Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
Wagner, J.C.; DeHart, M.D.
2000-03-01
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.
VVER-related burnup credit calculations
The calculations related to a VVER burnup credit calculational benchmark proposed to the Eastern and Central European research community in collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmark Working Group (working under WPNCS - Working Party on Nuclear Criticality Safety) are described. The results of a three-year effort by analysts from the Czech Republic, Finland, Germany, Hungary, Russia, Slovakia and the United Kingdom are summarized and commented on. (author)
A burnup credit calculation methodology for PWR spent fuel transportation
A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)
Study on the conservative factors for burnup credit criticality calculation
When applies the burnup credit technology to perform criticality safety analysis for spent fuel storage or transportation problems, it is important for one to confirm that all the conditions adopted are adequate to cover the severest conditions that may encounter in the engineering applications. Taking the OECD/NEA burnup credit criticality benchmarks as sample problems, we study the effect of some important factors that may affect the conservatism of' the results for spent fuel system criticality safety analysis. Effects caused by different nuclides credit strategy, different cooling time and axial burnup profile are studied by use of the STARBUCS module of SCALE5. 1 software package, and related conclusions about the conservatism of these factors are drawn. (authors)
Calculation study of TNPS spent fuel pool using burnup credit
Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)
Results of the isotopic concentrations of VVER calculational burnup credit benchmark No. 2(CB2)
Results of the nuclide concentrations are presented of VVER Burnup Credit Benchmark No. 2(CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is summarized. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2' code and other codes were used for the calculation of the spent fuel concentration. (author)
Burnup credit calculations on long-term disposal
One of the considered options for handling of irradiated nuclear fuel is the final disposal in some kind of repository. This necessitates the long-term investigation of subcriticality, heat production, public dose etc. NEA WPNCS Burnup Credit Expert Group defined a new benchmark to test the codes and data used for such problems. The effect of cooling time should be investigated. This implies that the decay data and not the cross sections influence the results. Composition of 4.5 % UO2 fuel with 50 MWd/kgU is given at the assembly removal from the core. Change of composition should be evaluated for 30 values of cooling time up to 1 million years. Keff should be evaluated with these compositions for a container housing 21 fuel assemblies. Initial concentration of 115 isotopes is given. For criticality calculations the usual 'burnup credit set' is used (14 actinides and 15 fission products). Results for additional isotopes is not presented now. The investigated fuel is 17 x 17 PWR UO2 type, with 25 guide tubes. The selected cooling times covers the time intervals of the usual handling procedures around the reactors (few years storing in storage pool, transport), interim storage (hundred years), and the long time scale of disposal up to 1 million years. Results: 1) For major actinides, ORIGEN and MULTICELL based keff results are practically identical up to 1000 years, far beyond the cooling times it was intended. 2) For actinides and fission products, the agreement is excellent up to 100 years, which covers the interim storage. 3) The difference of keff results about 0.02 at 1000 years. The reason is mainly the presence of Np-237, not considered in the previous case. It is produced from Am-241 by α-decay (432 years). Compositions calculated by ORIGEN and TIBSO results the same keff values for cooling times up to 1 million years. Changes in keff with cooling time have clear physical explanation. Compositions calculated by ORIGEN and MULTICELL results the same keff
Results of the isotopic concentrations of WWER calculation Burnup Credit Benchmark NO.2 (CB2)
The purpose of this document is to present the results of the nuclide concentrations of the WWER Burnup Credit Benchmark NO.2 (CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the WWER burnup credit benchmark is summarized in [1]. The CB2 benchmark focused on WWER burnup credit study proposed on the 97' Atomic Energy Research symposium [2]. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2'[3] code was used for the calculation of the spent fuel concentration. This work also comprises the results obtained by other codes [4]. (Author)
Results of the isotopic concentrations of VVER calculational burnup credit benchmark no. 2(cb2
The characterization of the irradiated fuel materials is becoming more important with the Increasing use of nuclear energy in the world. The purpose of this document is to present the results of the nuclide concentrations calculated Using Calculation VVER Burnup Credit Benchmark No. 2(CB2). The calculations were Performed in The Nuclear Technology Center of Cuba. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is Summarized in [1]. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium [2]. It should provide a comparison of the ability of various code systems And data libraries to predict VVER-440 spent fuel isotopes (isotopic concentrations) using Depletion analysis. This phase of the benchmark calculations is still in progress. CB2 should be finished by summer 1999 and evaluated results could be presented on the next AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and Cooling time. The depletion point ORIGEN2[3] code was used for the calculation of the spent Fuel concentration. The depletion analysis was performed using the VVER-440 irradiated fuel assemblies with in-core Irradiation time of 3 years, burnup of the 30000 mwd/TU, and an after discharge cooling Time of 0 and 1 year. This work also comprises the results obtained by other codes[4].
OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results
Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are 149Sm, 151Sm, and 155Gd
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results
In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)
Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-02-01
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)
Calculation of the CB1 burnup credit benchmark reaction rates with MCNP4B
The first calculational VVER-440 burnup credit benchmark CB1 in 1996. VTT Energy participated in the calculation of the CB1 benchmark with three different codes: CASMO-4, KENO-VI and MCNP4B. However, the reaction rates and the fission ν were calculated only with CASMO-4. Now, the neutron absorption and production reaction rates and the fission ν values have been calculated at VTT Energy with the MCNP4B Monte Carlo code using the ENDF60 neutron data library. (author)
Spent fuel pool storage calculations using the ISOCRIT burnup credit tool
Highlights: ► Depletion isotopics are needed for burnup credit in spent fuel pool analyses. ► We developed ISOCRIT to generate the isotopics using conservative depletion assumptions. ► ISOCRIT works in an automated fashion passing data between lattice physics and 3D Monte Carlo codes. ► Analyses to assess the impact of different depletion parameters on the reactivity of the spent fuel in pool conditions. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse’s state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.
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. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)
In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor keff (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations
Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual keff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data
Monte Carlo calculations of the REBUS critical experiment for validation of burnup credit
The application of burnup credit (BUC) to criticality safety analysis for Spent Nuclear Fuel (SNF) configurations requires the implementation of both estimation of the SNF composition with the aid of depletion calculation tools and estimation of the SNF reactivity with the aid of criticality calculation tools. Amongst the several experimental programs dedicated to the validation of both calculation tools, REBUS is distinguished by a combination of chemical analysis and critical experiment. In addition to detailed assays of irradiated fuel, the reactivity worth of the fuel rods under investigation is measured both before and after irradiation. Since a whole bundle of fuel rods is used in the experiment, the change in reactivity is significant enough to be observable by Monte Carlo calculations. Thus, the calculation tools which see the most widespread use in SNF critical safety applications can be validated directly. Apart from the effective neutron multiplication factor keff, REBUS also provides measurements of the flux and fission rate distributions. While the program comprises investigation of commercial UO2 fuel rods and mixed oxide (MOX) fuel from a research reactor, the presentation will focus on the commercial UO2 fuel with an overview of the experimental setup and first results from the analysis. (author)
The DECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide (UOX) fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The exercises consisted of inventory calculations of MOX fuels for two initial plutonium compositions. The depletion calculations were carried out using three representations of the MOX assemblies and their interface with UOX assemblies. This enabled the investigation of the spatial and spectral effects during the irradiation of the MOX fuels. (author)
Reverdy, L
1999-07-01
Nowadays optimization goes with everything. So French engineering firms try to demonstrate that fuel transport casks and storage pools are able to receive assemblies with higher {sup 235}U initial enrichments. Fuel Burnup distribution contributes to demonstrate it. This instruction has to elaborate a way to take credit of burnup effects on criticality safety designs. The calculation codes used are CESAR 4.21-APOLLO 1-MORET III. The assembly studied (UO{sub 2}) is irradiated in a French Pressurized Water Reactor like EDF nuclear power reactor: PWR 1300 MWe, 17 x 17 array. Its initial enrichment in {sup 235}U equals 4.5%. The studies exposed in this report have evaluated the effects of: (i) the 15 fission products considered in Burnup Credit ({sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 109}Ag, {sup 133}Cs, {sup 143}Nd, {sup 145}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd), (ii) the calculated abundances corrected or not by fixed factors, (iii) the choice of one cross sections library used by CESAR 4.21, (iv) the zone number elected in the axial burnup distribution zoning, (v) the kind of cut applied on (regular/optimized). Two axial distribution profiles are studied: one with 44 GWd/t average burnup, the other with 20 GWd/t average burnup. The second one considers a shallow control rods insertion in the upper limit of the assembly. The results show a margin in reactivity about 0.045 with consideration of the 6 most absorbent fission products ({sup 103}Rh, {sup 133}Cs, {sup 143}Nd, {sup 149}Sm, {sup 152}Sm, {sup 155}Gd), and about 0.06 for all Burnup Credit fission products whole. Those results have been calculated with an average burnup of 44 GWj/t. In a conservative approach, corrective factors must be apply on the abundance of some fission products. The cross sections library used by CESAR 4.21 (BBL 4) is sufficient and gives satisfactory results. The zoning of the assembly axial distribution burnup in 9
Burnup credit implementation in spent fuel management
The criticality safety analysis of spent fuel management systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. The concept of allowing reactivity credit for spent fuel offers economic incentives. Burnup Credit (BUC) could reduce mass limitation during dissolution of highly enriched PWR assemblies at the La Hague reprocessing plant. Furthermore, accounting for burnup credit enables the operator to avoid the use of Gd soluble poison in the dissolver for MOX assemblies. Analyses performed by DOE and its contractors have indicated that using BUC to maximize spent fuel transportation cask capacities is a justifiable concept that would result in public risk benefits and cost savings while fully maintaining criticality safety margins. In order to allow for Fission Products and Actinides in Criticality-Safety analyses, an extensive BUC experimental programme has been developed in France in the framework of the CEA-COGEMA collaboration. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Independent measurement systems, e.g. gamma spectrum detection systems, are needed to perform a true independent measurement of assembly burnup, without reliance on reactor records, using the gamma emission signatures fission products (mainly Cesium isotopes). (author)
The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (keff) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of keff. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of keff values calculated by the participants from the mean value is almost within the band of ±1%Δk/k. The deviations from the averaged calculated fission rate profiles are found to be within ±5% for most cases. (author)
Implementation of burnup credit in spent fuel management systems
Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate safety margin. The motivation for using burnup credit in criticality safety applications is based on economic considerations and additional benefits contributing to public health and safety and resource conservation. Interest in the implementation of burnup credit has been shown by many countries. In 1997, the International Atomic Energy Agency (IAEA) started a task to monitor the implementation of burnup credit in spent fuel management systems, to provide a forum to exchange information, to discuss the matter and to gather and disseminate information on the status of national practices of burnup credit implementation in the Member States. The task addresses current and future aspects of burnup credit. This task was continued during the following years. (author)
REBUS: A burnup credit experimental programme
An international programme called REBUS (REactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK-CEN and Belgonucleaire. At present it is sponsored by USNRC, EdF from France and VGB, representing German nuclear utilities. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark would qualify the codes to perform calculations of the burn-up credit. The benchmark exercise will investigate the following fuel types with associated burn-up. 1. Reference absorber test bundle, 2. Fresh commercial PWR UO2 fuel, 3. Irradiated commercial PWR UO2 fuel (50 GWd/tM), 4. Fresh PWR UO2 fuel, 5. Irradiated PWR UO2 fuel (30 GWd/tM). Reactivity effects will be measured in the critical facility VENUS. The accumulated burn-up of all rods will be measured non-destructively by gamma-spectrometry. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-18 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). The experimental implementation of the programme will start in 2000. (author)
The applications of burnup credit and the measurement techniques of burnup verification
The factors of influencing criticality safety, implementing criticality control conditions, the calculation methods for predicting criticality, casks design and cask loading graph are described. The problems in the application of burnup credit and the dominant error in burnup credit operation are analysed. In order to avoid the operation error, requirements of measurement techniques and the most suitable measurement method are introduced
COGEMA/TRANSNUCLEAIRE's experience with burnup credit
Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)
Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik
2014-06-15
In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
The OECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The report summarises and analyses the solutions to the specified exercises provided by 37 contributors from 10 countries. The exercises were based upon the calculation of infinite PWR fuel pin cell reactivity for fresh and irradiated MOX fuels with various MOX compositions, burn-ups and cooling times. In addition, several representations of the MOX fuel assembly were tested in order to check various levels of approximations commonly used in reactor physics calculations. (authors)
Future disposal burnup credit process and effort
The United States Department of Energy's Office of Civilian Radioactive Waste Management has developed a risk-informed, performance based methodology for disposal criticality analyses. The methodology is documented in the Disposal Criticality Analysis Methodology Topical Report, YMP/TR-004Q (YMP 2000). The methodology includes taking credit for the burnup of irradiated commercial light water reactor fuel in criticality analyses, i.e., burnup credit. This paper summarizes the ongoing and planned future burnup credit activities associated with the methodology. (author)
Burnup credit activities in the United States
This report covers progress in burnup credit activities that have occurred in the United States of America (USA) since the International Atomic Energy Agency's (IAEA's) Advisory Group Meeting (AGM) on Burnup Credit was convened in October 1997. The Proceeding of the AGM were issued in April 1998 (IAEA-TECDOC-1013, April 1998). The three applications of the use of burnup credit that are discussed in this report are spent fuel storage, spent fuel transportation, and spent fuel disposal. (author)
