高温气冷堆控制棒区不连续因子的计算与应用

应用不连续因子理论修正的扩散方程,对高温气冷堆侧反射层中控制棒区域的强吸收体和空洞区的局部均匀化进行处理.解决了均匀化区域不存在裂变中子源、边界存在强中子流所带来的的难题,并以输运计算的解为基准进行验算.结果表明:对控制棒区域可进行局部均匀化以及采用简化的边界条件计算不连续因子;用不连续因子修正的扩散方程可准确地处理控制棒.采用不连续因子修正的扩散方程计算高温气冷堆控制棒在计算精度、计算时间上均有优势.     

不连续因子应用于高温气冷堆三维扩散计算

高温气冷堆的三维燃耗计算、三维控制棒价值计算、堆芯功率重构以及芯外探测器响应分析都必须通过三维计算实现.由于高温气冷堆侧反射层中控制棒与吸收球区均为强吸收体,因此,在该区域无法直接用扩散方法计算,而用输运方法实现三维计算又过于耗时.根据不连续因子理论,利用二维(R,θ)几何下输运-扩散耦合计算,实现控制棒与吸收球区的局部均匀化,求得不连续因子和均匀化截面.在此基础上,实现带不连续因子的三维扩散计算.计算结果表明:常规的扩散计算会带来误差,采用不连续因子修正的扩散计算,不但对中子注量率分布改善明显,对本征值、控制棒价值等的改善也很明显,可逼近精细的输运方程的结果,而计算量明显减少.带不连续因子修正的扩散计算是实现高温气冷堆三维计算的有效途径.     

球床氟盐冷却高温堆堆芯热工水力模拟

本文对球床氟盐冷却高温堆堆芯热工流体现象进行了研究。采用计算流体动力学(CFD)方法进行了三维建模和计算,得到了燃料元件球表面温度分布和堆芯冷却剂速度场、温度场和压力的分布,验证了稳态工况下氟盐对堆芯的冷却能力,分析了氟盐的特殊热工流体力学性质对堆芯安全的影响,结果可用于球床氟盐冷却高温堆的初步设计。     

球床高温气冷堆吸收体价值计算的验证

VSOP程序广泛用于球床高温气冷堆的工程设计。对于被布置在堆芯侧反射层孔道中、用于反应性控制的吸收体,由于物理计算方法的限制,VSOP程序不具备计算其价值的功能,必须借助其他确定论程序进行外部耦合计算,涉及到几何的近似处理、截面的归并和转换,可能引入额外的误差。为此,本文采用蒙特卡罗程序建立了精细的堆芯模型,真实描述了堆芯活性区的球床结构、侧反射层的孔道结构、吸收体的形状和位置,在同样的堆芯状态下,比较了确定论耦合程序和MCNP程序计算得到的吸收体价值。结果表明：确定论耦合程序的计算结果是准确的,从设计角度上是偏保守的。     

高温气冷堆控制棒硼燃耗特性分析

控制棒价值及其燃耗规律是核反应堆物理设计关注的要点之一。球床式高温气冷堆控制棒位于侧反射层石墨孔道中,吸收体为圆环形的B₄C,其燃耗特性具有特殊性。采用MCNP耦合燃耗计算模块的方法,对控制棒吸收体进行精细划分,分析了各子区域硼的详细燃耗特性及控制棒价值的变化规律。计算结果表明,由于强烈的空间自屏效应,虽然吸收体外层硼燃耗很多,但吸收体内层硼燃耗很少,因此,反应堆运行寿期末控制棒价值减少很小。     

不连续因子计算强吸收体区域的改进

在球床高温气冷堆控制棒计算过程中,采用局部均匀化和不连续因子修正的扩散方法可取得较好的计算效果,但对具有强吸收性的吸收球区域的效果相对较差。经研究发现,在强吸收体模型中,强吸收体对附近石墨区域的中子通量分布仍具有较大影响,体现出扩散方法对强吸收体的不适应。因此,本文提出改进方案,不但计算吸收体区域边界处的不连续因子,且计算临近石墨区域的边界不连续因子。数值计算结果表明,改进方案可改善计算精度,在扩散计算框架下可达到精细结构的输运计算精度。     

