WorldWideScience

Sample records for performing criticality safety

  1. 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

  2. Critical review of safety performance metrics

    NARCIS (Netherlands)

    Karanikas, Nektarios

    2016-01-01

    Various tools for safety performance measurement have been introduced in order to fulfil the need for safety monitoring in organisations, which is tightly related to their overall performance and achievement of their business goals. Such tools include accident rates, benchmarking, safety culture and

  3. 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

  4. Performance Testing Methodology for Safety-Critical Programmable Logic Controller

    International Nuclear Information System (INIS)

    Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

    2009-01-01

    The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC

  5. 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

  6. Evaluating Models of Human Performance: Safety-Critical Systems Applications

    Science.gov (United States)

    Feary, Michael S.

    2012-01-01

    This presentation is part of panel discussion on Evaluating Models of Human Performance. The purpose of this panel is to discuss the increasing use of models in the world today and specifically focus on how to describe and evaluate models of human performance. My presentation will focus on discussions of generating distributions of performance, and the evaluation of different strategies for humans performing tasks with mixed initiative (Human-Automation) systems. I will also discuss issues with how to provide Human Performance modeling data to support decisions on acceptability and tradeoffs in the design of safety critical systems. I will conclude with challenges for the future.

  7. 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)

  8. 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

  9. 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.

  10. Dry critical experiments and analyses performed in support of the Topaz-2 Safety Program

    International Nuclear Information System (INIS)

    Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Loynstev, V.A.

    1994-01-01

    In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations

  11. 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

  12. Dry critical experiments and analyses performed in support of the TOPAZ-2 safety program

    International Nuclear Information System (INIS)

    Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Lobynstev, V.A.

    1995-01-01

    In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations. copyright 1995 American Institute of Physics

  13. Surgical procedures performed in the neonatal intensive care unit on critically ill neonates: feasibility and safety

    International Nuclear Information System (INIS)

    Mallick, M.S.; Jado, A.M.; Al-Bassam, A.R.

    2008-01-01

    Transferring unstable, ill neonates to and from the operating rooms carries significant risks and can lead to morbidity. We report on our experience in performing certain procedures in critically ill neonates in the neonatal intensive care unit (NICU). We examined the feasibility and safety for such an approach. All surgical procedures performed in the NICU between January 1999 and December 2005 were analyzed in terms of demographic data, diagnosis, preoperative stability of the patient, procedures performed, complications and outcome. Operations were performed at beside in the NICU in critically ill, unstable neonates who needed emergency surgery, in neonates of low birth weight (<1000 gm) and in neonates on special equipments like higher frequency ventilators and nitrous oxide. Thirty-seven surgical procedures were performed including 12 laparotomies, bowel resection and stomies, 7 repairs of congenital diaphragmatic hernias, 4 ligations of patent ductus arteriosus and various others. Birth weights ranged between 850 gm and 3500 gm (mean 2000 gm). Gestational age ranged between 25 to 42 weeks (mean, 33 weeks). Age at surgery was between 1 to 30 days (mean, 30 days). Preoperatively, 19 patients (51.3%) were on inotropic support and all were intubated and mechanically ventilated. There was no mortality related to surgical procedures. Postoperatively, one patient developed wound infection and disruption. Performing major surgical procedures in the NICU is both feasible and safe. It is useful in very low birth weight, critically ill neonates who have definite risk attached to transfer to the operating room. No special area is needed in the NICU to perform complication-free surgery, but designing an operating room within the NICU will be ideal. (author)

  14. Identifying the Critical Factors Affecting Safety Program Performance for Construction Projects within Pakistan Construction Industry

    Directory of Open Access Journals (Sweden)

    Zubair Ahmed Memon

    2013-04-01

    Full Text Available Many studies have shown that the construction industry one of the most hazardous industries with its high rates of fatalities and injuries and high financial losses incurred through work related accident. To reduce or overcome the safety issues on construction sites, different safety programs are introduced by construction firms. A questionnaire survey study was conducted to highlight the influence of the Construction Safety Factors on safety program implementation. The input from the questionnaire survey was analyzed by using AIM (Average Index Method and rank correlation test was conducted between different groups of respondents to measure the association between different groups of respondent. The finding of this study highlighted that management support is the critical factor for implementing the safety program on projects. From statistical test, it is concluded that all respondent groups were strongly in the favor of management support factor as CSF (Critical Success Factor. The findings of this study were validated on selected case studies. Results of the case studies will help to know the effect of the factors on implementing safety programs during the execution stage.

  15. Predicting fatigue and psychophysiological test performance from speech for safety critical environments

    Directory of Open Access Journals (Sweden)

    Khan Richard Baykaner

    2015-08-01

    Full Text Available Automatic systems for estimating operator fatigue have application in safety-critical environments. A system which could estimate level of fatigue from speech would have application in domains where operators engage in regular verbal communication as part of their duties. Previous studies on the prediction of fatigue from speech have been limited because of their reliance on subjective ratings and because they lack comparison to other methods for assessing fatigue. In this paper we present an analysis of voice recordings and psychophysiological test scores collected from seven aerospace personnel during a training task in which they remained awake for 60 hours. We show that voice features and test scores are affected by both the total time spent awake and the time position within each subject’s circadian cycle. However, we show that time spent awake and time of day information are poor predictors of the test results; while voice features can give good predictions of the psychophysiological test scores and sleep latency. Mean absolute errors of prediction are possible within about 17.5% for sleep latency and 5-12% for test scores. We discuss the implications for the use of voice as a means to monitor the effects of fatigue on cognitive performance in practical applications.

  16. 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?

  17. 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)

  18. 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)

  19. Assessment of criticality safety

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Heaberlin, S.W.; Clayton, E.D.; Carter, R.D.

    1979-01-01

    A study was made of 100 violations of criticality safety specifications reported over a 10-y period in the operations of fuel reprocessing plants. The seriousness of each rule violation was evaluated by assigning it a severity index value. The underlying causes or reasons, for the violations were identified. A criticality event tree was constructed using the parameters, causes, and reasons found in the analysis of the infractions. The event tree provides a means for visualizing the paths to an accidental criticality. Some 65% of the violations were caused by misinterpretation on the part of the operator, being attributed to a lack of clarity in the specification and insufficient training; 33% were attributed to lack of care, whereas only 2% were caused by mechanical failure. A fault tree was constructed by assembling the events that could contribute to an accident. With suitable data on the probabilities of contributing events, the probability of the accident's occurrence can be forecast. Estimated probabilities for criticality were made, based on the limited data available, that in this case indicate a minimum time span of 244 y of plant operation per accident ranging up to approx. 3000 y subject to the various underlying assumptions made. Some general suggestions for improvement are formulated based on the cases studied. Although conclusions for other plants may differ in detail, the general method of analysis and the fault tree logic should prove applicable. 4 figures, 8 tables

  20. 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

  1. Assessing Risk-Based Performance Indicators in Safety-Critical Systems for Nuclear Power Plants

    OpenAIRE

    TONT Gabriela

    2011-01-01

    The paper proposes framework for a multidisciplinary nuclear risk and safety assessment by modeling uncertainty and combining diverse evidence provided in such a way that it could be used to represent an entire argument about a system's dependability. The identified safety issues are being treated by means of probabilistic safety assessment (PSA). The behavior simulation of power plant in thepresence of risk factors is analyzed from the vulnerability, risk and functional safety viewpoints, hi...

  2. Spent fuel storage criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E M; Elmessiry, A M [National center of nuclear safety and radiation control atomic energy authority, (Egypt)

    1995-10-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs.

  3. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

    Amin, E.M.; Elmessiry, A.M.

    1995-01-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs

  4. 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.

  5. 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

  6. Licensing safety critical software

    International Nuclear Information System (INIS)

    Archinoff, G.H.; Brown, R.A.

    1990-01-01

    Licensing difficulties with the shutdown system software at the Darlington Nuclear Generating Station contributed to delays in starting up the station. Even though the station has now been given approval by the Atomic Energy Control Board (AECB) to operate, the software issue has not disappeared - Ontario Hydro has been instructed by the AECB to redesign the software. This article attempts to explain why software based shutdown systems were chosen for Darlington, why there was so much difficulty licensing them, and what the implications are for other safety related software based applications

  7. 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.)

  8. 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}.

  9. SCALE criticality safety verification and validation package

    International Nuclear Information System (INIS)

    Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

    1998-01-01

    Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

  10. 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)

  11. Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II Safety Program

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutoy, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N.

    1994-01-01

    Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated

  12. 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)

  13. Extending Critical Performativity

    DEFF Research Database (Denmark)

    Spicer, André; Alvesson, Mats; Kärreman, Dan

    2016-01-01

    In this article we extend the debate about critical performativity. We begin by outlining the basic tenets of critical performativity and how this has been applied in the study of management and organization. We then address recent critiques of critical performance. We note these arguments suffer...... of public importance; engaging with non-academic groups using dialectical reasoning; scaling up insights through movement building; and propagating deliberation...

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. Verification of safety critical software

    International Nuclear Information System (INIS)

    Son, Ki Chang; Chun, Chong Son; Lee, Byeong Joo; Lee, Soon Sung; Lee, Byung Chai

    1996-01-01

    To assure quality of safety critical software, software should be developed in accordance with software development procedures and rigorous software verification and validation should be performed. Software verification is the formal act of reviewing, testing of checking, and documenting whether software components comply with the specified requirements for a particular stage of the development phase[1]. New software verification methodology was developed and was applied to the Shutdown System No. 1 and 2 (SDS1,2) for Wolsung 2,3 and 4 nuclear power plants by Korea Atomic Energy Research Institute(KAERI) and Atomic Energy of Canada Limited(AECL) in order to satisfy new regulation requirements of Atomic Energy Control Boars(AECB). Software verification methodology applied to SDS1 for Wolsung 2,3 and 4 project will be described in this paper. Some errors were found by this methodology during the software development for SDS1 and were corrected by software designer. Outputs from Wolsung 2,3 and 4 project have demonstrated that the use of this methodology results in a high quality, cost-effective product. 15 refs., 6 figs. (author)

  19. Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II safety program

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutov, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N.; Chunyaev, E.I.; Marshall, A.C.; Sapir, J.L.; Pelowitz, D.B.

    1995-01-01

    Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated. copyright 1995 American Institute of Physics

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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)

  5. 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.)

  6. 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)

  7. 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)

  8. 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

  9. Safety performance indicators program

    International Nuclear Information System (INIS)

    Vidal, Patricia G.

    2004-01-01

    In 1997 the Nuclear Regulatory Authority (ARN) initiated a program to define and implement a Safety Performance Indicators System for the two operating nuclear power plants, Atucha I and Embalse. The objective of the program was to incorporate a set of safety performance indicators to be used as a new regulatory tool providing an additional view of the operational performance of the nuclear power plants, improving the ability to detect degradation on safety related areas. A set of twenty-four safety performance indicators was developed and improved throughout pilot implementation initiated in July 1998. This paper summarises the program development, the main criteria applied in each stage and the results obtained. (author)

  10. 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

  11. 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)

  12. Benchmarking criticality safety calculations with subcritical experiments

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1984-06-01

    Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

  13. 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

  14. 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

  15. Criticality safety research on nuclear fuel cycle facility

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-07-01

    This paper present d s current status and future program of the criticality safety research on nuclear fuel cycle made by Japan Atomic Energy Research Institute. Experimental research on solution fuel treated in reprocessing plant has been performed using two critical facilities, STACY and TRACY. Fundamental data of static and transient characteristics are accumulated for validation of criticality safety codes. Subcritical measurements are also made for developing a monitoring system for criticality safety. Criticality safety codes system for solution and power system, and evaluation method related to burnup credit are developed. (author)

  16. Neutron nuclear data measurements for criticality safety

    Directory of Open Access Journals (Sweden)

    Guber Klaus

    2017-01-01

    Full Text Available To support the US Department of Energy Nuclear Criticality Safety Program, neutron-induced cross section experiments were performed at the Geel Electron Linear Accelerator of the Joint Research Center Site Geel, European Union. Neutron capture and transmission measurements were carried out using metallic natural cerium and vanadium samples. Together with existing data, the measured data will be used for a new evaluation and will be submitted with covariances to the ENDF/B nuclear data library.

  17. 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)

  18. 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

  19. 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

  20. 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)

  1. 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...

  2. Software for safety critical applications

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Jurickova, M.; Chudy, R.

    2001-01-01

    The contribution gives an overview of the project of the software development for safety critical applications. This project has been carried out since 1997. The principal goal of the project was to establish a research laboratory for the development of the software with the highest requirements for quality and reliability. This laboratory was established at the department, equipped with proper hardware and software to support software development. A research team of predominantly young researchers for software development was created. The activities of the research team started with studying and proposing the software development methodology. In addition, this methodology was applied to the real software development. The verification and validation process followed the software development. The validation system for the integrated hardware and software tests was brought into being and its control software was developed. The quality of the software tools was also observed, and the SOSAT tool was used during these activities. National and international contacts were established and maintained during the project solution.(author)

  3. Microbiological performance of Hazard Analysis Critical Control Point (HACCP)-based food safety management systems: A case of Nile perch processing company

    NARCIS (Netherlands)

    Kussaga, J.B.; Luning, P.A.; Tiisekwa, B.P.M.; Jacxsens, L.

    2017-01-01

    This study aimed at giving insight into microbiological safety output of a Hazard Analysis Critical Control Point (HACCP)-based Food Safety Management System (FSMS) of a Nile perch exporting company by using a combined assessment, This study aimed at giving insight into microbiological safety output

  4. 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

  5. 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

  6. 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

  7. Criticality safety training at Westinghouse Hanford Company

    International Nuclear Information System (INIS)

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

    1983-01-01

    In 1972 the Westinghouse Hanford Company (WHC) established a comprehensive program to certify personnel who handle fissionable materials. As the quantity of fissionable material handled at WHC has increased so has the scope of training to assure that all employes perform their work in a safe manner. This paper describes training for personnel engaged in fuel fabrication and handling activities. Most of this training is provided by the Fissionable Material Handlers Certification Program. This program meets or exceeds all DOE requirements for training and has been attended by more than 475 employes. Since the program was instituted, the rate of occurrence of criticality safety limit violations has decreased by 50%

  8. 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)

  9. 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.)

  10. Evolvement of nuclear criticality safety programs

    International Nuclear Information System (INIS)

    Ketzlach, N.

    1992-01-01

    Nuclear criticality safety (NCS) has developed from a discipline requiring the services of personnel with only a background in reactor physics to that involving reactor physics, process engineering, and design as well as administration of the program to ensure all its requirements are implemented. When Oak Ridge National Laboratory (ORNL) was designed and constructed, the physicists at Los Alamos National Laboratory (LANL) were performing the criticality analyses. A physicist who had no chemical process or engineering experience was brought in from LANL to determine whether the facility would be safe. It was only because of his understanding of the reactor physics principles, scientific intuition, and some luck that the design and construction of the facility led to a safe plant. It took a number of years of experience with facility operations and the dedication of personnel for NCS to reach its present status as a recognized discipline

  11. 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.

  12. 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)

  13. 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...

  14. 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...

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. Nuclear criticality safety: 300 Area

    International Nuclear Information System (INIS)

    1991-01-01

    This Standard applies to the receipt, processing, storage, and shipment of fissionable material in the 300 Area and in any other facility under the control of the Reactor Materials Project Management Team (PMT). The objective is to establish practices and process conditions for the storage and handling of fissionable material that prevent the accidental assembly of a critical mass and that comply with DOE Orders as well as accepted industry practice

  5. 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

  6. Researches on nuclear criticality safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-10-01

    For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

  7. Researches on nuclear criticality safety evaluation

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi

    2003-01-01

    For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

  8. 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

  9. Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

    Energy Technology Data Exchange (ETDEWEB)

    Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-26

    The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. Safety KPIs - Monitoring of safety performance

    Directory of Open Access Journals (Sweden)

    Andrej Lališ

    2014-09-01

    Full Text Available This paper aims to provide brief overview of aviation safety development focusing on modern trends represented by implementation of Safety Key Performance Indicators. Even though aviation is perceived as safe means of transport, it is still struggling with its complexity given by long-term growth and robustness which it has reached today. Thus nowadays safety issues are much more complex and harder to handle than ever before. We are more and more concerned about organizational factors and control mechanisms which have potential to further increase level of aviation safety. Within this paper we will not only introduce the concept of Key Performance Indicators in area of aviation safety as an efficient control mechanism, but also analyse available legislation and documentation. Finally we will propose complex set of indicators which could be applied to Czech Air Navigation Service Provider.

  16. Memory Management for Safety-Critical Java

    DEFF Research Database (Denmark)

    Schoeberl, Martin

    2011-01-01

    Safety-Critical Java (SCJ) is based on the Real-Time Specification for Java. To simplify the certification of Java programs, SCJ supports only a restricted scoped memory model. Individual threads share only immortal memory and the newly introduced mission memory. All other scoped memories...... implementation is evaluated on an embedded Java processor....

  17. 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.)

  18. Safety Critical Java for Robotics Programming

    DEFF Research Database (Denmark)

    Thomsen, Bent; Luckow, Kasper Søe; Bøgholm, Thomas

    2015-01-01

    This paper introduces Safety Critical Java (SCJ) and argues its readiness for robotics programming. We give an overview of the work done at Aalborg University and elsewhere on SCJl, some of its implementations in the form of the JOP, FijiVM and HVM and some of the tools, especially WCA, Teta...

  19. Analysis of the criticality safety of a nuclear fuel deposit

    International Nuclear Information System (INIS)

    Landeyro, P.A.; Mincarini, M.

    1987-01-01

    In the present work a safety analysis from criticality accidents of nuclear fuel deposits is performed. The analysis is performed utilizing two methods derived from different physical principes: 1) superficial density method, obtained from experimental research; 2) solid angle method, derived from transport theory

  20. Recommendations for preparing the criticality safety evaluation of transportation packages

    International Nuclear Information System (INIS)

    Dyer, H.R.; Parks, C.V.

    1997-04-01

    This report provides recommendations on preparing the criticality safety section of an application for approval of a transportation package containing fissile material. The analytical approach to the evaluation is emphasized rather than the performance standards that the package must meet. Where performance standards are addressed, this report incorporates the requirements of 10 CFR Part 71. 12 refs., 6 figs., 8 tabs

  1. Computational methods for nuclear criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.

    1992-01-01

    Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

  2. 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

  3. 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)

  4. Safety critical application of fuzzy control

    International Nuclear Information System (INIS)

    Schildt, G.H.

    1995-01-01

    After an introduction into safety terms a short description of fuzzy logic will be given. Especially, for safety critical applications of fuzzy controllers a possible controller structure will be described. The following items will be discussed: Configuration of fuzzy controllers, design aspects like fuzzfiication, inference strategies, defuzzification and types of membership functions. As an example a typical fuzzy rule set will be presented. Especially, real-time behaviour a fuzzy controllers is mentioned. An example of fuzzy controlling for temperature control purpose within a nuclear reactor together with membership functions and inference strategy of such a fuzzy controller will be presented. (author). 4 refs, 17 figs

  5. SCALE system cross-section validation for criticality safety analysis

    International Nuclear Information System (INIS)

    Hathout, A.M.; Westfall, R.M.; Dodds, H.L. Jr.

    1980-01-01

    The purpose of this study is to test selected data from three cross-section libraries for use in the criticality safety analysis of UO 2 fuel rod lattices. The libraries, which are distributed with the SCALE system, are used to analyze potential criticality problems which could arise in the industrial fuel cycle for PWR and BWR reactors. Fuel lattice criticality problems could occur in pool storage, dry storage with accidental moderation, shearing and dissolution of irradiated elements, and in fuel transport and storage due to inadequate packing and shipping cask design. The data were tested by using the SCALE system to analyze 25 recently performed critical experiments

  6. 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.

  7. Critical enrichment and critical density of infinite systems for nuclear criticality safety evaluation

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Koyama, Takashi; Komuro, Yuichi

    1986-03-01

    Critical enrichment and critical density of homogenous infinite systems, such as U-H 2 O, UO 2 -H 2 O, UO 2 F 2 aqueous solution, UO 2 (NO 3 ) 2 aqueous solution, Pu-H 2 O, PuO 2 -H 2 O, Pu(NO 3 ) 4 aqueous solution and PuO 2 ·UO 2 -H 2 O, were calculated with the criticality safety evaluation computer code system JACS for nuclear criticality safety evaluation on fuel facilities. The computed results were compared with the data described in European and American criticality handbooks and showed good agreement with each other. (author)

  8. 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.

  9. Criticality Safety Evaluation of Standard Criticality Safety Requirements #1-520 g Operations in PF-4

    Energy Technology Data Exchange (ETDEWEB)

    Yamanaka, Alan Joseph Jr. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-13

    Guidance has been requested from the Nuclear Criticality Safety Division (NCSD) regarding processes that involve 520 grams of fissionable material or less. This Level-3 evaluation was conducted and documented in accordance with NCS-AP-004 (Ref. 1), formerly NCS-GUIDE-01. This evaluation is being written as a generic evaluation for all operations that will be able to operate using a 520-gram mass limit. Implementation for specific operations will be performed using a Level 1 CSED, which will confirm and document that this CSED can be used for the specific operation as discussed in NCS-MEMO-17-007 (Ref. 2). This Level 3 CSED updates and supersedes the analysis performed in NCS-TECH-14-014 (Ref. 3).

  10. Safety-Critical Java for Embedded Systems

    DEFF Research Database (Denmark)

    Rios Rivas, Juan Ricardo

    for Java aims at providing a reduced set of the Java programming language that can be used for systems that need to be certified at the highest levels of criticality. Safety-critical Java (SCJ) restricts how a developer can structure an application by providing a specific programming model...... and by restricting the set of methods and libraries that can be used. Furthermore, its memory model do not use a garbage-collected heap but scoped memories. In this thesis we examine the use of the SCJ specification through an implementation in a time-predictable, FPGA-based Java processor. The specification is now...

  11. 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

  12. Criticality safety considerations. Integral Monitored Retrievable Storage (MRS) Facility

    International Nuclear Information System (INIS)

    1986-09-01

    This report summarizes the criticality analysis performed to address criticality safety concerns and to support facility design during the conceptual design phase of the Monitored Retrievable Storage (MRS) Facility. The report addresses the criticality safety concerns, the design features of the facility relative to criticality, and the results of the analysis of both normal operating and hypothetical off-normal conditions. Key references are provided (Appendix C) if additional information is desired by the reader. The MRS Facility design was developed and the related analysis was performed in accordance with the MRS Facility Functional Design Criteria and the Basis for Design. The detailed description and calculations are documented in the Integral MRS Facility Conceptual Design Report. In addition to the summary portion of this report, explanatary notes for various terms, calculation methodology, and design parameters are presented in Appendix A. Appendix B provides a brief glossary of technical terms

  13. Security for safety critical space borne systems

    Science.gov (United States)

    Legrand, Sue

    1987-01-01

    The Space Station contains safety critical computer software components in systems that can affect life and vital property. These components require a multilevel secure system that provides dynamic access control of the data and processes involved. A study is under way to define requirements for a security model providing access control through level B3 of the Orange Book. The model will be prototyped at NASA-Johnson Space Center.

  14. Developing software for safety-critical applications

    International Nuclear Information System (INIS)

    Chudleigh, M.

    1989-01-01

    The effective implementation of many safety-critical systems involves microprocessors running software which needs to be of very high integrity. This article describes some of the problems of producing such software and the place of software within the total system. A development strategy is proposed based on three principles: the goal of defect-free development, the use of mathematical formalism, and the use of an independent team for testing. (author)

  15. Criticality safety study of shutdown diffusion cascade coolers

    International Nuclear Information System (INIS)

    Paschal, L.S.; Basoglu, B.; Bentley, C.L.; Dunn, M.E.

    1996-01-01

    Gaseous diffusion plants use cascade coolers in the production of highly enriched uranium (HEU) to remove heat from the enriched stream of UF 6 . The cascade coolers operate like shell and tube heat exchangers with the UF 6 on the shell side and Freon on the tube side. Recirculating cooling water (RCW) in condensers is used to cool the Freon. A criticality safety analysis was previously performed for cascade coolers during normal operation. The purpose of this paper is to evaluate several different hypothetical accidents regarding RCW ingress into the cooler to determine whether criticality safety concerns exist

  16. 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

  17. 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...

  18. 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

  19. 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

  20. Examining the Relationship of an All-Cause Harm Patient Safety Measure and Critical Performance Measures at the Frontline of Care.

    Science.gov (United States)

    Sammer, Christine; Hauck, Loran D; Jones, Cason; Zaiback-Aldinger, Julie; Li, Michael; Classen, David

    2018-02-07

    In 2015, the Institute of Medicine Vital Signs report called for a new patient safety composite measure to lessen the reporting burden of patient harm. Before this report, two patient safety organizations had developed an electronic all-cause harm measurement system leveraging data from the electronic health record, which identified and grouped harms into five broad categories and consolidated them into one all-cause harm outcome measure. The objective of this study was to examine the relationship between this all-cause harm patient safety measure and the following three performance measures important to overall hospital safety performance: safety culture, employee engagement, and patient experience. We studied the relationship between all-cause harm and three performance measures on eight inpatient care units at one hospital for 7 months. The findings demonstrated strong correlations between an all-cause harm measure and patient safety culture, employee engagement, and patient experience at the hospital unit level. Four safety culture domains showed significant negative correlations with all-cause harm at a P value of 0.05 or less. Six employee engagement domains were significantly negatively correlated with all-cause harm at a P value of 0.01 or less, and six of the ten patient experience measures were significantly correlated with all-cause harm at a P value of 0.05 or less. The results show that there is a strong relationship between all-cause harm and these performance measures indicating that when there is a positive patient safety culture, a more engaged employee, and a more satisfying patient experience, there may be less all-cause harm.This is an open-access article distributed under the terms of the Creative Commons Attribution-Non Commercial-No Derivatives License 4.0 (CCBY-NC-ND), where it is permissible to download and share the work provided it is properly cited. The work cannot be changed in any way or used commercially without permission from the

  1. A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

    International Nuclear Information System (INIS)

    Disney, R.K.

    1994-10-01

    The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE's) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE's within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ''site'' perception to a more uniform or ''national'' perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticals data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation

  2. Development of safety performance indicators for HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Ahn, Guk-Hoon; Lee, Kye-Hong; Lim, In-Cheol

    2007-01-01

    The nuclear facilities need an extensive basis for ensuring their safety. An operating organization should conduct its operation and utilization important to the safety in accordance with approved procedures and regulations. The general aims of a management system for nuclear facilities are to improve the safety performance through a planning, control and supervision of safety related activities and to foster a strong safety culture. The effectiveness of a management system can be monitored and measured to confirm the ability of its processes to achieve the intended safety performance by an assessment of the operational performance. The Operational Safety Performance Indicators, also known as SPI, help an organization define and measure a progress with regard to safety activity goals. The elements of a SPI are quantifiable measurements that reflect the critical success factors of an organizational safety. Since 1995, efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator program in nuclear power plants (NPP). IAEA-TECDOC-1141, 'Operational safety performance indicators for NPP' attempted to provide a frame work for an identification of performance indicators which have a relationship to the desired safety attributes, and therefore, to a safe plant operation. Three key attributes of a smooth operation, an operation with a low risk, and an operation with a positive safety attitude, were recommended, which are associated with a safe operation. Because these attributes cannot be directly measured, an indicator structure is expanded further until a level of easily quantifiable or directly measurable indicators is identified. The intention of this approach is to use quantitative information provided by the specific indicators and to analyze performance trends relative to established goals. The safety activities in HANARO have been continuously conducted to enhance its safe operation. HANARO

  3. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24

  4. 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

  5. Criticality safety of solvent extraction process

    International Nuclear Information System (INIS)

    Tachimori, Shoichi; Miyoshi, Yoshinori

    1987-01-01

    The article presents some comments on criticality safety of solvent extraction processes. When used as an extracting medium, tributyl phosphate extracts nitric acid and water, in addition to nitrates of U and Pu, into the organic phase. The amount of these chemical species extracted into the organic phase is dependent on and restricted by the concentrations of tributyl phosphate and other components. For criticality control, measures are taken to decrease the concentration of tributyl phosphate in the organic phase, in addition to control of the U and Pu concentrations in the feed water phase. It should be remembered that complexes of tributyl phosphate with nitrates of such metals as Pu(IV), Pu(VI), U(IV) and Th(IV) do not dissolve uniformly in the organic phase. In criticality calculation for solution-handling systems, U and Pu are generally assumed to have a valence of 6 and 4, respectively. In the reprocessing extraction process, however, U and Pu can have a valence of 4, and 3 and 6, respectively. The organic phase and aqueous phase contact in a counter-current flow. U and Pu will be accumulated if they are not brought out of the extraction system by this flow. (Nogami, K.)

