F. Jatuff

Neutronics investigations for the lower part of a Westinghouse SVEA-96+ assembly

Abstract Accurate critical experiments have been performed for the validation of total fission (Ftot) and 238U-capture (C8) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average 235U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing ~64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E) values have been compared in terms of C/E distributions. For Ftot, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C8. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of Ftot, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code).

LWR-PROTEUS verification of reaction rate distributions in modern 10 10 boiling water reactor fuel

Abstract HELIOS, CASMO-4, and MCNP4B calculations of reaction rate distributions in a modern, fresh 10 × 10 boiling water reactor fuel element have been validated using the experimental results of the LWR-PROTEUS Phase I project corresponding to full-density water moderation conditions (core 1B). The reaction rate distributions measured with a special gamma-scanning machine employing twin germanium detectors consisted of total fission Ftot and 238U-capture C8. The average statistical errors for the gamma scans were better than 0.5% for Ftot and 0.9% for C8. The rod-by-rod measurements were performed on 60 different fuel rods selected from the central part of a test zone consisting of actual, fresh SVEA-96+ fuel elements, thus gaining in realism by departing from conventional fuel rod mockups. In the case of Ftot, the root-mean-square (rms) of the rod-by-rod distribution of differences between calculational and experimental (C-E) values has been found to be 1.1% for HELIOS and for CASMO-4, and 1.3% for MCNP4B. For C8, the rms values of the (C-E) distributions are 1.0, 1.3, and 1.4% as obtained with HELIOS, CASMO-4, and MCNP4B, respectively. The effects of using different data libraries (ENDF/B-V, ENDF/B-VI, and JEF-2.2) with MCNP4B were also studied and have been found to be small.

Measurement and interpretation of delayed photoneutron effects in multizone criticals with partial D2O moderation

The effective fraction of delayed photoneutrons (ph) has been theoretically defined and experimentally determined in various different configurations of the LWR-PROTEUS critical facility. The peculiarity lies in the fact that the reactor has D2O in only one of the four fuelled zones, thus D(,n)H reactions take place mainly in this region. The work is divided into three parts. The first part is devoted to the description of the LWR-PROTEUS facility and to the measurements of ph. These experimental values are derived from standard inverse-kinetics analysis of neutron flux decay experiments for each of seven different configurations, with nine additional groups of neutron precursors to account for photoneutron effects. In the second part, the coupled neutron and gamma Boltzmann equations are reduced to exact point kinetics equations using the photon infinite-velocity approximation, and then to the point reactor model. Photoneutron-specific kinetics parameters, i.e. the photoneutron reactivity ph, the effective photoneutron delayed group fractions jph and the photoneutron effectiveness are rigorously defined and interpreted. In the third part, ph, jph and ph values have been estimated for the seven different configurations using only four MCNP4C photonic calculations with unitary photon sources in each of the four fuelled reactor regions. Comparisons have been made with the experiments, and the agreement obtained within 2 between the predicted and measured ph values is considered remarkably good in view of the simplicity of the models used, the approximations for the adjoint weighting and the complexity of the problem at hand

Validation of axial pin power distributions in a 10×10 BWR assembly across an enrichment boundary under full-density water moderation conditions

Critical experiments involving fuel pin gamma-scans have been performed to characterise total fission reaction rate axial distributions in a Westinghouse SVEA-96+ assembly under full-density water moderation [these investigations are part of the co-operation between PSI and Unterausschuss Kernenergie (UAK) or association of the Swiss nuclear operators, viz. Kernkraftwerk Leibstadt AG, Kernkraftwerk Gosgen-Daniken AG, BKW FMB Energie AG and Nordostschweizerische Kraftwerke AG]. The measurements have covered an 80-cm long region of the SVEA-96+ assembly involving two important axial heterogeneities: (a) the enrichment boundary separating the lower axial region (average 235U enrichment of 4.02 wt.%, 14 burnable absorber fuel pins) from the upper axial region (average 235U enrichment of 3.70 wt.%, 16 burnable absorber fuel pins), and (b) the presence of spacers at about 12 cm below this enrichment boundary. The experimental results have been compared to HELIOS/PRESTO-2 calculated values at three different levels, by means of radial point-wise, axial point-wise and axial node-wise comparisons. The results show good agreement (typically within 2%) between HELIOS/PRESTO-2 nodal pin powers and corresponding experimental values, with the largest discrepancy of 6-9% being highly localised in a pin subject to strong flux gradients and spectral interactions. Further improvements would require explicit three-dimensional lattice calculations and the development,of three-dimensional pin power reconstruction methods

