarks are related to the transport package of fissile materials such as spent fuel. The fissile nuclides in these benchmarks are 235 U and 239 Pu. The modules of PASC-3 which have been used for the calculations are BONAMI, NITAWL and KENO.5A. The final results for the experimental benchmarks do agree well with experimental data. For the calculational benchmarks the results presented here are in reasonable agreement with the results from other investigations. (author). 8 refs.; 20 figs.; 5 tabs

Consensus standards utilized and implemented for nuclear criticality safety in Japan

International Nuclear Information System (INIS)

Nomura, Yasushi; Okuno, Hiroshi; Naito, Yoshitaka

1996-01-01

The fundamental framework for the criticality safety of nuclear fuel facilities regulations is, in many advanced countries, generally formulated so that technical standards or handbook data are utilized to support the licensing safety review and to implement its guidelines. In Japan also, adequacy of the safety design of nuclear fuel facilities is checked and reviewed on the basis of licensing safety review guides. These guides are, first, open-quotes The Basic Guides for Licensing Safety Review of Nuclear Fuel Facilities,close quotes and as its subsidiaries, open-quotes The Uranium Fuel Fabrication Facility Licensing Safety Review Guidesclose quotes and open-quotes The Reprocessing Facility Licensing Safety Review Guides.close quotes The open-quotes Nuclear Criticality Safety Handbook close-quote of Japan and the Technical Data Collection are published and utilized to supply related data and information for the licensing safety review, such as for the Rokkasho reprocessing plant. The well-established technical standards and data abroad such as those by the American Nuclear Society and the American National Standards Institute are also utilized to complement the standards in Japan. The basic principles of criticality safety control for nuclear fuel facilities in Japan are duly stipulated in the aforementioned basic guides as follows: 1. Guide 10: Criticality control for a single unit; 2. Guide 11: Criticality control for multiple units; 3. Guide 12: Consideration for a criticality accident

Nuclear criticality safety staff training and qualifications at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Monahan, S.P.; McLaughlin, T.P.

1997-01-01

Operations involving significant quantities of fissile material have been conducted at Los Alamos National Laboratory continuously since 1943. Until the advent of the Laboratory's Nuclear Criticality Safety Committee (NCSC) in 1957, line management had sole responsibility for controlling criticality risks. From 1957 until 1961, the NCSC was the Laboratory body which promulgated policy guidance as well as some technical guidance for specific operations. In 1961 the Laboratory created the position of Nuclear Criticality Safety Office (in addition to the NCSC). In 1980, Laboratory management moved the Criticality Safety Officer (and one other LACEF staff member who, by that time, was also working nearly full-time on criticality safety issues) into the Health Division office. Later that same year the Criticality Safety Group, H-6 (at that time) was created within H-Division, and staffed by these two individuals. The training and education of these individuals in the art of criticality safety was almost entirely self-regulated, depending heavily on technical interactions between each other, as well as NCSC, LACEF, operations, other facility, and broader criticality safety community personnel. Although the Los Alamos criticality safety group has grown both in size and formality of operations since 1980, the basic philosophy that a criticality specialist must be developed through mentoring and self motivation remains the same. Formally, this philosophy has been captured in an internal policy, document ''Conduct of Business in the Nuclear Criticality Safety Group.'' There are no short cuts or substitutes in the development of a criticality safety specialist. A person must have a self-motivated personality, excellent communications skills, a thorough understanding of the principals of neutron physics, a safety-conscious and helpful attitude, a good perspective of real risk, as well as a detailed understanding of process operations and credible upsets

Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

International Nuclear Information System (INIS)

Tadano, Tomoaki

2016-01-01

The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

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

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

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

International Nuclear Information System (INIS)

Bess, John D.; Briggs, J. Blair; Nigg, David W.

2009-01-01

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

Burn-up credit in criticality safety of PWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

2014-12-15

Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

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

Nuclear Criticality Safety Organization qualification program. Revision 4

International Nuclear Information System (INIS)

Carroll, K.J.; Taylor, R.G.; Worley, C.A.

1997-01-01

The Nuclear Criticality Safety Organization (NCSO) is committed to developing and maintaining a staff of highly qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document defines the Qualification Program to address the NCSO technical and managerial qualification as required by the Y-12 Training Implementation Matrix (TIM). It is implemented through a combination of LMES plant-wide training courses and professional nuclear criticality safety training provided within the organization. This Qualification Program is applicable to technical and managerial NCSO personnel, including temporary personnel, sub-contractors and/or LMES employees on loan to the NCSO, who perform the NCS tasks or serve NCS-related positions as defined in sections 5 and 6 of this program

Nuclear criticality safety specialist training and qualification programs

International Nuclear Information System (INIS)

Hopper, C.M.

