Burnup Credit Research Articles

A possibility of highly efficient uranium utilization with a pebble bed fast reactor

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.

Calibration of Burnup Monitor Installed in Rokkasho Reprocessing Plant

Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid.

Analysis of the Burnup Credit Benchmark with an Updated WIMS-D Library

Analysis of the Burnup Credit Benchmark with an Updated WIMS-D Library

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel—I: Methodology Overview

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 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am 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 SNF created during burnup in the reactor core and subsequent decay. Chemical assay benchmarks are used for this purpose, in conjunction with a method for assessing the calculational bias and uncertainty for each isotope.2. Validate a computer code system to predict the subcritical multiplication factor keff of an SNF package by use of UO2 and UO2/PuO2 critical experiments. The method uses an upper safety limit on keff (which can be a function of trending parameters) to ensure that the calculated keff when increased for the bias and uncertainty is <0.95.3. Establish conditions for the SNF (depletion analysis) and package (criticality analysis) that bound keff. Bounding axial and horizontal profiles must be established to ensure that the “end effect” and “horizontal effect” are 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.5. Verify by measurement that SNF assemblies meet the package-loading criteria, and confirm proper assembly selection prior to loading.

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel–II: Validation

The calculation of isotopic concentrations in spent nuclear fuel (SNF) assemblies and the subcritical multiplication factor of SNF packages are two of the essential requirements of the actinide-only burnup credit methodology. To justify the accuracy of the computed values, the code systems used to perform the calculations must be validated. Here, the techniques used for actinide-only burnup credit isotopic and criticality validation are presented and demonstrated.Fifty-four chemical assays are included in the isotopic validation benchmark set. To perform the validation, the samples are analyzed to obtain isotopic concentrations for each of the isotopes included in the methodology (234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am). Correction factors are computed based on the measured and calculated values, which are then used to conservatively bias computed isotopic concentrations.For the criticality validation, 57 critical experiments are included in the benchmark set. The set is composed of 21 UO2 and 36 mixed-oxide experiments, which are analyzed to determine the bias and corresponding uncertainty, ultimately resulting in an upper safety limit. This limit represents the maximum computed keff value that would be considered subcritical.

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel–III: Bounding Treatment of Spatial Burnup Distributions

A flat, uniform axial burnup assumption, preferred for its computational simplicity, does not always conservatively estimate the pressurized water reactor spent-fuel-cask multiplication factors. Rather, the reactivity effect of the significantly underburned fuel ends, usually referred to as the “end effect,” can be properly treated by explicit modeling of the axial burnup distribution based on limiting axial burnup profiles. An alternative approach to this laborious explicit modeling is to augment the multiplication factor determined from an axially uniform calculation by an appropriate keff bias. Based on the observation that the end effect increases with a decrease in the cask size, conservative keff bias curves are determined by applying the limiting axial burnup profiles and assuming a single-assembly cask configuration. However, because of their conservative nature, the keff bias curves are not recommended unless there is a large reactivity margin in the particular cask of interest.The horizontal burnup distribution poses less reactivity concern simply because the limiting arrangement in a cask is an unlikely event. The possibility of two or more assemblies with low burnup zones placed inward and next to each other is small, while the underburned fuel ends will surely be next to each other. Regardless, the reactivity effect of the horizontal burnup distribution is bounded by assuming a conservative horizontal burnup gradient within individual assemblies and the most reactive arrangement of multiple assemblies in spent nuclear fuel casks. This approach can have a significant effect on small cask designs where the orientation of fuel assemblies has a substantial influence on the calculated multiplication factor because of the large radial neutron leakage.

Actinide-Only Burnup Credit for Spent Fuel Transport

A conservative methodology is described that would allow taking credit for burnup 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 burnup 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.

