First-collisions probabilities calculation of strong absorbers burnup

A method for calculating, using a matrix neutron first-collisions probabilities which is reconstructed at each step, the burnup of cells and fuel assemblies is presented. A method for reconstruction and correction of the first-collisions probabilities using average chords to the first collision of a neutron, which are calculated using the geometric module for reconstructing the stochastic trajectories of neutrons, is described. The results of a calculation of the multiplication coefficient of elementary cells with different material composition relative to the reference cell are presented. Computational results are presented for the burnup of a VVER fuel-assembly fragment with a consumable absorber are presented.     

燃耗截面基本库对输运燃耗耦合计算的影响分析

基于蒙特卡罗中子输运程序和ORIGEN2点燃耗程序的蒙特卡罗输运燃耗耦合计算方法应用广泛。但现有评价库中子连续截面的核素个数远小于燃耗计算涉及到的核素数量,即通过输运计算得到的燃耗截面不足以完全替代燃耗计算的基本库。采用经过栅元验证的蒙特卡罗燃耗程序MCBMPI,对最新的VERA燃耗计算基准题进行验证计算,对比分析不同的燃耗截面基本库对输运燃耗计算的影响。分析结果表明:1)在实际应用中尽量不要采用典型热中子截面库,会带来较大偏差;2)在燃耗计算核素替换较多的情况下,对该基准题而言,选取典型压水堆基本库还是典型快堆基本库,对结果影响不大,二者keff偏差在8‰以内,燃耗末期235U偏差在4‰以内,135Xe偏差在5‰左右;3)建议选取与研究对象能谱相近的基本库。     

Performance of Woodcock delta-tracking in lattice physics applications using the Serpent Monte Carlo reactor physics burnup calculation code

This paper presents the delta-tracking based geometry routine used in the Serpent Monte Carlo reactor physics burnup calculation code. The method is considered a fast and efficient alternative to the conventional surface-to-surface ray-tracing, and well suited to the lattice physics applications for which the code is mainly intended. The advantages and limitations of the routine are discussed and the applicability put to test in four example cases. It is concluded that the method performs well in LWR lattice applications, but really shows its efficiency when modeling HTGR particle fuels.     

关于深燃耗堆芯的反应性测量

介绍了在实验性PWR堆上完成的深燃耗条件下测量反应性概况。用实验结论剖析了在国外核电站堆芯上应用噪声分析法对慢化剂温度系数作全燃耗期监测研究中出现的测量结果与常规方法相差2～5倍的现象。从测量公式和堆芯扰动模型图入手所作的分析结果说明,没有消除随燃耗不断增长的强自发裂变中子源干扰是产生差异的根本原因。事实说明:在多种噪声分析技术中,只有能够清除自发裂变中子源干扰的方法才能成功地应用于燃耗后堆芯的反应性测量。     

Speed-up of cross-section collapsing with weight-window for subcritical burnup calculation

In the author’s group, a fusion–fission (FF) hybrid energy system has been analyzed using our own burnup calculation system consisting of Monte Carlo transport code MCNP-4C and point burnup code ORIGEN2.1. Since the neutron energy spectrum changes along with progress of burnup in a subcritical system, it is necessary to update one-group cross-section library in each burnup step. The one-group cross-sections are normally updated by collapsing the evaluated nuclear data such as JENDL and ENDF using a neutron flux calculated by an appropriate transport code such as MCNP. The collapsed cross-sections are handed over to ORIGEN, and the reaction rates for burnup of elements are thereafter estimated accurately.As well known, MCNP generates track-length (TL) data in the neutron transport calculation, which are base data to estimate the neutron flux. We thus use the track-length data directly instead of the calculated neutron flux, in order to evaluate the reaction rate as accurately as possible. However, the number of TLs becomes extremely large and thus it takes a longer computation time. We therefore reduce the number of TLs used in the cross-section collapsing process as far as the accuracy is conserved. However, in some energy region the number of TLs is inversely too small to conserve the original cross-section accuracy of the evaluated nuclear data files, because the number of TL data per unit energy is smaller than that of the nuclear data.In the present study, the weight-window (WW) technique of MCNP was applied to our burnup calculation system in order to control the number of TLs in such an energy region artificially and to complete the collapsing process accurately in the whole energy region. As a result, the variance of the calculated neutron flux thus deteriorates slightly, but the number of TLs could be successfully adjusted to conserve the accuracy of the nuclear data file in the whole energy region. And the accurate reaction rate estimation for burnup with MCNP was finally realized and simultaneously the computation time could be saved reasonably.     

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes.     

燃耗信任制临界计算中保守性因素研究

在运用燃耗信任制技术进行乏燃料储存、运输等环节的临界安全分析时,临界计算所采用的条件是否具有足够的包络性十分关键。本文借助于OECD/NEA发布的若干燃耗信任制临界安全基准题,使用SCALE5.1软件中的STARBUCS模块进行分析,对信任核素选取、乏燃料冷却时间以及端末效应等因素对乏燃料系统临界安全性的影响进行了研究,得出了各参数保守性的有关结论。     

基于蒙特卡罗方法的三维燃耗计算研究

采用通过编写连接MCNP程序和ORIGEN2程序的接口处理程序的方法进行快中子系统的燃耗计算。由MCNP、ORIGEN2、接口处理程序和截面文件组成的软件系统可用于燃料或堆芯非均匀布置快中子系统的燃料同位素成分和燃耗反应性损失计算,在燃耗反应性损失计算中采用了伪裂变产物的方法。介绍程序系统的研制情况,并给出用该软件系统计算中国实验快堆首炉堆芯和OECD/NEAMOX燃料快堆基准题的燃耗计算结果。     

Application of CANDLE burnup to block-type high temperature gas cooled reactor

The CANDLE burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a burnup control rod can be eliminated, and several merits are expected. This burnup may be realized by enriched uranium and burnable poison with large neutron absorption cross-section. With the fuel enrichment of 15%, gadolinium concentration of 3.0%, and fuel cell pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half-width of power density distribution of 1.5 m. When the concentration of natural gadolinium is higher, the burning region moving speed becomes slower and the burnup becomes higher, though the effective neutron multiplication factor becomes smaller. When U-235 enrichment is higher, the effective neutron multiplication factor becomes larger, the speed becomes slower, and the burnup becomes higher. When the pitch is wider, the effective neutron multiplication factor becomes larger, the speed becomes faster, and the burnup becomes higher.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and thesandwichmethod was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and the"sandwich"method was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

A fast and reliable calculation model for BWR fuel assembly burnup analyses

A code called superb has been developed for the BWR fuel assembly burnup analyses using a supercell model. Each of the characteristic heterogeneities of a BWR fuel assembly like water gap, poisoned pins, control blade etc. is treated by invoking the appropriate supercell concept. The burnup model of superb is so devised as to strike a balance between accuracy and speed. This is achieved by building isotopic densities in each fuel pin separately while the depletion equations are solved only in a few group of pins or burnup zones and the multigroup neutron spectra are differentiated in fewer group of pincell types. Multiple fuel ring burnup is considered only for Gd isotopes. A special empirical formula allows the microscopic cross section of Gd isotopes to be varied even during burnup integration.The supercell model has been tested against Monte-Carlo results for the fresh cold clean Tarapur fuel assembly with two Gd fuel pins. The burnup model of superb has been validated against one of the most sophisticated codes lwr-wims for a benchmark problem involving all the complexities of a BWR fuel assembly.The agreement of superb results with both Monte-Carlo and lwr-wims results is found to be excellent.     

An analysis of burnup reactivity credit for reactor RA spent fuel storage

The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described.     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。