Ganglin Yu

Asymptotic Wielandt Method and Superhistory Method for Source Convergence in Monte Carlo Criticality Calculation

Abstract In loosely coupled systems and large-scale systems, Monte Carlo criticality calculation suffers from slow fission source convergence because of the high dominance ratio (DR). In previous work, the Wielandt method and the superhistory method have been separately proposed to accelerate source convergence. However, although both methods decrease the DR, they are found not able to sufficiently accelerate fission source convergence. In this paper, the effective DR is defined and used to analyze the effectiveness of the Wielandt method and the superhistory method and to theoretically prove that they cannot reduce the computational time to converge the fission source. Accordingly, both methods are modified by adjusting the source population of inactive cycles, and their efficiency after adjustment is also compared. Moreover, the asymptotic Wielandt method (AWM) and the asymptotic superhistory method (ASM) are proposed, and the rules of deciding asymptotic parameters are also discussed. The new methods are implemented into the RMC code and validated by calculating loosely coupled problems and large-scale problems. Numerical calculation results show that AWM and ASM are practical and efficient for source convergence acceleration, which can save 75% to 90% of the computational time to reach a converged fission source.

RMC1.0: Development of Monte Carlo Code for Reactor Analysis

This paper describes a newly developed Monte Carlo code used for reactor analysis called RMC1.0, which is based on ACE format library. RMC1.0 is able to estimate criticality eigenvalue, and tally flux/spectrum with collision estimation method or tracking length method. Series of benchmarks and other examples are calculated for validation, which prove that RMC1.0 gives a good performance in both accuracy and efficiency compared with mcnp5. Despite its limitation in geometry processing, RMC1.0 has made a profitable attempt in self-development of Monte Carlo code for reactor analysis.Copyright

Research on Monte Carlo Perturbation Calculation Methods Applied in Reactor Physics

In the reactor design and analysis, there is often a need to calculate the effects caused by perturbations of temperature, components and even structure of reactors on reactivity. And in sensitivity studies, uncertainty analysis of target quantities and unclear data adjustment, perturbation calculations are also widely used. To meet the need of different types of reactors (complex, multidimensional systems), Monte Carlo perturbation methods have been developed. In this paper, several kinds of perturbation methods are investigated. Specially, differential operator sampling method and correlated tracking method are discussed in details. MCNP’s perturbation calculation capability is discussed by calculating certain problems, from which some conclusions are obtained on the capabilities of the differential operator sampling method used in the perturbation calculation model of MCNP. Also, a code using correlated tracking method has been developed to solve certain problems with cross-section changes, and the results generated by this code agree with the results generated by straightforward Monte Carlo techniques.© 2009 ASME

The Development and Validation of Nuclear Cross Section Processing Code for Reactor-RXSP

The accuracy of the nuclear cross section data is a prerequisite for the accuracy of reactor physics calculations. The RXSP(Reactor Cross Section Processing Code) which is developed by REAL (Reactor Engineering Analysis Laboratory) of Department of Engineering Physics in Tsinghua University, has changed the situation in China that nuclear cross section processing has been dependent of NJOY for a long time. The key methods such as fast Doppler broadening, thermal libraries interpolation, and OpenMP parallel acceleration, can be achieved with RXSP. This code is able to process the original data of ENDF/B (Evaluated Nuclear Data File/B) efficiently and accurately to produce the continuous energy point cross section data which is necessary for RMC. By comparing with NJOY, The microscopic and macroscopic verification shows that RXSP has the same accuracy as NJOY while RXSP has saved greatly the processing time to meet the efficient demand in the frequent reactor physics-thermal-hydraulic coupling calculations to solve the complex questions related on a large number of materials and temperature. In addition, RXSP make it available to process the resonance parameters of the R-matrix Limited format.Copyright

Neutron Generation Based Method for Monte Carlo Three-Dimensional Reactor Time-Dependent Simulation

