Implementation of BUC in a spent fuel management system requires application of a system specific loading criterion which indicates the requirements the spent fuel has to meet in order to be allowed to be loaded in the system. Any loading criterion used in BUC correlates the safety limit of a parameter appropriate to characterize the spent fuel with the initial enrichment of the fuel. The value of the safety limit of such a safety parameter at given initial enrichment depends on
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the used level of BUC (net fissile content, actinide-only, actinide-plus-fission-product level etc, cf. IAEA-TECDOC-1013, p.1) and
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the design of the spent fuel management system.
The safety parameter usually chosen to present the loading criterion is the average burnup of the spent fuel, i.e. the loading criterion is usually presented in form of a loading curve indicating the minimum average burnup necessary for fuel with a specific initial enrichment to be loaded in the spent fuel management system. For some systems, however, it is more convenient to present the loading criterion in a different form: For dissolver facilities in reprocessing plants, for instance, either the fissile content or the residual U235 enrichment of the spent fuel is chosen as safety parameter, i.e. the loading criterion indicates the maximum allowable fissile content or residual enrichment as a function of the initial enrichment.
The key steps in application of BUC to a spent fuel management system are the following:
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Safety assessment of the system including
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prediction of the spent fuel composition under bounding depletion conditions,
ocriticality calculation and evaluation of the fuel loading criterion for the
system.
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Application of the fuel loading criterion; this step consists in
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quantification and verification of the numerical value which the fuel to be loaded in the system has for the safety parameter chosen to present the loading criterion (e.g. average burnup, see above),
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implementation of a fuel loading procedure assuring compliance with the loading criterion.
Prediction of the isotopic inventory of the spent fuel by means of depletion calculations requires:
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definition of the fuel characteristics
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knowledge of the irradiation history of the fuel
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choice of the cooling time.
A loading criterion of a spent fuel management system designed for BUC usually applies to any fuel positions of the system. In fact a loading criterion can only be evaluated if no credit is taken for the real loading scheme. Therefore, evaluation of a loading criterion makes it necessary to look for a bounding irradiation history given by those depletion conditions (reactor operation conditions) which lead to the highest reactivity of the spent fuel. The depletion conditions are characterized by the following parameters or conditions:
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Fuel temperature
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Moderator temperature or density; void history (BWR reactor types)
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Presence of soluble boron (PWR reactor types)
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Use of fixed neutron absorbers in form of control rods or blades, control assemblies (VVER), burnable poison rods, axial shaping rods etc
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Operational strategies, reload patterns and core environment: specific power history, extended low power operation period at end of cycle, in-out/out-in strategies, presence of MOX fuel etc.
The depletion conditions are significantly impacted by fuel characteristics such as:
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the presence of integral burnable absorbers (in form of Gd or Er bearing fuel rods or so-called “IFBA rods” containing pellets with boron coating)
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axial zoning of initial enrichment and/or burnable absorbers, presence of axial blankets
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horizontally heterogeneous initial enrichment distributions (BWR, MOX fuel assemblies, VVER)
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presence of partial length fuel rods (BWR designs).
Criticality calculation and evaluation of the fuel loading criterion require:
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isotopic selection and validation
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validation of the criticality calculation code to be used,
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evaluation of the reactivity effect of axial and horizontal burnup profiles
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sensitivity studies on the reactivity effects of variations and tolerances in the parameters describing the characteristics of the spent fuel management systems.
By definition, for a specific initial enrichment the loading criterion provides a specific numerical value for the safety parameter chosen to present the loading criterion. This criterion must therefore cover the variety of reactivity effects due to the variety of axial and horizontal burnup profiles. So therefore, evaluation of a fuel loading criterion makes it necessary to look for a bounding axial burnup profile as well as for a bounding horizontal burnup profile. Since the shape of axial and horizontal burnup profiles and the reactivity effects due to these profiles change with the average burnup of these profiles the bounding axial burnup profile and the bounding horizontal burnup profile vary with the average burnup.