PB-FHR的控制棒布局设计及物理效应

反应性控制系统的设计是反应堆物理设计的主要内容之一。氟盐冷却高温球床堆(Pebble Bed-Fluoride salt-cooled High temperature Reactor,PB-FHR)用B4C吸收体的控制棒作为反应性控制的主要手段。所有控制棒分布于石墨反射层的孔道中,其空间布局、几何结构、中子吸收体的特性参数等是影响控制棒反应性控制的关键因素。本文基于SCALE6程序,以10 MW固态燃料钍基熔盐堆(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF1)(属于PB-FHR)设计模型为参考,系统研究了石墨反射层中控制棒径向位置、有效行程、棒体结构、吸收体长度、吸收体密度等因素对控制棒价值的影响。结果表明,控制棒的径向位置对控制棒价值影响较大;控制棒吸收体长度需综合考虑上下限位及极限下插限位对价值变化的影响;~(10)B的原子密度变化对控制棒价值影响较小。本研究为PB-FHR的反应性控制系统的设计及控制棒的制造加工提供理论参考。     

随机球床模拟方法对氟盐冷却高温堆中子学参数的影响分析

氟盐冷却高温堆（FHR）采用氟盐冷却球形燃料元件,其中子物理计算面临双重不均匀性问题：燃料球在堆芯内的随机排布和包覆燃料颗粒在燃料球中的随机排布。此问题是该堆型设计中面临的主要挑战之一。本文基于MCNP程序和固态燃料钍基熔盐堆（TMSR-SF1）模型完成了不同燃料球床与燃料球描述对关键中子学参数（如k_eff、堆芯能谱、控制棒价值和温度系数等）的影响分析。燃料球床描述使用随机序列添加（RSA）方法建立了随机球床模型与体心立方（BCC）结构的等效规则模型。包覆燃料颗粒描述则基于简立方（SC）等效模型利用MCNP程序中的URAN卡实现随机扰动。结果表明,包覆燃料颗粒随机分布的影响远小于燃料球随机分布的影响;尽管具有相同的总堆积密度,等效规则模型相比于随机球床模型会增加堆芯中子的泄漏,低估冷态满装载反应性约0.5%,高估控制棒价值约5%。     

氟盐冷却球床堆球栅元少群截面计算

氟盐冷却球床堆经10余年的发展,已逐步由预概念设计走向试验堆基准设计。本文采用确定论软件中的碰撞概率法模块对氟盐冷却球床堆球栅元建模,计算了其无穷增殖因数,少群均匀化总截面、俘获截面和裂变截面,并使用连续能量蒙特卡罗软件验证与分析。其中使用基于碰撞概率法的共振处理程序直接求解共振能区超精细群慢化方程,很好地处理了氟盐冷却球床堆的球型燃料元件所构成的双重非均匀系统。结果表明:确定论软件中碰撞概率法模块的计算结果与蒙特卡罗软件结果吻合,适用于对氟盐冷却球床堆进行少群截面加工。     

氟盐冷却球床堆与HTR-10和MSRE的定量相似性分析

氟盐冷却球床堆是当前国际上一种新的研究堆型,尚无已经建造完成的反应堆,因此,选择相似且具有运行经验的反应堆作为基准题有助于堆芯核设计软件适用性分析。利用国际上常采用的相似性分析软件,可对熔盐实验堆(Molten Salt Reactor Experiment,MSRE)及10 MW高温气冷堆(10 MW high-temperature gas-cooled test reactor,HTR-10)与氟盐冷却球床堆的相似性进行分析,定量判断它们作为基准题的合理性。分析结果表明,MSRE和氟盐冷却球床堆的能谱峰位能量接近且堆内元素种类相近,二者相似程度较高;常温临界HTR-10和氟盐冷却球床堆冷却剂不同,且能谱峰位能量差异较大,二者相似程度较低。因此,MSRE是氟盐冷却球床堆中子物理设计软件较理想的基准题。     