Phenomena and Parameters Important to Burnup Credit
Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given
Phenomena and parameters important to burnup credit
Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water- reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the United States and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given. (author)
Issues for effective implementation of burnup credit
In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)
2005 status and future of burnup credit in the USA
At the beginning of 2005 in the USA burnup credit is licensed for PWR and BWR spent fuel pools, is under license review for a transport cask, is under discussion for disposal criticality. Two basic approaches exist for burnup credit. The first approach, which is licensed for spent fuel pools, utilizes criticality experience with spent fuel that has not been chemically assayed. The second approach to burnup credit comes from utilizing chemical assay data to validate the depletion calculations and then clean critical experiments to validate the criticality calculation. A burnup credit standard (ANS/ANSI-8.27) is under development where the two approaches are actively discussed. Issues related to the two approaches are presented as well as possible ways of resolving the issues. (author)
Investigation of burnup credit implementation for BWR fuel
Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)
The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that keff and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)
Burnup credit issues in transportation and storage
Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed
Burnup credit issues in transportation and storage
Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the U.S. experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed. (author)
Final evaluation of the CB3+burnup credit benchmark addition
In 1966 a series of benchmarks focused on the application of burnup credit in WWER spent fuel management system was launched by L.Markova (1). The four phases of the proposed benchmark series corresponded to the phases of the Burnup Credit Criticality Benchmark organised by the OECD/NEA.These phases referred as CB1, CB2, CB3 and CB4 benchmarks were designed to investigate the main features of burnup credit in WWER spent fuel management systems. In the CB1 step, the multiplication factor of an infinite array of spent fuel rods was calculated taking the burnup, cooling time and different group of nuclides as parameters. The fuel compositions was given in the benchmark specification (Authors)
A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (author)
OECD/NEA Burnup Credit Criticality Benchmark
The report describes the final result of the phase-1A of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. The phase-1A benchmark problem is an infinite array of a simple PWR spent fuel rod. The analysis has been performed for the PWR spent fuels of 30 and 40 GWd/t after 1 and 5 years of cooling time. In total, 25 results from 19 institutes of 11 countries have been submitted. For the nuclides in spent fuel, 7 major actinides and 15 major fission products (FP) are selected for the benchmark calculation. In the case of 30 GWd/t burnup, it is found that the major actinides and the major FPs contribute more than 50% and 30% of the total reactivity loss due to burnup, respectively. Therefore, more than 80% of the reactivity loss can be covered by 22 nuclides. However, the larger deviation among the reactivity losses by participants has been found for cases including EPs than the cases with only actinides, indicating the existence of relatively large uncertainties in FP cross sections. The large deviation seen also in the case of the fresh fuel has been found to reduce sufficiently by replacing the cross section library from ENDF-B/IV with that from ENDF-B/V and taking the known bias of MONK6 into account. (author)
CB2 result evaluation (VVER-440 burnup credit benchmark)
The second portion of the four-piece international calculational benchmark on the VVER burnup credit (CB2) prepared in the collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmarks Working Group and proposed to the AER research community has been evaluated. The evaluated results of calculations performed by analysts from Cuba, the Czech Republic, Finland, Germany, Russia, Slovakia and the United Kingdom are presented. The goal of this study is to compare isotopic concentrations calculated by the participants using various codes and libraries for depletion of the VVER-440 fuel pin cell. No measured values were available for the comparison. (author)
Analysis of burnup credit on spent fuel storage
Chemical analyses were carried out on high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234U to 242Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)
The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)
Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu
2000-07-01
The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)
Transnucleaire's experience with burnup credit in transport operations
Facing a continued increase in fuel enrichment values, Transnucleaire has progressively implemented a burnup credit programme in order to maintain or, where possible, to improve the capacity of its transport packagings without physical modification. Many package design approvals, based on a notion of burnup credit, have been granted by the French competent authority for transport since the early eighties, and many of these approvals have been validated by foreign competent authorities. Up to now, these approvals are restricted to fuel assemblies made of enriched uranium and irradiated in pressurized water reactors (PWR). The characterization of the irradiated fuel and the reactivity of the package are evaluated by calculation, performed using qualified French codes developed by the CEA (Commisariat a l'Energie Atomique/French Atomic Energy Commission): CESAR as a depletion code and APOLO-MORET as a criticality code. The approvals are based on the hypothesis that the burnup considered is that applied on the least irradiated region of the fuel assemblies, the conservative approach being not to take credit for any axial profile of burnup along the fuel assembly. The most reactive configuration is calculated and the burnup credit is also restricted to major actinides only. On the operational side and in compliance with regulatory requirements, verification is made before transport, in order to meet safety objectives as required by the transport regulations. Besides a review of documentation related to the irradiation history of each fuel assembly, it consists of either a qualitative (go/no-go) verification or of a quantitative measurement, depending on the level of burnup credit. Thus the use of burnup credit is now a common practice with Transnucleaire's packages, particularly in France and Germany. New improvements are still in progress and qualifications of the calculation code are now well advanced, which will allow in the near future the use of six selected
Finnish contribution to the CB4 burnup credit benchmark
The CB4 phase of the WWER burnup credit benchmark series studies the effect of flat and realistic axial burnup profiles on the multiplication factor of a conceptual WWER cask loaded with spent fuel. The benchmark was calculated at VTT Energy with MCNP4C, using mainly ENDF/B-V1 cross sections. According to the calculation results the effect of the axial homogenization on the keff estimate is complex. At low burnups the use of a axial profile overestimates keff but at high burnups the reverse is the case. Ignoring fission products leads to conservative keff and the effect of axial homogenization on the multiplication factor is similar to a reduction of the burnup (Authors)
Integrated burnup calculation code system SWAT
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)
Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes
Burnup credit methodology validation against WWER experimental data
A methodology for criticality safety analyses with burnup credit application has been developed for WWER spent fuel management facilities. This methodology is based on two worldwide used code systems: SCALE 4.4 for depletion and criticality calculations and NESSEL-NUKO - for depletion calculations. The methodology is in process of extensive validation for WWER applications. The depletion code systems NESSEL-NUKO and SCALE4.4 (control module SAS2H) have been validated on the basis of comparison with the calculated results obtained by other depletion codes for the CB2 Calculational Burnup Credit Benchmark. The validation of these code systems for WWER-440 and WWER-1000 spent fuel assembly depletion analysis based on comparisons with appropriate experimental data commenced last year. In this paper some results from burnup methodology validation against measured nuclide concentration given in the ISTC project 2670 for WWER-440 and from ORNL publication for WWER-1000 are presented. (authors)
Strategies for Application of Isotopic Uncertainties in Burnup Credit
Gauld, I.C.
2002-12-23
Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103
Detailed Burnup Calculations for Testing Nuclear Data
Leszczynski, F.
2005-05-01
-section data for burnup calculations, using some of the main available evaluated nuclear data files (ENDF-B-VI-Rel.8, JEFF-3.0, JENDL-3.3), on an isotope-by-isotope basis as much as possible. The selected experimental burnup benchmarks are reference cases for LWR and HWR reactors, with analysis of isotopic composition as a function of burnup. For LWR (H2O-moderated uranium oxide lattices) four benchmarks are included: ATM-104 NEA Burnup credit criticality benchmark; Yankee-Rowe Core V; H.B.Robinson Unit 2 and Turkey Point Unit 3. For HWR (D2O-moderated uranium oxide cluster lattices), three benchmarks were selected: NPD-19-rod Fuel Clusters; Pickering-28-rod Fuel Clusters; and Bruce-37-rod Fuel Clusters. The isotopes with experimental concentration data included in these benchmarks are: Se-79, Sr90, Tc99, Ru106, Sn126, Sb125,1129, Cs133-137, Nd143, 145, Sm149-150, 152, Eul53-155, U234-235, 238, Np237, Pu238-242, Am241-243, and Cm242-248. Results and analysis of differences between calculated and measured absolute and/or relative concentrations of these isotopes for the seven benchmarks are included in this work.
Studies on future application of burnup credit in Hungary
This paper describes the present status of the fuel storage and the possible future applications of burnup credit in wet and dry storage systems in Hungary. It gives a survey of the activities planned in AERI concerning the burnup credit. Some part of these investigations dealing with the influence of the axial changing of the assembly burnup are given in more details. (author)
Fission product margin in burnup credit analyses
The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work
The impact of burn-up credit in criticality studies
Nowadays optimization goes with everything. So French engineering firms try to demonstrate that fuel transport casks and storage pools are able to receive assemblies with higher 235U initial enrichments. Fuel Burnup distribution contributes to demonstrate it. This instruction has to elaborate a way to take credit of burnup effects on criticality safety designs. The calculation codes used are CESAR 4.21-APOLLO 1-MORET III. The assembly studied (UO2) is irradiated in a French Pressurized Water Reactor like EDF nuclear power reactor: PWR 1300 MWe, 17 x 17 array. Its initial enrichment in 235U equals 4.5%. The studies exposed in this report have evaluated the effects of: i) the 15 fission products considered in Burnup Credit (95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 133Cs, 143Nd, 145Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 153Eu, 155Gd), ii) the calculated abundances corrected or not by fixed factors, iii) the choice of one cross sections library used by CESAR 4.21, iu) the zone number elected in the axial burnup distribution zoning, u) the kind of cut applied on (regular/optimized). Two axial distribution profiles are studied: one with 44 GWd/t average burnup, the other with 20 GWd/t average burnup. The second one considers a shallow control rods insertion in the upper limit of the assembly. The results show a margin in reactivity about 0.045 with consideration of the 6 most absorbent fission products (103Rh, 133Cs, 143Nd, 149Sm, 152Sm, 155Gd), and about 0.06 for all Burnup Credit fission products whole. Those results have been calculated with an average burnup of 44 GWj/t. In a conservative approach, corrective factors must be apply on the abundance of some fission products. The cross sections library used by CESAR 4.21 (BBL 4) is sufficient and gives satisfactory results. The zoning of the assembly axial distribution burnup in 9 regular zones grants a satisfying calculation time/result precision compromise. (author)
The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report
Burnup effects of MOX fuel pincells in PWR - OECD/NEA burnup credit benchmark analysis -
The burnup effects were analyzed for various cases of MOX fuel pincells of fresh and irradiated fuels by using the HELIOS, MCNP-4/B, CRX and CDP computer codes. The investigated parameters were burnup, cooling time and combinations of nuclides in the fuel region. The fuel compositions for each case were provided by BNFL (British Nuclear Fuel Limited) as a part of the problem specification so that the results could be focused on the calculation of the neutron multiplication factor. The results of the analysis show that the largest saving effect of the neutron multiplication factor due to burnup credit is 30 %. This is mainly due to the consideration of actinides and fission products in the criticality analysis
The US department of energy's transportation burnup credit program
Aspects of the U. S. Department of Energy's (DOE's) transportation burnup credit program, the Department's motivation for conducting the program, and the status of burnup credit activities are presented. The benefits, technical, and regulatory considerations associated with using burnup credit for transport of irradiated nuclear fuel are discussed. The methods used in the DOE's actinide-only topical report are described in terms of the technical and regulatory issues. (authors)
Application of burnup credit with partial boron credit to PWR spent fuel storage pools
The outcome of performing a burnup credit criticality safety analysis of a PWR spent fuel storage pool is the determination of burnup credit loading curves BLC=BLC(e) for the spent fuel storage racks designed for burnup credit, cp. Reference. A burnup credit loading curve BLC=BLC(e) specifies the loading criterion by indicating the minimum burnup BLC(e) necessary for the fuel assembly with a specific initial enrichment e to be placed in storage racks designed for burnup credit. (orig.)
Mueller, Don E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowen, Douglas G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-09-01
The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k_{eff}) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k_{eff} calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.
Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance
Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and
Investigation of Burnup Credit Issues in BWR Fuel
Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of 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, followed by criticality calculations to assess the value of keff for 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 used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel
Implementation of burnup credit in PWR spent fuel storage pools
Implementation of burnup credit in spent fuel storage of LWR fuel at nuclear power plants is approved in Germany since the beginning of 2000. The burnup credit methods applied have to comply with the newly developed German criticality safety standard DIN 25471 passed in November 1999 and published in September 2000, cp. (orig.)
Probabilistic assessment of dry transport with burnup credit
The general concept of probabilistic analysis and its application to the use of burnup credit in spent fuel transport is explored. Discussion of the probabilistic analysis method is presented. The concepts of risk and its perception are introduced, and models are suggested for performing probability and risk estimates. The general probabilistic models are used for evaluating the application of burnup credit for dry spent nuclear fuel transport. Two basic cases are considered. The first addresses the question of the relative likelihood of exceeding an established criticality safety limit with and without burnup credit. The second examines the effect of using burnup credit on the overall risk for dry spent fuel transport. Using reasoned arguments and related failure probability and consequence data analysis is performed to estimate the risks of using burnup credit for dry transport of spent nuclear fuel. (author)
Study on burn-up credit and minor actinide in post-irradiation analysis
Accuracy of burnup calculation for actinide is very important as to the study of burn-up credit. For minor-actinides such as Am243 and Cm244, however, typical burnup calculation codes are not accurate enough. The accuracy for both nuclides was studied by using the SWAT code. The study showed that the C/E values of both nuclides could be improved at the same time by changing the cross section of Pu242. A study of burnup calculation related to the cross section of Pu242 should be performed to improve the accuracy for both nuclides. (author)
Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses
Wagner, J.C.
2002-10-23
This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.