The complexity of fuel designs (modern BWR and VVER designs in particular), the
complexity of the depletion conditions (use of fixed neutron absorbers, operational strategies,
reload patterns and core environment), the complexity due to non-uniform burnup
distributions and the complexity of the design of spent fuel management systems require
careful choice of the BUC calculation methodology to be applied to assure sufficiently
accurate presentation of the physics of the problem to be solved. Accordingly, the first
technical topic on the agenda of the London TM, 2005, was:
• Technical topic 1: Principles of choosing the calculation methodology with respect to
the fuel design and the spent fuel management system.
The choice of the calculation methodology is inseparably linked with the isotopic selection and validation and the validation of the criticality calculation code chosen.
Isotopic validation and validation of the reactivity calculations are necessary conditions for demonstrating the adequacy of the chosen calculation methodology.
Therefore, the second technical topic on the agenda of the London TM was:
• Technical topic 2: Nuclear data and validation of depletion and reactivity
calculations.
Several papers were presented under topics 1 and 2. These papers addressed the following items:
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Establishment of a database of spent BWR fuel data in the USA including physical fuel data and reactor operating histories to support BUC analysis for BWR fuel; application of multivariate data analysis methods to identify data clusters and bounding conditions.
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BUC calculation methodologies for transport and storage casks:
A comprehensive survey of the US program to introduce actinide-plus-fission-product BUC in transport and storage casks was given. The activities relating to this program mainly address (a) the availability and applicability of existent isotopic assay data for validation of depletion calculations and existent critical experiments for validation of reactivity calculations, (b) the possibilities of increasing the size of isotopic assay data samples for the fission products relevant to BUC in particular, (c) the performance of new critical experiments with the relevant fission products, (d) the application of Sensitivity/Uncertainty tools for evaluating existing data and designing new experiments.
In a different paper the challenging idea is broadly outlined to apply BUC to actual cask loading schemes instead of generating a loading curve. This includes that the isotopic inventory and burnup distribution are calculated for each individual fuel assembly. This requires on-line core-following depletion calculation which is quasi-continuously recalibrated by means of in-core measurement. The paper gives no example, but it is obvious that the outlined procedure multiplies the complexity of BUC application with respect to burnup quantification and verification, depletion and reactivity calculation validation, implementation of cask loading.
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Studies of BUC methodology for spent fuel storage in the People’s Republic of China: A comprehensive program to introduce BUC in spent fuel storage has been initiated. This program includes validation of the calculational tools by means of experimental data and OECD/NEA BUC benchmarks as well as new critical experiments with spent fuel (from Qinshan nuclear power plant, for instance) planned to be performed at the China Institute of Atomic Energy.
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BUC application to the post-closure phase of a spent fuel repository:
Parametric criticality studies were presented for different fuel types, initial
enrichments, burnup values and scenarios to demonstrate the need for BUC to
minimize the probability of occurrence of the formation of critical and
supercritical configurations. The BUC calculation procedure used for these
studies has the capability of two and three dimensional calculations of pin power distributions, isotopic inventories and reactivity.
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Improvements, developments and validations of two and three dimensional depletion calculation codes:
The improvements in the TRITON depletion sequences of the SCALE system were presented: TRITON has been enhanced by the addition of depletion sequences which use KENO V.a/VI for three dimensional (3D) transport solutions. This enables the performance of direct 3D depletion calculations. The results of isotopic assay validation calculations performed by means of the one, two and three dimensional depletion sequences of the SCALE system were presented.
The development and validation of the integrated depletion code MVP-ORBURN were described. This code combines the continuous energy Monte Carlo code MVP with the point depletion code ORIGEN 2.
Results of BUC applications of MVP-ORBURN to PWR and BWR fuel were presented.
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Isotopic validation of BUC applications to VVER-440 spent fuel:
VVER-440 spent fuel isotopic assay data from RIAR Dimitrovgrad (ISTC project no. 2670): Radiochemical assay data for eight samples of spent VVER-440 fuel from Novovoronezh nuclear power plant unit 4 were presented. These data are the only spent VVER-440 fuel isotopic composition data publicly available outside of the Russian Federation which include fission products relevant to the actinide-plus-fission-product BUC level.
