Takamasa Mori

Validation of a Continuous-Energy Monte Carlo Burn-up Code MVP-BURN and Its Application to Analysis of Post Irradiation Experiment

In order to confirm the reliability of a continuous-energy Monte Carlo burn-up calculation code MVP- BURN, it was applied to the burn-up benchmark problems for a high conversion LWR lattice and a BWR lattice with burnable poison rods. The results of MVP-BURN have shown good agreements with those of a deterministic code SRAC95 for burn-up changes of infinite neutron multiplication factor, conversion ratio, power distribution, and number densities of major fuel nuclides. Serious propagation of statistical errors along burn-up was not observed even in a highly heterogeneous lattice. MVP-BURN was applied to the analysis of a post irradiation experiment for a sample fuel irradiated up to 34.1GWd/t, together with SRAC95 and SWAT. It was confirmed that the effect of statistical errors of MVP-BURN on a burned fuel composition was sufficiently small, and it could give a reference solution for other codes. In the analysis, the results of the three codes with JENDL-3.2 agreed with measured values within an error of 10% for most nuclides. However, large underestimation by about 20% was observed for 238Pu, 242mAm and 244Cm. It is probable that these discrepancies are a common problem for most current nuclear data files.

Error estimations and their biases in Monte Carlo eigenvalue calculations

AbstractBiases in the estimators of the variance and intercycle covariances in Monte Carlo eigenvalue calculations are analyzed. The relations among the “real” and “apparent” values of variances and intercycle covariances are derived, where real refers to a true value that is calculated from independently repeated Monte Carlo runs and apparent refers to the expected value of estimates from a single Monte Carlo run. Next, iterative methods based on the foregoing relations are proposed to estimate the standard deviation of the eigenvalue. The methods work well for the cases in which the ratios of the real to apparent values of variances are between 1.4 and 3.1. Even in the case where the foregoing ratio is >5, >70% of the standard deviation estimates fall within 40% from the true value.

Continuous energy Monte Carlo calculations of randomly distributed spherical fuels in high-temperature gas-cooled reactors based on a statistical geometry model

The method to treat randomly distributed spherical fuels in continuous energy Monte Carlo calculations has been established. In this method, the location of a spherical fuel is sampled probabilistically along the particle flight path from the spatial probability distribution of spherical fuels, called the nearest neighbor distribution. The necessary probability distribution was evaluated by a newly developed Monte Carlo hard sphere packing simulation code, which employs a random vector synthesis method to reduce overlaps of spherical fuels. The obtained probability distribution was validated by comparing a cross-section photograph of a real fuel compact and an X-ray diffraction experimental result. This method was installed in a Monte Carlo particle transport code and validated by an inventory check of spherical fuels and criticality calculations of ordered packing models. Also, an analysis of a critical assembly experiment was performed with the new code. As a result, it was confirmed that the method was applicable to practical reactor analysis. The method established is quite unique in the respect of probabilistically modeling the geometry of a great number of spherical fuels distributed randomly without any loss of the advantage of the continuous energy method.

Impact of perturbed fission source on the effective multiplication factor in Monte Carlo perturbation calculations

A new method to estimate a change in the effective multiplication factor due to the perturbed fission source distribution has been proposed for Monte Carlo perturbation calculations with the correlated sampling and differential operator sampling techniques. The method has been implemented into the MVP code for verification. Simple benchmark problems have been set up for fast and thermal systems and the applicability of the method has been verified with the problems. Consequently, it has been confirmed that the method is very effective to estimate the change. It has been also shown that there are some cases where the perturbed source effect is significant and the change in reactivity cannot be estimated accurately without taking the effect into account. Even in such cases, the new method can estimate the perturbed source effect and the estimation of the change in reactivity has been remarkably improved.

Analyses and Intercomparison for Phase I Fusion Integral Experiments at the FNS Facility

Phase I integral experiments of U.S./JAERI Collaborative Program on Fusion Breeder Neutronics which are carried out at the Fusion Neutronics Source (FNS) facility at JAERI ranged from D-T neutron source characterization experiments, tritium production rate (TPR) measurements in a reference Li/sub 2/0 assembly, first wall experiments with and without coolant simulation and beryllium neutron multiplier experiments in various configurations. Both U.S. and Japan have independently analyzed these experiments using their own data base and codes. Analytical predictions obtained by both countries are compared to measured values. Results of this intercomparison is presented in this paper.