  6. 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)

  7. Plutonium Finishing Plant (PFP) Safety Class and Safety Significant Commercial Grade Items (CGI) Critical Characteristic

    International Nuclear Information System (INIS)

    THOMAS, R.J.

    2000-01-01

    This document specifies the critical characteristics for Commercial Grade Items (CGI) procured for use in the Plutonium Finishing Plant as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to properly perform its safety function. There may be several manufacturers or models that meet the critical characteristics of any one item

  8. Multiprocessor Priority Ceiling Emulation for Safety-Critical Java

    DEFF Research Database (Denmark)

    Strøm, Torur Biskopstø; Schoeberl, Martin

    2015-01-01

    Priority ceiling emulation has preferable properties on uniprocessor systems, such as avoiding priority inversion and being deadlock free. This has made it a popular locking protocol. According to the safety-critical Java specication, priority ceiling emulation is a requirement for implementations....... However, implementing the protocol for multiprocessor systemsis more complex so implementations might perform worse than non-preemptive implementations. In this paper we compare two multiprocessor lock implementations with hardware support for the Java optimized processor: non-preemptive locking...

  9. 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)

  10. 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

  11. 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)

  12. 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)

  13. Technical bases for criticality safety standards

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1980-01-01

    An American National Standard implies a consensus of those substantially concerned with its scope and provisions. The technical basis, or foundation, on which the consensus rests, must in turn, be firmly established and documented for public review. The technical bases are discussed and reviewed of several standards in different stages of completion and acceptance: ANSI/ANS-8.12, 1978, Nuclear Criticality Control and Safety of Homogeneous Plutonium - Uranium Mixtures Outside Reactors (Approved July 17, 1978); ANS-815, Nuclear Criticality Control of Special Actinide Elements (Draft No. 5 of newly proposed standard); ANS-8.14, Use of Solutions of Neutron Absorbers for Criticality Control (Draft No. 4 of newly proposed standard); ANS-8.5 (Revision of N16.4, 1971), Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material (Draft No. 5 as a result of prescribed five-year review and update of old standard). In each of the preceding, the newly proposed (or revised) limits are based on the extension of experimental data via well established calculations, or by means of independent calculations with adequate margins for uncertainties. The four cases serve to illustrate the insight of the work group members in the establishment of the technical bases for the limits and the level of activity required on their part in the preparation of ANSI Standards. A time span of from four up to seven years has not been uncommon for the preparation, review, and acceptance of an ANSI Standard. 8 figures. 7 tables

  14. 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)

  15. Implications of Monte Carlo Statistical Errors in Criticality Safety Assessments

    International Nuclear Information System (INIS)

    Pevey, Ronald E.

    2005-01-01

    Most criticality safety calculations are performed using Monte Carlo techniques because of Monte Carlo's ability to handle complex three-dimensional geometries. For Monte Carlo calculations, the more histories sampled, the lower the standard deviation of the resulting estimates. The common intuition is, therefore, that the more histories, the better; as a result, analysts tend to run Monte Carlo analyses as long as possible (or at least to a minimum acceptable uncertainty). For Monte Carlo criticality safety analyses, however, the optimization situation is complicated by the fact that procedures usually require that an extra margin of safety be added because of the statistical uncertainty of the Monte Carlo calculations. This additional safety margin affects the impact of the choice of the calculational standard deviation, both on production and on safety. This paper shows that, under the assumptions of normally distributed benchmarking calculational errors and exact compliance with the upper subcritical limit (USL), the standard deviation that optimizes production is zero, but there is a non-zero value of the calculational standard deviation that minimizes the risk of inadvertently labeling a supercritical configuration as subcritical. Furthermore, this value is shown to be a simple function of the typical benchmarking step outcomes--the bias, the standard deviation of the bias, the upper subcritical limit, and the number of standard deviations added to calculated k-effectives before comparison to the USL

  16. Using fuzzy self-organising maps for safety critical systems

    International Nuclear Information System (INIS)

    Kurd, Zeshan; Kelly, Tim P.

    2007-01-01

    This paper defines a type of constrained artificial neural network (ANN) that enables analytical certification arguments whilst retaining valuable performance characteristics. Previous work has defined a safety lifecycle for ANNs without detailing a specific neural model. Building on this previous work, the underpinning of the devised model is based upon an existing neuro-fuzzy system called the fuzzy self-organising map (FSOM). The FSOM is type of 'hybrid' ANN which allows behaviour to be described qualitatively and quantitatively using meaningful expressions. Safety of the FSOM is argued through adherence to safety requirements-derived from hazard analysis and expressed using safety constraints. The approach enables the construction of compelling (product-based) arguments for mitigation of potential failure modes associated with the FSOM. The constrained FSOM has been termed a 'safety critical artificial neural network' (SCANN). The SCANN can be used for non-linear function approximation and allows certified learning and generalisation for high criticality roles. A discussion of benefits for real-world applications is also presented

  17. 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

  18. Software quality assurance plans for safety-critical software

    International Nuclear Information System (INIS)

    Liddle, P.

    2006-01-01

    Application software is defined as safety-critical if a fault in the software could prevent the system components from performing their nuclear-safety functions. Therefore, for nuclear-safety systems, the AREVA TELEPERM R XS (TXS) system is classified 1E, as defined in the Inst. of Electrical and Electronics Engineers (IEEE) Std 603-1998. The application software is classified as Software Integrity Level (SIL)-4, as defined in IEEE Std 7-4.3.2-2003. The AREVA NP Inc. Software Program Manual (SPM) describes the measures taken to ensure that the TELEPERM XS application software attains a level of quality commensurate with its importance to safety. The manual also describes how TELEPERM XS correctly performs the required safety functions and conforms to established technical and documentation requirements, conventions, rules, and standards. The program manual covers the requirements definition, detailed design, integration, and test phases for the TELEPERM XS application software, and supporting software created by AREVA NP Inc. The SPM is required for all safety-related TELEPERM XS system applications. The program comprises several basic plans and practices: 1. A Software Quality-Assurance Plan (SQAP) that describes the processes necessary to ensure that the software attains a level of quality commensurate with its importance to safety function. 2. A Software Safety Plan (SSP) that identifies the process to reasonably ensure that safety-critical software performs as intended during all abnormal conditions and events, and does not introduce any new hazards that could jeopardize the health and safety of the public. 3. A Software Verification and Validation (V and V) Plan that describes the method of ensuring the software is in accordance with the requirements. 4. A Software Configuration Management Plan (SCMP) that describes the method of maintaining the software in an identifiable state at all times. 5. A Software Operations and Maintenance Plan (SO and MP) that

  19. 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)

  20. 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.

  1. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo; Seong, Poong Hyun

    1997-01-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formed safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system

  2. SRTC criticality safety technical review of SRT-CMA-930039

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Review of SRT-CMA-930039, ''Nuclear Criticality Safety Evaluation (NCSE): DWPF Melter-Batch 1,'' December 1, 1993, has been performed by the Savannah River Technical Center (SRTC) Applied Physics Group. The NCSE is a criticality assessment of the Melt Cell in the DWPF. Additionally, this pertains only to Batch 1 operation, which differs from batches to follow. Plans for subsequent batch operations call for fissile material in the Salt Cell feed-stream, which necessitates a separate criticality evaluation in the future. The NCSE under review concludes that the process is safe from criticality events, even in the event that all lithium and boron neutron poisons are lost, provided uranium enrichments are less than 40%. Furthermore, if all the lithium and as much as 98% of the boron would be lost, uranium enrichments of 100% would be allowable. After a thorough review of the NCSE, this reviewer agrees with that conclusion. This technical review consisted of: an independent check of the methods and models employed, independent calculations application of ANSI/ANS 8.1, verification of WSRC Nuclear Criticality Safety Manual( 2 ) procedures

  3. 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.

  4. 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

  5. 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.

  6. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo

    1997-02-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

  7. 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

  8. 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

  9. 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

  10. 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)

  11. Automated tools for safety-critical software

    International Nuclear Information System (INIS)

    Lapassat, A.M.

    1993-01-01

    The regulatory (DSIN), the utilities (EDF, CEA..) and the CEA-Institute for Protection and Nuclear Safety (IPSN) work together at the French nuclear safety. This paper presents a tool, called CLAIRE, for simulation and tests of different nuclear safety system. (TEC)

  12. 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

  13. Training and qualification program for nuclear criticality safety technical staff. Revision 1

    International Nuclear Information System (INIS)

    Taylor, R.G.; Worley, C.A.

    1997-01-01

    A training and qualification program for nuclear criticality safety technical staff personnel has been developed and implemented. All personnel who are to perform nuclear criticality safety technical work are required to participate in the program. The program includes both general nuclear criticality safety and plant specific knowledge components. Advantage can be taken of previous experience for that knowledge which is portable such as performance of computer calculations. Candidates step through a structured process which exposes them to basic background information, general plant information, and plant specific information which they need to safely and competently perform their jobs. Extensive documentation is generated to demonstrate that candidates have met the standards established for qualification

  14. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Guinn, W.M.; Porter, N.J.

    1981-01-01

    The operator's role in nuclear safety is outlined and the concept of ''safety functions'' introduced. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. The plant safety evaluation uses four inputs in predicting the results of an event: the event initiator, the plant design, the initial plant conditions and setup, and the operator actions. If any of these inputs are not as assumed in the evaluation, confidence that the consequences will be as predicted is reduced. Based on the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results: Maintain plant setup in readiness to properly respond. Operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events. Monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of safety functions introduces this systematic approach and presents a hierarchy of protection. If the operator has difficulty identifying an event for any reason, the systematic safety function approach allows accomplishing the overall path of mitigating consequences. Ten functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions are identified

  15. 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.

  16. 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

  17. 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)

  18. 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

  19. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  20. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs

  1. 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

  2. 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)

  3. Review of studies on criticality safety evaluation and criticality experiment methods

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Yamamoto, Toshihiro; Misawa, Tsuyoshi; Yamane, Yuichi

    2013-01-01

    Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS. (author)

  4. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Porter, N.J.; Cross, M.T.; Guinn, W.M.

    1981-01-01

    The paper outlines the operator's role in nuclear safety and introduces the concept of ''safety functions''. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. ''An accident identical to that at Three Mile Island is not going to happen again'', said the Rogovin investigators. The concepts put forward in this paper are intended to help the operator avoid serious consequence from the next unexpected threat. On the basis of the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results. These three operator roles are: first, maintain plant setup in readiness to properly respond; second, operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events; third, the operator needs to monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of ''safety function'' introduces that systematic approach and prevents a hierarchy of protection. If the operator has difficulty in identifying an event for any reason, the systematic safety function approach allows ones to accomplish the overall path of mitigating consequences. There are ten identified functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions. The paper describes in detail the operator's role and the safety functions, and provides many examples of the use of alternative success paths to accomplish the safety function

  5. 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)

  6. 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)

  7. 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.

  8. 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.

  9. 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

  10. Safety performance indicators for the road network.

    NARCIS (Netherlands)

    Weijermars, W. Gitelman, V. Papadimitriou, E. Lima De & Azevedo, C.

    2010-01-01

    Within the 6th FP European project SafetyNet, a team has worked on the development of Safety Performance Indicators (SPIs) on seven road safety related areas. These SPIs reflect the operational conditions of the road traffic system that influence the system's safety performance. SPIs were developed

  11. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

    Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

  12. 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)

  13. 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

  14. 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

  15. 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

  16. Critical safety parameters: The logical approach to refresher training

    International Nuclear Information System (INIS)

    Johnson, A.R.; Pilkington, W.; Turner, S.

    1991-01-01

    Nuclear power plant managers must ensure that control room staff are able to perform effectively. This is of particular importance through the longer term after initial authorization. Traditionally refresher training has been based on delivery of fragmented training packages typically derived from the initial authorization training programs. Various approaches have been taken to provide a more integrated refresher training program. However, methods such as job and task analysis and subject matter expert derived training have tended to develop without a focused clear overall training objective. The primary objective of all control room staff training is to ensure a proper and safe response to all plant transients. At the Point Lepreau Nuclear Plant, this has defined the Critical Safety Parameter based refresher training program. The overall objective of the Critical Safety Parameter training program is to ensure that control room staff can monitor and control a discrete set of plant parameters. Maintenance of the selected parameters within defined boundaries assures adequate cooling of the fuel and containment of radioactivity. Control room staff need to be able to reliably respond correctly to plant transients under potentially high stress conditions,. utilizing the essential knowledge and skills to deal with such transients. The inference is that the knowledge and skills must be limited to that which can be reliably recalled. This paper describes how the Point Lepreau Nuclear Plant has developed a refresher training program on the basis of a limited number of Critical Safety Parameters. Through this approach, it has been possible to define the essential set of knowledge and skills which ensures a correct response to plant transients

  17. 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)

  18. 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)

  19. 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

  20. Maximizing Team Performance: The Critical Role of the Nurse Leader.

    Science.gov (United States)

    Manges, Kirstin; Scott-Cawiezell, Jill; Ward, Marcia M

    2017-01-01

    Facilitating team development is challenging, yet critical for ongoing improvement across healthcare settings. The purpose of this exemplary case study is to examine the role of nurse leaders in facilitating the development of a high-performing Change Team in implementing a patient safety initiative (TeamSTEPPs) using the Tuckman Model of Group Development as a guiding framework. The case study is the synthesis of 2.5 years of critical access hospital key informant interviews (n = 50). Critical juncture points related to team development and key nurse leader actions are analyzed, suggesting that nurse leaders are essential to maximize clinical teams' performance. © 2016 Wiley Periodicals, Inc.

  1. 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

  2. Fissile materials principles of criticality safety in handling and processing

    International Nuclear Information System (INIS)

    1976-01-01

    This Swedish Standard consists of the English version of the International Standard ISO 1709-1975-Nuclear energy. Fissile materials. Principles of criticality safety in handling and processing. (author)

  3. Process management - critical safety issues with focus on risk management

    International Nuclear Information System (INIS)

    Sanne, Johan M.

    2005-12-01

    Organizational changes focused on process orientation are taking place among Swedish nuclear power plants, aiming at improving the operation. The Swedish Nuclear Power Inspectorate has identified a need for increased knowledge within the area for its regulatory activities. In order to analyze what process orientation imply for nuclear power plant safety a number of questions must be asked: 1. How is safety in nuclear power production created currently? What significance does the functional organization play? 2. How can organizational forms be analysed? What consequences does quality management have for work and for the enterprise? 3. Why should nuclear power plants be process oriented? Who are the customers and what are their customer values? Which customers are expected to contribute from process orientation? 4. What can one learn from process orientation in other safety critical systems? What is the effect on those features that currently create safety? 5. Could customer values increase for one customer without decreasing for other customers? What is the relationship between economic and safety interests from an increased process orientation? The deregulation of the electricity market have caused an interest in increased economic efficiency, which is the motivation for the interest in process orientation. among other means. It is the nuclear power plants' owners and the distributors (often the same corporations) that have the strongest interest in process orientation. If the functional organization and associated practices are decomposed, the prerequisites of the risk management regime changes, perhaps deteriorating its functionality. When nuclear power operators consider the introduction of process orientation, the Nuclear Power Inspectorate should require that 1. The operators perform a risk analysis beforehand concerning the potential consequences that process orientation might convey: the analysis should contain a model specifying how safety is currently

  4. Critical Characteristics of Radiation Detection System Components to be Dedicated for use in Safety Class and Safety Significant System

    International Nuclear Information System (INIS)

    DAVIS, S.J.

    2000-01-01

    This document identifies critical characteristics of components to be dedicated for use in Safety Significant (SS) Systems, Structures, or Components (SSCs). This document identifies the requirements for the components of the common, radiation area, monitor alarm in the WESF pool cell. These are procured as Commercial Grade Items (CGI), with the qualification testing and formal dedication to be performed at the Waste Encapsulation Storage Facility (WESF) for use in safety significant systems. System modifications are to be performed in accordance with the approved design. Components for this change are commercially available and interchangeable with the existing alarm configuration This document focuses on the operational requirements for alarm, declaration of the safety classification, identification of critical characteristics, and interpretation of requirements for procurement. Critical characteristics are identified herein and must be verified, followed by formal dedication, prior to the components being used in safety related applications

  5. 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

  6. 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

  7. Criticality safety analyses in SKODA JS a.s

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.

    1999-01-01

    This paper describes criticality safety analyses of spent fuel systems for storage and transport of spent fuel performed in SKODA JS s.r.o.. Analyses were performed for different systems both at NPP site including originally designed spent fuel pool with a large pitch between assemblies without any special absorbing material, high density spent fuel pool with an additional absorption by boron steel, depository rack for fresh fuel assemblies with a very large pitch between fuel assemblies, a container for transport of fresh fuel into the reactor pool and a cask for transport and storage of spent fuel and container for final storage depository. required subcriticality has been proven taking into account all possible unfavourable conditions, uncertainties etc. In two cases, burnup credit methodology is expected to be used. (Authors)

  8. Robust optical sensors for safety critical automotive applications

    Science.gov (United States)

    De Locht, Cliff; De Knibber, Sven; Maddalena, Sam

    2008-02-01

    Optical sensors for the automotive industry need to be robust, high performing and low cost. This paper focuses on the impact of automotive requirements on optical sensor design and packaging. Main strategies to lower optical sensor entry barriers in the automotive market include: Perform sensor calibration and tuning by the sensor manufacturer, sensor test modes on chip to guarantee functional integrity at operation, and package technology is key. As a conclusion, optical sensor applications are growing in automotive. Optical sensor robustness matured to the level of safety critical applications like Electrical Power Assisted Steering (EPAS) and Drive-by-Wire by optical linear arrays based systems and Automated Cruise Control (ACC), Lane Change Assist and Driver Classification/Smart Airbag Deployment by camera imagers based systems.

  9. 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.

  10. 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....

  11. 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

  12. 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

  13. 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 ...

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. 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)

  16. NUSS safety standards: A critical assessment

    International Nuclear Information System (INIS)

    Minogue, R.B.

    1985-01-01

    The NUSS safety standards are based on systematic review of safety criteria of many countries in a process carefully defined to assure completeness of coverage. They represent an international consensus of accepted safety principles and practices for regulation and for the design, construction, and operation of nuclear power plants. They are a codification of principles and practices already in use by some Member States. Thus, they are not standards which describe methodologies at their present state of evolution as a result of more recent experience and improvements in technological understanding. The NUSS standards assume an underlying body of national standards and a defined technological base. Detailed design and industrial practices vary between countries and the implementation of basic safety standards within countries has taken approaches that conform with national industrial practices. Thus, application of the NUSS standards requires reconciliation with the standards of the country where the reactor will be built as well as with the country from which procurement takes place. Experience in making that reconciliation will undoubtedly suggest areas of needed improvement. After the TMI accident a reassessment of the NUSS programme was made and it was concluded that, given the information at that time and the then level of technology, the basic approach was sound; the NUSS programme should be continued to completion, and the standards should be brought into use. It was also recognized, however, that in areas such as probabilistic risk assessment, human factors methodology, and consideration of detailed accident sequences, more advanced technology was emerging. As these technologies develop, and become more amenable to practical application, it is anticipated that the NUSS standards will need revision. Ideally those future revisions will also flow from experience in their use

  17. 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)

  18. 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

  19. RECENT ADDITIONS OF CRITICALITY SAFETY RELATED INTEGRAL BENCHMARK DATA TO THE ICSBEP AND IRPHEP HANDBOOKS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Sartori

    2009-09-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

  20. Recent additions of criticality safety related integral benchmark data to the ICSBEP and IRPHEP handbooks

    International Nuclear Information System (INIS)

    Briggs, J. B.; Scott, L.; Rugama, Y.; Sartori, E.

    2009-01-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions. (authors)

  1. REcent Additions Of Criticality Safety Related Integral Benchmark Data To The Icsbep And Irphep Handbooks

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Scott, Lori; Rugama, Yolanda; Sartori, Enrico

    2009-01-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

  2. 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.)

  3. Proceedings of the Nuclear Criticality Technology and Safety Project Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, R.G. [comp.

    1994-01-01

    This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda.

  4. Proceedings of the Nuclear Criticality Technology and Safety Project Workshop

    International Nuclear Information System (INIS)

    Sanchez, R.G.

    1994-01-01

    This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda

  5. 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)

  6. 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)

  7. Criticality safety in high explosives dissolution

    International Nuclear Information System (INIS)

    Troyer, S.D.

    1997-01-01

    In 1992, an incident occurred at the Pantex Plant in which the cladding around a fissile material component (pit) cracked during dismantlement of the high explosives portion of a nuclear weapon. Although the event did not result in any significant contamination or personnel exposures, concerns about the incident led to the conclusion that the current dismantlement process was unacceptable. Options considered for redesign, dissolution tooling design considerations, dissolution tooling design features, and the analysis of the new dissolution tooling are summarized. The final tooling design developed incorporated a number of safety features and provides a simple, self-contained, low-maintenance method of high explosives removal for nuclear explosive dismantlement. Analyses demonstrate that the tooling design will remain subcritical under normal, abnormal, and credible accident scenarios. 1 fig

  8. Calculational study for criticality safety data of fissionable actinides

    International Nuclear Information System (INIS)

    Nojiri, Ichiro; Fukasaku, Yasuhiro.

    1997-01-01

    This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

  9. 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)

  10. Data-Centric Knowledge Discovery Strategy for a Safety-Critical Sensor Application

    Directory of Open Access Journals (Sweden)

    Nilamadhab Mishra

    2014-01-01

    Full Text Available In an indoor safety-critical application, sensors and actuators are clustered together to accomplish critical actions within a limited time constraint. The cluster may be controlled by a dedicated programmed autonomous microcontroller device powered with electricity to perform in-network time critical functions, such as data collection, data processing, and knowledge production. In a data-centric sensor network, approximately 3–60% of the sensor data are faulty, and the data collected from the sensor environment are highly unstructured and ambiguous. Therefore, for safety-critical sensor applications, actuators must function intelligently within a hard time frame and have proper knowledge to perform their logical actions. This paper proposes a knowledge discovery strategy and an exploration algorithm for indoor safety-critical industrial applications. The application evidence and discussion validate that the proposed strategy and algorithm can be implemented for knowledge discovery within the operational framework.

  11. Performance standards of road safety management

    Directory of Open Access Journals (Sweden)

    Čabarkapa Milenko R.

    2016-01-01

    Full Text Available Road safety management controlling means the process of finding out the information whether the road safety is improving in a measure to achieve the objectives. The process of control consists of three basic elements: definition of performances and standards, measurement of current performances and comparison with the set standards, and improvement of current performances, if they deviate from the set standards. The performance standards of road safety management system are focused on a performances measurement, in terms of their design and characteristics, in order to support the performances improvement of road safety system and thus, ultimately, improve the road safety. Defining the performance standards of road safety management system, except that determines the design of the system for performances measurement, directly sets requirements whose fulfillment will produce a road safety improvement. The road safety management system, based on the performance standards of road safety, with a focus on results, will produce the continuous improvement of road safety, achieving the long-term 'vision zero', the philosophy of road safety, that human life and health take priority over mobility and other traffic objectives of the road traffic.

  12. Nuclear criticality safety basics for personnel working with nuclear fissionable materials. Phase I

    International Nuclear Information System (INIS)

    Vausher, A.L.

    1984-10-01

    DOE order 5480.1A, Chapter V, ''Safety of Nuclear Facilities,'' establishes safety procedures and requirements for DOE nuclear facilities. The ''Nuclear Criticality Safety Basic Program - Phase I'' is documented in this report. The revised program has been developed to clearly illustrate the concept of nuclear safety and to help the individual employee incorporate safe behavior in his daily work performance. Because of this, the subject of safety has been approached through its three fundamentals: scientific basis, engineering criteria, and administrative controls. Only basics of these three elements were presented. 5 refs

  13. 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

  14. 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

  15. Test process for the safety-critical embedded software

    International Nuclear Information System (INIS)

    Sung, Ahyoung; Choi, Byoungju; Lee, Jangsoo

    2004-01-01

    Digitalization of nuclear Instrumentation and Control (I and C) system requires high reliability of not only hardware but also software. Verification and Validation (V and V) process is recommended for software reliability. But a more quantitative method is necessary such as software testing. Most of software in the nuclear I and C system is safety-critical embedded software. Safety-critical embedded software is specified, verified and developed according to V and V process. Hence two types of software testing techniques are necessary for the developed code. First, code-based software testing is required to examine the developed code. Second, after code-based software testing, software testing affected by hardware is required to reveal the interaction fault that may cause unexpected results. We call the testing of hardware's influence on software, an interaction testing. In case of safety-critical embedded software, it is also important to consider the interaction between hardware and software. Even if no faults are detected when testing either hardware or software alone, combining these components may lead to unexpected results due to the interaction. In this paper, we propose a software test process that embraces test levels, test techniques, required test tasks and documents for safety-critical embedded software. We apply the proposed test process to safety-critical embedded software as a case study, and show the effectiveness of it. (author)

  16. 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

  17. Is road safety management linked to road safety performance?

    Science.gov (United States)

    Papadimitriou, Eleonora; Yannis, George

    2013-10-01

    This research aims to explore the relationship between road safety management and road safety performance at country level. For that purpose, an appropriate theoretical framework is selected, namely the 'SUNflower' pyramid, which describes road safety management systems in terms of a five-level hierarchy: (i) structure and culture, (ii) programmes and measures, (iii) 'intermediate' outcomes'--safety performance indicators (SPIs), (iv) final outcomes--fatalities and injuries, and (v) social costs. For each layer of the pyramid, a composite indicator is implemented, on the basis of data for 30 European countries. Especially as regards road safety management indicators, these are estimated on the basis of Categorical Principal Component Analysis upon the responses of a dedicated road safety management questionnaire, jointly created and dispatched by the ETSC/PIN group and the 'DaCoTA' research project. Then, quasi-Poisson models and Beta regression models are developed for linking road safety management indicators and other indicators (i.e. background characteristics, SPIs) with road safety performance. In this context, different indicators of road safety performance are explored: mortality and fatality rates, percentage reduction in fatalities over a given period, a composite indicator of road safety final outcomes, and a composite indicator of 'intermediate' outcomes (SPIs). The results of the analyses suggest that road safety management can be described on the basis of three composite indicators: "vision and strategy", "budget, evaluation and reporting", and "measurement of road user attitudes and behaviours". Moreover, no direct statistical relationship could be established between road safety management indicators and final outcomes. However, a statistical relationship was found between road safety management and 'intermediate' outcomes, which were in turn found to affect 'final' outcomes, confirming the SUNflower approach on the consecutive effect of each layer

  18. Analyzing Software Requirements Errors in Safety-Critical, Embedded Systems

    Science.gov (United States)

    Lutz, Robyn R.

    1993-01-01

    This paper analyzes the root causes of safety-related software errors in safety-critical, embedded systems. The results show that software errors identified as potentially hazardous to the system tend to be produced by different error mechanisms than non- safety-related software errors. Safety-related software errors are shown to arise most commonly from (1) discrepancies between the documented requirements specifications and the requirements needed for correct functioning of the system and (2) misunderstandings of the software's interface with the rest of the system. The paper uses these results to identify methods by which requirements errors can be prevented. The goal is to reduce safety-related software errors and to enhance the safety of complex, embedded systems.

  19. Safety impacts of bicycle infrastructure: A critical review.

    Science.gov (United States)

    DiGioia, Jonathan; Watkins, Kari Edison; Xu, Yanzhi; Rodgers, Michael; Guensler, Randall

    2017-06-01

    This paper takes a critical look at the present state of bicycle infrastructure treatment safety research, highlighting data needs. Safety literature relating to 22 bicycle treatments is examined, including findings, study methodologies, and data sources used in the studies. Some preliminary conclusions related to research efficacy are drawn from the available data and findings in the research. While the current body of bicycle safety literature points toward some defensible conclusions regarding the safety and effectiveness of certain bicycle treatments, such as bike lanes and removal of on-street parking, the vast majority treatments are still in need of rigorous research. Fundamental questions arise regarding appropriate exposure measures, crash measures, and crash data sources. This research will aid transportation departments with regard to decisions about bicycle infrastructure and guide future research efforts toward understanding safety impacts of bicycle infrastructure. Copyright © 2017 Elsevier Ltd and National Safety Council. All rights reserved.

  20. 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)

  1. Validation of calculational methods for nuclear criticality safety - approved 1975

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, N16.1-1975, states in 4.2.5: In the absence of directly applicable experimental measurements, the limits may be derived from calculations made by a method shown to be valid by comparison with experimental data, provided sufficient allowances are made for uncertainties in the data and in the calculations. There are many methods of calculation which vary widely in basis and form. Each has its place in the broad spectrum of problems encountered in the nuclear criticality safety field; however, the general procedure to be followed in establishing validity is common to all. The standard states the requirements for establishing the validity and area(s) of applicability of any calculational method used in assessing nuclear criticality safety

  2. The International Criticality Safety Benchmark Evaluation Project on the Internet

    International Nuclear Information System (INIS)

    Briggs, J.B.; Brennan, S.A.; Scott, L.

