Takuya Umano

Development of a “best representativity” method for experimental data analysis and an application to the critical experiments at the Toshiba NCA facility

For nuclear critical experiments, it is essential to certify similarities of the experiment with the objective of the actual reactor conditions or actual reactor equipment. To judge the applicability of the experimental data, the concept of a “representativity factor” has recently been adopted in the critical experiment field, particularly for fast breeder reactors and future reactor studies. In this study, we extended this concept to the design of a light water reactor system. We developed a new numerical evaluation method and a calculation system. The method is based on a linear combination of the sensitivity coefficient vector of an experiment in which the representativity factor to the target system is maximized to utilize experimental data effectively. Simultaneously, using the measurement data of critical experiments, the method enables us to evaluate calculation errors caused by errors or uncertainties of physical parameters. The derivation of the new calculation method is explained first. We then qualify it with a sample calculation, presenting numerical results for three kinds of critical experiments conducted at the Toshiba Nuclear Critical Assembly facility. Finally, the results are compared with those of an extended bias factor method to clarify the performance of the new method.

Neutronics Analysis of Full MOX BWR Core Simulation Experiments FUBILA

A part of the experimental program FUBILA was dedicated to the study of core physics characteristics of full MOX BWR cores by testing five experimental cores: a core inserted with a B4C control blade, a core loaded with UO2 fuel rods, the core loaded with Gd2O3-UO2 fuel rods, a core loaded with 10 × 10 MOX assemblies and a core loaded with time-elapsed MOX fuel. The present article describes analysis results of the experimental data with deterministic analysis codes and a continuous energy Monte Carlo code coupled with major nuclear data libraries. The calculated critical k effs with the Monte Carlo calculations range from 0.999 to 1.007. Those of the transport calculations with sixteen energy groups are close to those of the Monte Carlo calculations while those of diffusion calculations with the same sixteen energy groups are systematically smaller by −0.3 to −0.5% Δk than those of the Monte Carlo calculations. The RMSs of differences between the calculated and measured core radial fission rates are 2 to 3%, 1 to 2%, and 1 to 2% for the diffusion, transport and Monte Carlo calculations, respectively. For the analysis of the Gd2O3-UO2 fuel rod loaded core, the C/Es of the radial fission rates were improved by adopting a detailed lattice calculation model and a newly measured thermal and resonance cross-sections of Gadolinium.

Burnup Sensitivity Analysis in a Fast Breeder Reactor—Part II: Prediction Accuracy of Burnup Characteristics

Evaluation quantitative de la precision des caracteristiques de combustion en utilisant des coefficients de sensibilite dans un grand surregenerateur refroidi au sodium et la matrice des covariances des donnees nucleaires

Analysis of Reactivity Measurements in Core Physics Experiments on Full-MOX BWR

A core physics experimental program FUBILA has been performed to study core physics characteristics of full-MOX BWR cores consisting of high Pu-enriched MOX assemblies for high burnups. The program includes the measurement of reactivity worth, which is essential in operating BWR cores. The reactivity worth is due to the reactivity caused by (1) changes in the in-channel void fraction of assemblies, (2) the insertion of a B4C control blade, (3) Gd2O3-UO2 rods and UO2 rods in assemblies, and (4) the mixing of boron in a moderator related to a stand-by liquid control system. The reactivity worth was measured by the modified neutron source multiplication method using the reactivity worth of a pilot rod as a reference worth. The measured reactivity worth was determined by processing count rates of neutron detectors taking into account the detector efficiency and effective neutron source intensity analyzed by three-dimensional transport calculations. Comparing the measured reactivity worth with the results obtained by deterministic calculation methods and a Monte Carlo calculation method, the accuracy of the calculations was evaluated. The calculated results generally well reproduce the measurements, except for the boron reactivity worth in the moderator, for which the calculations overestimate the reactivity worth.

Analysis of MISTRAL Experiments with JENDL-3.2

Nuclear Power Engineering Corporation has been analyzing measurement data of MISTRAL, a LWR MOX core physics experimental program, with a deterministic code system of the diffusion and the transport theories, SRAC, and a Monte Carlo transport code, MVP, based on JENDL-3.2. The program consists of one reference U02 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. The measurement parameters cover critical masses, boron concentrations, radial and axial core power distribution, spectrum indexes, conversion factors, effective delayed neutron fractions, iso-thermal temperature coefficients, boron efficiency, single absorber worth, water hole worth, absorber cluster worth and void worth. Critical keff calculations show a slight overestimate for MOX cores and a small systematic trend with aging of MOX fuel. Power distribution calculations show good agreement with the measurements in both homogeneous and mock-up cores. Measured spectrum indexes of fissile isotopes are well reproduced by the calculation. For other measurements, calculated results well reproduce the measurements within one or two times of the experimental error. The analysis verifies the validity of JENDL-3.2 and the analysis methods for the high moderation LWR MOX cores.

Intra-pellet neutron flux distribution measurements in LWR critical lattices

We have developed inexpensive and easy-handling measurement methods on intra-pellet neutron flux. A foil activation method with metallic foils, which were fabricated by punching out technique and etching technique to reduce fabrication error and positioning error, was used for the intra-pellet neutron flux distribution measurement. The developed method was applied to measure intra-pellet neutron flux distributions in a reduced–moderation light water reactor (LWR) lattices, and uncertainty of the distributions was estimated to be 1% to 2%. Measured values were analyzed with a continuous energy Monte Carlo code. Comparison of measurements and analyses revealed that the developed method is useful for the validation of an advanced fuel design method considering neutron behavior in fuel pellets.

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the “finite neutron multiplication factor”, k*, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and k* on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.