Yoshitaka Naito

The Sandwich Method for Determining Source Convergence in Monte Carlo Calculation

In slow source convergence problems, it is often difficult to ascertain whether the source iteration has converged or not. In order to solve this problem, a new “sandwich method” has been proposed. The essence of this method is that a finally converged eigenvalue keff is approached starting from two kinds of initial source guesses which give higher and lower neutron multiplication factors. It is especially important for evaluating nuclear criticality safety to know how to choose a biasing source to obtain an upper limit for keff . In this paper, (1) an example is shown to explain the difficulties in ascertaining the source convergence, (2) a method is proposed to obtain the upper and lower limit curves for keff by biasing the initial source distribution, (3) the sandwich method is applied to four benchmark problems proposed by the source convergence group of the OECD/NEA Working Party on Nuclear Criticality Safety. Our calculation results show that the sandwich method is an effective means to confirm source convergence in such slow convergence problems. Appendix is prepared to support the method theoretically.

Review of studies on criticality safety evaluation and criticality experiment methods

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS.

Study on the Criticality Safety Evaluation Method for Burnup Credit in JAERI

In relation to burnup credit, three tasks have been carried out at the Japan Atomic Energy Research Institute (JAERI) for establishing the evaluation method of criticality safety for a spent-fuel system, such as storage ponds and transports casks. The first task is to prepare a brenchmark database of criticality experiments and nuclide compositions of spent fuels. The database of nuclide composition is formed by data measured at JAERI and data collected from the literature. For the database of criticality experiments, the effective multiplication factor of a spent-fuel assembly has been measured at JAERI. The next task is to develop computer codes. The burnup and criticality codes have been developed and validated by analyzing a large number of benchmarks stored in the aferomentioned database. The last task needed to establish the methodology in order to confirm the subcriticality of a spent-fuel system applying burnup credit described. A reference fuel assembly is introduced so that the criticality of a system can be evaluated by using it, instead of modeling all fuel assemblies explicitly. To determine the nuclide composition of a spent fuel, a simple method is studied utilizing a large number of nuclide composition data stored in the database. Further, the effects of the axial burnup profile and calculation errors are discussed, and the remaining tasks are identified

Development of One Group Constant Library for the Computer Code ORIGEN-2 Using JENDL-3.2

Standard one group constant libraries for the ORIGEN-2 code are prepared using JENDL-3.2 for calculating the amount of nuclides generated in spent fuel of BWR, PWR and FBR, and the adaptability of the libraries is examined for the safety evaluation of the nuclear fuel cycle. This task is performed as a part of the activities of the nuclide generation evaluation working group of Japanese Nuclear Data Committee.

Monte Carlo calculations on high speed machines

Abstract 1. 1. Monte Carlo Calculations by a Modified Vector Processor. At Japan Atomic Energy Research Institute, four Monte Carlo codes KENO-IV, MORSE-DD, and VIM and MCNP were vectorized to examine the adaptability of vector processors for these codes. The performances and vectorization rates of the vectorized versions were not good except KENO-IV on SX-2 vector processor, on which the vectorized version attained three times faster speed compared to its scalar version. According to the experience with the vectorization, some additional features specialized for Monte Carlo calculations will improve the vectorization rates of the four codes. Functions of the features and anticipated effects are presented. 2. 2. Speedup of Monte Carlo Criticality Calculation by a Parallel Processor. The authors implemented the Monte Carlo code KENO-IV on a parallel processor system Topology 1000 with three processing units under control of SUN workstation. The Hansen-Roach 16-group cross section set was used for the calculation. A computation for a bare sphere of highly enriched uranium metal showed that the system attained 2.7 times speedup compared to that of one processing unit.

A computer program for solving the three-dimensional multigroup diffusion equation by finite element method with 20-node isoparametric element (FEM3DJAR)

Abstract The three-dimensional multigroup neutron diffusion equations are solved by the finite element method with 20 node isoparametric element. By this method, neutron diffusion problems in complex geometry have been able to be solved accurately. Benchmark problems are calculated and the results show that the scheme proposed here is very accurate even with a large mesh size.

Criticality Detection Method Based on Fission Product Gamma Radioactivity Measurement

Abstract In this paper, a method is proposed to evaluate the extent of subcriticality of an accident-damaged nuclear reactor. With this method the activity ratio of two fission product (FP) rare gas nuclides and is measured. From the measured value, the value of the nuclides in the fuel region is estimated by correcting for the time lag incurred when the gases diffuse from the fuel region to the measuring point. A simple expression for an effective multiplication factor has been derived that uses the corrected -to-activity ratio and the -to-fission yield ratios of and but that does not require information on the amount or distribution of the fuel material, making the proposed method very simple. This method has the advantage that FP rare gases can easily leak from the reactor core through many openings and gaps, reaching germanium counters without reacting with other materials.

Development of a Method to Calculate keff Satisfying a Convergence Criterion of Source Iteration in Monte Carlo Calculation

A method to judge whether the source iteration in Monte Carlo calculations is converged or not by applying a “sandwich method” is described in this report. A procedure is also introduced to determine the number of skip generations in the source iteration satisfying the convergence criterion. Moreover, the effectiveness of this method is demonstrated by calculating using the fresh fuel vault model in the OECD source convergence benchmark problems.

Overview of Recent Research Activities of Monte Carlo Simulation in Japan

This paper describes recent progresses of the Monte Carlo simulation technology in nuclear energy field in Japan. Radiation shielding solution method using the Monte Carlo had been validated as a reliable tool through the discussion of “Radiation Shielding Safety Demonstration Analysis Group” of Japan Atomic Energy Research Institute. Since 1996, “Monte Carlo Simulation Working Group” has been accumulating use experiences of Monte Carlo codes in the wide range of nuclear energy field. This working group is planing to publish “Guideline of Monte Carlo Simulations” during FY-99. This “Guideline” is expected to be a first Japanese practical textbook of Monte Carlo calculation. In 1998, the first full-scale topical conference on Monte Carlo simulation was held in Tokyo. “Research Committee on Particle Simulation with the Monte Carlo Method” was established in Atomic Energy Society of Japan in 1998. This committee is composed of more than seventy members from many fields of nuclear energy research in Japan. This committee is expected to be a core that will drive the research and development activity of Monte Carlo calculation in Japan.

Fault-Tree Analysis of Criticality in a Pulsed Column of a Typical Reprocessing Facility

Scenario identification, preparation of reliability data, and fault-tree construction were conducted for a criticality in a pulsed column of a typical model of a reprocessing facility to find a wea...