热管式锂冷空间快堆中子学计算分析

采用MCNP程序对锂冷热管式锂冷空间快堆建立中子学计算模型,对其中子通量密度和功率分布、有效增殖因子等进行了计算,采用分区燃料布置,得到满足长寿命运行要求的分区装载方案,利用ORIGEN2程序进行燃耗校核,计算了转鼓的价值和转鼓转角随运行时间的变化情况。模型分析结果表明:分区装料后的堆芯满足临界安全设计和不均匀系数要求;堆芯的过剩反应性足够7年不换料满功率运行;意外发射失败掉入湿沙或海水中,由于有谱移吸收体铼,堆芯仍然保持足够的次临界度;转鼓的价值可以保证堆芯在整个寿期内安全的停堆和正常的启动;热管式锂冷空间快堆基本物理特性合理,满足设计要求。     

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。     

加速器驱动系统的燃料选择

对装载不同增殖材料的现实加速器驱动系统(ADS)的安全及嬗变超铀核素特性进行研究。分别 以(U,TRU)O2和(Th,TRU)O2作为堆芯燃料,先用LAHET和MCNP程序对ADS进行稳态模拟计 算,再耦合MCNP和ORIGEN2程序计算燃耗过程中的核素密度变化。结果显示,装载钍基燃料的 ADS对超铀核素的嬗变效果较好,且在燃耗过程中其反应性和质子流强波动较小;装载铀基燃料的 ADS则具有更安全的多普勒效应和缓发中子有效份额。总体来看,如果需要堆长时间安全嬗变超铀核 素,装载钍基燃料会取得更好的效果。     

加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

运用MCNP与ORIGEN2耦合计算程序COUPLE，对加速器驱动的次临界系统（ADS）钠冷金属燃料快堆堆芯进行稳态与燃耗计算，比较分析次锕系核素（MA）非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明，对比均匀布置，非均匀布置具有更高的MA嬗变率与嬗变支持比，在反应性参数方面导致多普勒效应与有效缓发中子分额降低，钠空泡效应增大，在堆芯功率分布与加速器束流功率方面没有明显变化。     

气冷快堆燃料组件均匀化初步研究

气冷快堆是第4代核能系统候选方案之一,具有高温多用途、能增殖等优点。本工作以一气冷快堆的设计方案为研究对象,针对单组件模型和全堆芯模型,采用MCNP耦合ORIGEN的方法,计算了有关临界、燃耗过程的几个重要物理特性,比较了采用精细化结构和组件均匀化方法在计算精度、计算时间等的差别,说明了采用组件均匀化方法进行气冷快堆全堆燃耗计算的必要性和可行性。     

基于CITATION-ORIGEN2球床堆平衡态计算程序的实现

基于确定论中子扩散软件CITATION和点燃耗软件ORIGEN2,编写了球床堆分析程序COBBLE,以实现指定燃料球加载策略下的球床堆平衡态燃耗计算。COBBLE程序采用谱区能谱修正方法,通过迭代求解得到球床堆堆芯平衡态参数。本文选取简化的球床模块高温气冷堆(PBMR)堆芯进行建模,计算其功率分布及燃耗分布,并使用基于蒙特卡罗方法的球床堆燃耗计算程序PBRE进行了验证与分析。结果表明,COBBLE程序适用于球床堆的平衡态燃耗计算。     

基于MCNP-ORIGEN2耦合程序的小型行波堆堆芯概念设计

研究设计了基于中国实验快堆(China Experimental Fast Reactor,CEFR)的小型"行波"概念堆。采用中子输运程序MCNP和点燃耗程序ORIGEN2的耦合程序进行堆芯设计,重点研究了不同点火组件的富集度和不同布料方案对小型堆的物理参数的影响,设计堆芯寿期为30 a,并给出相应的倒料方案。不同点火组件富集度对比结果表明,小堆需要选取合适的富集度,富集度太低无法维持临界,而太高会影响堆芯增殖效应;而低泄漏和棋盘式布料两种方式对比结果表明,后者的增殖组件增殖效应明显高于前者。最终确定倒料周期为8 a,倒料三次,堆芯实现较长寿期,且整个寿期内反应性变化小,各组件燃耗深度相对均匀,组件平均卸料燃耗深度约为238 MWD/kg HM。     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

