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A series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio). Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keff values overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%. Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values. [ABSTRACT FROM AUTHOR]

AbstractA series of simulated experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency to simulate a light water reactor core with a tight lattice cell containing highly enriched mixed-oxide fuel with a fissile plutonium (Pu) ratio >15%. The prediction accuracy of the neutron computation codes and nuclear data libraries in a wide range of neutron spectra was evaluated by constructing three experimental configurations of the FCA-XXII-1 assembly with different void fractions (45%, 65%, and 95%) of the moderator material (foamed polystyrene). The hydrogen-to–nuclear fuel atomic ratio was systematically varied from 0.1 to 0.8. In a previous paper, we reported the criticality and reactivity worths measured in these experiments.This technical note provides the experimental results for the central reaction rate ratios and fission distributions as follows. The fission rate ratios of uranium (U) (238U) and 239Pu relative to 235U were measured at the core centers using three calibrated fission chambers, and the 238U capture reaction rate ratio relative to 235U fission was measured using depleted U foils. The reaction rate distributions were also obtained by traversing four micro fission chambers of highly enriched U, natural U, Pu, and neptunium through each core region in the radial and axial directions. The experimental analyses were performed using detailed models of the Monte Carlo code MVP3 with the Japanese evaluated nuclear data library of JENDL-4.0. Most calculation results agreed well with the experiments, whereas those for the fission rate ratio of 239Pu to 235U were underestimated by up to 6% with the softening neutron spectrum. [ABSTRACT FROM AUTHOR]