1993-01-01

Since the beginning of the Nuclear Criticality Safety Division of the American Nuclear Society (ANS) in 1967, the nuclear criticality safety (NCS) community has sought to provide an exchange of information at a national level to facilitate the education and development of NCS specialists. In addition, individual criticality safety organizations within government contractor and licensed commercial nonreactor facilities have developed training and qualification programs for their NCS specialists. However, there has been substantial variability in the content and quality of these program requirements and personnel qualifications, at least as measured within the government contractor community. The purpose of this paper is to provide a brief, general history of staff training and to describe the current direction and focus of US DOE guidance for the content of training and qualification programs designed to develop NCS specialists

The Analysis of SBWR Critical Power Bundle Using Cobrag Code

Directory of Open Access Journals (Sweden)

Yohannes Sardjono

2013-03-01

Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

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

A critical flow model for the Cathena thermalhydraulic code

International Nuclear Information System (INIS)

Popov, N.K.; Hanna, B.N.

1990-01-01

The calculation of critical flow rate, e.g., of choked flow through a break, is required for simulating a loss of coolant transient in a reactor or reactor-like experimental facility. A model was developed to calculate the flow rate through the break for given geometrical parameters near the break and fluid parameters upstream of the break for ordinary water, as well as heavy water, with or without non- condensible gases. This model has been incorporated in the CATHENA, one-dimensional, two-fluid thermalhydraulic code. In the CATHENA code a standard staggered-mesh, finite-difference representation is used to solve the thermalhydraulic equations. This model compares the fluid mixture velocity, calculated using the CATHENA momentum equations, with a critical velocity. When the mixture velocity is smaller than the critical velocity, the flow is assumed to be subcritical, and the model remains passive. When the fluid mixture velocity is higher than the critical velocity, the model sets the fluid mixture velocity equal to the critical velocity. In this paper the critical velocity at a link (momentum cell) is first estimated separately for single-phase liquid, two- phase, or single-phase gas flow condition at the upstream node (mass/energy cell). In all three regimes non-condensible gas can be present in the flow. For single-phase liquid flow, the critical velocity is estimated using a Bernoulli- type of equation, the pressure at the link is estimated by the pressure undershoot method

Administrative practices for nuclear criticality safety, ANSI/ANS-8.19-1996

International Nuclear Information System (INIS)

Smith, D.R.

1996-01-01

American National Standard, open-quotes Administrative Practices for Nuclear Criticality Safety,close quotes American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.19-1996, addresses the responsibilities of management, supervision, and the criticality safety staff in the administration of an effective criticality safety program. Characteristics of operating procedures, process evaluations, material control procedures, and emergency plans are discussed

Nuclear criticality safety. Chapter 0530 of AEC manual

International Nuclear Information System (INIS)

2006-01-01

The programme objectives of this chapter of the U.S. Atomic Energy Commission manual on nuclear criticality safety are to protect the health and safety of the public and of the government and contractor personnel working in plants that handle fissionable material and to protect public and private property from the consequences of a criticality accident occurring in AEC-owned plants and other AEC-contracted activities involving fissionable materials

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...

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

CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

International Nuclear Information System (INIS)

C.E. Sanders

2005-01-01

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

Criticality Safety Basics for INL Emergency Responders

Energy Technology Data Exchange (ETDEWEB)

Valerie L. Putman

2012-08-01

This document is a modular self-study guide about criticality safety principles for Idaho National Laboratory emergency responders. This guide provides basic criticality safety information for people who, in response to an emergency, might enter an area that contains much fissionable (or fissile) material. The information should help responders understand unique factors that might be important in responding to a criticality accident or in preventing a criticality accident while responding to a different emergency.

This study guide specifically supplements web-based training for firefighters (0INL1226) and includes information for other Idaho National Laboratory first responders. However, the guide audience also includes other first responders such as radiological control personnel.

For interested readers, this guide includes clearly marked additional information that will not be included on tests. The additional information includes historical examples (Been there. Done that.), as well as facts and more in-depth information (Did you know …).

INL criticality safety personnel revise this guide as needed to reflect program changes, user requests, and better information. Revision 0, issued May 2007, established the basic text. Revision 1 incorporates operation, program, and training changes implemented since 2007. Revision 1 increases focus on first responders because later responders are more likely to have more assistance and guidance from facility personnel and subject matter experts. Revision 1 also completely reorganized the training to better emphasize physical concepts behind the criticality controls that help keep emergency responders safe. The changes are based on and consistent with changes made to course 0INL1226.