Feasibility Assessment of Burnup Credit in the Criticality Analysis of Shipping Casks with Boiling Water Reactor Spent Fuel

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 basketmore » 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.« less

Validation of SCALE-4 for Burnup Credit Applications

In the past, criticality analysis of pressurized water reactor (PWR) fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at Oak Ridge National Laboratory (ORNL) in support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation approach of criticality safety methods to be used in burnup credit cask design. The date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The SCALE code package is a well-established code system that has been widely used in away from reactor applications. Criticality safety analyses are performed via the criticality safety analysis sequences (CSAS) and spent-fuel characterization via the shielding analysis sequence (QSAS) and spent-fuel characterization via the shielding analysis sequence (SAS2H). The SCALE 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data has been used for all calculations. The American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors of correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications using measured critical configurations includes: 1. fresh fuel critical experiments having geometric and nuclear characteristics similar to PWR spent fuel in storage and transport configurations 2. commercial PWR hot-zero-power and hot-full-power reactor critical configurations. The ability to closely predict reactor critical conditions is important in the validation of a methodology for spent-fuel applications because input data are determined based on relatively little detail of reactor core operation. Such limited information is expected to be representative of data available when burnup credit calculations are being performed in the determination of optimum cask loadings. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

Burnup Credit Experiences with the GA-4 Cask

General Atomics (GA) is developing two legal weight truck casks for shipping spent fuel from both pressurized water reactors (PWRs) and boiling water reactors (BWRs). The GA-4 (4PWR) and GA-9 (9BWR) casks are high-capacity legal weight truck casks designed to transport light water reactor spent-fuel assemblies. The GA-9 cask can meet the criticality safety requirements using the fresh fuel assumption. To maintain a capacity of four PWR spent-fuel assemblies, the GA-4 cask uses burnup credit as part of the criticality control for initial enrichments >2.9 wt% 235 U. Using the U.S. Department of Energy Burnup Credit Program as a basis, GA has performed burnup credit analysis for the GA-4 cask. The approach to calculating the minmum burnup requirement takes into account all of the key parameters affecting k eff . It is based on technically sound principles and conservatively increases the burnup requirement for a given enrichment to account for all uncertainties and biases associated with the calculations

Evaluation of Burnup Credit for Fuel Storage Analysis—Experience in Spain

Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (reracking) and the implementation of new analysis approaches with reduced conservatism (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); another BWR storage analysis is being developed at this moment.The elimination of the “fresh fuel assumption” increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduce new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis.The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible.

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 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 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 Credit in Spent Fuel Transport to Cogema LA Hague Reprocessing Plant

Among the spent fuel elements received yearly at the COGEMA La Hague reprocessing plant, an ever growing number benefit from burnup credit for transport or storage purposes. The burnup credit level is determined by the package or storage rack criticality assessment and approved by the French safety authorities. Depending on this level, the fuel irradiation is controlled either by qualitative or quantitative measuring techniques. A description of the documentary and operational system developed by COGEMA and NTL for fuel identification and burnup control is presented.

Canister Filling Materials - Design Requirements and Evaluation of Candidate Materials

AbstractSKB has been evaluating a copper/steel canister for use in the disposal of spent nuclear reactor fuel. The canister consists of an outer layer of about 50 mm thickness of copper to provide corrosion resistance and an inner layer of similar size to provide structural strength. Once the canister is breached by corrosion, it is possible that the void volume inside the canister might fill with water. Water inside the canister would moderate the energy of the neutrons emitted by spontaneous fission in the fuel. If the space in the canister between and around the fuel pins is occupied by canister filling materials, the potential for criticality is avoided. We have developed a set of design requirements for canister filling material for the case where it is to be used alone, with no credit for burnup of the fuel or other measures, such as the use of neutron absorbers. Requirements were divided into three classes: essential requirements, desirable features, and undesirable features. The essential requirements are that the material fill at least 60% of the original void space, that the solubility of the filling material be less than 100 mg/l in pure water or expected repository waters at 50°C, and that the material not compact under its own weight by more than 10%. In this paper we review the reasons for these requirements, the desirable and undesirable features, and evaluate 11 candidate materials with respect to the design requirements and features. The candidate materials are glass beads, lead shot, copper spheres, sand, olivine, hematite, magnetite, crushed rock, bentonite, other clays, and concrete. Emphasis is placed on the determination of whether further work is needed to eliminate uncertainties in the evaluation of the ability of a particular filling material to be successfully used under actual conditions, and on the ability to predict the long-term performance of the material under the repository conditions.

Feasibility and Incentives for Burnup Credit in Spent-Fuel Transport Casks

Feasibility and Incentives for Burnup Credit in Spent-Fuel Transport Casks