Validation studies of computational scheme for high-fidelity fluence estimations of the Swiss BWRs

Abstract

The paper presents recent activities conducted at the Paul Scherrer Institut (PSI) in relation to the development and validation of an integral calculation methodology based on CASMO-4/SIMULATE-3/MCNPX for accurate estimations of the fast neutron fluence (FNF) accumulated on reactor pressure vessels and internals of the operating Swiss BWRs. With this computational scheme, the default neutron source is set up at the pin-by-pin level with realistic spectrum specifications based on the actual reactor cycle-specific data from validated reference CASMO-4/SIMULATE-3 core analysis models. On this basis, MCNPX models are then applied for optimized calculations of the fast neutron flux at the RPV or at any other location of interest including e.g. at surveillance dosimeters. In that framework, the validation studies conducted so far have included one dosimeter set irradiated in a BWR/6 reactor during two relatively recent operating cycles. Although this first analysis revealed a satisfactory performance when comparing the calculation results to measured data, it was considered necessary to proceed with further sensitivity/optimization studies combined with an enlarged validation basis (i.e. using additional dosimeter sets) in order to strengthen the overall confidence in the scheme both at the qualitative and quantitative level. A summary of the recent progress achieved in these directions is presented in this paper. To start, recalling that BWRs are characterized by very complex and heterogeneous fuel assembly and core designs (e.g. pins with different enrichments and burnable absorber loading, partial length rods, fuel assemblies of different types in the core), the impact of such heterogeneities on FNF estimations is under investigation in order to determine the level of modeling details required for accurate computational schemes to be used for long-term evaluations of modern BWR core designs. Next, additional validation studies based on experimental dosimeter data obtained from the same BWR/6 reactor are presented. These enlarged validation studies involve the analysis of four dosimeter sets, each irradiated during one cycle (including the 3 first reactor operation cycles), and subsequently analyzed at the PSI Hot Lab shortly after the dosimeters extraction. All these additional validation studies are conducted using both the JEFF-3.1.1 and the ENDF/B-VII.0 continuous-energy neutron data libraries in order to assess the sensitivity of the PSI BWR computational scheme also upon the employed nuclear data.

Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies

“…Details and some illustrative examples on the topic can be found in [60][61][62][63]. For the case of the problems discussed in this paper the main parameter of interest is the fast neutron flux, i.e., the flux of neutrons above the energy threshold of 1 MeV (the fraction of neutrons with energy above~20 MeV is negligible in nuclear reactors).…”

“…Secondly, several "short-term" fluence monitors were irradiated during shorter operation periods (during single reactor cycles). As in the case of the PWR, specific activities of the fluence monitors have been later measured (at the PSI Hotlab) and used for the validation studies [62]. Hereafter all considered sources of BWR experimental data for FNF modeling validation will be called "Dosimeters (D)".…”

“…It can be noted that typically the C/E-1 values observed from the relevant validation studies in the dosimetry domain fall within~±20% (see e.g., [62,67,68]) and this level of accuracy is considered normally as acceptable [1,60]. In more details, only a rather general and approximate assessment could be given here for demonstration purpose, based e.g., on the results published in [62,68]: for the 93 Nb fluence monitor the average (C/E-1) values obtained with the PWR-GP, PWR-ST and BWR-D validation models (with ENDF/B-VII.0 in [62] and ENDF/B-VII.1 in [68] as the base ND and CM library) were respectively~5%,~5% and~18%.…”

“…In more details, only a rather general and approximate assessment could be given here for demonstration purpose, based e.g., on the results published in [62,68]: for the 93 Nb fluence monitor the average (C/E-1) values obtained with the PWR-GP, PWR-ST and BWR-D validation models (with ENDF/B-VII.0 in [62] and ENDF/B-VII.1 in [68] as the base ND and CM library) were respectively~5%,~5% and~18%. For the same models and the 54 Fe fluence monitors the (C/E-1) values were~−1%,~3% and~9%.…”

Assessment of representativity of the PETALE experiments for validation of Swiss LWRs ex-core dosimetry calculations

“…The original purpose of the program is to measure the neutron transport through materials typical for "heavy reflectors" as found, for example, in the EPR reactor design and thus to provide validation data for the selected materials cross-sections and reflector properties. However, the PETALE setup can also be useful for providing experimental data applicable for validation of calculation methodologies for typical LWR reactor dosimetry applications, like the one being developed and validated at PSI [1], [2], [3]. Thus, a pertinent question which can * e-mail: alexander.vasiliev@psi.ch be posed is: how representative are the planned PETALE measurements for the validation of FNF predictions in LWR structures?…”