WorldWideScience

Sample records for fuel criticality safety

  1. 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

  2. Study of criticality safety of fresh fuel

    International Nuclear Information System (INIS)

    Pesic, M.; Dasic, N.

    1998-01-01

    A study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements, in the original metal containers (OMC) placed in the special metal protection 'cage' and in a designed wooden storage case (WSC), at the RB research reactor is carried out by using the well-known MCNP TM computer code. It is shown for the first time that these HEU and LEU fresh fuel elements, stored in the WSC or in the OMCs, are far away from criticality, even in a case of unexpected flooding of the storage by light water. (author)

  3. Nuclear fuel technology - Administrative criteria related to nuclear criticality safety

    International Nuclear Information System (INIS)

    2004-01-01

    An effective nuclear criticality-safety programme includes cooperation among management, supervision, and the nuclear criticality-safety staff and, for each employee, relies upon conformance with operating procedures. Although the extent and complexity of safety-related activities may vary greatly with the size and type of operation with fissile material, certain safety elements are common. This International Standard represents a codification of such elements related to nuclear criticality safety. General guidance for nuclear criticality safety may be found in ISO 1709. The responsibilities of management, supervision, and the nuclear criticality-safety staff are addressed. The Objectives and characteristics of operating and emergency procedures are included in this International Standard. ISO 14943 was prepared by Technical Committee ISO/TC 85, Nuclear energy, Subcommittee SC 5, Nuclear fuel technology

  4. 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

  5. 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)

  6. Criticality safety assessment of WWER-1000 spent fuel cask

    International Nuclear Information System (INIS)

    Apostolov, T.; Manolova, M.; Prodanova, R.

    2001-01-01

    A methodology that allows taking credit for burnup in the criticality safety analysis of WWER spent fuel casks is presented. It is based on the two world well known and used code systems:NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The results of criticality calculations of WWER-1000 spent fuel storage and transportation cask, applying burnup credit is shown. The depletion calculations have been carried out for three types of WWER-1000 fuel assemblies (with enrichment of 3.0%, 3.3% and 3.3% profiled) by modelling the real operational history of the first three fuel cycles at unit 6, Kozloduy NPP. The criticality calculational model has been developed on the basis of real fuel cask, designed by the Izorskie zavody. The results obtained show that the criticality safety criterion K eff < 0.95 is satisfied for both fresh and spent fuel. Besides, the implementation of burnup credit accounts for the reduced reactivity of spent fuel and allows evaluating the conservatism of the fresh fuel assumption. (author)

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

    International Nuclear Information System (INIS)

    Chad Pope

    2007-01-01

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

  8. 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

  9. 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

  10. 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)

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

    International Nuclear Information System (INIS)

    Marshall, William J.; Wagner, John C.

    2012-01-01

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

  12. Credit to fuel burnup for criticality safety evaluations in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    1998-01-01

    The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analysis characteristics specific to burnup credit are described, namely the need to increase the experimental data to allow for a more detailed validation of the depletion codes, and of the criticality codes when applied to spent fuel. Reactivity effects that arise in burnup credit analysis, such as axial and radial effects, fuel irradiation history and others are revised. The methods used to address them in the approved methodologies are outlined. Finally, the regulatory approach used to accept these new analytical methodologies is described. (author)

  13. 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

  14. Criticality safety assessor training at British Nuclear Fuels, Sellafield

    International Nuclear Information System (INIS)

    Gunston, K.J.

    1993-01-01

    In accordance with company policy agreed to in April 1986, graduate new entrants to British Nuclear Fuels (BNFL) Sellafield join a management trainee/appraisal scheme. The purpose of this scheme is that while doing a real job, the trainee should undergo structured training and be given the opportunity to develop both personally and professionally. As part of this scheme, each trainee has a structured training program that is devised to fulfill the requirements of the individual, the department, the site and the professional body to which the trainee aspires. This paper outlines the management trainee training/appraisal system and the structured training program that is used to train criticality safety assessors at BNFL Sellafield

  15. 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

  16. 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.)

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

    International Nuclear Information System (INIS)

    Tang, J.S.

    1980-11-01

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

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

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1999-01-01

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

  19. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based

  20. 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.

  1. 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)

  2. Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

    International Nuclear Information System (INIS)

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

    1993-01-01

    The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality

  3. Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Lake, W.H. (USDOE, Washington, DC (USA)); Sanders, T.L. (Sandia National Labs., Albuquerque, NM (USA)); Parks, C.V. (Oak Ridge National Lab., TN (USA))

    1990-01-01

    This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs.

  4. Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

    International Nuclear Information System (INIS)

    Lake, W.H.; Sanders, T.L.; Parks, C.V.

    1990-01-01

    This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs

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

    International Nuclear Information System (INIS)

    Richard, R.F.

    1995-01-01

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

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

    Science.gov (United States)

    Baranko, Benjamin T.

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

  7. Study on burnup credit evaluation method at JAERI towards securing criticality safety rationale for management of spent fuel

    International Nuclear Information System (INIS)

    Nomura, Y.

    1998-01-01

    Lately, due to massive accumulation of spent fuel discharged from light water reactors in Japan, it is gradually demanded to introduce the so-called burnup credit methodology into criticality safety design for nuclear fuel cycle facilities, such as spent fuel storage pools and transport casks. In order to save space in the spent fuel storage pool of the Rokkasho Reprocessing Plant, the burnup credit design has been firstly implemented for its criticality safety evaluation. Here, its design conditions and operational control procedures are briefly shown and research using burned fuel at JAERI is explained to support its licensing safety review, focusing on the relevant content of the Nuclear Criticality Safety Handbook of Japan, which has been prepared so far and planned in the near future. Finally, international co-operation for study on burnup credit issues practiced by JAERI is addressed. (author)

  8. Consideration on the partial moderation in criticality safety analysis of LWR fresh fuel storage

    International Nuclear Information System (INIS)

    Tanaka, S.; Tanimoto, R.; Suzuki, K.; Ishitobi, M.

    1987-01-01

    In criticality safety analyses of fuel fabrication facilities, neutron effective multiplication factor (k eff ) of a storage vault has been calculated assuming ''partial moderation'' in whole space (hereafter reffered to as unlimited partial moderation). Where the enrichment of fuels to be stored is about 3.5 % or less, calculated k eff is usually low enough to show subcriticality even in unlimited partial moderation. However, it is scheduled to elevate LWR fuels enrichment for economical higher burnup and the unlimited partial moderation would require to introduce neutron absorbers to maintain subcriticality. It is clear that this causes economical disadvantages, and hence we reconsidered this assumption to avoid such a condition. Reconsideration of the unlimited partial moderation was carried out in following steps. (1) Water quantity to be assumed in atmosphere to obtain criticality was revealed too much to realize. (2) Typical realistic water quantity in atmosphere was estimated to apply as an alternative assumption. (3) A fresh fuel assembly storage was chosen as a model array and calculations with lattice code WIMS-D 1 and Monte Calro code KENO-IV 2 were performed to compare new alternative assumption with the unlimited one. As results of the above calculations, maximum k eff of the array under the new assumption was remarkably reduced to the value less than 0.95 though the maximum k eff under the unlimited one was higher than 1.0. (author)

  9. Criticality safety evaluation report for 300 Area N Reactor fuel fabrication and storage facility

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1993-12-01

    Criticality safety support calculations for the 300 Area Fuels Manufacturing Facility have been performed to update values currently found in the criticality prevention specifications. The old safe masses and associated dimensions table has been updated with new values, which were produced using more modern computer codes that comply with Software Quality Assurance (SQA) requirements. Monte Carlo calculations, using MCNP (Carter 1991), were performed to spot check these results, obtained using the WIMS-E code (Gubbins et. al. 1982). In addition, the criticality considerations associated with certain accident, or upset, conditions were analyzed. These scenarios include fire, the bringing together of multiple safe masses into one neutronically coupled system, misstacking, and accidental interspersed moderation

  10. Criticality safety study of the MSRE Fuel Drain Tank Cell in Building 7503

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

    This report provides a criticality safety study of the molten salt reactor fuel currently being stored in the Fuel Drain Tank (FDT) Cell of the Molten Salt Reactor Experiment (MSRE) facility (Building 7503) located in the Melton Valley area of the Oak Ridge National Laboratory. The FDTs contain approximately 36 kg of uranium consisting primarily of {sup 233}U, some plutonium, and fission products in a solidified fluoride salt mixture. The nominal composition of the fluoride salt mixture in the FDTs is 42.16 wt % LiF, 35.79 wt % BeF{sub 2}, 21.01 wt % ZrF{sub 4}, 1.02 wt % UF{sub 4}, and 0.02 wt % PuF{sub 3}. The historic criticality safety study does not meet current standards. This work is in support of a new nuclear criticality safety analysis and approval update. Questions concerning the degree of subcriticality associated with the material in its current state and in its most reactive credible upset condition are addressed. The safety study consists of two parts. In the first part, the FDT Cell was modeled using KENO V.a and analyzed using a variety of cross-section sets. The base FDT Cell model was then modified to represent the most reactive credible upset conditions and analyzed. The second part consists of establishing a benchmark for the FDT Cell. Because of the lack of any other relevant benchmark experiments, the original MSRE was also modeled in KENO V.a and analyzed. The results of the reactor model were then compared with documented MSRE reactor conditions. The analysis shows that even under the most reactive credible upset conditions, the MSRE FDT Cell is significantly subcritical.

  11. Criticality safety criteria for the handling, storage, and transportation of LWR fuel outside reactors: ANS-8.17-1984

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1991-01-01

    The potential for criticality accidents during the handling, storage, and transportation of fuel for nuclear reactors represents a health and safety risk to personnel involved in these activities, as well as to the general public. Appropriate design of equipment and facilities, handling procedures, and personnel training can minimize this risk. Even though the focus of the American National Standard, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, ANSI/ANS-8.1-1983, is general criteria for the ensurance of criticality safety, Standard ANS-8.17-1984 provides additional guidance applicable to handling, storage, and transportation of light water reactor (LWR) nuclear fuel units in any phase of the fuel cycle outside the reactor core. The intended application of the standard focused on the criticality safety considerations in the design of facilities or operations in fuel cycle facilities where the fissile material is in the form of rods or assemblies of rods. The standard consists of two major sections. The first provides general safety criteria. The second major section provides the criteria to establish subcriticality. In performing criticality safety evaluations, it is necessary to ensure consideration of the most reactive credible conditions based on the fuel design procedures, and moderation and reflection conditions

  12. 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

  13. Criticality safety strategy and analysis summary for the fuel cycle facility electrorefiner at Argonne National Laboratory West

    International Nuclear Information System (INIS)

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

    1996-01-01

    As part of the termination activities of Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory (ANL) West, the spent metallic fuel from EBR-II will be treated in the fuel cycle facility (FCF). A key component of the spent-fuel treatment process in the FCF is the electrorefiner (ER) in which the actinide metals are separated from the active metal fission products and the reactive bond sodium. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt, and refined uranium or uranium/plutonium products are deposited at cathodes. The criticality safety strategy and analysis for the ANL West FCF ER is summarized. The FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. To show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOEs) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOEs, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that will verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality

  14. 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)

  15. 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

  16. 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.

  17. Criticality safety criteria for the handling, storage, and transportation of LWR fuel outside reactors: ANS-8.17-1984

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1996-01-01

    The potential for criticality accidents during the handling, storage, and transportation of fuel for nuclear reactors represents a health and safety risk to personnel involved in these activities, as well as to the general public. Appropriate design of equipment and facilities, handling procedures, and personnel training can minimize this risk. Even though the focus of the American National Standard, 'Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,' ANSI/ANS-8.1-1983, is general criteria for the ensurance of criticality safety, ANS-8.17-1984, provides additional guidance applicable to handling, storage, and transportation of light-water- reactor (LWR) nuclear fuel units in any phase of the fuel cycle outside the reactor core. ANS-8.17 had its origin in the late 1970s when a work group consisting of representatives from private industry, personnel from government contractor facilities, and scientists and engineers from the national laboratories was established. The work of this group resulted in the issuance of ANSI/ANS-8.17 in January 1984. This document provides a discussion of this standard

  18. Development of advanced spent fuel management process / criticality safety analysis for integrated mockup and metallized spent fuel storage

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Shin, Hee Sung; Shin, Young Joon; Bae, Kang Mok

    1999-02-01

    Benchmark calculation for SCALE4.3 CSAS6 module and burnup credit criticality analysis performed by CSAS6 module are described in this report. Calculation biases by the SCALE4.3 CSAS6 module for PWR spent fuel, metallized spent fuel and aqueous nuclear materials have been determined on the basis of the benchmark to be 0.011, 0.023 and 0.010, respectively. The maximum allowable multiplication factor for an integrated mockup and metallized spent fuel storage is conservatively determined to be 0.927. With the aid of this code system, K eff values as a function of metallization ratio for the integrated mockup have been calculated. The maximum values of K eff for normal and hypothetical accident conditions are 0.346 and 0.598, respectively, much less than the maximum allowable multiplication factor of 0.927. Besides, burnup credit criticality analysis has been performed for infinite arrays of square and hexagonal canisters containing metallized spent fuel rods with different canister wall thickness, canister surface-to-surface distance and water content. It is revealed that the effective multiplication factor for canister arrays as mentioned above is well below the subcritical limit regardless of external conditions when its wall thickness is over 9 mm. (Author). 37 refs., 27 tabs., 64 figs

  19. Criticality impacts on LWR fuel storage efficiency

    International Nuclear Information System (INIS)

    Napolitano, D.

    1992-01-01

    This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

  20. Framatome-ANP France UO2 fuel fabrication - criticality safety analysis in the light of the 1999' Tokay Mura accident

    International Nuclear Information System (INIS)

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

    2004-01-01

    In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The Framatome-ANP production of its French low enriched (5 w/o) UO 2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO 2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach Framatome-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (authors)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. Criticality safety assessment by assuming spent fuel burnup distribution. Examination of various methods for setting burnup (1) (Contract research)

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Okuno, Hiroshi; Miyoshi, Yoshinori

    2004-03-01

    Firstly, concerning the methods to set burnup for depletion calculation linked with criticality safety evaluation taking burnup credit into consideration, the upper 50 cm averaged burnups approved by regulations in European countries and USA are comparatively evaluated. Secondary, errors produced by different shapes of axial spent fuel burnup distribution assumed for criticality calculation, bias errors associated with depletion calculation compensated by correction factors applied to calculated nuclide isotopic composition, and statistic errors exerted by variation of irradiation history parameters used as input data for depletion calculation, are separately evaluated by performing criticality analyses with the spent fuel transport cask model of OECD/NEA Burnup Credit Criticality Benchmark. As a result, methods are proposed to set equivalent burnups reduced from a given burnup so as to compensate these errors to obtain criticality calculation results on the conservative side. Finally, 'Equivalent Uniform Burnup' and Equivalent Initial Enrichment' which are derived by incorporating these errors synthetically, are described to mention possibility of their common usage irrespective of difference in spent fuel transport cask specification. (author)

  3. The Key-Role of shielding analysis in advanced Candu Fuel bundles nuclear safety improvement for some accidental criticality scenarios

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Rizoiu, A.; Olteanu, G.

    2008-01-01

    The paper aims to present the source term and photon dose rates estimation for advanced Candu fuel bundles in some accidental criticality scenarios. As reference, the Candu standard fuel bundle has been used. The scenarios take into account for a very short-time irradiated or spent fuel bundles for some configurations closed to criticality. In order to estimate irradiated fuel characteristic parameters and radiation doses, the ORNL's SCALE 5 codes Origin-S and Monte Carlo MORSE-SGC have been used. The paper includes the irradiated fuel characteristic parameters comparison for the considered Candu fuel bundles, providing also a comparison between the corresponding radiation doses

  4. 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

  5. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  6. Criticality safety assessment on the RSG-GAS spent fuel storage for anticipating the next core conversion program

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Kuntoro, Iman; Zuhair; Liem, Peng Hong

    2003-01-01

    Criticality assessment on the spent fuel storage racks of the RSG-GAS multipurpose reactor has been conducted to support the undergoing core conversion program, in which higher uranium fuel densities of silicide (up to 4.8 gU.cm -3 ) and molybdenum (up to 8.3 gU.cm -3 ) fuel elements are adopted to enhance the reactor performance, core cycle length and reactor utilization. In the assessment, the k eff of the rack as a function of fuel density is calculated for fresh fuel elements which is a very conservative approach recommended by IAEA. Besides fuel densities, effects of water densities due to pool water temperature variation, and the fuel elements' orientation on the k eff are analyzed as well. The criticality calculations are all carried out by using MNCP4B2 Monte Carlo code with ENDF/B-VI library. For the library sensitivity, JENDL-3.3 library is also used and compared. The calculation results show the most reactive condition is for the case when the spent fuel racks are filled with fresh U-6Mo fuel element with meat density of 8.30 gU.cm -3 . For all fuel types, density and operating condition, the calculated k eff with 3 times standard deviations are confirmed less than the allowable value of 0.95. It can be concluded that the existing spent fuel storage racks can be safely used for storing the planned high density uranium fuels. (author)

  7. 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

  8. 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)

  9. Assessment of criticality safety

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1983-01-01

    A study was undertaken to determine concerns and possible trends in nuclear criticality safety in DOE facilities. The information gathered from specialists in the field did not indicate that serious concerns or extends exist. The program was terminated before completion

  10. Criticality safety issues associated with the introduction of low void reactivity fuel in the Bruce reactors - a management and technical overview

    International Nuclear Information System (INIS)

    Thompson, J.W.; Austman, G.; Iglesias, F.; Schmeing, H.; Elliott, C.; Archinoff, G.

    2004-01-01

    The concept of criticality for operating reactor staff, particularly in a natural uranium-fuelled reactor, is relatively benign - the reactor is controlled at the critical condition by the regulating system. That is, issues related to criticality exist only within the reactor, in a set of carefully managed circumstances. With the introduction of enriched Low Void Reactivity Fuel (LVRF) into this operating environment comes a new 'concept of criticality', one which, although physically the same, cannot be treated in the same fashion. It may be the case that criticality can be achieved outside the reactor, albeit with a set of very pessimistic assumptions. Such 'inadvertent criticality' outside the reactor, should it occur, cannot be controlled. The consequences of such an inadvertent criticality could have far-reaching effects, not only in terms of severe health effects to those nearby, but also in terms of the negative impact on Bruce Power, and the Canadian nuclear industry in general. Thus the introduction of LVRF in the Bruce B reactors, and therefore the introduction of this new hazard, inadvertent criticality, warrants the development of a governance structure for its management. Such a program will consist of various elements, including the establishment of a framework to administer the criticality safety program, analytical assessment to support the process design, the development of operational procedures, the development of enhanced emergency procedures if necessary, and the implementation of a criticality safety training program. The entire package must be sufficient to demonstrate to station management, and the regulator, that the criticality safety risks associated with the implementation of enriched fuel have been properly evaluated, and that all necessary steps have been taken to effectively manage these risks. A well-founded Criticality Safety Program will offer such assurance. In this paper, we describe the establishment of a Criticality Safety

  11. 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)

  12. 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.)

  13. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  14. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  15. 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

  16. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  17. 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)

  18. Criticality safety calculations of storage canisters

    International Nuclear Information System (INIS)

    Agrenius, L.

    2002-04-01

    In the planned Swedish repository for deep disposal of spent nuclear fuel the fuel assemblies will be stored in storage canisters made of cast iron and copper. To assure safe storage of the fuel the requirement is that the normal criticality safety criteria have to be met. The effective neutron multiplication factor must not exceed 0.95 in the most reactive conditions including different kinds of uncertainties. In this report it is shown that the criteria could be met if credit for the reactivity decrease due to the burn up of the fuel is taken into account. The criticality safety criteria are based on the US NRC regulatory requirements for transportation and storage of spent fuel

  19. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2004-01-01

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging

  20. Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities (5). Evaluation method and trial evaluation of criticality accident

    International Nuclear Information System (INIS)

    Yamane, Yuichi; Abe, Hitoshi; Nakajima, Ken; Hayashi, Yoshiaki; Arisawa, Jun; Hayami, Satoru

    2010-01-01

    A special committee of 'Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)' was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for the Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objectives of this research are to obtain information useful for establishing quantitative performance objectives and to demonstrate risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the consequence analysis method for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution and fire (including the rapid decomposition of TBP complexes), resulting in the release of radioactive materials to the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this report, the evaluation methods of criticality accident, such as simplified methods, one-point reactor kinetics codes and quasi-static method, were investigated and their features were summarized to provide information useful for the safety evaluation of NFFs. In addition, several trial evaluations were performed for a hypothetical scenario of criticality accident using the investigated methods, and their results were compared. The release fraction of volatile fission products in a criticality accident was also investigated. (author)

  1. 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)

  2. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2004-09-10

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the

  3. Metal fuel safety performance

    International Nuclear Information System (INIS)

    Miles, K.J. Jr.; Tentner, A.M.

    1988-01-01

    The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

  4. Nuclear Criticality Safety Handbook, Version 2. English translation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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)

  5. Developing safety in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Brown, M.L.

    1996-01-01

    The nuclear fuel cycle had its origins in the new technology developed in the 1940s and 50s involving novel physical and chemical processes. At the front end of the cycle, mining, milling and fuel fabrication all underwent development, but in general the focus of process development and safety concerns was the reprocessing stage, with radiation, contamination and criticality the chief hazards. Safety research is not over and there is still work to be done in advancing technical knowledge to new generation nuclear fuels such as Mixed Oxide Fuel and in refining knowledge of margins and of potential upset conditions. Some comments are made on potential areas for work. The NUCEF facility will provide many useful data to aid safety analysis and accident prevention. The routine operations in such plants, basically chemical factories, requires industrial safety and in addition the protection of workers against radiation or contamination. The engineering and management measures for this were novel and the early operation of such plants pioneering. Later commissioning and operating experience has improved routine operating safety, leading to a new generation of factories with highly developed worker protection, engineering safeguards and safety management systems. Ventilation of contamination control zones, remote operation and maintenance, and advanced neutron shielding are engineering examples. In safety management, dose control practices, formally controlled operating procedures and safety cases, and audit processes are comparable with, or lead, best industry practice in other hazardous industries. Nonetheless it is still important that the knowledge and experience from operating plants continue to be gathered together to provide a common basis for improvement. The NEA Working Group on Fuel Cycle Safety provides a forum for much of this interchange. Some activities in the Group are described in particular the FINAS incident reporting system. (J.P.N.)

  6. 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.)

  7. Spent fuel storage options: a critical appraisal

    International Nuclear Information System (INIS)

    Singh, K.P.; Bale, M.G.