In a different paper these RIAR data served for validation of depletion calculations performed by means of the SAS2H sequence (employing ORIGEN-S) and the TRITON control module (using NEWT) of the SCALE system. These validation calculations included comparisons with 12 samples of actinide assay data taken at the Kurchatov Institute, Moscow, from irradiated fuel from Novovoronezh nuclear power plant unit 4. The calculations also included comparisons with the Takahama-3 assay data and the OECD/NEA VVER CB2-benchmark data.
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Evaluation of the REBUS experiments on PWR fuels: A feature of paramount importance in the REBUS program is that this program was aimed at providing an experimental database jointly usable for validation of depletion and reactivity calculation in such a way that a direct validation of the calculational tools commonly used in BUC criticality safety analysis is enabled (i.e.
estimation of keff rather than reactivity perturbation calculations). The REBUS program therefore included integral reactivity worth measurements using fuel bundles from commercial PWR samples and, afterwards, radiochemical assay of the irradiated fuel. Two papers were presented providing preliminary analysis results obtained with a couple of different calculation codes.
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Sensitivity and uncertainty studies of the applicability of critical experiments to BUC criticality calculations: Two papers addressing this item were presented:
In one paper the applicability of critical MOX experiments selected
from the International Handbook of Evaluated Criticality Safety
Benchmark Experiments (IHECSBE) to the use of BUC for the
compact storage installation at the VVER-440 Paks nuclear power
plant, Hungary, and the VVER-440 fuel transport casks C-30 and TK-6
was studied. Fission rates, capture rates and neutron fluxes were
calculated in five broad energy groups, and the relative importance of the uranium and plutonium isotopes was analyzed. From the comparisons of the results obtained for the experiments and the application cases it was concluded that a “stand-alone” use of the selected MOX experiments is not sufficient for validating the use of BUC for the application cases.The other paper presented outcomes of a study of the applicability of more than 1000 critical benchmark configurations primarily taken from the IHECSBE to the Generic Burnup Credit GBC-32 prototypical high capacity US rail cask assumed to be loaded with 32 fuel assemblies of the 17x17-25 type with an initial enrichment of 4.0 wt.-% U235, an average burnup of 40 MWd/kg U and a cooling time of 5 years. This study was performed with the aid of the Sensitivity/Uncertainty (S/U) analysis sequences TSUNAMI of the SCALE system at Oak Ridge National Laboratory (ORNL), USA. It was found that only MOX configurations from the French Haut Taux de Combustion (HTC) experimental series are applicable: 156 HTC MOX configurations were considered in the study, 143 were found to be applicable.+) None of the 978 analyzed non-HTC experiment configurations including high-, intermediate and low enriched uranium as well as plutonium and non-HTC MOX configurations was identified as applicable. Only 45 of the non-HTC MOX configurations were classified as marginally applicable.
However, an important aspect of assessing these outcomes is to consider whether all the BUC nuclides used in the application case (the GBC-32 cask) are really represented in the experiments. If some nuclides are missing in a critical configuration and if these missing nuclides have a significant reactivity worth in the application case, then this critical configuration is obviously tending to be rejected. The TSUNAMI procedure is therefore capable of calculating sensitivity profiles for any desired nuclide as a function of neutron energy. The degree of agreement between the nuclide-specific sensitivity profiles for experiments and the application case is described as “coverage”.
Coverage is provided by an experiment wherever the sensitivity profile of the experiment covers the sensitivity profile of the application case.
Experiments which were identified as “not applicable” in total can have significant degrees of coverage for specific nuclides. Examples for such cases are given in the presented paper. Therefore, methods are under development at ORNL which utilize the relevant information from such experiments.