Impact of New Gadolinium Cross Sections on Reaction Rate Distributions in 10 × 10 BWR Assemblies

Abstract Radial distributions of the total fission rate and the 238U-capture-to-total-fission (C8/Ftot) ratio were measured in SVEA-96+ and SVEA-96 Optima2 assemblies during the LWR-PROTEUS program. Fission rates predicted using MCNPX with JEFF-3.1 cross sections underestimated the measured values in the gadolinium-poisoned pins of the SVEA-96 Optima2 assembly; similarly, C8/Ftot ratios were overestimated in some gadolinium-poisoned pins of the SVEA-96+ assembly. A considerable effort was invested at the Paul Scherrer Institut to explain the discrepancies in gadolinium pins, without success. Recently, gadolinium cross sections were measured at the Rensselaer Polytechnic Institute by Leinweber et al. and differed significantly from current library values. ENDF/B-VII.0 gadolinium cross sections have currently been modified to include the new measurements, and these data have been processed with NJOY to yield files usable by MCNPX. Fission rates in the gadolinium-poisoned fuel pins of the SVEA-96 Optima2 pins were increased by 1.4 to 2.0% using the newly produced cross sections, yielding to a better agreement with the experimental values. Predicted C8/Ftot ratios were decreased on average by 1.7% in both clustered and unclustered groups of gadolinium-poisoned fuel pins of the SVEA-96+ assembly correcting the overpredictions previously reported in the clustered gadolinium pins. Earlier reported discrepancies observed in PROTEUS integral experiments, between measured and calculated reaction rates in the gadolinium-poisoned pins, might thus be due to inaccurate gadolinium cross sections. The PROTEUS results support the new thermal and epithermal gadolinium data measured by Leinweber et al.

Within-pin reaction rate distributions: CASMO-4 and HELIOS compared against tomographic measurements at the PROTEUS reactor

Abstract In the framework of the LWR-PROTEUS project—an extended validation program for advanced light water reactor core analysis tools conducted at the Paul Scherrer Institute—the radial, internal variations of the total fission rate (Ftot) and the capture rate in 238U (C8) have been calculated for zero-burnup pins of a Westinghouse SVEA-96+ boiling water reactor fuel assembly using two codes, namely, CASMO-4 and HELIOS. While Ftot distributions predicted by CASMO-4 and HELIOS are in good agreement, C8 distributions show significant inconsistencies (20 to 30%). The calculations are compared with experimental results obtained using single photon emission computerized tomography for several SVEA-96+ pins irradiated in the zero-power reactor PROTEUS. The comparisons confirm the predicted shape of the Ftot distributions within UO2 pins and clearly indicate that HELIOS within-pin predictions for C8 are more reliable than CASMO-4 results. This is important for the derivation of gamma-ray self-absorption corrections when pin-integrated reaction rates are to be determined using the gamma-scanning technique. Thus, the use of CASMO-4-type within-pin distributions would lead to 3 to 4% discrepancies in the absolute, self-absorption-corrected pin-integrated values deduced for C8 and hence for C8/Ftot. For relative C8 distributions, the discrepancy would be much smaller, namely, up to ~1% if pins containing a burnable absorber are involved.

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

Abstract 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-4.