    2000-01-01

    The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in October 1992 by the US Department of Energy's (DOE's) defense programs and is documented in the Transactions of numerous American Nuclear Society and International Criticality Safety Conferences. The work of the ICSBEP is documented as an Organization for Economic Cooperation and Development (OECD) handbook, International Handbook of Evaluated Criticality Safety Benchmark Experiments. The ICSBEP Internet site was established in 1996 and its address is http://icsbep.inel.gov/icsbep. A copy of the ICSBEP home page is shown in Fig. 1. The ICSBEP Internet site contains the five primary links. Internal sublinks to other relevant sites are also provided within the ICSBEP Internet site. A brief description of each of the five primary ICSBEP Internet site links is given

  3. Computational Methods for Sensitivity and Uncertainty Analysis in Criticality Safety

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Childs, R.L.; Rearden, B.T.

    1999-01-01

    Interest in the sensitivity methods that were developed and widely used in the 1970s (the FORSS methodology at ORNL among others) has increased recently as a result of potential use in the area of criticality safety data validation procedures to define computational bias, uncertainties and area(s) of applicability. Functional forms of the resulting sensitivity coefficients can be used as formal parameters in the determination of applicability of benchmark experiments to their corresponding industrial application areas. In order for these techniques to be generally useful to the criticality safety practitioner, the procedures governing their use had to be updated and simplified. This paper will describe the resulting sensitivity analysis tools that have been generated for potential use by the criticality safety community

  4. 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

  5. Critical safety function guidelines for experimental fusion facilities

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1989-01-01

    As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented. 10 refs., 6 figs

  6. Critical safety function guidelines for experimental fusion facilities

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1989-01-01

    As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented

  7. Safety performance monitoring of autonomous marine systems

    International Nuclear Information System (INIS)

    Thieme, Christoph A.; Utne, Ingrid B.

    2017-01-01

    The marine environment is vast, harsh, and challenging. Unanticipated faults and events might lead to loss of vessels, transported goods, collected scientific data, and business reputation. Hence, systems have to be in place that monitor the safety performance of operation and indicate if it drifts into an intolerable safety level. This article proposes a process for developing safety indicators for the operation of autonomous marine systems (AMS). The condition of safety barriers and resilience engineering form the basis for the development of safety indicators, synthesizing and further adjusting the dual assurance and the resilience based early warning indicator (REWI) approaches. The article locates the process for developing safety indicators in the system life cycle emphasizing a timely implementation of the safety indicators. The resulting safety indicators reflect safety in AMS operation and can assist in planning of operations, in daily operational decision-making, and identification of improvements. Operation of an autonomous underwater vehicle (AUV) exemplifies the process for developing safety indicators and their implementation. The case study shows that the proposed process leads to a comprehensive set of safety indicators. It is expected that application of the resulting safety indicators consequently will contribute to safer operation of current and future AMS. - Highlights: • Process for developing safety indicators for autonomous marine systems. • Safety indicators based on safety barriers and resilience thinking. • Location of the development process in the system lifecycle. • Case study on AUV demonstrating applicability of the process.

  8. Nursing home safety: does financial performance matter?

    Science.gov (United States)

    Oetjen, Reid M; Zhao, Mei; Liu, Darren; Carretta, Henry J

    2011-01-01

    This study examines the relationship between financial performance and selected safety measures of nursing homes in the State of Florida. We used descriptive analysis on a total sample of 1,197. Safety information was from the Online Survey, Certification and Reporting (OSCAR) data of 2003 to 2005, while the financial performance measures were from the Medicare cost reports of 2002 to 2004. Finally, we examined the most frequently cited deficiencies as well as the relationship between financial performance and quality indicators. Nursing homes in the bottom quartile of financial performance perform poorly on most resident-safety measures of care; however, nursing homes in the top two financial categories also experienced a higher number of deficiencies. Nursing homes in the next to lowest quartile of financial performance category best perform on most of these safety measures. The results reinforce the need to monitor nursing home quality and resident safety in US nursing homes, especially among facilities with poor overall financial performance.

  9. Criticality safety for TMI-2 canister storage at INEL

    International Nuclear Information System (INIS)

    Jones, R.R.; Briggs, J.B.; Ayers, A.L. Jr.

    1986-01-01

    Canisters containing Three Mile Island Unit 2 (TMI-2) core debris will be researched, stored, and prepared for final disposition at the Idaho National Engineering Laboratory (INEL). The canisters will be placed into storage modules and assembled into a storage rack, which will be located in the Test Area North (TAN) storage pool. Criticality safety calculations were made (a) to ensure that the storage rack is safe for both normal and accident conditions and (b) to determine the effects of degradation of construction materials (Boraflex and polyethylene) on criticality safety

  10. Nuclear Criticality Safety Organization training implementation. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-05-19

    The Nuclear Criticality Safety Organization (NCSO) 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. This document provides a listing of the roles and responsibilities of NCSO 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 Training Implementation document is applicable to all technical and managerial NCSO personnel, including temporary personnel, sub-contractors and/or LMES employees on loan to the NCSO, who are in a qualification program.

  11. Nuclear Criticality Safety Organization training implementation. 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 qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document provides a listing of the roles and responsibilities of NCSO 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 Training Implementation document is applicable to all technical and managerial NCSO personnel, including temporary personnel, sub-contractors and/or LMES employees on loan to the NCSO, who are in a qualification program

  12. 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)

  13. Safety analysis of the Los Alamos critical experiments facility

    International Nuclear Information System (INIS)

    Paxton, H.C.

    1975-10-01

    The safety of Pajarito Site critical assembly operations depends upon protection built into the facility, upon knowledgeable personnel, and upon good practice as defined by operating procedures and experimental plans. Distance, supplemented by shielding in some cases, would protect personnel against an extreme accident generating 10 19 fissions. During the facility's 28-year history, the direct cost of criticality accidents has translated to a risk of less than $200 per year

  14. Merger of Nuclear Data with Criticality Safety Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Derrien, H.; Larson, N.M.; Leal, L.C.

    1999-09-20

    In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes � measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations � which in the past have been treated independently.

  15. Merger of Nuclear Data with Criticality Safety Calculations

    International Nuclear Information System (INIS)

    Derrien, H.; Larson, N.M.; Leal, L.C.

    1999-01-01

    In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations which in the past have been treated independently

  16. Indicators to monitor NPP operational safety performance

    International Nuclear Information System (INIS)

    Gomez-Cobo, Ana

    2002-01-01

    Since December 1995 the IAEA activities on safety performance indicators focused on the elaboration of a framework for the establishment of an operational safety performance indicator programme. The development of this framework began with the consideration of the concept of NPP operational safety performance and the identification of operational safety attributes. For each operational safety attribute, overall indicators, envisioned as providing an overall evaluation of relevant aspects of safety performance, were established. Associated with each overall indicator is a level of strategic indicators intended to provide a bridge from overall to specific indicators. Finally each strategic indicator was supported by a set of specific indicators, which represent quantifiable measures of performance. The programme development was enhanced by pilot plant studies, conducted over a 15 month period from January 1998 to March 1999. The result of all this work is compiled in the IAEA-TECDOC-1141, to be published shortly. This paper presents a summary of this IAEA TECDOC. It describes the operational safety performance indicator framework proposed and discusses the results of and lessons learned from the pilot studies. Despite the efforts described, it is clear that additional research is still necessary in areas such as plant-specific adaptation of proposed frameworks in order to suit individual data collection systems and plant characteristics, indicator selection, indicator definition, goal setting, action thresholds, analysis of trends, indicator display systems, analysis of overall safety performance (i.e., aggregation or combination of indicators), safety culture indicators, qualitative indicators, and use of additional indicators to address issues such as industrial safety attitude and performance, staff welfare, and environmental compliance. This is the rationale for a new IAEA Coordinated Research Project on 'Development and application of indicators to monitor NPP

  17. 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)

  18. Experience of safety and performance improvement for fuel handling equipment

    International Nuclear Information System (INIS)

    Gyoon Chang, Sang; Hee Lee, Dae

    2014-01-01

    The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

  19. Private Memory Allocation Analysis for Safety-Critical Java

    DEFF Research Database (Denmark)

    Dalsgaard, Andreas E.; Hansen, René Rydhof; Schoeberl, Martin

    2012-01-01

    Safety-critical Java (SCJ) avoids garbage collection and uses a scope based memory model. This memory model is based on a restricted version of RTSJ [2] style scopes. The scopes form a clear hierarchy with different lifetimes. Therefore, references between objects in different scopes are only...

  20. Chip-Multiprocessor Hardware Locks for Safety-Critical Java

    DEFF Research Database (Denmark)

    Strøm, Torur Biskopstø; Puffitsch, Wolfgang; Schoeberl, Martin

    2013-01-01

    and may void a task set's schedulability. In this paper we present a hardware locking mechanism to reduce the synchronization overhead. The solution is implemented for the chip-multiprocessor version of the Java Optimized Processor in the context of safety-critical Java. The implementation is compared...

  1. 14 CFR 417.121 - Safety critical preflight operations.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Safety critical preflight operations. 417.121 Section 417.121 Aeronautics and Space COMMERCIAL SPACE TRANSPORTATION, FEDERAL AVIATION... surveillance. A launch operator must implement its hazard area surveillance and clearance plan, of § 417.111(j...

  2. Nuclear Criticality Safety Assessment for Tank 38H Salt Dissolution

    International Nuclear Information System (INIS)

    Davis, P.L.

    1996-01-01

    This assessment report of sample results of the accumulating insoluble solids from Tank 38H demonstrates that an inherent subcritical condition for nuclear criticality safety exists during saltcake dissolution. This report also defines criteria for future sampling of Tank 38H for continued verification of the inherent subcritical condition as saltcake dissolution proceeds

  3. Safety prediction for basic components of safety critical software based on static testing

    International Nuclear Information System (INIS)

    Son, H.S.; Seong, P.H.

    2001-01-01

    The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, both of which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

  4. Safety prediction for basic components of safety-critical software based on static testing

    International Nuclear Information System (INIS)

    Son, H.S.; Seong, P.H.

    2000-01-01

    The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

  5. Nuclear criticality safety calculational analysis for small-diameter containers

    International Nuclear Information System (INIS)

    LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.

    1995-11-01

    This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can (open-quotes F-canclose quotes); (4) 5.25-inch-diameter by 15-inch-high steel can (open-quotes Z-canclose quotes); and (5) 5.0-inch-diameter by 9-inch-high polybottle (open-quotes CO-4close quotes). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO 2 F 2 +H 2 O and UF 4 +oil materials at 100% and 10% enrichments and U 3 O 8 , and H 2 O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k eff + 2σ < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant

  6. Prospective safety performance evaluation on construction sites.

    Science.gov (United States)

    Wu, Xianguo; Liu, Qian; Zhang, Limao; Skibniewski, Miroslaw J; Wang, Yanhong

    2015-05-01

    This paper presents a systematic Structural Equation Modeling (SEM) based approach for Prospective Safety Performance Evaluation (PSPE) on construction sites, with causal relationships and interactions between enablers and the goals of PSPE taken into account. According to a sample of 450 valid questionnaire surveys from 30 Chinese construction enterprises, a SEM model with 26 items included for PSPE in the context of Chinese construction industry is established and then verified through the goodness-of-fit test. Three typical types of construction enterprises, namely the state-owned enterprise, private enterprise and Sino-foreign joint venture, are selected as samples to measure the level of safety performance given the enterprise scale, ownership and business strategy are different. Results provide a full understanding of safety performance practice in the construction industry, and indicate that the level of overall safety performance situation on working sites is rated at least a level of III (Fair) or above. This phenomenon can be explained that the construction industry has gradually matured with the norms, and construction enterprises should improve the level of safety performance as not to be eliminated from the government-led construction industry. The differences existing in the safety performance practice regarding different construction enterprise categories are compared and analyzed according to evaluation results. This research provides insights into cause-effect relationships among safety performance factors and goals, which, in turn, can facilitate the improvement of high safety performance in the construction industry. Copyright © 2015 Elsevier Ltd. All rights reserved.

  7. Criticality safety enhancements for SCALE 6.2 and beyond

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Bekar, Kursat B.; Celik, Cihangir; Clarno, Kevin T.; Dunn, Michael E.; Hart, Shane W.; Ibrahim, Ahmad M.; Johnson, Seth R.; Langley, Brandon R.; Lefebvre, Jordan P.; Lefebvre, Robert A.; Marshall, William J.; Mertyurek, Ugur; Mueller, Don; Peplow, Douglas E.; Perfetti, Christopher M.; Petrie Jr, Lester M.; Thompson, Adam B.; Wiarda, Dorothea; Wieselquist, William A.; Williams, Mark L.

    2015-01-01

    SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. Since 1980, regulators, industry, and research institutions around the world have relied on SCALE for nuclear safety analysis and design. SCALE 6.2 provides several new capabilities and significant improvements in many existing features for criticality safety analysis. Enhancements are realized for nuclear data; multigroup resonance self-shielding; continuous-energy Monte Carlo analysis for sensitivity/uncertainty analysis, radiation shielding, and depletion; and graphical user interfaces. An overview of these capabilities is provided in this paper, and additional details are provided in several companion papers.

  8. Criticality studies: One of the two pillars of criticality safety at the Belgonucleaire MOX plant

    International Nuclear Information System (INIS)

    Lance, B.; Maldague, T.; Evrard, G.; Renard, A.; Kockerols, P.

    2001-01-01

    The present paper focuses on the criticality studies performed by the Engineering Division of Belgonucleaire. These are one of the two pillars of the criticality prevention implemented for the Belgonucleaire MOX producing plant. (author)

  9. 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 – Ck's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.

  10. 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.

  11. SRTC criticality safety technical review: Nuclear criticality safety evaluation 94-02, uranium solidification facility pencil tank module spacing

    International Nuclear Information System (INIS)

    Rathbun, R.

    1994-01-01

    Review of NMP-NCS-94-0087, ''Nuclear Criticality Safety Evaluation 94-02: Uranium Solidification Facility Pencil Tank Module Spacing (U), April 18, 1994,'' was requested of the SRTC Applied Physics Group. The NCSE is a criticality assessment to show that the USF process module spacing, as given in Non-Conformance Report SHM-0045, remains safe for operation. The NCSE under review concludes that the module spacing as given in Non-Conformance Report SHM-0045 remains in a critically safe configuration for all normal and single credible abnormal conditions. After a thorough review of the NCSE, this reviewer agrees with that conclusion

  12. 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

  13. Evaluation for nuclear safety-critical software reliability of DCS

    International Nuclear Information System (INIS)

    Liu Ying

    2015-01-01

    With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I and C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make besement for evaluating the reliability and safety of DCS. (author)

  14. From Safety Critical Java Programs to Timed Process Models

    DEFF Research Database (Denmark)

    Thomsen, Bent; Luckow, Kasper Søe; Thomsen, Lone Leth

    2015-01-01

    frameworks, we have in recent years pursued an agenda of translating hard-real-time embedded safety critical programs written in the Safety Critical Java Profile [33] into networks of timed automata [4] and subjecting those to automated analysis using the UPPAAL model checker [10]. Several tools have been...... built and the tools have been used to analyse a number of systems for properties such as worst case execution time, schedulability and energy optimization [12–14,19,34,36,38]. In this paper we will elaborate on the theoretical underpinning of the translation from Java programs to timed automata models...... and briefly summarize some of the results based on this translation. Furthermore, we discuss future work, especially relations to the work in [16,24] as Java recently has adopted first class higher order functions in the form of lambda abstractions....

  15. Cluster monte carlo method for nuclear criticality safety calculation

    International Nuclear Information System (INIS)

    Pei Lucheng

    1984-01-01

    One of the most important applications of the Monte Carlo method is the calculation of the nuclear criticality safety. The fair source game problem was presented at almost the same time as the Monte Carlo method was applied to calculating the nuclear criticality safety. The source iteration cost may be reduced as much as possible or no need for any source iteration. This kind of problems all belongs to the fair source game prolems, among which, the optimal source game is without any source iteration. Although the single neutron Monte Carlo method solved the problem without the source iteration, there is still quite an apparent shortcoming in it, that is, it solves the problem without the source iteration only in the asymptotic sense. In this work, a new Monte Carlo method called the cluster Monte Carlo method is given to solve the problem further

  16. Determination of Safety Performance Grade of NPP Using Integrated Safety Performance Assessment (ISPA) Program

    International Nuclear Information System (INIS)

    Chung, Dae Wook

    2011-01-01

    Since the beginning of 2000, the safety regulation of nuclear power plant (NPP) has been challenged to be conducted more reasonable, effective and efficient way using risk and performance information. In the United States, USNRC established Reactor Oversight Process (ROP) in 2000 for improving the effectiveness of safety regulation of operating NPPs. The main idea of ROP is to classify the NPPs into 5 categories based on the results of safety performance assessment and to conduct graded regulatory programs according to categorization, which might be interpreted as 'Graded Regulation'. However, the classification of safety performance categories is highly comprehensive and sensitive process so that safety performance assessment program should be prepared in integrated, objective and quantitative manner. Furthermore, the results of assessment should characterize and categorize the actual level of safety performance of specific NPP, integrating all the substantial elements for assessing the safety performance. In consideration of particular regulatory environment in Korea, the integrated safety performance assessment (ISPA) program is being under development for the use in the determination of safety performance grade (SPG) of a NPP. The ISPA program consists of 6 individual assessment programs (4 quantitative and 2 qualitative) which cover the overall safety performance of NPP. Some of the assessment programs which are already implemented are used directly or modified for incorporating risk aspects. The others which are not existing regulatory programs are newly developed. Eventually, all the assessment results from individual assessment programs are produced and integrated to determine the safety performance grade of a specific NPP

  17. Hardware Support for Safety-critical Java Scope Checks

    DEFF Research Database (Denmark)

    Rios Rivas, Juan Ricardo; Schoeberl, Martin

    2012-01-01

    Memory management in Safety-Critical Java (SCJ) is based on time bounded, non garbage collected scoped memory regions used to store temporary objects. Scoped memory regions may have different life times during the execution of a program and hence, to avoid leaving dangling pointers, it is necessary...... in terms of execution time for applications where cross-scope references are frequent. Our proposal was implemented and tested on the Java Optimized Processor (JOP)....

  18. Life extension decision making of safety critical systems: An overview

    OpenAIRE

    Shafiee, Mahmood; Animah, I.

    2017-01-01

    In recent years, the concept of “asset life extension” has become increasingly important to safety critical industries including nuclear power, offshore oil and gas, petrochemical, renewable energy, rail transport, aviation, shipping, electricity distribution and transmission, etc. Extending the service life of industrial assets can offer a broad range of economic, technical, social and environmental benefits as compared to other end-of-life management strategies such as decommissioning and r...

  19. A Test Suite for Safety-Critical Java using JML

    DEFF Research Database (Denmark)

    Ravn, Anders Peter; Søndergaard, Hans

    2013-01-01

    Development techniques are presented for a test suite for the draft specification of the Java profile for Safety-Critical Systems. Distinguishing features are: specification of conformance constraints in the Java Modeling Language, encoding of infrastructure concepts without implementation bias......, and corresponding specifications of implicitly stated behavioral and real-time properties. The test programs are auto-generated from the specification, while concrete values for test parameters are selected manually. The suite is open source and publicly accessible....

  20. Use of modern software - based instrumentation in safety critical systems

    International Nuclear Information System (INIS)

    Emmett, J.; Smith, B.

    2005-01-01

    Many Nuclear Power Plants are now ageing and in need of various degrees of refurbishment. Installed instrumentation usually uses out of date 'analogue' technology and is often no longer available in the market place. New technology instrumentation is generally un-qualified for nuclear use and specifically the new 'smart' technology contains 'firmware', (effectively 'soup' (Software of Uncertain Pedigree)) which must be assessed in accordance with relevant safety standards before it may be used in a safety application. Particular standards are IEC 61508 [1] and the British Energy (BE) PES (Programmable Electronic Systems) guidelines EPD/GEN/REP/0277/97. [2] This paper outlines a new instrument evaluation system, which has been developed in conjunction with the UK Nuclear Industry. The paper concludes with a discussion about on-line monitoring of Smart instrumentation in safety critical applications. (author)

  1. 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

  2. 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

  3. Safety performance of preliminary KALIMER conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

    1999-07-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  4. Safety performance of preliminary KALIMER conceptual design

    International Nuclear Information System (INIS)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

    1999-01-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  5. 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

  6. 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.

  7. 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

  8. Criticality safety issues in the disposition of BN-350 spent fuel

    International Nuclear Information System (INIS)

    Schaefer, R. W.; Klann, R. T.; Koltyshev, S. M.; Krechetov, S.

    2000-01-01

    A criticality safety analysis has been performed as part of the BN-350 spent fuel disposition project being conducted jointly by the DOE and Kazakhstan. The Kazakhstan regulations are reasonably consistent with those of the DOE. The high enrichment and severe undermoderation of this fast reactor fuel has significant criticality safety consequences. A detailed modeling approach was used that showed some configurations to be safe that otherwise would be rejected. Reasonable requirements for design and operations were needed, and with them, all operations were found to be safe

  9. Definition and means of maintaining the criticality detectors and alarms portion of the PFP safety envelope

    International Nuclear Information System (INIS)

    White, W.F.

    1997-01-01

    The Criticality Alarm System (CAS) provides continuous detection for high radiation (criticality) events and automatically initiates an evacuation signal to affected personnel. The Safety Envelope (SE) for PFP includes the necessary equipment and the required procedures to ensure the CAS is capable of performing its intended function. This document provides the definition and means of maintaining the SE for PFP related to the CAS. This document also identifies and provides a justification for those portions of the CAS excluded from the PFP Safety Envelope

  10. Nuclear Criticality Safety Organization guidance for the development of continuing technical training. Revision 1

    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 at the Oak Ridge Y-12 Plant and throughout the DOE complex. Continuing technical training is training outside of the initial qualification program to address identified organization-wide needs. Typically, this training is used to improve organization performance in the conduct of business. This document provides guidelines for the development of the technical portions of the Continuing Training Program. It is not a step-by-step procedure, but a collection of considerations to be used during the development process

  11. Performance of balanced bellows safety relief valves

    International Nuclear Information System (INIS)

    Lai, Y.S.

    1992-01-01

    By the nature of its design, the set point and lift of a conventional spring loaded safety relief valve are sensitive to back pressure. One way to reduce the adverse effects of the back pressure on the safety relief valve function is to install a balanced bellows in a safety relief valve. The metallic bellows has a rather wide range of manufacturing tolerance which makes the design of the bellows safety relief valve very complicated. The state-of-the-art balanced bellows safety relief valve can only substantially minimize, but cannot totally eliminate the back pressure effects on its set point and relieving capacity. Set point change is a linear function of the back pressure to the set pressure ratio. Depending on the valve design, the set point correction factor can be either greater or smaller than unity. There exists an allowable back pressure and critical back pressure for each safety relief valve. When total back pressure exceeds the R a , the relieving capacity will be reduced mainly resulting from the valve lift being reduced by the back pressure and the capacity reduction factor should be applied in valve sizing. Once the R c is exceeded, the safety relief valve becomes unstable and loses its over pressure protection capability. The capacity reduction factor is a function of system overpressure, but their relationship is non-linear in nature. (orig.)

  12. 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)

  13. Operation, Safety and Human: Critical Factors for the Success of Railway Transportation

    NARCIS (Netherlands)

    Rajabali Nejad, Mohammadreza; Martinetti, Alberto; van Dongen, Leonardus Adriana Maria

    2016-01-01

    This paper focuses on three categories of performance indicators for railway transportation: the excellence of operation, system safety and human factors. These are among the most critical indicators for delivering high quality services. This paper discusses the main issues, challenges and future

  14. 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

  15. High level issues in reliability quantification of safety-critical software

    International Nuclear Information System (INIS)

    Kim, Man Cheol

    2012-01-01

    For the purpose of developing a consensus method for the reliability assessment of safety-critical digital instrumentation and control systems in nuclear power plants, several high level issues in reliability assessment of the safety-critical software based on Bayesian belief network modeling and statistical testing are discussed. Related to the Bayesian belief network modeling, the relation between the assessment approach and the sources of evidence, the relation between qualitative evidence and quantitative evidence, how to consider qualitative evidence, and the cause-consequence relation are discussed. Related to the statistical testing, the need of the consideration of context-specific software failure probabilities and the inability to perform a huge number of tests in the real world are discussed. The discussions in this paper are expected to provide a common basis for future discussions on the reliability assessment of safety-critical software. (author)

  16. Criticality safety of low-density storage arrays

    International Nuclear Information System (INIS)

    Bauer, T. H.; Nuclear Engineering Division

    2005-01-01

    This paper proposes a straightforward bounding method for the criticality safety analysis of fissionable materials configured into large arrays of standard containers. While criticality-safe storage limits have been well established for single containers, even under flooded conditions, it is also necessary to rule out any potential for criticality arising from neutronic interactions among multiple containers that might build up over long distances in a large array. Traditionally, the array problem has been approached by individual Monte Carlo analyses of explicit arrangements of single units and their surroundings. Deemphasizing specific configurations, the present technique takes advantage of low average density of fissionable material in typical storage arrays to separate neutron interactions that take place in the neutron's 'birth unit' from subsequent interactions in a dilute array. Numerous explicit Monte Carlo analyses show that array effects may be conservatively calculated by analyses that homogenize fissionable contents and depend only on the overall array shape, size, and reflective boundary

  17. Criticality safety of low-density storage arrays

    International Nuclear Information System (INIS)

    Bauer, T.H.

    1996-01-01

    This paper proposes a straightforward bounding method for the criticality safety analysis of fissionable materials configured into large arrays of standard containers. While criticality-safe storage limits have been well established for single containers, even under flooded conditions, it is also necessary to rule out any potential for criticality arising from neutronic interactions among multiple containers that might build up over long distances in a large array. Traditionally, the array problem has been approached by individual Monte Carlo analyses of explicit arrangements of single units and their surroundings. Deemphasizing specific configurations, the present technique takes advantage of low average density of fissionable material in typical storage arrays to separate neutron interactions that take place in the neutron's open-quotes birth unitclose quotes from subsequent interactions in a dilute array. Numerous explicit Monte Carlo analyses show that array effects may be conservatively calculated by analyses that homogenize fissionable contents and depend only on the overall array shape, size, and reflective boundary

  18. Ending on a positive: Examining the role of safety leadership decisions, behaviours and actions in a safety critical situation.

    Science.gov (United States)

    Donovan, Sarah-Louise; Salmon, Paul M; Horberry, Timothy; Lenné, Michael G

    2018-01-01

    Safety leadership is an important factor in supporting safe performance in the workplace. The present case study examined the role of safety leadership during the Bingham Canyon Mine high-wall failure, a significant mining incident in which no fatalities or injuries were incurred. The Critical Decision Method (CDM) was used in conjunction with a self-reporting approach to examine safety leadership in terms of decisions, behaviours and actions that contributed to the incidents' safe outcome. Mapping the analysis onto Rasmussen's Risk Management Framework (Rasmussen, 1997), the findings demonstrate clear links between safety leadership decisions, and emergent behaviours and actions across the work system. Communication and engagement based decisions featured most prominently, and were linked to different leadership practices across the work system. Further, a core sub-set of CDM decision elements were linked to the open flow and exchange of information across the work system, which was critical to supporting the safe outcome. The findings provide practical implications for the development of safety leadership capability to support safety within the mining industry. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Development of safety performance indicators in Japan

    International Nuclear Information System (INIS)

    Ohashi, H.; Tamao, S.; Tanaka, J.; Sawayama, T.

    2001-01-01

    For the purpose of safety regulations of operating nuclear power stations in Japan, the regulatory authorities utilize two types of regulations. One is the direct regulation, such as periodical inspection to inspect the function and performance of equipment important to safety, and the other is the audit type regulation such as preservation inspection to audit the compliance with the safety preservation rules. As performance indicators are expected to be an effective tool to evaluate the activities by audit type regulations, NUPEC is studying a comprehensive set of operational performance indicators to meet the effective evaluation method for the safety preservation activities in the audit type regulations under the frame of current safety regulation system. The study includes the establishment of comprehensive operational performance indicators applicable in Japan, the effective application of performance indicators to the current Japanese regulation, the clarification of the applicable scope of utilization, the possibility of applying the performance indicators. This report describes the present status of our performance indicator studies. After the completion of these studies the regulatory authorities will evaluate if and how the new set of comprehensive performance indicators could be introduced to Japanese regulatory scheme. (author)

  20. Characteristics of safety critical organizations . work psychological perspective

    International Nuclear Information System (INIS)

    Oedewald, P.; Reiman, T.