The REBUS experimental programme for burn-up credit
An international programme called REBUS for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Centre SCK·CEN and Belgonucleaire with the support of EdF and IRSN from France and VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. Recently also ORNL from the U.S. jointed the programme. The programme aims to establish a neutronic benchmark for reactor physics codes in order to qualify the codes for calculations of the burn-up credit. The benchmark exercise investigate the following fuel types with associated burn-up: reference fresh 3.3% enriched UO2 fuel, fresh commercial PWR UO2 fuel and irradiated commercial PWR UO2 fuel (54 GWd/tM), fresh PWR MOX fuel and irradiated PWR MOX fuel (20 GWd/tM). The experiments on the three configurations with fresh fuel have been completed. The experiments show a good agreement between calculation and experiments for the different measured parameters: critical water level, reactivity effect of the water level and fission-rate and flux distributions. In 2003 the irradiated BR3 MOX fuel bundle was loaded into the VENUS reactor and the associated experimental programme was carried out. The reactivity measurements in this configuration with irradiated fuel show a good agreement between experimental and preliminary calculated values. (author)
Aiming at validation of depletion codes against WWER-440 spent fuel data some calculations of isotope composition of WWER-440 spent fuel assembly have been carried out by the NESSEL-NUKO code system. The initial data and data for the comparisons were taken from the ISTS burn up credit project data, recently published in the ISTC report 'Radiochemical Assays of Irradiated WWER-440 Fuel for Use in Spent Fuel Burnup Credit Activities. The specific work scope included the destructive assay (DA) of spent fuel assembly rod segments with an - -38.5 MWd/KgU burn up from a single WWER-440 fuel assembly from the Novovorenezh reactor in Russia (Authors)
Addressing the Axial Burnup Distribution in PWR Burnup Credit Criticality Safety
This paper summarizes efforts related to developing a technically justifiable approach for addressing the axial burnup distribution in PWR burnup-credit criticality safety analyses. The paper reviews available data on the axial variation in burnup and the effect of axial burnup profiles on reactivity in a SNF cask. A publicly available database of profiles is examined to identify profiles that maximize the neutron multiplication factor, keff, assess its adequacy for general PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. For this assessment, a statistical evaluation of the keff values associated with the profiles in the axial burnup profile database was performed that identifies the most reactive profiles as statistical outliers that are not representative of typical discharged SNF assemblies. The impact of these bounding profiles on the neutron multiplication factor for a high-density burnup credit cask is quantified. Finally, analyses are presented to quantify the potential reactivity consequence of assemblies with axial profiles that are not bounded by the existing database. The paper concludes with findings for addressing the axial burnup distribution in burnup credit analyses
Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit
Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL
2015-05-01
Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.
Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired keff for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program
The implementation of burnup credit in VVER-440 spent fuel
The countries using Russian reactors VVER-440 cooperate in reactor physics in Atomic Energy Research (AER). One of topic areas is 'Physical Problems of Spent Fuel, Radwaste and Decommissioning' (Working Group E). In this article, in the first part is an overview about our activity for numerical and experimental verification of codes which participants use for calculation of criticality, isotopic concentration, activity, neutron and gamma sources and shielding is shown. The set of numerical benchmarks (CB1, CB2, CB3 and CB4) is very similar (the same idea, the VVER-440) to the OECD/NEA/NSC Burnup Credit Criticality Benchmarks, Phases 1 and 2. In the second part, verification of the SCALE 4.4 system (only criticality and nuclide concentrations) for VVER-440 fuel is shown. In the third part, dependence of criticality on burnup (only actinides and actinides + fission products) for transport cask C30 with VVER-440 fuel by optimal moderation is shown. In the last part, current status in implementation burnup credit in Slovakia is shown. (author)
Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages
The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the
Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2
None, None
1998-09-01
The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the
Burnup credit demands for spent fuel management in Ukraine
In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)
燃耗信任制临界计算中保守性因素研究%Study on the conservative factors for burnup credit criticality calculation
刘驰; 蒋校丰; 张少泓
2012-01-01
When applies the burnup credit technology to perform criticality safety analysis for spent fuel storage or transportation problems, it is important for one to confirm that all the conditions adopted are adequate to cover the severest conditions that may encounter in the engineering applications. Taking the OECD/NEA burnup credit criticality benchmarks as sample problems, we study the effect of some important factors that may affect the conservatism of the results for spent fuel system criticality safety analysis. Effects caused by different nuclides credit strategy, different cooling time and axial burnup profile are studied by use of the STARBUCS module of SCALE5. 1 software package, and related conclusions about the conservatism of these factors are%在运用燃耗信任制技术进行乏燃料储存、运输等环节的临界安全分析时,临界计算所采用的条件是否具有足够的包络性十分关键.本文借助于OECD/NEA发布的若干燃耗信任制临界安全基准题,使用SCALE5.1软件中的STARBUCS模块进行分析,对信任核素选取、乏燃料冷却时间以及端末效应等因素对乏燃料系统临界安全性的影响进行了研究,得出了各参数保守性的有关结论.
Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias
Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)
EPRI R and D perspective on burnup credit
'Burnup credit' refers to taking credit for the burnup of nuclear fuel in the performance of criticality safety analyses. Historically, criticality safety analyses for transport of spent nuclear fuel have assumed the fuel to be unirradiated (i.e. 'fresh' fuel). In 1999, the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office issued Interim Staff Guidance - 8 (ISG-8) with recommendations for the use of burnup credit in storage and transportation of pressurized water reactor (PWR) spent fuel. The use of burnup credit offers an opportunity to reduce the number of spent nuclear fuel shipments by ∼30%. A simple analysis shows that the increased risk of a criticality event associated with properly using burnup credit is negligible. Comparing this negligible risk component with the reduction in common transport risks due to the reduced number of spent fuel shipments (higher capacity casks for transporting PWR spent fuel) leads to the conclusion that using 'burnup credit' is preferable to using the 'fresh fuel' assumption. A specific objective of the EPRI program is to support the Goals of the U.S. Industry. These goals are consistent with the original U.S. Department of Energy (DOE) goal defined in 1988: a burnup credit methodology that takes credit for the negative reactivity that is practical (all fissile actinides, most neutron absorbing actinides, and a subset of the fission products that account for the majority of the available credit from all fission products). The determination of the optimum number of fission products to consider in a practical burnup credit methodology validates the approach advocated by researchers from France to first focus on a handful of isotopes that include Sm-149; Rh-103; Nd-143; Gd-155; and Sm-152. (author)
Burnup calculation code system COMRAD96
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
A guide introducing burnup credit, preliminary version. Contract research
NONE
2001-07-01
It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)
A guide introducing burnup credit, preliminary version. Contract research
It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)
Burnup credit considerations in dry spent-fuel storage licensing
Burnup credit has been allowed in reactor basin spent-fuel storage at pressurized water reactors for a number of years. However, such storage occurs under strict administrative, procedural, and design controls. In recent years, dry spent-fuel storage cask vendors have expressed interest in designing cask fuel baskets with allowance for burnup credit. At last year's American Nuclear Society Winter Meeting, an ad hoc session was organized and authorized on burnup credit for dry storage and transportation casks. It has become clear that some utilities are interested in burnup credit for dry storage designs. Given this, the US Nuclear Regulatory Commission (NRC) staff is examining the technical issues involved in allowing burnup credit. Analytical work focused on the development of branch technical positions for determination of burnup credit for dry spent-fuel storage technology designs has begun. Procedural and administrative issues will be examined, based on licensing experience, and will also be the subject of branch technical positions. At an appropriate time, preparation of regulatory guides will be considered
A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit
The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years
Taking burnup credit for interim storage and transportation system for BWR fuels
In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)
Siemens PWR burnup credit criticality analysis methodology: Depletion code and verification methods
Application of burnup credit requires knowledge of the reactivity state of the irradiated fuel for which burnup credit is taken. The isotopic inventory of the irradiated fuel has to be calculated, therefore, by means of depletion codes. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the code package SAV. This code package is based on the first principles approach, i.e., avoids cycle or reactor specific fitting or adjustment parameters. This approach requires a general and comprehensive qualification of SAV by comparing experimental with calculational results. In the paper on hand the attention is focused mainly on the evaluation of chemical assay data received from different experimental programmes. (author)
Disposal criticality analysis methodology's principal isotope burnup credit
This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)
Determination of axial profit performed burnup credit by SCALE 4.3-system
SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs
Given a trend towards higher burnup power reactor fuel, the IAEA began an active programme in burnup credit (BUC) with major meetings in 1997 (IAEA-TECDOC-1013), 2000 (IAEA-TECDOC-1241) and 2002 (IAEA-TECDOC-1378) exploring worldwide interest in using BUC in spent fuel management systems. This publication contains the proceedings of the IAEA's 4th major BUC meeting, held in London. Sixty participants from 18 countries addressed calculation methodology, validation and criticality, safety criteria, procedural compliance with safety criteria, benefits of BUC applications, and regulatory aspects in BUC. This meeting encouraged the IAEA to continue its activities on burnup credit including dissemination of related information, given the number of Member States having to deal with increased spent fuel quantities and extended durations. A 5th major meeting on burnup credit is planned 2008. Burnup credit is a concept that takes credit for the reduced reactivity of fuel discharged from the reactor to improve loading density of irradiated fuel assemblies in storage, transportation, and disposal applications, relative to the assumption of fresh fuel nuclide inventories in loading calculations. This report has described a general four phase approach to be considered in burnup credit implementation. Much if not all of the background research and data acquisition necessary for successful burnup credit development in preparation for licensing has been completed. Many fuel types, facilities, and analysis methods are encompassed in the public knowledge base, such that in many cases this guidance will provide a means for rapid development of a burnup credit program. For newer assembly designs, higher enrichment fuels, and more extensive nuclide credit, additional research and development may be necessary, but even this work can build on the foundation that has been established to date. Those, it is hoped that this report will serve as a starting point with sufficient reference to
Parametric neutronic analyses related to burnup credit cask design
The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of keff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models
This paper describes the experience gained in Germany in implementing burnup credit in wet storage and dry transport systems of spent PWR, BWR, and MOX fuel. It gives a survey of the levels of burnup credit presently used, the regulatory status and activities planned, the fuel depletion codes and criticality calculation codes employed, the verification methods used for validating these codes, the modeling assumptions made to ensure that the burnup credit criticality analysis is based on a fuel irradiation history which leads to bounding neutron multiplication factors, and the implementation of procedures used for fuel loading verification. (author)
Program package for 2D burnup calculation
The program package for 2 dimension burnup calculation was developed for TRIGA Mark III reactor. The package consists of 3 modules: PRESIX, SIXTUS-2, and BURN; 1 library, and 2 input files. PRESIX module prepared cross sections for diffusion calculation. SIXTUS-2 module, a two dimensional diffusion code in hexagonal geometry, calculates keff, neutron fluxes and power distributions. BURN module performs the burnup of fuel elements and stored the result in the ELEM.DAT file. PRESIX.LIB is two group cross section library for major reactor core components prepared using WIMS-D4 code. PRES.INP, the first input file, reads information on reactor power and core loading pattern. ELEM.DAT, the second input file, is prepared for specific TRIGA reactor and dependent on operation history. To verify the reactor model and computational methods, the calculated excess reactivities were compared to the measurement. The results are in good agreement. (author)
Validation issues for depletion and criticality analysis in burnup credit
This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technical community (national laboratories, licensees, and regulators) that have been exploring the use of burnup credit. There is not necessarily agreement on the importance of the various issues, which sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data (radiochemical assays and critical experiments) covering the full range and characteristics of spent nuclear fuel in away-from-reactor systems. The paper will also introduce recent efforts initiated at Oak Ridge National Laboratory (ORNL) to provide technical information that can help better assess the value of different experiments. The focus of the paper is on experience with validation issues related to use of burnup credit for transport and dry storage applications. (author)
Burnup credit in nuclear waste transport: An economic analysis
The US DOE is responsible for transporting nuclear spent fuel from commercial reactors to monitored retrievable storage (MRS) facilities and/or to repositories. Current plans call for approximately 110,000 metric tons uranium (MTU) to be transported over approximately 40 years beginning in 1998. Because of the large volume of spent fuel to be transported, new generations of spent fuel transportation casks are being planned. These casks will embody the latest technology and will be designated to accommodate the spent fuel in a way that maximizes the overall efficiency of the cask. In planning for the new generation of transport casks, the DOE is investigating the possibility of tailoring the cask design for the extent to which spent fuel has been used in the reactors, or, for spent fuel burnup. Granting design credit for burnup would allow one to fabricate casks with relatively larger capacities than would be possible otherwise. The remainder of the paper discusses the economic implications of using burnup credit in cask design, discusses the approach used in analyzing the economics of burnup credit, describes the results of the analysis, and offers some conclusions about the economic value of the burnup credit option
TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES
A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. Fifty-seven UO2, UO2/Gd2O3, and UO2/PuO2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on keff (which can be a function of the trending parameters) such that the biased keff, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading
Actinide-only burnup credit for spent fuel transport
A conservative methodology is described that would allow taking credit for burn up in the criticality safety analysis of spent nuclear fuel packages. Requirements for its implementation include isotopic and criticality validation, generation of package loading criteria using limiting parameters, and assembly burn up verification by measurement. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentrations with burnup. No credit for fission product neutron absorbers is taken. Analyses are included regarding the methodology's financial benefits and conservative margin. It is estimated that the proposed actinide-only burnup credit methodology would save 20% of the transport costs. Nevertheless, the methodology includes a substantial margin. Conservatism due to the isotopic correction factors, limiting modelling parameters, limiting axial profiles and exclusion of the fission products ranges from 10 to 25% k. (author)
Technical Development on Burn-up Credit for Spent LWR Fuel
Gauld, I.C.
2001-12-26
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.
Technical Development on Burn-up Credit for Spent LWR Fuel
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report
ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT
The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent 235U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU)
ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT
A.H. Wells
2004-11-17
The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).