Handbook on criticality. Vol. 1. Criticality and nuclear safety; Handbuch zur Kritikalitaet. Bd. 1. Kritikalitaet und nukleare Sicherheit

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-04-15

This handbook was prepared primarily with the aim to provide information to experts in industry, authorities or research facilities engaged in criticality-safety-related problems that will allow an adequate and rapid assessment of criticality safety issues already in the planning and preparation of nuclear facilities. However, it is not the intention of the authors of the handbook to offer ready solutions to complex problems of nuclear safety. Such questions have to remain subject to an in-depth analysis and assessment to be carried out by dedicated criticality safety experts. Compared with the previous edition dated December 1998, this handbook has been further revised and supplemented. The proven basic structure of the handbook remains unchanged. The handbook follows in some ways similar criticality handbooks or instructions published in the USA, UK, France, Japan and the former Soviet Union. The expedient use of the information given in this handbook requires a fundamental understanding of criticality and the terminology of nuclear safety. In Vol. 1, ''Criticality and Nuclear Safety'', therefore, first the most important terms and fundamentals are introduced and explained. Subsequently, experimental techniques and calculation methods for evaluating criticality problems are presented. The following chapters of Vol. 1 deal i. a. with the effect of neutron reflectors and absorbers, neutron interaction, measuring methods for criticality, and organisational safety measures and provide an overview of criticality-relevant operational experience and of criticality accidents and their potential hazardous impact. Vol. 2 parts 1 and 2 finally compile criticality parameters in graphical and tabular form. The individual graph sheets are provided with an initially explained set of identifiers, to allow the quick finding of the information of current interest. Part 1 includes criticality parameters for systems with {sup 235}U as fissile material, while part

Nuclear Criticality Technology and Safety Project parameter study database

International Nuclear Information System (INIS)

Toffer, H.; Erickson, D.G.; Samuel, T.J.; Pearson, J.S.

1993-03-01

A computerized, knowledge-screened, comprehensive database of the nuclear criticality safety documentation has been assembled as part of the Nuclear Criticality Technology and Safety (NCTS) Project. The database is focused on nuclear criticality parameter studies. The database has been computerized using dBASE III Plus and can be used on a personal computer or a workstation. More than 1300 documents have been reviewed by nuclear criticality specialists over the last 5 years to produce over 800 database entries. Nuclear criticality specialists will be able to access the database and retrieve information about topical parameter studies, authors, and chronology. The database places the accumulated knowledge in the nuclear criticality area over the last 50 years at the fingertips of a criticality analyst

Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

International Nuclear Information System (INIS)

Williams, R.A.

1995-01-01

The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

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)

How to interpret safety critical failures in risk and reliability assessments

International Nuclear Information System (INIS)

Selvik, Jon Tømmerås; Signoret, Jean-Pierre

2017-01-01

Management of safety systems often receives high attention due to the potential for industrial accidents. In risk and reliability literature concerning such systems, and particularly concerning safety-instrumented systems, one frequently comes across the term ‘safety critical failure’. It is a term associated with the term ‘critical failure’, and it is often deduced that a safety critical failure refers to a failure occurring in a safety critical system. Although this is correct in some situations, it is not matching with for example the mathematical definition given in ISO/TR 12489:2013 on reliability modeling, where a clear distinction is made between ‘safe failures’ and ‘dangerous failures’. In this article, we show that different interpretations of the term ‘safety critical failure’ exist, and there is room for misinterpretations and misunderstandings regarding risk and reliability assessments where failure information linked to safety systems are used, and which could influence decision-making. The article gives some examples from the oil and gas industry, showing different possible interpretations of the term. In particular we discuss the link between criticality and failure. The article points in general to the importance of adequate risk communication when using the term, and gives some clarification on interpretation in risk and reliability assessments.

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.

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.

Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

Directory of Open Access Journals (Sweden)

V. V. Galchenko

2010-12-01

Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

Criticality qualification of a new Monte Carlo code for reactor core analysis

International Nuclear Information System (INIS)

Catsaros, N.; Gaveau, B.; Jaekel, M.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.; Zisis, Th.

2009-01-01

In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

Validation of nuclear criticality safety software and 27 energy group ENDF/B-IV cross sections. Revision 1

International Nuclear Information System (INIS)

Lee, B.L. Jr.; D'Aquila, D.M.

1996-01-01

The original validation report, POEF-T-3636, was documented in August 1994. The document was based on calculations that were executed during June through August 1992. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This revision is written to clarify the margin of safety being used at Portsmouth for nuclear criticality safety calculations. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Lockheed Martin Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. For calculations of Portsmouth systems using the specified codes and systems covered by this validation, a maximum k eff including 2σ of 0.9605 or lower shall be considered as subcritical to ensure a calculational margin of safety of 0.02. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25

Safety culture and subcontractor network governance in a complex safety critical project

International Nuclear Information System (INIS)