A Simple Method for Control Rod Worth Evaluation with Diffusion Codes

To present a simple method for calculating the worth of control rods in multiregion reactor cores, the well model approximation is applied to the finite difference diffusion code. The results of calculation are compared with experimental results obtained on the ōzenji Critical Facility, and also with calculations using transport equivalent constants. With its simplicity, the method proposed is accurate enough to suffice amply for most cases where only the reactivity is the end result sought, although the application of this method is limited to control rods inserted in the core region, and gives no information about the detailed neutron flux distribution around the control rod.     

SPH等效均匀化方法研究

SPH等效均匀化方法是一种通过调整截面参数使得均匀化前后反应率保持守恒的均匀化方法。近年来,由于计算机硬件条件的飞速发展和对更高精度堆芯分析能力的要求,基于Pin-by-Pin的全堆芯计算越来越受到业界的关注。SPH均匀化方法不需要保存额外的均匀化参数(不连续因子),是栅元均匀化的首选方法之一。本文研究了SPH因子的求解方法及其应用,证明了在组件层面上SPH修正后均匀的扩散计算能够完全恢复非均匀输运计算结果。本文对由UO2燃料和MOX燃料组成的Colorset问题进行了检验,数值结果表明,与传统的通量-体积权重的均匀化方法相比,基于SPH均匀化方法的细网计算可以更好的预测控制棒价值和燃料棒功率分布。     

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.     

微型中子源堆低浓化堆芯的探讨

通过对２３５Ｕ富集度为１９．９％的ＵＯ２和Ｕ３Ｓｉ２－Ａｌ的弥散体２种燃料进行物理计算，从中筛选出了优化的堆芯方案，并对其静态物理参数，诸如有效倍增因子、绝对中子通量密度、上铍反射层反应性价值、反应性温度系数、控制棒价值等进行了计算。     

Benchmark analysis of SPERT IV reactor

Benchmark calculations have been performed for SPERT IV D-12/25 core. Experimental data of the core was provided by International Atomic Energy Agency (IAEA). Combination of WIMS/D4 and CITATION codes has been used for performing the neutronic analysis of the reactor. Lattice calculations have been performed through WIMS/D4 while 3-dimensional reactor core has been modeled in CITATION. Ten energy groups were considered for these calculations. Energy wise microscopic cross-sections were generated for fuel, control absorber, control follower, guide tube, grid plate, reflector and structural regions separately of the core using WIMS/D4. Thermal neutron flux profiles at different axial and radial locations of the core were evaluated. Critical position of the control rods, excess reactivity, shut down margin, control rod worth, reactivity feed back coefficients and kinetic parameters of the core were estimated. Reasonable agreement has been found between experimentally determined and the calculated parameters.     

Calculation of the Cruciform Control Rod Worth

A new method of calculating the worth of cruciform control rods arranged in the D-lattice is presented. The lattice is divided into two regions: a transport region near the control rods, and a diffusion region removed from the rods. The neutron behavior in the former region is treated by collision probability method and in the latter by diffusion approximation. The prompt decay constants calculated by the present method are in very good agreement with experimental data, which evidences the accuracy of the present method for evaluating cruciform control rod worth.     

原型微堆低浓化初步研究

利用蒙特卡罗计算程序,对高浓铀为燃料的原型微堆的有效增殖因数、控制棒价值、上铍反射层价值以及辐照座内的中子注量率等参数进行了计算。将计算值与实验结果进行了比较,两者基本相符。在原型微堆堆芯尺寸保持不变的情况下,将堆芯燃料元件芯体用富集度为12.5%UO₂替换UAl和用锆包壳替换铝包壳,对堆芯燃料低浓化方案进行了计算,给出了方案的计算结果。并利用RELAP5程序计算了原型微堆低浓铀堆芯阶跃引入4.0 mk反应性情况下反应堆的相关参数。     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.     

Dependence of control rod worth on fuel burnup

One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.