[en] The ENDF-6 format used for evaluation files allows different interpolation modes between distributions related to discrete incident energies. It is the case for angular distributions, energy distributions, energy-angle distributions and continuous photon energy spectra. The linear-linear interpolation is very often used particularly in the JEFF and ENDF/B libraries. This has no major impact when the distribution covers all the range (e.g. for angular distributions and for prompt fission neutron spectra), but, in all the other cases it could induce a violation of kinematic energy conservation. Certain codes for nuclear data processing (e.g. NJOY) and particle transport (e.g. MCNP) reinterpret data of evaluation files by changing the interpolation mode, in order to improve compliance with the kinematics of the physical reactions, even if this approach does not allow the evaluators to subsequently test the quality of their work. A new option for the TRIPOLI-4"® code offers the users the opportunity to choose between the raw evaluation and a more physical treatment of these data for shielding and criticality benchmarks using neutron or coupled neutron-photon calculations. Thus, it is now possible to quantify the impact of such an interpretation, moreover this allows the evaluators to really test the relevance of their evaluation. The most sensitive configurations appear to be high energy neutron and coupled neutron-photon calculations. (author)

[en] Neutronic analyses rely on accurate cross-sections from nuclear data libraries. The differences in the cross-sections from the JEFF-3.1 and ENDF/B-VII.0 nuclear data libraries are evaluated in this paper by estimating their impact on the eigenvalue of Sodium-cooled Fast Reactors (SFR). This analysis is performed with three SFR cores characterized by a variety of power ranges, fuel types (uranium and thorium) and fuel forms (oxide and metal). It was observed that the reactivity value is consistently larger when JEFF-3.1 is used instead of ENDF/B-VII.0 and the discrepancy can be as high as 1,900 pcm. A perturbation analysis was performed with the ERANOS2.2 code system to evaluate the contributions by isotope, reaction and energy range to the observed reactivity changes. It was found that the capture cross sections of Th-232 and U-233 are both responsible for as high as 700 pcm of the discrepancy in the case of the thorium-fueled core. Plutonium and sodium are the two main contributors to the observed reactivity change in the case of uranium-fueled cores. The reactivity impact due to the differences in the cross-sections from ENDF/B-VII.0 and JEFF3.1 is also compared to the uncertainties calculated respectively with the COMMARA-2.0 and the COMAC.V0 covariance matrixes. (author)

[en] New generations of simulation tools responding to the challenges brought by the advanced features of both 3"r"d"+ generation Pressurized Water Reactor (PWR) cores and 4"t"h generation sodium fast neutron reactor (SFR) cores are taking shape. The developments of new simulations tools are also motivated by strict requirements of nuclear safety authorities. The new tools have the objective of setting new reference standards for neutronic prediction and will take advantage of innovative algorithms which have been implemented in existing CEA codes, such as ERANOS (fast reactors) and APOLLO2 (PWR); the new codes should at the same time remove remaining calculation errors. Although innovative algorithms have been filling the gaps which did exist 40 years ago between tools specifically dedicated to either thermal neutron cores or fast neutron ones, there remains a series of algorithms which deserve particular attention: the treatment of leakage in cell calculations. This paper describes methods for treating neutron leakage in self-shielding calculations with the sub-group method, and in the cell balance calculation. Applications of the MOC method of solution to treat neutron leakage are described. The application of the MOC can eliminate approximations at the cell interfaces while maintaining precise neutron leakage treatment. The new APOLLO3"® code, presently under development at CEA, is candidate for hosting such algorithms. (author)

[en] Surveillance and diagnostics of core barrel vibrations has been performed in the Swedish Ringhals PWRs for several years, with main focus on the pendular motion (beam mode). The monitoring of the beam mode showed that the amplitude of the corresponding peak in the ex-core neutron spectra increases along the cycle, and decreases after refueling. Previous investigations on the reason of this behaviour, i.e. whether it is due to the increase of the core barrel vibration amplitude or to the increase of the neutron physics coupling between vibrations and neutron noise, were not decisive. The objective of the work reported here is to clarify this question. From frequency analysis, two modes of vibration have been identified in the frequency range of the beam mode. Several results coming from the trend analysis performed during recent years indicate that one of the modes is due to the core barrel motion itself and remains constant during cycle, and the other is due to the individual flow induced vibrations of the fuel elements, showing an increasing trend during the cycle. In this work, the method to separate the contributions from the two modes has been refined, and the results of this approach to the latest measurements are presented. The results confirm the origin of the two vibration modes and show constant amplitude of the core barrel motion throughout the cycle. (author)