加速器驱动的次临界系统的燃耗分析计算和堆芯优化设计

以加速器驱动的次临界系统（ADS）在事故情况下仍处于次临界、k_eff随燃耗时间变化的最大范围不超过1.5%和包壳材料HT9钢可承受的最大辐照损伤的前提下,将堆芯燃料区分为嬗变区和增殖区,并将整个过程保持嬗变区的燃料成分不变。通过对ADS燃料的组成成分、堆芯布置和堆芯功率分布等方面的研究,在Pu的外层富集度与内层富集度之比为1.0～1.5范围内,调整增殖区的燃料成分,并利用MCNP和ORIGEN耦合的COUPLED2程序计算k_eff随燃耗时间的变化。同时,综合考虑功率展平、次锕系核素的嬗变率和燃耗深度等因素,建立1套符合工程实际的次临界系统。     

热管式空间反应堆燃耗计算研究

针对热管式空间反应堆,基于OpenMC程序产生均匀化截面参数,并由确定论快堆分析程序SARAX进行堆芯输运及燃耗计算。以蒙特卡罗程序（MCNP）的输运计算结果以及MVP程序的燃耗计算结果作为参考解,通过对比稳态输运计算和燃耗计算的结果,证明了耦合的OpenMC和SARAX程序系统对于空间堆中子学分析和燃耗分析的适用性和高效性。为热管式空间反应堆的设计分析提供了参考。      

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.     

Comparisons of HELIOS,ORIGEN2, and Monteburns Calculated 241 Am and 243 Am Concentrations to Measured Values for PWR,BWR, and VVER Spent Fuel

A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating ²⁴¹ Am and ²⁴³Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The ²⁴¹Am and ²⁴³Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for ²⁴¹Am and ±20% for ²⁴³Am) and the Garigliano measurements (within ±12% for ²⁴¹ Am and ±20% for ²⁴³Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for ²⁴¹Am and ±12% for ²⁴³Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials.     

球床高温堆平衡态燃耗计算程序的开发

基于MCNP5和ORIGEN2耦合方法,开发了平衡态下球床高温堆的燃耗计算程序PBRE,用于堆的性能价值分析。为节省蒙特卡罗计算时间,对迭代收敛的方法进行优化,使之可在10个迭代步内收敛。使用PBRE对清华大学HTR-10进行建模计算,得到的平均卸料燃耗深度与文献报道值一致,表明PBRE程序适用于球床堆平衡态的燃耗分析。     

基于CRAM的压水堆燃耗程序开发及验证

本文基于Cinder90燃耗数据库开发了燃耗求解程序MCRAM,并耦合MCNP程序对重要的锕系核素和裂变产物核素的反应截面进行了修正。以OECD/NEA乏燃料成分基准数据库中的Takahama-3压水堆燃料组件为基准题,对MCRAM程序的计算结果进行了验证,并与其他程序的计算结果进行了比较。结果表明,MCRAM程序对重要裂变产物和主要锕系核素的计算结果相对偏差小于5%,计算精度与ORIGEN2程序的相当。与此同时,同一例题的计算效率MCRAM较之MCNTRANS程序提高了近200倍。     

Effect of fission products yield library on the calculation of the nuclide composition of the fuel in sodium-cooled fast reactors

The calculation of the composition of irradiated fuel for different degrees of burnup is a basic problem in the analysis of nuclear-radiological safety of objects holding spent fuel assemblies. The yield of fission products is one of the important initial indicators in burnup calculations. Methods for compiling libraries of fission products yield on the basis of the ENDF/B up-to-date evaluated nuclear data files are described. The nuclide composition of uranium oxide and uranium-plutonium-zirconium metal fuel in sodium-cooled fast reactors is analyzed by means of high-precision calculations performed with different fission product yields libraries using different computer codes MONTEBURNS–MCNP5–ORIGEN2 and the results are presented.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Requirements on fuel management for a safe and optimum operation of the German research reactor FRJ-2

In case of a failure of a coarse control arm (CCAs) at FRJ-2, reactivity is added to the reactor. The amount of this reactivity depends on the efficiency of the individual CCAs which has been measured as 180% of the average reactivity of the six arms for the central arm. For this design basis accident, it is required that only four out of five residual arms are capable of shutting down the reactor. This minimum shutdown reactivity is provided by an optimum fuel management including an experimental reactivity determination. Calculation of fuel burnup and material densities is performed by the depletion code SUSAN, which has been verified by separate calculations using ORIGEN. The difference in the reactivity values (between calculation and measurement) is mainly a consequence of the limitation of the inverse kinetic method, which is incapable of covering the effects of the flux deformation and interaction of the CCAs and core in the process of reactor scram.