Investigations of 238U Captures to Total Fissions in a Westinghouse SVEA-96+ Assembly

Abstract The reaction-rate ratio C8/Ftot, neutron captures in 238U to total fissions, has been measured in 80 out of 96 fuel rods of a Westinghouse SVEA-96+ boiling water reactor fuel assembly. High-resolution gamma spectroscopy was performed on individual fuel rods, withdrawn from the SVEA-96+ assembly after irradiation at low power in the center of the LWR-PROTEUS reactor core. Absolute experimental errors of 1.7% and relative errors of 0.6% (for rod-to-rod ratios) were achieved. The experimental results were used as a database for validation of four different calculational tools: CASMO-4 and HELIOS as commercial assembly codes, the Paul Scherrer Institute in-house code BOXER, and the Monte Carlo transport code MCNPX. In general, on the level of a few percent, there is good agreement between experiment and calculations, the use of a recently proposed 239Np gamma-ray emission probability improving even further the agreement. However, the highly heterogeneous design of the SVEA-96+ assembly (both in terms of material compositions and neutron moderation conditions) causes some problems. Clear deviations from assembly mean values are found among the burnable absorber fuel rods that are grouped in clusters (direct neighbors), a unique feature of this assembly design. For these rods the codes overpredict C8/Ftot by several percent, including MCNPX. Additional trends, not present in the results from the Monte Carlo calculation which generally shows the best overall agreement with experiment, are identified for the deterministic codes.

Development and application of a decomposition methodology for interpretation of reactivity effect discrepancies

Abstract With reactivity being the most important integral reactor physics quantity—and simultaneously the one that can be measured with the highest accuracy—there is a great interest in understanding how possible space- and energy-dependent data and/or modeling discrepancies may propagate into a calculated reactivity change, and with which magnitude this occurs. In the context of pin removal reactivity effects in a light water reactor assembly, for example, it is illustrative to carry out, for any arbitrary localized material composition perturbation, a decomposition of the total effect into individual space- and energy-dependent contributions of the different unit cells in the assembly. If this decomposition is normalized to +100% in the case of a positive reactivity effect and to – 100% in the case of a negative reactivity effect, an importance map is established that indicates the relative contribution (in percent) of each individual contributing cell to the total reactivity effect caused by the localized material composition change. Such an importance map can be interpreted as a sensitivity matrix that quantifies the final discrepancy in a calculated reactivity effect, with respect to its reference value, as a weighted sum of the complete collection of cell-wise data and/or modeling discrepancies. The current paper outlines the basic theory and gives certain practical applications of the proposed decomposition methodology. Thus, it is found that the developed methodology offers in-depth, quantitative explanations for calculational discrepancies observed in the analysis of fuel pin removal experiments conducted in the framework of the LWR-PROTEUS program at the Paul Scherrer Institute.

Methods/Data Testing for Advanced LWR Fuel Designs and Operational Modes in the PROTEUS Facility

In the framework of the LWR-PROTEUS programme, integral experiments are being performed for the validation of reactor physics methods/data applied to the analysis of modem light water reactor (LWR) fuel assembly designs. One of the important aspects pertaining to this programme is the investigation of the sensitivity of integral quantities, such as reaction rate spatial distributions, local reaction rate ratios and reactivity effects, to different nuclear data sets, e.g. ENDF/B-V, ENDF/B-VI and JEF-2.2. Data sensitivity studies are currently reported, largely based on continuous-energy Monte Carlo calculations at PROTEUS-core, BWR fuel assembly and pin-cell levels. Data effects could thus be investigated systematically, independent of the inherent limitations of deterministic codes such as geometrical simplification and resonance shielding approximations. The results show a significant sensitivity of system reactivity to the data set employed, calculated values being larger by several hundreds of pcm with JEF-2.2 than with ENDF/B-V or B-VI. As regards detailed reaction rate distributions in BWR assemblies (investigated in LWR-PROTEUS Phase I), there appears to be only a weak dependence on the data library used. In the context of the investigations planned with highly burnt fuel rod segments in the Phase II experiments, certain studies made at pincell level have indicated that the effects of basic data differences are likely to be much lower than uncertainties in fuel composition predictions.