Revised SWAT. The integrated burnup calculation code system
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Revised SWAT. The integrated burnup calculation code system
Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel
A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951
Actinide-only burnup credit methodology for PWR spent nuclear fuel
A conservative methodology is presented that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel (SNF) packages. The method is based on the assumption that the isotopic concentration in the SNF and cross sections of each isotope for which credit is taken must be supported by validation experiments. The method allows credit for the changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps: 1. Validate a computer code system to calculate isotopic concentrations of spent nuclear fuel created during burnup in the reactor core and subsequent decay. 2. Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package by use of UO2 and UO2/Puo2 critical experiments. 3. Establish conditions for the SNF (depletion analysis) and package (criticality analysis) which bounds keff. 4. Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). 5. Verify by measurement that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. (author)
Burnup credit implementation plan and preparation work at JAERI
Application of the burnup credit concept is considered to be very effective to the design of spent fuel transport and storage facilities. This technology is all the more important when considering construction of the intermediate spent fuel storage facility, which is to be commissioned by 2010 due to increasing amount of accumulated spent fuel in Japan. Until reprocessing and recycling all the spent fuel arising, they will be stored as an energy stockpile until such time as they can be reprocessed. On the other hand, the burnup credit has been partly taken into account for the spent fuel management at Rokkasho Reprocessing Plant, which is to be commissioned in 2005. They have just finished the calibration tests for their burnup monitor with initially accepted several spent fuel assemblies. Because this monitoring system is employed with highly conservative safety margin, it is considered necessary to develop the more rational and simplified method to confirm burnup of spent fuel. A research program has been instituted to improve the present method employed at the spent fuel management system for the Spent Fuel Receiving and Storage Pool of Rokkasho Reprocessing Plant. This program is jointly performed by Japan Nuclear Fuel Limited (JNFL) and JAERI.This presentation describes the current status of spent fuel accumulation discharged from PWR and BWR in Japan and the recent incentive to introduce burnup credit into design of spent fuel storage and transport facilities. This also includes the content of the joint research program initiated by JNFL and JAERI. The relevant study has been continued at JAERI. The results by these research programs will be included in the Burnup Credit Guide Original Version compiled by JAERI. (author)
Taking burnup credit for interim storage and transportation system for BWR fuels
In the back-end issues of nuclear fuel cycle, selection of reprocessing or one-through is a big issue. For both of the cases, a reasonable interim storage and transportation system is required. This study proposes an advanced practical monitoring and evaluation system. The system features the followings: (l) Storage racks and transportation casks taking credit for burnup. (2) A burnup estimation system using a compact monitor with Cd- Te detectors and fission chambers. (3) A neutron emission-rate evaluation methodology, especially important for high burnup MOX fuels. (4) A nuclear materials management system for safeguards. Current storage system and transport casks are designed on the basis of a fresh fuel assumption. The assumption is too conservative. Taking burnup credit gives a reasonable design while keeping conservatism. In order to establish a reasonable burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of some modules such as TGBLA, ORIGEN, CITATION, MCNP and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. The code takes operational history such as, power density, void fraction into account. This code is applied to the back-end issues for a more accurate design of a storage and a transportation system. The ORIGEN code is well-known one-point isotope depletion code. In the calculation system, the code calculates isotope compositions using libraries generated from the TGBLA code. The CITATION code, the MCNP code, and the KENO code are three dimensional diffusion code, continuous energy Monte Carlo code, discrete energy Monte Carlo code, respectively. Those codes calculate k- effective of the storage and transportation systems using isotope compositions generated from the ORIGEN code. The CITATION code and the KENO code are usually used for practical designs. The MCNP code is used for reference
An economic evaluation of a storage system for casks with burnup credit
It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)
Overview of the burnup credit activities at OECD/NEA/NSC
This article summarizes activities of the OECD/NEA Burnup Credit Expert Panel, a subordinate group to the Working Party on Nuclear Criticality Safety (WPNCS). The WPNCS of the OECD/NEA coordinates and carries out work in the domain of criticality safety at the international level. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burnup credit. The activities of the expert panel are aimed toward improving safety and identifying economic solutions to issues concerning the back-end of the fuel cycle. The main objective of the activities of the OECD/NEA Burnup Credit Expert Panel is to demonstrate that the available criticality safety calculational tools are appropriate for application to burned fuel systems and that a reasonable safety margin can be established. The method established by the expert panel for investigating the physics and predictability of burnup credit is based on the specification and comparison of calculational benchmark problems. A wide range of fuel types, including PWR, BWR, MOX, and VVER fuels, has been or are being addressed by the expert panel. The objective and status of each of these benchmark problems is reviewed in this article. It is important to note that the focus of the expert panel is the comparison of the results submitted by each participant to assess the capability of commonly used code systems, not to quantify the physical phenomena investigated in the comparisons or to make recommendations for licensing action. (author)
Application of scale-4 depletion/criticality sequences in burnup credit analyses
The concept of allowing reactivity credit for the transmuted state of spent fuel complicates the criticality analysis by requiring the specification of the fuel mixture to potentially include large numbers of isotopes representative of the fuel conditions. These conditions include the initial enrichment, local or average burnup conditions depending on the analysis approach, and the post-shutdown cooling time. In the development of an analysis methodology to evaluate spent fuel shipping and transport casks (flasks) based on this burnup credit, commercial reactor critical configurations were evaluated as potential experimental spent fuel criticals. This paper describes how the SCALE-4 depletion sequences (SAS2H), the cross-section processing sequence (CSASN), and the criticality module (KENO V.a) were used to evaluate these reactor criticals. A description of a newly developed sequence for linking SAS2H calculated burnup-dependent isotopics to KENO V.a mixing tables [SAS2H Nuclide Inventories for KENO Runs (SNIKR)] is also included
A validated methodology for evaluating burnup credit in spent fuel casks
The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the U.S. Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in keff. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. (Author)
A validated methodology for evaluating burnup credit in spent fuel casks
The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in keff. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs
Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs
Wagner, J.C.
2002-12-17
This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).
The high density spent fuel storage rack Boraflex was known to experience changes of its physical property and to dissolve under exposure to radiation in an aqueous environment for long period of time. In this study, the criticality evaluation for spent fuel storage rack of Ulchin Unit 2 under normal condition was performed assuming complete loss of 10B from the Boraflex and applying burnup credit. Criticality evaluation code KENO-V.a. from SCALE4.4 system was benchmarked against critical experiments to obtain the calculation bias and bias uncertainties. The manufacturing tolerances of nuclear fuel and storage rack and their reactivity uncertainties were derived, as well. Considering those bias and uncertainties of calculation, the criticality of spent fuel storage under normal condition was conservatively evaluated. The criticality evaluation result using burnup credit can be presented as a spent fuel loading curve that indicates the acceptable burnup domain in spent fuel storage pool. The spent fuels with various initial enrichments and discharge fuel burnup can be safely accommodated in the storage without taking any boron credit from Boraflex, provided the combination falls within the acceptable domain in the loading curve. The spent fuel with initial enrichment of 5.0w/o was evaluated to meet the subcritical safety if its burnup is over 43.0GWD/MTU. The criticality evaluation result also showed that spent fuels with the initial enrichment less than 1.6w/o were able to be stored in the storage pool regardless of their burnup. Conclusively, in the Region 2 of the spent fuel storage pool, the maximum keff , considering all uncertainties, was calculated as 0.94818
Advances In Burnup Credit Criticality Safety Analysis Methods And Applications
An International Workshop on “Advances in Applications of Burnup Credit for Spent Fuel Storage, Transport, Reprocessing, and Disposition” organized by the Nuclear Safety Council of Spain (CSN) in cooperation with the International Atomic Energy Agency (IAEA) was held at Córdoba, Spain, on October 27– 30, 2009. The objectives of this workshop were to identify the benefits that accrue from recent improvements of the burnup credit (BUC) analysis methodologies, to analyze the implications of applying improved BUC methodologies, focusing on both the safety-related and operational aspects, and to foster the exchange of international experience in licensing and implementation of BUC applications. In the paper on hand the attention is focused on the improvements of BUC analysis methodologies. (author)
Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit
The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States (U.S.) Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized water reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% Δk. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they
Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit
Wagner, J.C.
2001-09-28
The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs
Value of 236U to actinide-only burnup credit
The US Department of Energy (DOE) submitted a topical report to the US Nuclear Regulatory Commission (NRC) in May 1995 in order to gain approval of a method for criticality analysis of transport packages that takes account for the change in actinide isotopes with burnup [pressurized water reactors (PWRs) only]. Historically, the NRC has conservatively assumed that the fuel was in its initial conditions (without any burnable absorbers). In order to permit credit for the changes in actinide content, the NRC has required validation of the depletion and criticality codes for spent nuclear fuel, justification of conservative depletion modeling, and finally confirmation measurements before loading. The NRC requested additional information on March 22, 1996. The DOE responded by a revision of the topical report in May 1997. The NRC again responded with another set of requests of additional information in April 1998. In that set of questions, the NRC challenged the use of 236U in burnup credit. Uranium-236 is not found in any significant amount in any available critical experiments. The authors explore the value of 236U to actinide-only burnup credit
Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL
2015-01-01
Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k_{eff}) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup
The increase of PWR-UOX fuel initial enrichment and the extensive needs for spent fuel storage or cask capacities reinforce the interest in taking burnup credit into account in criticality calculations. However, this utilization of credit for fuel burnup requires the definition of a methodology that ensures the conservatism of calculations. In order to guarantee the conservatism of the spent fuel inventory calculation, a depletion calculation scheme for burnup credit is under development. This paper presents the studies on the main parameters which have an effect on nuclides concentration: the presence of control rods during depletion and the fuel assembly environment, particularly the presence of MOx fuels around the UO2 assembly. Reactivity effects which are relevant to these parameters are then presented, and physics phenomena are identified. (author)
Practical issues with implementation of burnup credit in the USA for storage and transportation
The US NRC issued an interim staff guidance (ISG8 rev1) allowing for burnup credit applications for storage and transport casks in July of 1999. In over two and a half years there has still not been a license submittal using burnup credit. ISG8 rev1 does not provide sufficient burnup credit to allow loading of 5 wt% enriched fuel in a 32 PWR assembly cask without the addition of absorber rod inserts. Pressure to allow all assemblies to contain inserts from the utility, force continued investigation into alternative levels of burnup credit. Utilities do not wish to measure to confirm burnup. This measurement costs, which range form $10 000 to $50 000 per cask and must be done prior to loading. Since burnup credit is actually only needed for transport, and transport is not expected for many years, many utilities are considering keeping the money in the bank until the time of transport. In order to address the need perceived for additional burnup credit beyond actinide-only burnup credit (ISG8), investigations have moved beyond into assuming moderator exclusion during transport and the use of burnup credit to cover a beyond design basis accident assumption of flooding. Burnup credit analysis requirements for a beyond design basis accident should be less than that for criticality control for normal operation. It is proposed that burnup credit analysis to cover the beyond design basis accident of flooding should be consistent with the beyond design basis dilution event in PWR spent fuel pools. The US NRC precedence for this type of burnup credit allows for all isotopes, a 5% reduction in the delta k of burnup, and an allowable keff of less than 1.0 after biases and uncertainties. (author)
Evaluation of burnup credit for accommodating PWR spent nuclear fuel in high-capacity cask designs
This paper presents an evaluation of the amount of burnup credit needed for high-density casks to transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic 32-assembly cask and the current regulatory guidance were used as bases for this evaluation. By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based loading curves, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of SNF assemblies in high-capacity storage and transportation casks. The impact of varying selected calculational assumptions is also investigated, and considerable improvement in effectiveness is shown with the inclusion of the principal fission products (FPs) and minor actinides and the use of a bounding best-estimate approach for isotopic validation. Given sufficient data for validation, the most significant component that would improve accuracy, and subsequently enhance the utilization of burnup credit, is the inclusion of FPs. (author)
Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.
Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan
2015-09-15
Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.
DeHart, M.D.
1996-05-01
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. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.
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. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports
One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13% k/k.
Theory analysis and simple calculation of travelling wave burnup scheme
Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)
Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library
Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a 'best estimate' value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 fission products (FPs) of PWR-MOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF- 3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections. (authors)
Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library
Lecarpentier D.
2013-03-01
Full Text Available Burnup Credit (BUC is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a “best estimate” value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library. Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.