    1990-01-01

    The delayed decisions on nuclear fuel reprocessing strategies in the USA and other countries have forced the development of new long-term irradiated fuel storage techniques, to allow a larger volume of fuel to be held on the nuclear station site after removal from the reactor. The nuclear power industry has responded to the challenge by developing several viable options for long-term onsite storage, which can be employed individually or in tandem. They are: densification of storage in the existing spent fuel pool; building another fuel pool facility at the plant site; onsite cask park, and on site vault clusters. Desirable attributes of a storage option are: Safety: minimise the number of fuel handling steps; Economy: minimise total installed, and O and M cost; Security: protection from anti-nuclear protesters; Site adaptability: available site space, earthquake characteristics of the region and so on; Non-intrusiveness: minimise required modifications to existing plant systems; Modularisation: afford the option to adapt a modular approach for staged capital outlays; and Maturity: extent of industry experience with the technology. A critical appraisal is made of each of the four aforementioned storage options in the light of these criteria. (2 figures, 1 table, 4 references) (Author)

  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. Investigations and Recommendations on the Use of Existing Experiments in Criticality Safety Analysis of Nuclear Fuel Cycle Facilities for Weapons-Grade Plutonium

    International Nuclear Information System (INIS)

    Rearden, B.T.

    2002-01-01

    report is given in Sect. 2. This report pertains to two of the five AOAs identified by the licensee [Duke, Cogema, Stone and Webster (DCS)] for the validation of criticality codes in the design of the Mixed-Oxide Fuel Fabrication Facility (MFFF). The five AOAs are as follows: (1) Pu-nitrate aqueous solutions (homogeneous systems), (2) Mixed-oxide (MOX) pellets, fuel rods and fuel assemblies (heterogeneous systems), (3) PuO 2 powders, (4) MOX powders, and (5) Aqueous solutions of Pu compounds (Pu-oxalate solutions). This report addresses a S/U analysis pertaining to AOA 3, PuO 2 powders, and AOA 4, MOX powders. AOA 3 and AOA 4 are the subject of this report since the other AOAs (solutions and heterogeneous systems) appear to be well represented in the documented benchmark experiments used in the criticality safety community. Prior to this work, DCS used traditional criticality validation techniques to identify numerous experimental benchmarks that are applicable to AOAs 3 and 4. Traditional techniques for selection of applicable benchmark experiments essentially consist of evaluating the area of applicability for important design parameters (e.g., Pu content or average neutron energy) and ensuring experiments have similar characteristics that bound or nearly bound the range of conditions requiring design analysis. DCS provided ORNL with compositions and dimensions for critical systems used to establish preliminary mass limits for facility powder and fuel pellet handling areas corresponding to AOAs 3 and 4. ORNL has reviewed existing critical experiments to identify those, which, in addition to those provided by DCS, may be applicable to the criticality code validation for AOAs 3 and 4. A S/U analysis was then performed to calculate the integral parameters used to determine the similarity of each critical experiment to each design system provided by DCS. This report contains a review of the S/U theory, a description of the design systems, a brief description of the

  10. Spent fuel packaging and its safety analysis

    International Nuclear Information System (INIS)

    Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

    1983-01-01

    An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

  11. 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)

  12. 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.)

  13. 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)

  14. 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)

  15. Nuclear criticality safety in Canada

    International Nuclear Information System (INIS)

    Shultz, K.R.

    1980-01-01

    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

  16. Criticality safety in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Krug, H.; Thomas, W.

    1980-01-01

    Criticality safety is of concern in processing, handling and storage of nuclear fuel outside reactors. Considerable experience has been gained in this area in the last twenty years. No criticality safety problems can be performed by applying a great variety of simple or sophisticated computation techniques. Calculated criticality data have been collected and issued in a handbook. A review is given of criticality safety philosophy, common practice in safety design, application of limitations and standards and accident analysis. As recent developments efforts have been made to investigate concrete reflection, the possible use of gadolinium as homogeneous poison, the suitability of hafnium as neutron absorber and nondestructive in-line measurement of fissile content concentrations. Recently a critical safe design has been established for a cast-iron cask for dry storage of spent fuel elements taking account of accidental moderation of the fuel. 8 figures

  17. 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.

  18. 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.

  19. Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi

    2001-01-01

    The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

  20. Criticality control during conditioning of spent nuclear fuel in the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Khalil, H.S.

    1994-01-01

    Spent nuclear fuel may be unacceptable for direct repository storage because of composition, enrichment, form, physical condition, or the presence of undesirable materials such as sodium. Fuel types which are not acceptable for direct storage must be processed or conditioned to produce physical forms which can safely be stored in a repository. One possible approach to conditioning is the pyroprocess implemented in the Fuel Cycle Facility (FCF) at Argonne National Laboratory-West. Conditioning of binary (U-Zr) and ternary (U-Pu-Zr) metallic fuels from the EBR-2 reactor is used to demonstrate the process. Criticality safety considerations limit batch sizes during the conditioning steps and provide one constraint on the final form of conditioned material. Criticality safety during conditioning is assured by the integration of criticality safety analysis, equipment design, process development, a measurement program, accountability procedures, and a computerized Mass Tracking System. Criticality issues related to storage and shipment of conditioned material have been examined

  1. 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

  2. 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)

  3. Lecture notes for criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. 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

  5. Lecture notes for criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. 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

  7. 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.

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-11

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

  10. University education and nuclear criticality safety professionals

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, R.E. [Safe Sites of Colorado, Golden, CO (United States); Stachowiak, R.V. [Kaiser-Hill Co., LLC, Golden, CO (United States); Knief, R.A. [Ogden Environmental and Energy Services, Albuquerque, NM (United States)

    1996-12-31

    The problem of developing a productive criticality safety specialist at a nuclear fuel facility has long been with us. The normal practice is to hire a recent undergraduate or graduate degree recipient and invest at least a decade in on-the-job training. In the early 1980s, the U.S. Department of Energy (DOE) developed a model intern program in an attempt to speed up the process. The program involved working at assigned projects for extended periods at a working critical mass laboratory, a methods development group, and a fuel cycle facility. This never gained support as it involved extended time away from the job. At the Rocky Flats Environmental Technology Site, the training method is currently the traditional one involving extensive experience. The flaw is that the criticality safety staff turnover has been such that few individuals continue for the decade some consider necessary for maturity in the discipline. To maintain quality evaluations and controls as well as interpretation decisions, extensive group review is used. This has proved costly to the site and professionally unsatisfying to the current staff. The site contractor has proposed a training program to remedy the basic problem.

  11. 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%

  12. Transport safety of irradiated fuel

    International Nuclear Information System (INIS)

    Rosa Giménez, R. de la

    2016-01-01

    The complication of the transport of spent fuel is significant not only because of the danger of the transported good itself but also for the size of the package. The number of this kind of expeditions are supposed to increase considerably in the coming years, for that reason is necessary for specialized companies such as ETSA be prepared. To this end, ETSA has already implemented most of the measures necessary to ensure safety - security of transport, not only during its execution but throughout the preparation.

  13. 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.)

  14. Spent Nuclear Fuel Project Safety Management Plan

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities

  15. 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?

  16. Summary of fuel safety research meeting 2004

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Hidaka, Akihide; Nakamura, Jinichi; Suzuki, Motoe; Nagase, Fumihisa; Sasajima, Hideo; Fujita, Misao; Otomo, Takashi; Kudo, Tamotsu; Amaya, Masaki; Sugiyama, Tomoyuki; Ikehata, Hisashi; Iwasaki, Ryo; Ozawa, Masaaki; Kida, Mitsuko

    2004-10-01

    Fuel Safety Research Meeting 2004, which was organized by the Japan Atomic Energy Research Institute, was held on March 1-2, 2004 at Toranomon Pastoral, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meeting covered the status of fuel safety research activities, fuel behavior under RIA and LOCA conditions, high burnup fuel behavior, and radionuclides release under severe accident conditions. This summary contains all the abstracts and OHP sheets presented in the meeting. (author)

  17. 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

  18. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  19. 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)

  20. Proceedings of fuel safety research specialists' meeting

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2002-08-01

    Fuel Safety Research Specialists' Meeting, which was organized by Japan Atomic Energy Research Institute, was held on March 4-5, 2002 at JAERI in Tokai Establishment. Purposes of the Meeting are to exchange information and views on LWR fuel safety topics among the specialist participants from domestic and foreign organizations, and to discuss the recent and future fuel research activities in JAERI. In the Meeting, presentations were given and discussions were made on general report of fuel safety research activities, fuel behaviors in normal operation and accident conditions, FP release behaviors in severe accident conditions, and JAERI's ''Advanced LWR Fuel Performance and Safety Research Program''. A poster exhibition was also carried out. The Meeting significantly contributed to planning future program and cooperation in fuel research. This proceeding integrates all the pictures and papers presented in the Meeting. The 10 of the presented papers are indexed individually. (J.P.N.)

  1. Hydrogen and Gaseous Fuel Safety and Toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader; J. Sephen Herring

    2007-06-01

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

  2. 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)

  3. Issues related to criticality safety analysis for burnup credit applications

    International Nuclear Information System (INIS)

    DeHart, M.D.; Parks, C.V.

    1995-01-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design

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

    International Nuclear Information System (INIS)

    DUNCAN, D.R.

    1999-01-01

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

  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. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    The nuclear fuel cycle covers the procurement and preparation of fuel for nuclear power reactors, its recovery and recycling after use and the safe storage of all wastes generated through these operations. The facilities associated with these activities have an extensive and well documented safety record accumulated over the past 40 years by technical experts and safety authorities. This report constitutes an up-to-date analysis of the safety of the nuclear fuel cycle, based on the available experience in OECD countries. It addresses the technical aspects of fuel cycle operations, provides information on operating practices and looks ahead to future activities

  7. Safety-Critical Java for Embedded Systems

    DEFF Research Database (Denmark)

    Rios Rivas, Juan Ricardo

    Safety-critical systems are real-time systems whose failure can have severe or catastrophic consequences, possibly endangering human life. Many safety-critical systems incorporate embedded computers used to control different tasks. Software running on safety-critical systems needs to be certified...... before its deployment and the most time-consuming step of this process is the testing and verification phase. Due to the increasing complexity in safety-critical systems there is a need for new technologies that can facilitate testing and verification activities. The safety-critical specification...... 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...

  8. 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

  9. TAPS safety evaluation criteria for reload fueling

    International Nuclear Information System (INIS)

    Mahendra Nath; Veeraraghavan, N.

    1976-01-01

    To improve operating performance of Tarapur reactors, several proposals are under consideration such as core expansion, change-over to an improved fuel design with lower heat rating, extension of fuel cycle lengths etc., which have a bearing on overall plant operating characteristics and reactor safety. For evaluating safety implications of the various proposals, it is necessary to formulate safety evaluation criteria for reload fuelling. Salient features of these criteria are discussed. (author)

  10. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  11. 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)

  12. 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...

  13. Proceedings of the Nuclear Criticality Technology Safety

    International Nuclear Information System (INIS)

    Sanchez, Renee G.

    1998-01-01

    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

  14. 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)

  15. 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)

  16. Safety and Licensing of Spent Fuel Storage and Transport — Safety Issues Within Spent Fuel Transport

    International Nuclear Information System (INIS)

    Brut, S.; Derlot, F.; Milet, L.

    2015-01-01

    We can consider the different safety issues within French fuel transport as follows: (a) the proof as regards the leaking fuel assembly transport with hydrogen generation coming from potential in leakage water inside fuel rods; ( b) the measures taken to enforce the new design as well as the new manufacturing which have been decided since January 1 st 2007 in the frame of the 96 IAEA Regulation as regards the full water penetration as compared to the 85 IAEA Regulation, the latter allowing partial water penetration on certain conditions; and (c) the obligation of implementing various risk controls on exploitation site in order to take into account the possible human failure which are intrinsically increasing the permissible doses rates for workers. Even quite recently the leaking fuel assembly transport has been considered with no specific measure as regards the radiolysis phenomenon or the quality of drying cask holds. All these measures were sufficiently in accordance to rule out this issue. Lately, the leaking fuel assembly transport needs the implementation of equipment controls involved in nuclear power plants as regards the hydrogen rate before loading departure in order to determine on the evolution law, the maximum duration authorized for the transportation to not exceed the lower limit of inflammable status. As regards the proof of the criticality-safety casks, the main justification to be held on the irradiated fuel assembly on drop accident conditions could find a key in the hypothesis of the important damage of the fuel but should be in this matter, compensated by a limit of containment penetration for safety reason. For this case, the application of the 96 IAEA Regulation involves the use of independent leak tightness barriers. TN International is introducing different examples in France linked to the selection of multiple barriers. When limited in leakage quantity of water inside the cask is considered for the criticality studies, the French Competent

  17. 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

  18. 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.

  19. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

  20. Safety evaluation of a conceptual fuel recycle complex

    International Nuclear Information System (INIS)

    Hodges, M.E.

    1980-01-01

    A conceptual design integration study for an integrated Fuel Recycle Complex (FRC) has been completed. A safety evaluation of the radiation shielding, fire precautions, handling of nonradioactive hazardous materials, criticality hazards, operating errors, and the influence of natural phenomena on the FRC shows that all federal regulations are met or exceeded

  1. 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.

  2. Summary of fuel safety research meeting 2005

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Nakamura, Takehiko; Nagase, Fumihisa; Nakamura, Jinichi; Suzuki, Motoe; Sasajima, Hideo; Sugiyama, Tomoyuki; Amaya, Masaki; Kudo, Tamotsu; Chuto, Toshinori; Tomiyasu, Kunihiko; Udagawa, Yutaka; Ikehata, Hisashi; Kida, Mitsuko; Ikatsu, Nobuhiko; Hosoyamada, Ryuji; Hamanishi, Eizou; Iwasaki, Ryo; Ozawa, Masaaki

    2006-03-01

    Fuel Safety Research Meeting 2005, which was organized by the Japan Atomic Energy Agency (Establishment of the new organization in Oct. 1, 2005 integrated of JAERI and JNC) was held on March 2-3, 2005 at Toshi Center Hotel, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meting covered the status of fuel safety research activities, fuel behavior under Reactivity Initiated Accident (RIA) and Loss of coolant accident (LOCA) conditions, high fuel behavior, and radionuclide release under severe accident conditions. This summary contains all the abstracts and sheets of viewgraph presented in the meeting. (author)

  3. Criticality and Its Uncertainty Analysis of Spent Fuel Storage Rack for Research Reactor

    International Nuclear Information System (INIS)

    Han, Tae Young; Park, Chang Je; Lee, Byung Chul

    2011-01-01

    For evaluating the criticality safety of spent fuel storage rack in an open pool type research reactor, a permissible upper limit of criticality should be determined. It can be estimated from the criticality upper limit presented by the regulatory guide and an uncertainty of criticality calculation. In this paper, criticalities for spent fuel storage rack are carried out at various conditions. The calculation uncertainty of MCNP system is evaluated from the calculation results for the benchmark experiments. Then, the upper limit of criticality is determined from the uncertainties and the calculated criticality of the spent fuel storage rack is evaluated

  4. 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)

  5. Safety assessment for Dragon fuel element production

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1963-11-01

    This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

  6. Recent development in safety regulation of nuclear fuel cycle activities

    International Nuclear Information System (INIS)

    Kato, S.

    2001-01-01

    Through the effort of deliberation and legislation over five years, Japanese government structure was reformed this January, with the aim of realizing simple, efficient and transparent administration. Under the reform, the Agency for Nuclear and Industrial Safety (ANIS) was founded in the Ministry of Economy, Trade and Industry (METI) to be responsible for safety regulation of energy-related nuclear activities, including nuclear fuel cycle activities, and industrial activities, including explosives, high-pressure gasses and mining. As one of the lessons learned from the JCO criticality accident of September 1999, it was pointed out that the government's inspection function was not enough for fuel fabrication facilities. Accordingly, new statutory regulatory activities were introduced, namely, inspection of observance of safety rules and procedures for all kinds of nuclear operators and periodic inspection of fuel fabrication facilities. In addition, in order to cope with insufficient safety education and training of workers in nuclear facilities, licensees of nuclear facilities are required by law to specify safety education and training for their workers. ANIS is committed to enforce these new regulatory activities effectively and efficiently. In addition, it is going to be prepared, in its capacity as safety regulatory authority, for future development of Japanese fuel cycle activities, including commissioning of JNFL Rokkasho reprocessing plant and possible application for licenses for JNFL MOX fabrication plant and for spent fuel interim storage facilities. (author)

  7. Recent development in safety regulation of nuclear fuel cycle activities

    International Nuclear Information System (INIS)

    Kato, S.

    2002-01-01

    Through the effort of deliberation and legislation over five years, Japanese government structure was reformed this January, with the aim of realizing simple, efficient and transparent administration. Under the reform, the Agency for Nuclear and Industrial Safety (ANIS) was founded in the Ministry of Economy, Trade and Industry (METI) to be responsible for safety regulation of energy-related nuclear activities, including nuclear fuel cycle activities, and industrial activities, including explosives, high-pressure gasses and mining. As one of the lessons learned from the JCO criticality accident of September 1999, it was pointed out that government's inspection function was not enough for fuel fabrication facilities. Accordingly, new statutory regulatory activities were introduced, namely, inspection of observance of safety rules and procedures for all kinds of nuclear operators and periodic inspection of fuel fabrication facilities. In addition, in order to cope with insufficient safety education and training of workers in nuclear facilities, licensees of nuclear facilities are required by law to specify safety education and training for their workers. ANIS is committed to enforce these new regulatory activities effectively and efficiently. In addition, it is going to be prepared for, in its capacity of safety regulatory authority, future development of Japanese fuel cycle activities, including commissioning of JNFL Rokkasho reprocessing plant and possible application for licenses for JNFL MOX fabrication plant and for spent fuel interim storage facilities. (author)

  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. Safety recycling of reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Weinlaender, W.

    Additionally to the measures, which descent from the conventional safety techniques, a series of supplementary protective measures have to be taken in connection with the Atomic Energy Law, the Radiation Protection Ordinance and the nuclear-technical practice, which in particular guarantee a safe enclosure and a safe residual heat rejection of the handled radioactive material and an avoidance of nuclear chain reactions. The most important plant malfunctions to be considered within the scope of the plant safety control according to the atomic law are, the radioactivity release due to mechanical damage of fuel elements, containment leakage, explosions in process equipment and/or vessels, burning of run out organic solvents, criticality malfunctions, and the already mentioned accidental failure of after-heat removal. If we let alone the extremely low probabilities for the occurrence of such accidents due to the selected methods, the layout of the equipment and by taking the required quality warranty measures into consideration, and infer such accidents in spite of this, the resulting radiation doses outside the plant are in all cases much lower than 5 rem, which is the design limit according to the regulations for radiation protection. (orig./HP) [de

  10. 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.

  11. Review of fuel safety criteria in France

    Energy Technology Data Exchange (ETDEWEB)

    Boutin, Sandrine; Graff, Stephanie; Foucher-Taisne, Aude; Dubois, Olivier [Institut de Radioprotection et du Surete Nucleaire, Fontenay-aux-Roses (France)

    2018-01-15

    Fuel safety criteria for the first barrier, based on state-of-the-art at the time, were first defined in the 1970s and came from the United States, when the French nuclear program was initiated. Since then, there has been continuous progress in knowledge and in collecting experimental results thanks to the experiments carried out by utilities and research institutes, to the operating experience, as well as to the generic R and D programs, which aim notably at improving computation methodologies, especially in Reactivity-Initiated accident and Loss-of-Coolant Accident conditions. In this context, the French utility EDF proposed new fuel safety criteria, or reviewed and completed existing safety demonstration covering the normal operating, incidental and accidental conditions of Pressurised Water Reactors. IRSN assessed EDF's proposals and presented its conclusions to the Advisory Committee for Reactors Safety of the Nuclear Safety Authority in June 2017. This review focused on the relevance of historical limit values or parameters of fuel safety criteria and their adequacy with the state-of-the-art concerning fuel physical phenomena (e.g. Pellet-Cladding Mechanical Interaction in incidental conditions, clad embrittlement due to high temperature oxidation in accidental conditions, clad ballooning and burst during boiling crisis and fuel melting).

  12. 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...

  13. 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...

  14. 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

  15. Criticality safety calculations for three types of final disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    2005-07-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent nuclear fuel has been studied with the MCNP4C Monte Carlo code. Three types of spent fuel disposal canisters have been analysed. The differences between the canisters result from the size and geometry of the spent fuel assemblies to be disposed of in them. One canister type has been designed to contain 12 hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant ('VVER canister'). The second type is for 12 square BWR fuel bundles used at the Olkiluoto 1 and 2 units ( B WR canister ) and the third type is for four fuel assemblies of the Olkiluoto 3 unit to be constructed in the near future ( E PR canister ) . Each canister type is of similar size in the radial direction, but the axial lengths vary significantly. A spent fuel disposal canister must meet the normal criticality safety criteria. The effective multiplication factor must be less than 0.95 also when the canister is in the most reactive credible configuration (optimum moderation and close reflection). Uncertainties in the calculation methods may necessitate the use of an even lower reactivity limit. However, no systematic uncertainty analysis was carried out during this study. It has been proved in an earlier study that a version of the VVER canister loaded with twelve similar fresh VVER-440 assemblies with the initial enrichment of 4.2% fulfils the criticality safety criteria. Also an earlier design of the BWR canister loaded with twelve fresh BWR assemblies of so-called ATRIUM 10x10-9Q type with the initial enrichment of 3.8% and without burnable absorbers has been proved to meet the safety criteria. Therefore, in this study only a few calculations have been carried out for the present versions of VVER and BWR canisters and the results are in good agreement with the previous ones. The main emphasis of this study has been on the EPR canister. This new canister type fulfils the

  16. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  17. 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

  18. 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

  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 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

  1. 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.)

  2. 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

  3. Nuclear Criticality Safety Handbook (English translation)

    International Nuclear Information System (INIS)

    1995-09-01

    This is an English translation of 'Nuclear Criticality Safety Handbook' compiled by Nuclear Materials Regulation Division in Nuclear Safety Bureau of Science and Technology Agency of Japan and published by Nikkan Shobo in 1988, which was originally written in Japanese. (author)

  4. 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

  5. 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

  6. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  7. 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

  8. Safety features of subcritical fluid fueled systems

    Energy Technology Data Exchange (ETDEWEB)

    Bell, C.R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

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

  9. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, Charles R.

    1995-01-01

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

  10. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1994-01-01

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

  11. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  12. Safety of fuel cycle facilities. Topical issues paper no. 3

    International Nuclear Information System (INIS)

    Ranguelova, V.; Niehaus, F.; Delattre, D.

