VVER nuclear fuel burnup with different absorbers

A computational study is performed of the fuel burnup in VVER-1000 using different absorbers in open and closed fuel cycles. It is shown that mixtures of plutonium isotopes (energy and others) can give the same effect as gadolinium, which is currently used. Fuel burnup increases. When neptunium, americium, and curium isotopes are used as a consumable absorber in a closed fuel cycle, the accompanying effect is elimination of long-lived α-emitting radionuclides which have accumulated in long-term repositories.     

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。     

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。     

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。     

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.     

基于SCALE/TRITON的单流双区熔盐堆燃耗计算方法

美国橡树岭国家实验室开发的SCALE/TRITON程序广泛用于反应堆临界安全、中子物理、辐射屏蔽和灵敏度与不确定度等方面的计算分析。基于SCALE/TRITON程序,采用等效体积、均匀混合和平均截面等三种外部耦合方法,处理单流双区熔盐堆的燃耗计算,解决了SCALE/TRITON程序在计算中不能精确反映流动燃料周期性均匀混合的问题。研究表明平均截面法与均匀混合法的计算结果几乎完全一致,与橡树岭文献结果也能很好符合,等效体积法因为没有考虑堆芯分区结构的差异而导致计算结果与其他两种方法偏离较大。基于SCALE/TRITON发展的平均截面法,放宽了对步长的要求,具有准确性好、计算效率高的优点,适用于熔盐堆两区(或多区)的堆芯设计与燃耗性能分析,具有重要的应用意义。     

Criticality and burnup analyses of a PBMR-400 full core using Monte Carlo calculation method

In this study, the criticality and burnup analyses have been performed for full core model of Pebble Bed Modular Reactors, such as PBMR-400, using the computer codes MCNP5.1.4 and MONTEBURNS 2.0. Three different pebble distributions, namely; Body Centered Cubic (BCC) (packing fraction = 68%), Random Packing (RP) (packing fraction = 61%) and Simple Cubic (SC) (packing fraction = 52%) were selected for the analyses. The calculated core effective multiplication factor, k_eff, for BCC, RP and SC came to be 1.2395, 1.2357 and 1.2223, respectively. The core life for these distributions were calculated as ～1200, 1000, and 800 Effective Full Power Days (EFPDs), whereas, the corresponding burnups came out to be ～99,000, ～92,000 and ～86,000 MWD/T, respectively, for end of life k_eff set equal to 1.02.     

A co-ordinate transformation technique for studying flux perturbations induced by strong vibrating absorbers

Calculation of the neutron noise, induced by small amplitude vibrations of a strong absorber is a difficult task because the traditional linearisation technique cannot be applied. Two methods, based on two different representations of the absorber, were developed earlier to solve the problem. In both methods the rod displacements are described by a Taylor expansion, such that the boundary condition needs only be considered at the surface of a static rod. Only one of the methods is applicable in two dimensions. In this paper an alternative method is developed and used for the solution of the problem. The essence of the method is a variable transformation by which the moving boundary is transformed into a static one without Taylor expansion. The corresponding equations are solved in a linear manner and the solution is transformed back to the original parameter space. The method is equally applicable in one- and two dimensions. The solutions are in complete agreement with those of the previous methods.     

Heterogeneous calculation of hexagonal lattices by the method of passage probabilities

A method is proposed for determining the characteristics of heterogeneous hexagonal cells using first-collision probabilities—probabilities of passage, escape, and collisions, as well as calculations of the flows in cell zones on the basis of balance equations for the current. The spatial-angular dependence of the current at the sides of a cell is described in a discrete approximation, and the nonuniformity of the sources and flows in zones in described using spatial harmonics. The results obtained in one-group calculations of model lattices by the passage probability and Monte Carlo methods are compared. 3 figures, 5 tables. 5 references. IPé, Belarussian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 330–335, November, 1999     

Curve fitting to defect size distributions for the calculation of failure probabilities

The imperfection size distribution of pipe welds reported by Raussi and Tiainen [1] is analysed. A lognormal distribution and a Weibull distribution are shown to be compatible with the data.     

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.     

Algorithms - Optimal algorithm for collision probabilities calculation in 3-D rectangular-cylindrical geometries

In 3D geometry, the problems of the collision probability (CP) method related to high memory and execution time requirements increase rapidly with the number of regions. The paper shows an attempt to minimise these requirements at the level of probability matrix evaluation. The mixed method (MM) for tracking and probability calculations, and an improved module of CPs contribution calculation, were developed. Finally, the new schema has approximately 50% reduced memory and execution time compared to the old version.     

Increasing fuel burnup in VVER

The results of investigations of fuel burnup increase in VVER are presented. The influence of the costs of different technological stages, changes in the number of refuelings and run time, fuel enrichment and waste, and the consumption of natural uranium on increasing burnup is examined. An analysis taking account of the uncertainty of future prices is performed. The price for natural uranium up to 2020 is estimated using a model. The results presented in this article show that the cost reduction in the fuel component with an increase of VVER fuel burnup in an open fuel cycle can be 6%. __________ Translated from Atomnaya énergiya, Vol. 104, No. 3, pp. 137–141, March, 2008.