    2006-02-01

    This book deals with organizations that operate in high hazard industries, such as the nuclear power, aviation, oil and chemical industry organisations. The society puts a great strain on these organisations to rigorously manage the risks inherent in the technology they use and the products they produce. In this book, an organisational psychology view is taken to analyse what are the typical challenges of daily work in these environments. The analysis is based on a literature review about human and organisational factors in safety critical industries, and on the interviews of Finnish safety experts and safety managers from four different companies. In addition to this, personnel interviews conducted in the Finnish nuclear power plants are utilised. The authors come up with eight themes that seem to be common organizational challenges cross the industries. These include e.g. how does the personnel understand the risks and what is the right level for rules and procedures to guide the work activities. The primary aim of this book is to contribute to the Finnish nuclear safety research and safety management discussion. However, the book is equally suitable for risk management, organizational development and human resources management specialists in different industries. The purpose is to encourage readers to consider how the human and organizational factors are seen in the field they work in. (orig.)

  1. Critical safety issues in the design of fusion machines

    International Nuclear Information System (INIS)

    Kramer, W.

    1991-01-01

    In the course of developing fusion machines both general safety considerations and safety assessments for the various components and systems of actual machines increase in number and become more and more coherent. This is particularly true for the NET/ITER projects where safety analysis plays an increasing role for the design of the machine. Since in a D/T tokamak the radiological hazards will be dominant basic radiological safety objectives are discussed. Critical safety issues as identified in particular by the NET/ITER community are reviewed. Subsequently, issues of major concern are considered both for normal operation and for conceivable accidents. The following accidents are considered to be crucial: Loss of cooling in plasma facing components, loss of vacuum, tritium system failure, and magnet system failure. To mitigate accident consequences a confinement concept based on passive features and multiple barriers including detritiation and filtering has to be applied. The reactor building as final barrier needs special attention to cope with both internal and external hazards. (orig.)

  2. Special characteristics of safety critical organizations. Work psychological perspective

    Energy Technology Data Exchange (ETDEWEB)

    Oedewald, P.; Reiman, T.

    2007-03-15

    This book deals with organizations that operate in high hazard industries, such as the nuclear power, aviation, oil and chemical industry organizations. The society puts a great strain on these organizations to rigorously manage the risks inherent in the technology they use and the products they produce. In this book, an organizational psychology view is taken to analyse what are the typical challenges of daily work in these environments. The analysis is based on a literature review about human and organizational factors in safety critical industries, and on the interviews of Finnish safety experts and safety managers from four different companies. In addition to this, personnel interviews conducted in the Finnish nuclear power plants are utilised. The authors come up with eight themes that seem to be common organizational challenges cross the industries. These include e.g. how does the personnel understand the risks and what is the right level for rules and procedures to guide the work activities. The primary aim of this book is to contribute to the nuclear safety research and safety management discussion. However, the book is equally suitable for risk management, organizational development and human resources management specialists in different industries. The purpose is to encourage readers to consider how the human and organizational factors are seen in the field they work in. (orig.)

  3. Evaluating safety-critical organizations - emphasis on the nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Reiman, Teemu; Oedewald, Pia (VTT, Technical Research Centre of Finland (Finland))

    2009-04-15

    - it is understood that safety is a complex phenomenon. Safety is understood as a property of an entire system and not just absence of incidents - people feel personally responsible for the safety of the entire system, they feel they can have an effect on safety - the organizations aims for understanding the hazards and anticipating the risks in their activities - the organization is alert to the possibility of an unanticipated event - good prerequisites for carrying out the daily work exist. An organizational evaluation should aim at reasoning the: - sources of effectiveness in the organizational dimensions - sources of ineffectiveness in the organization dimensions - social processes in the organization - psychological outcomes of the current organization on a personnel level, e.g. motivation, understanding of hazards and sense of control. When drawing inferences from the organizational evaluations and defining development initiatives, it is important to consider actions that will promote and maintain the strengths of the organization as well as actions that will address and develop the weak areas. Issues associated with data collection and choice of methods has been a topic of much discussion in the field of evaluation of safety-critical organizations. We argue that the problem of collecting data is not the most important problem in terms of facilitating valid evaluations. A more important problem concerns the criteria that are used, as well as the operationalization of criteria into something measurable. Too much effort has been spent on methods and too little on contemplating the question of valid evaluation criteria and a valid means of deducing from the data whether the criteria are fulfilled. In order to accomplish this, a valid evaluation framework is needed, which incorporates the idea of organization as a complex sociotechnical system. This report has been an attempt to illustrate the premises and key issues to consider in organizational evaluations. No

  4. Evaluating safety-critical organizations - emphasis on the nuclear industry

    International Nuclear Information System (INIS)

    Reiman, Teemu; Oedewald, Pia

    2009-04-01

    understood that safety is a complex phenomenon. Safety is understood as a property of an entire system and not just absence of incidents - people feel personally responsible for the safety of the entire system, they feel they can have an effect on safety - the organizations aims for understanding the hazards and anticipating the risks in their activities - the organization is alert to the possibility of an unanticipated event - good prerequisites for carrying out the daily work exist. An organizational evaluation should aim at reasoning the: - sources of effectiveness in the organizational dimensions - sources of ineffectiveness in the organization dimensions - social processes in the organization - psychological outcomes of the current organization on a personnel level, e.g. motivation, understanding of hazards and sense of control. When drawing inferences from the organizational evaluations and defining development initiatives, it is important to consider actions that will promote and maintain the strengths of the organization as well as actions that will address and develop the weak areas. Issues associated with data collection and choice of methods has been a topic of much discussion in the field of evaluation of safety-critical organizations. We argue that the problem of collecting data is not the most important problem in terms of facilitating valid evaluations. A more important problem concerns the criteria that are used, as well as the operationalization of criteria into something measurable. Too much effort has been spent on methods and too little on contemplating the question of valid evaluation criteria and a valid means of deducing from the data whether the criteria are fulfilled. In order to accomplish this, a valid evaluation framework is needed, which incorporates the idea of organization as a complex sociotechnical system. This report has been an attempt to illustrate the premises and key issues to consider in organizational evaluations. No method can

  5. Criticality safety analysis of a calciner exit chute

    International Nuclear Information System (INIS)

    Haught, C.F.; Basoglu, B.; Brewer, R.W.; Hollenback, D.F.; Wilkinson, A.D.; Dodds, H.L.

    1994-01-01

    Calcination of uranyl nitrate into uranium oxide is part of normal operations of some enrichment plants. Typically, a calciner discharges uranium oxide powder (U 3 O 8 ) into an exit chute that directs the powder into a receiving can located in a glove box. One possible scenario for a criticality accident is the exit chute becoming blocked with powder near its discharge. The blockage restricts the flow of powder causing the exit chute to become filled with the powder. If blockage does occur, the height of the powder could reach a level that would not be safe from a criticality point of view. In this analysis, the subcritical height limit is examined for 98% enriched U 3 O 8 in the exit chute with full water reflection and optimal water moderation. The height limit for ensuring criticality safety during such an accumulation is 28.2 cm above the top of the discharge pipe at the bottom of the chute. Chute design variations are also evaluated with full water reflection and optimal water moderation. Subcritical configurations for the exit chute variation are developed, but the configurations are not safe when combined with the calciner. To ensure criticality safety, modifications must be made to the calciner tube or safety measures must be implemented if these designs are to be utilized with 98% enriched material. A geometrically safe configuration for the exit chute is developed for a blockage of 20% enriched powder with full water reflection and optimal water moderation, and this configuration is safe when combined with the existing calciner

  6. Food safety performance indicators to benchmark food safety output of food safety management systems

    NARCIS (Netherlands)

    Jacxsens, L.; Uyttendaele, M.; Devlieghere, F.; Rovira, J.; Oses Gomez, S.; Luning, P.A.

    2010-01-01

    There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses.

  7. Analysing context-dependent deviations in interacting with safety-critical systems

    International Nuclear Information System (INIS)

    Paterno, Fabio; Santoro, Carmen

    2006-01-01

    Mobile technology is penetrating many areas of human life. This implies that the context of use can vary in many respects. We present a method that aims to support designers in managing the complex design space when considering applications with varying contexts and help them to identify solutions that support users in performing their activities while preserving usability and safety. The method is a novel combination of an analysis of both potential deviations in task performance and most suitable information representations based on distributed cognition. The originality of the contribution is in providing a conceptual tool for better understanding the impact of context of use on user interaction in safety-critical domains. In order to present our approach we provide an example in which the implications of introducing new support through mobile devices in a safety-critical system are identified and analysed in terms of potential hazards

  8. Exemption, exception and other criteria for transport criticality safety

    International Nuclear Information System (INIS)

    Mennerdahl, D.

    2004-01-01

    Many strange concepts, requirements and specifications related to criticality safety are present in the Regulations. Some earlier problems have been corrected but, going back to 1961 and the first edition of the Regulations, it seems as many changes have been to the worse. Fissile material was defined correctly as a material that could consist of or contain fissile nuclides. Materials consisting of pure fissile nuclides don't exist but are important in package designs. 238 Pu was included as a fissile nuclide only as an emergency, because there was no alternative, but this caused some people to think that all nuclides supporting criticality are fissile. Neutron interaction between different (non-identical) packages had to be evaluated, making the transport index or allowable number of packages a credible safety control. That is not true anymore. The 15 gram exception limit for fissile nuclides was combined with a transport mode limit, similar to but more restrictive than the current consignment limit. The confinement system was introduced to help with formulation of a single requirement for safety of the containment system but is becoming something very different. Controls before the first use of a packaging have become controls of the first use of a package, supporting multiple shipments of the same package. The lack of exemption limits for fissile material essentially makes all radioactive materials fissile (all radioactive material contains some fissile atoms). Radioactive material seems to be defined without consideration of the criticality hazard of the material. LSA materials are defined with consideration of criticality, but only relates to quantities in fissile exceptions when other properties can be equally or more important. In July 2004, a number of proposals to IAEA have been submitted by Sweden to improve and expand the criticality safety control of the Regulations. Essential is the introduction of the fissionable nuclide and material concepts in

  9. Exemption, exception and other criteria for transport criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Mennerdahl, D. [E Mennerdahl Systems, Taeby (Sweden)

    2004-07-01

    Many strange concepts, requirements and specifications related to criticality safety are present in the Regulations. Some earlier problems have been corrected but, going back to 1961 and the first edition of the Regulations, it seems as many changes have been to the worse. Fissile material was defined correctly as a material that could consist of or contain fissile nuclides. Materials consisting of pure fissile nuclides don't exist but are important in package designs. {sup 238}Pu was included as a fissile nuclide only as an emergency, because there was no alternative, but this caused some people to think that all nuclides supporting criticality are fissile. Neutron interaction between different (non-identical) packages had to be evaluated, making the transport index or allowable number of packages a credible safety control. That is not true anymore. The 15 gram exception limit for fissile nuclides was combined with a transport mode limit, similar to but more restrictive than the current consignment limit. The confinement system was introduced to help with formulation of a single requirement for safety of the containment system but is becoming something very different. Controls before the first use of a packaging have become controls of the first use of a package, supporting multiple shipments of the same package. The lack of exemption limits for fissile material essentially makes all radioactive materials fissile (all radioactive material contains some fissile atoms). Radioactive material seems to be defined without consideration of the criticality hazard of the material. LSA materials are defined with consideration of criticality, but only relates to quantities in fissile exceptions when other properties can be equally or more important. In July 2004, a number of proposals to IAEA have been submitted by Sweden to improve and expand the criticality safety control of the Regulations. Essential is the introduction of the fissionable nuclide and material

  10. Mining Behavior Based Safety Data to Predict Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey C. Joe

    2010-06-01

    The Idaho National Laboratory (INL) operates a behavior based safety program called Safety Observations Achieve Results (SOAR). This peer-to-peer observation program encourages employees to perform in-field observations of each other's work practices and habits (i.e., behaviors). The underlying premise of conducting these observations is that more serious accidents are prevented from occurring because lower level “at risk” behaviors are identified and corrected before they can propagate into culturally accepted “unsafe” behaviors that result in injuries or fatalities. Although the approach increases employee involvement in safety, the premise of the program has not been subject to sufficient empirical evaluation. The INL now has a significant amount of SOAR data on these lower level “at risk” behaviors. This paper describes the use of data mining techniques to analyze these data to determine whether they can predict if and when a more serious accident will occur.

  11. Characterization strategy report for the criticality safety issue

    International Nuclear Information System (INIS)

    Doherty, A.L.; Doctor, P.G.; Felmy, A.R.; Prichard, A.W.; Serne, R.J.

    1997-06-01

    High-level radioactive waste from nuclear fuels processing is stored in underground waste storage tanks located in the tank farms on the Hanford Site. Waste in tank storage contains low concentrations of fissile isotopes, primarily U-235 and Pu-239. The composition and the distribution of the waste components within the storage environment is highly complex and not subject to easy investigation. An important safety concern is the preclusion of a self-sustaining neutron chain reaction, also known as a nuclear criticality. A thorough technical evaluation of processes, phenomena, and conditions is required to make sure that subcriticality will be ensured for both current and future tank operations. Subcriticality limits must be based on considerations of tank processes and take into account all chemical and geometrical phenomena that are occurring in the tanks. The important chemical and physical phenomena are those capable of influencing the mixing of fissile material and neutron absorbers such that the degree of subcriticality could be adversely impacted. This report describes a logical approach to resolving the criticality safety issues in the Hanford waste tanks. The approach uses a structured logic diagram (SLD) to identify the characterization needed to quantify risk. The scope of this section of the report is limited to those branches of logic needed to quantify the risk associated with a criticality event occurring. The process is linked to a conceptual model that depicts key modes of failure which are linked to the SLD. Data that are needed include adequate knowledge of the chemical and geometric form of the materials of interest. This information is used to determine how much energy the waste would release in the various domains of the tank, the toxicity of the region associated with a criticality event, and the probability of the initiating criticality event

  12. 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.

  13. 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

  14. 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

  15. Performance management in healthcare: a critical analysis.

    Science.gov (United States)

    Hewko, Sarah J; Cummings, Greta G

    2016-01-01

    Purpose - The purpose of this paper is to explore the underlying theoretical assumptions and implications of current micro-level performance management and evaluation (PME) practices, specifically within health-care organizations. PME encompasses all activities that are designed and conducted to align employee outputs with organizational goals. Design/methodology/approach - PME, in the context of healthcare, is analyzed through the lens of critical theory. Specifically, Habermas' theory of communicative action is used to highlight some of the questions that arise in looking critically at PME. To provide a richer definition of key theoretical concepts, the authors conducted a preliminary, exploratory hermeneutic semantic analysis of the key words "performance" and "management" and of the term "performance management". Findings - Analysis reveals that existing micro-level PME systems in health-care organizations have the potential to create a workforce that is compliant, dependent, technically oriented and passive, and to support health-care systems in which inequalities and power imbalances are perpetually reinforced. Practical implications - At a time when the health-care system is under increasing pressure to provide high-quality, affordable services with fewer resources, it may be wise to investigate new sector-specific ways of evaluating and managing performance. Originality/value - In this paper, written for health-care leaders and health human resource specialists, the theoretical assumptions and implications of current PME practices within health-care organizations are explored. It is hoped that readers will be inspired to support innovative PME practices within their organizations that encourage peak performance among health-care professionals.

  16. Fault tree synthesis for software design analysis of PLC based safety-critical systems

    International Nuclear Information System (INIS)

    Koo, S. R.; Cho, C. H.; Seong, P. H.

    2006-01-01

    As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

  17. Safety performance evaluation using proactive indicators in a selected industry

    Directory of Open Access Journals (Sweden)

    Abolfazl Barkhordari

    2015-03-01

    Full Text Available Background & Objectives: Quality and effectiveness of safety systems are critical factors in achieving their goals. This study was aimed to represent a method for performance evaluation of safety systems by proactive indicators using different updated models in the field of safety which will be tested in a selected industry. Methods: This study is a cross-sectional study. Proactive indicators used in this study were: Unsafe acts rate, Safety Climate, Accident Proneness, and Near-miss incident rate. The number of in 1473 safety climate questionnaires and 543 Accident Proneness questionnaires was completed. Results: The minimum and maximum safety climate score were 56.88 and 58.2, respectively, and the minimum and maximum scores of Accident Proneness were 98.2 and 140.7, respectively. The maximum number of Near-miss incident rate were 408 and the minimum of that was 196. The maximum number of unsafe acts rate was 43.8 percent and the minimum of that was 27.2 percent. In nine dimensions of Safety climate the eighth dimension (personal perception of risk with the score of 4.07 has the lowest score and the fourth (laws and safety regulations dimension with 8.05 has the highest score. According to expert opinions, the most important indicator in the assessment of safety performance was unsafe acts rate, while near-miss incident rate was the least important one. Conclusion: The results of this survey reveal that using proactive (Prospective indicators could be an appropriate method in organizations safety performance evaluation.

  18. Criticality safety for deactivation of the Rover dry headend process

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1995-01-01

    The Rover dry headend process combusted Rover graphite fuels in preparation for dissolution and solvent extraction for the recovery of 235 U. At the end of the Rover processing campaign, significant quantities of 235 U were left in the dry system. The Rover Dry Headend Process Deactivation Project goal is to remove the remaining uranium bearing material (UBM) from the dry system and then decontaminate the cells. Criticality safety issues associated with the Rover Deactivation Project have been influenced by project design refinement and schedule acceleration initiatives. The uranium ash composition used for calculations must envelope a wide range of material compositions, and yet result in cost effective final packaging and storage. Innovative thinking must be used to provide a timely safety authorization basis while the project design continues to be refined

  19. Criticality safety analysis of the NPP Krsko storage racks

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2002-01-01

    NPP Krsko is going to increase the capacity of the spent fuel storage pool by replacement of the existing racks with high-density racks. This will be the second reracking campaign since 1983 when storage was increased from 180 to 828 storage locations. The pool capacity will increase from 828 to 1694 with partial reracking by the spring 2003. The installed capacity will be sufficient for the current design plant lifetime. Complete reracking of the spent fuel pool will additionally increase capacity to 2321 storage locations. The design, rack manufacturing and installation has been awarded to the Framatome ANP GmbH. Burnup credit methodology, which was approved by the Slovenian Nuclear Safety Administration in previous licensing of existing racks, will be again implemented in the licensing process with the recent methodology improvements. Specific steps of the criticality safety analysis and representative results are presented in the paper.(author)

  20. 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.)

  1. 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.)

  2. 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

  3. Criticality safety of spent fuel casks considering water inleakage

    International Nuclear Information System (INIS)

    Osgood, N.L.; Withee, C.J.; Easton, E.P.

    2004-01-01

    A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

  4. A safety-critical java technology compatibility kit

    DEFF Research Database (Denmark)

    Søndergaard, Hans; Korsholm, Stephan E.; Ravn, Anders Peter

    2014-01-01

    In order to claim conformance with a given Java Specification Request (JSR), a Java implementation has to pass all tests in an associated Technology Compatibility Kit (TCK). This paper presents development of test cases and tools for the draft Safety-Critical Java (SCJ) specification. In previous...... work we have shown how the Java Modeling Language (JML) is applied to specify conformance constraints for SCJ, and how JML-related tools may assist in generating and executing tests. Here we extend this work with a layout for concrete test cases including checking of results in a simplified version...

  5. Safety-critical Java on a Java processor

    DEFF Research Database (Denmark)

    Schoeberl, Martin; Rios Rivas, Juan Ricardo

    2012-01-01

    The safety-critical Java (SCJ) specification is developed within the Java Community Process under specification request number JSR 302. The specification is available as public draft, but details are still discussed by the expert group. In this stage of the specification we need prototype...... implementations of SCJ and first test applications that are written with SCJ, even when the specification is not finalized. The feedback from those prototype implementations is needed for final decisions. To help the SCJ expert group, a prototype implementation of SCJ on top of the Java optimized processor...

  6. Safety-critical Java for low-end embedded platforms

    DEFF Research Database (Denmark)

    Søndergaard, Hans; Korsholm, Stephan E.; Ravn, Anders Peter

    2012-01-01

    We present an implementation of the Safety-Critical Java profile (SCJ), targeted for low-end embedded platforms with as little as 16 kB RAM and 256 kB flash. The distinctive features of the implementation are a combination of a lean Java virtual machine (HVM), with a bare metal kernel implementing...... hardware objects, first level interrupt handlers, and native variables, and an infrastructure written in Java which is minimized through program specialization. The HVM allows the implementation to be easily ported to embedded platforms which have a C compiler as part of the development environment...

  7. Patterns for Safety-Critical Java Memory Usage

    DEFF Research Database (Denmark)

    Rios Rivas, Juan Ricardo; Nilsen, Kelvin; Schoeberl, Martin

    2012-01-01

    Scoped memories are introduced in real-time Java profiles in order to make object allocation and deallocation time and space predictable. However, explicit scoping requires care from programmers when dealing with temporary objects, passing scope-allocated objects as arguments to methods, and retu......Scoped memories are introduced in real-time Java profiles in order to make object allocation and deallocation time and space predictable. However, explicit scoping requires care from programmers when dealing with temporary objects, passing scope-allocated objects as arguments to methods...... are illustrated by implementations in the safety-critical Java profile....

  8. 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.)

  9. Instructional games and activities for criticality safety training

    International Nuclear Information System (INIS)

    Bullard, B.; McBride, J.

    1993-01-01

    During the past several years, the Training and Management Systems Division (TMSD) staff of Oak Ridge Institute for Science and Education (ORISE) has designed and developed nuclear criticality safety (NCS) training programs that focus on high trainee involvement through the use of instructional games and activities. This paper discusses the instructional game, initial considerations for developing games, advantages and limitations of games, and how games may be used in developing and implementing NCS training. It also provides examples of the various instructional games and activities used in separate courses designed for Martin Marietta Energy Systems (MMES's) supervisors and U.S. Nuclear Regulatory Commission (NRC) fuel facility inspectors

  10. Criticality safety of high-level tank waste

    International Nuclear Information System (INIS)

    Rogers, C.A.

    1995-01-01

    Radioactive waste containing low concentrations of fissile isotopes is stored in underground storage tanks on the Hanford Site in Washington State. The goal of criticality safety is to ensure that this waste remains subcritical into the indefinite future without supervision. A large ratio of solids to plutonium provides an effective way of ensuring a low plutonium concentration. Since the first waste discharge, a program of audits and appraisals has ensured that operations are conducted according to limits and controls applied to them. In addition, a program of surveillance and characterization maintains watch over waste after discharge

  11. Reliability assessment for safety critical systems by statistical random testing

    International Nuclear Information System (INIS)

    Mills, S.E.

    1995-11-01

    In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs

  12. Reliability assessment for safety critical systems by statistical random testing

    Energy Technology Data Exchange (ETDEWEB)

    Mills, S E [Carleton Univ., Ottawa, ON (Canada). Statistical Consulting Centre

    1995-11-01

    In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs.

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

    International Nuclear Information System (INIS)

    Richard, R.F.

    1997-01-01

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

  14. Criticality safety aspects of K-25 Building uranium deposit removal

    International Nuclear Information System (INIS)

    Haire, M.J.; Jordan, W.C.; Ingram, J.C. III; Stinnet, E.C. Jr.

    1995-01-01

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate 235 U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report

  15. Criticality safety aspects of K-25 Building uranium deposit removal

    Energy Technology Data Exchange (ETDEWEB)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Ingram, J.C. III; Stinnet, E.C. Jr. [Oak Ridge K-25 Site, TN (United States)

    1995-12-31

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate {sup 235}U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report.

  16. 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

  17. University of New Mexico short course in nuclear criticality safety: Training for new NCS [nuclear criticality safety] specialists

    International Nuclear Information System (INIS)

    Busch, R.D.

    1990-01-01

    Since 1973, the University of New Mexico (UNM) has given ten short courses in nuclear criticality safety (NCS). Generally, thee have been given every other year, although in 1989 it was decided to offer the course on an annual basis. This decision was primarily based on the large demand for NCS specialists and a large turnover rate in the industry. The purpose of the course is to provide a 1-week overview of NCS. The typical student has been involved in NCS for <1 yr, although it many cases they have been associated with the nuclear industry in other capacities for many years. The short course is conducted at several levels. Carefully prepared lectures provide the information framework for selected topics. The following topics are covered in the course: basic reactor theory, criticality accidents and consequences, hand calculations, administration of a criticality safety program, regulators and their processes, computer methods and applications, experimental methods and correlations, overview of some process operations, and transportation and storage issues in NCS

  18. WSRC approach to validation of criticality safety computer codes

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. Study on criticality safety evaluation of a system where flood will never occur

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Yamamoto, Toshihiro; Komuro, Yuichi; Itahara, Kuniyuki.

    1995-03-01

    Criticality safety evaluation for a single unit containing nuclear fuel has usually been performed on the assumption that there is a fully thick water reflector around the unit. For a system where flood will never occur, however, the thick reflector assumption is usually not applied recently. In such cases, a method is proposed, which models surrounding structural material and branch pipes as 2.5cm thick water reflector. This report shows that reactivity worth of structural material and branch pipes is, in many cases, less than that of 2.5cm thick water reflector. Further, another method is shown to evaluate criticality safety for a multiple unit system, using computed results with surrounding structural material and branch pipes neglected. And it is shown with many sample calculations that the method with 2.5cm thick water reflector in place of structural material and pipes gives safety side results to similar systems to real reprocessing plants. (author)

  20. 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)

  1. Diversity requirements for safety critical software-based automation systems

    International Nuclear Information System (INIS)

    Korhonen, J.; Pulkkinen, U.; Haapanen, P.

    1998-03-01

    System vendors nowadays propose software-based systems even for the most critical safety functions in nuclear power plants. Due to the nature and mechanisms of influence of software faults new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)' various safety assessment methods and tools for software based systems are developed and evaluated. This report first discusses the (common cause) failure mechanisms in software-based systems, then defines fault-tolerant system architectures to avoid common cause failures, then studies the various alternatives to apply diversity and their influence on system reliability. Finally, a method for the assessment of diversity is described. Other recently published reports in OHA-report series handles the statistical reliability assessment of software based (STUK-YTO-TR 119), usage models in reliability assessment of software-based systems (STUK-YTO-TR 128) and handling of programmable automation in plant PSA-studies (STUK-YTO-TR 129)

  2. Criticality safety evaluation report for K Basin filter cartridges

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1995-01-01

    A criticality safety evaluation of the K Basin filter cartridge assemblies has been completed to support operations without a criticality alarm system. The results show that for normal operation, the filter cartridge assembly is far below the safety limit of k eff = 0.95, which is applied to plutonium systems at the Hanford Site. During normal operating conditions, uranium, plutonium, and fission and corrosion products in solution are continually accumulating in the available void spaces inside the filter cartridge medium. Currently, filter cartridge assemblies are scheduled to be replaced at six month intervals in KE Basin, and at one year intervals in KW Basin. According to available plutonium concentration data for KE Basin and data for the U/Pu ratio, it will take many times the six-month replacement time for sufficient fissionable material accumulation to take place to exceed the safety limit of k eff = 0.95, especially given the conservative assumption that the presence of fission and corrosion products is ignored. Accumulation of sludge with a composition typical of that measured in the sand filter backwash pit will not lead to a k eff = 0.95 value. For off-normal scenarios, it would require at least two unlikely, independent, and concurrent events to take place before the k eff = 0.95 limit was exceeded. Contingencies considered include failure to replace the filter cartridge assemblies at the scheduled time resulting in additional buildup of fissionable material, the loss of geometry control from the filter cartridge assembly breaking apart and releasing the individual filter cartridges into an optimal configuration, and concentrations of plutonium at U/Pu ratios less than measured data for KE Basin, typically close to 400 according to extensive measurements in the sand filter backwash pit and plutonium production information

  3. Criticality safety margins for mixtures of fissionable materials

    International Nuclear Information System (INIS)

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

    1992-01-01

    In the determination of criticality safety margins, approximations for combinations of fissile and fissionable isotopes are sometimes used that go by names such as the rule of fractions or equivalency relations. Use of the rule of fractions to ensure criticality safety margins was discussed in an earlier paper. The purpose of this paper is to correct errors and to clarify some of the implications. Deviations of safety margins from those calculated by the rule of fractions are still noted; however, the deviations are less severe. Caution in applying such rules is still urged. In general, these approximations are based on American National Standard ANSI/ANS-8.15, Sec. 5.2. This section allows that ratios of material masses to their limits may be summed for fissile nuclides in aqueous solutions. It also allows the addition of nonfissile nuclides if an aqueous moderator is present and addresses the effects of infinite water or equivalent reflector. Water-reflected binary combinations of aqueous solutions of fissile materials, as well as binary combinations of fissile and fissionable metals, were considered. Some combinations were shown to significantly decrease the margin of subcriticality compared to the single-unit margins. In this study, it is confirmed that some combinations of metal units in an optimum geometry may significantly decrease the margin of subcriticality. For some combinations of aqueous solutions of fissile materials, the margin of subcriticality may also be reduced by very small amounts. The conclusion of Ref. 1 that analysts should be careful in applying equivalency relations for combining materials remains valid and sound advice. The ANSI/ANS standard, which allows the use of ratios of masses to their limits, applies to aqueous, fully water-reflected, single-unit solutions. Extensions to other situations should be considered with extreme care

  4. Criticality safety evaluation of the fuel cycle facility electrorefiner

    International Nuclear Information System (INIS)

    Lell, R.M.; Mariani, R.D.; Fujita, E.K.; Benedict, R.W.; Turski, R.B.

