Motoo Aoyama

Neutronic Analysis Code for Fuel Assembly Using a Vectorized Monte Carlo Method

AbstractA fuel assembly analysis code, VMONT, in which a multigroup neutron transport calculation is combined with a burnup calculation, has been developed for comprehensive design work use. The neutron transport calculation is performed with a vectorized Monte Carlo method that can realize speeds >10 times faster than those of a scalar Monte Carlo method. The validity of the VMONT code is shown through test calculations against continuous energy Monte Carlo calculations and the PROTEUS tight lattice experiment.

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.

A New Direct Calculation Method of Response Matrices Using a Monte Carlo Calculation

A novel direct calculation method of response matrices on heterogeneous lattices by using the Monte Carlo method is proposed. These direct response matrices (DRMs) can be used in core calculations in place of the conventional homogenized lattice constants. The DRMs are formalized by four sub response matrices (sub-RMs) in order to respond to a core eigenvalue, k; thus the DRMs can be re-evaluated on each outer iteration in the core calculations. The sub-RMs can be evaluated by analyzing each neutrons trajectory from ordinary lattice calculations with the Monte Carlo code. Since these sub-RMs are calculated directly under an actual complex assembly geometry, i.e., without a homogenization process, intra-assembly heterogeneous effects can be reflected on global partial current balance calculations. With using two of the sub-RMs, which deal with neutron production probabilities for each fuel pin, and the obtained partial current balance, pin-wise neutron production distributions can also be reconstructed. T...

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.

Development of an advanced core analysis system for boiling water reactor designs

A core analysis system has been developed for the recent advanced designs of boiling water reactors. This system consists of a fuel assembly analysis code VMONT and a three-dimensional core simulator COSNEX. To cope with heterogeneous structures found in the recent high-performance fuel, VMONT employs a Monte Carlo neutron transport calculation method. COSNEX is based on a three-group nodal expansion method to treat spectral interactions among fuel assemblies. Both codes are vectorized to meet timing requirements as design tools. The analysis system is verified by the tracking of recent plant operations. Although the analyzed cores are highly heterogeneous in the multienrichment configuration, the system gives sufficient accuracy both in critical eigen values and thermal power distributions.

ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

Abstract An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor.

Improvement of Monte Carlo Lattice Burnup Calculation Performance with the Correlated Sampling Method

For next generation reactor designs, which are attempting wide variations of assembly configurations, the flexibility Monte Carlo method holds is attractive, but still costly for repetitive design study works. This paper presents an advanced correlated sampling (ACS) method which was developed to speed up Monte Carlo lattice burnup calculations. The ACS method is the combination of the correlated sampling method and a pseudo-scattering technique. All burnup steps are considered as consecutive perturbed problems using the same neutron collision history, which is pre-calculated based on a selected unperturbed problem. Since neutron weights can be adjusted on every collision point, rather than along paths between them, the perturbed calculation is very fast and the neutron collision history is light enough to be stored in memory or physical storage, which is an indispensable feature for consecutive perturbed calculations. The presented theory shows that the ACS method has good potential to work for a wide range of neutron absorption variations, the dominant perturbation in the lattice burnup. In an example calculation on a BWR lattice, the ACS calculation results of 600,000 neutrons/step agree well with the independent Monte Carlo runs of 20,000,000 neutrons/step within 0.1%dk/k in terms of kœ throughout 95 steps (~50GWd/t). Average calculation time of neutron tracking with the former method is 3.4 s/step with 600,000 neutron histories on a single processor of an Alpha21164-600 MHz, and the speed-up factor against the Monte Carlo calculation turns out to be about 100.

A Conceptual Design of a Fuel Bundle for Extended Burnup in Boiling Water Reactors

A new design concept of a boiling water reactor (BWR) fuel bundle for extended burnup was proposed to improve the capacity factor without increasing the fuel cycle cost. Some effects, which are raised from higher burnup, such as strong pellet-cladding interaction due to enhanced fuel swelling and changes in neutronic characteristics due to increased fuel enrichment, are minimized by a reduction in the maximum fuel temperature to below 1200/sup 0/C and an increase in the moderator-to-fuel ratio. To realize these concepts, a 9 X 9 lattice design with a reduced fuel rod diameter and annular pellets was proposed. The proposed fuel bundle design offers advantages in fuel cycle improvements through extension of achievable burnup and reduction of fuel inventory. The core, loaded with the proposed fuel bundles which achieve 30% higher burnup by the full power month, has a potential for natural uranium savings of about 20% per unit power and a reduction in the amount of reprocessing of about 40% per unit power, compared with the current BWR design when coupled with other improvements such as refueling pattern optimization, natural uranium axial blankets, and spectral shift with flow control.

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.