A novel Monte Carlo time-dependent simulation method, named neutron generation based method (NGBM), is proposed for three-dimensional reactor dynamic analysis. Different from the traditional direct simulation method (DSM) based on neutron history, the new method, originating from the process of Monte Carlo criticality calculation, is based on neutron generation. In order to turn the original static calculation into a dynamic one, the time mark and time-dependent flux tally are added, the weight of neutron is adjusted while accumulating the flux estimator and the criteria for ending simulation is set. This new method is of higher computing efficiency than the direct simulation method for super-critical time-dependent situation, because it is able to take the advantage of Monte Carlo criticality calculation to keep the number of neutrons per generation approximately constant while the direct simulation method cannot stop the exponential increase of neutron population. The new method was integrated into RMC (Reactor Monte Carlo code developed by Tsinghua University). A numerical experiment was performed. The results demonstrate the feasibility and accuracy of the neutron generation based method for reactor dynamic analysis. The relative deviation of the time-dependent neutron flux tells that the accuracy of the neutron generation based method is enough for routine reactor safety analysis. And the experiment also shows the high efficiency of this method for super-critical reactor systems, since in the experiment, RMC runs nearly 7 times faster than MCNP which uses the direct simulation method.Copyright

A New Method on Flux Mapping by Function Expansion

The core flux (power) distribution is very important to safe and economical operation of nuclear reactor. It can be obtained by many methods depending on the desired accuracy and execution time. For on-line core surveillance and regulation, we need to get the real-time flux distribution. If the true local parameters such as fuel temperature, coolant temperature and material density were known, the solution of the diffusion equation with instantaneous parameters could, in principle, provide the necessary spatial details. However, in reality, it is impossible to obtain the operational “readings” of these parameters for each fuel cell. The detector results at certain locations can be applied to improve the results of the only diffusion calculations by Flux Mapping methods. Function expansion method is employed to express the approximate real distribution by the combination of several Flux Mapping method results as the expansion basis functions. The Harmonics Synthesis Method (HSM) and Least-Square method are combined to get a new Flux Mapping method in this paper. The simulation results show that the new method can be used for Flux Mapping and get better results.Copyright

A Method of Optimized Utilization of Point-Wise Data Format in Monte Carlo Code

Monte Carlo codes are powerful and accurate tools for reactor core calculation. Most Monte Carlo codes use the point-wise data format, in which the data are given as tables of energy-cross section pairs. When calculating the cross sections at an incident energy value, it should be determined which grid interval the energy falls in. This procedure is repeated so frequently in Monte Carlo codes that its contribution in the overall calculation time can become quite significant. In this paper, the time distribution of Monte Carlo method is analyzed to illustrate the time consuming of cross section calculation. By investigation on searching and calculating cross section data in Monte Carlo code, a new search algorithm called hash table is elaborately designed to substitute the traditional binary search method in locating the energy grid interval. The results indicate that in the criticality calculation, hash table can save 5%∼17% CPU time, depending on the number of nuclides in the material, as well as complexity of geometry for particles tracking.Copyright

Neutronics analysis of the target-moderator-reflector (TMR) configuration for compact long pulsed neutron sources

The compact pulsed neutron source facility can play an important role in the research, education, user training, and development of the advanced neutron scattering instruments. The materials for the target, moderator, reflector (TMR) and their configurations must be optimized to get the optimal yield of neutrons with energy in the range of 1 meV to eV order which depends on the proton energy and its nuclear reaction. Several kinds of materials of the TMR, their configurations, and their dimensions are investigated by the Monte Carlo simulation and optimized for developing the compact pulsed neutron source. The results would contribute to the construction of the Compact Pulsed Hadron Source (CPHS) of Tsinghua University.Copyright

THERMAL NEUTRON CROSS SECTIONS OF COLD-MODERATORS

This paper reviews some moderators for cold neutron source. As a good cold moderator, solid methane was studied and evaluated using the new synthetic frequency spectrum theory (SFS). Due to its high proton density and easy handling, mesitylene (not included in ENDF/B) has also been considered as a very good moderator for cold neutron source. Evaluation of this material in different crystalline phases was done by using a preliminary frequency spectra built combining experimental and synthetic contributions. As a result, we have generated ACE format scattering data files by the NJOY code, with validations of comparing total cross sections from experiments.Copyright

Research on the Radioactivity of Thorium Fuel in Different Reactors

It’s very important to estimate the mass and radiotoxicity of isotopes in spent fuel of thorium fuel cycle, which will benefit the application of the thorium fuel. Much research work has been done on the spent fuel and radioactivity of thorium-based fuel before, yet the difference in usage is always ignored. This paper studies the raise of isotopes in spent fuel of thorium-based fuel cycle in pressurized water reactors and fast neutron reactors, focus on the radioactivity level of actinides and fission products, the important nuclides which have long term radiological impact or give the highest contribution to the total dose on short term after different decay periods. The paper discuss the important nuclides in the measurement of thorium fuel burnup, the conclusion will benefit the actual application of thorium fuel.Copyright