Analysis of neutronics parameters measured in Phase-II experiments of the JAERI/US collaborative program on fusion blanket neutronics. Part I: Source characteristics and reaction rate distributions

Fusion blanket neutronics parameters measured in the Phase II assembly have been analyzed at both JAERI and the US. Both parties have analyzed the experiments independently by using different nuclear data and calculational methods based on 3-D Monte Carlo and 2-D Sn codes. This part includes the results of the analysis on the source characteristics in the assembly and the reaction rate distributions in the test zone consisting of Li2O with and without a beryllium multiplier. The source characterization has been made by measuring the neutron spectrum and various reaction rates. These reactions include 58Ni(n,2n), 58Ni(n,p), 27Al(n,α), 93Nb(n,2n), 197Au(n,2n), and 197Au(n,γ). The ratios of calculated to measured values are compared among both countries and the different nuclear data used. Considerable discrepancies have been observed for the 58Ni(n,2n), 58Ni(n,p) and 93Nb(n,2n) reactions depending on which nuclear data was used, while good agreement is seen for the reactions 197Au(n,2n) and 27Al(n,α). The distributions of these reaction rates in the test zone have also been analyzed to examine the prediction accuracy of neutronics parameters in a breeder zone. Using the recently measured cross sections at the FNS resulted in a significant reduction in the discrepancies for most reaction rates.

Comparison of vectorization methods used in a Monte Carlo code

This paper examines vectorization methods used in Monte Carlo codes for particle transport calculations. Event and zone selection methods developed from conventional all-zone and one-zone algorithms have been implemented in a general-purpose vectorized code, GMVP. Moreover, a vectorization procedure to treat multiple-lattice geometry has been developed using these methods. Use of lattice geometry can reduce the computation cost for a typical pressurized water reactor fuel subassembly calculation, especially when the zone selection method is used. Sample calculations for external and fission source problems are used to compare the performances of both methods with the results of conventional scalar codes. Though the speedup resulting from vectorization depends on the problem solved, a factor of 7 to 10 is obtained for practical problems on the FACOM VP-100 computer compared with the conventional scalar code, MORSE-CG.

Analysis of neutronics parameters measured in Phase II experiments of the JAERI/US collaborative program on fusion blanket neutronics. Part II: Tritium production and in-system spectrum

Tritium Production Rate (TPR) and in-system spectrum measurements were performed on a Li2O test assembly (with and without beryllium multiplier) at the FNS facility at JAERI in addition to other neutronics parameters that include source neutron characterization and in-system reaction rates discussed in a companion paper. These activities are part of an on-going collaborative program between JAERI and the US. Calculations for the neutronic parameters were performed independently by both parties based on various 3-D Monte Carlo and 2-D discrete ordinales codes and data libraries. Local and zonal TPRs were measured by various techniques and comparisons were made to predictions. The calculated to measured values, C/E, for integrated zonal TPR from natural lithium are typically C/E = 0.97–1.07 but local TPR has larger uncertainties (C/E = 0.85–1.13). Predicted enhancement in tritium breeding upon including the Be multiplier is in the order of 7–8% while measurements indicate an approximately 10% increase in integrated TPR. The in-system spectra are well-predicted at En > 10 MeV by various codes as compared to NE213 measurements, although calculation tends to overpredict the 14.1 MeV peak. Large discrepancies, however, were observed in the energy range 1.01 MeV

Monte Carlo calculations on vector supercomputers using GMVP

Abstract A multigroup general purpose Monte Carlo code GMVP has been developed. The vectorization algorithm is based on a stack-driven zone selection method. GMVP can treat repeated rectangular and hexagonal lattices together with combinatorial geometry which is quite useful to achieve a high gain by vectorization. The performance of the code was evaluated by solving various types of problems. In addition, the continuous energy code is under development and the performance is compare with conventional codes. The code was installed on other four different supercomputers to investigate portability and computer dependence of code performance.

Fusion Integral Experiments and Analysis and the Determination of Design Safety Factors — II: Application to the Prediction Uncertainty of Tritium Production Rate from the U.S. DOE/JAERI Collaborative Program on Fusion Blanket Neutronics

Many fusion integral experiments were performed during the last decade within a well-established collaboration between the United States and Japan on fusion breeder neutronics. These experiments started in 1983 and aimed at verifying the prediction accuracy of key neutronics parameters based on the state-of-the-art neutron transport codes and basic nuclear databases. The tritium production rate (TPR) has the prime focus among other reactions. The experimental and calculational data sets of local TPR in each experiment were interpolated to give an estimate of the prediction uncertainty, u i , and the standard deviation, σ i of the line-integrated TPR, a quantity that is closely related to the total breeding ratio (TBR) in the test assembly. A novel methodology developed during the collaboration was applied to arrive at estimates to design safety factors that fusion blanket designers can use to ensure that the achievable TBR in a blanket does not fall below a minimum required value. Associated with each safety factor is a confidence level, designers may choose to have, that calculated TPR will not exceed the actual measured value. Higher confidence levels require larger safety factors. Tabular and graphical forms for these factors are given, as derived independently for TPR from Li-6 (T 6 ), Li-7 (T 7 ), and natural lithium (T n ). Furthermore, distinction was made between safety factors based on the technique applied, discrete ordinates methods, and Monte Carlo methods in the U.S. calculations, JAERIs calculations, and in both calculations considered simultaneously. The derived factors are applicable to TPR in Li 2 O breeding material ; nevertheless, the results can be used as initial guidance to assist in resolving the tritium self-sufficiency issue in other breeding media.