Oedewald, Pia; Gotcheva, Nadezhda

2015-01-01

In safety critical industries many activities are currently carried out by subcontractor networks. Nevertheless, there are few studies where the core dimensions of resilience would have been studied in safety critical network activities. This paper claims that engineering resilience into a system is largely about steering the development of culture of the system towards better ability to anticipate, monitor, respond and learn. Thus, safety culture literature has relevance in resilience engineering field. This paper analyzes practical and theoretical challenges in applying the concept of safety culture in a complex, dynamic network of subcontractors involved in the construction of a new nuclear power plant in Finland, Olkiluoto 3. The concept of safety culture is in focus since it is widely used in nuclear industry and bridges the scientific and practical interests. This paper approaches subcontractor networks as complex systems. However, the management model of the Olkiluoto 3 project is to a large degree a traditional top-down hierarchy, which creates a mismatch between the management approach and the characteristics of the system to be managed. New insights were drawn from network governance studies. - Highlights: • We studied a relevant topical subject safety culture in nuclear new build project. • We integrated safety science challenges and network governance studies. • We produced practicable insights in managing safety of subcontractor networks

Criticality safety validation: Simple geometry, single unit {sup 233}U systems

Energy Technology Data Exchange (ETDEWEB)

Putman, V.L.

1997-06-01

Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL {sup 233}U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in {sup 233}U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed {sup 233}U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k{sub eff} calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va.

48 CFR 209.270 - Aviation and ship critical safety items.

Science.gov (United States)

2010-10-01

... Requirements 209.270 Aviation and ship critical safety items. ... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Aviation and ship critical safety items. 209.270 Section 209.270 Federal Acquisition Regulations System DEFENSE ACQUISITION...

Criticality safety benchmark evaluation project: Recovering the past

Energy Technology Data Exchange (ETDEWEB)

Trumble, E.F.

1997-06-01

A very brief summary of the Criticality Safety Benchmark Evaluation Project of the Westinghouse Savannah River Company is provided in this paper. The purpose of the project is to provide a source of evaluated criticality safety experiments in an easily usable format. Another project goal is to search for any experiments that may have been lost or contain discrepancies, and to determine if they can be used. Results of evaluated experiments are being published as US DOE handbooks.

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

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.

Product Engineering Class in the Software Safety Risk Taxonomy for Building Safety-Critical Systems

Science.gov (United States)

Hill, Janice; Victor, Daniel

2008-01-01

When software safety requirements are imposed on legacy safety-critical systems, retrospective safety cases need to be formulated as part of recertifying the systems for further use and risks must be documented and managed to give confidence for reusing the systems. The SEJ Software Development Risk Taxonomy [4] focuses on general software development issues. It does not, however, cover all the safety risks. The Software Safety Risk Taxonomy [8] was developed which provides a construct for eliciting and categorizing software safety risks in a straightforward manner. In this paper, we present extended work on the taxonomy for safety that incorporates the additional issues inherent in the development and maintenance of safety-critical systems with software. An instrument called a Software Safety Risk Taxonomy Based Questionnaire (TBQ) is generated containing questions addressing each safety attribute in the Software Safety Risk Taxonomy. Software safety risks are surfaced using the new TBQ and then analyzed. In this paper we give the definitions for the specialized Product Engineering Class within the Software Safety Risk Taxonomy. At the end of the paper, we present the tool known as the 'Legacy Systems Risk Database Tool' that is used to collect and analyze the data required to show traceability to a particular safety standard

The Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Henderson, B.D.; Meade, R.A.; Pruvost, N.L.

1999-01-01

The Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory (LANL) is a program jointly funded by the U.S. Department of Energy (DOE) and the U.S. Nuclear Regulatory Commission (NRC) in conjunction with the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2. The goal of CSIRC is to preserve primary criticality safety documentation from U.S. critical experimental sites and to make this information available for the benefit of the technical community. Progress in archiving criticality safety primary documents at the LANL archives as well as efforts to make this information available to researchers are discussed. The CSIRC project has a natural linkage to the International Criticality Safety Benchmark Evaluation Project (ICSBEP). This paper raises the possibility that the CSIRC project will evolve in a fashion similar to the ICSBEP. Exploring the implications of linking the CSIRC to the international criticality safety community is the motivation for this paper

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

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

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

Recommendations relating to safety-critical real-time software in nuclear power plants

International Nuclear Information System (INIS)

1992-01-01

The Advisory Committee on Nuclear Safety (ACNS) has reviewed safety issues associated with the software for the digital computers in the safety shutdown systems for the Darlington NGS. From this review the ACNS has developed four recommendations for safety-critical real-time software in nuclear power plants. These recommendations cover: the completion of the present efforts to develop an overall standard and sub-tier standards for safety-critical real-time software; the preparation of schedules and lists of responsibilities for this development; the concentration of AECB efforts on ensuring the scrutability of safety-critical real-time software; and, the collection of data on reliability and causes of failure (error) of safety-critical real-time software systems and on the probability and causes of common-mode failures (errors). (9 refs.)

