Tetsushi Hino

Three-Dimensional Direct Response Matrix Method Using a Monte Carlo Calculation

A three-dimensional direct response matrix method using the Monte Carlo calculation has been developed. The direct response matrix is formalized by four sub-response matrices in order to respond to a core eigenvalue k and thus can be recomposed at each outer iteration in core analysis. The sub-response matrices can be evaluated by ordinary single fuel assembly calculations with the Monte Carlo method in three dimensions. Since these sub-response matrices are calculated for the actual geometry of the fuel assembly, the effects of intra- and inter-assembly heterogeneities can be reflected on global partial neutron current balance calculations in core analysis. For the purpose of verification of this method, the calculations for heterogeneous systems are performed. As a result, the results obtained using this method agree well with those obtained using direct calculations with the Monte Carlo method. This means that this method accurately reflects the effects of intra- and inter-assembly heterogeneities and can be used for core analysis.

BWR Core Simulator Using Three-Dimensional Direct Response Matrix and Analysis of Cold Critical Experiments

A new core analysis method has been developed in which neutronic calculations using a three-dimensional direct response matrix (3D-DRM) method are coupled with thermal-hydraulic calculations. As it requires neither a diffusion approximation nor a homogenization process of lattice constants, a precise representation of the neutronic heterogeneity effect in an advanced core design is possible. Moreover, the pin-by-pin power distribution can be directly evaluated, which enables precise evaluations of core thermal margins. Verification of the neutronic calculation using the 3D-DRM method was examined by analyses of cold criticality experiments of commercial power plants. The standard deviations and maximum differences in predicted neutron multiplication factors were 0.07%Δk and 0.19%Δk for a BWR5 plant, and 0.11%Δk and 0.25%Δk for an ABWR plant, respectively. A coupled analysis of the 3D-DRM method and thermal-hydraulic calculations for a quarter ABWR core was done, and it was found that the thermal power and coolant-flow distributions were smoothly converged.

Analysis of the MOX critical experiments BASALA by pin-by-pin core analysis method using three-dimensional direct response matrix

An analysis of the MOX critical experiments BASALA was performed to verify the pin-by-pin core analysis method using a three-dimensional direct response matrix. The BASALA experiments simulate full MOX BWR cores, and they were carried out in the EOLE critical facility of the French Atomic Energy Commission (CEA) by the Nuclear Power Engineering Corporation (NUPEC) in collaboration with CEA. The BASALA experimental cores are very heterogeneous because their size is much smaller than that of commercial power plants. The main features of the pin-by-pin core analysis method using the three-dimensional direct response matrix are that the response matrix can reflect the intra-assembly heterogeneous effect, the diffusion approximation is not involved, and the fuel rod fission rate can be directly evaluated. The maximum difference of the critical k-effective values among all nine cores analyzed was about 0.4% Δk. The root mean square differences between the calculated and measured radial fuel rod fission rate distributions in the test assembly of all cores were within 1.8% and nearly comparable to measurement error. The calculated results of the reactivity worth agreed with the measured results within 9%. These good agreements mean that the pin-by-pin core analysis method using the three-dimensional direct response matrix accurately reflects the effects of the intra- and inter-assembly heterogeneities in heterogeneous systems like the BASALA experimental cores.

Development of Spectral History and Pin Power Correction Methods for Pin-by-Pin Core Analysis Method Using Three-Dimensional Direct Response Matrix

Spectral history and pin power correction methods have been developed for the pin-by-pin core analysis method using the three-dimensional direct response matrix (3D-DRM). The direct response matrix is formalized using four subresponse matrices in order to respond to a core eigenvalue k and thus it can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the historical effect, which is related to spectral heterogeneity. The spectral history method is used to evaluate the nodal burn-up spectrum obtained by using the outgoing neutron current instead of the nodal flux because the 3D-DRM method does not use the nodal flux. The pin power correction method corrects the fuel rod neutron production rates obtained in the pin-by-pin calculation. These two methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error and nodal neutron production rate errors can be reduced by half during burn-up. The root-mean-square differences between the relative fuel rod neutron production rate distributions and the maximum error of relative fuel rod production rate can also be reduced by half. This means that the developed methods can reflect the effects of intra- and interassembly heterogeneities during burn-up and can be used for core analysis.

Doppler Reactivity Worth Uncertainties due to Errors of Resolved Resonance Parameters

Errors of the resolved resonance parameters for the evaluated nuclear data file JENDL-3.2 were evaluated on the basis of Breit-Wigner Multi-level formula. For the Reich-Moore resonance parameters, the errors equivalent to the Breit-Wigner resonance parameters were obtained. Reactivity uncertainties of Doppler reactivity worth are estimated by the sensitivity coefficients of the infinitely diluted cross section and resonance self-shielding factor to the changes of resonance parameters of interest. The resonance self-shielding factor based on NR-approximation was analytically described. Total uncertainty of Doppler reactivity worth p for whole resonance was estimated by means of error propagation law.

Application of the Resource-Renewable Boiling Water Reactor for TRU Management and Long-Term Energy Supply

The RBWR (resource-renewable boiling water reactor) is an innovative BWR that has a capability to breed and burn trans-uranium elements (TRUs) using a multi-recycling process. The RBWR can be used as a long-term energy supply, and it reduces the negative environmental impact that TRUs cause as they are otherwise long-lived radioactive wastes. Various design concepts of the RBWR core have been proposed. The RBWR-AC is a break-even reactor and the RBWR-TB and RBWR-TB2 are TRU burners. The RBWR-TB is designed to burn TRUs from the RBWR-TB itself and to burn almost all the TRUs by repeating their recycling. The RBWR-TB is assumed to be applied for a nuclear power phase-out scenario. The RBWR-TB2 is intended to burn TRUs from LWR spent fuels. The RBWR-TB2 is assumed to be applied for reducing the amount of TRUs to be managed in storage facilities. The RBWR cores achieve their TRU multi-recycling capability under the constraint that the void reactivity coefficient must be negative by introducing the parfait core concept. This chapter reviews details of the specific design and core characteristics of the RBWR.

Analysis of Mixed Oxide Fuel Critical Experiments EPICURE and MISTRAL with Nuclear Analysis Code for BWR

Kazuya ISHII1,*, Atsushi FUSHIMI1, Tetsushi HINO1, Hiromi MARUYAMA2, Sadayuki IZUTSU2, Masaru SASAGAWA2 and Yutaka IWATA3 1Power & Industrial Systems R & D Laboratory , Hitachi, Ltd., 7-2-1 Omika-cho, Hitachi-shi, Ibaraki 319-1221, Japan 2-Global Nuclear Fuel -Japan Co., Ltd., 2-3-1 Uchikawa, Yokosuka-shi, Kanagawa 239-0836, Japan 3Hitachi Works, Hitachi, Ltd., 3-1-1 Saiwai-cho, Hitachi-shi, Ibaraki 317-8511, Japan (Received June 6, 2005 and accepted in revised form November 8, 2005)