    2001-01-01

    A wide range of nuclear fuel cycle facilities are in operation. These installations process, use, store and dispose of radioactive material and cover: mining and milling, conversion, enrichment, fuel fabrication (including mixed oxide fuel), reactor, interim spent fuel storage, reprocessing, waste treatment and waste disposal facilities. For the purposes of this paper, reactors and waste disposal facilities are not considered. The term 'fuel cycle facilities' covers only the remainder of the installations listed above. The IAEA Secretariat maintains a database of fuel cycle facilities in its Member States. Known as the Nuclear Fuel Cycle Information System (NFCIS), it is available as an on-line service through the Internet. More than 500 such facilities have been reported under this system. The facilities are listed by facility type and operating status. Approximately one third of all of the facilities are located in developing States. About half of all facilities are reported to be operating, of which approximately 40% are operating in developing States. In addition, some 60 facilities are either in the design stage or under construction. Although the radioactive source term for most fuel cycle facilities is lower than the source term for reactors, which results in less severe consequences to the public from potential accidents at these fuel cycle installations, recent events at some fuel cycle facilities have given rise to public concern which has to be addressed adequately by national regulatory bodies and at the international level. Worldwide, operational experience feedback warrants improvements in the safety of these facilities. Some of the hazards are similar for reactor and non-reactor facilities. However, the differences between these installations give rise to specific safety concerns at fuel cycle facilities. In particular, these concerns include: criticality, radiation protection of workers, chemical hazards, fire and explosion hazards. It is recognized

  13. Chemical process safety at fuel cycle facilities

    International Nuclear Information System (INIS)

    Ayres, D.A.

    1997-08-01

    This NUREG provides broad guidance on chemical safety issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not exhaustive, for addressing chemical process safety in the safe storage, handling, and processing of licensed nuclear material. It expounds to license holders and applicants a general philosophy of the role of chemical process safety with respect to NRC-licensed materials; sets forth the basic information needed to properly evaluate chemical process safety; and describes plausible methods of identifying and evaluating chemical hazards and assessing the adequacy of the chemical safety of the proposed equipment and facilities. Examples of equipment and methods commonly used to prevent and/or mitigate the consequences of chemical incidents are discussed in this document

  14. Use of PCs and workstations for criticality safety analysis

    International Nuclear Information System (INIS)

    Watmough, M.H.; Evans, A.M.; Smith, N.R.

    1993-01-01

    Making effective use of the best available computer technology has long been a feature of U.K. criticality safety analyses. In recent years, the industry's drive for improvements in design safety assessment practice has been supplemented by increasing commercial awareness, bringing further emphasis to the consideration of cost-effectiveness in criticality computation. Consequently, there has been an evolution of computing facilities seen by U.K. criticality assessors from the mainframe terminals of the 1970s through minicomputers in the 1980s to graphical workstations and, most recently, personal computers (PCs) in the 1990s. These moves have been initiated by the availability of hardware capable of providing adequate performance and facilitated by cooperation and subsequently formal collaboration between British Nuclear Fuels Limited (BNFL) and the Atomic Energy Authority (AEA) in the field of software development

  15. 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

  16. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. 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)

  18. Fuel elements and safety engineering goals

    International Nuclear Information System (INIS)

    Schulten, R.; Bonnenberg, H.

    1990-01-01

    There are good prospects for silicon carbide anti-corrosion coatings on fuel elements to be realised, which opens up the chance to reduce the safety engineering requirements to the suitable design and safe performance of the ceramic fuel element. Another possibility offered is combined-cycle operation with high efficiencies, and thus good economic prospects, as with this design concept combining gas and steam turbines, air ingress due to turbine malfunction is an incident that can be managed by the system. This development will allow economically efficient operation also of nuclear power reactors with relatively small output, and hence contribute to reducing CO 2 emissions. (orig./DG) [de

  19. 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...

  20. 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

  1. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    The purpose of this design calculation is to demonstrate that the handling operations of transportation casks performed in the Transportation Cask Receipt and Return Facility (TCRRF) and Buffer Area meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC [Bechtel SAIC Company] 2004 [DIRS 171599], Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''Transportation Cask Receipt/Return Facility Description Document'' (BSC 2004 [DIRS 170217], Section 3.2.3). Specific scope of work contained in this activity consists of the following items: (1) Evaluate criticality effects for both dry and fully flooded conditions pertaining to TCRRF and Buffer Area operations for defense in depth. (2) Evaluate Category 1 and 2 event sequences for the TCRRF as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). This evaluation includes credible fuel reconfiguration conditions. In addition to the scope of work listed above, an evaluation was also performed of modeling assumptions for commercial spent nuclear fuel (CSNF) regarding inclusion of plenum and end regions of the active fuel. This calculation is limited to CSNF and US Department of Energy (DOE) SNF. it should be mentioned that the latter waste form is evaluated more in depth in the ''Canister Handling Facility Criticality Safety Calculations (BSC 2004 [DIRS 167614]). Further, the design and safety analyses of the naval SNF canisters are the responsibility of the US 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 TCRRF and Buffer Area and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will have little or no impact on the criticality results and/or conclusions

  2. 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.

  3. OECD/NEA working party on nuclear criticality safety: Challenge of new realities

    International Nuclear Information System (INIS)

    Nomura, Y.; Brady, M.C.; Briggs, J.B.; Sartori, E.

    1998-01-01

    New issues in criticality safety continue to emerge as spent fuel storage facilities reach the saturation point, fuel enrichments and burn-ups increase and new types of plutonium-carrying fuels are being developed. The new challenges related to the manipulation, transportation and storage of fuel demand further work to improve models predicting behavior through new experiments, especially where there is a lack of data in the present databases. This article summarizes the activities of the OECD/NEA working groups that coordinate and carry out work in the domain of criticality safety. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burn-up credit. This is aimed toward improving safety and identifying economic solutions to issues concerning the back end of the fuel cycle

  4. OECD/NEA working party on nuclear criticality safety: challenge of new realities

    International Nuclear Information System (INIS)

    Nomura, Y.; Brady, M.C.; Briggs, J.B.; Sartori, E.

    1998-01-01

    New issues in critically safety continue to emerge as spent fuel storage facilities reach the saturation point, fuel enrichments and burn-ups increase and new types of plutonium-carrying fuels are being developed. The new challenges related to the manipulation, transportation and storage of fuel demand further work to improve models predicting behaviour through new experiments, especially where there is a lack of data the present databases. This article summarizes the activities of the OECD/NEA working groups that co-ordinate and carry out work in the domain of criticality safety. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burn-up credit. This is aimed toward improving safety and identifying economic solutions to issues concerning the back end of the fuel cycle. (authors)

  5. 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.

  6. 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

  7. 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

  8. Progress of criticality experiments and nuclear design studies on Erbia-bearing super high burnup fuel

    International Nuclear Information System (INIS)

    Yamasaki, Masatoshi; Kuroishi, Takeshi; Takeda, Toshikazu; Yamamoto, Akio; Unesaki, Hironobu; Sano, Tadafumi; Mori, Masaaki

    2009-01-01

    To cope with the '5wt% enrichment barrier', the development project on Erbia bearing super high-burnup fuel with high uranium enrichment, such as critical experiments with fully Erbia loaded core in KUCA, criticality safety analyses against simple shapes, physicochemical properties measurement and so on, is under going. Following the three experiments of fully-Er-loaded core, the result of the fourth core of higher content of Erbia is presented. As a second topic, fuel design using isotopically modified Erbia, such as a) Er-166 removed from Erbium content, b) Only Er-167 used, is shown. In use of isotopically modified Erbia, the ECOS diagram, which shows minimum Erbia contents that should be mixed into above 5wt% enriched fuel, so that the criticality safety will be equivalent to that with the conventional below 5wt% enriched fuel, should be modified in order to correspond to various Erbia isotope vectors. (author)

  9. 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.

  10. Neutron nuclear data measurements for criticality safety

    Science.gov (United States)

    Guber, Klaus; Paradela, Carlos; Heyse, Jan; Kopecky, Stefan; Schillebeeckx, Peter; Siegler, Peter

    2017-09-01

    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.

  11. 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)

  12. Present status of International Criticality Safety Benchmark Evaluation Project (ICSBEP)

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2000-01-01

    The International Criticality Safety Evaluation Project, ICSBEP was designed to identify and evaluate a comprehensive set of critical experiment benchmark data. Compilation of the data into a standardized format are made by reviewing original and subsequently revised documentation for calculating each experiment with standard criticality safety codes. Five handbooks of evaluated criticality safety benchmark experiments have been published since 1995. (author)

  13. New approach to managing nuclear criticality risk at Nuclear Fuel Services, Inc

    International Nuclear Information System (INIS)

    Green, R.; Droke, R.; Paine, D.

    1992-01-01

    The negative aspects of having a nuclear criticality accident at a fuel fabrication facility have substantially increased in recent years. Although ensuring that the facility is designed and operating in a critically safe manner is a high management priority, practices of managing the risk of a criticality accident have not significantly changed. The method of evaluating risk based on quantitative analysis can enable management to adapt to the increased consequences of a nuclear criticality accident. Additional controls may be placed on high-risk areas within a facility to ensure safe operation of the plant. Areas where controls are in place that impede the productivity of the facility and have negligible impact on criticality safety may be removed or replaced. Management can also streamline the safety analysis efforts applied to facility upgrades by demonstrating that proposed design changes do not compromise criticality safety. Future expansion of quantitative analysis techniques will also allow risk-based management decisions on industrial, radiological, and environmental safety

  14. A desktop 3D printer in safety-critical Java

    DEFF Research Database (Denmark)

    Strøm, Tórur Biskopstø; Schoeberl, Martin

    2012-01-01

    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...... 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...

  15. 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)

  16. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  17. 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

  18. CRISTAL: a french criticality code package to assess nuclear installation criticality safety

    International Nuclear Information System (INIS)

    Michel, E Doucet; Comte, N.; Jean Michel Gomit; Eric Letang; Chrsitine Magnaud; Christophe Venard; Herve Toubon; Cogema

    2005-01-01

    For more than thirty years, CEA, IRSN and the French nuclear industry have been combining their efforts to finance, develop and validate computer codes to assess the criticality safety concerns of nuclear installations, transport casks, and reprocessing facilities. As one of the major world fuel vendors, Framatome ANP is deeply involved in defining code developments which incorporate feedback from both users and customers. The result of these continuous efforts is the evolutionary CRISTAL code. The CRISTAL package was developed as an easy-to-use system using cross-section libraries (JEF 2.2 and CEA93), well-established computer codes (APOLLO2, MORET 4 and TRIPOLI-4) and including a Graphical User-Friendly Interface. The APOLLO2 computer code, a spectral code used for evaluating the basic characteristics of fuel assemblies, has been upgraded to perform criticality safety calculations. The MORET 4 computer code is a neutron simulation code in three dimensions which uses the multigroup formalism for cross-sections and the Monte Carlo method to solve the Boltzmann equation. Through the years, the CRISTAL package has been improved to take into account both the growth of its validation database and the increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating the comprehensive APOLLO2 and MORET 4 computer codes; CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and on the implementation of user-friendly graphical interfaces. Thanks to its broad validation database, CRISTAL V0 provides outstanding accuracy of criticality evaluation for configurations covering the entire fuel cycle life (i.e. from fuel enrichment, pellet/assembly fabrication and transport casks to fuel reprocessing). With more than a thousand benchmark/calculation comparisons, uncertainties can be deduced for various file media, fissile shapes, fissile process interactions, neutron-poisoning screens and material

  19. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  20. 14 CFR 417.121 - Safety critical preflight operations.

    Science.gov (United States)

    2010-01-01

    ... the launch of an unguided suborbital rocket, in addition to meeting the other requirements of this... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION LICENSING LAUNCH SAFETY Launch Safety Responsibilities § 417.121 Safety critical preflight operations. (a) General. A launch operator must perform safety critical...

  1. Development of the DOE Nuclear Criticality Safety Program Web site for the nuclear criticality safety professional

    International Nuclear Information System (INIS)

    Lee, C.K.; Huang, S.; Morman, J.A.; Garcia, A.S.

    2000-01-01

    Development of the US Department of Energy (DOE) Nuclear Criticality Safety (NCS) Program (NCSP) Web site is the result of the efforts of many members of the NCS community and is maintained by Lawrence Livermore National Laboratory (LLNL) under the direction of the NCSP Management Team. This World-Wide-Web resource was developed as part of the DOE response to Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. The NCSP Web site provides information of interest to NCS professionals and includes links to other sites actively involved in the collection and dissemination of criticality safety information. To the extent possible, the hyperlinks on this Web site direct the user to the original source of the referenced material to ensure access to the latest, most accurate version. This site is intended to provide a central location for access to relevant NCS information in a user-friendly environment for the criticality safety community

  2. TRIGA R a widely used fuel - Assessments, validation and application of criticality standards

    International Nuclear Information System (INIS)

    Comte, N.; Thievenaz, B.; Doucet, M.; Torlini, E.; Pin, T.

    2009-01-01

    The TRIGA R (Training, Research, Isotope production, General Atomics), reactor is the most widely used non-power nuclear reactor in the world. Sixty six TRIGA R reactors are running or under construction in 24 countries. These reactors are used in many diverse applications, including production of radioisotopes for medicine and industry, treatment of tumors, non destructive testing, basic research on the properties of matter, and for education and training. TRIGA R International, a joint venture company between General Atomics and CERCA, a wholly-owned subsidiary of AREVA NP, manufactures and sells TRIGA R fuel to research reactor. The CERCA Factory is currently the only Company in the world to manufacture this type of fuel. Criticality safety evaluations implement requirements to justify sufficient sub critical margins for fuel fabrication plants. Basic criticality data (i. e., criticality standards) are used to determine the sub critical margins for all processes involving enriched uranium. Safety analysis help to identify and update those standards to bound normal, abnormal and accidental conditions. The French criticality code package CRISTAL V1 was used to perform the calculations. After a brief description of the fuel design and the calculation scheme, this paper presents the results for the different types and media of TRIGA R fuel (powder, elements). Experimental benchmarks data were used for validation as well as comparative calculations with Monte-Carlo codes MORET 4 and MCNP. The criticality section of the safety analysis report supporting the actual TRIGA workshops was reviewed to ensure the update of the criticality standards does not harm the safety of the installations. (authors)

  3. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

  4. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    International Nuclear Information System (INIS)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae

    2016-01-01

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed

  5. Critical review of nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kuster, N.

    1996-01-01

    Transmutation of long-lived radionuclides is considered as an alternative to the in-depth disposal of spent nuclear fuel, in particular, on the final stage of the nuclear fuel cycle. The majority of conclusions is the result of the common work of the Karlsruhe FZK and the Commissariat on nuclear energy of France (CEA)

  6. Estimation of the Fuel Depletion Code Bias and Uncertainty in Burnup-Credit Criticality Analysis

    International Nuclear Information System (INIS)

    Kim, Jong Woon; Cho, Nam Zin; Lee, Sang Jin; Bae, Chang Yeal

    2006-01-01

    In the past, criticality safety analyses for commercial light-water-reactor (LWR) spent nuclear fuel (SNF) storage and transportation canisters assumed the spent fuel to be fresh (unirradiated) fuel with uniform isotopic compositions. This fresh-fuel assumption provides a well-defined, bounding approach to the criticality safety analysis that eliminates concerns related to the fuel operating history, and thus considerably simplifies the safety analysis. However, because this assumption ignores the inherent decrease in reactivity as a result of irradiation, it is very conservative. The concept of taking credit for the reduction in reactivity due to fuel burnup is commonly referred to as burnup credit. Implementation of burnup credit requires the computational prediction of the nuclide inventories (compositions) for the dominant fissile and absorbing nuclide species in spent fuel. In addition to that, the bias and uncertainty in the predicted concentration of all nuclides used in the analysis be established by comparisons of calculated and measured radiochemical assay data. In this paper, three methods for considering the bias and uncertainty will be reviewed. The estimated bias and uncertainty that the results of 3rd method are presented

  7. Feasibility of performing criticality experiments with spent LWR fuel

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1989-01-01

    Considering fuel operating histories, i.e., burnup, in criticality analyses and control involves two distinctly different issues. One of these issues is concerned with the design and fabrication of fuel shipping and storage systems. The other issue is concerned with the verification of actual operating practices in the field. Both design and verification activities would benefit from physical measurement data as opposed to relying on administrative controls and calculations; however, it is the authors intention to address only the feasibility of obtaining experimental spent fuel criticality data in support of the design effort. The author discusses the background, obstacles, feasibility conclusions reached when applying burnup concepts to spent fuels

  8. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1984-01-01

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  9. Nuclear Fuel Safety Criteria Technical Review - Second edition

    International Nuclear Information System (INIS)

    Beck, Winfried; Blanpain, Patrick; Fuketa, Toyoshi; Gorzel, Andreas; Hozer, Zoltan; Kamimura, Katsuichiro; Koo, Yang-Hyun; Maertens, Dietmar; Nechaeva, Olga; Petit, Marc; Rehacek, Radomir; Rey-Gayo, Jose Maria; Sairanen, Risto; Sonnenburg, Heinz-Guenther; Valach, Mojmir; Waeckel, Nicolas; Yueh, Ken; Zhang, Jinzhao; Voglewede, John

    2012-01-01

    Most of the current nuclear fuel safety criteria were established during the 1960's and early 1970's. Although these criteria were validated against experiments with fuel designs available at that time, a number of tests were based on unirradiated fuels. Additional verification was performed as these designs evolved, but mostly with the aim of showing that the new designs adequately complied with existing criteria, and not to establish new limits. In 1996, the OECD Nuclear Energy Agency (NEA) reviewed existing fuel safety criteria, focusing on new fuel and core designs, new cladding materials and industry manufacturing processes. The results were published in the Nuclear Fuel Safety Criteria Technical Review of 2001. The NEA has since re-examined the criteria. A brief description of each criterion and its rationale are presented in this second edition, which will be of interest to both regulators and industry (fuel vendors, utilities)

  10. Criticality analysis of PWR spent fuel storage facilities inside nuclear power plants

    International Nuclear Information System (INIS)

    Neuber, J.C.

    1999-01-01

    This paper describes some of the main features of the actinide plus fission product burnup credit methodology used by Siemens for criticality safety design analysis of wet PWR storage pools with soluble boron in the pool water. Application of burnup credit requires knowledge of the isotopic inventory of the irradiated fuel for which burnup credit is taken. This knowledge is gained by using depletion codes. The results of the depletion analysis are a necessary input to the criticality analysis. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the Siemens standard design procedure SAV90. The quality of this procedure relies on statistics on the differences between calculation and measurement extracted from in-core measurement data and chemical assay data. Siemens performs criticality safety calculations with the aid of the criticality calculation modules of the SCALE code package. These modules are verified many times with the aid of various kinds of critical experiments and configurations: Application of these modules to spent LWR fuel assembly storage pools was verified by analyzing critical experiments simulating such storage pools. Actinide plus fission product burnup credit applications of these modules were verified by analyzing PWR reactor critical configurations. The result of performing a burnup credit analysis is the determination of a burnup, credit loading curve for the spent fuel storage racks designed for burnup credit. This curve specifies the loading criterion by indicating the minimum burnup necessary for the fuel assembly with a specific initial enrichment to be placed in the storage racks designed for burnup credit. The loading of the spent fuel storage racks designed for burnup credit requires the implementation of controls to ensure that the loading curve is met. The controls include the determination of fuel assembly burnup based on reactor records. (author)

  11. 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.

  12. 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.

  13. 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...... are thread private. Furthermore, the notation of a maximum backing store requirement enables implementation of the scoped memories without fragmentation issues. In this paper we explore the implications of this new scoped memory model and possible simplifications in the implementation. It is possible...... to unify the three memory area types and provide a single class to represent all three memory areas of SCJ. The knowledge of the maximum storage requirements allows using nested backing stores in the implementation of the memory area representation. The proposed design of an SCJ compliant scope...

  14. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  15. The criticality implications of taking credit for fuel burn-up

    International Nuclear Information System (INIS)

    Clemson, P.D.; Thorne, P.R.

    1989-01-01

    The first problem usually facing a criticality engineer is data specification. The design engineer and operator need a design and operating envelope of data which allows practical operation, but the criticality engineer must ensure that key parameters remain within specific limits to maintain criticality safety. Given the high safety standards demanded in the nuclear industry, a naturally cautious view is taken and pessimistic assumptions of parameter values are therefore made. This explains why assessments of spent fuel transport and storage assume unirradiated fissile compositions (the fresh fuel assumption). Looking ahead, the pressure to achieve more economic designs and the need to extend design specifications is likely to increase. The large decrease in spent fuel reactivity resulting from fuel burn-up provides a potential way of meeting these requirements. The challenge is to develop ways of taking credit for fuel burn-up with no reduction in safety. This raises various issues which are discussed briefly in this paper, together with a scoping study to investigate the scale of reactivity change with burn-up for an existing BNFL cask design

  16. Quality and ecological safety of motor fuels

    OpenAIRE

    Boychenko, Sergey; Vovk, Oksana; Chernyak, Larisa; Akinina, Ksenya

    2007-01-01

    У статті розглянуто взаємозвязок якості палив для транспортних засобів та екологічної безпеки. Запропоновано головні напрямки вирішення эколого- енергетичних проблем системи "чоловік – навколишнє середовище – паливо – транспортний засіб". Article considering interconnection of fuel quality for transport and ecological safety. Main directions were offered of solving ecological-energy problem system "human-environment-fuel-transport"....

  17. Criticality implications of extended fuel burnup

    International Nuclear Information System (INIS)

    Eng, R.

    1985-01-01

    The advantages (and disadvantages) of extended fuel burnup to the operating nuclear utility are well documented. Niagara Mohawk found that increasing the refueling interval from 12 months to 24 months led to a 60% increase in full power days per refueling outage day. Today's incentives include decreasing spent fuel accumulation, increasing plant capacity factors through longer cycles, reducing uranium ore requirements, reducing radiation exposures of workers during refueling, reduced disposal requirements, reduced number of heat-up and cooldown transients, reduced plant security burden (during refueling outages), and reduced manpower time for regulatory review

  18. 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

  19. 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

  20. Prevention of criticality accidents. Fuel elements storage

    International Nuclear Information System (INIS)

    Canavese, S.I.; Capadona, N.M.

    1990-01-01

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author) [es

  1. Safety assessment of ammonia as a transport fuel

    DEFF Research Database (Denmark)

    Duijm, N.J.; Markert, Frank; Paulsen, Jette Lundtang

    2005-01-01

    of transport of ammonia to the refuelling stations and safety of the activities at the refuelling station (unloading and refuelling). Comparisons are made between the safety of using ammonia and the safety of otherexisting or alternative fuels. The conclusion is that the hazards in relation to ammonia need...... to be controlled by a combination of technical and regulatory measures. The most important requirements are: - Advanced safety systems in the vehicle -Additional technical measures and regulations are required to avoid releases in maintenance workshops and unauthorised maintenance on the fuel system. - Road...... transport of ammonia to refuelling stations in refrigerated form - Sufficient safety zonesbetween refuelling stations and residential or otherwise public areas. When these measures are applied, the use of ammonia as a transport fuel wouldn’t cause more risks than currently used fuels (using current practice)....