Whilst in the study of the applicability of MOX experiments to VVER-440 spent fuel systems the experiments were selected in the traditional way (expert’s judgment: comparisons of materials, geometries, gross integral parameters such as moderation ratio, lethargy of average neutron energy causing fission), S/U evaluation methods like TSUNAMI allow detailed quantitative comparisons of the similarity of nuclear system with respect to the underlying nuclear data characterizing the isotopic compositions in all the material zones of the
+) Different from the recommendation reported in section 3.2.4.4 of this summary a criterion of ck ≥ 0.9 was used in the presented paper. Experiments with 0.8 < ck < 0.9 are identified as “marginally acceptable”.
systems and hence their impact on the neutron spectra and the reactivity. Even though covariance matrices containing all the information about the “uncertainties” (variances and correlations) of the nuclear data (cf. section 3.2.4.4 of this summary) are still under development (for BUC analyses in particular), and even though no theory based rationale has been found until now for the decision criterion whether an experiment can be regarded as acceptable or not, S/U evaluation methods such as TSUNAMI provide a powerful tool for the selection of experiments. In addition, because of the capability of calculating nuclear specific sensitivity profiles, S/U methods can be used to design new experiments such that coverage with application cases is achieved.
Once an adequate BUC calculation methodology is chosen and validated, the task is to determine a criticality safety acceptance criterion from which the fuel loading criterion for the application case, i.e. the spent fuel management system to be analyzed, can be derived. Determination of a criticality safety acceptance criterion requires a consistent calculation route
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to evaluate the experimental information from chemical assay data in order to consider the isotopic bias in the keff value of the spent fuel management system,
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to evaluate the experimental information from reactivity worth measurements and critical benchmark experiments to take account of the calculated bias in the keff value of the spent fuel management system due to the criticality calculation procedure applied,
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to determine bounding irradiation histories including specific depletion conditions (e.g. use of control rods).
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to evaluate the reactivity effects of axial and horizontal burnup profiles under the conditions of the spent fuel management system,
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to cover the variability in keff value of the system due to variations and tolerances in the parameters describing the characteristics of the system.
There is a wealth of ways to come to a consistent calculation route. For instance, the isotopic bias can:
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either be covered by applying nuclide specific number density correction factors derived from comparisons of predicted isotopic concentrations to chemical assay data;
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or explicitly calculated by means of sensitivity analyses of the impact of the bias of the concentrations of the individual isotopes on the keff value of the spent fuel management system.
The bias in the prediction of the BUC nuclide reactivity worths can:
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either be conservatively covered by applying penalty factors to the isotopic number densities used in the calculation of the keff value of the spent fuel management system;
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or evaluated by means of sensitivity based criticality validation techniques.
Obviously, different ways of establishing a consistent calculation route result in different degrees of conservatism maintained in the k
effvalue calculated for the spent fuel management system and hence in different safety margins. The third technical topic on the agenda of the London TM was accordingly described as “criticality safety criteria”:
• Technical topic 3: Criticality safety criteria
Several presentations addressing this topic were given. All these presentations together reflected the wealth of methods of establishing a consistent calculation route.
All aspects of determining a criticality safety acceptance criterion were covered for several fuel types and designs (PWR, BWR, VVER-440 and VVER-1000) and spent fuel managements systems (wet storage installations, transport and storage casks, dissolver facilities in reprocessing plants):
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Determination of bounding depletion conditions;
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Consideration of isotopic bias;
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Consideration of the reactivity bias in the reactivity calculations;
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Determination of bounding axial and horizontal burnup profiles and determination of the reactivity effects related to axial and horizontal burnup profiles and isotopic number density distributions;
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Evaluation of manufacturing tolerances and variations in the parameters describing fuel designs and spent fuel management systems.
In addition, the impact of the degree of conservatism due to the chosen depletion conditions and possible enhanced by applying conservatively derived isotopic correction factors to the number densities obtained from the depletion calculations on the estimates of reactivity effects due to non-uniform burnup distributions and manufacturing tolerances was described. This impact can be significant as exemplified for the reactivity effect due to axial burnup distribution.
The objective of a BUC criticality safety analysis of a spent fuel management system is to express the criticality safety acceptance criterion, derived for the system, in terms of a BUC loading criterion. Application of the loading criterion requires implementation of control and fuel handling procedures which assure compliance with the loading criterion. The fourth technical topic on the agenda of the London TM therefore was:
• Technical topic 4: Procedural compliance with the safety criteria