Investigation of ADS-Type Heterogeneities in the MUSE4 Critical Configuration

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Characterisation of radial reaction rate distributions across the 92-pin section of a SVEA-96 Optima2 assembly

Characterisation of the SVEA-96 Optima2 boiling water reactor assembly, in terms of the radial distributions of normalised total fission and 238U capture rates, is reported at its central elevation, i.e. at the 92-pin section, where the one-third part-length pins are replaced by water. Measurements performed in the PROTEUS facility are compared with MCNPX predictions. The calculation model included the measured locations of the SVEA-96 Optima2 assemblies and sub-assemblies, within the PROTEUS test zone. Predicted and experimental fission and 238U capture rates are found to agree, respectively, within 3.5% and 4% for all pins. Fission rates in the burnable-absorber UO2-Gd2O3 fuel pins have been predicted without bias using the ENDF/B-VI data library but show an average 1.4% under-prediction with the JEFF-3.1 data library. A slight overestimation of the total fission rate in the pins located at the periphery of the assemblies was observed and has been attributed to an inaccurate modelling of the pin positions. However, there was no systematic bias observed due to the absence of the one-third pins at the corners of the assembly.[All rights reserved Elsevier]

Importance of the MUSE experiments for emerging ADS concepts from the nuclear data viewpoint

The current investigation, conducted in the general framework of the MUSE program ("MUltiplication avec une Source Externe"), considers the representativity of a specific configuration of its fourth phase (M4SC2), which is driven by an external D(d,n)He3 or T(d,n)He4 neutron source, with respect to current concepts of eXperimental Accelerator Driven Systems (XADSs) with gas (He), Na and Pb/Bi coolants. The study has been carried out from the nuclear data viewpoint, with the external source being accounted for in an appropriate manner. In this context, data sensitivity/uncertainty analyses based on first-order perturbation theory calculations have been performed using the deterministic code ERANOS (Version 2.0) in conjunction with its adjusted nuclear data library ERALIB-1. It is found that the M4SC2 configuration, independent of the external source, is quite representative of the different XADSs for actinide capture reactions at the centre of the fuel zone, relative to 239Pu fission at the same location. For the case of a threshold fission reaction, such as that in 238U, the sensitivity to the external source is significantly higher. With respect to the corresponding spectral index, M4SC2 with the D(d,n)He3 source remains quite representative of the He- and Na-cooled XADSs. For the system with Pb/Bi coolant, on the other hand, effects of uncertainties associated with the data for these two nuclides and their low content in the MUSE configuration result in significantly lower associated representativity factors. A better overall representativity of the Pb/Bi-cooled XADS is expected to be achieved by the new MUSE_Na/Pb configuration. [All rights reserved Elsevier

Importance of the MUSE experiments for emerging ADS concepts from the nuclear data viewpoint

International audienceThe current investigation, conducted in the general framework of the MUSE program(‘‘MUltiplication avec une Source Externe’’), considers the representativity of a specific configurationof its fourth phase (M4SC2), which is driven by an external D(d,n)He3 or T(d,n)He4neutron source, with respect to current concepts of eXperimental Accelerator Driven Systems(XADSs) with gas (He), Na and Pb/Bi coolants. The study has been carried out from thenuclear data viewpoint, with the external source being accounted for in an appropriate manner.In this context, data sensitivity/uncertainty analyses based on first-order perturbation theorycalculations have been performed using the deterministic code ERANOS (Version 2.0) inconjunction with its adjusted nuclear data library ERALIB-1.It is found that the M4SC2 configuration, independent of the external source, is quite representativeof the different XADSs for actinide capture reactions at the centre of the fuel zone,relative to 239Pu fission at the same location. For the case of a threshold fission reaction, suchas that in 238U, the sensitivity to the external source is significantly higher. With respect to the corresponding spectral index, M4SC2 with the D(d,n)He3 source remains quite representativeof the He- and Na-cooled XADSs. For the system with Pb/Bi coolant, on the other hand,effects of uncertainties associated with the data for these two nuclides and their low contentin the MUSE configuration result in significantly lower associated representativity factors.A better overall representativity of the Pb/Bi-cooled XADS is expected to be achieved bythe new MUSE_Na/Pb configuration

Comparison of 3D reaction rate distributions measured in an optima2 BWR assembly with MCNPX predictions