  2. Critical Issues for Particle-Bed Reactor Fuels

    Science.gov (United States)

    Evans, Robert S.; Husser, Dewayne L.; Jensen, Russell R.; Kerr, John M.

    1994-07-01

    Particle-Bed Reactors (PBRs) potentially offer performance advantages for nuclear thermal propulsion, including very high power densities, thrust-to-weight ratios, and specific impulses. A key factor in achieving all of these is the development of a very-high-temperature fuel. The critical issues for all such PBR fuels are uranium loading, thermomechanical and thermochemical stability, compatibility with contacting materials, fission product retention, manufacturability, and operational tolerance for particle failures. Each issue is discussed with respect to its importance to PBR operation, its status among current fuels, and additional development needs. Mixed-carbide-based fuels are recommended for further development to support high-performance PBRs.

  3. The preliminary edition of nuclear criticality safety handbook of Japan, (5)

    International Nuclear Information System (INIS)

    Sakai, T.; Naito, Y.

    1987-01-01

    Full reflector thickness and isolation thickness of water and ordinary concrete were calculated with the nuclear criticality safety evaluation code system JACS for nuclear criticality safety on fuel facilities. Since these two values depend on physical/chemical property of nuclear fuel and shape/size of array units, it is unreasonable to apply only one value for all neutron interaction problems. In this paper, a simple evaluation method is examined to obtain full reflector thickness and isolation thickness for various cases. The following results are obtained; these two values are expressed as a simple function of neutron migration length of reflector, and this relation is well explained by one-group diffusion theory. With this relation, we proposed a new method to determine full reflector thickness and isolation thickness which are able to be used for criticality safety design. (author)

  4. Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-11-01

    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.

  5. Overview of the Activities of the OECD/NEA/NSC Working Party on Nuclear Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    Rugama, Y. [OECD Nuclear Energy Agency, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France); Blomquist, R. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Brady Raap, M. [PNNL, 902 Battelle Blvd, P.O. Box 999, MSIN: K8-34, Richland, Washington 99352 (United States); Briggs, B. [INL, P.O. Box 1625, MS-3860 2525 North Fremont, Idaho Falls, ID 83415-3860 (United States); Gulliford, J. [NEXIA SOLUTIONS B168 Curie Av. Harwell Business Centre Didcot Oxfordshire OX110QT (United Kingdom); Miyoshi, Y.; Suyama, K. [JAEA 2-4 Shirakata-Shirane, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan); Ivanova, T. [IRSN, BP17 - 92262 Fontenay-Aux-Roses Cedex (France)

    2008-07-01

    Over the years, substantial progress has been made in developing nuclear data and computer codes to evaluate criticality safety for nuclear fuel handling. This state-of-the-art knowledge also has an economic impact. Increased understanding of uncertainties in safety margins allow rational and more economical designs for manipulation, storage and transportation of fissile materials. In the mid-nineties, several activities related to criticality-safety were grouped together into the Working Party on Nuclear Criticality Safety. Six expert groups co-ordinate various activities that run the gamut from experimental evaluations to code and data intercomparisons, for the study of static and transient criticality behaviors. The various reports produced by the expert groups attempt to establish practical rules and identify applicable tools when appropriate. (authors)

  6. 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.

  7. Critical experiments on minimal-content gadolinia for above-5wt% enrichment fuels in Toshiba NCA

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Watanabe, Shouichi; Yoshioka, Kenichi; Mitsuhashi, Ishi; Kumanomido, Hironori; Sugahara, Satoshi; Hiraiwa, Kouji

    2009-01-01

    A concept of 'minimal-content gadolinia' with a content of less than several hundred ppm mixed in the 'above-5wt% enrichment UO 2 fuel' for super high burnup is proposed for ensuring the criticality safety in the UO 2 fuel fabrication facility for light water reactors (LWRs) without increase in investment cost. Required gadolinia contents calculated were from 53 to 305 ppm for enrichments of UO 2 powders for boiling water reactor (BWR) fuel from 6 to 10 wt%. It is expected that the minimal-content gadolinia yields an acceptable reactivity suppression at the beginning of operating cycle and no reactivity penalty at the end of operating cycle due to no residual gadolinium. A series of critical experiments were carried out in the Toshiba Nuclear Critical Assembly (NCA). Reactivity effects of the gadolinia were measured to clarify the nuclear characteristics, and the measured values and the calculated values agreed within 5%. (author)

  8. 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

  9. General principles of the nuclear criticality safety for handling, processing and transportation fissile materials in the USSR

    International Nuclear Information System (INIS)

    Vnukov, V.S.; Rjazanov, B.G.; Sviridov, V.I.; Frolov, V.V.; Zubkov, Y.N.

    1991-01-01

    The paper describes the general principles of nuclear criticality safety for handling, processing, transportation and fissile materials storing. Measures to limit the consequences of critical accidents are discussed for the fuel processing plants and fissile materials storage. The system of scientific and technical measures on nuclear criticality safety as well as the system of control and state supervision based on the rules, limits and requirements are described. The criticality safety aspects for various stages of handling nuclear materials are considered. The paper gives descriptions of the methods and approaches for critical risk assessments for the processing facilities, plants and storages. (Author)

  10. 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

  11. A relative risk comparison of criticality control strategies based on fresh fuel and burnup credit design bases

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1989-01-01

    The fresh fuel design basis provides some margin of safety, i.e., criticality safety is almost independent of loading operations if fuel designs do not change significantly over the next 40 years. However, the design basis enrichment for future nuclear fuel will most likely vary with time. As a result, it cannot be guaranteed that the perceived passivity of the concept will be maintained over the life cycle of a future cask system. Several options are available to ensure that the reliability of a burnup credit system is comparable to or greater than that of a system based on a fresh fuel assumption. Criticality safety and control reliability could increase with burnup credit implementation. The safety of a burnup credit system could be comparable to that for a system based on the fresh fuel assumption. A burnup credit philosophy could be implemented without any cost-benefit tradeoff. A burnup credit design basis could result in a significant reduction in total system risk as well as economic benefits. These reductions occur primarily as a result of increased cask capacities and, thus, fewer shipments. Fewer shipments also result in fewer operations over the useful life of a cask, and opportunities for error decrease. The system concept can be designed such that only benefits occur. These benefits could include enhanced criticality safety and the overall reliability of cask operations, as well as system risk and economic benefits. Thus, burnup credit should be available as an alternative for the criticality design of spent fuel shipping casks

  12. A critical assessment of fuel cell technology

    International Nuclear Information System (INIS)

    Lindstroem, O.

    1994-01-01

    Cold combustion is a promised technology to mankind since the middle of the last century. The fuel cell may at last become the energy machine of the one to come after a long journey on a road bordered with expectations, successes and disappointments. Ten billion people will need the cell for their well-being. The progress and the state-of-art is assessed by means of figures of merit for performance, normalized to standard conditions, life and variability. State-of-art current densities for multi-kW stacks operating on atmospheric pressure air at 0.74 V cell voltage (50% efficiency, HHV) are estimated to be 150 mA/cm 2 for MCFC, 160 mA/cm 2 for AFC, 239 mA/cm 2 for PEFC and 270 mA/cm 2 for SOFC. PAFC gives 260 mA/cm 2 at 0.66 V and DMFC 100 mA/cm 2 at 0.37 V. Decay rates are about 1%/1000 h for PEFC, PAFC and SOFC compared to 2%/1000 h for AFC and 3%/1000 h for MCFC. Coefficients of variation for cell voltages amount to about 1% for all options, except for MCFC with 3-4%. Improvement of cell performance after 1975 is nil to moderate, except for SOFC with a consistent annual improvement of about 10%. There is room for further development of terrestrial AFCs towards 300-400 mA/cm 2 considering the figure 800 mA/cm 2 for oxygen AFCs. Life and cost will decide the future of the fuel cell. Prospects are not as good as they could be. The fuel cell community lacks understanding of the basics of fuel processing, as demonstrated by the widespread misbelief ('the CO 2 syndrome') that CO 2 cannot be removed cost effectively from a hydrogen feed (which is practiced in every NH 3 plant around the world). The competition, read the gas turbine, has to be taken very seriously. Emphasis has to be shifted from premature demonstrations to R and D on fundamental problems, which have been around too long. 34 refs

  13. 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

  14. 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)

  15. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

  16. Safety aspects of solvent nitration in HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Wilbourn, R.G.

    1977-06-01

    Reprocessing of HTGR fuels requires evaporative concentration of uranium and thorium nitrate solutions. The results of a bench-scale test program conducted to assess the safety aspects of planned concentrator operations are reported

  17. Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

    International Nuclear Information System (INIS)

    Eberle, C.S.; Dean, E.M.; Angelo, P.L.

    1995-01-01

    A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

  18. Methodology for Determination of the Upper Safety Limit for Criticality Calculations for Criticality Safety Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.H.; Keener, H.J.; DeClue, J.F.; Krass, A.W.

    2001-04-01

    This report considers the methods for determination of an upper safety limit, and incorporating uncertainty and margin into the safety limit, provides comparisons, and recommends a preferred method for determining the Upper Safety Limit (USL). A USL is developed for CSAS25 from SCALE4.4a. The USL is applicable for the CSAS25 control module from the SCALE 4.4a computer code system for use in evaluating nuclear criticality safety of enriched uranium systems. The benchmark calculation results used for this report are documented in Y/DD-896. The statistical evaluation is documented in CCG-380. The 27-group ENDF/B-IV, 44-group ENDF/B-V, and 238-group ENDF/B-V cross-section libraries were used. Numerical methods for applying margins are described, but the determination of appropriate correlating parameters and values for additional margin, applicable to a particular analysis, must be determined as part of a process analysis. As such, this document does not specify final upper subcritical limits as has been done in the past. No correlation between calculation results and neutron energy causing fission was found for the critical experiment results. Analysts using these results are responsible for exercising sound engineering judgment using strong technical arguments to develop ''a margin in k{sub eff} or other correlating parameter that is sufficiently large to ensure that conditions (calculated by this method to be subcritical by this margin) will actually be subcritical.'' Documentation of area of applicability and determination and justification of the appropriate margin in the analyst's evaluation, in conjunction with this report, will constitute the complete Validation Report in accordance with ANSI/ANS-8.1-1998, Section 4.3.6(4).

  19. Critical survey on electrode aging in molten carbonate fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, K.

    1979-12-01

    To evaluate potential electrodes for molten carbonate fuel cells, we reviewed the literature pertaining to these cells and interviewed investigators working in fuel cell technology. In this critical survey, the effect of three electrode aging processes - corrosion or oxidation, sintering, and poisoning - on these potential fuel-cell electrodes is presented. It is concluded that anodes of stabilized nickel and cathodes of lithium-doped NiO are the most promising electrode materials for molten carbonate fuel cells, but that further research and development of these electrodes are needed. In particular, the effect of contaminants such as H/sub 2/S and HCl on the nickel anode must be investigated, and methods to improve the physical strength and to increase the conductivity of NiO cathodes must be explored. Recommendations are given on areas of applied electrode research that should accelerate the commercialization of the molten carbonate fuel cell. 153 references.

  20. 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

  1. 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

  2. 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

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Linderoth, C.E.

    1984-03-01

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

  5. Safety assessment of ammonia as a transport fuel

    Energy Technology Data Exchange (ETDEWEB)

    Duijm, N.J.; Markert, F.; Lundtang paulsen, Jette

    2005-02-01

    This report describes the safety study performed as part of the EU supported project 'Ammonia Cracking for Clean Electric Power Technology' The study addresses the following activities: safety of operation of the ammonia-powered vehicle under normal and accident (collision) conditions, safety of transport of ammonia to the refuelling stations and safety of the activities at the refuelling station (unloading and refuelling). Comparisons are made between the safety of using ammonia and the safety of other existing or alternative fuels. The conclusion is that the hazards in relation to ammonia need to be controlled by a combination of technical and regulatory measures. The most important requirements are: - Advanced safety systems in the vehicle - Additional technical measures and regulations are required to avoid releases in maintenance workshops and unauthorised maintenance on the fuel system - Road transport of ammonia to refuelling stations in refrigerated form - Sufficient safety zones between refuelling stations and residential or otherwise public areas. When these measures are applied, the use of ammonia as a transport fuel wouldnt cause more risks than currently used fuels (using current practice). (au)

  6. Safety analysis of IFR fuel processing in the Argonne National Laboratory Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Charak, I; Pedersen, D.R.; Forrester, R.J.; Phipps, R.D.

    1993-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory (ANL) includes on-site processing and recycling of discharged core and blanket fuel materials. The process is being demonstrated in the Fuel Cycle Facility (FCF) at ANL's Idaho site. This paper describes the safety analyses that were performed in support of the FCF program; the resulting safety analysis report was the vehicle used to secure authorization to operate the facility and carry out the program, which is now under way. This work also provided some insights into safety-related issues of a commercial IFR fuel processing facility. These are also discussed

  7. 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)

  8. 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

  9. Safety aspects of the IFR pyroprocess fuel cycle

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  10. Review of the current criticality data validation requirements within the British Nuclear Fuel industry

    International Nuclear Information System (INIS)

    Broome, P.E.; Thorne, P.R.

    1993-01-01

    British Nuclear Fuels plc (BNFL) has commercial involvements throughout the whole of the nuclear fuel cycle, from fuel fabrication through to irradiated fuel transport and pond storage, chemical reprocessing, waste treatment, handling and storage, and the isolation and storage of purified UO 3 and PuO 2 products. The wide range of materials and compositions handled and the variety of environments encountered results in a requirement to address a large number of criticality safety problems. With some problems requiring novel approaches to their solution, there are inevitably some areas where further validation data would be desirable. The principal calculational tool used in Britain for criticality safety assessment work is the Monte Carlo computer code MONK6B. This has already been validated against a wide range of critical experiments, and work is under way to increase the validation data base even further. As previously indicated, however, there are areas where improvements could be made, and these are outlined in the following paragraphs. The power of the MONK6B geometry package is such that virtually any conceivable geometric configuration can be accurately modeled in detail. It remains, therefore, that in most instances, the limiting factor on the accuracy and reliability of the final answer rests with the quality of the nuclear data and this, of course, requires to be validated

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. 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.

  16. Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL

    2010-06-01

    Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNF compositions caused by cross-section data uncertainty and by approximations in the calculational method. The validation is based on comparison between radiochemical assay (RCA) data and calculated isotopic concentrations for fuel samples representative of SNF inventory. The criticality analysis methodology for commercial SNF disposal allows burnup credit for 14 actinides and 15 fission product isotopes in SNF compositions. The neutronic-depletion method for disposal criticality analysis employing burnup credit is the two-dimensional (2-D) depletion sequence TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)/NEWT (New ESC-based Weighting Transport code) and the 44GROUPNDF5 crosssection library in the Standardized Computer Analysis for Licensing Evaluation (SCALE 5.1) code system. The SCALE 44GROUPNDF5 cross section library is based on the Evaluated Nuclear Data File/B Version V (ENDF/B-V) library. The criticality calculation code for disposal criticality analysis employing burnup credit is General Monte Carlo N-Particle (MCNP) Transport Code. The purpose of this calculation report is to determine the bias on the calculated effective neutron multiplication factor, k{sub eff}, due to the bias and bias uncertainty associated with

  17. 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)

  18. Criticality calculation for cluster fuel bundles using grey Dancoff factor

    International Nuclear Information System (INIS)

    Hyeong Heon Kim; Nam Zin Cho

    1999-01-01

    This paper applies the grey Dancoff factor calculated by Monte Carlo method to the criticality calculation for cluster fuel bundles. Dancoff factors for five symmetrically different pin positions of CANDU37 and CANFLEX fuel bundles in full three-dimensional geometry are calculated by Monte Carlo method. The concept of equivalent Dancoff factor is introduced to use the grey Dancoff factor in the resonance calculation based on equivalence theorem. The equivalent Dancoff factor which is based on the realistic model produces an exact fuel collision probability and can be used in the resonance calculation just as the black Dancoff factor. The infinite multiplication factors based on the black Dancoff factors calculated by collision probability or Monte Carlo method are overestimated by about 2 mk for normal condition and 4 mk for void condition of CANDU37 and CANFLEX fuel bundles in comparison with those based on the equivalent Dancoff factors

  19. 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

  20. 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

  1. 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)

  2. Safety and Regulatory Issues of the Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

    2014-02-01

    Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

  3. Improving BWR fuel critical power without increasing bundle pressure drop

    International Nuclear Information System (INIS)

    Matzner, B.; Shiraishi, L.M.; Danielson, D.W.; Congdon, S.P.

    2004-01-01

    It has been almost axiomatic that BWR fuel bundle critical power performance could not be improved without an accompanying increase in bundle pressure drop. It appeared that in order to increase the bundle dryout resistance it was necessary to perturb the bundle coolant flow paths in some fashion. This resulted in an unacceptable bundle pressure drop increase. However, by adding part length rods to decrease bundle pressure drop and by inserting an extra spacer with rearranged spacer pitch and flow trippers on the channel wall at the top of the bundle to increase critical power it was possible to achieve the goal of increased bundle critical power without pressure drop increase. (author)

  4. 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)

  5. 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)

  6. 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

  7. 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).

  8. Safety evaluation of a hydrogen fueled transit bus

    Energy Technology Data Exchange (ETDEWEB)

    Coutts, D.A.; Thomas, J.K.; Hovis, G.L.; Wu, T.T. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1997-12-31

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

  9. Safety physics inter-comparison of advanced concepts of critical reactors and ADS

    International Nuclear Information System (INIS)

    Slessarev, I.

    2001-01-01

    Enhanced safety based on the principle of the natural ''self-defence'' is one of the most desirable features of innovative nuclear systems (critical or sub-critical) regarding both TRU transmutation and ''clean'' energy producer concepts. For the evaluation of the ''self-defence'' domain, the method of the asymptotic reactivity balance has been generalised. The promising option of Hybrids systems (that use a symbiosis of fission and spallation in sub-critical cores) which could benefit the advantages of both Accelerated Driven Systems of the traditional type and regular critical systems, has been advocated. General features of Hybrid dynamics have been presented and analysed. It was demonstrated that an external neutron source of Hybrids can expand the inherent safety potential significantly. This analysis has been applied to assess the safety physics potential of innovative concepts for prospective nuclear power both for energy producers and for transmutation. It has been found, that safety enhancement goal defines a choice of sub-criticality of Hybrids. As for energy producers with Th-fuel cycle, a significant sub-criticality level is required due to a necessity of an improvement of neutronics together with safety enhancement task. (author)

  10. Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.

    1987-01-01

    Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

  11. 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 ...

  12. 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)

  13. 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...

  14. Control of criticality risk in the manufacture of fuel elements for research reactors

    International Nuclear Information System (INIS)

    Friedenthal, M.; Cardenas Yucra, H.R.; Marajofsky, A.; La Gamma de Batistoni, A.M.

    1987-01-01

    The control of criticality risk in a chemical plant adopts different forms according to the quantities of fissile material and the type of compounds used. This work presents the treatment of the critical excursion risk adopted in production plants of U 3 O 8 and manufacturing plants of fuel elements for research reactors, located in Constituyentes Atomic Center. The possible events and accidents related to the fissile material control are analyzed, and the systems of administrative control and intrinsic safety through engineering are described. (Author)

  15. 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....

  16. 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

  17. 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.

  18. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

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

    1981-12-30

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

  19. 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)

  20. A fuel for sub-critical fast reactor

    Science.gov (United States)

    Moiseenko, V. E.; Chernitskiy, S. V.; Ågren, O.; Noack, K.

    2012-06-01

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and 238U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  1. Fire and blast safety manual for fuel element manufacture

    International Nuclear Information System (INIS)

    Ensinger, U.; Koehler, B.; Mester, W.; Riotte, H.G.; Sehrbrock, H.W.

    1988-01-01

    The manual aims to enable people involved in the planning, operation, supervision, licensing or appraisal of fuel element factories to make a quick and accurate assessment of blast safety. In Part A, technical plant principles are shown, and a summary lists the flammable materials and ignition sources to be found in fuel element factories, together with theoretical details of what happens during a fire or a blast. Part B comprises a list of possible fires and explosions in fuel element factories and ways of preventing them. Typical fire and explosion scenarios are analysed more closely on the basis of experiments. Part B also contains a list and an assessment of actual fires and explosions which have occurred in fuel element factories. Part C contains safety measures to protect against fire and explosion, in-built fire safety, fire safety in plant design, explosion protection and measures to protect people from radiation and other hazards when fighting fires. A distinction is drawn between UO 2 , MOX and HTR fuel elements. (orig./DG) [de

  2. 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

  3. Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

    Directory of Open Access Journals (Sweden)

    Herrero J.J.

    2017-01-01

    Full Text Available In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

  4. Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

    Science.gov (United States)

    Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

    2017-09-01

    In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

  5. 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

  6. Safety research in nuclear fuel cycle at PNC

    International Nuclear Information System (INIS)

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF 6 , uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  7. 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

  8. Key technical issues relating to safety of spent fuel dry storage in vaults: CASCAD system

    International Nuclear Information System (INIS)

    Berge, F.

    1994-01-01

    The operating CASCAD Facility at the Cadarashe site (FR) was commissioned in May 1990. Fuel is received in tight canisters which are transferred to storage pits in the vault and scheduled to be stored for up to 50 years. Canistering operations are performed in a cell of the reactor building.The paper describes the main functions of the facility as: cask receipt and shipping; fuel unloading; fuel conditioning; canisters emplacements in storage location; fuel storage; fuel retrieving and shipping at the end of the storage period; operation system and operation organization. Safety characteristics of the facility discussed are: fuel decay heat removal; subcriticality control and radiological protection. The fuel decay heat removal has two main purposes: 1) maintaining rod cladding temperature below a set limit in order to keep the fuel in its as received condition; 2) maintaining structures and equipment performing a safety function below the design temperature. The features of the sub-criticality control in the storage vault are such that sub-criticality in normal and accidental conditions is provided by the arrangement of pits in the vault. Radiological protection is based on limiting collective and individual annual dose equivalent to ALARA levels ensuring that they remain in any case below the set limits. Radiological protection system described consists in: confinement of radioactive materials for protection against its dissemination; radiation shielding for protection against irradiation. It is pointed out that all technical solutions presented are based on or adapted from proven technologies used in operating facilities in France or in other countries. The solution not only benefits from the experience of SGN in the design, construction and start-up of facilities for fuel or high level waste handling and storage, but also from the experience of the CEA and COGEMA groups in operating such facilities. 2 figs., 1 ref

  9. 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

  10. 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.