    1993-01-01

    The integral Fast Reactor (IFR) being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal cooled reactors and a closed-loop fuel cycle. Some of the primary advantages are passive safety for the reactor and resistance to diversion for the heavy metal in the fuel cycle. in addition, the IFR pyroprocess recycles all the long-lived actinide activation products for casting into new fuel pins so that they may be burned in the reactor. A key component in the Fuel Cycle Facility (FCF) recycling process is the electrorefiner (ER) in which the actinides are separated from the fission products. In the process, the metal fuel is electrochemically dissolved into a high-temperature molten salt, and electrorefined uranium or uranium/plutonium products are deposited at cathodes. This report addresses the new and innovative aspects of the criticality analysis ensuing from processing metallic fuel, rather than metal oxide fuel, and from processing the spent fuel in batch operations. in particular, the criticality analysis employed a mechanistic approach as opposed to a probabilistic one. A probabilistic approach was unsuitable because of a lack of operational experience with some of the processes, rendering the estimation of accident event risk factors difficult. The criticality analysis also incorporated the uncertainties in heavy metal content attending the process items by defining normal operations envelopes (NOES) for key process parameters. The goal was to show that reasonable process uncertainties would be demonstrably safe toward criticality for continuous batch operations provided the key process parameters stayed within their NOES. Consequently the NOEs became the point of departure for accident events in the criticality analysis

  5. Proceedings of the workshop on integral experiment covariance data for critical safety validation

    Energy Technology Data Exchange (ETDEWEB)

    Stuke, Maik (ed.)

    2016-04-15

    For some time, attempts to quantify the statistical dependencies of critical experiments and to account for them properly in validation procedures were discussed in the literature by various groups. Besides the development of suitable methods especially the quality and modeling issues of the freely available experimental data are in the focus of current discussions, carried out for example in the Expert Group on Uncertainty Analysis for Criticality Safety Assessment (UACSA) of the OECD-NEA Nuclear Science Committee. The same committee compiles and publishes also the freely available experimental data in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Most of these experiments were performed as series and might share parts of experimental setups leading to correlated results. The quality of the determination of these correlations and the underlying covariance data depend strongly on the quality of the documentation of experiments.

  6. Proceedings of the workshop on integral experiment covariance data for critical safety validation

    International Nuclear Information System (INIS)

    Stuke, Maik

    2016-04-01

    For some time, attempts to quantify the statistical dependencies of critical experiments and to account for them properly in validation procedures were discussed in the literature by various groups. Besides the development of suitable methods especially the quality and modeling issues of the freely available experimental data are in the focus of current discussions, carried out for example in the Expert Group on Uncertainty Analysis for Criticality Safety Assessment (UACSA) of the OECD-NEA Nuclear Science Committee. The same committee compiles and publishes also the freely available experimental data in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Most of these experiments were performed as series and might share parts of experimental setups leading to correlated results. The quality of the determination of these correlations and the underlying covariance data depend strongly on the quality of the documentation of experiments.

  7. A study on quantitative V and V of safety-critical software

    International Nuclear Information System (INIS)

    Eom, H. S.; Kang, H. G.; Chang, S. C.; Ha, J. J.; Son, H. S.

    2004-03-01

    Recently practical needs have required quantitative features for the software reliability for Probabilistic Safety Assessment which is one of the important methods being used in assessing the overall safety of nuclear power plant. But the conventional assessment methods of software reliability could not provide enough information for PSA of NPP, therefore current assessments of a digital system which includes safety-critical software usually exclude the software part or use arbitrary values. This paper describes a Bayesian Belief Networks based method that models the rule-based qualitative software assessment method for a practical use and can produce quantitative results for PSA. The framework was constructed by utilizing BBN that can combine the qualitative and quantitative evidence relevant to the reliability of safety-critical software and can infer a conclusion in a formal and a quantitative way. The case study was performed by applying the method for assessing the quality of software requirement specification of safety-critical software that will be embedded in reactor protection system

  8. Safety analysis report for the Hanford Critical Mass Laboratory: Supplement No. 2. Experiments with heterogeneous assemblies

    International Nuclear Information System (INIS)

    Gore, B.F.; Davenport, L.C.

    1981-04-01

    Factors affecting the safety of criticality experiments using heterogeneous assemblies are described and assessed. It is concluded that there is no substantial change in safety from experiments already being routinely performed at the Critical Mass Laboratory (CML), and that laboratory and personnel safety are adequately provided by the combination of engineered and administrative safety limits enforced at the CML. This conclusion is based on the analysis of operational controls, potential hazards, and the consequences of accidents. Contingencies considered that could affect nuclear criticality include manual changes in fuel loadings, water flooding, fire, explosion, loss of services, earthquake, windstorm, and flood. Other potential hazards considered include radiation exposure to personnel, and potential releases within the Assembly Room and outside to the environment. It is concluded that the Maximum Credible Nuclear Burst of 3 x 10 18 fissions (which served as the design basis for the CML) is valid for heterogeneous assemblies as well as homogeneous assemblies. This is based upon examination of the results of reactor destructive tests and the results of the SL-1 reactor destructive accident. The production of blast effects which might jeopardize the CML critical assembly room (of thick reinforced concrete) is not considered credible due to the extreme circumstances required to produce blast effects in reactor destructive tests. Consequently, it is concluded that, for experiments with heterogeneous assemblies, the consequences of the Maximum Credible Burst are unchanged from those previously estimated for experiments with homogeneous systems

  9. Performance and safety of hydraulic turbines

    International Nuclear Information System (INIS)

    Brekke, H

    2010-01-01

    The first part of the paper contains the choice of small turbines for run of the river power plants. Then a discussion is given on the optimization of the performance of different types of large turbines. Finally a discussion on the safety and necessary maintenance of turbines is given with special attention to bolt connections.

  10. Criticality safety of low-density storage arrays

    International Nuclear Information System (INIS)

    Bauer, T.H.

    1996-01-01

    This note proposes a straightforward and simple method for the criticality safety analysis of fissionable materials configured into large arrays of standard containers. While criticality-safe storage limits have been well-established for standard containers--even under flooded conditions, it is also necessary to rule out the potential for criticality arising from neutronic interactions among multiple containers that might build up over long distances in a large array. Traditionally, the array problem has been approached by individual Monte Carlo analyses of explicit arrangements of single units and their surroundings. Here, the authors show how multiple Monte Carlo analyses can be usefully combined for wide-ranging general application. The technique takes advantage of low average density of fissionable material in typical storage arrays to separate neutron interactions that take place in the neutron's ''birth unit'' from subsequent interactions in a highly dilute array. Effects of array size, in particular, are conservatively calculated by straightforward analyses which simply smear array contents uniformly across the extent of the array. For given unit loadings in standard containers, practical expressions for neutron multiplication depend only on overall array shape, size and reflective boundary

  11. Licensing process for safety-critical software-based systems

    Energy Technology Data Exchange (ETDEWEB)

    Haapanen, P. [VTT Automation, Espoo (Finland); Korhonen, J. [VTT Electronics, Espoo (Finland); Pulkkinen, U. [VTT Automation, Espoo (Finland)

    2000-12-01

    System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications

  12. Licensing process for safety-critical software-based systems

    International Nuclear Information System (INIS)

    Haapanen, P.; Korhonen, J.; Pulkkinen, U.

    2000-12-01

    System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications. Many of the

  13. 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

  14. Nuclear criticality safety practices in digestion systems of the large scale production facility of the Department of Energy at Fernald

    International Nuclear Information System (INIS)

    Dolan, L.C.

    1982-01-01

    Nuclear criticality safety practices used at the Feed Materials Production Center at Fernald, Ohio in conjunction with its metal dissolving and nonmetal, e.g., ash and ore concentrates, digesting operations are reviewed. Operating procedures with several different types of dissolver or digestor systems, i.e., metal dissolver, continuous, drum and safe geometry, are discussed. Calculations performed to verify the criticality safety of the operations are described

  15. 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

  16. Food safety performance indicators to benchmark food safety output of food safety management systems.

    Science.gov (United States)

    Jacxsens, L; Uyttendaele, M; Devlieghere, F; Rovira, J; Gomez, S Oses; Luning, P A

    2010-07-31

    There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses. Validation was conducted on the basis of an extensive microbiological assessment scheme (MAS). The assumption behind the food safety performance diagnosis is that food businesses which evaluate the performance of their food safety management system in a more structured way and according to very strict and specific criteria will have a better insight in their actual microbiological food safety performance, because food safety problems will be more systematically detected. The diagnosis can be a useful tool to have a first indication about the microbiological performance of a food safety management system present in a food business. Moreover, the diagnosis can be used in quantitative studies to get insight in the effect of interventions on sector or governmental level. Copyright 2010 Elsevier B.V. All rights reserved.

  17. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  18. 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)

  19. Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

    International Nuclear Information System (INIS)

    Briggs, J.B.

    1991-02-01

    Intraplant shipment of small quantities of plutonium and uranium at the Rocky Flats Plant (RFP) are made in one-gallon shipping containers. Criticality safety calculations have been performed to provide an analytical basis upon which handling, storage, and transportation limits on these containers are based. The calculations and results are documented in this report. This analysis was categorized as Quality Level A (according to the EG ampersand G Idaho Quality Manual) in that it is a service whose failure could cause undue risks to employees or public health and safety. It is intended to comply with NQA-1. 7 refs., 7 figs., 12 tabs

  20. SGHWR fuel performance, safety and reliability

    International Nuclear Information System (INIS)

    Pickman, D.O.; Inglis, G.H.

    1977-05-01

    The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

  1. Safety-critical Java on a time-predictable processor

    DEFF Research Database (Denmark)

    Korsholm, Stephan E.; Schoeberl, Martin; Puffitsch, Wolfgang

    2015-01-01

    For real-time systems the whole execution stack needs to be time-predictable and analyzable for the worst-case execution time (WCET). This paper presents a time-predictable platform for safety-critical Java. The platform consists of (1) the Patmos processor, which is a time-predictable processor......; (2) a C compiler for Patmos with support for WCET analysis; (3) the HVM, which is a Java-to-C compiler; (4) the HVM-SCJ implementation which supports SCJ Level 0, 1, and 2 (for both single and multicore platforms); and (5) a WCET analysis tool. We show that real-time Java programs translated to C...... and compiled to a Patmos binary can be analyzed by the AbsInt aiT WCET analysis tool. To the best of our knowledge the presented system is the second WCET analyzable real-time Java system; and the first one on top of a RISC processor....

  2. Software Reliability Issues Concerning Large and Safety Critical Software Systems

    Science.gov (United States)

    Kamel, Khaled; Brown, Barbara

    1996-01-01

    This research was undertaken to provide NASA with a survey of state-of-the-art techniques using in industrial and academia to provide safe, reliable, and maintainable software to drive large systems. Such systems must match the complexity and strict safety requirements of NASA's shuttle system. In particular, the Launch Processing System (LPS) is being considered for replacement. The LPS is responsible for monitoring and commanding the shuttle during test, repair, and launch phases. NASA built this system in the 1970's using mostly hardware techniques to provide for increased reliability, but it did so often using custom-built equipment, which has not been able to keep up with current technologies. This report surveys the major techniques used in industry and academia to ensure reliability in large and critical computer systems.

  3. Training and qualification program for nuclear criticality safety technical staff

    International Nuclear Information System (INIS)

    Taylor, R.G.; Worley, C.A.

    1996-01-01

    A training and qualification program for nuclear criticality safety technical staff personnel has been developed and implemented. The program is compliant with requirements and provides evidence that a systematic approach has been taken to indoctrinate new technical staff. Development involved task analysis to determine activities where training was necessary and the standard which must be attained to qualify. Structured mentoring is used where experienced personnel interact with candidates using checksheets to guide candidates through various steps and to provide evidence that steps have been accomplished. Credit can be taken for the previous experience of personnel by means of evaluation boards which can credit or modify checksheet steps. Considering just the wealth of business practice and site specific information a new person at a facility needs to assimilate, the program has been effective in indoctrinating new technical staff personnel and integrating them into a productive role. The program includes continuing training

  4. An evaluation of safety-critical Java on a Java processor

    OpenAIRE

    Rios Rivas, Juan Ricardo; Schoeberl, Martin

    2014-01-01

    The safety-critical Java (SCJ) specification provides a restricted set of the Java language intended for applications that require certification. In order to test the specification, implementations are emerging and the need to evaluate those implementations in a systematic way is becoming important. In this paper we evaluate our SCJ implementation which is based on the Java Optimized Processor JOP and we measure different performance and timeliness criteria relevant to hard real-time systems....

  5. Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

    International Nuclear Information System (INIS)

    Shaw, M.E.

    1991-12-01

    Criticality safety calculations have been performed to provide an analytical basis for handling, storage and transport of Rocky Flats Plant (RFP) one-gallon shipping containers. A mass limit was establish for metal (solid uranium or plutonium) and slurries (undissolved U or Pu solids in a ''mud,'' ''sludge,'' or ''slurry''). A separate volume limit was developed for plutonium solutions (liquids, either aqueous or organic, containing no visible undissolved solids)

  6. Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, M.E.

    1991-12-01

    Criticality safety calculations have been performed to provide an analytical basis for handling, storage and transport of Rocky Flats Plant (RFP) one-gallon shipping containers. A mass limit was establish for metal (solid uranium or plutonium) and slurries (undissolved U or Pu solids in a ``mud,`` ``sludge,`` or ``slurry``). A separate volume limit was developed for plutonium solutions (liquids, either aqueous or organic, containing no visible undissolved solids).

  7. Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, M.E.

    1991-12-01

    Criticality safety calculations have been performed to provide an analytical basis for handling, storage and transport of Rocky Flats Plant (RFP) one-gallon shipping containers. A mass limit was establish for metal (solid uranium or plutonium) and slurries (undissolved U or Pu solids in a mud,'' sludge,'' or slurry''). A separate volume limit was developed for plutonium solutions (liquids, either aqueous or organic, containing no visible undissolved solids).

  8. Quantification of Safety-Critical Software Test Uncertainty

    International Nuclear Information System (INIS)

    Khalaquzzaman, M.; Cho, Jaehyun; Lee, Seung Jun; Jung, Wondea

    2015-01-01

    The method, conservatively assumes that the failure probability of a software for the untested inputs is 1, and the failure probability turns in 0 for successful testing of all test cases. However, in reality the chance of failure exists due to the test uncertainty. Some studies have been carried out to identify the test attributes that affect the test quality. Cao discussed the testing effort, testing coverage, and testing environment. Management of the test uncertainties was discussed in. In this study, the test uncertainty has been considered to estimate the software failure probability because the software testing process is considered to be inherently uncertain. A reliability estimation of software is very important for a probabilistic safety analysis of a digital safety critical system of NPPs. This study focused on the estimation of the probability of a software failure that considers the uncertainty in software testing. In our study, BBN has been employed as an example model for software test uncertainty quantification. Although it can be argued that the direct expert elicitation of test uncertainty is much simpler than BBN estimation, however the BBN approach provides more insights and a basis for uncertainty estimation

  9. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  10. Student research in criticality safety at the University of Arizona

    International Nuclear Information System (INIS)

    Hetrick, D.L.

    1997-01-01

    A very brief progress report on four University of Arizona student projects is given. Improvements were made in simulations of power pulses in aqueous solutions, including the TWODANT model. TWODANT calculations were performed to investigate the effect of assembly shape on the expansion coefficient of reactivity for solutions. Preliminary calculations were made of critical heights for the Los Alamos SHEBA assembly. Calculations to support French experiments to measure temperature coefficients of dilute plutonium solutions confirmed feasibility

  11. Safety demonstration analyses on criticality for severe accident during overland transport of fresh nuclear fuel

    International Nuclear Information System (INIS)

    Takahashi, Satoshi; Okuno, Hiroshi; Yamada, Kenji; Watanabe, Kouji; Nomura, Yasushi; Miyoshi, Yoshinori

    2005-01-01

    Criticality safety analysis was performed for transport packages of uranium dioxide powder or of fresh PWR fuel involved in a severe accident during overland transportation, and as a result, sub-criticality was confirmed against impact accident conditions such as loaded by a drop from high position to a concrete or asphalt surface, and fire accident conditions such as caused by collisions with an oil tank trailer carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside an unventilated tunnel. (author)

  12. 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

  13. Criticality safety philosophy for the Sellafield MOX plant

    International Nuclear Information System (INIS)

    Edge, Jane; Gulliford, Jim

    2003-01-01

    The Sellafield MOX Plant (SMP) has been operational since 2001, blending plutonium dioxide from THORP reprocessing operations, with uranium dioxide to produce Mixed Oxide (MOX) fuel elements. In handling the quantities of fuel associated with a commercial fuel fabrication plant, it is necessary to impose criticality controls. Plutonium dioxide (PuO 2 ), uranium dioxide (UO 2 ) and recycled MOX are mixed together in batches. An Engineered Protection System (EPS) prevents the production of MOX powder in excess of 20w/o Pu(fissile)/(Pu+U), achieved through the combination of a weight-based' system and a diverse 'neutron monitoring' radiometric system. The 'neutron monitoring' component of the EPS determines the fissile enrichment of the batch of MOX powder, based on pessimistic isotopic requirements of the PuO 2 feedstock powder. Guaranteeing the maximum MOX enrichment of 20w/o Pu(fissile)/(Pu + U) at an early stage of the fuel manufacturing process enables the criticality safety assessor to demonstrate that normal operations are deterministically safe. This paper describes in detail the EPS at the front end of plant and the engineered and operational protection in downstream areas. In addition plant operational experience in producing the first fuel assemblies is discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  15. Criticality safety of transuranic storage arrays at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Boyd, W.A.; Fecteau, M.W.

    1993-01-01

    The Waste Isolation Pilot Plant (WIPP) facility is designed to store transuranic waste that will consist mainly of surface contaminate articles and sludge. The fissile material in the waste is predominantly 239 Pu. The waste is grouped into two categories: contact-handled waste, which will be stored in 55-gal steel drums or in steel boxes, and remote-handled waste, which will be stored in specially designed cylindrical steel canisters. To show that criticality safety will be acceptable, criticality analyses were performed to demonstrate that a large number of containers with limiting loadings of fissile material could be stored at the site and meet a k eff limit of 0.95. Criticality analyses based on the classic worst-case moderated plutonium sphere approach would severely limit the capacity for storage of waste at the facility. Therefore, these analyses use realistic or credible worst-case assumptions to better represent the actual storage situation without compromising the margin of safety. Numerous sensitivity studies were performed to determine the importance of various parameters on the criticality of the configuration. It was determined that the plutonium loading has the dominant effect on the system reactivity. Nearly all other reactivity variations from the sensitivity studies were found to be relatively small. The analysis shows that criticality of the contact-handled waste storage drums and boxes and the remote-handled canisters is prevented by restrictions on maximum fissile loading per container and on the size of handling/storage areas

  16. Operational safety performance indicators for nuclear power plants

    International Nuclear Information System (INIS)

    2000-05-01

    Since the late 1980s, the IAEA has been actively sponsoring work in the area of indicators to monitor nuclear power plant (NPP) operational safety performance. The early activities were mainly focused on exchanging ideas and good practices in the development and use of these indicators at nuclear power plants. Since 1995 efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator programme. The result of this work, compiled in this publication, is intended to assist NPPs in developing and implementing a monitoring programme, without overlooking the critical aspects related to operational safety performance. The framework proposed in this report was presented at two IAEA workshops on operational safety performance indicators held in Ljubljana, Slovenia, in September 1998 and at the Daya Bay NPP, Szenzhen, China, in December 1998. During these two workshops, the participants discussed and brainstormed on the indicator framework presented. These working sessions provided very useful insights and ideas which where used for the enhancement of the framework proposed. The IAEA is acknowledging the support and contribution of all the participants in these two activities. The programme development was enhanced by pilot plant studies. Four plants from different countries with different designs participated in this study with the objective of testing the applicability, usefulness and viability of this approach

  17. Collegiate Aviation Research and Education Solutions to Critical Safety Issues. UNO Aviation Monograph Series. UNOAI Report.

    Science.gov (United States)

    Bowen, Brent, Ed.

    This document contains four papers concerning collegiate aviation research and education solutions to critical safety issues. "Panel Proposal Titled Collegiate Aviation Research and Education Solutions to Critical Safety Issues for the Tim Forte Collegiate Aviation Safety Symposium" (Brent Bowen) presents proposals for panels on the…

  18. 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

  19. 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

  20. A safety-critical decision support system evaluation using situation awareness and workload measures

    International Nuclear Information System (INIS)

    Naderpour, Mohsen; Lu, Jie; Zhang, Guangquan

    2016-01-01

    To ensure the safety of operations in safety-critical systems, it is necessary to maintain operators' situation awareness (SA) at a high level. A situation awareness support system (SASS) has therefore been developed to handle uncertain situations [1]. This paper aims to systematically evaluate the enhancement of SA in SASS by applying a multi-perspective approach. The approach consists of two SA metrics, SAGAT and SART, and one workload metric, NASA-TLX. The first two metrics are used for the direct objective and subjective measurement of SA, while the third is used to estimate operator workload. The approach is applied in a safety-critical environment called residue treater, located at a chemical plant in which a poor human-system interface reduced the operator's SA and caused one of the worst accidents in US history. A counterbalanced within-subjects experiment is performed using a virtual environment interface with and without the support of SASS. The results indicate that SASS improves operators' SA, and specifically has benefits for SA levels 2 and 3. In addition, it is concluded that SASS reduces operator workload, although further investigations in different environments with a larger number of participants have been suggested. - Highlights: • The suitability of a cognitive decision support system is investigated. • An evaluation approach considering situation awareness and workload measures is proposed. • A computerized system based on the proposed approach is implemented. • The implemented system is used in a safety-critical environment.

  1. SGHWR fuel performance, safety and reliability

    International Nuclear Information System (INIS)

    Pickman, D.O.; Inglis, G.H.

    1977-01-01

    The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of performance, safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith S.G.H.W.R. that the necessary bias toward safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications and has to be paid for. By a detailed evaluation of S.G.H.W.R. fuel defects it is shown that very few defects can be shown to be related to design, rating or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that must be met to achieve the desirable reliability level. It is possible that large-scale experience with S.G.H.W.R. fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to S.G.H.W.R. fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. The philosophy and supporting experimental work programme are outlined and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to new 60 pin fuel element to be used in the commercial S.G.H.W.R.'s and how it compares in design and performance aspects with the 36 pin element that has been used to date in the Winfrith S.G.H.W.R

  2. Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Baji, C.; Vamos, G.; Toth, J.

    2001-01-01

    The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

  3. Validation of Nuclear Criticality Safety Software and 27 energy group ENDF/B-IV cross sections

    International Nuclear Information System (INIS)

    Lee, B.L. Jr.

    1994-08-01

    The validation documented in this report is based on calculations that were executed during June through August 1992, and was completed in June 1993. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Martin Marietta 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 IBM 3090 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. When the validation results are treated as a single group, there is 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated K eff > 0.9616. Based on this result the Portsmouth Nuclear Criticality Safety Department has adopted the calculational acceptance criteria that a k eff + 2σ ≤ 0.95 is safety subcritical. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25

  4. Potential safety features and safety analysis aspects for high performance light water reactor (HPLWR)

    International Nuclear Information System (INIS)

    Aksan, N.; Schulenberg, T.; Squarer, D.

    2003-01-01

    Research Activities are ongoing worldwide to develop advanced nuclear power plants with high thermal efficiency for the purpose to improve their economical competitiveness. Within the 5th Framework Programme of the European Commission, a project has been launched with the main objective to assess the technical and economical feasibility of a high efficiency LWR operating at super critical pressure conditions. Several European research institutions, industrial partners and the University of Tokyo participated and worked in this common research project. Within the aims of the development of the HPLWR is to use both passive and active safety systems for performing safety related functions in the event of transients or accidents. Consequently substantial effort has been invested in order to define the safety features of the plant in a European environment, as well as to incorporate passive safety features into the design. Throughout this process, the European Utility Requirements (EUR) and requirements known from Generation IV initiative were considered as a guideline in general terms in order to include further advanced ideas. The HPLWR general features were compared to both requirements, indicating a potential to meet these. Since, the supercritical HPLWR represents a challenge for best-estimate safety codes like RELAP5, CATHARE and TRAB due to the fact that these codes were developed for two-phase or single-phase coolant at pressures far below critical point, work on the preliminary assessment of the appropriateness of these codes have been performed for selected relevant phenomena, and application of the codes to the selected transients on the basis of defined 'reference design'. An overview on their successful upgrade to supercritical pressures and application to some plant safety analysis are provided in the paper. Further elaborations in relation to future needs are also discussed. (author)

  5. System Predicts Critical Runway Performance Parameters

    Science.gov (United States)

    Millen, Ernest W.; Person, Lee H., Jr.

    1990-01-01

    Runway-navigation-monitor (RNM) and critical-distances-process electronic equipment designed to provide pilot with timely and reliable predictive navigation information relating to takeoff, landing and runway-turnoff operations. Enables pilot to make critical decisions about runway maneuvers with high confidence during emergencies. Utilizes ground-referenced position data only to drive purely navigational monitor system independent of statuses of systems in aircraft.

  6. Natural Language Interface for Safety Certification of Safety-Critical Software

    Science.gov (United States)

    Denney, Ewen; Fischer, Bernd

    2011-01-01

    Model-based design and automated code generation are being used increasingly at NASA. The trend is to move beyond simulation and prototyping to actual flight code, particularly in the guidance, navigation, and control domain. However, there are substantial obstacles to more widespread adoption of code generators in such safety-critical domains. Since code generators are typically not qualified, there is no guarantee that their output is correct, and consequently the generated code still needs to be fully tested and certified. The AutoCert generator plug-in supports the certification of automatically generated code by formally verifying that the generated code is free of different safety violations, by constructing an independently verifiable certificate, and by explaining its analysis in a textual form suitable for code reviews.

  7. 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

  8. Safety issues in cultural heritage management and critical infrastructures management

    Science.gov (United States)

    Soldovieri, Francesco; Masini, Nicola; Alvarez de Buergo, Monica; Dumoulin, Jean

    2013-12-01

    This special issue is the fourth of its kind in Journal of Geophysics and Engineering , containing studies and applications of geophysical methodologies and sensing technologies for the knowledge, conservation and security of products of human activity ranging from civil infrastructures to built and cultural heritage. The first discussed the application of novel instrumentation, surface and airborne remote sensing techniques, as well as data processing oriented to both detection and characterization of archaeological buried remains and conservation of cultural heritage (Eppelbaum et al 2010). The second stressed the importance of an integrated and multiscale approach for the study and conservation of architectural, archaeological and artistic heritage, from SAR to GPR to imaging based diagnostic techniques (Masini and Soldovieri 2011). The third enlarged the field of analysis to civil engineering structures and infrastructures, providing an overview of the effectiveness and the limitations of single diagnostic techniques, which can be overcome through the integration of different methods and technologies and/or the use of robust and novel data processing techniques (Masini et al 2012). As a whole, the special issue put in evidence the factors that affect the choice of diagnostic strategy, such as the material, the spatial characteristics of the objects or sites, the value of the objects to be investigated (cultural or not), the aim of the investigation (knowledge, conservation, restoration) and the issues to be addressed (monitoring, decay assessment). In order to complete the overview of the application fields of sensing technologies this issue has been dedicated to monitoring of cultural heritage and critical infrastructures to address safety and security issues. Particular attention has been paid to the data processing methods of different sensing techniques, from infrared thermography through GPR to SAR. Cascini et al (2013) present the effectiveness of a

  9. Performance of dynamic safety barriers-Structuring, modelling and visualization

    OpenAIRE

    Wikdahl, Olga

    2014-01-01

    The main objective of this master thesis is to discuss performance of dynamic safety barriers. A comprehensive literature review is performed in order to get understanding what dynamic safety barrier is. Three different concepts of dynamic safety barriers based on various meanings of dynamic were derived from the literature review: - dynamic safety barriers related to motion or physical force - dynamic safety barriers as updated barriers from dynamic risk analysis - dynamic safety ...