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

Tank waste remediation system nuclear criticality safety program management review

International Nuclear Information System (INIS)

BRADY RAAP, M.C.

1999-01-01

This document provides the results of an internal management review of the Tank Waste Remediation System (TWRS) criticality safety program, performed in advance of the DOE/RL assessment for closure of the TWRS Nuclear Criticality Safety Issue, March 1994. Resolution of the safety issue was identified as Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-40-12, due September 1999

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

Agility in Development of Safety-Critical Software: A Conceptual Model

DEFF Research Database (Denmark)

Tordrup Heeager, Lise; Nielsen, Peter Axel

2018-01-01

Safety-critical information systems are being used increasingly as we see applications in new areas such as personal medical devices, traffic control and detection of pathogens. A current research debate is whether safety-critical systems must be developed with traditional waterfall processes...

Modification and application of the ATHLET-SC code to trans-critical simulations

International Nuclear Information System (INIS)

Fu, S.-W.; Zhou, C.; Xu, Z.-H.; Liu, X.-J.; Yang, Y.-H.; Cheng, H.

2011-01-01

In the simulation of trans-critical transients of Supercritical water cooled reactor (SCWR), calculation will terminate because of the sudden change in void fraction across the critical point. To solve this problem, a pseudo two-phase method is proposed with a virtual region of latent heat at pseudo-critical temperatures. A smooth variation of void fraction can be realized by using liquid-field conservation equations at temperatures lower than the pseudo-critical temperature, and vapor-field conservation equations at temperatures higher than the pseudo-critical temperature. Using this method, the system code ATHLET is modified to ATHLET-SC mod 2 on the basic of the previous modified version ATHLET-SC by Shanghai Jiao Tong University. The results of tests are verified that the calculation error with the pseudo two-phase method for supercritical fluid is acceptable, when the virtual region of latent heat is kept small. Moreover, the ATHLET-SC mod 2 code is used to simulate the pressurization and depressurization process of a single flow channel with the pressure transition as well as blowdown process. The results indicate a good applicability of the modified code. (author)

CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

C.E. Sanders

2005-04-07

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

Development and assessment of a sub-channel code applicable for trans-critical transient of SCWR

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2013-01-01

Highlights: • A new sub-channel code COBRA-SC for SCWR is developed. • Pseudo two-phase method is employed to realize trans-critical transient calculation. • Good suitability of COBRA-SC is demonstrated by preliminary assessment. • The calculation results of COBRA-SC agree well with ATHLET code. -- Abstract: In the last few years, extensive R and D activities have been launched covering various aspects of supercritical water-cooled reactor (SCWR), especially the thermal-hydraulic analysis. Sub-channel code plays an indispensable role to predict the detail thermal-hydraulic behavior of the SCWR fuel assembly. This paper develops a new version of sub-channel code COBRA-SC based on the previous COBRA-IV code. The supercritical water property and heat transfer/pressure drop correlations under supercritical pressure are implemented to this code. Moreover, in order to simulate the trans-critical transient (the pressure undergo a decrease from the supercritical pressure to the subcritical pressure), pseudo two-phase method is employed in COBRA-SC code. This work is completed by introduction of a virtual two-phase region near the pseudo-critical line. A smooth transition of void fraction can be realized. In addition, several heat transfer correlations right underneath the critical point are introduced into this code to capture the heat transfer behavior during the trans-critical transient. Some experimental data from simple geometry, e.g. the single tube, small rod bundle, is used to validate and evaluate this new developed COBRA-SC code. The predicted results show a good agreement with the experimental data, demonstrating good feasibility of this code for SCWR condition. A code to code comparison between COBRA-SC and ATHLET for a blowdown transient of a small fuel assembly is also presented and discussed in this paper

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)

Criticality codes migration to workstations at the Hanford site

International Nuclear Information System (INIS)

Miller, E.M.

1993-01-01

Westinghouse Hanford Company, Hanford Site Operations contractor, Richland, Washington, currently runs criticality codes on the Cray X-MP EA/232 computer but has recommended that US Department of Energy DOE-Richland replace the Cray with more economical workstations

FAST: An advanced code system for fast reactor transient analysis

International Nuclear Information System (INIS)

Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

2005-01-01

One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

Nuclear data for criticality safety

International Nuclear Information System (INIS)

Westfall, R.M.

1994-01-01

A brief overview is presented on emerging requirements for new criticality safety analyses arising from applications involving nuclear waste management, facility remediation, and the storage of nuclear weapons components. A derivation of criticality analyses from the specifications of national consensus standards is given. These analyses, both static and dynamic, define the needs for nuclear data. Integral data, used primarily for analytical validation, and differential data, used in performing the analyses, are listed, along with desirable margins of uncertainty. Examples are given of needs for additional data to address systems having intermediate neutron energy spectra and/or containing nuclides of intermediate mass number

Influence of safeguards and fire protection on criticality safety

International Nuclear Information System (INIS)

Six, D.E.