  11. Safety Issues with Hydrogen as a Vehicle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; J. S. Herring

    1999-09-01

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

  12. Safety Issues with Hydrogen as a Vehicle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles; Herring, James Stephen

    1999-10-01

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

  13. An Approach to Modeling Software Safety in Safety-Critical Systems

    OpenAIRE

    Ben S. Medikonda; Seetha R. Panchumarthy

    2009-01-01

    Software for safety-critical systems has to deal with the hazards identified by safety analysis in order to make the system safe, risk-free and fail-safe. Software safety is a composite of many factors. Problem statement: Existing software quality models like McCalls and Boehms and ISO 9126 were inadequate in addressing the software safety issues of real time safety-critical embedded systems. At present there does not exist any standard framework that comprehensively addresses the Factors, Cr...

  14. Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-01-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included

  15. 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)

  16. Experiments for IFR fuel criticality in ZPPR-21

    International Nuclear Information System (INIS)

    Olsen, D.N.; Collins, P.J.; Carpenter, S.G.

    1991-01-01

    A series of benchmark measurements was made in ZPPR-21 to validate criticality calculations for fuel processing operations for Argonne's Integral Fast Reactor program. Six different mixtures of Pu/U/Zr fuel with a graphite reflector were built and criticality was determined by period measurements. The assemblies were isolated from room return neutrons by a lithium hydride shield. Analysis was done using a fully-detailed model with the VIM Monte Carlo code and ENDF/B-V.2 data. Sensitivity analysis was used to validate the measurements against other benchmark data. A simple RZ model was defined and used with the KENO code. Corrections to the RZ model were provided by the VIM calculations with low statistical uncertainty. (Author)

  17. Experiments for IFR fuel criticality in ZPPR-21

    International Nuclear Information System (INIS)

    Olsen, D.N.; Collins, P.J.; Carpenter, S.G.

    1991-01-01

    A series of benchmark measurements was made in ZPPR-21 to validate criticality calculations for fuel operations in Argonne's Integral Fast Reactor. Six different mixtures of Pu/U/Zr fuel with a graphite reflector were built and criticality was determined by period measurements. The assemblies were isolated from room return problems by a lithium hydride shield. Analysis was done using a fully-detailed model with the VIM Monte Carlo code and ENDF/B-V.2 data. Sensitivity analysis was used to validate the measurements against other benchmark data. A simple RZ model was defined the used with the KENO code. Corrections to the RZ model were provided by the VIM calculations with low statistical uncertainty. 7 refs., 5 figs., 5 tabs

  18. Integration of Several Elements of the DOE Nuclear Criticality Safety Program

    International Nuclear Information System (INIS)

    Valentine, T.E.

    2001-01-01

    The U. S. Department of Energy established the Nuclear Criticality Safety Program (NCSP) to maintain the infrastructure and expertise in nuclear criticality safety to support line criticality safety programs at various DOE sites. The seven tasks of the NCSP include critical experiments, benchmarking, nuclear data, analytical methods, applicable ranges of bounding curves and data, information preservation and dissemination, and training and qualification. The goals of this program are to improve the knowledge, tools, data, guidance, and information available to the nuclear criticality safety community. In addition various elements of the NCSP are integrated together to provide the nuclear criticality safety community with the most precise nuclear data for criticality safety analyses. This paper describes how several elements of the NCSP were integrated together in the evaluation of the silicon nuclear data. Silicon is frequently encountered in decontamination and decommissioning efforts, process sludge and settling tanks, in situ vitrification, and waste remediation efforts (including waste storage, retrieval, characterization, volume reduction, and stabilization). Silicon was also identified as an important isotope for addressing concerns associated with the storage of spent nuclear fuels in a geologic repository. The inadequacy of the silicon nuclear data in the intermediate energy region mandated that additional neutron capture cross-section measurements had to be performed that encompassed the resolved resonance region. An evaluation was performed that included analysis of the most recent neutron capture and existing transmission cross-section measurements performed at the Oak Ridge Electron Linear Accelerator. Critical experiments were performed at the Institute of Physics and Power Engineering in Obninsk, Russia because of the lack of critical experiment data for analysis of storage of nuclear material in a geologic repository. These critical experiments were

  19. Safety in connection with the request for approval of the installation alteration in the fuel reprocessing facilities of Power Reactor and Nuclear Fuel Development Corporation (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report to the Prime Minister by the Nuclear Safety Commission was presented concerning the safety in the installation alteration of the fuel reprocessing facilities, as PNC had requested its approval to the Prime Minister. The safety was confirmed. The items of examination on the safety made by the committee on Examination of Nuclear Fuel Safety of NSC were the aseismic design of liquid waste storage, uranium denitration facility, intermediate gate and radioactive solid waste storage; the criticality safety design of the denitration facility; the radiation shielding design of the liquid waste storage, denitration facility and solid waste storage; the function of radioactive material containment of the liquid waste storage and denitration facility; the radiation control in the liquid waste storage, denitration facility and solid waste storage; the waste management in the liquid waste storage and denitration facility; fire and explosion prevention in the liquid waste storage; exposure dose from the liquid waste storage and denitration facility. (Mori, K.)

  20. Prevention of criticality accidents in a fuel cycle plant

    International Nuclear Information System (INIS)

    Gatti, A.M.; Canavese, S.I.; Capadona, N.M.

    1990-01-01

    This work reports the basic considerations on criticality accidents applied to an uranium dioxide fuel cycle production plant. The different fabrication stages are briefly described, with the identification of the neutronically isolated areas. Once the areas have been defined, an evaluation is made, setting up the control parameters to be used in each of them and their variation ranges; normal operation limitations based on experimental data or validating calculations, applied specifically to 5% enriched uranium, are established. Afterwards, defined parameters deviations are analyzed due to incidental conditions in order to prevent criticality accidents under normal conditions and maintenance operations. (Author) [es

  1. Evaluation of a potential nuclear fuel repository criticality: Lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, J.R.; Evans, D.

    1995-10-01

    This paper presents lessons learned from a Probabilistic Risk Assessment (PRA) of the potential for a criticality in a repository containing spent nuclear fuel with high enriched uranium. The insights gained consisted of remarkably detailed conclusions about design issues, failure mechanisms, frequencies and source terms for events up to 10,000 years in the future. Also discussed are the approaches taken by the analysts in presenting this very technical report to a nontechnical and possibly antagonistic audience.

  2. Evaluation of a potential nuclear fuel repository criticality: Lessons learned

    International Nuclear Information System (INIS)

    Wilson, J.R.; Evans, D.

    1995-01-01

    This paper presents lessons learned from a Probabilistic Risk Assessment (PRA) of the potential for a criticality in a repository containing spent nuclear fuel with high enriched uranium. The insights gained consisted of remarkably detailed conclusions about design issues, failure mechanisms, frequencies and source terms for events up to 10,000 years in the future. Also discussed are the approaches taken by the analysts in presenting this very technical report to a nontechnical and possibly antagonistic audience

  3. Summary of NRC LWR safety research programs on fuel behavior, metallurgy/materials and operational safety

    International Nuclear Information System (INIS)

    Bennett, G.L.

    1979-09-01

    The NRC light-water reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions. Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials research program provides independent confirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear power plants. The topics currently being addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics

  4. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovitchev, A.M.

    2001-09-28

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

  6. Hot Fuel Examination Facility/North Facility safety report

    Energy Technology Data Exchange (ETDEWEB)

    Adams, R.M.; Hampson, D.C.; Ferguson, K.R.; Hylsky, E.

    1975-02-01

    Design and safety-related construction features of the Hot Fuel Examination Facility/North, located on the Argonne--West site at the Idaho National Engineering Laboratory are described. The proposed operations, the organizational structure, and emergency plans are given. Evaluations of potential accident situations are presented and it is concluded that HFEF/N can be operated safely and without undue hazard.

  7. Standard model for safety analysis report of fuel reprocessing plants

    International Nuclear Information System (INIS)

    1979-12-01

    A standard model for a safety analysis report of fuel reprocessing plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  8. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    1980-09-01

    A standard model for a safety analysis report of fuel fabrication plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  9. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Tsuchida, Noboru; Shiraishi, Tadao; Takahashi, Yutaka; Inada, Seiji; Saito, Minoru; Futamura, Yoshiaki; Kitano, Kyoshiro.

    1992-10-01

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  10. 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

  11. Criticality Safety Basics for INL FMHs and CSOs

    International Nuclear Information System (INIS)

    Putman, V.L.

    2012-01-01

    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, and additional information

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. Fuel cycle safety research. Role and development in future

    International Nuclear Information System (INIS)

    Matsumoto, Shiro

    2005-01-01

    The report stresses important roles played by risk informed methodology in chemical process safety problems as encountered in fuel cycle facility such as a reprocessing plant. Abnormal situation management is a safety issue - The lesson learned from the pesticide manufacturing facility in Bhopal, India. Little attention has been given to understanding the issues regarding performance during normal versus abnormal situation. The first step is in abnormal situation management is to define what really is abnormal. The second step is to ensure that everyone understands the difference between normal and abnormal, and the root causes of abnormal events. The third step is to be aware of current practices that support abnormal situation management, and the procedures, practices, and techniques used to respond to abnormal conditions. Dynamic simulation will become to useful techniques for abnormal situation management as well as probabilistic safety assessment for process facilities including nuclear fuel cycle. (S. Ohno)

  18. 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)

  19. Criticality Safety Code Validation with LWBR’s SB Cores

    Energy Technology Data Exchange (ETDEWEB)

    Putman, Valerie Lee

    2003-01-01

    The first set of critical experiments from the Shippingport Light Water Breeder Reactor Program included eight, simple geometry critical cores built with 233UO2-ZrO2, 235UO2-ZrO2, ThO2, and ThO2-233UO2 nuclear materials. These cores are evaluated, described, and modeled to provide benchmarks and validation information for INEEL criticality safety calculation methodology. In addition to consistency with INEEL methodology, benchmark development and nuclear data are consistent with International Criticality Safety Benchmark Evaluation Project methodology.Section 1 of this report introduces the experiments and the reason they are useful for validating some INEEL criticality safety calculations. Section 2 provides detailed experiment descriptions based on currently available experiment reports. Section 3 identifies criticality safety validation requirement sources and summarizes requirements that most affect this report. Section 4 identifies relevant hand calculation and computer code calculation methodologies used in the experiment evaluation, benchmark development, and validation calculations. Section 5 provides a detailed experiment evaluation. This section identifies resolutions for currently unavailable and discrepant information. Section 5 also reports calculated experiment uncertainty effects. Section 6 describes the developed benchmarks. Section 6 includes calculated sensitivities to various benchmark features and parameters. Section 7 summarizes validation results. Appendices describe various assumptions and their bases, list experimenter calculations results for items that were independently calculated for this validation work, report other information gathered and developed by SCIENTEC personnel while evaluating these same experiments, and list benchmark sample input and miscellaneous supplementary data.

  20. Effect of mixing state on criticality safety evaluation in MOX powder and additive

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori

    2005-01-01

    Criticality safety analyses are discussed in which MOX powder and additive (e.g. zinc-stearate) are mixed in a powder treatment process of MOX fuel fabrication. The multiplication factor k eff is largely affected by how they are mixed, i.e., how the density and volume change with the mixing. In general, k eff increases when MOX powder is mixed with zinc-stearate. However, plutonium content and density of MOX powder make a difference in the k eff 's changes. Especially, MOX powder with a higher plutonium content and a higher density is not always unsafe in terms of criticality if it is mixed with zinc-stearate. (author)

  1. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  2. Safety problems in fuel reprocessing plants

    International Nuclear Information System (INIS)

    Amaury, P.; Jouannaud, C.; Niezborala, F.

    1979-01-01

    The document first situates the reprocessing in the fuel cycle as a whole. It shows that a large reprocessing plant serves a significant number of reactors (50 for a plant of 1500 tonnes per annum). It then assesses the potential risks with respect to the environment as well as with respect to the operating personnel. The amounts of radioactive matter handled are very significant and their easily dispersible physical form represents very important risks. But the low potential energy likely to bring about this dispersion and the very severe and plentiful confinement arrangements are such that the radioactive risks are very small, both with respect to the environment and the operating personnel. The problems of the interventions for maintenance or repairs are mentioned. The intervention techniques in a radioactive environment are perfected, but they represent the main causes of operating personnel irradiation. The design principle applied in the new plants take this fact into account, involving a very significant effort to improve the reliability of the equipment and ensuring the provision of devices enabling the failing components to be replaced without causing irradiation of the personnel [fr

  3. 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

  4. 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

  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. Criticality safety implementation, operations, and training for TMI-2 defueling

    International Nuclear Information System (INIS)

    Knief, R.A.; Fergus, I.E.

    1986-01-01

    The nuclear criticality safety of the accident-damaged Three Mile Island Unit 2 (TMI-2) reactor has depended primarily on the use of soluble neutron poison. Detailed calculations have been performed to establish the boric acid concentrations needed for subcriticality of postulated core configurations related to recovery and defueling activities. Based on the minimum acceptable boron concentration, operating requirements were developed, safety reviews conducted, strategies and procedures implemented, and training conducted

  7. 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

  8. 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.

  9. 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

  10. Critical Ethnography, Cultural Safety, and International Nursing Research

    OpenAIRE

    Jean N. Harrowing PhD; Judy Mill PhD; Jude Spiers PhD; Judith Kulig PhD; Walter Kipp PhD

    2010-01-01

    Critical qualitative methodology provides a strategy to examine the human experience and its relationship to power and truth. Cultural safety is a concept that has been applied to nursing education and practice and refers to interactions that acknowledge and respect the unique cultural background of patients. It recognizes power inequities between caregivers who belong to dominant cultures and patients who may belong to oppressed groups. Culture is interpreted from a critical constructivist p...

  11. 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.

  12. 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

  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. Fuel safety criteria technical review - Results of OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria

    International Nuclear Information System (INIS)

    Hollasky, N.; Valtonen, K.; Hache, G.; Gross, H.; Bakker, K.; Recio, M.; Bart, G.; Zimmermann, M.; Van Doesburg, W.; Killeen, J.; Meyer, R.O.; Speis, T.

    2000-01-01

    With the advent of advanced fuel and core designs, the adoption of more aggressive operational modes and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a concern if safety margins have remained adequate. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel that was available at that time, mostly with unirradiated specimens. Verification was of course performed as designs progressed in later years, however mostly with the aim to be able to prove that these designs adequately complied with existing criteria, and not to establish new limits. The OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) was therefore given the mandate to technically review the existing fuel safety criteria, focusing on the 'new design' elements (new fuel and core design, cladding materials, manufacturing processes, high burnup, MOX, etc.) introduced by the industry. It should also identify if additional efforts may be required (experimental, analytical) to ensure that the basis for fuel safety criteria is adequate to address the relevant safety issues. In this report, fuel-related criteria are discussed without attempting to categorize them according to event type or risk significance. For each of these 20 criteria, we present a brief description of the criterion as it is used in several applications along with the rationale for having such a criterion. New design elements, such as different cladding materials, higher burnup, and the use of MOX fuels, can affect fuel-related margins and, in some cases, the criteria themselves. Some of the more important effects are mentioned in order to indicate whether the criteria need to be re-evaluated. The discussion may not cover all possible effects, but should be sufficient to identify those criteria that need to be addressed. A summary of these discussions is given in Section 7. As part

  15. Probabilistic safety analysis of transportation of spent fuel

    International Nuclear Information System (INIS)

    Subramaniam, Chitra

    1999-11-01

    The report presents the results of the study carried out to estimate the accident risk involved in the transport of spent fuel from Rajasthan Atomic Power Station near Kota to the fuel reprocessing plant at Tarapur. The technique of probabilistic safety analysis is used. The fuel considered is the Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally-cooled shipping cask over a distance of 822 km by rail. The Indian rail accident statistics are used to estimate the basic rail accident frequency. The possible ways in which a release of radioactive material can occur from the spent fuel cask are identified by the fault tree analysis technique. The release sequences identified are classified into eight accident severity categories, and release fractions are assigned to each. The consequences resulting from the release are estimated by the computer code RADTRAN 4. Results of the risk analysis indicate that the accident risk values are very low and hence acceptable. Parametric studies show that the risk would continue to be small even if the controlling parameters were to simultaneously take extreme adverse values. (author)

  16. The safety concept and classification for HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The principle subjects of this study is to determine the safety concept and classification of HANARO FTL for the design of systems. HANARO FTL should be approved by KINS for installation and operation. And these irradiation for facilities will be designed and installed based upon the safety concept and classification results. (author). 14 refs., 12 tabs., 10 figs.

  17. 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)

  18. 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)

  19. Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report

    International Nuclear Information System (INIS)

    MORGAN, R.G.

    1999-01-01

    The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building)

  20. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadi, Ashgar; Omidvari, Nima [Iran Radioactive Waste Management Company, Tehran (Iran, Islamic Republic of); Hassanzadeh, Mostafa [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-12-15

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k{sub eff}) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  1. Economic incentives for additional critical experimentation applicable to fuel dissolution

    International Nuclear Information System (INIS)

    Mincey, J.F.; Primm, R.T. III; Waltz, W.R.

    1981-01-01

    Fuel dissolution operations involving soluble absorbers for criticality control are among the most difficult to establish economical subcritical limits. The paucity of applicable experimental data can significantly hinder a precise determination of a bias in the method chosen for calculation of the required soluble absorber concentration. Resorting to overly conservative bias estimates can result in excessive concentrations of soluble absorbers. Such conservatism can be costly, especially if soluble absorbers are used in a throw-away fashion. An economic scoping study is presented which demonstrates that additional critical experimentation will likely lead to reductions in the soluble absorber (i.e., gadolinium) purchase costs for dissolution operations. The results indicate that anticipated savings maybe more than enough to pay for the experimental costs

  2. TRIGA FUEL PHASE I AND II CRITICALITY CALCULATION

    International Nuclear Information System (INIS)

    Angers, L.

    1999-01-01

    The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside a five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP containing the five SRS HLW and DOE SNF canisters in various stages of degradation. This calculation will support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository (MGR)

  3. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Vieno, T.

    1994-04-01

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  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. Nuclear power performance and safety. V.5. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1988-01-01

    The International Conference on Nuclear Power Performance and Safety, organized by the International Atomic Energy Agency, was held at the Austria Centre Vienna (ACV) in Vienna, Austria, from 28 September to 2 October 1987. The objective of the Conference was to promote an exchange of worldwide information on the current trends in the performance and safety of nuclear power and its fuel cycle, and to take a forward look at the expectations and objectives for the 1990s. Policy decisions for waste management have already been taken in many countries and the 1990s should be a period of demonstration and implementation of these policies. As ilustrated by data presented from a number of countries, many years of experience in radioactive waste management have been achieved and the technology exists to implement the national plans and policies that have been developed. The establishment of criteria, the development of safety performance methodology and site investigation work are key activities essential to the successful selection, characterization and construction of geological repositories for the final disposal of radioactive waste. Considerable work has been done in these areas over the last ten years and will continue into the 1990s. However, countries that are considering geological disposal for high level waste now recognize the need for relating the technical aspects to public understanding and acceptance of the concept and decision making activities. The real challenge for the 1990s in waste disposal will be successfully to integrate technological activities within a process which responds to institutional and public concern. Volume 5 of the Proceedings comprehends the contributions on waste management in the 1990s. Decontamination and decommissioning, waste management, treatment and disposal, nuclear fuel cycle - present and future. Enrichment services and advanced reactor fuels, improvements in reactor fuel utilization and performance, spent fuel management

  6. 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...

  7. 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

  8. Safety and economic comparison of fusion fuel cycles

    International Nuclear Information System (INIS)

    Brereton, S.J.; Kazimi, M.S.

    1987-08-01

    The DT, DD and DHe fusion fuel cycles are compared on the basis of safety and economics. The designs for the comparison employ HT-9 structure and helium coolant; liquid lithium is used as the tritium breeder for the DT fuel cycle. The reactors are pulsed superconducting tokamaks, producing 4000 MW thermal power. The DT and DD designs are developed utilizing a plasma beta of 5%, 10% and 20%, assuming first stability scaling laws; a single value of 10% for beta is used for the DHe design. Modest extrapolations of current day technology are employed, providing a reference point for the relative ranking of the fuel cycles. Technological advances and improved understanding of the physics involved may alter the relative positions from what has been determined here. 92 figs., 59 tabs

  9. 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.

  10. Criticality safety enhancements for SCALE 6.2 and beyond

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

    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.

  11. TVO-92 safety analysis of spent fuel disposal

    International Nuclear Information System (INIS)

    Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H.

    1993-08-01

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

  12. The SCALE criticality safety analysis sequences: Status and future directions

    International Nuclear Information System (INIS)

    Parks, C.V.

    1993-01-01

    The Standardized Computer Analyses for Licensing Evaluation (SCALE) code system. Was originally conceived and developed in the late 1970s for the US Nuclear Regulatory Commission. The goal was to provide easy-to-use, yet accurate, analysis capabilities for use in evaluating the criticality safety, shielding, and heat transfer aspects of transportation packages for radioactive material. The Criticality Safety Analysis Sequences (CSAS) for SCALE were developed to ''automate'' problem-dependent cross-section and material processing prior to execution of the wellestablished XSDRNPM or KENO codes for calculation of k eff . The criticality analysis sequences provided in SCALE-4 are summarized. The SCALE system continues to be maintained and enhanced by staff of the Computing Applications Division at Oak Ridge National Laboratory (ORNL). The purpose of this paper is to discuss recent work to improve system portability and user interfaces and to provide information on ongoing work to enhance the analysis capabilities

  13. 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.

  14. Implementation of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Stewart, L.; Tonkay, D.

    2004-01-01

    This paper discusses the implementation of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The Joint Convention: establishes a commitment with respect to safe management of spent nuclear fuel and radioactive waste; requires the Parties to ''take appropriate steps'' to ensure the safety of their spent fuel and waste management activities, but does not delineate standards the Parties must meet; and seeks to attain, through its Contracting Parties, a higher level of safety with respect to management of their spent nuclear fuel, disused sealed sources, and radioactive waste

  15. ENSURING ADEQUATE SAFETY WHEN USING HYDROGEN AS A FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Coutts, D

    2007-01-22

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

  16. 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

  17. Experimental research on safety assurance of advanced WWER fuel cycles

    International Nuclear Information System (INIS)

    Krainov, Ju.; Kukushkin, Ju.