Study on the criticality safety evaluation method for burnup credit in JAERI
In relation to burnup credit, three tasks have been carried out at the Japan Atomic Energy Research Institute (JAERI) for establishing the evaluation method of criticality safety for a spent-fuel system, such as storage age ponds and transport casks. The first task is to prepare a benchmark database of criticality experiments and nuclide compositions of spent fuels. The database of nuclide composition is formed by data treasured at JAERI and data collected from the literature. For the database of criticality experiments, the effective multiplication factor of a spent-fuel assembly has been measured at JAERI and data collected from the literature. For the database of criticality experiments, the effective multiplication factor of a spent-fuel assembly has been measured at JAERI. The next task is to develop computer codes. The burnup and criticality codes have been developed and validated by analyzing a large number of benchmarks stored in the aforementioned database. The last task needed to establish the methodology in order to confirm the subcriticality of a spent-fuel system applying burnup credit is described. A reference fuel assembly is introduced so that the criticality of a system can be evaluated by using it, instead of modeling all fuel assemblies explicitly. To determine the nuclide composition of a spent fuel, a simple method is studied utilizing a large number of nuclide composition data stored in the database. Further, the effects of the axial burnup profile and calculation errors are discussed, and the remaining tasks are identified
Burnup calculation methodology in the serpent 2 Monte Carlo code
This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)
An overview of burnup credit application in spent nuclear fuel management
The current status of burnup credit application has been overviewed for spent nuclear fuel management. It was revealed that the use of burnup credit is practically limited to spent nuclear fuel storage, for which selected actinides-only are taken into account
A survey of previous and current industry-wide efforts regarding burnup credit
Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey
Burn-up credit in criticality safety of PWR spent fuel
Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B4C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, keff, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The keff was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, keff was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up
Benefits of the delta K of depletion benchmarks for burnup credit validation
Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, keff. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)
Benchmark calculation with MOSRA-SRAC for burnup of a BWR fuel assembly
The Japan Atomic Energy Agency has developed the Modular Reactor Analysis Code System MOSRA to improve the applicability of neutronic characteristics modeling. The cell calculation module MOSRA-SRAC is based on the collision probability method and is one of the core modules of the MOSRA system. To test the module on a real-world problem, it was combined with the benchmark program 'Burnup Credit Criticality Benchmark Phase IIIC.' In this program participants are requested to submit the neutronic characteristics of burnup calculations for a BWR fuel assembly containing fuel rods poisoned with gadolinium (Gd2O3), which is similar to the fuel assembly at TEPCO's Fukushima Daiichi Nuclear Power Station. Because of certain restrictions of the MOSRA-SRAC burnup calculations part of the geometry model was homogenized. In order to verify the validity of MOSRA-SRAC, including the effects of the homogenization, the calculated burnup dependent infinite multiplication factor and the nuclide compositions were compared with those obtained with the burnup calculation code MVP-BURN which had already been validated for many benchmark problems. As a result of the comparisons, the applicability of MOSRA-SRAC module for the BWR assembly has been verified. Furthermore, it can be shown that the effects of the homogenization are smaller than the effects due to the calculation method for both multiplication factor and compositions. (author)
For adopting burnup credit in transport or storage of spent fuel (SF), development of a reliable burnup calculation code is crucial. For this purpose, data of Post Irradiation Examination (PIE) have been extensively analyzed to evaluate accuracy of burnup calculation codes for a 14*14 or 15*15 PWR fuel assembly. This study shows results of analysis of this latest PIE with SWAT and ORIGEN2.1. SWAT is an integrated burnup code system for a 17*17 PWR fuel assembly that has been developed by Tohoku University and JAERI. The results show that SWAT can more precisely predict nuclide composition of latest PWR assembly than ORIGEN2.1. (O.M.)
This paper has presented the results of a computational benchmark and independent calculations to verify the benchmark calculations for the estimation of the additional reactivity margin available from fission products and minor actinides in a PWR burnup credit storage/transport environment. The calculations were based on a generic 32 PWR-assembly cask. The differences between the independent calculations and the benchmark lie within 1% for the uniform axial burnup distribution, which is acceptable. The Δk for KENO - MCNP results are generally lower than the other Δk values, due to the fact that HELIOS performed the depletion part of the calculation for both the KENO and MCNP results. The differences between the independent calculations and the benchmark for the non-uniform axial burnup distribution were within 1.1%
Use of burnup credit in criticality evaluation for spent fuel storage pool
Boraflex is a polymer based material which is used as matrix to contain a neutron absorber material, boron carbide. In a typical spent fuel pool the irradiated Boraflex has been known as a significant source of silica. Since 1996, it was reported that elevated silica levels were measured in the Ulchin Unit 2 spent fuel pool water. Therefore, the Ulchin Unit 2 spent fuel storage racks were needed to be reanalyzed to allow storage of fuel assemblies with normal enrichments up to 5.0w/o U-235 in all storage cell locations using credit for burnup. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels. In region 2, the calculations were performed by assuming in an infinite radial array of storage cells. No credit is taken for axial or radial neutron leakage. The water in the spent fuel storage pool was assumed to be pure. In the evaluation of the Ulchin Unit 2 spent fuel storage pool, criticality analyses were performed with the CASMO-3 code. A reactivity uncertainty in the fuel depletion calculations was combined with other calculational uncertainty. The manufacturing tolerances were considered, as well. From the calculation, the acceptable burnup domain in region 2 of the spent fuel storage pool. where the curve identifies conditions of equal reactivity for various initial enrichments between 1.6w/o and 5.0w/o, was evaluated. In region 2, the maximum keff including all uncertainties, is 0.94648 for the enrichment-burnup combination from loading curve. (author)
TRIGA criticality experiment for testing burn-up calculations
Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz [Jozef Stefan Institute, Reactor Physics Division, Ljubljana (Slovenia)
1999-07-01
A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)
Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses
The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., eff of a spent fuel cask
This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses
Wagner, J.C.
2001-08-02
This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.
The burn-up credit physics and the 40. Minerve anniversary
The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)
It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab
A calculational benchmark focused on VVER-440 burnup credit, similar to that of the OECD/NEA/NSC Burnup Credit Benchmark Working Group, was proposed on the 96'AER Symposium. Its first part, CB1, was specified there whereas the second part, CB2, was specified a year later, on 97'AER Symposium in Zittau. A final statistical evaluation is presented of CB1 results and summarizes the CB2 results obtained to date. Further, the effect of an axial burnup profile of VVER-440 spent fuel on criticality ('end effect') is proposed to be studied in the CB3 benchmark problem of an infinite array of VVER-440 spent fuel rods. (author)
Calculation Study of TNPS Spent Fuel Pool Using Burnup Credit%田湾核电站乏燃料水池采用燃耗信任制的计算研究
夏兆东; 周小平; 李晓波; 吕牛; 郑继业
2013-01-01
Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks ,the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied .The MONK9A code was used to analyze kef of dif-ferent enrichment fuels at different burnups .A reference loading curve was proposed in accordance with the system kef ’s changing with the burnup of different initially enriched nuclear fuels .The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC .%以田湾核电站（TNPS）2×5排列的贮存格架构成的乏燃料水池为例，研究采用燃耗信任制技术的密集贮存和临界安全问题。采用M ONK9A程序计算分析不同富集度、不同燃耗的乏燃料装载情况下系统的 ke f 。根据系统 ke f随不同初始富集度燃料的燃耗变化情况给出了水池的参考装载曲线。采用燃耗信任制技术的密集贮存方案能提高贮存能力31％。
Burnup credit application in criticality analysis of storage casks with spent RBMK-1500 nuclear fuel
Nuclear criticality safety analysis of two types of the casks CASTOR RBMK-1500 and CONSTOR RBMK-1500 was performed using the SCALE 4.3 computer code system. These casks are planned for an interim dry storage of spent nuclear fuel at Ignalina nuclear power plant. Effective neutron multiplication factor keff was calculated for different density of the water inside the casks for unfavorable operational cases and for assumed hypothetical accident conditions when fuel in the system is fresh and fuel is depleted (i.e. burnup credit taken into account). Results show that for all cases effective neutron multiplication factor keff is less then allowable value 0.95. (author)
Use of burnup credit in criticality safety design analysis of spent fuel storage systems
Full text: It is well known that the use of Burnup Credit (BUC) in criticality safety design analysis of spent fuel storage systems significantly impacts the design of the system. BUC is defined as the consideration of the change in the fuel's isotopic composition and hence in its reactivity due to the irradiation of the fuel. Using BUC means to identify that isotopic composition and hence that burnup which just results in the maximum neutron multiplication factor allowable for the system, including all mechanical and calculational uncertainties. This burnup is the minimum burnup necessary for fuel to be loaded in the system. Since the isotopic composition at given burnup depends on the initial enrichment of the fuel, the minimum burnup is usually given as a function of the initial enrichment. The graph of this function is commonly named as 'loading curve'. Thus, application of BUC to a spent fuel storage system consists in implementation of three key steps: Determination of the isotopic composition as a function of burnup and initial enrichment; Criticality calculation and evaluation of the loading curve; Quantification and verification of the actual burnup of the fuel to be loaded into the system. The main considerations of the first and the second step will be discussed. The isotopic composition is predicted by means of depletion calculations. To perform such calculations the parameters describing the fuel design characteristics and the fuel depletion conditions have to be defined. In addition the cooling time that may be credited (e.g., in BUC applications to spent fuel storage/transport cask systems) has to be specified. These parameters will be discussed with particular attention being given to the sensitivity of the neutron multiplication factor of the storage system to variations in the parameters and conditions characterizing the depletion conditions. These parameters and conditions are: Specific power and operating history, fuel temperature, moderator
Impact of Integral Burnable Absorbers on PWR Burnup Credit Criticality Safety Analyses
The concept of taking credit for the reduction in reactivity of burned or spent nuclear fuel (SNF) due to fuel burnup is commonly referred to as burnup credit. The reduction in reactivity that occurs with fuel burnup is due to the net reduction of fissile nuclide concentrations and the production of actinide and fission-product neutron absorbers. The change in the inventory of these nuclides with fuel burnup, and the consequent reduction in reactivity, is dependent upon the depletion environment. Therefore, the use of burnup credit necessitates consideration of all possible fuel operating conditions, including the use of integral burnable absorbers (IBAs). The Interim Staff Guidance on burnup credit [1] issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends licensees restrict the use of burnup credit to assemblies that have not used burnable absorbers (e.g., IBAs or burnable poison rods, BPRs). This restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. The reason for this restriction is that the presence of burnable absorbers during depletion hardens the neutron spectrum, resulting in lower 235U depletion and higher production of fissile plutonium isotopes. Enhanced plutonium production has the effect of increasing the reactivity of the fuel at discharge and beyond. Consequently, an assembly exposed to burnable absorbers may have a slightly higher reactivity for a given burnup than an assembly that has not been exposed to burnable absorbers. This paper examines the effect of IBAs on reactivity for various designs and enrichment/poison loading combinations as a function of burnup. The effect of BPRs, which are typically removed during operation, is addressed elsewhere [2
The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately
Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel
An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO2 fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)
TRIGA fuel burn-up calculations and its confirmation
The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.
Burnup credit implementation in WWER spent fuel management systems: Status and future aspects
This paper describes the motivation for possible burnup credit implementation in WWER spent fuel management systems in Bulgaria. The activities being done are described, namely: the development and verification of a 3D few-group diffusion burnup model; the application of the KORIGEN code for evaluation of WWER fuel nuclear inventory during reactor core lifetime and after spent fuel discharge; using the SCALE modular system (PC Version 4.1) for criticality safety analyses of spent fuel storage facilities. Future plans involving such important tasks as validation and verification of computer systems and libraries for WWER burnup credit analysis are shown. (author)
Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis
Enercon Services, Inc.
2011-03-14
Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in
Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis
Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in
OECD/NEA burnup credit criticality benchmark. Result of phase IIA
The report describes the final result of the Phase IIA of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. In the Phase IIA benchmark problems, the effect of an axial burnup profile of PWR spent fuels on criticality (end effect) has been studied. The axial profiles at 10, 30 and 50 GWd/t burnup have been considered. In total, 22 results from 18 institutes of 10 countries have been submitted. The calculated multiplication factors from the participants have lain within the band of ± 1% Δk. For the irradiation up to 30 GWd/t, the end effect has been found to be less than 1.0% Δk. But, for the 50 GWd/t case, the effect is more than 4.0% Δk when both actinides and FPs are taken into account, whereas it remains less than 1.0% Δk when only actinides are considered. The fission density data have indicated the importance end regions have in the criticality safety analysis of spent fuel systems. (author)
OECD/NEA burnup credit criticality benchmark. Result of phase IIA
Takano, Makoto; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1996-02-01
The report describes the final result of the Phase IIA of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. In the Phase IIA benchmark problems, the effect of an axial burnup profile of PWR spent fuels on criticality (end effect) has been studied. The axial profiles at 10, 30 and 50 GWd/t burnup have been considered. In total, 22 results from 18 institutes of 10 countries have been submitted. The calculated multiplication factors from the participants have lain within the band of {+-} 1% {Delta}k. For the irradiation up to 30 GWd/t, the end effect has been found to be less than 1.0% {Delta}k. But, for the 50 GWd/t case, the effect is more than 4.0% {Delta}k when both actinides and FPs are taken into account, whereas it remains less than 1.0% {Delta}k when only actinides are considered. The fission density data have indicated the importance end regions have in the criticality safety analysis of spent fuel systems. (author).
In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. In the conventional criticality safety evaluation of UNF where the fresh fuel composition is conservatively assumed, neutron multiplication factor is becoming overestimated as the fuel enrichment increases. The recent development of higher-enrichment fuel has therefore enhanced the benefit of the application of burnup credit. When applying the burnup credit to the criticality safety analysis of the disposed fuel system, the safe-side estimation of the reactivity is required taking into account the factors which affect the neutron multiplication factor of the burnt fuel system such as the nuclide composition uncertainties. In this report, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study. (author)
Validation of IRBURN calculation code system through burnup benchmark analysis
Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.
The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for
Validation of BGCore System for Burnup Calculations
BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. BGCore interfaces the Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating fuel composition during irradiation and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. The objective of this study is validation of the BGCore system against well established and verified, state of the art computer codes for thermal and fast spectrum lattices
OECD-NEA criticality working group - a status report and the burnup credit challenge
A Working Group established by the organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA), Paris, has examined the validity of computational methods used for calculations that evaluate the nuclear criticality safety issues involved in the storage, handling and transportation of fissile materials. The basic goal of this Working Group is to attempt to define and implement a procedure that can be shown to demonstrate the validity of the various computational methods used to make criticality safety calculations. The current activities of the Working Group involve an effort to establish the validity of computational methods used to evaluate the criticality safety of the storage, handling, and transportation of spent light-water-reactor fuel elements in which one seeks to take credit for the fissile material burnup and/or buildup of fission products. (J.P.N.)