1980-01-01

There are several positive influences of safeguards and fire protection on criticality safety. Experts in each discipline must be aware of regulations and requirements of the others and work together to ensure a fault-tree design. EG and G Idaho, Inc., routinely uses an Occupancy-Use Readiness Manual to consider all aspects of criticality safety, fire protection, and safeguards. The use of the analytical tree is described

Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

Energy Technology Data Exchange (ETDEWEB)

Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

1995-12-31

The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

Challenges in the application of burn-up credit to the criticality safety of the THORP reprocessing plant

International Nuclear Information System (INIS)

Mayson, R.T.H.; Gunston, K.J.

1999-01-01

Since 1991 BNFL has made a significant investment in the development of the burn-up credit method and the application to its operations. It has recently demonstrated that using this method for the THORP dissolvers, it is possible to justify operating safety with reduced neutron poison concentrations and this has now been submitted to the regulators. The continued challenges the criticality safety community is facing are to show that we are not reducing safety levels because we are using burn-up credit. The burn-up credit method that has been developed can be summarized as follows. It consists of performing reactivity calculations for irradiated fuel using compositions generated by and inventory prediction code, generally in order to determine the limiting burn-up required for that fuel in a particular environment. In addition, it has always been envisaged that a confirmatory measurement of burn-up would be required to be made prior to certain operations such as the sharing of fuel into a dissolver. The burn-up credit method therefore relies upon three key components of inventory prediction, reactivity calculation code and the quantification and verification of burn-up. (J.P.N.)

Applications of PRA in nuclear criticality safety

International Nuclear Information System (INIS)

McLaughlin, T.P.

1992-01-01

Traditionally, criticality accident prevention at Los Alamos National Laboratory (LANL) has been based on a thorough review and understanding of proposed operations or changes to operations involving both process supervision and criticality safety staff. The outcome of this communication was usually an agreement, based on professional judgment, that certain accident sequences were credible and had to be precluded by design; others were incredible and thus did not warrant expenditures to further reduce their likelihood. The extent of documentation was generally in proportion to the complexity of the operation but never as detailed as that associated with quantified risk assessments. During the last 3 yr, nuclear criticality safety-related probabilistic risk assessments (PRAs) have been performed on operations in two LANL facilities. Both of these were conducted in order to better understand the cost/benefit aspects of PRAs as they apply to largely hands-on operations with fissile material

HELIOS: Application for criticality limits assessment

International Nuclear Information System (INIS)

Simeonov, T.

2011-01-01

In the early years, after the discovery of fission, the criticality safety assessment and the established safety limits, have been mainly based on direct experiments. Later, following the advances in the theory, computational methods and computer hardware, theoretical methods have been elaborated to the level to become reliable assessment tools. The computer codes started replacing the experiments, while the experimental data became a valuable validation source for their models. An application of the two-dimensional transport theory code HELIOS for assessment of criticality limits is presented in this paper. The effect of the enrichment, the system dimensions, H/U5 ration and different reflectors were studied in heterogeneous and homogenized systems. Comparisons with published experimental data and evaluated safety limits are made here to demonstrate the range of HELIOS applicability and limitations. (Author)

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)

Validation testing of safety-critical software

International Nuclear Information System (INIS)

Kim, Hang Bae; Han, Jae Bok

1995-01-01

A software engineering process has been developed for the design of safety critical software for Wolsung 2/3/4 project to satisfy the requirements of the regulatory body. Among the process, this paper described the detail process of validation testing performed to ensure that the software with its hardware, developed by the design group, satisfies the requirements of the functional specification prepared by the independent functional group. To perform the tests, test facility and test software were developed and actual safety system computer was connected. Three kinds of test cases, i.e., functional test, performance test and self-check test, were programmed and run to verify each functional specifications. Test failures were feedback to the design group to revise the software and test results were analyzed and documented in the report to submit to the regulatory body. The test methodology and procedure were very efficient and satisfactory to perform the systematic and automatic test. The test results were also acceptable and successful to verify the software acts as specified in the program functional specification. This methodology can be applied to the validation of other safety-critical software. 2 figs., 2 tabs., 14 refs. (Author)

Overview of Risk Mitigation for Safety-Critical Computer-Based Systems

Science.gov (United States)

Torres-Pomales, Wilfredo

2015-01-01

This report presents a high-level overview of a general strategy to mitigate the risks from threats to safety-critical computer-based systems. In this context, a safety threat is a process or phenomenon that can cause operational safety hazards in the form of computational system failures. This report is intended to provide insight into the safety-risk mitigation problem and the characteristics of potential solutions. The limitations of the general risk mitigation strategy are discussed and some options to overcome these limitations are provided. This work is part of an ongoing effort to enable well-founded assurance of safety-related properties of complex safety-critical computer-based aircraft systems by developing an effective capability to model and reason about the safety implications of system requirements and design.