[en] This paper concerns with how to optimally determine enrichments of fuel assembly (FA) batches of beginning-of-life (BOL) and reload cycle cores of a pressurized water reactor (PWR) plant which runs on a multi-batch, multi-cycle fuel management scheme. As a way to determine the optimum FA enrichments, a multi-cycle, multi-objective FA loading pattern (LP) optimization problem for the transient cycle cores involving the BOL and the reload cycle cores of the PWR plant is solved by the adaptively constrained discontinuous penalty function-based (ACDPF-based) multi-objective simulated annealing (MOSA) algorithm in combination with the commercial core neutronics design code ASTRA (Advanced Static and Transient Reactor Analyzer). The applicability and the effectiveness of the ACDPF-based MOSA algorithm is examined in terms of its solution to the first three transient cycle FA LP optimization problem of Yonggwang Nuclear Unit 4 (YGN4) a PWR plant in Korea. The practicality and usefulness of the ACDPF-based MOSA algorithm as an optimizer to determine optimum enrichments of BOL and reload cycle cores are discussed. (author)

[en] This paper reviews the history of development of AEGIS/SCOPE2, an advanced in-core fuel management code for PWRs. The initial project, development of a proto-type code, was started in 1996 as a feasibility study of the advanced calculation method/algorithm for advanced computation environments such as distributed parallel computers like PC-clusters which are commonly used nowadays. With success of development of the prototype code, a production-level advanced core calculation code, SCOPE2, was developed followed by AEGIS, an advanced assembly calculation code. These codes have been developed on the basis of the object-oriented programming approach and the agile software development. The authors extracted the key factors for success of the project as good practices from the viewpoint of code design, implementation, project management and verification and validation. Those practices are universal and may be applicable to any projects in the future. (author)

[en] The possible delay of decades for the deployment of fourth generation reactors brings up new issues. Countries like France which are storing Plutonium for years in prediction of starting FBR could have to adapt their plutonium management strategy. We have compared different strategies for plutonium recycling in PWR reactors. We have chosen to limit the scope of this study to PWRs considering the standard uranium cycle, and we investigated the influence of the heterogeneity in the assembly that would contain the recycled plutonium. The comparison of different strategies is made at steady-state. This paper present a specific method developed to allow complete and detailed studies of equilibrium scenarios and how it has been implemented and integrated inside the open-source MURE package. We will then discuss of the results obtained through this method apply to PWRs loaded with homogeneous and heterogeneous assemblies using the following criteria: resource consumption, Pu inventories in the cycle and waste production. (author)

[en] This paper presents an assessment conducted at the Paul Scherrer Institut (PSI) of two Monte Carlo-based codes, namely Monteburns and Serpent, for LWR depletion calculations. First, a validation of predicted spent fuel compositions is presented on the basis of a selected Swiss spent fuel sample from the LWR-PROTEUS Phase II program. Second, an additional assessment is performed for an OECD/NEA burnup credit benchmark which is also related to isotopic predictions. In both cases, equivalent pin-cell models are applied and the results are presented in terms of relative isotopic concentrations between measured/benchmark and predicted values. Additional comparisons are also made to previously published PSI results using the MCNPX/CINDER and CASMO-4E codes. The results show that all codes provide an overall good agreement with experimental data although a few specific trends are observed for some nuclides which are pointed out and briefly discussed. (author)

[en] Several years ago, the OECD/NEA Nuclear Science Committee (NSC) established the Expert Group on Uncertainty Analysis in Modeling (UAM-LWR) after thorough discussions of the demands from nuclear research, industry, safety and regulation to provide the best estimate predictions of nuclear systems parameters with their confidence bounds. UAM objectives include among others, the quantification of uncertainties of neutronic calculations with respect to their value for the multi-physics analysis. Since the kinetics parameters and their uncertainties are of particular interest for these studies the deterministic approaches for analysis of uncertainties in nuclear reactor kinetic parameters (neutron generation lifetime and delayed neutron effective fraction) have been developed in frame of the UAM-LWR. The approach uses combination of generalization of perturbation theory to reactivity analysis, and the Generalized Perturbation Theory (GPT) for sensitivity computation. It has been applied to the UAM complementary fast neutron SNEAK test case that has unique set of experimental data for β_e_f_f. In this example, the covariance matrices of nuclear data have been derived from COMMARA library by Bayesian adjustment upon the set of the fast neutron integral benchmarks with the BERING code package. Then, the kinetic parameters uncertainties with their correlations have been applied to simplified model of a reactivity insertion transient where relative uncertainty of power peak was taken as figure of merit. The results demonstrate that the uncertainties due to nuclear data impact significantly the energy release in a coupled transient modeling. It was also found that such uncertainties become higher if the correlations between uncertainties of different lumped parameters are taken into account. (author)

[en] The systematic method with least squares method to evaluate high accurate equivalence factors, and non-linear “Equivalent Pu-239” formula which is applicable to broad range of Pu isotopic composition, have been developed. And the method to evaluate each MOX fuel rod type's equivalence factors with high accurate equivalence for local power peaking has been developed. The applicability of the methods was confirmed for BWR MOX fuel. (author)