Status of burnup credit for transport of SNF in the United States
Allowing credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transportation, and disposal of spent nuclear fuel (SNF) while maintaining a subcritical margin sufficient to establish an adequate safety basis. This paper reviews the current status of burnup credit applied to the design and transport of SNF casks in the United States. The existing U.S. regulatory guidance on burnup credit is limited to pressurized-water-reactor (PWR) fuel and to allowing credit only for actinides in the SNF. By comparing loading curves against actual SNF discharge data for U.S. reactors, the potential benefits that can be realized using the current regulatory guidance with actinide-only burnup credit are illustrated in terms of the inventory allowed in high-capacity casks and the concurrent reduction in SNF shipments. The additional benefits that might be realized by extending burnup credit to credit for select fission products are also illustrated. The curves show that, although fission products in SNF provide a small decrease in reactivity compared with actinides, the additional negative reactivity causes the SNF inventory acceptable for transportation to increase from roughly 30% to approximately 90% when fission products are considered. A savings of approximately $150M in transport costs can potentially be realized for the planned inventory of the repository. Given appropriate experimental data to support code validation, a realistic best-estimate analysis of burnup credit that includes validated credit for fission products is the enhancement that will yield the most significant impact on future transportation plans
As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity
Pore pressure calculation of the UO2 high burnup structure
Highlights: • Pore pressure is calculated based on local burnup, density and porosity. • Ronchi's equations of state are used instead of van der Waals’ equation. • Pore pressure increases as HBS transformation begins and then stays constant. • A best approximated parameter used for pore pressure calculation is recommended. -- Abstract: UO2 high burnup structure has an important impact on fuel behavior, especially in case of reactivity initiated accident (RIA). Pore relaxation enhances local fuel swelling and puts additional load to the fuel cladding, which makes fuel more susceptible to pellet–cladding mechanical interaction induced failure. Therefore, pore pressure calculation becomes vital when evaluating the fuel failure. In this paper pore pressure is calculated as a function of pellet radial local burnup based on the basic characteristics of HBS using Ronchi's correlation. The results indicate that pore pressure will approach a stable value as HBS is developing. A best approximated C value of 55 N/m is recommended for pore pressure calculation
U.S. Regulatory Research Program for Implementation of Burnup Credit in Transport Casks
In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to support the development of technical bases and guidance that would facilitate the implementation of burnup credit into licensing activities for transport and dry cask storage. This paper reviews the following major areas of investigation: (1) specification of axial burnup profiles, (2) assumption on cooling time, (3) allowance for assemblies with fixed and removable neutron absorbers, (4) the need for a burnup margin for fuel with initial enrichments over 4 wt %, and (5) evaluation of assay data and critical experiments. The capabilities of a new computational tool that facilitates the performance and coupling of the depletion and criticality analyses needed for burnup credit are also discussed
New burnup calculation of TRIGA IPR-R1 reactor
The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
New burnup calculation of TRIGA IPR-R1 reactor
Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
Burnup calculations using serpent code in accelerator driven thorium reactors
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
Burnup calculations using serpent code in accelerator driven thorium reactors
Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.
2013-07-15
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
An analysis of burnup reactivity credit for reactor RA spent fuel storage
The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described. (author)
The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit
The use of burnup credit in criticality control for the Korean spent fuel management program
More than 25% k-eff saving effect is observed in this burnup credit analysis. This mainly comes from the adoption of actinide nuclides and fission products in the criticality analysis. By taking burnup credit, the high capacity of the storage and transportation can be more fully utilized, reducing the space of storage and the number of shipments. Larger storage and fewer shipments for a given inventory of spent fuel result should in remarkable cost savings and more importantly reduce the risks to the public and occupational workers for the Korean Spent Fuel Management Program
Burnup credit methodology in the NPP Krzko spent fuel pool reracking project
NPP Krzko is going to increase the capacity of the spent fuel storage pool by replacement of the existing racks with high-density racks. The design, rack manufacturing and installation has been awarded to the Framatome ANP GmbH. Burnup credit methodology, which has been already approved by the Slovenian Nuclear Safety Administration in previous licensing of existing racks, will be again implemented in the licensing process with the recent methodology improvements. Specific steps of the criticality analysis and representative results are presented in the paper showing also the current national practice of the burnup credit implementation. (author)
The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation
Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA
For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)
Specific application of burnup credit for MOX PWR fuels in the rotary dissolver
In prospect of a Mixed OXide spent fuels processing in the rotary dissolver in COGEMA/La Hague plant, it is interesting to quantify the criticality-safety margins from the burnup credit. Using the current production computer codes and considering a minimal fuel irradiation of 3 200 megawatt-day per ton, this paper shows the impact of burnup credit on industrial parameters such as the permissible concentration in the dissolution solution or the permissible oxide mass in the rotary dissolver. Moreover, the burnup credit is broken down into five sequences in order to quantify the contribution of fissile nuclides decrease and of minor actinides and fission products formation. The implementation of the burnup credit in the criticality-safety analysis of the rotary dissolver may lead to workable industrial conditions for the particular MOX fuel studied. It can eventually be noticed that minor actinides contribution is negligible and that considering only the six major fission products is sufficient, owing to the weak fuel irradiation contemplated. (author)
Effect of Self-Shielding on Burn-Up Calculation of ETRR-2 Reactor
There exist two approaches for burn-up calculation. The first on is to use cell parameters generated using cell calculation code at different degrees of burn-up. The other is to use microscopic cross sections with self-shielding in order to compensate for the variation of spectrum at different degree of burn-up. The effect of using different forms of self-shielding factors on burn-up calculation for ETRR-2 reactor has been determined. The results of the two approaches are inter-compared up to 50% burn-up
Development and verification of Monte Carlo burnup calculation system
Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)
New Burnup Calculation System for Fusion-Fission Hybrid System
Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise
40 CFR 89.207 - Credit calculation.
2010-07-01
... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Credit calculation. 89.207 Section 89... Trading Provisions § 89.207 Credit calculation. (a) Requirements for calculating NO X credits from Tier 1 engines rated at or above 37 kW. (1) For each participating engine family, emission credits (positive...
A calculational benchmark focused on WWER-440 burnup credit, simular to that of the OECD/NEA/NSC Burnup Credit Criticality Benchmark Working Group, was proposed on the 96'AER Symposium. Its first part, CB1, was specified there whereas the second part, CB2, was specified a year later, on 97'AER Symposium in Zittau. This paper brings a final statistical evaluation of CB1 results and summarizes the CB2 results obtained to date. Further, the effect of an axial burnup profile of WWER-440 spent fuel on criticality ('end effect') is proposed to be studied in the CB3 benchmark problem of an infinite array of WWER-440 spent fuel rods as specified in the paper. (Authors)
Non destructive assay of nuclear LEU spent fuels for burnup credit application
Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron
SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2
1 - Description of program or function: SWAT evaluates isotopic composition of spent nuclear fuel, especially for burnup credit issues by driving codes SRAC95 and ORIGEN2.1 or ORIGEN2. SWAT is an automated driver code system. At the initial development phase, it was constructed by combining source programs of SRAC and ORIGEN2. To overcome the problem associated with code updates, SWAT chose to use system function of UNIX operating system to execute SRAC95 and ORIGEN2. So that, SWAT is independent of development and modification of SRAC95 and ORIGEN2.1. In SWAT, ORIGEN2(82) or ORIGEN2.1 is used for burnup calculations using the matrix exponential method. An updated decay library is included in the distribution. SWAT uses SRAC95 for neutron spectrum and effective cross section calculation in 107 groups, using the collision probability method for given geometry and isotopic composition. One or two dimensional cell geometries are supported in SRAC95. NEA-1698/02: The main purpose of new package is to run SWAT on several machines not supported in previous package (IA64 under Linux, Windows with cygwin and Sun,...) and several commercial FORTRAN compiler (Intel, PGI, Fujitsu). 2 - Methods: In calculating the problem-dependent cross section in SWAT, the total burnup history is divided into 'burnup steps'. Power, boric acid concentration, temperature of each region, and void ratio of coolant are given as history data. For each burnup step, the neutron spectrum and effective cross section are evaluated by SRAC95 using the information given in previous burnup calculation and cell geometry information. The user can select geometry options for the collision probability method in SRAC95. 3 - Restrictions on the complexity of the problem: Resonance absorption calculation with ultra-fine group cross section can not be directly applicable for 2D geometry
40 CFR 91.1307 - Credit calculation.
2010-07-01
... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Credit calculation. 91.1307 Section 91...) CONTROL OF EMISSIONS FROM MARINE SPARK-IGNITION ENGINES In-Use Credit Program for New Marine Engines § 91.1307 Credit calculation. For each participating engine family, emission credits (positive or...
Burnup calculations using the ORIGEN code in the CONKEMO computing system
This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)
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
Feasibility and incentives for burnup credit in spent-fuel casks
The spent-fuel carrying capacities of previous-generation spent-fuel shipping casks have been primarily thermal and/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced considerably and cask capacities become criticality limited. Using burnup credit in the design of future casks can result in increased cask capacities as well as reduced environmental impacts and savings in time and money
Details on an actinide-only burnup credit application in the USA
Details on the Actinide-Only burnup credit assumptions that will be used for the CASTOR X/32 S cask are presented. Preliminary results show that using a conservative set of assumptions the cask will allow most fuel to be loaded without the addition of any additional reactivity control. With the addition of 8 control rod elements it is possible to load most of the rest of the fuel. (author)
Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel
This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan
Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel
Parks, C V; DeHart, M D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2000-03-13
This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.
The role of ORIGEN-S in the design of burnup credit spent fuel casks
Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative fresh fuel assumption be used in the criticality analysis. The U.S. Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process
Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors
Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package MTRPC system, using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTRPC Package, Empirical Formula For Fuel Burn-Up.
Miniature neutron source reactor burnup calculations using IRBURN code system
Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.
Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit
Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL
2015-01-01
Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k_{eff}) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in
This article summarises activities of the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) Expert Group on Burnup Credit Criticality, a subordinate group to the Working Part on Nuclear Criticality Safety (WPNCS). The WPNCS of the OECD/NEA coordinates and carries out work in the domain of criticality safety at the international level. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burnup credit. The activities of the expert group are aimed toward improving safety and identifying economic solutions to issues concerning the back-end of the fuel cycle. The main objective of the activities of the OECD/NEA Expert Group on Burnup Credit Criticality is to demonstrate that the available criticality safety calculational tools are appropriate for application to irradiated (burned) nuclear fuel systems and that a reasonable safety margin can be established. The method established by the expert group for investigating the physics and predictability of burnup credit is based on the specification and comparison of calculational benchmark problems. A wide range of fuel types, including PWR, BWR, MOX, and VVER fuels, has been or is being addressed by the expert group. The objective and status of each of these benchmark problems is reviewed in this article. It is important to note that the focus of the expert group is the comparison of the results submitted by each participant to assess the capability of commonly used code systems, not to quantify the physical phenomena investigated in the comparisons or to make recommendations for licensing action. (author)
Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments
The burnup of 17 fuel elements of the TRIGA Mark-II reactor in Vienna was measured. Different types of fuel elements had been simultaneously used for several years. The measured burnup values are compared with those calculated on the basis of core configuration and reactor operation history records since the beginning of operation. A one-dimensional, two-group diffusion computer code TRIGAP was used for the calculations. Comparison with burnup values determined by γ-scanning is also made. (orig./HP)
Development and validation of Monte-Carlo burnup calculation code MCNTRANS
A new nuclear fuel burnup calculation code MCNTRANS based on MCNP was introduced in this paper. The neutronics calculation parameter was extracted from the MCNP5 reaction rate tally result, while a graph theory algorithm was implemented to track the burnup chain and the analytic solution of the Bateman equation was given. At the same time, the detailed physical process was considered to improve the accuracy and serviceability of this code, and prediction-correction method was used to allow a large burnup step. The OECD/NEA and JAERI pin cell benchmark problems were used to validate the code MCNTRANS while a reference result was given by other code. It can be concluded that the calculation results of MCNTRANS are generally consistent with the experimental result and that of the other burnup codes, and part of the actinides and fission products calculation result show better accuracy. (authors)
The on-site wet-type spent fuel storage facility ISF-1 is currently used for interim storage of spent nuclear fuel removed from Chernobyl NPP power units. The results of ISF-1 preliminary criticality analyses demonstrated the need for using the burnup credit principle in nuclear safety analysis. This paper provides results from the selection and testing of computer codes for determining the isotopic composition of RBMK spent fuel. Assessment is carried out and conclusions are made on conservative approaches to fuel burnup credit in subsequent ISF-1 safety assessment. (author)
Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask
Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2008-01-31
The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in k_{eff} of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in k_{eff} of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% ^{235}U enrichment results in an increase in k_{eff}of--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.
COREBN: A core burn-up calculation module for SRAC2006
COREBN is an auxiliary code of the SRAC system for multi-dimensional core burn-up calculation based on the diffusion theory and interpolation of macroscopic cross-sections tabulated to local parameters such as burn-up degree, moderator temperature and so on. The macroscopic cross-sections are prepared by cell burn-up calculations with the collision probability method of SRAC. SRAC and COREBN have wide applicability for various types of cell and core geometries. They have been used mainly for the purpose of core burn-up management of research reactors in Japan Atomic Energy Agency. The report is a revision of the users manual for the latest version of COREBN served with the SRAC released in 2006. (author)
Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor
Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs
Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)
Boyarinov, V. F.; Davidenko, V. D.; Polismakov, A. A.; Tsibulsky, V. F. [Russian Research Center Kurchatov Inst., Nuclear Reactor Inst., 123182, Moscow (Russian Federation)
2006-07-01
Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)
Reactivity effects of nonuniform axial burnup distributions on spent fuel
When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup
Reactivity effects of nonuniform axial burnup distributions on spent fuel
Leary, R.W. II; Parish, T.A. [Texas A & M Univ., College Station, TX (United States)
1995-12-01
When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup.