As part of a joint research programme between the Paul Scherrer Institute (PSI) and swissnuclear, with the co-operation of the Leibstadt nuclear power plant in Switzerland and fuel suppliers Westinghouse Sweden, measurements and calculations have been made of the axial and radial distributions of fission and 238U capture rates in the fuel rods of a Westinghouse SVEA-96 Optima2 boiling water reactor assembly. The measurements, made in the zero-energy research reactor PROTEUS at PSI, have been compared with calculations carried out using the Monte Carlo code MCNPX. The results reported are for the regions near the ends of the part-length fuel rods, which are a feature of SVEA-96 Optima2 assemblies. The sudden increase in moderation above the ends of the part-length rods leads to power peaking in the adjacent rods. Careful attention needs to be given to this phenomenon in the deployment of such fuel, the present paper providing experimental evidence for the ability of a stochastic code to predict such effects. [All rights reserved Elsevier]

Validation of Monte Carlo predictions of LWR-PROTEUS safety parameters using an improved whole-reactor model

The recent experimental programme conducted in the PROTEUS research reactor at the Paul Scherrer Institute (PSI) has concerned detailed investigations of advanced light water reactor (LWR) fuels. More than fifteen different configurations of the multi-zone critical facility have been studied, each of them requiring accurate estimation of operational safety parameters, in particular the critical driver loadings, shutdown rod worths and the effective delayed neutron fraction eff. The current paper presents a full-scale 3D Monte Carlo model for the facility, set up using the MCNPX code, which has been employed for calculation of the operational characteristics for seven different LWR-PROTEUS configurations. Thereby, a variety of nuclear data libraries (viz. ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JEFF3.1, JENDL3.2, and JENDL3.3) have been used, and predictions of keff and shutdown rod worths compared with experimental values. Even though certain library-specific trends have been observed, the keff predictions are generally very satisfactory, viz. with discrepancies of 0.5% between calculation (C) and experiment (E). The results also confirm the consistent determination of reactivity variations, the C/E values for the shutdown (safety) rod worths being always within 5% of unity. In addition, the MCNP modelling of the multi-zone reactor has yielded interesting results for the delayed neutron fraction (eff) in the different configurations, a breakdown being made possible in each case in terms of delayed neutron group, fissioning nuclide, and reactor region. [All rights reserved Elsevier]

Radial and azimuthal 235U fission and 238U capture distributions in BWR UO2 pins: CASMO-4 and MCNP4C versus activation foil measurements

In the context of the LWR-PROTEUS program, radial and azimuthal 235U fission (F5) and 238U capture (C8) rate distributions have been calculated for zero-burnup pins of a Westinghouse SVEA-96 Optima2 boiling water reactor fuel assembly using the stochastic MCNP4C and the deterministic CASMO-4 codes. The within-pin F5 distributions predicted by the two codes are in very good agreement; the C8 distributions are more pronounced, and there are significant discrepancies between the codes, both azimuthally and radially. The calculations have been compared with experimental results obtained from activation foil measurements in two pins of the assembly irradiated in the center of the PROTEUS test zone. The measurements confirm that the two codes can accurately predict the radial and azimuthal F5 distributions but that MCNP4C within-pin C8 distributions are much more accurate than those of CASMO-

Validation of an MCNP4B whole-reactor model for LWR-PROTEUS using ENDF/B-V, ENDF/B-VI, and JEF-2.2 cross-section libraries

A detailed three-dimensional, continuous-energy MCNP4B model of the LWR-PROTEUS critical facility has been developed for the analysis of whole-reactor characteristics using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section sets. The model has been applied to the determination of the critical loading, as well as the evaluation of reactivity worths for safety/shutdown rods, control rods, and individual driver-region fuel rods. The initially obtained results for the first configuration investigated (Core 1B) indicated that, for the same geometrical and materials specifications, the ENDF/B-V data library yields the closest critical prediction (discrepancy of 64040 pcm), followed by ENDF/B-VI (98040 pcm) and JEF-2.2 (134040 pcm). The differences in results between the three data libraries were studied by considering the contributions of individual materials to the neutron balance. 235U and 238Pu cross-sections from JEF-2.2, for example, explain an effect of ~400 pcm. Refinement of the materials specifications in the MCNP4B whole-reactor model, in particular the impurities assumed for the graphite driver of the driver and reflector regions, has been shown to reduce the final discrepancy of the ENDF/B-V based keff result to ~0.2%. The MCNP4B results for relative reactivity effects, such as control rod worths, are found to agree within experimental errors with the measured value