Modified uniform-fission-site algorithm in Monte Carlo simulation of reactor criticality problem

Based on the research of the uniform fission site algorithm, the uniform tally density algorithm and the uniform track number density algorithm are proposed and compared with the original uniform fission site algorithm in this paper for seeking high performance of global tallying in Monte Carlo criticality calculation. Because reducing the largest uncertainties to an acceptable level simply by running a large number of neutron histories is often prohibitively expensive, the researches are indispensable for the calculation to reach the goal of practical application (the so called 95/95 standard). Using the global volume-averaged cell flux tally and energy deposition tally of the pin-by-pin model of Dayawan nuclear reactor as two examples, these new algorithms show better results. Although the uniform tally density algorithm has the best performance, the uniform track number density algorithm still has the advantage of being applicable to any type of tally, which is based on the track length estimator without any modification. All the algorithms are realized in a recently developed parallel Monte Carlo particle transport code JMCT.

Algorithm researches for efficient global tallying in criticality calculation of Monte Carlo method

Danhua¹,

Deng²,

Li³

et al. 2016

Simulation calculation of weapon-grade plutonium production in pressurized water reactor

Xu¹,

Fu²,

Zhu³

et al. 2017

The nuclear nonproliferation is a common objective for the international society, of which one of the most important issues is the nonproliferation of weapon-grade nuclear material. Plutonium is a by-product when nuclear reactors are operated. If a commercial power nuclear reactor operates without counting its economic benefits, it is possible that weapon-grade plutonium (WGPu) would be produced in the nuclear reactor with using uranium as nuclear fuel. In the paper, we quantitatively study the plutonium isotopic composition and yield of the WGPu produced in a pressurized water reactor (PWR), and thereby investigate the proliferation risk of commercial nuclear reactors. The properties of plutonium produced in the PWR are calculated by MCORGS, which is developed by us to link MCNP and ORIGENS for calculating the transport-burnup. For evaluating the changing behavior of plutonium isotopic ratio dependent on the cooling time after being discharged from a PWR, we add the model of calculating the depletion and decay properties of nuclear fuel into the MCORGS code system. In order to calculate the yield of WGPu produced in the PWR, we carry out the neutron and burnup calculations by using five reactor models. The simulation models and operation history are based on the configuration and parameters of Japanese Takahama-3 unit. According to the positions and proportions of UO2 fuel rods, burnable poison rods and guide tubes in Takahama-3 PWR, we build a PWR model of an infinite heterogeneous 66 pin cell lattice, carry out simulation calculation and explore the condition for WGPu existing in the two kinds of fuel rods. When the burnup of a UO2 fuel rod is no more than 4.7 MWd/kgU, it contains WGPu. When the burnup of a burnable poison rod is no more than 2.7 MWd/kgU, it contains WGPu. Therefore, the issue of WGPu production in PWR is transformed into the research of the spatial distribution of PWR burnup. In order to obtain the axial PWR burnup, we build an infinite fuel pin cell model in which the PWR is divided into 20 equal zones in the axial direction, and calculate PWR axial burnup distribution when it is operated at 9 typical powers of Takahama-3 PWR. It is found that the burnup value of the two ends of 1/20 section is worth 1/3 of the two middle ones. Based on the principle of neutron leakage in a PWR and the simulation results of a fuel assembly, we build a special PWR mode, in which the PWR is divided into 10 zones in radial direction, and obtain the radial distribution of PWR burnup after the first, the second and the third fuel cycle. Based on the WGPu existing condition and the spatial distribution of a PWR burnup, in this paper we present the exact position of WGPu contained in PWR core and the yield of WGPu in UO2 fuel rods. The calculation results indicate that the spent nuclear fuel with low burnup brings huge proliferation risk, of which the supervision should be strengthened.

Monte Carlo global variance reduction method combining source bias and weight window

Zhang,

Liu,

Wei

et al. 2024

The global tally problem has a wide range of application scenarios in major research fields such as Monte Carlo simulations of pin-by-pin reactor models and time-dependent particle transport problems in multiphysics coupling calculations. Due to the uneven power distribution of the simulated system, the statistical errors of all tallies are unevenly distributed, which leads to some low global efficiency. For this kind of problem with global characteristics, it is necessary to develop global variance reduction techniques to accurately obtain the distribution of target tallies in the entire space. A large number of global variance reduction algorithms have been studied based on the consideration of flattening global tally error distribution, so as to improve global efficiency. This work focuses on the combination of two efficient global variance reduction algorithms, namely, the uniform tally density algorithm and the weight window algorithm which belongs to source bias and transport process bias, respectively. In total, a method is proposed to adjust the weight window parameters using the bias factor of the uniform tally density algorithm. Then, the weight window method will be used to reduce the weight fluctuation caused by the uniform tally density method. 12 By this way, an organic combination of these two algorithms can be realized. A series of comparative tests are carried out based on the Hoogenboom-Martin pressurized water reactor benchmark, and it is verified that the hybrid global variance reduction algorithm proposed in this paper is better than the single weight window algorithm or the uniform tally density algorithm. In terms of reducing the maximum error, the global efficiency of the hybrid algorithm is 2.6 times and 3 times that of the weight window algorithm and the uniform tally density algorithm, respectively. In addition, through comparative analysis of computational asymmetry degree and computational efficiency, it is verified that the uniform tally density algorithm has better performance than the classical uniform fission site algorithm, and the performance advantages of the uniform tally density algorithm are quantitatively evaluated based on some new indicators. The results show that the hybrid global variance reduction algorithm proposed in this paper can solve the global tally problem efficiently, further promoting research in related fields.