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
Analysis of the burnup credit benchmark with an updated WIMS-D Library
The OECD/NEA Burnup Credit Benchmark was analyzed with the WIMSD5B code using a fully updated library based on ENDF/B-VI Revision 5 data. Parts-1A and 1B were considered. The criticality prediction tested in Part-1A was in very good agreement with the reference result. A slight trend to overestimate the absorption rate by the fission products was noted, which can be explained by spectral effects resulting from the coarseness of the WIMS-D 69-group energy grid. The isotopic composition prediction tested in Part-1B was within the uncertainty interval of the reference results, except for 109 Ag at lower burnup and 155 Gd in all the cases. For 109 Ag the cause of the discrepancy was the use of old fission yield data in generating the reference solution. Similarly for 155 Gd the difference was due to old 155 Eu capture cross sections. Compared to the measurements, a serious underprediction of Sm isotopes is observed. This could be due to problems in the measured values or in the nuclear data of Sm precursors. We conclude that our processing methods do not introduce significant errors to the basic nuclear data. Care should be taken in the interpretation of the reference average benchmark solution due to a possible bias towards the ENDF/B-V evaluated nuclear data files
Computational fluid dynamics analysis for K24B cask design with burnup credit
Korea Nuclear Engineering Service Corp. (KONES) has designed K24B cask for the storage and the transportation of 24 (CE-type 16x16) PWR assemblies. K24B cask is designed with considering burnup credit of spent fuel. In order to remove heat from the fuel assemblies effectively, the flow channels in the upper and the lower part of fuel assemblies are set up to promote the natural convection. Computational fluid dynamics analysis is carried out to estimate and assure the thermal integrity of K24B cask. Conduction and radiation heat transfer through the cask components and the natural convective heat transfer in the cask are simulated. As a result of the analysis, the maximum temperatures of the cask components are maintained below the operating temperature for the safety. Therefore, the design of K24B cask can satisfy the safety limit. (author)
Incentives for the allowance of ''burnup credit'' in the design of spent nuclear fuel shipping casks
An analysis has been completed which indicates that the consideration of spent fuel histories ('burnup credit') in the criticality design of spent fuel shipping casks could result in significant public risk benefits and cost savings in the transport of spent nuclear fuel. Capacities of casks could be increased considerably in some cases. These capacity increases result in lower public and occupational exposures to ionizing radiation due to the reduced number of shipments necessary to transport a given amount of fuel. Additional safety benefits result from reduced non-radiological risks to both public and occupational sectors. In addition, economic benefits result from lower in-transit shipping costs, reduced transportation fleet capital costs, and fewer cask handling requirements at both shipping and receiving facilities
Incentives for the allowance of burnup credit in the design of spent nuclear fuel shipping casks
An analysis has been completed which indicates that the consideration of spent fuel histories ('burnup credit') in the criticality design of spent fuel shipping casks could result in considerable public risk benefits and cost savings in the transport of spent nuclear fuel. Capacities of casks could be increased considerably in some cases. These capacity increases result in lower public and occupational exposures to ionizing radiation due to the reduced number of shipments necessary to transport a given amount of fuel. Additional safety benefits result from reduced non-radiological risks to both public and occupational sectors. In addition, economic benefits result from lower in-transit shipping costs, reduced transportation fleet capital costs, and fewer cask handling requirements at both shipping and receiving facilities
Kinetic parameter calculation as function of burn-up of candu reactor
Kinetic parameter calculation as function of burn-up of candu reactor. Kinetic marameter calculation as function of burp-up of CANDU reactor with Canflex fuel type-CANDU has been done. This type of fuel is currently being develop, so kinetic parameter such as effective delay neutron fraction (.......), delay neutron decay constant ( .... ) and prompt neutron generation time ( ...... ) are very important for analysis of reactor operation safety. WIMS-CRNL code was used to generate macroscopic cross section and reaction rate based on transport theory. Fast and thermal neutron velocity and macroscopic cross section fission product of the unit cell were determined by KINETIC Code. The result of calculation showed that the value of effective delay neutron fraction was 7,785616 x 10-3 at the beginning of operation at burn-up of 0 MWD/T and after the reactor operated at burn-up of 7,2231 x 10-3 MWD/T was 4,962766 x 10-3, or reduced by 36%. The value of prompt generation time was 9,982703 x 10-4 s at the beginning of operation at burn-up of 0 MWD/T and 8,965416 x 10-4 s after the reactor operated at burn-up of 7,2231 x 103 MWD/T, or reduced by 10%. The result of calculation showed that the values of effective delay neutron fraction and prompt neutron generation time are still great enough
Burnup of low enriched metal uranium fuel of the RA reactor is described by two chain reactions. Energy balance and material changes in the fuel are described by systems of differential equations. Numerical integration of these equations is base on the the reactor operation data. Neutron flux and percent of Uranium-235 or more frequently yield of epithermal neutrons in the neutron flux, is determined by iteration from the measured contents of 106Ru, 134Cs and 137Cs in the irradiated fuel. The computer program was written in FORTRAN-IV. Burnup is calculated by using the measured activities of fission products. Burnup results are absolute values
Study of the acceleration of nuclide burnup calculation using GPU with CUDA
The computation costs of neutronics calculation code become higher as physics models and methods are complicated. The degree of them in neutronics calculation tends to be limited due to available computing power. In order to open a door to the new world, use of GPU for general purpose computing, called GPGPU, has been studied [1]. GPU has multi-threads computing mechanism enabled with multi-processors which realize mush higher performance than CPUs. NVIDIA recently released the CUDA language for general purpose computation which is a C-like programming language. It is relatively easy to learn compared to the conventional ones used for GPGPU, such as OpenGL or CG. Therefore application of GPU to the numerical calculation became much easier. In this paper, we tried to accelerate nuclide burnup calculation, which is important to predict nuclides time dependence in the core, using GPU with CUDA. We chose the 4.-order Runge-Kutta method to solve the nuclide burnup equation. The nuclide burnup calculation and the 4.-order Runge-Kutta method were suitable to the first step of introduction CUDA into numerical calculation because these consist of simple operations of matrices and vectors of single precision where actual codes were written in the C++ language. Our experimental results showed that nuclide burnup calculations with GPU have possibility of speedup by factor of 100 compared to that with CPU. (authors)
Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry
A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs
Burnup calculation capability in the PSG2 / Serpent Monte Carlo reactor physics code
The PSG continuous-energy Monte Carlo reactor physics code has been developed at VTT Technical Research Centre of Finland since 2004. The code is mainly intended for group constant generation for coupled reactor simulator calculations and other tasks traditionally handled using deterministic lattices physics codes. The name was recently changed from acronym PSG to 'Serpent', and the capabilities have been extended by implementing built-in burnup calculation routines that enable the code to be used for fuel cycle studies and the modelling of irradiated fuels. This paper presents the methodology used for burnup calculation. Serpent has two fundamentally different options for solving the Bateman depletion equations: 1) the Transmutation Trajectory Analysis method (TTA), based on the analytical solution of linearized depletion chains and 2) the Chebyshev Rational Approximation Method (CRAM), an advanced matrix exponential solution developed at VTT. The first validation results are compared to deterministic CASMO-4E calculations. It is also shown that the overall running time in Monte Carlo burnup calculation can be significantly reduced using specialized calculation techniques, and that the continuous-energy Monte Carlo method is becoming a viable alternative to deterministic assembly burnup codes. (authors)
A simplified burnup calculation strategy with refueling in static molten salt reactor
Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)
Applicability of the cross section adjustment method based on random sampling (RS) technique to burnup calculations is investigated. The cross section adjustment method is a technique for reduction of prediction uncertainties in reactor core analysis and has been widely applied to fast reactors. As a practical method, the cross section adjustment method based on RS technique is newly developed for application to light water reactors (LWRs). In this method, covariance among cross sections and neutronics parameters are statistically estimated by the RS technique and cross sections are adjusted without calculation of sensitivity coefficients of neutronics parameters, which are necessary in the conventional cross section adjustment method. Since sensitivity coefficients are not used, the RS-based method is expected to be practically applied to LWR core analysis, in which considerable computational costs are required for estimation of sensitivity coefficients. Through a simple pin-cell burnup calculation, applicability of the present method to burnup calculations is investigated. The calculation results indicate that the present method can adequately adjust cross sections including burnup characteristics. (author)
A higher initial 235U enrichment is currently required in the nuclear fuel fabrication specification to realize higher fuel burnup. Traditionally, in the criticality safety design of spent fuel (SF) storage and transportation (S/T) casks or facilities, the fuel is usually assumed to be at its full initial enrichment (so called fresh fuel assumption) to provide a large safety allowance, which is sometimes excessively given, for example requiring unnecessarily large space between fuel assemblies. The burnup credit taken for criticality safety design is firstly implemented to the SF Storage Rack of Rokkasho Reprocessing Facility, which is completed and expected for operation soon. Except for that, no burnup credit has been taken in criticality safety design for SF S/T casks or intermediate storage facilities in Japan. Since in the near future it is considered inevitable to handle spent fuel massively, it is desired to implement the rational S/T design saving safety and economy by taking into account the fuel burnup in the criticality safety control. Computer codes and data which are vital to assess criticality safety in the design stage of nuclear fuel cycle facility have been developed and prepared to constitute a Japanese criticality safety handbook at JAERI
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
Development of burnup calculation function in reactor Monte Carlo code RMC
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (authors)
The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C-E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA-AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies. (author)
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations
Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)
2008-04-15
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.
Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications
Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)
2016-06-15
Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.
Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))
1999-12-15
Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)
ANS/ENS tutorial session: Burnup credit issues in spent fuel transportation: Overview and objectives
A number of opportunities exist to increase the efficiency of the next generation of spent fuel shipping casks. Improving cask efficiency will not only reduce life cycle transportation costs, but also is consistent with maintaining public and occupational radiological risks and, more importantly, total risks (radiological and nonradiological) within the guidelines of the ''as low as reasonably achievable'' (ALARA) philosophy. Increases in cask capacities will reduce both the total number of shipments required to transport a given amount of fuel and the number of handling operations at both shipping and receiving facilities. Additional capacity increases can be achieved by implementing various design strategies based on new concepts and/or the actual characteristics of the majority of the spent fuel to be shipped in the future. For example, it has been determined that additional capacity increases can be achieved by taking credit for burnup, the reduced reactivity that results when fuel has been used to produce power in a nuclear reactor. That is, as the fuel is used the atoms of fissile material decrease, and neutron absorbers (or ''poisons'') that tend to retard the fission process are produced. 7 refs., 1 fig
The International Atomic Energy Agency convened a technical committee Meeting on Requirements, Practices and Developments in Burnup Credit (BUC) Applications in Madrid, Spain, from 22 to 26 April 2002. The purpose of this meeting was to explore the progress and status of international activities related to the BUC applications for spent nuclear fuel. This meeting was the third major meeting on the uses of BUC for spent fuel management systems held since the IAEA began to monitor the uses of BUC in spent fuel management systems in 1997. The first major meeting was an Advisory Group meeting (AGM), which was held in Vienna, in October 1997. The second major meeting was a technical committee meeting (TCM), which was held in Vienna, in July 2000. Several consultants meetings were held since 1997 to advise and assist the IAEA in planning and conducting its BUC activities. The proceedings of the 1997 AGM were published as IAEA-TECDOC-1013, and the proceedings of the 2000 TCM as IAEA-TECDOC-1241. BUC for wet and dry storage systems, spent fuel transport, reprocessing and final disposal is needed in many Member States to allow for increased enrichment, and to increase storage capacities, cask capacities and dissolver capacities avoiding the need for extensive modifications. The use of BUC is a necessity for spent fuel disposal
TRIGA fuel burn-up calculations supported by gamma scanning
High resolution gamma-ray spectroscopy based non-destructive methods is employed to measure spent fuel parameters. By this method, the axial distribution of Cesium-137 has been measured which results in an axial burn up profiles. Knowing the exact irradiation history of the fuel, four spent TRIGA fuel elements have been selected for on-site gamma scanning using a special shielded scanning device developed at the Atominstitute. Each selected fuel element was transferred into the fuel inspection unit using the standard fuel transfer cask. Each fuel element was scanned in one centimetre steps of its active fuel length and the Cesium-137 activity was determined as a proved burn up indicator. The absolute activity of each centimetre was measured and compared with the reactor physics code ORIGEN2.2 results. This code was used to calculate average burn up and isotopic composition of fuel element. The comparison between measured and calculated results shows good agreement. (author)
Considering spent fuel management (storage, transport and reprocessing), the approach using 'fresh fuel assumption' in criticality-safety studies results in a significant conservatism in the calculated value of the system reactivity. The concept of Burnup Credit (BUC) consists in considering the reduction of the spent fuel reactivity due to its burnup. A careful BUC methodology, developed by CEA in association with AREVA-NC was recently validated and written up for PWR-UOx fuels. However, 22 of 58 French reactors use MOx fuel, so more and more irradiated MOx fuels have to be stored and transported. As a result, why industrial partners are interested in this concept is because taking into account this BUC concept would enable for example a load increase in several fuel cycle devices. Recent publications and discussions within the French BUC Working Group highlight the current interest of the BUC concept in PWR-MOx spent fuel industrial applications. In this case of PWR-MOx fuel, studies show in particular that the 15 FPs selected thanks to their properties (absorbing, stable, non-gaseous) are responsible for more than a half of the total reactivity credit and 80% of the FPs credit. That is why, in order to get a conservative and physically realistic value of the application keff and meet the Upper Safety Limit constraint, calculation biases on these 15 FPs inventory and individual reactivity worth should be considered in a criticality-safety approach. In this context, thanks to an exhaustive literature study, PWR-MOx fuels particularities have been identified and by following a rigorous approach, a validated and physically representative BUC methodology, adapted to this type of fuel has been proposed, allowing to take fission products into account and to determine the biases related to considered isotopes inventory and to reactivity worth. This approach consists of the following studies: - isotopic correction factors determination to guarantee the criticality
Verification of a Multi-group Cross Section Library for Burnup Calculation
Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of); Joo, Hang Yu [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
Despite satisfying the estimation of the neutronic parameters without depletion to some extent, it still requires detailed investigation of the behavior of a fuel with strong neutron absorber over its operating life time by nTRACER, the direct whole core calculation code with the conventional semi Predictor-Corrector method. This study is mainly focused on the verification of the newly generated multi-group library for burnup calculation by nTRACER through the analysis of its performance of depletion calculation of UO{sub 2} fuel with strong neutron absorbers such as Gadolinium. Firstly, the depletion calculation results of nTRACER are presented by comparing the evolution of k-inf and the inventories of commonly found important isotopes as a function of burnup in the cases of gadolinia(GAD)-bearing fuel pin and fuel assembly (FA) with those of MCNPX-version.2.6.0. The newly generated multi-group library for burnup calculation by nTRACER was verified through GAD-bearing fuel after the new approach of resonance treatment had been employed. Though very good agreement in the overall effect reflected on the multiplication factor of FA at BOC, the evolution of k-inf along fuel irradiation history was systematically well underestimated by nTRACER when compared to Monte Carlo results.
