Munenari Yamamoto

Pin Power Reconstruction Methods of the Few-Group BWR Core Simulator NEREUS

An improved pin power reconstruction method has been incorporated in the few-group nodal BWR core simulator NEREUS, which is based on the analytic polynomial nodal method. With the analytic polynomial nodal method, accurate node surface fluxes are available, which are used later to reconstruct pin powers. The intranodal homogeneous thermal flux is corrected using a semi-empirical proportional relation between surface transition components of the homogeneous and heterogeneous fluxes. This correction method is effective for BWR calculations, especially for controlled assemblies or mixed loading of off-set assemblies. A unified model accounting for effects of spectral histories, caused by spectral interactions between fuel assemblies and the control blade insertion, was also developed for intranodal burnup correction. The change in pin powers due to the control blade history can be predicted well, without laborious assembly depletion calculations with control blade insertion. R-factors used in critical power...

Advanced Nodal Methods of the Few-Group BWR Core Simulator NEREUS

A few-group nodal BWR core simulator NEREUS, which is based on the analytic polynomial nodal method, has been developed. The analytic polynomial nodal method is applicable to multigroups and has good accuracy for analysis of cores having large spectral mismatch between fuel assemblies, since the intranodal thermal flux can be correctly represented. In NEREUS, the following advanced methods are adopted to overcome the shortcomings of the conventional analytic polynomial nodal method. The flux solution iteration matrix is cast into the form of finite difference, by using the flux discontinuity factors which correct the finite difference error as well as the homogenization error. The intranodal burnup and spectral history distributions are considered in the method, and source moments are obtained by orthogonal expansions. Burnup calculations are made by looking-up exposure dependent tables for macroscopic cross sections, which are prepared by single assembly spectrum and depletion calculations. A unified mod...

An Improved One-and-a-Half Group BWR Core Simulator for a New-Generation Core Management System

An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good.

Effect of Subchannel Void Fraction Distribution on Lattice Physics Parameters for Boiling Water Reactor Fuel Bundles

In boiling water reactor (BWR) cores, the radial void distribution in fuel bundles is thought to deviate from uniform distribution. The effect of heterogeneity in the subchannel void fraction distribution, caused by the presence of Gd-poisoned and cold surfaces as well as control blades on BWR lattice physics parameters, has been evaluated. The cross-sectional void distributions in each axial plane of a fuel bundle are calculated using the subchannel thermal hydraulics code COBRAG. The rod power distributions to be fed to COBRAG are calculated using either the Monte Carlo code MCNP4C or the fuel lattice code TGBLA. Iterative methods consisting of COBRAG and MCNP4C (or TGBLA) are established. A set of test cases was generated for a typical BWR 8 × 8 fuel bundle. The results of the coupled MCNP4C and COBRAG method reveal that both Gd rods and control blade increase in worth due to the void heterogeneity, showing a maximum decrease in K∞ of ∼0.7%Δk/K∞ and ∞1.5%Δk/K∞, respectively, at moderator density conditions equivalent to ∼40% void fraction. With the capability of the coupled TGBLA and COBRAG method to deplete fuel bundles, the acceleration of Gd depletion was evaluated. The impacts on the burnup characteristics of K∞ reached ±0.6%ΔK∞ at maximum.

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO2 assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of 235U, well reproduce those of 239Pu and 240Pu, yet underestimate those of 236U, 237Nd, 238Pu, 241Pu, and 242Pu. The inventories of minor actinides are underestimated by the present analysis except for 241Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO2 assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO2 assemblies were well reproduced by the burnup calculations of MVP-BURN.

Validation of the HELIOS.HX Code for High Conversion Light Water Reactor Lattice Analysis

The HELIOS.HX code has been developed for the design study of high conversion light water reactor (HCLWR) lattices. Analysis of the PROTEUS critical experiments at the Swiss Federal Institute for Reactor Research has been carried out as the first step toward validation of the HELIOS.HX code, and indications are that the accuracy may be at a higher or comparable level compared to that of WIMS-D, EPRI-CPM, and SRAC. In addition, comparisons with Monte Carlo calculations have also been performed for an HCLWR fuel assembly benchmark problem, showing that the accuracy is passable in the prediction of important nuclear characteristics, thereby indicating the validity of various approximations involved in the physics methods. These numerical results indicate that the code has basic potential as a tool for HCLWR lattice analysis, but covers only limited HCLWR lattice conditions.

Improved Intermediate Resonance Approximation in Heterogeneous System

An improvement of the IR (Intermediate Resonance) approximation is proposed. The effective resonance integrals of the infinite cylindrical fuel lattices are calculated by the analytical method, efficiently and with reasonably good accuracy. The fuel escape probability Pf(E) is based on the two-term rational approximation with an accurate Bell factor [Abar]. The IR parameters λ, κ and μ are determined taking into account the interference between the resonance and potential scatterings, and the effect of the fuel temperature. The effective resonance integrals of the main resonance levels of 238U were calculated using the improved IR approximations, and the accuracy of the method was ascertained by comparing the results with those obtained by the RICM code, which solves the neutron slowing-down equations numerically. Simple approximate expressions for J(ξ β) and J(ξ, σ, β) functions, which are useful in the IR approximations, are also given together with a table of J(ξ, β) and the errors of the simple approx...

Analysis of Rod-by-Rod FP Inventory Distributions in BWR 8 × 8 UO2 Assemblies Using Lattice Physics Method

Fuel rod gamma-ray spectrometry was performed for one-, two-, three-, and five-cycle irradiated BWR 8 × 8-4 fuel assembles, and relative rod-by-rod FP inventory distributions of 137Cs, 134Cs, 106Ru, and 95Zr wereobtained for the upper and lower axial height nodes of the assemblies. The measured data was corrected for the difference in gamma-ray transmission between UO2 and Gd2O3-UO2 rods. These distributions were compared with those calculated using the collision probability method, Pij, of the SRAC code system with aninfinite lattice model of the fuel assembly. The calculated results generally well reproduce the measured distributions, and the accuracy of the analysis method of the present study was evaluated to be 1 to 2% for the inventory distributions of 137Cs, 2 to 3% for 134Cs, 2 to 3% for 106Ru, and 2 to 3% for 95Zr, which represent the distributions of the burn-up, the thermal flux multiplied by the burn-up, the buildup of 239Pu, and the fission rate at the end of the fuel discharge cycle, respectively. One of the notable results of the analysis of this study is that the FP inventories for the Gd2O3-UO2 rods were underestimated in most cases.

Method for calculating thermal neutron spectrum in BWR lattice.

A new method for calculating space-dependent thermal neutron spectrum in a BWR lattice has been developed primarily aiming at its application to a lattice design code. Since the method has been constructed from combination of the cell-by-cell THERMOS calculation with leakage boundary condition and the coarse-mesh diffusion calculation for the cell leakage coupling, it is considerably economical as compared with various other transport theory approaches. Validation of the method has been made through comparisons with Monte Carlo calculations for a typical 8x8 BWR lattice. The discrepancies with the reference Monte Carlo calculations are mostly in the ranges of ±1.7 and ±3.0% for the rod-by-rod 235U fission cross section and the absorption rate distribution, respectively. The accuracy is acceptable and ensures the validity of various simplifications and assumptions involved in the present method. The present method is expected to be applicable to the BWR lattice analysis over the practical range of design p...