  10. Barrier performance researches for the safety evaluation

    International Nuclear Information System (INIS)

    Niibori, Yuichi

    2004-01-01

    So far, many researches were conducted to propose a scientific evidence (a safety case) for the realization of geological disposal in Japan. In order to regulate the geological disposal system of radioactive wastes, on the other hand, we need also a holistic approach to integrate various data related for the performance evaluations of the engineered barrier system and the natural barrier system. However, the scientific bases are not sufficient to establish the safety regulation for such a natural system. For example, we often apply the specific probability density function (PDF) to the uncertainty of barrier system due to the essential heterogeneity. However, the applicability is not clear in the regulation point of view. A viewpoint to understand such an applicability of PDFs has been presented. (author)

  11. DOE spent nuclear fuel -- Nuclear criticality safety challenges and safeguards initiatives

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1994-01-01

    The field of nuclear criticality safety is confronted with growing technical challenges and the need for forward-thinking initiatives to address and resolve issues surrounding economic, safe and secure packaging, transport, interim storage, and long-term disposal of spent nuclear fuel. These challenges are reflected in multiparameter problems involving optimization of packaging designs for maximizing the density of material per package while ensuring subcriticality and safety under variable normal and hypothetical transport and storage conditions and for minimizing costs. Historic and recently revealed uncertainties in basic data used for performing nuclear subcriticality evaluations and safety analyses highlight the need to be vigilant in assessing the validity and range of applicability of calculational evaluations that represent extrapolations from ''benchmark'' data. Examples of these uncertainties are provided. Additionally, uncertainties resulting from the safeguarding of various forms of fissionable materials in transit and storage are discussed

  12. Concepts and techniques: Active electronics and computers in safety-critical accelerator operation

    International Nuclear Information System (INIS)

    Frankel, R.S.

    1995-01-01

    The Relativistic Heavy Ion Collider (RHIC) under construction at Brookhaven National Laboratory, requires an extensive Access Control System to protect personnel from Radiation, Oxygen Deficiency and Electrical hazards. In addition, the complicated nature of operation of the Collider as part of a complex of other Accelerators necessitates the use of active electronic measurement circuitry to ensure compliance with established Operational Safety Limits. Solutions were devised which permit the use of modern computer and interconnections technology for Safety-Critical applications, while preserving and enhancing, tried and proven protection methods. In addition a set of Guidelines, regarding required performance for Accelerator Safety Systems and a Handbook of design criteria and rules were developed to assist future system designers and to provide a framework for internal review and regulation

  13. Concepts and techniques: Active electronics and computers in safety-critical accelerator operation

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, R.S.

    1995-12-31

    The Relativistic Heavy Ion Collider (RHIC) under construction at Brookhaven National Laboratory, requires an extensive Access Control System to protect personnel from Radiation, Oxygen Deficiency and Electrical hazards. In addition, the complicated nature of operation of the Collider as part of a complex of other Accelerators necessitates the use of active electronic measurement circuitry to ensure compliance with established Operational Safety Limits. Solutions were devised which permit the use of modern computer and interconnections technology for Safety-Critical applications, while preserving and enhancing, tried and proven protection methods. In addition a set of Guidelines, regarding required performance for Accelerator Safety Systems and a Handbook of design criteria and rules were developed to assist future system designers and to provide a framework for internal review and regulation.

  14. CSER 96-014: criticality safety of project W-151, 241-AZ-101 retrieval system process test

    Energy Technology Data Exchange (ETDEWEB)

    Vail, T.S., Fluor Daniel Hanford

    1997-02-06

    This Criticality Safety Evaluation Report (CSER) documents a review of the criticality safety implications of a process test to be performed in tank 241-AZ-101 (101-AZ). The process test will determine the effectiveness of the retrieval system for mobilization of solids and the practicality of the system for future use in the underground storage tanks at Hanford. The scope of the CSER extends only to the testing and operation of the mixer pumps and does not include the transfer of waste from the tank. Justification is provided that a nuclear criticality is extremely unlikely, if not impossible, in this tank.

  15. A study on methodologies for assessing safety critical network's risk impact on Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lim, T. J.; Lee, H. J.; Park, S. K.; Seo, S. J.

    2006-08-01

    The objectives of this project is to investigate and study existing reliability analysis techniques for communication networks in order to develop reliability analysis models for Nuclear Power Plant's safety-critical networks. It is necessary to make a comprehensive survey of current methodologies for communication network reliability. Major outputs of the first year study are design characteristics of safety-critical communication networks, efficient algorithms for quantifying reliability of communication networks, and preliminary models for assessing reliability of safety-critical communication networks

  16. Nuclear criticality safety evaluation of Spray Booth Operations in X-705, Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Sheaffer, M.K.; Keeton, S.C.

    1993-01-01

    This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested

  17. Reliability estimation of safety-critical software-based systems using Bayesian networks

    International Nuclear Information System (INIS)

    Helminen, A.

    2001-06-01

    Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of software-based safety-critical automation systems in nuclear power plants. In the research project 'Programmable automation system safety integrity assessment (PASSI)', belonging to the Finnish Nuclear Safety Research Programme (FINNUS, 1999-2002), various safety assessment methods and tools for software based systems are developed and evaluated. The project is financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT). In this report the applicability of Bayesian networks to the reliability estimation of software-based systems is studied. The applicability is evaluated by building Bayesian network models for the systems of interest and performing simulations for these models. In the simulations hypothetical evidence is used for defining the parameter relations and for determining the ability to compensate disparate evidence in the models. Based on the experiences from modelling and simulations we are able to conclude that Bayesian networks provide a good method for the reliability estimation of software-based systems. (orig.)

  18. 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

  19. Lithium safety and tolerability in mood disorders: a critical review

    Directory of Open Access Journals (Sweden)

    Ivan Aprahamian

    2014-04-01

    Full Text Available Background : Lithium is a first-line treatment for bipolar disorder in all phases, also indicated as add-on drug for unipolar depression and suicide prevention. This study encompasses a broad critical review on the safety and tolerability of lithium for mood disorders. Methods : A computerized search for English written human studies was made in MEDLINE, using the keywords “lithium” and “mood disorders”, starting from July 1993 through July 2013 (n = 416. This initial search aimed to select clinical trials, prospective data, and controlled design studies of lithium treatment for mood disorders reporting adverse effects (n = 36. The final selection yielded 91 studies. Results : The most common general side effects in patients on lithium treatment were thirst, frequent urination, dry mouth, weight gain, fatigue and cognitive complaints. Lithium users showed a high prevalence of hypothyroidism, hyperparathyroidism, and decrease in urinary concentration ability. Reduction of glomerular filtration rate in patients using lithium was also observed, but in a lesser extent. The evidence of teratogenicity associated with lithium use is not well established. Anti-inflammatory non-steroidal drugs, thiazide diuretics, angiotensin-converting enzyme inhibitors, and alprazolam may increase serum lithium and the consequent risk for intoxication. Discussion : Short-term lithium treatment is associated with mild side effects. Medium and long-term lithium treatment, however, might have effects on target organs which may be prevented by periodical monitoring. Overall, lithium is still a safe option for the treatment of mood disorders.

  20. Critical Reflections on Conservatism in Nuclear Safety Regulation

    International Nuclear Information System (INIS)

    Choi, Young Sung; Choi, Kwang Sik

    2007-01-01

    A recent report published by the Committee on Nuclear Regulatory Activities (CNRA) of the OECD Nuclear Energy Agency (NEA) says that a fundamental principle for safety regulators is the practice of conservative decision making. Nuclear regulators frequently face challenging issues surrounded by uncertainties or lack of data and information. No matter what efforts will be made to collect the available information and to assess the issues, nobody can clear all the uncertainties and make absolutely certain decision. More often than not, the regulators have to make a decision in light of continuing uncertainties and limited information. It is at this point that the principle of conservatism should play a role. However the principle comes in many diverse forms such as default conservatism, precautionary principle, defense in depth and realistic conservatism. These different forms of conservatism have different roles and meanings that will take a decision maker to drastically different results. This paper reviews different forms of conservatism in critical way, presents analytical framework for decision making under uncertainty and suggests future research works needed

  1. Nuclear criticality safety program for environmental restoration projects

    International Nuclear Information System (INIS)

    Marble, R.C.; Brown, T.D.

    1994-05-01

    The Fernald Environmental Management Project (FEMP), formerly known as the Feed Materials Production Center (FMPC), is located on a 1050 acre site approximately twenty miles northwest of Cincinnati, Ohio. The production area of the site covers approximately 136 acres in the central portion of the site. Surrounding the core production area is a buffer consisting of leased grazing land, reforested land, and unused areas. The uranium processing facility was designed and constructed in the early 1950s. During the period from 1952 to 1989 the site produced uranium feed material and uranium products used in the United States weapons complex. Production at the site ended in 1989, when the site was shut down for what was expected to be a short period of time. However, the FUTC was permanently shut down in 1991, and the site's mission was changed from production to environmental restoration. The objective of this paper is to give an update on activities at the Fernald Site and to describe the Nuclear Criticality Safety issues that are currently being addressed

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  3. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  4. 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

  5. Safety Performance Index Industri Batik Tulis Berdasarkan Kriteria Majemuk

    Directory of Open Access Journals (Sweden)

    Nachnul Ansori

    2015-12-01

    Full Text Available Accident generally occurs due to the activities which is done in unsafe conditions or even unsafe behavior. These conditions can influence workers productivity. In batik industries, those workers use toxic material and work in non ergonomic atmosphere. Moreover, they also do not take care of the environment and do not use personal protective equipments (PPE. Workers at Madura batik SMEs have not fully realized the significance of occupational health and safety (OHS in their working areas. The aims of this research is to evaluate OHS performance based on the indicators of safety performance index (SPI, which is multicriteria, in that industri. The safety performance attributes were obtained from factor analysis from the previous study. The index is calculated based on the weighted evaluation results of critical behavior checklist (CBC, integrated with analytical hierarchy process (AHP. As the results, we should give priority to improve the knowledge and experience toward OHS of the workers and also improve the working environment of the Madura’s batik industries. Additionally, the performance of OHS in coloring process is also on the threshold of unsafe condition, further development on the coloring process is needed.

  6. Operational Safety Performance Indicators and Balanced Scorecard in HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Ahn, Guk-Hoon; Lee, Kye-Hong; Lim, In-Cheol; Kim, Hark-Rho

    2007-01-01

    Research reactors need an extensive basis for ensuring their safety. The importance of a safety management in nuclear facilities and activities has been emphasized. The safety activities in HANARO have been continuously conducted to enhance its safe operation. Last year, HANARO prepared two indicator sets to measure and assess the safety status of the reactor's operation and utilization. One is Safety Performance Indicators (SPI) and the other is Balanced Scorecard (BSC). Through reviewing these indicators, we can obtain the following information; - Plant safety status - Safety parameter trends - Safety information, for example, reactor operation status and radiation safety HANARO will continuously pursue the trends of SPI and BSC

  7. CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

    International Nuclear Information System (INIS)

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

    1982-01-01

    Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

  8. Nuclear critical safety analysis for UX-30 transport of freight package

    International Nuclear Information System (INIS)

    Quan Yanhui; Zhou Qi; Yin Shenggui

    2014-01-01

    The nuclear critical safety analysis and evaluation for UX-30 transport freight package in the natural condition and accident condition were carried out with MONK-9A code and MCNP code. Firstly, the critical benchmark experiment data of public in international were selected, and the deflection and subcritical limiting value with MONK-9A code and MCNP code in calculating same material form were validated and confirmed. Secondly, the neutron efficiency multiplication factors in the natural condition and accident condition were calculated and analyzed, and the safety in transport process was evaluated by taking conservative suppose of nuclear critical safety. The calculation results show that the max value of k eff for UX-30 transport freight package is less than the subcritical limiting value, and the UX-30 transport freight package is in the state of subcritical safety. Moreover, the critical safety index (CSI) for UX-30 package can define zero based on the definition of critical safety index. (authors)

  9. Formal model-based development for safety-critical embedded software

    International Nuclear Information System (INIS)

    Kim, Jin Hyun; Choi, Jin Young

    2005-01-01

    Safety-critical embedded software for nuclear I and C system is developed under the safety and reliability regulation. Programmable logic controller(PLC) is a computer system for instrumentation and control (I and C) system of nuclear power plants. PLC consists of various I and C logics in software, including real-time operating system (RTOS). Hence, errors related with RTOS should be detected and eliminated in development processes. Practically, the verification and validation for errors in RTOS is performed in test procedure, in which a lot of tasks for testing are embedded in RTOS and are running under a test environments. But the test process can not be enough to guarantee the safety and reliability of RTOS. Therefore, in this paper, we introduce to applying formal methods with the development of software for the PLC. We particularity apply formal methods to a development of RTOS for PLC, which is a safety critical level. In this development, we use the state charts of I-Logix to specify and verification and model checking to verify the specification

  10. Formal model-based development for safety-critical embedded software

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyun; Choi, Jin Young [Korea University, seoul (Korea, Republic of)

    2005-11-15

    Safety-critical embedded software for nuclear I and C system is developed under the safety and reliability regulation. Programmable logic controller(PLC) is a computer system for instrumentation and control (I and C) system of nuclear power plants. PLC consists of various I and C logics in software, including real-time operating system (RTOS). Hence, errors related with RTOS should be detected and eliminated in development processes. Practically, the verification and validation for errors in RTOS is performed in test procedure, in which a lot of tasks for testing are embedded in RTOS and are running under a test environments. But the test process can not be enough to guarantee the safety and reliability of RTOS. Therefore, in this paper, we introduce to applying formal methods with the development of software for the PLC. We particularity apply formal methods to a development of RTOS for PLC, which is a safety critical level. In this development, we use the state charts of I-Logix to specify and verification and model checking to verify the specification.

  11. Formal verification and validation of the safety-critical software in a digital reactor protection system

    International Nuclear Information System (INIS)

    Kwon, K. C.; Park, G. Y.

    2006-01-01

    This paper describes the Verification and Validation (V and V) activities for the safety-critical software in a Digital Reactor Protection System (DRPS) that is being developed through the Korea nuclear instrumentation and control system project. The main activities of the DRPS V and V process are a preparation of the software planning documentation, a verification of the software according to the software life cycle, a software safety analysis and a software configuration management. The verification works for the Software Requirement Specification (SRS) of the DRPS consist of a technical evaluation, a licensing suitability evaluation, a inspection and traceability analysis, a formal verification, and preparing a test plan and procedure. Especially, the SRS is specified by the formal specification method in the development phase, and the formal SRS is verified by a formal verification method. Through these activities, we believe we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the nuclear safety-critical software in a DRPS. (authors)

  12. 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.)

  13. 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

  14. Some problems of neutron source multiplication method for site measurement technology in nuclear critical safety

    International Nuclear Information System (INIS)

    Shi Yongqian; Zhu Qingfu; Hu Dingsheng; He Tao; Yao Shigui; Lin Shenghuo

    2004-01-01

    The paper gives experiment theory and experiment method of neutron source multiplication method for site measurement technology in the nuclear critical safety. The measured parameter by source multiplication method actually is a sub-critical with source neutron effective multiplication factor k s , but not the neutron effective multiplication factor k eff . The experiment research has been done on the uranium solution nuclear critical safety experiment assembly. The k s of different sub-criticality is measured by neutron source multiplication experiment method, and k eff of different sub-criticality, the reactivity coefficient of unit solution level, is first measured by period method, and then multiplied by difference of critical solution level and sub-critical solution level and obtained the reactivity of sub-critical solution level. The k eff finally can be extracted from reactivity formula. The effect on the nuclear critical safety and different between k eff and k s are discussed

  15. Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Leotlela, Mosebetsi J. [Witwatersrand Univ., Johannesburg (South Africa). School of Physics; Koeberg Operating Unit, Johannesburg (South Africa). Regulations and Licensing; Malgas, Isaac [Koeberg Nuclear Power Station, Duinefontein (South Africa). Nuclear Engineering Analysis; Taviv, Eugene [ASARA consultants (PTY) LTD, Johannesburg (South Africa)

    2015-11-15

    In nuclear criticality safety analysis it is essential to ascertain how various components of the nuclear system will perform under certain conditions they may be subjected to, particularly if the components of the system are likely to be affected by environmental factors such as temperature, radiation or material composition. It is therefore prudent that a sensitivity analysis is performed to determine and quantify the response of the output to variation in any of the input parameters. In a fissile system, the output parameter of importance is the k{sub eff}. Therefore, in attempting to prevent reactivity-induced accidents, it is important for the criticality safety analyst to have a quantified degree of response for the neutron multiplication factor to perturbation in a given input parameter. This article will present the results of the perturbation of the parameters that are important to nuclear criticality safety analysis and their respective correlation equations for deriving the sensitivity coefficients.

  16. Safety and security profiles of industry networks used in safety- critical applications

    Directory of Open Access Journals (Sweden)

    Mária FRANEKOVÁ

    2008-01-01

    Full Text Available The author describes the mechanisms of safety and security profiles of industry and communication networks used within safety – related applications in technological and information levels of process control recommended according to standards IEC 61784-3,4. Nowadays the number of vendors of the safety – related communication technologies who guarantees besides the standard communication, the communication amongst the safety – related equipment according to IEC 61508 is increasing. Also the number of safety – related products is increasing, e. g. safety Fieldbus, safety PLC, safety curtains, safety laser scanners, safety buttons, safety relays and other. According to world survey the safety Fieldbus denoted the highest growth from all manufactured safety products.The main part of this paper is the description of the safety-related Fieldbus communication system, which has to guaranty Safety Integrity Level.

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

    International Nuclear Information System (INIS)

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

    1980-05-01

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

  18. Criticality safety studies involved in actions to improve conditions for storing 'RA' research reactor spent fuel

    International Nuclear Information System (INIS)

    Matausek, M.; Marinkovic, N.

    1998-01-01

    A project has recently been initiated by the VINCA Institute of Nuclear Sciences to improve conditions in the spent fuel storage pool at the 6.5 MW research reactor RA, as well as to consider transferring this spent fuel into a new dry storage facility built for the purpose. Since quantity and contents of fissile material in the spent fuel storage at the RA reactor are such that possibility of criticality accident can not be a priori excluded, according to standards and regulations for handling fissile material outside a reactor, before any action is undertaken subcriticality should be proven under normal, as well as under credible abnormal conditions. To perform this task, comprehensive nuclear criticality safety studies had to be performed. (author)

  19. 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

  20. Performance and Job Satisfaction - A Critical Review

    Directory of Open Access Journals (Sweden)

    Devaki Devi Puvada

    2012-04-01

    Full Text Available The human aspect in the present context has acquired immense importance. The needbased motivational theories relate need and psychological gratification to motivation and jobsatisfaction; they consider performance as the endresult. However, Lawler and Porter postulatereversely that job performance leads to job satisfaction. The study focuses on testing the degree anddirection of the relationship between Performance and Job Satisfaction with intervening variablessuch as job relations, commitment, role conflict, value system, motivation and organizational climateand with socio-economic variables. The study is conducted on 928 employees drawn from 13 publicsector and 5 private sector organizations using simple random sampling and males as the matchingsample in the State of Andhra Pradesh, India. It isevident from the study that the performance levelof the employees is significantly lesser than theirjob satisfaction level. The analysis leads to state thatall those performing well are satisfied and all those satisfied do not perform well indicating thatperformance leads to job satisfaction, job satisfaction does not lead necessarily to performance andthe relationship is intertwined. This empirical evidence supports the theory of Lawler and Porter andsets direction for future studies at micro level.

  1. 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...

  2. Nuclear data needs within the U. S. Nuclear Criticality Safety program

    International Nuclear Information System (INIS)

    McKnight, R.D.; Dunn, M.E.; Little, R.C.; Felty, J.R.; McKamy, J.N.

    2008-01-01

    This paper will present the nuclear data needs currently identified within the US Nuclear Criticality Safety Program (NCSP). It will identify the priority data needs; it will describe the process of prioritizing those needs; and it will provide brief examples of recent data advances which have successfully addressed some of the priority criticality safety data needs.

  3. Tank waste remediation system nuclear criticality safety inspection and assessment plan

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This plan provides a management approved procedure for inspections and assessments of sufficient depth to validate that the Tank Waste Remediation System (TWRS) facility complies with the requirements of the Project Hanford criticality safety program, NHF-PRO-334, ''Criticality Safety General, Requirements''

  4. 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...

  5. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    International Nuclear Information System (INIS)

    Montierth, Leland M.

    2016-01-01

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  6. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  7. Design of safety-critical systems using the complementarities of success and failure domains with a case study

    International Nuclear Information System (INIS)

    Ahmed, Rizwan; Koo, June Mo; Jeong, Yong Hoon; Heo, Gyunyoung

    2011-01-01

    A safety-critical system has to qualify the performance-related requirements and the safety-related requirements simultaneously. Conceptually, design processes should consider both of them simultaneously but the practices do not and/or cannot follow such a theoretical approach due to the limitation of design resources. From our experience, we found that safety-related functions must be simultaneously resolved with the development of performance-related functions, particularly, in case of safety-critical systems. Since, success and failure domain analyses are essential for the investigation of performance-related and safety-related requirements, respectively, we articulated our perception to Axiomatic Design (AD), Fault Tree Analysis (FTA), and TRIZ. A design evolution procedure considering feedbacks from AD to identify functional couplings, TRIZ methodology to explore uncoupling solutions and FTA to improve reliability in a systematic way is presented here. A case study regarding design of safety injection tank installed in a nuclear power plant is also included to illustrate the proposed framework. It is expected that several iterations between AD-TRIZ-FTA would result into an optimized design which could be tested against the desired performance and safety criteria.

  8. Design of safety-critical systems using the complementarities of success and failure domains with a case study

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Rizwan; Koo, June Mo [Department of Nuclear Engineering, Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of); Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.k [Department of Nuclear Engineering, Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2011-01-15

    A safety-critical system has to qualify the performance-related requirements and the safety-related requirements simultaneously. Conceptually, design processes should consider both of them simultaneously but the practices do not and/or cannot follow such a theoretical approach due to the limitation of design resources. From our experience, we found that safety-related functions must be simultaneously resolved with the development of performance-related functions, particularly, in case of safety-critical systems. Since, success and failure domain analyses are essential for the investigation of performance-related and safety-related requirements, respectively, we articulated our perception to Axiomatic Design (AD), Fault Tree Analysis (FTA), and TRIZ. A design evolution procedure considering feedbacks from AD to identify functional couplings, TRIZ methodology to explore uncoupling solutions and FTA to improve reliability in a systematic way is presented here. A case study regarding design of safety injection tank installed in a nuclear power plant is also included to illustrate the proposed framework. It is expected that several iterations between AD-TRIZ-FTA would result into an optimized design which could be tested against the desired performance and safety criteria.

  9. Safety implications of anomalous effects of neutron absorbers on criticality

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1987-04-01

    A number of ''anomalies'' in nuclear criticality have been disclosed in recent years, and as new data have become available additional anomalies have come to light. Application of existing data, without familiarity with the anomalies could lead to diminished criticality control, or more costly less efficient control. As neutron absobers are frequently used for criticality control, this paper briefly presents and discusses six apparent anomalies pertaining to the effect of neutron absorbers on the criticality of fissionable material

  10. Critical incidents related to cardiac arrests reported to the Danish Patient Safety Database

    DEFF Research Database (Denmark)

    Andersen, Peter Oluf; Maaløe, Rikke; Andersen, Henning Boje

    2010-01-01

    Background Critical incident reports can identify areas for improvement in resuscitation practice. The Danish Patient Safety Database is a mandatory reporting system and receives critical incident reports submitted by hospital personnel. The aim of this study is to identify, analyse and categorize...... critical incidents related to cardiac arrests reported to the Danish Patient Safety Database. Methods The search terms “cardiac arrest” and “resuscitation” were used to identify reports in the Danish Patient Safety Database. Identified critical incidents were then classified into categories. Results One...

  11. An evaluation of safety-critical Java on a Java processor

    DEFF Research Database (Denmark)

    Rios Rivas, Juan Ricardo; Schoeberl, Martin

    2014-01-01

    The safety-critical Java (SCJ) specification provides a restricted set of the Java language intended for applications that require certification. In order to test the specification, implementations are emerging and the need to evaluate those implementations in a systematic way is becoming important....... In this paper we evaluate our SCJ implementation which is based on the Java Optimized Processor JOP and we measure different performance and timeliness criteria relevant to hard real-time systems. Our implementation targets Level 0 and Level1 of the specification and to test it we use a series of micro...

  12. Deconstructing Global Markets through Critical Performative Experiences in Puerto Rico

    Science.gov (United States)

    Medina, Carmen Liliana; Weltsek, Gustave J.

    2013-01-01

    Critical Performative Pedagogies, the idea that "The nature of drama as a once removed creative experience turns non-critical implicit classroom identity formation into explicit identity performance as it asks participants to actively reflect upon how identity is created and engaged within fictional social interactions." (Weltsek and…

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

    Energy Technology Data Exchange (ETDEWEB)

    Craft, A.E., E-mail: aaron.craft@inl.gov [Center for Space Nuclear Research (CSNR), INL, Idaho Falls, ID (United States); O’Brien, R.C., E-mail: Robert.OBrien@inl.gov [Center for Space Nuclear Research (CSNR), INL, Idaho Falls, ID (United States); Howe, S.D., E-mail: Steven.Howe@inl.gov [Center for Space Nuclear Research (CSNR), INL, Idaho Falls, ID (United States); King, J.C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Metallurgical and Materials Engineering Department, Colorado School of Mines, Golden, CO 80401 (United States)

    2014-07-01

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

  14. American National Standard administrative practices for nuclear criticality safety, ANSI/ANS-8.19

    International Nuclear Information System (INIS)

    Smith, D.R.; Carson, R.W.

    1991-01-01

    American National Standard Administrative Practices for Nuclear Criticality Safety, ANSI/ANS-8.19, provides guidance for the administration of an effective program to control the risk of nuclear criticality in operations with fissile material outside reactors. The several sections of the standard address the responsibilities of management, supervisory personnel, and the criticality safety staff, as well as requirements and suggestions for the content of operating procedures, process evaluations, material control procedures, and emergency procedures

  15. Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks

    Energy Technology Data Exchange (ETDEWEB)

    ROGERS, C.A.

    2000-02-17

    This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

  16. Criticality Safety Evaluation of Hanford Site High-Level Waste Storage Tanks

    International Nuclear Information System (INIS)

    ROGERS, C.A.

    2000-01-01

    This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions

  17. 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

  18. An aspect-oriented approach for designing safety-critical systems

    Science.gov (United States)

    Petrov, Z.; Zaykov, P. G.; Cardoso, J. P.; Coutinho, J. G. F.; Diniz, P. C.; Luk, W.

    The development of avionics systems is typically a tedious and cumbersome process. In addition to the required functions, developers must consider various and often conflicting non-functional requirements such as safety, performance, and energy efficiency. Certainly, an integrated approach with a seamless design flow that is capable of requirements modelling and supporting refinement down to an actual implementation in a traceable way, may lead to a significant acceleration of development cycles. This paper presents an aspect-oriented approach supported by a tool chain that deals with functional and non-functional requirements in an integrated manner. It also discusses how the approach can be applied to development of safety-critical systems and provides experimental results.

  19. 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.)

  20. Dynamic modeling of the tradeoff between productivity and safety in critical engineering systems

    International Nuclear Information System (INIS)

    Cowing, Michelle M.; Elisabeth Pate-Cornell, M.; Glynn, Peter W.

    2004-01-01

    Short-term tradeoffs between productivity and safety often exist in the operation of critical facilities such as nuclear power plants, offshore oil platforms, or simply individual cars. For example, interruption of operations for maintenance on demand can decrease short-term productivity but may be needed to ensure safety. Operations are interrupted for several reasons: scheduled maintenance, maintenance on demand, response to warnings, subsystem failure, or a catastrophic accident. The choice of operational procedures (e.g. timing and extent of scheduled maintenance) generally affects the probabilities of both production interruptions and catastrophic failures. In this paper, we present and illustrate a dynamic probabilistic model designed to describe the long-term evolution of such a system through the different phases of operation, shutdown, and possibly accident. The model's parameters represent explicitly the effects of different components' performance on the system's safety and reliability through an engineering probabilistic risk assessment (PRA). In addition to PRA, a Markov model is used to track the evolution of the system and its components through different performance phases. The model parameters are then linked to different operations strategies, to allow computation of the effects of each management strategy on the system's long-term productivity and safety. Decision analysis is then used to support the management of the short-term trade-offs between productivity and safety in order to maximize long-term performance. The value function is that of plant managers, within the constraints set by local utility commissions and national (e.g. energy) agencies. This model is illustrated by the case of outages (planned and unplanned) in nuclear power plants to show how it can be used to guide policy decisions regarding outage frequency and plant lifetime, and more specifically, the choice of a reactor tripping policy as a function of the state of the

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

    International Nuclear Information System (INIS)

    Marshall, William J.; Wagner, John C.