    2002-01-01

    The paper presents the results of experimental investigations on substantiation of implementation of a modernized butt joint for the WWER-440 reactor, carried out in the critical test facility 'P' in the RRC 'Kurchatov Institute'. The comparison results of the calculation and experimental data obtained in the physical startup of Volgodonsk NPP-1 with the WWER-1000 are also given. In the implementation of four-year fuel cycle in the WWER-440 with the average enrichment of fuel makeup 3.82% it was solved to conduct experimental research of power distribution in the vicinity of control rod butt junction. Moreover, it was assumed that adequate actions should be applied to eliminate inadmissible power jumps, if necessary. It is not available to measure their values in NPP conditions. Therefore, the power distribution near the butt joint was studied in a 19-rod bank installed in the critical test facility 'P' first for the normal design of the joint when surrounding fuel assemblies enrichment goes up. Then a set of calculation and tests was fulfilled to optimize a butt junction design. On the base of this research the composition of a butt junction was advanced by placing Hf plates into the junction. The effectiveness of modernized butt joint design was experimentally confirmed. In Volgodonsk NPP-1 with WWER-1000 the four-year fuel cycle is being implemented. During the physical startup of the reactor the measurements of the reactivity effects and coefficients were measured at the minimum controlled flux level, and the parameters of a number of critical states were recorded. The data obtained were compared with the calculation. The validity of the certified code package for forecasting the neutronic characteristics of WWER-1000 cores in the implementation of a four year fuel cycle has been supported (Authors)

  18. A study of criticality in a spent fuel repository based on current canister designs

    International Nuclear Information System (INIS)

    Hicks, T.; Prescott, A.

    2000-01-01

    SKB's concept for underground disposal of spent fuel includes requirements on the repository's barrier system that are aimed at ensuring long-term safety. This report is concerned with the requirement on disposal canister design that there is no risk of criticality in the event of water entering the canister after disposal. In particular, recent changes in disposal canister design have led to the need to re-evaluate the possibility of a criticality excursion occurring after disposal. This report presents the results of criticality calculations that have been performed based on current disposal canister designs, and presents the findings of reviews of previous criticality studies that were undertaken as part of the Swedish repository development programme. Under the current reference design, the canister will comprise an outer layer of copper and a cylindrical insert of cast nodular iron. Channels for the fuel assemblies will be created by casting steel pipes into the nodular iron. The pipes will be welded together to a cassette and placed in the centre of the casting mould. Recently, the steel cassette has been modified by the introduction of cooling tubes to improve the casting process. Criticality calculations have been undertaken for spent fuel disposal canisters under repository conditions. The calculations involved determining the neutron multiplication factor for various disposal configurations, depending on the type of canister and fuel assemblies, the initial fuel enrichment, the amount of fuel burn-up, and the amount of burnable poison present. In particular, canisters were assumed to fail at some time after disposal, such that water entered the canister and filled the voids, including the cooling tubes in the canister. Canisters containing various Boiling-Water Reactor (BWR) and Pressurized-Water Reactor (PWR) spent fuel assemblies were assessed. The reference case for each fuel type assumed a burnup of 40,000 MWd/tHM (megawatt days per tonne of heavy

  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. 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

  1. 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

  2. Safety Aspects of Radioactive Waste Management in Different Nuclear Fuel Cycle Policies, a Comparative Study

    International Nuclear Information System (INIS)

    Gad Allah, A.A.

    2009-01-01

    With the increasing demand of energy worldwide, and due to the depletion of conventional natural energy resources, energy policies in many countries have been devoted to nuclear energy option. On the other hand, adopting a safe and reliable nuclear fuel cycle concept guarantees future nuclear energy sustain ability is a vital request from environmental and economic point of views. The safety aspects of radioactive waste management in the nuclear fuel cycle is a topic of great importance relevant to public acceptance of nuclear energy and the development of nuclear technology. As a part of nuclear fuel cycle safety evaluation studies in the department of nuclear fuel cycle safety, National Center for Nuclear Safety and Radiation Control (NCNSRC), this study evaluates the radioactive waste management policies and radiological safety aspects of three different nuclear fuel cycle policies. The once-through fuel cycle (OT- fuel cycle) or the direct spent fuel disposal concept for both pressurized light water reactor ( PWR) and pressurized heavy water reactor (PHWR or CANDU) systems and the s elf-generated o r recycling fuel cycle concept in PWR have been considered in the assessment. The environmental radiological safety aspects of different nuclear fuel cycle options have been evaluated and discussed throughout the estimation of radioactive waste generated from spent fuel from these fuel cycle options. The decay heat stored in the spent fuel was estimated and a comparative safety study between the three fuel cycle policies has been implemented

  3. Private Memory Allocation Analysis for Safety-Critical Java

    DEFF Research Database (Denmark)

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

    2012-01-01

    allowed from objects allocated in scopes with a shorter lifetime to objects allocated in scopes with a longer lifetime. To ensure memory safety, programmers are required to either manually annotate the application with complex annotations, rely on a runtime test of each reference assignment, or statically......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...

  4. Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

    2015-01-01

    Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

  5. Fuel Handling Equipment Maintenance for Critical Path Time Savings

    Energy Technology Data Exchange (ETDEWEB)

    Saville, M.; Williams, A.

    2015-07-01

    By sharing lessons learned and operating experience gained by AREVA Stearns RogerTM Services from more than 45 years of servicing, maintaining, and upgrading Fuel Handling Equipment (FHE) and as the original equipment manufacturer to 56% of domestic U.S. FHE (PWR and BWR) as well as 19 units overseas, this paper presents trends and market forces that have led to the neglect of FHE, the risks of not adequately maintaining FHE, and the financial benefits of proactively maintaining FHE. The benefit to audiences is to come to a better understanding of how critical path delays can be avoided and thus reduce nuclear power plant operating costs. Note that statistics and monetary values given herein are based on recent typical experiences of AREVA Stearns RogerTM Services. Examples discussed are based on actual lessons learned. For the purposes of this paper, upgrades are considered a part of equipment maintenance unless specifically discussed separately. (Author)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. 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

  8. Nuclear criticality safety program development using necessary and sufficient standards

    Energy Technology Data Exchange (ETDEWEB)

    Croucher, D.W.; Stachowiak, R.V. [Kaiser-Hill Co., LLC, Golden, CO (United States); Wilson, R.E. [Safe Sites of Colorado, Golden, CO (United States)

    1996-12-31

    The U.S. Department of Energy`s (DOE`s) Necessary and Sufficient Standards Closure Process has been used to develop a new criticality, safety program manual for the Rocky Flats Environmental Technology Site (RFETS). Standards define and communicate the expectations for performance of work. The purpose of the necessary and sufficient standards closure process is to apply standards determined to be necessary and sufficient for protecting the workers, the public, and the environment. This ensures that the applied standards add value to the performance of the activity; work effectiveness is increased. The purpose of this paper is to briefly describe the process and the results for the selection of national criticality safety standards for use at the Rocky Flats facilities.

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Barilo, Nick F.

    2014-09-29

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

  11. 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...

  12. 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)....

  13. Safety-Critical Software: Status Report and Annotated Bibliography

    Science.gov (United States)

    1993-06-01

    software in place of hardware in safety-critical sys- tems are the Therac 25 (a therapeutic linear accelerator) and nuclear reactor shutdown sys- tems...Leveson and Turner [141], is the Therac 25 radiation treatment machine. A predecessor to the Therac 25, the Therac 20, had a number of hardware Interlocks...to stop an undesirable behavior. Much of the software in the Therac 25 was similar to that of the Therac 20 and the software in both cases contained

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

    DEFF Research Database (Denmark)

    Ravn, Anders P.; 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....

  15. Plenary keynote: Monitoring Safety Critical Infrastructure with Mobile Robots

    OpenAIRE

    Sattar, TP

    2017-01-01

    Reliable Non Destructive Testing (NDT) is vital to the integrity and performance management of capital assets in safety critical industries such as aerospace, transportation, pipelines, petro-chemical processing, and power generation [ ]. The structures that are to be inspected are usually very large and located in remote and hazardous environments. The NDT system has to be deployed by first providing very expensive access, requiring the erection of scaffolding and lengthy preparation before ...

  16. 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)

  17. Safety analysis of LWR irradiated fuel element pool storages before reprocessing

    International Nuclear Information System (INIS)

    Lefort, G.; Leclerc, J.; Hoffman, A.; Frejaville, C.; Domage, M.

    1984-01-01

    The protection of operators and environment requires imperatively that the safety must be taken into account as early as the design of the pools takes place and working conditions are defined. The analysis of criticality, irradiation, contamination, external or internal aggression hazards... allows to draw the main constraints which must be retained in the sizing of these pools: the criticality risk needs distances between fuel elements which results in a not very good utilization of the available area which leads to the utilization of neutron shieldings or requires a safe knowledge of the fuel elements burn up; the irradiation and contamination risks require a special quality of the pool water (temperature, activity, purity...) a good tightness of the basins to locate and to isolate the dubions fuel elements; the external or internal aggression risks such as earthquakes, missiles or loads drops, explosion, imply the civil engineering and involve the use of special technical devices. A brief presentation of the pool storages of the next UP2-800 and UP3 A reprocessing plants allows to show how the requirement drawn by safety analysis have been enforced, while carrying out civil engineering works without equivalent in the world, in this field. The foreseeable evolution of the uranium enrichment rate and burn-up of next PWR fuel elements have an effect upon the risk evaluations; a device apparatus, developed in CEA, for the measurement of burn up and cooling time is presented. At least, a short presentation of the mechanical structure durability studies of the reception and storage spent fuels installations are allowed to improve our knowledge in working conditions and in case of serious accidents

  18. Agile Methods for Open Source Safety-Critical Software.

    Science.gov (United States)

    Gary, Kevin; Enquobahrie, Andinet; Ibanez, Luis; Cheng, Patrick; Yaniv, Ziv; Cleary, Kevin; Kokoori, Shylaja; Muffih, Benjamin; Heidenreich, John

    2011-08-01

    The introduction of software technology in a life-dependent environment requires the development team to execute a process that ensures a high level of software reliability and correctness. Despite their popularity, agile methods are generally assumed to be inappropriate as a process family in these environments due to their lack of emphasis on documentation, traceability, and other formal techniques. Agile methods, notably Scrum, favor empirical process control, or small constant adjustments in a tight feedback loop. This paper challenges the assumption that agile methods are inappropriate for safety-critical software development. Agile methods are flexible enough to encourage the rightamount of ceremony; therefore if safety-critical systems require greater emphasis on activities like formal specification and requirements management, then an agile process will include these as necessary activities. Furthermore, agile methods focus more on continuous process management and code-level quality than classic software engineering process models. We present our experiences on the image-guided surgical toolkit (IGSTK) project as a backdrop. IGSTK is an open source software project employing agile practices since 2004. We started with the assumption that a lighter process is better, focused on evolving code, and only adding process elements as the need arose. IGSTK has been adopted by teaching hospitals and research labs, and used for clinical trials. Agile methods have matured since the academic community suggested they are not suitable for safety-critical systems almost a decade ago, we present our experiences as a case study for renewing the discussion.

  19. 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

  20. Agile Methods for Open Source Safety-Critical Software

    Science.gov (United States)

    Enquobahrie, Andinet; Ibanez, Luis; Cheng, Patrick; Yaniv, Ziv; Cleary, Kevin; Kokoori, Shylaja; Muffih, Benjamin; Heidenreich, John

    2011-01-01

    The introduction of software technology in a life-dependent environment requires the development team to execute a process that ensures a high level of software reliability and correctness. Despite their popularity, agile methods are generally assumed to be inappropriate as a process family in these environments due to their lack of emphasis on documentation, traceability, and other formal techniques. Agile methods, notably Scrum, favor empirical process control, or small constant adjustments in a tight feedback loop. This paper challenges the assumption that agile methods are inappropriate for safety-critical software development. Agile methods are flexible enough to encourage the right amount of ceremony; therefore if safety-critical systems require greater emphasis on activities like formal specification and requirements management, then an agile process will include these as necessary activities. Furthermore, agile methods focus more on continuous process management and code-level quality than classic software engineering process models. We present our experiences on the image-guided surgical toolkit (IGSTK) project as a backdrop. IGSTK is an open source software project employing agile practices since 2004. We started with the assumption that a lighter process is better, focused on evolving code, and only adding process elements as the need arose. IGSTK has been adopted by teaching hospitals and research labs, and used for clinical trials. Agile methods have matured since the academic community suggested they are not suitable for safety-critical systems almost a decade ago, we present our experiences as a case study for renewing the discussion. PMID:21799545

  1. Formalization and Validation of Safety-Critical Requirements

    Directory of Open Access Journals (Sweden)

    Alessandro Cimatti

    2010-03-01

    Full Text Available The validation of requirements is a fundamental step in the development process of safety-critical systems. In safety critical applications such as aerospace, avionics and railways, the use of formal methods is of paramount importance both for requirements and for design validation. Nevertheless, while for the verification of the design, many formal techniques have been conceived and applied, the research on formal methods for requirements validation is not yet mature. The main obstacles are that, on the one hand, the correctness of requirements is not formally defined; on the other hand that the formalization and the validation of the requirements usually demands a strong involvement of domain experts. We report on a methodology and a series of techniques that we developed for the formalization and validation of high-level requirements for safety-critical applications. The main ingredients are a very expressive formal language and automatic satisfiability procedures. The language combines first-order, temporal, and hybrid logic. The satisfiability procedures are based on model checking and satisfiability modulo theory. We applied this technology within an industrial project to the validation of railways requirements.

  2. Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

    2005-01-01

    structure was found to be intact to maintain leak tightness even in the severe fire accidents. In addition, criticality safety was assessed by using the continuous energy Monte Carlo code MVP and the nuclear data library JENDL-3.2 for the cask in consideration of the mechanical damages and thermal failure resulted from the above analyses. As a conclusion, integrity of the packages of fresh nuclear fuel materials can be maintained even in the case of the severe accident, and criticality safety can also be secured to prevent radioactive material from releasing into environment. (author)

  3. 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.)

  4. Guide to verification and validation of the SCALE-4 criticality safety software

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

    Whenever a decision is made to newly install the SCALE nuclear criticality safety software on a computer system, the user should run a set of verification and validation (V&V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V&V in that it specifies test cases to run and gives expected results. The report describes the V&V that has been performed for the nuclear criticality safety software in a version of SCALE-4. The verification problems specified by the code developers have been run, and the results compare favorably with those in the SCALE 4.2 baseline. The results reported in this document are from the SCALE 4.2P version which was run on an IBM RS/6000 workstation. These results verify that the SCALE-4 nuclear criticality safety software has been correctly installed and is functioning properly. A validation has been performed for KENO V.a utilizing the CSAS25 criticality sequence and the SCALE 27-group cross-section library for {sup 233}U, {sup 235}U, and {sup 239}Pu fissile, systems in a broad range of geometries and fissile fuel forms. The experimental models used for the validation were taken from three previous validations of KENO V.a. A statistical analysis of the calculated results was used to determine the average calculational bias and a subcritical k{sub eff} criteria for each class of systems validated. Included the statistical analysis is a means of estimating the margin of subcriticality in k{sub eff}. This validation demonstrates that KENO V.a and the 27-group library may be used for nuclear criticality safety computations provided the system being analyzed falls within the range of the experiments used in the validation.

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. Safety criteria for spent-fuel transport. Final report

    International Nuclear Information System (INIS)

    Goldmann, K.; Gekler, W.C.

    1986-10-01

    The focus of this study is on the question, ''Do current regulations provide reasonable assurance of safety for a transport scenario of spent fuel, as presently anticipated by the Department of Energy, under the Nuclear Waste Policy Act.'' This question has been addressed by developing a methodology for identifying the expected frequency of Accidents Which Exceed Regulatory Conditions in Severity (AWERCS) for spent fuel transport casks and then assessing the health effects resulting from that frequency. By applying the methodology to an illustrative case of road transports, it was found that the accidental release of radioactive material from impact AWERCS would make negligible contributions to health effects associated with spent fuel transports by road. It is also concluded that the current regulatory drop test requirements in 10 CFR 71.51 which form the basis for cask design and were used to establish AWERCS screening criteria for this study are adequate, and that no basis was found to conclude that cask performance under expected road accident conditions represents an undue risk to the public

  11. Transport safety of high burnup fuel elements and HAW moulds

    International Nuclear Information System (INIS)

    Baier, G.; Hoermann, E.; Winter, M.

    1991-09-01

    The effect of changes of the peripheral conditions laid down in the project 'Safety studies on waste management' (PSE), on the assessment of transport safety of the transport container CASTOR IIa was analysed. Changes as against the PSE are due to the reprocessing of fuel elements abroad; increased burnup of fuel elements, and changes of transport conditions by rail. The higher burnup of fuel elements has only limited effects on the activity, thermal output, neutron and gamma radiation of the inventory. The analysis of inventory data shows that increased burnup has no serious influence on the container design. In the area of neutron shielding, toughening is required which, however, is possible by simply adding further polyethylene rods. An essential effect is the notable reduction of the transport volume from 155 transports in the past to 125 transports altogether now, with an increase of the medium burnup from 40 to 50 Gwd/tSM. That entails a reduction by 18% of the collective dose during normal operation, because above all the dose-intensive handling operations are reduced. The accident risk also decreases with increasing burnup, because the effect of a reduced transport volume leading to a lower accident frequency is stronger than that of inventory increase. Reprocessing abroad has only little influence of the transport distances to be covered altogether, and so on the normal loads and incident risks proportionate to the overall distance. With regard to the HAW glass forms, a slightly higher number of forms results from the lower loading of the glass with fission products prescribed in France. The number of HAW transports therefore rises from 16 to 19 (both by road and rail). The technical changes in the area of rail transport are characterized by higher speeds of goods trains and passenger trains. However, this does not yet apply to transports with heavy containers for which a speed limit of 100 km/h continues to be valid. (orig.) [de

  12. 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

  13. 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)

  14. 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.)

  15. 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.)

  16. 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

  17. 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)

  18. Some investigations on criticality safety using the Monte Carlo code OMEGA

    International Nuclear Information System (INIS)

    Seifert, E.

    1991-01-01

    The Monte Carlo method has proved very useful in solving problems of criticality safety. In the ZfK Rossendorf, the code OMEGA was developed by use of which the calculations presented in this paper were carried out. On the example of the RFR fuel transport container it has been studied which maximum value k eff reaches if ingress of water cannot be excluded. In this case the consideration of the detailed geometrical structure of the bulk of the container proves essential which is with no problems possible by using the OMEGA code. On the example of the experimentally critical facility RAKE it is shown that using the diffusion approximation may lead to noticeable errors. This cause of error will be eliminated by the Monte Carlo method from the first. (orig.) [de

  19. A Safety-Critical Java Technology Compatibility Kit

    DEFF Research Database (Denmark)

    Søndergaard, Hans; Korsholm, Stephan Erbs; Ravn, Anders P.

    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...

  20. Safety-Critical Java for Low-End Embedded Platforms

    DEFF Research Database (Denmark)

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

    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...

  1. 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....

  2. 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...

  3. 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...

  4. 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.

  5. 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

  6. 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

  7. Safety reassessment of the old installations involved in fuel cycle

    International Nuclear Information System (INIS)

    Guillard, M.

    2003-01-01

    Since the early 1990's, CEA (French atomic energy commission) has been preparing a plan for the renovation of a laboratory situated at Cadarache and dedicated to the study of irradiated materials and fuels. The main aim of this renovation was the improvement of the seismic behaviour of the laboratory since it was not built according to the para-seismic rules now in force. The solution chosen, given the different projects studied, the provisional unavailability of the plant and the related costs, was a partial reinforcement of the building in association with a limited plant life time and the reduction of activities in the oldest part of the installation. Another aim of this renovation was a global upgrading of the safety concerning: -) radioactive material containment (upgrade of the first static barrier by reinforcing cell leak-proofing, installation of a second level of very high efficiency filtration at the cell outputs, and separation of cell and general building ventilation networks; -) fire protection (fire sectoring with the isolation of the premises involving safety-important equipment, replacement of the automatic fire detection system, and definition of a new piloting of ventilation in case of fire); -) power cut risks (installation of permanent sources for the power supply of safety-important equipment); and -) earthquake behaviour (addition of reinforced connections between the 3 parts of the building, strengthening of peripheral walls, widening of joints between cells and building, and reinforcement of the foundation of the concrete cells). (A.C.)

  8. 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

  9. LPG diesel dual fuel engine – A critical review

    Directory of Open Access Journals (Sweden)

    B. Ashok

    2015-06-01

    Full Text Available The engine, which uses both conventional diesel fuel and LPG fuel, is referred to as ‘LPG–diesel dual fuel engines’. LPG dual fuel engines are modified diesel engines which use primary fuel as LPG and secondary fuel as diesel. LPG dual fuel engines have a good thermal efficiency at high output but the performance is less during part load conditions due to the poor utilization of charges. This problem can be overcome by varying factors such as pilot fuel quantity, injection timing, composition of the gaseous fuel and intake charge conditions, for improving the performance, combustion and emissions of dual fuel engines. This article reviews about the research work done by the researchers in order to improve the performance, combustion and emission parameters of a LPG–diesel dual fuel engines. From the studies it is shown that the use of LPG in diesel engine is one of the capable methods to reduce the PM and NOx emissions but at same time at part load condition there is a drop in efficiency and power output with respect to diesel operation.

  10. Investigation regarding the safety of handling the fuel assemblies for the nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    It was concluded previously that the general inspection of safety and the repair of shielding can be carried out as the fuel assemblies are charged, and the safety can be secured sufficiently. According to the decision by the meeting of cabinet ministers concerned with the nuclear ship ''Mutsu'', the Mutsu General Inspection and Repair Technology Investigation Committee investigated on the basic concept regarding the method and the safety of taking out, transporting and preserving the fuel assemblies. 112 fuel rods and 9 burnable poison rods are arranged into the square grid of 11 x 11 in a fuel assembly, and 32 fuel assemblies are employed. The contents of the investigation are the outline of the fuel assemblies, the present states of nuclear fission products, surface dose rate and soundness of the fuel assemblies, the safety of taking out, transporting and preserving the fuel assemblies, the measures required for securing the safety, and the place for taking out the fuel assemblies. In case of taking out, transporting and preserving the fuel assemblies, it is considered in view of the present state of the fuel assemblies that the safety can be secured sufficiently if the works are carried out carefully by taking the methods and conditions investigated into consideration. Also the committee reached already the conclusion described at the outset. (Kako, I.)