Pore pressure calculation of the UO{sub 2} high burnup structure
Gao, Lijun, E-mail: lijungaothu@gmail.com [Tsinghua University, 100084 Beijing (China); Science and Technology on Reactor System Design Technology Laboratory, P.O. Box 622-500, 610041 Chengdu (China); Chen, Bingde [Nuclear Power Institute of China, 610041 Chengdu (China); Xiao, Zhong [National Energy R and D Center on Advanced Nuclear Fuel, 610041 Chengdu (China); Jiang, Shengyao; Yu, Jiyang [Tsinghua University, 100084 Beijing (China)
2013-07-15
Highlights: • Pore pressure is calculated based on local burnup, density and porosity. • Ronchi's equations of state are used instead of van der Waals’ equation. • Pore pressure increases as HBS transformation begins and then stays constant. • A best approximated parameter used for pore pressure calculation is recommended. -- Abstract: UO{sub 2} high burnup structure has an important impact on fuel behavior, especially in case of reactivity initiated accident (RIA). Pore relaxation enhances local fuel swelling and puts additional load to the fuel cladding, which makes fuel more susceptible to pellet–cladding mechanical interaction induced failure. Therefore, pore pressure calculation becomes vital when evaluating the fuel failure. In this paper pore pressure is calculated as a function of pellet radial local burnup based on the basic characteristics of HBS using Ronchi's correlation. The results indicate that pore pressure will approach a stable value as HBS is developing. A best approximated C value of 55 N/m is recommended for pore pressure calculation.
Fuel burnup calculation for HEU and LEU cores of Ghana MNSR
Fuel burnup calculations have been performed using a computer program developed as part of this research work for both Highly Enriched Uranium (90.2 % U-235) and Low Enriched Uranium (12.6 % U-235) cores for Ghana Research Reactor-1 (GHARR-1). Fuel depletion analyses of the GHARR-1 core was also performed which provided an inventory of the actinides formed as a result of burnup. The effect of the production of plutonium isotopes with burnup on reactor operation was also estimated. A FORTRAN 95 code was written based on the three group model approach namely fast, resonance and slow (thermal) neutron reactions. The time rate of change of each fuel isotope density is given by a first order differential equation. A general solution for each fuel isotope rate equation was used as input for the computer code. These results are particularized to the case of constant power during a short time interval, during which the slow (thermal) neutron flux is considered constant. The results obtained for the HEU were in good agreement with those found in literature. Therefore, this code can be used to estimate the burnup of LEU fuel for core conversion from HEU to LEU. (au)
Evaluation of RSG-GAS Core Management Based on Burnup Calculation
Evaluation of RSG-GAS Core Management Based on Burnup Calculation. Presently, U3Si2-Al dispersion fuel is used in RSG-GAS core and had passed the 60th core. At the beginning of each cycle the 5/1 fuel reshuffling pattern is used. Since 52nd core, operators did not use the core fuel management computer code provided by vendor for this activity. They use the manually calculation using excel software as the solving. To know the accuracy of the calculation, core calculation was carried out using two kinds of 2 dimension diffusion codes Batan-2DIFF and SRAC. The beginning of cycle burn-up fraction data were calculated start from 51st to 60th using Batan-EQUIL and SRAC COREBN. The analysis results showed that there is a disparity in reactivity values of the two calculation method. The 60th core critical position resulted from Batan-2DIFF calculation provide the reduction of positive reactivity 1.84 % Δk/k, while the manually calculation results give the increase of positive reactivity 2.19 % Δk/k. The minimum shutdown margin for stuck rod condition for manual and Batan-3DIFF calculation are -3.35 % Δk/k dan -1.13 % Δk/k respectively, it means that both values met the safety criteria, i.e <-0.5 % Δk/k. Excel program can be used for burn-up calculation, but it is needed to provide core management code to reach higher accuracy. (author)
Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code
Full text: In this study, some neutronic calculations of first and second core cycles of 5 MW pool type TR-2 Research Reactor have been performed using Multi-Step Monte Carlo Burnup Code System MONTEBURNS and the results were compared with the values of experiments and other codes. Time dependent keff distribution and burnup ratios belong to first and second core cycles of TR-2 Research Reactor were compared and quite good consistence in the results were observed. After modeling the first and second core cycles of TR-2 with MCNP5 Monte Carlo code, MCNP5 used in MONTEBURNS code has been parallelized in 8 HP ProLiant BL680C G5 systems with 4 quad-core Intel Xeon E7330 CPU, utilizing the MPI parallel protocol and simulations were performed on the 128 cores Linux parallel computing machine system. The computation time was reduced by parallelization of MONTEBURNS which uses MCNP in many steps. (authors)
UK regulatory perspective on the application of burn-up credit to the BNFL thorp head end plant
In the UK the Health and Safety Executive, which incorporates the Nuclear Installations Inspectorate (NII), is responsible for regulation of safety on nuclear sites. This paper reports progress made in the application and development of a UK regulatory position for assessing licensee's plant safety caes which invoke the use of Burn-up Credit for criticality applications. The NII's principles and strategy for the assessment of this technical area have been developed over a period of time following expressions of interest from UK industry and subsequent involvement in the international collaborations and debate in this area. This experience has now been applied to the first main plant safety case application claiming Burn-up Credit. This case covers the BNFL Thermal Oxide Reprocessing Plant (THORP) dissolver at Sellafield, where dissolved gadolinium neutron poison is used as a criticality control. The case argues for a reduction in gadolinium content by taking credit for the burn-up of input fuel. The UK regulatory process, assessment principles and criteria are briefly outlined, showing the regulatory framework used to review the case. These issues include the fundamental requirement in UK Health and Safety law to demonstrate that risks have been reduced to as low as reasonably practicable (ALARP), the impact on safety margins, compliance and operability procedures, and the need for continuing review. Novel features of methodology, using a ''Residual Enrichment'' and ''Domain Boundary'' approach, were considered and accepted. The underlying validation, both of criticality methodology and isotopic determination, was also reviewed. Compliance was seen to rely heavily on local in-situ measurements of spent fuel used to determine ''Residual Enrichment'' and other parameters, requiring review of the development and basis of the correlations used to underpin the measurement process. Overall, it was concluded that the case as presented was adequate. Gadolinium reduction
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
PWR AXIAL BURNUP PROFILE ANALYSIS
The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)
PWR AXIAL BURNUP PROFILE ANALYSIS
J.M. Acaglione
2003-09-17
The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).
Several currently available methods to set burnup for depletion calculation are reviewed and discussed about its adequacy for criticality safety assessment of spent fuel (SF) transport casks by taking burnup credit (BC) into accounts. Various errors associated with BC criticality analyses are evaluated and converted to equivalent burnup to compare each other. Methods are proposed to use some reduced burnups equivalent to compensation of these associated errors. Effects of assumption of axial burnup distribution on criticality calculation and irradiation history parameter variation on depletion calculation are evaluated with OECD/NEA BC international benchmark data. (author)
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar
2005-07-01
Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)
Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment
Highlights: • MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. • Coupled system MCORE uses “modified predictor corrector” approach. • MCORE can use different depletion schemes and simulate fuel shuffling. • MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. • MCORE is also assessed by a fast reactor benchmark problem. - Abstract: An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA). The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800 pcm at the end of the burn-up for LEU assembly (40 MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values
Activity ratio measurement and burnup analysis for high burnup PWR fuels
Applying burnup credit to spent fuel transportation and storage system is beneficial. To take burnup credit to criticality safety design for a spent fuel transportation cask and storage rack, the burnup of target fuel assembly based on core management data must be confirmed by experimental methods. Activity ratio method, in which measured the ratio of the activity of a nuclide to that of another, is one of the ways to confirm burnup history. However, there is no public data of gamma-ray spectrum from high burnup fuels and validation of depletion calculation codes is not sufficient in the evaluation of the burnup or nuclide inventories. In this study, applicability evaluation of activity ratio method was carried out for high burnup fuel samples taken from PWR lead use assembly. In the gamma-ray measurement experiments, energy spectrum was taken in the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA), and 134Cs/137Cs and 154Eu/137Cs activity ratio were obtained. With the MVP-BURN code, the activity ratios were calculated by depletion calculation tracing the operation history. As a result, 134Cs/137Cs and 154Eu/137Cs activity ratios for UO2 fuel samples show good agreements and the activity ratio method has good applicability to high burnup fuels. 154Eu/134Cs activity ratio for Gd2O3+UO2 fuels also shows good agreements between calculation results and experimental results as well as the activity ratio for UO2 fuels. It also becomes clear that it is necessary to pay attention to not only burnup but also axial burnup distribution history when confirming the burnup of UO2+Gd2O3 fuel with 134Cs/137Cs activity ratios. (author)
For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95. percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input. Particularly it was shown that during the burnup, the variances when considering all the parameters uncertainties is equivalent to the sum of variances if the parameter uncertainties are sampled separately
Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core
In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)
Dufek, Jan; Anglart, Henryk
2013-01-01
Numerically stable Monte Carlo burnup calculations of nuclear fuel cycles are now possible with the previously derived Stochastic Implicit Euler method based coupling scheme. In this paper, we show that this scheme can be easily extended to include the thermal-hydraulic feedback during the Monte Carlo burnup simulations, while preserving its unconditional stability property. At each time step, the implicit solution (for the end-of-step neutron flux, fuel nuclide densities and thermal-hydrauli...
One of the traditional methods for determining the burnup of irradiated Light Water Reactor (LWR) fuel is the 148Nd method according to ASTM E-321. Probably one of the largest sources for systematic errors in this method is the assumed fission yield, requiring knowledge of the fraction of fissions occurring in different fissile nuclides. Another traditional method for burnup determination is based on the uranium and plutonium isotopic composition; however, this method is rarely used for LWR fuel due to its rather simplified and rough assumptions regarding the neutron spectrum and fission fractions. However, modern physics codes like CASMO and HELIOS are instead able to calculate the amount of fission products and actinides formed or consumed during reactor operation in a much more sophisticated way. Isotopic Dilution Analysis with chemical separation of elements of interest, followed by isotopic analysis with a Thermal Ionization Mass Spectrometer (TIMS) is a well established method for determining the content of selected isotopes in samples of dissolved irradiated fuel. This method normally provides very accurate and precise results. High Performance Liquid Chromatography (HPLC) for elemental separations, combined with Inductively Coupled Plasma Mass Spectrometry (ICP-MS) has become a much faster alternative. In general, this method is somewhat less precise. This disadvantage is at least partly compensated by the possibility of analyzing a larger number of nuclides and samples. The local pellet burnup of a well characterised fuel sample irradiated in the Swedish Boiling Water Reactor Forsmark 3 to about 60 MWd/kgU was determined. Weight ratios of neodymium isotopes relative to 238U, analysed by Isotope Dilution Analysis applying HPLC-ICP-MS as well as 235U and 239Pu abundance values were compared to corresponding values calculated by a single-assembly CASMO-4 simulation. Input data were generated by CASMO-4/POLCA7 core tracking calculations. The overall result
Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II
The isotopic compositions of 5 UO2 samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostly predicted within ±10%, the two codes giving quite different results, except for 242Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)
Graphical abstract: - Highlights: • Continuous Energy Monte-Carlo burnup code. • Instabilities of depletion calculation in loosely coupled system. • Advanced step model for burnup calculations. • Xenon profile oscillation in thermal reactor. • Parametrical study of instabilities. - Abstract: In this paper we use the Continuous Energy Monte-Carlo tool to expose the problem of burnup instabilities occurring in 1D and 2D systems based on PWR geometry. The intensity of power profile oscillations is studied as a function of geometry properties and time step length. We compare two step models for depletion procedure: classic staircase step model and stochastic implicit Euler method, that belongs to the family of predictor–corrector schemes. What is more, we consider the usage of better neutron source intensity value than beginning-of-step approximation. Required methodology was implemented into MCB5 simulation code. The practical conclusions about depletion calculations were formulated and the efficiency of advanced step model was confirmed