    2012-09-01

    This report documents work performed for the Department of Energy's Office of Nuclear Energy (DOENE) Fuel Cycle Technologies Used Fuel Disposition Campaign to assess the impact of fuel reconfiguration due to fuel failure on the criticality safety of used nuclear fuel (UNF) in storage and transportation casks. This work was motivated by concerns related to the potential for fuel degradation during extended storage (ES) periods and transportation following ES, but has relevance to other potential causes of fuel reconfiguration. Commercial UNF in the United States is expected to remain in storage for longer periods than originally intended. Extended storage time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications for virtually all aspects of a UNF storage and transport system's performance. The potential impact of fuel reconfiguration on the safety of UNF in storage and transportation is dependent on the likelihood and extent of the fuel reconfiguration, which is not well understood and is currently an active area of research. The objective of this work is to assess and quantify the impact of postulated failed fuel configurations on the criticality safety of UNF in storage and transportation casks. Although this work is motivated by the potential for fuel degradation during ES periods and transportation following ES, it has relevance to fuel reconfiguration due to the effects of high burnup. Regardless of the ultimate disposition path, UNF will need to be transported at some point in the future. To investigate and quantify the impact of fuel reconfiguration on criticality safety limits, which are given in terms of the effective neutron multiplication factor, a set of failed fuel

  2. Nuclear criticality safety analysis of a spent fuel waste package in a tuff repository

    International Nuclear Information System (INIS)

    Weren, B.H.; Capo, M.A.; O'Neal, W.C.

    1983-12-01

    An assessment has been performed of the criticality potential associated with the disposal of spent fuel in a tuff geology above the water table. Eleven potential configurations were defined which cover a vast range of geometries and conditions from the nominal configuration at emplacement to a hypothetical configuration thousands of years after emplacement in which the structure is gone, the fuel pellets disintegrated and the borehole flooded. Of these eleven configurations, four have been evaluated at this time. The results of this evaluation indicate that even with very conservative assumptions (4.5 w/o fresh fuel), criticality is not a problem for the nominal configuration either dry or fully flooded. In the cases where the condition of the waste package is assumed to have severely deteriorated, over long times, calculations were performed with less conservative assumptions (depleted fuel). An assessment of these calculations indicates that criticality safety could be demonstrated if the depletion of the fissile inventory during fuel irradiation is taken into account. A detailed discussion of the calculations performed is presented in this report. Also included are a description of the configurations which were considered, the analytical methods and models used, and a discussion of additional related work which should be performed. 15 references, 11 figures, 8 tables

  3. Critical review of glass performance modeling

    International Nuclear Information System (INIS)

    Bourcier, W.L.

    1994-07-01

    Borosilicate glass is to be used for permanent disposal of high-level nuclear waste in a geologic repository. Mechanistic chemical models are used to predict the rate at which radionuclides will be released from the glass under repository conditions. The most successful and useful of these models link reaction path geochemical modeling programs with a glass dissolution rate law that is consistent with transition state theory. These models have been used to simulate several types of short-term laboratory tests of glass dissolution and to predict the long-term performance of the glass in a repository. Although mechanistically based, the current models are limited by a lack of unambiguous experimental support for some of their assumptions. The most severe problem of this type is the lack of an existing validated mechanism that controls long-term glass dissolution rates. Current models can be improved by performing carefully designed experiments and using the experimental results to validate the rate-controlling mechanisms implicit in the models. These models should be supported with long-term experiments to be used for model validation. The mechanistic basis of the models should be explored by using modern molecular simulations such as molecular orbital and molecular dynamics to investigate both the glass structure and its dissolution process

  4. Critical review of glass performance modeling

    Energy Technology Data Exchange (ETDEWEB)

    Bourcier, W.L. [Lawrence Livermore National Lab., CA (United States)

    1994-07-01

    Borosilicate glass is to be used for permanent disposal of high-level nuclear waste in a geologic repository. Mechanistic chemical models are used to predict the rate at which radionuclides will be released from the glass under repository conditions. The most successful and useful of these models link reaction path geochemical modeling programs with a glass dissolution rate law that is consistent with transition state theory. These models have been used to simulate several types of short-term laboratory tests of glass dissolution and to predict the long-term performance of the glass in a repository. Although mechanistically based, the current models are limited by a lack of unambiguous experimental support for some of their assumptions. The most severe problem of this type is the lack of an existing validated mechanism that controls long-term glass dissolution rates. Current models can be improved by performing carefully designed experiments and using the experimental results to validate the rate-controlling mechanisms implicit in the models. These models should be supported with long-term experiments to be used for model validation. The mechanistic basis of the models should be explored by using modern molecular simulations such as molecular orbital and molecular dynamics to investigate both the glass structure and its dissolution process.

  5. Criticality safety analysis of accelerator transmutation waste system

    International Nuclear Information System (INIS)

    Landeyro, P.A.; Cepraga, D.G.; Orazi, A.

    1993-01-01

    The Accelerator Transmutation Waste system (ATW) is under development at the Los Alamos National Laboratory. It consists of a particle accelerator producing a proton beam having an energy of 1.5 GeV. These particles are introduced into the upper part of a molten Pb-Bi column and they produce, by a spallation reaction, a high strength neutron flux, 1.0x10 16 n/(square centimeters sec). The neutrons enter a heavy water blanket where actinides and long-lived fission products circulate in vertical tubes. The goal of this research effort is to perform an independent verification of the feasibility of actinide burning in the ATW system. The work is divided into four tasks: a) production of an actinide and long-lived fission product cross section library from JEF 2.2; b) simulation, using MCNP and KENO IV Monte Carlo codes, of the ATW configurations existing in literature; c) validation of the cross sections by comparison of Keff and reaction rate results, calculated with MCNP and KENO IV, with experimental benchmarks and intercomparison between calculations of a PWR unit cell and the computations carried out with various codes and cross section libraries (NEACRF criticality working group data); d) simulation of the ATW configuration. The two first tasks are almost complete with excellent agreement between this study's results and those of Los Alamos

  6. GT-MHR design, performance, and safety

    International Nuclear Information System (INIS)

    Neylan, A.J.; Shenoy, A.; Silady, F.A.; Dunn, T.D.

    1994-11-01

    The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a low power density passively safe modular reactor with key technology developments in the U.S. during the last decade: large industrial gas turbines; large active magnetic bearings; and compact, highly effective plate-fin heat exchangers. This is accomplished through the unique use of the Brayton cycle to produce electricity with the helium as primary coolant from the reactor directly driving the gas turbine electrical generator. This cycle can achieve a high net efficiency in the range of 45% to 48%. In the design of the GT-MHR the desirable inherent characteristics of the inert helium coolant, graphite core, and the coated fuel particles are supplemented with specific design features such as passive heat removal to achieve the safety objective of not disturbing the normal day-to-day activities of the public even for beyond design basis rare accidents. Each GT-MHR plant consists of four modules. The GT-MHR module components are contained within steel pressure vessels: a reactor vessel, a power conversion vessel, and a connecting cross vessel. All vessels are sited underground in a concrete silo, which serves as an independent vented low pressure containment structure. By capitalizing on industrial and aerospace gas turbine development, highly effective heat exchanger designs, and inherent gas cooled reactor temperature characteristics, the passively safe GT-MHR provides a sound technical, monetary, and environmental basis for new nuclear power generating capacity. This paper provides an update on the status of the design, which has been under development on the US-DOE program since February 1993. An assessment of plant performance and safety is also included

  7. Safety Injection Tank Performance Analysis Using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Oan; Lee, Jeong Ik; Nietiadi Yohanes Setiawan [KAIST, Daejeon (Korea, Republic of); Addad Yacine [KUSTAR, Abu Dhabi (United Arab Emirates); Bang, Young Seok; Yoo, Seung Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    This may affect the core cooling capability and threaten the fuel integrity during LOCA situations. However, information on the nitrogen flow rate during discharge is very limited due to the associated experimental measurement difficulties, and these phenomena are hardly reflected in current 1D system codes. In the current study, a CFD analysis is presented which hopefully should allow obtaining a more realistic prediction of the SIT performance which can then be reflected on 1D system codes to simulate various accident scenarios. Current Computational Fluid Dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study aims to find a better CFD prediction and more accurate modeling to predict the system performance during accident scenarios. The safety injection tank with fluidic device was analyzed using commercial CFD. A fine resolution grid was used to capture the vortex of the fluidic device. The calculation so far has shown good consistency with the experiment. Calculation should complete by the conference date and will be thoroughly analyzed to be discussed. Once a detailed CFD computation is finished, a small-scale experiment will be conducted for the given conditions. Using the experimental results and the CFD model, physical models can be validated to give more reliable results. The data from CFD and experiments will provide a more accurate K-factor of the fluidic device which can later be applied in system code inputs.

  8. Analyzing Software Errors in Safety-Critical Embedded Systems

    Science.gov (United States)

    Lutz, Robyn R.

    1994-01-01

    This paper analyzes the root causes of safty-related software faults identified as potentially hazardous to the system are distributed somewhat differently over the set of possible error causes than non-safety-related software faults.

  9. Taking ownership of safety. What are the active ingredients of safety coaching and how do they impact safety outcomes in critical offshore working environments?

    Science.gov (United States)

    Krauesslar, Victoria; Avery, Rachel E; Passmore, Jonathan

    2015-01-01

    Safety coaching interventions have become a common feature in the safety critical offshore working environments of the North Sea. Whilst the beneficial impact of coaching as an organizational tool has been evidenced, there remains a question specifically over the use of safety coaching and its impact on behavioural change and producing safe working practices. A series of 24 semi-structured interviews were conducted with three groups of experts in the offshore industry: safety coaches, offshore managers and HSE directors. Using a thematic analysis approach, several significant themes were identified across the three expert groups including connecting with and creating safety ownership in the individual, personal significance and humanisation, ingraining safety and assessing and measuring a safety coach's competence. Results suggest clear utility of safety coaching when applied by safety coaches with appropriate coach training and understanding of safety issues in an offshore environment. The current work has found that the use of safety coaching in the safety critical offshore oil and gas industry is a powerful tool in managing and promoting a culture of safety and care.

  10. Safety performance evaluation of converging chevron pavement markings : final report.

    Science.gov (United States)

    2014-12-01

    The objectives of this study were (1) to perform a detailed safety analysis of converging chevron : pavement markings, quantifying the potential safety benefits and developing an understanding of the : incident types addressed by the treatment, and (...

  11. Collegiate Aviation Research and Education Solutions to Critical Safety Issues

    Science.gov (United States)

    Bowen, Brent (Editor)

    2002-01-01

    This Conference Proceedings is a collection of 6 abstracts and 3 papers presented April 19-20, 2001 in Denver, CO. The conference focus was "Best Practices and Benchmarking in Collegiate and Industry Programs". Topics covered include: satellite-based aviation navigation; weather safety training; human-behavior and aircraft maintenance issues; disaster preparedness; the collegiate aviation emergency response checklist; aviation safety research; and regulatory status of maintenance resource management.

  12. Definition and Means of Maintaining the Criticality Prevention Design Features Portion of the PFP Safety Envelope

    International Nuclear Information System (INIS)

    RAMBLE, A.L.

    2000-01-01

    The purpose of this document is to record the technical evaluation of the Operational Safety Requirements described in the Plutonium Finishing Plant Final (PFP) Operational Safety Requirements, WHC-SD-CP-OSR-010. Rev. 0-N , Section 3.1.1, ''Criticality Prevention System.'' This document, with its appendices, provides the following: (1) The results of a review of Criticality Safety Analysis Reports (CSAR), later called Criticality Safety Evaluation Reports (CSER), and Criticality Prevention Specifications (CPS) to determine which equipment or components analyzed in the CSER or CPS are considered as one of the two unlikely, independent, and concurrent changes before a criticality accident is possible. (2) Evaluations of equipment or components to determine the safety boundary for the system (Section 4). (3) A list of essential drawings that show the safety system or component (Appendix A). (4) A list of the safety envelope (SE) equipment (Appendix B). (5) Functional requirements for the individual safety envelope equipment (Sections 3 and 4). (6) A list of the operational and surveillance procedures necessary to maintain the system equipment within the safety envelope (Section 5)

  13. SRTC criticality safety technical review: Phase 1 criticality analysis for the 9972-9975 family of shipping casks: (SRT-CMA-940003)

    International Nuclear Information System (INIS)

    Rathbun, R.

    1994-01-01

    Review of SRT-CMA-940003, ''Phase I Criticality Analysis For The 9972-9975 Family Of Shipping Casks (U). (SRT-CMA-940003).'' January 22, 1994, has been performed by the SRTC Applied Physics Group. The NCSE is a criticality assessment of the 9972-9975 family of shipping casks. This work is a follow-on of a previous criticality safety evaluation, with the differences between this and the previous evaluation are that now wall tolerances are modeled and more sophisticated analytical methods are applied. The NCSE under review concludes that, with one exception, the previously specified plutonium and uranium mass limits for 9972-9975 family of shipping casks do ensure that WSRC Nuclear Criticality Safety Manual requirements (ref. 1) are satisfied. The one exception is that the plutonium mass limit for the 9974 cask had to be reduced from 4.4 to 4.3 kg. In contrast, the 7.5 kg uranium mass limit for the 9974 cask was raised to 14.5 kg, making the uranium mass identical for all casks in this family. This technical review consisted of an independent check of the methods and models employed, application of ANSI/ANS 8.1 and 8.15, and verification of WSRC Nuclear Criticality Safety Manual procedures

  14. Communication's Role in Safety Management and Performance for the Road Safety Practices

    OpenAIRE

    Salim Keffane (s)

    2014-01-01

    Communication among organizations could play an important role in increasing road safety. To get in-depth knowledge of its role, this study measured managers' and employees' perceptions of the communication's role on six safety management and performance criteria for road safety practices by conducting a survey using a questionnaire among 165 employees and 135 managers. Path analysis using AMOS-19 software shows that some of the safety management road safety practices have high correlation wi...

  15. American National Standards and the DOE - A cooperative effort to promote nuclear criticality safety

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1996-01-01

    The U.S. Department of Energy's (DOE's) new criticality safety order, DOE Order 420.1 (open-quotes Facility Safety,close quotes October 13, 1995), Sec. 4.3 (open-quotes Nuclear Criticality Safetyclose quotes), invokes, as an integral part, 12 appropriate American National Standards Institute/American Nuclear Society (ANSI/ANS) Series-8 standards for nuclear criticality safety, but with modifications. (The order that 420.1/4.3 replaced also invoked some ANSI/ANS Series-8 standards.) These modifications include DOE operation-specific exceptions to the standards and elaborations on some of the wording in the standards

  16. 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.

  17. I. Reactor safety (including comments on criticisms of WASH-1400)

    International Nuclear Information System (INIS)

    1976-01-01

    A major concern in any nuclear power programme is a reactor accident resulting in a large release of radioactivity to the environment. Serious reactor accidents are possible and the risk of such accidents cannot be reduced to zero i.e. absolute safety cannot be assured. All that can be expected is that the measures used to ensure safety in the design and operation of a reactor are such that the risk of accident is reduced to acceptably low levels. No member of the general public is known to have died or been injured as a result of an accident in over 1000 commercial nuclear power reactor-years. Some accidents in power reactors in operation today have come close enough to an environmental release of radioactivity to cause serious public concern about future safety. Apparent inadequacies in safety practices disclosed by former members of the nuclear power industry have added to this concern. To obtain an objective appraisal of the reactor safety issue this report examines the measures taken in the design and operation of nuclear reactors to reduce the probability of accident to acceptably low levels

  18. Conditions for Critical Performativity in a Polycontextural Society

    DEFF Research Database (Denmark)

    Knudsen, Morten

    2017-01-01

    of sentences to a societal level, and taking inspiration from the failure of Marxist notions of the relation between theory and practice, the paper suggests thinking critical performativity under the conditions of differentiation. This idea is qualified by means of Niklas Luhmann and his theory...... elements of a critical research strategy – under polycontextural conditions - and four guidelines for a critical science are suggested....

  19. Effect of fissile isotope burnup on criticality safety for stored disintegrated fuel rods

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Selby, G.P.

    1978-09-01

    If the fuel rods were to disintegrate and water added, a criticality could occur in a 13-in. PWR canister with fresh fuel enriched to 3.5 wt % 235 U. The question is, ''If credit could be taken for burnup, could this indicate a subcritical condition.'' In attempting to answer this question, a series of calculations were performed. A set of isotopic concentrations were generated for 5,000, 10,000, 15,000, and 20,000 MWD/MTU burnup levels. Four reflector materials, water, concrete and two types of soil, were considered. Results indicate that allowing credit for fissile isotope burnup does not completely remove the concern for criticality safety in the event of rod disintegration. Reactivities which are ''subcritical'' (k/sub eff/ = 0.95) would not occur for three of the four reflector materials at even the 20,000 MWD/MTU burnup level in the 13-in. canister. The water reflected canister would achieve the k/sub eff/ = 0.95 level near 18,000 MWD/MTU. A smaller canister could be postulated. If a quarter inch gap is allowed, a Westinghouse 17 x 17 PWR assembly requires a 12 1 / 4 inch diameter canister. For such a canister with water reflection the ''subcritical'' (k/sub eff/ = 0.95) level would be reached near 15,000 MWD/MTU. The soil reflected canisters would reach this level between 18,000 and 19,000 MWD/MTU. Considering the difficulties in taking credit for burnup, such modest gains in apparent safety are not encouraging. This situation might be improved, however, if credit were also taken for neutron absorption by fission product poisons produced during burnup. It is strongly recommended that other approaches to a solution of the criticality safety problem be considered

  20. Benchmarking road safety performances of countries.

    NARCIS (Netherlands)

    Wegman, F.C.M. & Oppe, S.

    2014-01-01

    In order to obtain political interest in road safety problems and to learn from other countries’ ‘good practices’, it is often helpful to compare one’s own safety situation with that of other countries. In a number of projects tools have been developed for such comparisons. These tools range from

  1. NMC and A and nuclear criticality safety systems integration: A prospective way for enhancement of the nuclear industry facilities safety

    International Nuclear Information System (INIS)

    Ryazanov, Boris G.; Sviridov, Victor I.; Frolov, Vladimir V.; Shvedov, Maxim O.; Mclaughlin, Thomas P.; Pruvost, Norman L.

    2003-01-01

    A considerable body of data has now been acquired about the principles, parameters and consequences of nuclear (criticality) accidents at facilities of the atomic industry in Russia, the United States, Great Britain and Japan. The total number of such accidents stands at 22. Russian and US specialists have prepared a rather extensive survey and analysis of these accidents. The final and important section of this survey is the lessons implied by the results of analysis of these 22 accidents. Among these lessons is the necessity of unconditional enforcement of control over the movement and transformations of special nuclear materials (SNM), and in particular fissile materials, (those SNMs with criticality accident concerns) during production and processing. Inadequacies in such control have been among the causes of most of the accidents that have occurred. Nuclear materials control and accounting (MC and A) for the purpose of ensuring storage reliability and nonproliferation safeguards is a major task of nuclear facilities in any nation. MC and A systems use the latest techniques and hardware for periodic control of SNM in specifically organized material balance areas. Immediate checking, periodic inventory of SNM, and measurements of the parameters of SNM at key points are the main sources of data for these systems. Data about the presence and sites of location of SNM in material balance areas that are acquired in inventories can be used for objective assessment of the status of nuclear safety. On the other hand, the inventory itself involves performance of operations that are unlike routine process engineering, and require special consideration of nuclear safety. Use of the techniques and hardware of MC and A systems not only for purposes of storage reliability, but also to ensure nuclear safety, will reduce the risk of nuclear accidents. This paper gives a concise overview of nuclear accidents that have occurred due to inadequacies in MC and A, and demonstrates

  2. Critical evaluation of nuclear safety reports Pt. 1

    International Nuclear Information System (INIS)

    Egely, Gy.

    1987-01-01

    Licensing procedures of siting, commissioning and operation of nuclear power plants in the USA, FRG, France and Japan are compared. The standard format and content of nuclear safety analysis reports including the general description of the plant, the presentation of the characteristics of siting, building structures, components, facilities, the reactors, the cooling system, the safety system, the measuring and control system, the power supply system, the auxilliary system, the energy transformation system, etc. are discussed in detail by the example of the US procedure. (V.N.)

  3. 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.

  4. Safety performance indicators used by the Russian Safety Regulatory Authority in its practical activities on nuclear power plant safety regulation

    International Nuclear Information System (INIS)

    Khazanov, A.L.

    2005-01-01

    The Sixth Department of the Nuclear, Industrial and Environmental Regulatory Authority of Russia, Scientific and Engineering Centre for Nuclear and Radiation Safety process, analyse and use the information on nuclear power plants (NPPs) operational experience or NPPs safety improvement. Safety performance indicators (SPIs), derived from processing of information on operational violations and analysis of annual NPP Safety Reports, are used as tools to determination of trends towards changing of characteristics of operational safety, to assess the effectiveness of corrective measures, to monitor and evaluate the current operational safety level of NPPs, to regulate NPP safety. This report includes a list of the basic SPIs, those used by the Russian safety regulatory authority in regulatory activity. Some of them are absent in list of IAEA-TECDOC-1141 ('Operational safety performance indicators for nuclear power plants'). (author)

  5. Critical Thinking and EFL Learners' Performance on Different Writing Modes

    Science.gov (United States)

    Golpour, Farhad

    2014-01-01

    The essential function of critical thinking in education is obvious by many studies done in this field. The main purpose of this article is to find the relationship between critical thinking levels of Iranian EFL learners and their performance on different modes of writing. The sample of the study selected among those who studying English at the…

  6. Critical Arts-Based Research in Education: Performing Undocumented Historias

    Science.gov (United States)

    Bagley, Carl; Castro-Salazar, Ricardo

    2012-01-01

    The article seeks to elucidate and academically position the genre of critical arts-based research in education. The article fuses Critical Race Theory (CRT), life history and performance, alongside work with undocumented American students of Mexican origin, to show how a politicised qualitative paradigmatic re envisioning can occur in which…

  7. Indicators of safety culture - selection and utilization of leading safety performance indicators

    Energy Technology Data Exchange (ETDEWEB)

    Reiman, Teemu; Pietikaeinen, Elina (VTT, Technical Research Centre of Finland (Finland))

    2010-03-15

    Safety indicators play a role in providing information on organizational performance, motivating people to work on safety and increasing organizational potential for safety. The aim of this report is to provide an overview on leading safety indicators in the domain of nuclear safety. The report explains the distinction between lead and lag indicators and proposes a framework of three types of safety performance indicators - feedback, monitor and drive indicators. Finally the report provides guidance for nuclear energy organizations for selecting and interpreting safety indicators. It proposes the use of safety culture as a leading safety performance indicator and offers an example list of potential indicators in all three categories. The report concludes that monitor and drive indicators are so called lead indicators. Drive indicators are chosen priority areas of organizational safety activity. They are based on the underlying safety model and potential safety activities and safety policy derived from it. Drive indicators influence control measures that manage the socio technical system; change, maintain, reinforce, or reduce something. Monitor indicators provide a view on the dynamics of the system in question; the activities taking place, abilities, skills and motivation of the personnel, routines and practices - the organizational potential for safety. They also monitor the efficacy of the control measures that are used to manage the socio technical system. Typically the safety performance indicators that are used are lagging (feedback) indicators that measure the outcomes of the socio technical system. Besides feedback indicators, organizations should also acknowledge the important role of monitor and drive indicators in managing safety. The selection and use of safety performance indicators is always based on an understanding (a model) of the socio technical system and safety. The safety model defines what risks are perceived. It is important that the safety

  8. Indicators of safety culture - selection and utilization of leading safety performance indicators

    International Nuclear Information System (INIS)

    Reiman, Teemu; Pietikaeinen, Elina

    2010-03-01

    Safety indicators play a role in providing information on organizational performance, motivating people to work on safety and increasing organizational potential for safety. The aim of this report is to provide an overview on leading safety indicators in the domain of nuclear safety. The report explains the distinction between lead and lag indicators and proposes a framework of three types of safety performance indicators - feedback, monitor and drive indicators. Finally the report provides guidance for nuclear energy organizations for selecting and interpreting safety indicators. It proposes the use of safety culture as a leading safety performance indicator and offers an example list of potential indicators in all three categories. The report concludes that monitor and drive indicators are so called lead indicators. Drive indicators are chosen priority areas of organizational safety activity. They are based on the underlying safety model and potential safety activities and safety policy derived from it. Drive indicators influence control measures that manage the socio technical system; change, maintain, reinforce, or reduce something. Monitor indicators provide a view on the dynamics of the system in question; the activities taking place, abilities, skills and motivation of the personnel, routines and practices - the organizational potential for safety. They also monitor the efficacy of the control measures that are used to manage the socio technical system. Typically the safety performance indicators that are used are lagging (feedback) indicators that measure the outcomes of the socio technical system. Besides feedback indicators, organizations should also acknowledge the important role of monitor and drive indicators in managing safety. The selection and use of safety performance indicators is always based on an understanding (a model) of the socio technical system and safety. The safety model defines what risks are perceived. It is important that the safety

  9. Possibilities and Limitations of Applying Software Reliability Growth Models to Safety- Critical Software

    International Nuclear Information System (INIS)

    Kim, Man Cheol; Jang, Seung Cheol; Ha, Jae Joo

    2006-01-01

    As digital systems are gradually introduced to nuclear power plants (NPPs), the need of quantitatively analyzing the reliability of the digital systems is also increasing. Kang and Sung identified (1) software reliability, (2) common-cause failures (CCFs), and (3) fault coverage as the three most critical factors in the reliability analysis of digital systems. For the estimation of the safety-critical software (the software that is used in safety-critical digital systems), the use of Bayesian Belief Networks (BBNs) seems to be most widely used. The use of BBNs in reliability estimation of safety-critical software is basically a process of indirectly assigning a reliability based on various observed information and experts' opinions. When software testing results or software failure histories are available, we can use a process of directly estimating the reliability of the software using various software reliability growth models such as Jelinski- Moranda model and Goel-Okumoto's nonhomogeneous Poisson process (NHPP) model. Even though it is generally known that software reliability growth models cannot be applied to safety-critical software due to small number of expected failure data from the testing of safety-critical software, we try to find possibilities and corresponding limitations of applying software reliability growth models to safety critical software

  10. 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)

  11. Performance management when innovation and learning become critical performance indicators

    NARCIS (Netherlands)

    Molleman, E.; Timmerman, H.

    2003-01-01

    If the organization's leading performance indicators shift towards innovation and the creation of knowledge, this will result in more non-routine work and a higher level of interdependency among workers. We argue that a contingent performance management (PM) system has to focus on Learning and group

  12. Criticality safety engineering at the Savannah River Site - the 1990s

    International Nuclear Information System (INIS)

    Chandler, J.R.; Apperson, C.E. Jr.