  11. Fuel Receiving and Storage Station. Nuclear Regulatory Commission's safety evaluation report

    International Nuclear Information System (INIS)

    1976-01-01

    The safety evaluation report covers design of structures, components, equipment, and systems; nuclear criticality safety; radiological safety; accident analysis; conduct of operations; quality assurance; common defense and security; financial qualifications; financial protection and indemnity requirements; and technical specifications

  12. Outline of results of safety research (in nuclear fuel cycle field in fiscal year 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The safety research in Power Reactor and Nuclear Fuel Development Corporation in fiscal year 1996 has been carried out based on the basic plan of safety research (from fiscal year 1996 to 2000) which was decided in March, 1996. In this report, on nuclear fuel cycle field, namely all the subjects in the fields of nuclear fuel facilities, environmental radioactivity and waste disposal, and the subjects related to nuclear fuel facilities among the fields of aseismatic and probabilistic safety assessments, the results of research in fiscal year 1996, the first year of the 5-year project, are summarized together with the outline of the basic plan of safety research. The basic policy, objective and system for promotion of the safety research are described. The objectives of the safety research are the advancement of safety technology, the safety of facilities, stable operation techniques, the safety design and the evaluation techniques of next generation facilities, and the support of transferring nuclear fuel cycle to private businesses. The objects of the research are uranium enrichment, fuel fabrication and reprocessing, and waste treatment and storage. 52 investigation papers of the results of the safety research in nuclear fuel cycle field in fiscal year 1996 are collected in this report. (K.I.)

  13. Existing experimental criticality data applicable to nuclear-fuel-transportation systems

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1983-02-01

    Analytical techniques are generally relied upon in making criticality evaluations involving nuclear material outside reactors. For these evaluations to be accepted the calculations must be validated by comparison with experimental data for a known set of conditions having physical and neutronic characteristics similar to those conditions being evaluated analytically. The purpose of this report is to identify those existing experimental data that are suitable for use in verifying criticality calculations on nuclear fuel transportation systems. In addition, near term needs for additional data in this area are identified. Of the considerable amount of criticality data currently existing, that are applicable to non-reactor systems, those particularly suitable for use in support of nuclear material transportation systems have been identified and catalogued into the following groups: (1) critical assemblies of fuel rods in water; (2) critical assemblies of fuel rods in water containing soluble neutron absorbers; (3) critical assemblies containing solid neutron absorber; (4) critical assemblies of fuel rods in water with heavy metal reflectors; and (5) critical assemblies of fuel rods in water with irregular features. A listing of the current near term needs for additional data in each of the groups has been developed for future use in planning criticality research in support of nuclear fuel transportation systems. The criticality experiments needed to provide these data are briefly described and identified according to priority and relative cost of performing the experiments

  14. 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

  15. Safety aspects of spent nuclear fuel interim storage installations

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2007-01-01

    Nowadays safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. In this kind of storage it is necessary to consider the physical security. Nuclear installations have become more vulnerable. New types of accidents must be considered in the design of these installations, which in the early days were not considered like: fissile material stolen, terrorists' acts and war conflicts, and traditional accidents concerning the transport of the spent fuel from the reactor to the storage location, earthquakes occurrence, airplanes crash, etc. Studies related to airplane falling had showed that a collision of big commercials airplanes at velocity of 800 km/h against SNF storage and specially designed concrete casks, do not result in serious structural injury to the casks, and not even radionuclides liberation to the environment. However, it was demonstrated that attacks with modern military ammunitions, against metallic casks, are calamitous. The casks could not support a direct impact of this ammo and the released radioactive materials can expose the workers and public as well the local environment to harmful radiation. This paper deals about the main basic aspects of a dry SNF storage installation, that must be physically well protected, getting barriers that difficult the access of unauthorized persons or vehicles, as well as, must structurally resist to incidents or accidents caused by unauthorized intrusion. (author)

  16. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Tsibulya, Anatoly; Rozhikhin, Yevgeniy

    2012-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

  17. 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

  18. 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

  19. A safety assessment of the SEAFP fuel cycle systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.; Ciattaglia, S.; Pace, L. di

    1995-01-01

    CFFTP and ENEA participated in a joint safety assessment of the fuel cycle design developed for the SEAFP fusion power reactor study (SEAFP: Safety and Environmental Assessment of Fusion Power). The assessment considered both conventional (deflagation/detonation) and radioactive hazards associated with the handling of significant quantities of hydrogen isotopes (H, D and T). Accordingly, the assessment focused on systems or equipment where either the flow rate, or inventory, of hydrogen isotopes was large. A systematic and thorough assessment of initiating events that can lead to an accidental release of tritium into the environment was the first step of the analysis process. This review demonstrated that, in all cases, there are at least two lines of defence available for mitigating the consequences of such accidents -i.e., secondary confinement (glove box, second pipe, caisson, etc.) and the building confinement, backed-up by an air detritiation capability. Therefore, large releases of tritium to the environment will occur only at very low frequencies. (orig.)

  20. Feasibility and incentives for the consideration of spent fuel operating histories in the criticality analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Westfall, R.M.; Jones, R.H.

    1987-08-01

    Analyses have been completed that indicate the consideration of spent fuel histories (''burnup credit'') in the design of spent fuel shipping casks is a justifiable concept that would result in cost savings and public risk benefits in the transport of spent nuclear fuel. Since cask capacities could be increased over those of casks without burnup credit, the number of shipments necessary to transport a given amount of fuel could be reduced. Reducing the number of shipments would increase safety benefits by reducing public and occupational exposure to both radiological and nonradiological risks associated with the transport of spent fuel. Economic benefits would include lower in-transit shipping, reduced transportation fleet capital costs, and reduced numbers of cask handling operations at both shipping and receiving facilities. 44 refs., 66 figs., 28 tabs

  1. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Paul; Neall, Fiona; Snellman, Margit; Pastina, Barbara; Nordman, Henrik; Johnson, Lawrence; Hjerpe, Thomas

    2008-03-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific RandD and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic 'difference analysis' of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the

  2. Safety-Critical Java on a Time-predictable Processor

    DEFF Research Database (Denmark)

    Korsholm, Stephan Erbs; 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....

  3. 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....

  4. New Criticality Safety Analysis Capabilities in SCALE 5.1

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, Stephen M [ORNL; DeHart, Mark D [ORNL; Dunn, Michael E [ORNL; Goluoglu, Sedat [ORNL; Horwedel, James E [ORNL; Petrie Jr, Lester M [ORNL; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL

    2007-01-01

    Version 5.1 of the SCALE computer software system developed at Oak Ridge National Laboratory, released in 2006, contains several significant enhancements for nuclear criticality safety analysis. This paper highlights new capabilities in SCALE 5.1, including improved resonance self-shielding capabilities; ENDF/B-VI.7 cross-section and covariance data libraries; HTML output for KENO V.a; analytical calculations of KENO-VI volumes with GeeWiz/KENO3D; new CENTRMST/PMCST modules for processing ENDF/B-VI data in TSUNAMI; SCALE Generalized Geometry Package in NEWT; KENO Monte Carlo depletion in TRITON; and plotting of cross-section and covariance data in Javapeno.

  5. 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.

  6. 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

  7. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel rods

    International Nuclear Information System (INIS)

    Hoovler, G.S.; Baldwin, M.N.

    1981-04-01

    Criticla arrays of 2.5%-enriched UO 2 fuel rods that simulate underwater rod storage of spent power reactor fuel are being constructed. Rod storage is a term used to describe a spent fuel storage concept in which the fuel bundles are disassembled and the rods are packed into specially designed cannisters. Rod storage would substantially increase the amount of fuel that could be stored in available space. These experiments are providing criticality data against which to benchmark nuclear codes used to design tightly packed rod storage racks

  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. Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

    International Nuclear Information System (INIS)

    Kelly, J.E.

    1995-01-01

    Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results

  10. Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.E.

    1995-02-08

    Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results.

  11. Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

    International Nuclear Information System (INIS)

    Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

    2014-01-01

    Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

  12. 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…

  13. 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

  14. OECD/NEA WGFCS Workshop: Safety Assessment of Fuel Cycle Facilities - Regulatory Approaches and Industry Perspectives

    International Nuclear Information System (INIS)

    2013-01-01

    Nuclear fuel is produced, processed, and stored mainly in industrial-scale facilities. Uranium ores are processed and refined to produce a pure uranium salt stream, Uranium is converted and enriched, nuclear fuel is fabricated (U fuel and U/Pu fuel for the closed cycle option); and spent fuel is stored and reprocessed in some countries (close cycle option). Facilities dedicated to the research and development of new fuel or new processes are also considered as Fuel Cycle Facilities. The safety assessment of nuclear facilities has often been led by the methodology and techniques initially developed for Nuclear Power Plants. As FCFs cover a wide diversity of installations the various approaches of national regulators, and their technical support organizations, for the Safety Assessment of Fuel Cycle Facilities are also diverse, as are the approaches by their industries in providing safety justifications for their facilities. The objective of the Working Group on Fuel Cycle Safety is to advance the understanding for both regulators and operators of relevant aspects of nuclear fuel cycle safety in member countries. A large amount of experience is available in safety assessment of FCFs, which should be shared to develop ideas in this field. To contribute to this task, the Workshop on 'Safety Assessment of Fuel Cycle Facilities - Regulatory Approaches and Industry Perspectives' was held in Toronto, on 27 - 29 September 2011. The workshop was hosted by Canadian Nuclear Safety Commission. The current proceedings provide summary of the results of the workshop with the text of the papers given and presentations made

  15. 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

  16. Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

    International Nuclear Information System (INIS)

    Taylor, L. L.; Loo, H. H.

    1999-01-01

    Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable

  17. 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)

  18. 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

  19. Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the ''fresh fuel'' assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions

  20. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  1. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  2. Sensitivity studies of some important parameters in the criticality of nuclear fuel pool

    International Nuclear Information System (INIS)

    Pina, C.

    1981-01-01

    The criticality study of spent fuels storage aiming to verify this security. A small description of the optimal moderation phenomena as well as the effect of stainless steel plates as neutron absorbed material, arre presented [pt

  3. Modeling the critical safety functions status tree of a NPP using FPGA

    International Nuclear Information System (INIS)

    Farias, Marcos Santana; Oliveira, Mauro Vitor de; Jaime, Guilherme Dutra Gonzaga; Almeida, Jose Carlos Soares de; Augusto, Silas Cordeiro

    2013-01-01

    Field Programmable Gate Arrays (FPGAs) based systems and equipment are beginning to appear in new plants I and C applications, as well as in retrofits for operating plants, in particular for safety applications due to their capability to face the systems obsolescence since they are circuit independent. The circuits implemented can be portable to different FPGAs architectures. Moreover, they reduce complexity for regulatory approval as compared to conventional microprocessor-based systems. Critical safety function (CSF) is the most significant design concept for prioritize operator actions for NPP based on the potential threat to the three barriers (fuel cladding, primary coolant system boundary, and containment) and allows the operator to respond to these threats prior to event diagnosis. CSF has a hierarchical information structure that organizes the system variables affecting the plant safety in terms of goal-means relations. This paper describes the application of FPGA in the implementation of the CSFs status tree logic for a Westinghouse 3-loops NPP simulator. (author)

  4. HSI for monitoring the critical safety functions status tree of a NPP

    International Nuclear Information System (INIS)

    Oliveira, Mauro Vitor de; Almeida, Jose Carlos Soares de; Augusto, Silas Cordeiro; Jaime, Guilherme Dutra Gonzaga

    2013-01-01

    Critical safety function (CSF) is the most significant design concept for prioritize operator actions based on the potential threat to the three barriers (fuel cladding, primary coolant system boundary, and containment) and allows the operator to respond to these threats prior to event diagnosis. CSF has a hierarchical information structure that organizes the system variables affecting the plant safety in terms of goal-means relations. It is important that the operator should be aware of various success paths associated with each CSF in order to respond to unanticipated system failures quickly. When an emergency occurs in NPPs, the operator should monitor CSFs periodically and identify possible success paths as necessary, and try to stabilize or safely shut down the plant using emergency operating procedure (EOP) that includes steps to check the CSFs. This implies that safety function status check may become a cognitively burdensome task that needs to be supported by proper information display. The advanced human-system interface (HSI) in nuclear power plants provides an information environment that supports the operators' burdensome cognitive tasks. This paper describes a CSFs interface design for supporting the operator's tasks to monitor and identify the associated success path for Westinghouse 3-loops NPP. (author)

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-09-20

    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.

  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. Radiation safety evaluation of the HANARO research reactor spent fuel disposal system

    Directory of Open Access Journals (Sweden)

    Kook Dong-Hak

    2011-01-01

    Full Text Available HANARO is a multi-purpose Korean research reactor with a high neutron flux and unique fuel types. Recently, a new conceptual disposal system for HANARO spent fuel has been suggested as a solution to the current spent fuel accumulation problem in the near future. This paper investigates HANARO spent fuel characteristics, evaluates radiation intensity by generating its own cross-section library and, finally, estimates the radiation safety of its disposal system. By comparing the absorbed dose of the engineering barrier with the recommended value, it was concluded that the proposed disposal system has a large enough safety margin.

  9. The Criticality Safety Information Resource Center at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, B.D.; Meade, R.A. [Los Alamos National Lab., NM (United States); Pruvost, N.L. [Galaxy Computer Services, Inc., Santa Fe, NM (United States)

    1997-05-01

    The mission of the Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory (LANL) is the preservation of primary documentation supporting criticality safety. In many cases, but not all, this primary documentation consists of experimentalists` logbooks. Experience has shown that the logbooks and other primary information are vulnerable to being discarded. Destruction of these logbooks results in a permanent loss to the criticality safety community.

  10. Safety risks of hydrogen fuel for applications in transportation vehicles.

    Science.gov (United States)

    2009-04-01

    Combustion of hydrocarbon fuels in many practical applications produces pollutants that are harmful to human health and environment. Hydrogen fuel is considered to be a potential answer to the clean energy demands, especially with the advances in fue...

  11. Critical assessment of power trains with fuel-cell systems and different fuels

    Science.gov (United States)

    Höhlein, B.; von Andrian, S.; Grube, Th; Menzer, R.

    Legal regulations (USA, EU) are a major driving force for intensifying technological developments with respect to the global automobile market. In the future, highly efficient vehicles with very low emission levels will include low-temperature fuel-cell systems (PEFC) as units of electric power trains. With alcohols, ether or hydrocarbons used as fuels for these new electric power trains, hydrogen as PEFC fuel has to be produced on board. These concepts including the direct use of methanol in fuel-cell systems, differ considerably in terms of both their development prospects and the results achieved so far. Based on process engineering analyses for net electricity generation in PEFC-powered power trains, as well as on assumptions for electric power trains and vehicle configurations, different fuel-cell performances and fuel processing units for octane, diesel, methanol, ethanol, propane and dimethylether have been evaluated as fuels. The possible benefits and key challenges for different solutions of power trains with fuel-cell systems/on-board hydrogen production and with direct methanol fuel-cell (DMFC) systems have been assessed. Locally, fuel-cell power trains are almost emission-free and, unlike battery-powered vehicles, their range is comparable to conventional vehicles. Therefore, they have application advantages cases of particularly stringent emission standards requiring zero emission. In comparison to internal combustion engines, using fuel-cell power trains can lead to clear reductions in primary energy demand and global, climate-relevant emissions providing the advantage of the efficiency of the hydrogen/air reaction in the fuel cell is not too drastically reduced by additional conversion steps of on-board hydrogen production, or by losses due to fuel supply provision.

  12. 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

  13. Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

    International Nuclear Information System (INIS)

    2012-12-01

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

  14. 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

  15. Integrated Design and Analysis Environment for Safety Critical Human-Automation Systems Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Flight deck systems, like many safety critical systems, often involve complex interactions between multiple human operators, automated subsystems, and physical...

  16. Investigating the safety of medication administration in adult critical care settings.

    Science.gov (United States)

    Mansour, Mansour; James, Veronica; Edgley, Alison

    2012-01-01

    Medication errors are recognized causes of patient morbidity and mortality in hospital settings, and can occur at any stage of the medication management process. Medication administration errors are reported to occur more frequently in critical care settings, and can be associated with severe consequences. However, patient safety research tends to focus on accident causations rather than organizational factors which enhance patient safety and health care resilience to unsafe practice. The Organizational Safety Space Model was developed for high-risk industries to investigate factors that influence organizational safety. Its application in health care settings may offer a unique approach to understand organizational safety in the health care context, particularly in investigating the safety of medication administration in adult critical care settings. This literature review explores the development and use of the Organizational Safety Space Model in the industrial context, and considers its application in investigating the safety of medication administration in adult critical care settings. SEARCH STRATEGIES (INCLUSION AND EXCLUSION CRITERIA): CINAHL, Medline, British Nursing Index (BNI) and PsychInfo databases were searched for peer-reviewed papers, published in English, from 1970 to 2011 with relevance to organizational safety and medication administration in critical care, using the key words: organization, safety, nurse, critical care and medication administration. Archaeological searching, including grey literature and governmental documents, was also carried out. From the identified 766 articles, 51 studies were considered relevant. The Organizational Safety Space Model offers a productive, conceptual system framework to critically analyse the wider organizational issues, which may influence the safety of medication administration and organizational resilience to accidents. However, the model needs to be evaluated for its application in health care settings in

  17. Design of fuel handling and storage systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2003-01-01

    The purpose of this Safety Guide is to provide recommendations on the design of fuel handling and storage systems for nuclear power plants. It presents recommendations on how to fulfil the requirements established in the Safety Requirements publication Safety of Nuclear Power Plants: Design. The scope of this Safety Guide is primarily the design of handling and storage systems for fuel assemblies associated with thermal nuclear reactors that are land based. It addresses all stages of fuel handling and storage, which include: the safe receipt of fuel at the nuclear power plant; the storage and inspection of fuel before use; the transfer of fresh fuel into the reactor; the removal of irradiated fuel from the reactor; the reinsertion of irradiated fuel when required; the storage, inspection and repair of the irradiated fuel and its preparation for removal from the reactor pool; the handling of the transport casks. Limited consideration is given to the handling and storage of certain core components, such as reactivity control devices. The recommendations of this Safety Guide also apply to other reactor types as appropriate, such as gas cooled reactors and reactors that are designed for on-load refuelling. Reference provides recommendations on the design of storage facilities for spent fuel, which are not an integral part of an operating nuclear power plant, although such facilities may be located on the same site. Such spent fuel storage facilities provide for the safe storage of spent nuclear fuel after it has been removed from the reactor pool and before it is reprocessed or disposed of as radioactive waste

  18. Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Univ. of New Mexico, Albuquerque, NM (United States). Nuclear Engineering Dept.

    2016-09-20

    These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such as MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.

  19. Development of nuclear fuel cycle technologies - bases of long-term provision of fuel and environmental safety of nuclear power

    International Nuclear Information System (INIS)

    Solonin, M.I.; Polyakov, A.S.; Zakharkin, B.S.; Smelov, V.S.; Nenarokomov, E.A.; Mukhin, I.V.

    2000-01-01

    To-day nuclear power is one of the options, however, to-morrow it may become the main source of the energy, thus, providing for the stable economic development for the long time to come. The availability of the large-scale nuclear power in the foreseeable future is governed by not only the safe operation of nuclear power plants (NPP) but also by the environmentally safe management of spent nuclear fuel, radioactive waste conditioning and long-term storage. More emphasis is to be placed to the closing of the fuel cycle in view of substantial quantities of spent nuclear fuel arisings. The once-through fuel cycle that is cost effective at the moment cannot be considered to be environmentally safe even for the middle term since the substantial build-up of spent nuclear fuel containing thousands of tons Pu will require the resolution of the safe management problem in the nearest future and is absolutely unjustified in terms of moral ethics as a transfer of the responsibility to future generations. The minimization of radioactive waste arisings and its radioactivity is only feasible with the closed fuel cycle put into practice and some actinides and long-lived fission radionuclides burnt out. The key issues in providing the environmentally safe fuel cycle are efficient processes of producing fuel for NPP, radionuclide after-burning included, a long-term spent nuclear fuel storage and reprocessing as well as radioactive waste management. The paper deals with the problems inherent in producing fuel for NPP with a view for the closed fuel cycle. Also discussed are options of the fuel cycle, its effectiveness and environmental safety with improvements in technologies of spent nuclear fuel reprocessing and long-lived radionuclide partitioning. (authors)

  20. 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

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-10-01

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

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

    International Nuclear Information System (INIS)

    Shedrow, C.B.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

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

  5. SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

    International Nuclear Information System (INIS)

    D. A. Thomas

    1996-01-01

    The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

  6. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    International Nuclear Information System (INIS)

    Madhuresh, R.; Nagarajan, R.; Jit, I.; Sanatkumar, A.

    1996-01-01

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D 2 0- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

  7. Safety and operations of hydrogen fuel infrastructure in northern climates : a collaborative complex systems approach.

    Science.gov (United States)

    2010-10-07

    "This project examined the safety and operation of hydrogen (H2) fueling system infrastructure in : northern climates. A multidisciplinary team lead by the University of Vermont (UVM), : combined with investigators from Zhejiang and Tsinghua Universi...

  8. Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

    2003-01-01

    In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

  9. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  10. Evaluation of safety, performance and emissions of synthetic fuel blends in a Cessna Citation II

    NARCIS (Netherlands)

    Snijders, T.A.; Melkert, J.A.

    2011-01-01

    Prior to being used in aviation, alternative fuels have to be tested thoroughly to ensure safe operation. At Delft University of Technology, a test programme was performed to evaluate the safety, performance and emissions of synthetic fuel blends. During test preparations, compatibility of the

  11. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

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

    International Nuclear Information System (INIS)

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

    1980-12-01

    The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents

  13. 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.

  14. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

    International Nuclear Information System (INIS)

    Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

    2008-12-01

    The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments that lie outside the

  15. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

    Energy Technology Data Exchange (ETDEWEB)

    Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

    2008-12-15

    The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments

  16. Nuclear safety philosophy and its general application to fuel management and handling - a regulator's viewpoint

    International Nuclear Information System (INIS)

    Petty, I.C.

    1995-01-01

    The Nuclear Safety Division (NSD) of the Health and Safety Executive (HSE) informs the UK Nuclear Industry of the principles that it applies in assessing whether licensees have demonstrated that their nuclear plants are as safe as is reasonably practicable. The paper commences with a discussion of the non-prescriptive approach to health and safety regulation which is the basis of the regulatory activities of NSD's operating arm -the Nuclear Installations Inspectorate (NII). It then describes in broad terms the overall approach used by NII for analysing the safety of nuclear plant, including fuel, which will cover both deterministic and probabilistic methodologies. The paper then introduces the sections of the Safety Assessment Principles which apply to nuclear fuel safety (both fuel handling and management). Most of these principles are of a general nature and do not just apply to fuel. The paper explains how safety cases might relate to the SAPs and offers some views on how a licensee might interpret them in developing his safety case. Particular emphasis is placed on the importance of submitting a high quality safety case and the type of information that should be in it. The advantages of the approach proposed, to the licensee as well as to the regulator, are identified. (author)

  17. Safety aspects related to burnup increase and mixed oxide fuel

    International Nuclear Information System (INIS)

    Thomas, W.