Method of V ampersand V for safety-critical software in NPPs

International Nuclear Information System (INIS)

Kim, Jang-Yeol; Lee, Jang-Soo; Kwon, Kee-Choon

1997-01-01

Safety-critical software is software used in systems in which a failure could affect personal or equipment safety or result in large financial or social loss. Examples of systems using safety-critical software are systems such as plant protection systems in nuclear power plants (NPPs), process control systems in chemical plants, and medical instruments such as the Therac-25 medical accelerator. This paper presents verification and validation (V ampersand V) methodology for safety-critical software in NPP safety systems. In addition, it addresses issues related to NPP safety systems, such as independence parameters, software safety analysis (SSA) concepts, commercial off-the-shelf (COTS) software evaluation criteria, and interrelationships among software and system assurance organizations. It includes the concepts of existing industrial standards on software V ampersand V, Institute of Electrical and Electronics Engineers (IEEE) Standards 1012 and 1059. This safety-critical software V ampersand V methodology covers V ampersand V scope, a regulatory framework as part of its acceptance criteria, V ampersand V activities and task entrance and exit criteria, reviews and audits, testing and quality assurance records of V ampersand V material, configuration management activities related to V ampersand V, and software V ampersand V (SVV) plan (SVVP) production

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

Review of WHC criticality safety audit findings for 1970-1981

International Nuclear Information System (INIS)

Rogers, C.A.; Paglieri, J.N.

1984-01-01

At Westinghouse Hanford Company (WHC) all fissionable material handling must meet DOE requirements for safety. This necessitates a program of regular audits by the Safety group to verify compliance with criticality safety limits and controls and to alert facility management to observed discrepancies and potential problems. Audits of fissionable material facilities by Safety are required at least once every 6 months, but in practice are conducted more frequently. This paper summarizes findings from over 400 criticality safety audits conducted by Safety between July 1970 and July 1981 in seven fissionable material facilities to show their types and frequencies of occurrence. All limit violations occurring during this period are summarized, including those found by the operating group. 1 ref., 1 tab

Developing guidance in the nuclear criticality safety assessment for fuel cycle facilities

International Nuclear Information System (INIS)

Galet, C.; Evo, S.

2012-01-01

In this poster IRSN (Institute for radiation protection and nuclear safety) presents its safety guides whose purpose is to transmit the safety assessment know-how to any 'junior' staff or even to give a view of the safety approach on the overall risks to any staff member. IRSN has written a first version of such a safety guide for fuel cycle facilities and laboratories. It is organized into several chapters: some refer to types of assessments, others concern the types of risks. Currently, this guide contains 13 chapters and each chapter consists of three parts. In parallel to the development of criticality chapter of this guide, the IRSN criticality department has developed a nuclear criticality safety guide. It follows the structure of the three parts fore-mentioned, but it presents a more detailed first part and integrates, in the third part, the experience feedback collected on nuclear facilities. The nuclear criticality safety guide is online on the IRSN's web site

Research on neutron source multiplication method in nuclear critical safety

International Nuclear Information System (INIS)

Zhu Qingfu; Shi Yongqian; Hu Dingsheng

2005-01-01

The paper concerns in the neutron source multiplication method research in nuclear critical safety. Based on the neutron diffusion equation with external neutron source the effective sub-critical multiplication factor k s is deduced, and k s is different to the effective neutron multiplication factor k eff in the case of sub-critical system with external neutron source. The verification experiment on the sub-critical system indicates that the parameter measured with neutron source multiplication method is k s , and k s is related to the external neutron source position in sub-critical system and external neutron source spectrum. The relation between k s and k eff and the effect of them on nuclear critical safety is discussed. (author)

Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

International Nuclear Information System (INIS)

Dunn, M.E.; Greene, N.M.; Leal, L.C.

1999-01-01

Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

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

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

Analysis of Critical Characteristics for Safety Graded Personnel Computers in the KNICS Architecture

International Nuclear Information System (INIS)

Lee, Hyun Chul; Lee, Dong Young

2009-01-01

Critical characteristics analysis of a safety related item is to identify characteristics to be verified to replace an original item with the dedicated item. It is sure that the dedicated item meeting critical characteristics would perform its intended safety function instead of the specified item. KNICS project developed two safety systems: IDiPS RPS (Reactor Protection System) and IDiPS ESF-CCS (Engineered Safety Features-Component Control System). Two safety systems of IDiPS are equipped with personnel computers, so-called COMs (Cabinet Operator Modules), in their cabinets. The personnel computers, COMs, are responsible for safety system monitoring, testing, and maintaining. Even though two safety systems are safety critical system, the personnel computers of two systems, i.e. COMs, are not graded as safety-graded items. Regulation requirements are expected to be strengthened, and the functions of the personnel computer may be enhanced to include safety-related functions and safety functions, it would be necessary that the grade of the personnel computers is adjusted to a higher level, the safety grade. To try to upgrade a non safety system, i.e. COMs, to a safety system, its safety functions and requirements, i.e. critical characteristics, must be identified and verified. This paper describes the process of the identification of critical characteristics and the results of analysis

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)

Criticality safety training at the Hot Fuel Examination Facility

International Nuclear Information System (INIS)

Garcia, A.S.; Courtney, J.C.; Thelen, V.N.