    1996-01-01

    The privatization and downsizing effort that is ongoing within the U.S. Department of Energy (DOE) is requiring a change in the management of criticality safety engineering resources at the Savannah River Site (SRS). Downsizing affects the number of criticality engineers employed by the prime contractor, Westinghouse Savannah River Company (WSRC), and privatization affects the manner in which business is conducted. In the past, criticality engineers at the SRS have been part of the engineering organizations that support each facility handling fissile material. This practice led to different criticality safety engineering organizations dedicated to fuel fabrication activities, reactor loading and unloading activities, separation and waste management operations, and research and development

  13. 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

  14. Failure Modes Effects and Criticality Analysis, an Underutilized Safety, Reliability, Project Management and Systems Engineering Tool

    Science.gov (United States)

    Mullin, Daniel Richard

    2013-09-01

    The majority of space programs whether manned or unmanned for science or exploration require that a Failure Modes Effects and Criticality Analysis (FMECA) be performed as part of their safety and reliability activities. This comes as no surprise given that FMECAs have been an integral part of the reliability engineer's toolkit since the 1950s. The reasons for performing a FMECA are well known including fleshing out system single point failures, system hazards and critical components and functions. However, in the author's ten years' experience as a space systems safety and reliability engineer, findings demonstrate that the FMECA is often performed as an afterthought, simply to meet contract deliverable requirements and is often started long after the system requirements allocation and preliminary design have been completed. There are also important qualitative and quantitative components often missing which can provide useful data to all of project stakeholders. These include; probability of occurrence, probability of detection, time to effect and time to detect and, finally, the Risk Priority Number. This is unfortunate as the FMECA is a powerful system design tool that when used effectively, can help optimize system function while minimizing the risk of failure. When performed as early as possible in conjunction with writing the top level system requirements, the FMECA can provide instant feedback on the viability of the requirements while providing a valuable sanity check early in the design process. It can indicate which areas of the system will require redundancy and which areas are inherently the most risky from the onset. Based on historical and practical examples, it is this author's contention that FMECAs are an immense source of important information for all involved stakeholders in a given project and can provide several benefits including, efficient project management with respect to cost and schedule, system engineering and requirements management

  15. Contractors’ Attitude towards Enhancing Safety Performance: Case Study on Construction Firms in Penang

    Directory of Open Access Journals (Sweden)

    Ulang N. Md

    2014-01-01

    Full Text Available A qualitative study was conducted to investigate the contractors’ attitude towards enhancing the safety performance in construction site. Despite the fact that there are many safety initiatives established by the government, the rates of accidents are still in a critically high condition. Thus the purpose of this research is to study the contractors’ attitude towards enhancing the implementation of safety management system in construction site in order to increase the safety awareness of construction practitioners in construction site and improve the safety condition of construction sites. This study is conducted through oral interviews with the construction practitioners, and visual inspection of construction sites. The attitudes of contractors are evaluated from 3 aspects: Contractors’ efforts in implement and enforce the safety rules, Contractors efforts in overcoming the rate of accidents, and Reasons given by the contractors for not implement safety law.

  16. Studies of safety and critical work situations in nuclear power plants: A human factors perspective

    International Nuclear Information System (INIS)

    Jacobsson Kecklund, L.

    1998-05-01

    The purpose of this thesis was to develop and apply different approaches for analyzing safety in critical work situations in real work settings in nuclear power plants, and also to identify safety enhancing measures by using the framework of interaction between human, organizational and technical subsystems. A Cognitive Psychology as well as a Stress Psychology framework was used. All studies were related to the annual outage operational state where the need for coping with many infrequent tasks, often carried out under high time pressure, puts great strain on the staff and organisation of the plant. In three studies the natural variations in the plant state, normal operation and annual outage operation, were used to explore human performance, work-related factors as well as coping and the operators' own resources and the relationship between them. In the annual outage condition high work demands, decreased sleepiness at night shift, more errors and less satisfaction with work performance quality was reported by maintenance as well as by control room operators. A relationship between high work demands and more organizational problems and reports of more frequent human errors and lower satisfactions with work performance quality was also identified in the annual outage condition. Moreover, a relationship between increased sleepiness during night shift, more frequent use of coping strategies and a higher frequency of human errors was reported. In two studies the Event and Barrier Function Model was applied to analyze the safety of barrier function systems inserted into work process sequences to protect the systems from the negative consequences of failures and errors. The model was also used to assess safety in relation to a technical and organizational change. The last study addressed changes in work performance and work-related factors in relation to a technical and organizational change of a safety significant work process involving increased automation and new

  17. An overview of criticality safety research at the All-Russian Research Institute of Experimental Physics

    Energy Technology Data Exchange (ETDEWEB)

    Kuvshinov, M.I.; Voinov, A.M.; Yuferev, V.I. [All-Russian Research Institute of Experimental Physics, Arzamas (Russian Federation)] [and others

    1997-06-01

    This paper presents a summary of experimental and calculational activities conducted at VNIIEF from the late 1940s to now to study the critical conditions of systems as part of a nuclear safety program. 9 refs., 1 tab.

  18. Seafood safety: economics of hazard analysis and Critical Control Point (HACCP) programmes

    National Research Council Canada - National Science Library

    Cato, James C

    1998-01-01

    .... This document on economic issues associated with seafood safety was prepared to complement the work of the Service in seafood technology, plant sanitation and Hazard Analysis Critical Control Point (HACCP) implementation...

  19. An overview of criticality safety research at the All-Russian Research Institute of Experimental Physics

    International Nuclear Information System (INIS)

    Kuvshinov, M.I.; Voinov, A.M.; Yuferev, V.I.

    1997-01-01

    This paper presents a summary of experimental and calculational activities conducted at VNIIEF from the late 1940s to now to study the critical conditions of systems as part of a nuclear safety program. 9 refs., 1 tab

  20. A Practical Risk Assessment Methodology for Safety-Critical Train Control Systems

    Science.gov (United States)

    2009-07-01

    This project proposes a Practical Risk Assessment Methodology (PRAM) for analyzing railroad accident data and assessing the risk and benefit of safety-critical train control systems. This report documents in simple steps the algorithms and data input...

  1. Single parameter controls for nuclear criticality safety at the Oak Ridge Y-12 Plant

    International Nuclear Information System (INIS)

    Baker, J.S.; Peek, W.M.

    1995-01-01

    At the Oak Ridge Y-12 Plant, there are numerous situations in which nuclear criticality safety must be assured and subcriticality demonstrated by some method other than the straightforward use of the double contingency principle. Some cases are cited, and the criticality safety evaluation of contaminated combustible waste collectors is considered in detail. The criticality safety evaluation for combustible collectors is based on applying one very good control to the one controllable parameter. Safety can only be defended when the contingency of excess density is limited to a credible value based on process knowledge. No reasonable single failure is found that will result in a criticality accident. The historically accepted viewpoint is that this meets double contingency, even though there are not two independent controls on the single parameter of interest

  2. Benchmarking Global Food Safety Performances: The Era of Risk Intelligence.

    Science.gov (United States)

    Valleé, Jean-Charles Le; Charlebois, Sylvain

    2015-10-01

    Food safety data segmentation and limitations hamper the world's ability to select, build up, monitor, and evaluate food safety performance. Currently, there is no metric that captures the entire food safety system, and performance data are not collected strategically on a global scale. Therefore, food safety benchmarking is essential not only to help monitor ongoing performance but also to inform continued food safety system design, adoption, and implementation toward more efficient and effective food safety preparedness, responsiveness, and accountability. This comparative study identifies and evaluates common elements among global food safety systems. It provides an overall world ranking of food safety performance for 17 Organisation for Economic Co-Operation and Development (OECD) countries, illustrated by 10 indicators organized across three food safety risk governance domains: risk assessment (chemical risks, microbial risks, and national reporting on food consumption), risk management (national food safety capacities, food recalls, food traceability, and radionuclides standards), and risk communication (allergenic risks, labeling, and public trust). Results show all countries have very high food safety standards, but Canada and Ireland, followed by France, earned excellent grades relative to their peers. However, any subsequent global ranking study should consider the development of survey instruments to gather adequate and comparable national evidence on food safety.

  3. Road safety performance indicators for the interurban road network.

    NARCIS (Netherlands)

    Yannis, G. Weijermars, W.A.M. Gitelman, V. Vis, M. Chazirisa, A. Papadimitriou, E. & Lima Azevedo, C.

    2013-01-01

    Various road safety performance indicators (SPIs) have been proposed for different road safety research areas, mainly as regards driver behaviour (e.g. seat belt use, alcohol, drugs, etc.) and vehicles (e.g. passive safety); however, no SPIs for the road network and design have been developed. The

  4. Use of Opioid Medications for Employees in Critical Safety or Security Positions and Positions with Safety Sensitive Duties

    Science.gov (United States)

    2017-01-30

    can cause harm) to the physical well-being of or jeopardize the security of the employee , co-workers, customers or the general public through a lapse...DEPARTMENT OF THE ARMY US ARMY PUBLIC HEALTH CENTER 5158 BLACKHAWK ROAD ABERDEEN PROVING GROUND MARYLAND 21010-5403 Directorate of Clinical... Employees in Critical Safety or Security Positions and Positions with Safety Sensitive Duties. 1. REFERENCES. A. Army Regulation 40-5, Preventive

  5. Critical Care Organizations: Business of Critical Care and Value/Performance Building.

    Science.gov (United States)

    Leung, Sharon; Gregg, Sara R; Coopersmith, Craig M; Layon, A Joseph; Oropello, John; Brown, Daniel R; Pastores, Stephen M; Kvetan, Vladimir

    2018-01-01

    New, value-based regulations and reimbursement structures are creating historic care management challenges, thinning the margins and threatening the viability of hospitals and health systems. The Society of Critical Care Medicine convened a taskforce of Academic Leaders in Critical Care Medicine on February 22, 2016, during the 45th Critical Care Congress to develop a toolkit drawing on the experience of successful leaders of critical care organizations in North America for advancing critical care organizations (Appendix 1). The goal of this article was to provide a roadmap and call attention to key factors that adult critical care medicine leadership in both academic and nonacademic setting should consider when planning for value-based care. Relevant medical literature was accessed through a literature search. Material published by federal health agencies and other specialty organizations was also reviewed. Collaboratively and iteratively, taskforce members corresponded by electronic mail and held monthly conference calls to finalize this report. The business and value/performance critical care organization building section comprised of leaders of critical care organizations with expertise in critical care administration, healthcare management, and clinical practice. Two phases of critical care organizations care integration are described: "horizontal," within the system and regionalization of care as an initial phase, and "vertical," with a post-ICU and postacute care continuum as a succeeding phase. The tools required for the clinical and financial transformation are provided, including the essential prerequisites of forming a critical care organization; the manner in which a critical care organization can help manage transformational domains is considered. Lastly, how to achieve organizational health system support for critical care organization implementation is discussed. A critical care organization that incorporates functional clinical horizontal and

  6. The Qualification Experiences for Safety-critical Software of POSAFE-Q

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jang Yeol; Son, Kwang Seop; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Programmable Logic Controllers (PLC) have been applied to the Reactor Protection System (RPS) and the Engineered Safety Feature (ESF)-Component Control System (CCS) as the major safety system components of nuclear power plants. This paper describes experiences on the qualification of the safety-critical software including the pCOS kernel and system tasks related to a safety-grade PLC, i.e. the works done for the Software Verification and Validation, Software Safety Analysis, Software Quality Assurance, and Software Configuration Management etc.

  7. 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.

  8. NuSEE: an integrated environment of software specification and V and V for PLC based safety-critical systems

    International Nuclear Information System (INIS)

    Koo, Seo Ryong; Seong, Poong Hyun; Yoo, Jun Beom; Cha, Sung Deok; Youn, Cheong; Han, Hyun Chul

    2006-01-01

    As the use of digital systems becomes more prevalent, adequate techniques for software specification and analysis have become increasingly important in Nuclear Power Plant (NPP) safety-critical systems. Additionally, the importance of software Verification and Validation (V and V) based on adequate specification has received greater emphasis in view of improving software quality. For thorough V and V of safety-critical systems, V and V should be performed throughout the software lifecycle. However, systematic V and V is difficult as it involves many manual-oriented tasks. Tool support is needed in order to more conveniently perform software V and V. In response, we developed four kinds of Computer Aided Software Engineering (CASE) tools to support system specification for a formal-based analysis according to the software lifecycle. In this work, we achieved optimized integration of each tool. The toolset, NuSEE, is an integrated environment for software specification and V and V for PLC based safety-critical systems. In accordance with the software lifecycle, NuSEE consists of NuSISRT for the concept phase, NuSRS for the requirements phase, NuSDS for the design phase and NuSCM for configuration management. It is believed that after further development our integrated environment will be a unique and promising software specification and analysis toolset that will support the entire software lifecycle for the development of PLC based NPP safety-critical systems

  9. Quantitative assessment of safety barrier performance in the prevention of domino scenarios triggered by fire

    International Nuclear Information System (INIS)

    Landucci, Gabriele; Argenti, Francesca; Tugnoli, Alessandro; Cozzani, Valerio

    2015-01-01

    The evolution of domino scenarios triggered by fire critically depends on the presence and the performance of safety barriers that may have the potential to prevent escalation, delaying or avoiding the heat-up of secondary targets. The aim of the present study is the quantitative assessment of safety barrier performance in preventing the escalation of fired domino scenarios. A LOPA (layer of protection analysis) based methodology, aimed at the definition and quantification of safety barrier performance in the prevention of escalation was developed. Data on the more common types of safety barriers were obtained in order to characterize the effectiveness and probability of failure on demand of relevant safety barriers. The methodology was exemplified with a case study. The results obtained define a procedure for the estimation of safety barrier performance in the prevention of fire escalation in domino scenarios. - Highlights: • We developed a methodology for the quantitative assessment of safety barriers. • We focused on safety barriers aimed at preventing domino effect triggered by fire. • We obtained data on effectiveness and availability of the safety barriers. • The methodology was exemplified with a case study of industrial interest. • The results showed the role of safety barriers in preventing fired domino escalation

  10. A systematic approach for safety evidence collection in the safety-critical domain

    NARCIS (Netherlands)

    Lin, H.; Wu, Ji; Yuan, C.; Luo, Y.; Brand, van den M.G.J.; Engelen, L.J.P.

    2015-01-01

    In order to show that the required safety objectives are met, it is necessary to collect safety evidence in the form of consistent and complete data. However, manual safety evidence collection is usually tedious and time-consuming, due to a large number of artifacts and implicit relations between

  11. The Dynamics of Agile Practices for Safety-Critical Software Development

    DEFF Research Database (Denmark)

    Nielsen, Peter Axel; Tordrup Heeager, Lise

    2017-01-01

    This short paper reports from a case study of the agile development of safety-critical software. It utilizes a framework of dynamic relationships between agile practices with the purpose of demonstrating the utility of the framework to understand a case in its context, and it shows significant...... dynamics. The study is concluded by pointing at which further research on the framework is required to use the framework in managing the agile development of safety-critical software....

  12. CTMCONTROL: Addressing the MC/DC Objective for Safety-Critical Automotive Software

    OpenAIRE

    Mjeda , Anila; Hinchey , Mike

    2013-01-01

    International audience; We propose a method tailored to the requirements of safety-critical embedded automotive software, named CTMCONTROL. CTMCONTROL has a par-ticular focus on the specification-based control logic of the system under test and offers improvements in testing coverage metrics over a classic method which is routinely used in industry. The proposed method targets the Modified Condition/ Decision Coverage (MC/DC) objective for automotive safety-critical software. CTMCONTROL is va...

  13. 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)

  14. Use of safety management practices for improving project performance.

    Science.gov (United States)

    Cheng, Eddie W L; Kelly, Stephen; Ryan, Neal

    2015-01-01

    Although site safety has long been a key research topic in the construction field, there is a lack of literature studying safety management practices (SMPs). The current research, therefore, aims to test the effect of SMPs on project performance. An empirical study was conducted in Hong Kong and the data collected were analysed with multiple regression analysis. Results suggest that 3 of the 15 SMPs, which were 'safety committee at project/site level', 'written safety policy', and 'safety training scheme' explained the variance in project performance significantly. Discussion about the impact of these three SMPs on construction was provided. Assuring safe construction should be an integral part of a construction project plan.

  15. ASIC-based design of NMR system health monitor for mission/safety-critical applications.

    Science.gov (United States)

    Balasubramanian, P

    2016-01-01

    N-modular redundancy (NMR) is a generic fault tolerance scheme that is widely used in safety-critical circuit/system designs to guarantee the correct operation with enhanced reliability. In passive NMR, at least a majority (N + 1)/2 out of N function modules is expected to operate correctly at any time, where N is odd. Apart from a conventional realization of the NMR system, it would be useful to provide a concurrent indication of the system's health so that an appropriate remedial action may be initiated depending upon an application's safety criticality. In this context, this article presents the novel design of a generic NMR system health monitor which features: (i) early fault warning logic, that is activated upon the production of a conflicting result by even one output of any arbitrary function module, and (ii) error signalling logic, which signals an error when the number of faulty function modules unfortunately attains a majority and the system outputs may no more be reliable. Two sample implementations of NMR systems viz. triple modular redundancy and quintuple modular redundancy with the proposed system health monitoring are presented in this work, with a 4-bit ALU used for the function modules. The simulations are performed using a 32/28 nm CMOS process technology.

  16. Verification of MCNP6.2 for Nuclear Criticality Safety Applications

    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)

    2017-05-10

    Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.

  17. A Method to Select Test Input Cases for Safety-critical Software

    International Nuclear Information System (INIS)

    Kim, Heeeun; Kang, Hyungook; Son, Hanseong

    2013-01-01

    This paper proposes a new testing methodology for effective and realistic quantification of RPS software failure probability. Software failure probability quantification is important factor in digital system safety assessment. In this study, the method for software test case generation is briefly described. The test cases generated by this method reflect the characteristics of safety-critical software and past inputs. Furthermore, the number of test cases can be reduced, but it is possible to perform exhaustive test. Aspect of software also can be reflected as failure data, so the final failure data can include the failure of software itself and external influences. Software reliability is generally accepted as the key factor in software quality since it quantifies software failures which can make a powerful system inoperative. In the KNITS (Korea Nuclear Instrumentation and Control Systems) project, the software for the fully digitalized reactor protection system (RPS) was developed under a strict procedure including unit testing and coverage measurement. Black box testing is one type of Verification and validation (V and V), in which given input values are entered and the resulting output values are compared against the expected output values. Programmable logic controllers (PLCs) were used in implementing critical systems and function block diagram (FBD) is a commonly used implementation language for PLC

  18. Ontario Hydro experience in the identification and mitigation of potential failures in safety critical software systems

    International Nuclear Information System (INIS)

    Huget, R.G.; Viola, M.; Froebel, P.A.

    1995-01-01

    Ontario Hydro has had experience in designing and qualifying safety critical software used in the reactor shutdown systems of its nuclear generating stations. During software design, an analysis of system level hazards and potential hardware failure effects provide input to determining what safeguards will be needed. One form of safeguard, called software self checks, continually monitor the health of the computer on line. The design of self checks usually is a trade off between the amount of computing resources required, the software complexity, and the level of safeguarding provided. As part of the software verification activity, a software hazards analysis is performed, which identifiers any failure modes that could lead to the software causing an unsafe state, and which recommends changes to mitigate that potential. These recommendations may involve a re-structuring of the software to be more resistant to failure, or the introduction of other safeguarding measures. This paper discusses how Ontario Hydro has implemented these aspects of software design and verification into safety critical software used in reactor shutdown systems

  19. 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.

  20. 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

  1. 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)

  2. 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)

  3. Model checking of safety-critical software in the nuclear engineering domain

    International Nuclear Information System (INIS)

    Lahtinen, J.; Valkonen, J.; Björkman, K.; Frits, J.; Niemelä, I.; Heljanko, K.

    2012-01-01

    Instrumentation and control (I and C) systems play a vital role in the operation of safety-critical processes. Digital programmable logic controllers (PLC) enable sophisticated control tasks which sets high requirements for system validation and verification methods. Testing and simulation have an important role in the overall verification of a system but are not suitable for comprehensive evaluation because only a limited number of system behaviors can be analyzed due to time limitations. Testing is also performed too late in the development lifecycle and thus the correction of design errors is expensive. This paper discusses the role of formal methods in software development in the area of nuclear engineering. It puts forward model checking, a computer-aided formal method for verifying the correctness of a system design model, as a promising approach to system verification. The main contribution of the paper is the development of systematic methodology for modeling safety critical systems in the nuclear domain. Two case studies are reviewed, in which we have found errors that were previously not detected. We also discuss the actions that should be taken in order to increase confidence in the model checking process.

  4. 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

  5. Criticality accident studies and research performed in the Valduc criticality laboratory, France

    International Nuclear Information System (INIS)

    Barbry, F.; Fouillaud, P.

    2001-01-01

    In 1967, the IPSN (Institut de Protection et de Surete Nucleaire - Nuclear Protection and Safety Institute) started studies and research in France on criticality accidents, with the objective of improving knowledge and modelling of accidents in order to limit consequences to the public, the environment and installations. The criticality accident is accompanied by an intense emission of neutronic and gamma radiation and releases of radioactive products in the form of gas and aerosols, generating irradiation and contamination risks. The main objectives of the studies carried out, particularly using the CRAC installation and the SILENE reactor at Valduc (France), were to model the physics of criticality accidents, to estimate the risks of irradiation and radioactive releases, to elaborate an accident detection system and to provide information for intervention plans. This document summarizes the state of knowledge in the various fields mentioned above. The results of experiments carried out in the Valduc criticality laboratory are used internationally as reference data for the qualification of calculation codes and the assessment of the consequences of a criticality accident. The SILENE installation, that reproduces the various conditions encountered during a criticality accident, is also a unique international research tool for studies and training on those matters. (author)

  6. Design criteria for the 218-group criticality safety reference library

    International Nuclear Information System (INIS)

    Westfall, R.M.; Ford, W.E. III; Webster, C.C.

    1978-01-01

    The generation of a 218-group neutron cross-section library from ENDF/B-IV data is described. Experience in selecting broad-group subsets and applying them in the analysis of critical experiments is related. Recommendations on the use of the 218-group library are made. 3 figures, 5 tables

  7. An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, Roger J.

    1989-09-01

    This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.

  8. Estimating Impact and Frequency of Risks to Safety and Mission Critical Systems Using CVSS

    NARCIS (Netherlands)

    Houmb, S.H.; Nunes Leal Franqueira, V.; Engum, E.A.

    2008-01-01

    Many safety and mission critical systems depend on the correct and secure operation of both supportive and core software systems. E.g., both the safety of personnel and the effective execution of core missions on an oil platform depend on the correct recording storing, transfer and interpretation of

  9. Classification for Safety-Critical Car-Cyclist Scenarios Using Machine Learning

    NARCIS (Netherlands)

    Cara, I.; Gelder, E.D.

    2015-01-01

    The number of fatal car-cyclist accidents is increasing. Advanced Driver Assistance Systems (ADAS) can improve the safety of cyclists, but they need to be tested with realistic safety-critical car-cyclist scenarios. In order to store only relevant scenarios, an online classification algorithm is

  10. 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

  11. Safety critical systems handbook a straightforward guide to functional safety : IEC 61508 (2010 edition) and related standards

    CERN Document Server

    Smith, David J

    2010-01-01

    Electrical, electronic and programmable electronic systems increasingly carry out safety functions to guard workers and the public against injury or death and the environment against pollution. The international functional safety standard IEC 61508 was revised in 2010, and this is the first comprehensive guide available to the revised standard. As functional safety is applicable to many industries, this book will have a wide readership beyond the chemical and process sector, including oil and gas, power generation, nuclear, aircraft, and automotive industries, plus project, instrumentation, design, and control engineers. * The only comprehensive guide to IEC 61508, updated to cover the 2010 amendments, that will ensure engineers are compliant with the latest process safety systems design and operation standards* Helps readers understand the process required to apply safety critical systems standards* Real-world approach helps users to interpret the standard, with case studies and best practice design examples...

  12. Job Demands-Control-Support model and employee safety performance.

    Science.gov (United States)

    Turner, Nick; Stride, Chris B; Carter, Angela J; McCaughey, Deirdre; Carroll, Anthony E

    2012-03-01

    The aim of this study was to explore whether work characteristics (job demands, job control, social support) comprising Karasek and Theorell's (1990) Job Demands-Control-Support framework predict employee safety performance (safety compliance and safety participation; Neal and Griffin, 2006). We used cross-sectional data of self-reported work characteristics and employee safety performance from 280 healthcare staff (doctors, nurses, and administrative staff) from Emergency Departments of seven hospitals in the United Kingdom. We analyzed these data using a structural equation model that simultaneously regressed safety compliance and safety participation on the main effects of each of the aforementioned work characteristics, their two-way interactions, and the three-way interaction among them, while controlling for demographic, occupational, and organizational characteristics. Social support was positively related to safety compliance, and both job control and the two-way interaction between job control and social support were positively related to safety participation. How work design is related to employee safety performance remains an important area for research and provides insight into how organizations can improve workplace safety. The current findings emphasize the importance of the co-worker in promoting both safety compliance and safety participation. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

  13. Decomobil, Deliverable 3.6, Human Centred Design for Safety Critical Transport Systems

    OpenAIRE

    PAUZIE, Annie; MENDOZA, Lucile; SIMOES, Anabela; BELLET, Thierry; MOREAU, Fabien

    2014-01-01

    The scientific seminar on 'Human Centred Design for Safety Critical Transport Systems' organized in the framework of DECOMOBIL has been held the 8th of September 2014 in Lisbon, Portugal, hosted by ADI/ISG. The aims of the event were to present the scientific problematic related to the safety of the complex transport systems and the increasing importance of human-­centred design, with a specific focus on Resilience Engineering concept, a new approach to safety management in highly complex sys...

  14. Optimal Braking Patterns and Forces in Autonomous Safety-Critical Maneuvers

    OpenAIRE

    Fors, Victor

    2018-01-01

    The trend of more advanced driver-assistance features and the development toward autonomous vehicles enable new possibilities in the area of active safety. With more information available in the vehicle about the surrounding traffic and the road ahead, there is the possibility of improved active-safety systems that make use of this information for stability control in safety-critical maneuvers. Such a system could adaptively make a trade-off between controlling the longitudinal, lateral, and ...

  15. The implementation of a burnup credit based criticality safety assessment in the THORP head end plant

    International Nuclear Information System (INIS)

    Gulliford, J.; Edge, J.A.; Gracey, J.; Harris, N.

    2003-01-01

    A new criticality safety assessment based on Actinide-Only Burnup Credit has been developed to cover operations in BNFL's Thermal Oxide Reprocessing Plant (THORP). Reduction of the gadolinium concentration leads to significant reduction in active waste volumes. Detailed description of the methodology was presented at ICNC 1999 and the basic components of the approved safety case have remained unchanged from those proposed then. This paper presents a brief summary of the new methodology, and describes further analyses carried out to quantify additional safety margins. These additional margins are not credited in the derivation of the operating limits, but provide further evidence of the fault tolerance inherent in the new regime. As part of the arrangements to monitor the overall performance of the plant and instrumentation under the new regime, various analyses of plant data are made, including 'on-line' cross checks of measured versus expected fuel parameters (i.e. in addition to the checks on Residual Enrichment). Statistical analyses of data are made and compared with similar data from earlier batches. A summary of analyses made on some of the early fuel batches is presented here. A summary of the likely further development in the Burnup Credit methodology is given in this paper. (author)

  16. Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); No, Young Hun; Song, Seung Hwan [POSCON, Seoul (Korea, Republic of)

    2006-07-01

    The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants.

  17. Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

    International Nuclear Information System (INIS)

    Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young; No, Young Hun; Song, Seung Hwan

    2006-01-01

    The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants

  18. Sensitivity-Uncertainty Techniques for Nuclear Criticality Safety

    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)

    2017-08-07

    The sensitivity and uncertainty analysis course will introduce students to keff sensitivity data, cross-section uncertainty data, how keff sensitivity data and keff uncertainty data are generated and how they can be used. Discussion will include how sensitivity/uncertainty data can be used to select applicable critical experiments, to quantify a defensible margin to cover validation gaps and weaknesses, and in development of upper subcritical limits.

  19. Cultural safety and the challenges of translating critically oriented knowledge in practice.

    Science.gov (United States)

    Browne, Annette J; Varcoe, Colleen; Smye, Victoria; Reimer-Kirkham, Sheryl; Lynam, M Judith; Wong, Sabrina

    2009-07-01

    Cultural safety is a relatively new concept that has emerged in the New Zealand nursing context and is being taken up in various ways in Canadian health care discourses. Our research team has been exploring the relevance of cultural safety in the Canadian context, most recently in relation to a knowledge-translation study conducted with nurses practising in a large tertiary hospital. We were drawn to using cultural safety because we conceptualized it as being compatible with critical theoretical perspectives that foster a focus on power imbalances and inequitable social relationships in health care; the interrelated problems of culturalism and racialization; and a commitment to social justice as central to the social mandate of nursing. Engaging in this knowledge-translation study has provided new perspectives on the complexities, ambiguities and tensions that need to be considered when using the concept of cultural safety to draw attention to racialization, culturalism, and health and health care inequities. The philosophic analysis discussed in this paper represents an epistemological grounding for the concept of cultural safety that links directly to particular moral ends with social justice implications. Although cultural safety is a concept that we have firmly positioned within the paradigm of critical inquiry, ambiguities associated with the notions of 'culture', 'safety', and 'cultural safety' need to be anticipated and addressed if they are to be effectively used to draw attention to critical social justice issues in practice settings. Using cultural safety in practice settings to draw attention to and prompt critical reflection on politicized knowledge, therefore, brings an added layer of complexity. To address these complexities, we propose that what may be required to effectively use cultural safety in the knowledge-translation process is a 'social justice curriculum for practice' that would foster a philosophical stance of critical inquiry at both the

  20. 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)