    1992-01-01

    The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

  18. The safety of storage facilities for LMFBR spent fuels

    International Nuclear Information System (INIS)

    Lefort, G.; Puit, J.C.

    1982-04-01

    Storage conditions for the cooling of fast neutron spent fuel assembly, before reprocessing, will be adapted specifically to this fuel cycle, but in a preliminary testing period it is better to take advantage of the industrial storage conditions already used for spent fuel of light water reactors, even if no extrapolation is possible in the future. In this aim fuel assemblies are dismantled, as soon as thermal conditions allow it, and pins are gathered in leak proof containers in a gas atmosphere and put in a cooling pool. This solution gives good results but studies and experiments are resumed, for future processing, for safe, easy and economical transport and storage conditions [fr

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  20. R and D activities at INR pitesti related to safety and reliability of CANDU type fuel

    International Nuclear Information System (INIS)

    Horhoianu, G.

    2003-01-01

    The focus of Nuclear Fuel R and D Program of Institute for Nuclear Research (INR) Pitesti is to maintain and improve the reliability, economics and safety of 37-element natural uranium CANDU fuel bundles in Cernavoda Nuclear Generating Station (CNGS). The second requirement is to improve the CANDU fuel design and to develop 43-element advanced fuel bundle that will reduce capital and fuelling cost, increase the operating and safety margins, improve natural - uranium utilization, and provide synergy with other reactor systems to improve resource utilization and spent fuel management. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history etc. has been obtained using in-pile measurements and PIE results of CANDU fuel elements irradiated in the TRIGA Material Testing Reactor (MTR) of INR Pitesti. In last time the data base was updated to include the results of Power Pulse Tests performed in TRIGA - Annular Core Pulse Reactor (ACPR) of INR Pitesti. One of the current research objective of our fuel bahaviour studies is to investigate the reliability behaviour of CANDU type fuel during power cycling operation condition. The INR research programme also include the out pile separate effects experiments to evaluate properties of the UO2 and cladding and development of computer models to describe sheath deformation and gas release processes. A program for LOCA simulating in-reactor tests is in progress at INR Pitesti to provide a database for verification of transient fuel performance codes and demonstrate that the significant fuel behaviour phenomena have all been included in the models.This data base is used extensively for the validation of the fuel behaviour codes. This paper summarizes R and D activities of INR Pitesti, related to safety and reliability of CANDU type fuel and presents some of the recent results obtained from in reactor tests. (author)

  1. Final safety analysis report for the irradiated fuels storage facility

    International Nuclear Information System (INIS)

    Bingham, G.E.; Evans, T.K.

    1976-01-01

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

  2. A relative risk comparison of criticality control strategies based on fresh fuel and burnup credit design bases

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1988-01-01

    The proposed use of burnup credit in spent fuel cask design and operation represents a departure from current regulatory practice, and creates technical issues that ultimately must be resolved for the concept to be implemented. Issues related to specific technical considerations can generally be resolved conclusively. However, an underlying perception may still exist that the use of burnup credit compromises criticality safety. In practice, individual casks are designed to satisfy regulatory requirements in a generally conservative manner. The designer's application of the regulatory requirements involves some engineering judgement, as does the regulator's implementation of them. This does not have an adverse effect on safety, but does make it difficult to objectively compare new or alternative designs and/or operating approaches. 5 refs., 7 figs., 2 tabs

  3. Reassessment of gadolinium odd isotopes neutron cross sections: scientific motivations and sensitivity-uncertainty analysis on LWR fuel assembly criticality calculations

    Directory of Open Access Journals (Sweden)

    Rocchi Federico

    2017-01-01

    Full Text Available Gadolinium odd isotopes cross sections are crucial in assessing the neutronic performance and safety features of a light water reactor (LWR core. Accurate evaluations of the neutron capture behavior of gadolinium burnable poisons are necessary for a precise estimation of the economic gain due to the extension of fuel life, the residual reactivity penalty at the end of life, and the reactivity peak for partially spent fuel for the criticality safety analysis of Spent Fuel Pools. Nevertheless, present gadolinium odd isotopes neutron cross sections are somehow dated and poorly investigated in the high sensitivity thermal energy region and are available with an uncertainty which is too high in comparison to the present day typical industrial standards and needs. This article shows how the most recent gadolinium cross sections evaluations appear inadequate to provide accurate criticality calculations for a system with gadolinium fuel pins. In this article, a sensitivity and uncertainty analysis (S/U has been performed to investigate the effect of gadolinium odd isotopes nuclear cross sections data on the multiplication factor of some LWR fuel assemblies. The results have shown the importance of gadolinium odd isotopes in the criticality evaluation, and they confirmed the need of a re-evaluation of the neutron capture cross sections by means of new experimental measurements to be carried out at the n_TOF facility at CERN.

  4. Prediction of critical heat flux in fuel assemblies using a CHF table method

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; Hwang, Dae Hyun; Bang, Je Geon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Baek, Won Pil; Chang, Soon Heung [Korea Advance Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor. 8 refs., 3 figs., 3 tabs. (Author)

  5. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management

    International Nuclear Information System (INIS)

    1997-01-01

    The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was adopted on 5 September 1997 by a Diplomatic Conference convened by the IAEA from 1 to 5 September 1997. The Joint Convention was opened for signature at Vienna on 29 September 1997 during the forty-first session of the General Conference of the IAEA. This document reproduces the text of the Convention

  6. Fuel safety criteria in NEA member countries - Compilation of responses received from member countries

    International Nuclear Information System (INIS)

    2003-03-01

    In 2001 the Committee on the Safety of Nuclear Installations (CSNI) issued a report on Fuel Safety Criteria Technical Review. The objective was to review the present fuel safety criteria and judge to which extent they are affected by the 'new' design elements, such as different cladding materials, higher burnup, the use of MOX fuels, etc. The report stated that the current framework of fuel safety criteria remains generally applicable, being largely unaffected by the 'new' or modern design elements. The levels (numbers) in the individual safety criteria may, however, change in accordance with the particular fuel and core design features. Some of these levels have already been - or are continuously being - adjusted. The level adjustments of several other criteria (RIA, LOCA) also appears to be needed, on the basis of experimental data and the analysis thereof. As a follow-up, among its first tasks, the CSNI Special Expert Group on Fuel Safety Margins (SEG FSM) initiated the collection of information on the present fuel safety criteria used in NEA member states with the objective to solicit national practices in the use of fuel safety criteria, in particular to get information on their specific national levels/values, including their recent adjustments, and to identify the differences and commonalties between the different countries. Two sources of information were used to produce this report: a compilation of responses to a questionnaire prepared for the June 2000 CNRA meeting, and individual responses from the SEGFSM members to the new revised questionnaire issued by the task Force preparing this report. In accordance with the latter, the fuel safety criteria discussed in this report were divided into three categories: (A) safety criteria - criteria imposed by the regulator; (B) operational criteria - specific to the fuel design and provided by the fuel vendor as part of the licensing basis; (C) design criteria - limits employed by vendors and/or utilities for fuel

  7. 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)

  8. 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...

  9. 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.

  10. 30 CFR 75.1904 - Underground diesel fuel tanks and safety cans.

    Science.gov (United States)

    2010-07-01

    ... spout; (3) Provided with a vent valve designed to open and close simultaneously and automatically with... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Underground diesel fuel tanks and safety cans. 75.1904 Section 75.1904 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR...

  11. 30 CFR 75.1903 - Underground diesel fuel storage facilities and areas; construction and safety precautions.

    Science.gov (United States)

    2010-07-01

    ... HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS... storage; and (4) Maintained to prevent the accumulation of water. (c) Welding or cutting other than that... contained diesel fuel, these practices shall be followed: (1) Cutting or welding shall not be performed on...

  12. Fuel and canister process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  13. Fuel and canister process report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    Werme, Lars; Lilja, Christina

    2010-12-01

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  14. Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

    Energy Technology Data Exchange (ETDEWEB)

    Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

  15. Safety concerning the partial alteration of the fuel reprocessing facility in Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    A report of the Committee on Examination of Nuclear Fuel Safety was presented to the Atomic Energy Commission of Japan, which is concerned with the safety in the partial alteration of the fuel reprocessing facility in the Power Reactor and Nuclear Fuel Development Corporation. Its safety was confirmed. The alteration is as follows. In order to raise the storage capacity of low-level radioactive solid wastes, the second storage is installed. The building for low-level solid waste storage is a ferro-concrete structure, total floor space is 1,260 m 2 , two-storied with one basement floor. The basement floor is for storage, the first floor for receiving, storage, etc., and the second floor for storage and machinery room. The capacity of the second storage is for about 9,500 200 l drums (equivalent to about four years waste discharge). The contents of the safety examination are safety of facilities such as aseismatic property and radiation protection, and effects on the surrounding environment. (Mori, K

  16. Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

    2006-01-01

    According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

  17. Technology Development for Integrated Safety Test of Spent Nuclear Fuel Transportation and Storage System

    International Nuclear Information System (INIS)

    Seo, Kiseog; Seo, J. S.; Lee, J. C.

    2012-05-01

    A dedicated review on the U. S. NRC Regulation 10 CFR Part 72 and regulatory guide NUREG/1536 has been performed. The safety requirements for spent nuclear fuel dry storage cask are analyzed and summarized in structural, thermal, shielding, criticality, materials, tests and maintenance aspects. Also a guideline for preparing the safety analysis report is provided. The heat flow analysis was performed by varying the dimensions of the heat flow test facility. From the heat flow analysis for the test facility, as the test facility became test facility. From the heat flow analysis for the test facility, as the test facility became bigger; the thermal effect became smaller. Therefore, the dimensions of the heat flow test facility was designed with 5m Χ 5m Χ 6m(H). Analyses of heat transfer characteristics and mechanism for spent PWR fuel assemblies, option study for production of the effective thermal conductivity and option study for effective thermal conductivity test have been performed to obtain the basic data for production of the effective thermal conductivity. It became clear that the diffusion coefficient of chloride ion of concrete remarkably increases along with the temperature rise, and that there is a linear relation between the logarithm values of the diffusion coefficients and the reciprocal of the temperature. It is understood to be able to express the temperature dependency of the diffusion coefficient roughly by an Arrhenius equation as the velocity coefficient is provided as the diffusion coefficient. The specifications and characteristics of storage facilities under operation including dual purpose casks were investigated. Components subject to material degradation were examined. Based on literature survey, investigating a drop analysis incorporating with material degradation, the basic data to develop an analysis methodology was obtained

  18. The fourth country report on agreement of safety supervision on radiation waste and management of spent fuel

    International Nuclear Information System (INIS)

    2011-10-01

    This report covered the agreement of safety supervision on radiation waste and management of spent fuel. It listed the stipulation, the common law, the coverage and the amount of stock on spent fuel. Also, it indicated law and regulations and restriction on the related the agency, general safety regulations, policy on guarantee of quality, emergency method, dismantling, management of safety control on spent fuel including a process of establishment and safety requirements, regulations of conveyance between countries and improvement of safety of spent fuel.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. Aviation Fuel System Reliability and Fail-Safety Analysis. Promising Alternative Ways for Improving the Fuel System Reliability

    Directory of Open Access Journals (Sweden)

    I. S. Shumilov

    2017-01-01

    Full Text Available The paper deals with design requirements for an aviation fuel system (AFS, AFS basic design requirements, reliability, and design precautions to avoid AFS failure. Compares the reliability and fail-safety of AFS and aircraft hydraulic system (AHS, considers the promising alternative ways to raise reliability of fuel systems, as well as elaborates recommendations to improve reliability of the pipeline system components and pipeline systems, in general, based on the selection of design solutions.It is extremely advisable to design the AFS and AHS in accordance with Aviation Regulations АП25 and Accident Prevention Guidelines, ICAO (International Civil Aviation Association, which will reduce risk of emergency situations, and in some cases even avoid heavy disasters.ATS and AHS designs should be based on the uniform principles to ensure the highest reliability and safety. However, currently, this principle is not enough kept, and AFS looses in reliability and fail-safety as compared with AHS. When there are the examined failures (single and their combinations the guidelines to ensure the AFS efficiency should be the same as those of norm-adopted in the Regulations АП25 for AHS. This will significantly increase reliability and fail-safety of the fuel systems and aircraft flights, in general, despite a slight increase in AFS mass.The proposed improvements through the use of components redundancy of the fuel system will greatly raise reliability of the fuel system of a passenger aircraft, which will, without serious consequences for the flight, withstand up to 2 failures, its reliability and fail-safety design will be similar to those of the AHS, however, above improvement measures will lead to a slightly increasing total mass of the fuel system.It is advisable to set a second pump on the engine in parallel with the first one. It will run in case the first one fails for some reasons. The second pump, like the first pump, can be driven from the

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

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1995-11-01

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

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1995-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-06-01

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

  7. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    International Nuclear Information System (INIS)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

  8. Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities (4). Investigation of safety evaluation method for fire and explosion incidents

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Tashiro, Shinsuke; Ueda, Yoshinori

    2010-01-01

    A special committee on 'Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)' was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to the establishment of quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution, and fire (including rapid decomposition of TBP complexes), resulting in the release of radio active materials into the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this technical report, the research results about basic experimental data and the method for safety evaluation of fire and explosion incidents were summarized. (author)

  9. Models for Microbial Fuel Cells: A critical review

    Science.gov (United States)

    Xia, Chengshuo; Zhang, Daxing; Pedrycz, Witold; Zhu, Yingmin; Guo, Yongxian

    2018-01-01

    Microbial fuel cells (MFCs) have been widely viewed as one of the most promising alternative sources of renewable energy. A recognition of needs of efficient development methods based on multidisciplinary research becomes crucial for the optimization of MFCs. Modeling of MFCs is an effective way for not only gaining a thorough understanding of the effects of operation conditions on the performance of power generation but also becomes of essential interest to the successful implementation of MFCs. The MFC models encompass the underlying reaction process and limiting factors of the MFC. The models come in various forms, such as the mathematical equations or the equivalent circuits. Different modeling focuses and approaches of the MFC have emerged. In this study, we present a state of the art of MFCs modeling; the past modeling methods are reviewed as well. Models and modeling methods are elaborated on based on the classification provided by Mechanism-based models and Application-based models. Mechanisms, advantages, drawbacks, and application fields of different models are illustrated as well. We exhibit a complete and comprehensive exposition of the different models for MFCs and offer further guidance to promote the performance of MFCs.

  10. Management concepts and safety applications for nuclear fuel facilities

    Energy Technology Data Exchange (ETDEWEB)

    Eisner, H.; Scotti, R.S. [George Washington Univ., Washington, DC (United States). School of Engineering and Applied Science; Delicate, W.S. [KEVRIC Co., Inc., Silver Spring, MD (United States)

    1995-05-01

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

  11. Management concepts and safety applications for nuclear fuel facilities

    International Nuclear Information System (INIS)

    Eisner, H.; Scotti, R.S.

    1995-05-01

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

  12. Achievements and Perspectives of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Louvat, D.; Lacoste, A.C.

    2006-01-01

    The Joint Convention on the Safety of Spent Fuel management and on the Safety of Radioactive Waste Management is the first legal instrument to directly address the safety of spent fuel and radioactive waste management on a global scale. The Joint Convention entered into force in 2001. This paper describes its process and its main achievements to date. The perspectives to establish of a Global Waste Safety Regime based on the Joint Convention are also discussed. (authors)

  13. Influence of single-phase heat transfer correlations on safety analysis of research reactors with narrow rectangular fuel channels

    International Nuclear Information System (INIS)

    Rawashdeh, A.; Altamimi, R.; Lee, B.; Chung, Y. J.; Park, S.

    2013-01-01

    The influence of different single-phase heat transfer correlations on the fuel temperature and minimum critical heat flux ratio (MCHFR) during a typical accident of a 5 MW research reactor is investigated. A reactor uses plate type fuel, of which the cooling channels have a narrow rectangular shape. RELAP5/MOD3.3 tends to over-predict the Nusselt number (Nu) at a low Reynolds number (Re) region, and therefore the correlation set is modified to properly describe the thermal behavior at that region. To demonstrate the effect of Nu at a low-Re region on an accident analysis, a two-pump failure accident was chosen as a sample problem. In the accident, the downward core flow decreases by a pump coast-down, and then reverses upward by natural convection. During the pump coast-down and flow reversal, the flow undergoes a laminar flow regime which has a different Nu with respect to the correlation sets. Compared to the results by the original RELAP5/MOD3.3, the modified correlation set predicts the fuel temperature to be a little higher than the original value, and the MCHFR to be a little lower than the original value. Although the modified correlation set predicts the fuel temperature and the MCHFR to be less conservative than those calculated from the original correlation of RELAP5/MOD3.3, the maximum fuel temperature and the MCHFR still satisfy the safety acceptance criteria

  14. The regulatory process, nuclear safety research and the fuel cycle in the United Kingdom

    International Nuclear Information System (INIS)

    Watson, P.

    1996-01-01

    The main legislation governing the safety of nuclear installations in the United Kingdom is the Health and Safety at Work Act 1974 (HSWA) and the associated relevant statutory provisions of the Nuclear Installations Act 1965 (as amended). The HSWA sought to simplify and unify all industrial safety legislation and set in place the Health and Safety Commission (HSC) and its executive arm, the Health and Safety Executive (HSE). The Health and Safety Executive's Nuclear Safety Division (NSD) regulates the nuclear activities on such sites through HM Nuclear Installations Inspectorate (NII). Under the Nuclear Installations Act (NIA) no corporate body may use any site for the purpose of installing or operating any reactor, other than such a reactor comprised in a means of transport, or other prescribed installation unless the operator has been granted a nuclear site licence by the Health and Safety Executive. Nuclear fuel cycle facilities are examples of such prescribed installations. (J.P.N.)

  15. 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)

  16. 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.

  17. Hodel fuels controversy: critics say campaign remarks damage waste program

    International Nuclear Information System (INIS)

    Harris, T.K.

    1984-01-01

    Critics of the federal effort to seek a disposal site for radioactive wastes complain that Energy Secretary Hodel politicized the process during the 1984 election campaign when he spoke on behalf of Texas Senator Phil Gramm, who opposes Texas as the first site for development. Officials from other potential sites argue that the decision should be based on technical and objective criteria, not political pressure. Besides the political tone, Hodel's comments in Texas imply that a presidential decision can be affected by state objections, which misinterprets the Nuclear Waste Policy Act of 1982. States which are potential repository sites have raised issues which can form the basis for future litigation and taken other steps that could lead to prolonged debate over monitored retrieval storage (MRS). The basic issue, however, is whether or not the US can deal with the nuclear waste problem

  18. Analysis of differences in fuel safety criteria for WWER and western PWR nuclear power plants

    International Nuclear Information System (INIS)

    2003-11-01

    In 2001 the OECD issued a report of the NEA/CSNI (Committee on the Safety of Nuclear Installations) Task Force on the existing safety criteria for reactor fuel for western LWR nuclear power plants (both for PWRs and BWRs) under new design elements. Likewise in 2001, the IAEA released a report by a Working Group on the existing safety criteria for reactor fuel for WWER nuclear power plants under new design requirements. However, it was found that it was not possible to compare the two sets of criteria on the basis upon which they had been established. Therefore, the IAEA initiated an assessment of the common features and differences in fuel safety criteria between plants of eastern and western design, focusing on western PWRs and eastern WWER reactors. Between October 2000 and November 2001, the IAEA organized several workshops with representatives from eastern and western European countries in which the current fuel safety related criteria for PWR and WWER reactors were reviewed and compared. The workshops brought together expert representatives from the Russian Federation, from the Ukraine and from western countries that operate PWRs. The first workshop focused on a general overview of the fuel safety criteria in order for all representatives to appreciate the various criteria and their respective bases. The second workshop (which involved one western and one eastern expert) concentrated on addressing and explaining the differences observed, and documenting all these results in preparation for a panel discussion. This panel discussion took place during the third workshop, where the previously obtained results were reviewed in detail and final recommendations were made. This report documents the findings of the workshops. It highlights the common features and differences between PWR and WWER fuel, and may serve as a general basis for the safety evaluation of these fuels. Therefore, it will be very beneficial for licensing activities for PWR and WWER plants, as it

  19. Safety criteria for wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    Vitkova, M.; Gorinov, I.; Dacheva, D.

    2006-01-01

    An overview of the safety criteria for both type of storage (wet and dry) is presented in this paper. The main criteria - thermal, shielding, radiological - are discussed in details. Some requirement to the materials and the operational regimes are also discussed. An attempt for comparison of the safety criteria and general problems for these both type of storage is made

  20. National report of the Slovak Republic compiled in terms of the Join on the safety of spent fuel management and on the safety of radwaste management

    International Nuclear Information System (INIS)

    Jurina, V.; Viktory, D.; Petrik, T.; Sovcik, J.; Suess, J.; Tomek, J.; Lukacovic, J.; Ivan, J.; Ziakova, M.; Metke, E.; Pospisil, M.; Turner, M.; Homola, J.; Vaclav, J.; Bystricka, S.; Barbaric, M.; Horvath, J.; Betak, J.; Mihaly, B.; Adamovsky, V.; Baloghova, A.; Orihel, M.; Vasina, D.; Balaz, J.; Misovicova, D.; Vrtoch, M.; Mlcuch, J.; Granak, P.; Meleg, J.; Bardy, M.; Gogoliak, J.

    2011-08-01

    The National safety report of the Slovak Republic on the safety of spent fuel management and on the safety of radwaste management in 2011 is presented. These activities in the safety of spent fuel management and radioactive waste management in the Slovak Republic are reported under the headings: (A) Introduction; B) Concept for spent nuclear fuel management (SNF) and radwaste management (RAW); (C) Scope of application of the convention; (D) Spent fuel management and radioactive waste (RAW) management facilities; (E) Legislation and regulation; (F) General safety provisions; (G) Safety of spent fuel management; (H) Safety of radioactive waste (RAW) management; (I) Transboundary movement of spent nuclear fuel and radioactive waste; (J) Disused sealed sources; (K) Planned measures to improve safety; (L) Communication with the public; (M) Annexes. Annexes consists of following parts: I. List of nuclear facilities for spent fuel and RAW management. II. Limits of radioactive material discharges into atmosphere and hydrosphere. III. List of nuclear installations in decommissioning. IV. Inventory of stored spent nuclear fuel. V. Inventory of stored RAW. VI. List of national laws, decrees and guidelines. VII. List of international expert reports (including safety reports). VIII. List of authors.