1983-01-01

HFEF comprises four hot cells and out-of-cell support facilities for the US breeder program. The HFEF criticality safety program includes training in the basic theory of criticality and in specific criticality hazard control rules that apply to HFEF. A professional staff-member oversees the implementation of the criticality prevention program

Design Information from the PSA for Digital Safety-Critical Systems

International Nuclear Information System (INIS)

Kang, Hyun Gook; Jang, Seung Cheol

2005-01-01

Many safety-critical applications such as nuclear field application usually adopt a similar design strategy for digital safety-critical systems. Their differences from the normal design for the non-safety-critical applications could be summarized as: multiple-redundancy, highly reliable components, strengthened monitoring mechanism, verified software, and automated test procedure. These items are focusing on maintaining the capability to perform the given safety function when it is requested. For the past several decades, probabilistic safety assessment (PSA) techniques are used in the nuclear industry to assess the relative effects of contributing events on plant risk and system reliability. They provide a unifying means of assessing physical faults, recovery processes, contributing effects, human actions, and other events that have a high degree of uncertainty. The applications of PSA provide not only the analysis results of already installed system but also the useful information for the system under design. The information could be derived from the PSA experience of the various safety-critical systems. Thanks to the design flexibility, the digital system is one of the most suitable candidates for risk-informed design (RID). In this article, we will describe the feedbacks for system design and try to develop a procedure for RID. Even though the procedure is not sophisticated enough now, it could be the start point of the further investigation for developing more complete and practical methodology

Quantitative reliability assessment for safety critical system software

International Nuclear Information System (INIS)

Chung, Dae Won; Kwon, Soon Man

2005-01-01

An essential issue in the replacement of the old analogue I and C to computer-based digital systems in nuclear power plants is the quantitative software reliability assessment. Software reliability models have been successfully applied to many industrial applications, but have the unfortunate drawback of requiring data from which one can formulate a model. Software which is developed for safety critical applications is frequently unable to produce such data for at least two reasons. First, the software is frequently one-of-a-kind, and second, it rarely fails. Safety critical software is normally expected to pass every unit test producing precious little failure data. The basic premise of the rare events approach is that well-tested software does not fail under normal routine and input signals, which means that failures must be triggered by unusual input data and computer states. The failure data found under the reasonable testing cases and testing time for these conditions should be considered for the quantitative reliability assessment. We will present the quantitative reliability assessment methodology of safety critical software for rare failure cases in this paper

SACS2: Dynamic and Formal Safety Analysis Method for Complex Safety Critical System

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun

2009-01-01

Fault tree analysis (FTA) is one of the most widely used safety analysis technique in the development of safety critical systems. However, over the years, several drawbacks of the conventional FTA have become apparent. One major drawback is that conventional FTA uses only static gates and hence can not capture dynamic behaviors of the complex system precisely. Although several attempts such as dynamic fault tree (DFT), PANDORA, formal fault tree (FFT) and so on, have been made to overcome this problem, they can not still do absolute or actual time modeling because they adapt relative time concept and can capture only sequential behaviors of the system. Second drawback of conventional FTA is its lack of rigorous semantics. Because it is informal in nature, safety analysis results heavily depend on an analyst's ability and are error-prone. Finally reasoning process which is to check whether basic events really cause top events is done manually and hence very labor-intensive and timeconsuming for the complex systems. In this paper, we propose a new safety analysis method for complex safety critical system in qualitative manner. We introduce several temporal gates based on timed computational tree logic (TCTL) which can represent quantitative notion of time. Then, we translate the information of the fault trees into UPPAAL query language and the reasoning process is automatically done by UPPAAL which is the model checker for time critical system

Applications of probabilistic risk analysis in nuclear criticality safety design

International Nuclear Information System (INIS)

Chang, J.K.

1992-01-01

Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

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

Role of criticality models in ANSI standards for nuclear criticality safety

International Nuclear Information System (INIS)

Thomas, J.T.

1976-01-01

Two methods used in nuclear criticality safety evaluations in the area of neutron interaction among subcritical components of fissile materials are the solid angle and surface density techniques. The accuracy and use of these models are briefly discussed

Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations

International Nuclear Information System (INIS)

Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.

1979-01-01

The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements