Gd burnable poison system for reactivity control of the first cycle of a PWR

The reactivity control of a PWR core may be performed by a system of burnable poison (BP) rods. In such a case, the soluble B system may be eliminated and the BP rods will be responsible for the excess reactivity provided for fuel depletion and fission products accumulation. A strong negative moderator temperature coefficient is a desirable safety feature, inherent to a poison-free moderator. The design objective of a PWR core controlled completely by a system of BP rods is achieved by utilization of Gd as the poison material and annular geometry of a BP rod. The proposed concept is tested as a retrofittable option for the current generation, as well as new PWR plants. A plausible incore fuel-management scheme is demonstrated, with planar power distribution, close to an acceptable range. The fuel-cycle penalty due to the residual poison content at EOC is relatively small.     

Whole Core Calculations of Power Reactors by Use of Monte Carlo Method

Abstract

Whole core calculations have been performed for a commercial size PWR and a prototype LMFBR by using vectorized Monte Carlo codes. Geometries of cores were precisely represented in a pin by pin model. The calculated parameters were k _eff, control rod worth, power distribution and so on. Both multigroup and continuous energy models were used and the accuracy of multigroup approximation was evaluated through the comparison of both results. One million neutron histories were tracked to considerably reduce variances. It was demonstrated that the high speed vectorized codes could calculate k _eff, assembly power and some reactivity worths within practical computation time. For pin power and small reactivity worth calculations, the order of 10 million histories would be necessary. It would be difficult for the conventional scalar code to solve such large scale problems while the present codes consumed computation time less than 30 min for a PWR and 1 hour for an LMFBR. Required number of histories to achieve target design accuracy were estimated for those neutronic parameters.     

固定棒位法测量控制棒总价值

控制棒价值测量的准确度与效率对核电厂的安全性与经济性具有重要影响。在动态刻棒等反应性测量工作中,本底与中子源对探测器有显著影响,致使根据实测电流计算得到的反应性显著偏离真实值。基于点堆逆动态方程,通过对本底与中子源影响的分析,利用固定棒位状态下的测量数据计算反应性并得到控制棒总价值,给出了一种不受本底与中子源影响的简便的控制棒总价值测量计算方法,并在零功率实验装置上进行验证。结果表明,该方法可有效避免本底和中子源组件对反应性探测的影响,并简化了离线理论计算,其与周期法计算结果的相对偏差在1%以内。     

超级均匀化方法用于球床氟盐冷却高温堆扩散计算

球床氟盐冷却高温堆的控制棒位于侧反应射层内,存在无裂变中子源且受堆芯泄漏谱强烈影响的强吸收体区域扩散计算难题。超级均匀化方法(Super Homogenization,SPH)被用于对氟盐球冷却床堆侧反射层中控制棒区域的强吸收体进行等效均匀化处理,同时堆芯除控制棒区域外采用谱修正方法(Spectra Modification,SM),将输运计算的结果作为基准进行验算。结果表明,SM-SPH模型能有效地计算球床氟盐冷却高温堆反射层控制棒价值及通量分布,并且较常规的SPH方法能更好地处理棒间干涉效应。     

Investigation of axial power gradients near a control rod tip

Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations.     

Control Rod Worth in High Conversion PWR

In a High Conversion PWR (HCPWR), the neutron spectrum is hardened in comparison with the conventional PWR and the nuclear characteristics of the control rod is much different. We have calculated the control rod worth in a HCPWR for several materials of strong neutron absorbers and selected B₄C for the control rod material. For the B₄C control rod, we have investigated the various nuclear characteristics such as the spatial and energy dependency of neutron absorption, the dependency of the control rod worth on moderator to fuel volume ratio and Pu enrichment etc., and obtained the useful data for the nulear design of control rods in a HCPWR.     

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.     

Estimation of irradiated control rod worth

When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber’s burnup. The suggested methodology is based on measurements of the rod’s worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.     

Interpretation of Spatial Kinetics for Rod Drop Experiment in Loosely Coupled-Core Systems

Abstract

The coupled two-core reactor systems with various degrees of spatial coupling were constructed in the Kyoto University Critical Assembly (KUCA) to study the spatial kinetics observed in the control rod drop experiment. By applying the two-mode and the two-point kinetic models to the space-dependent rod worths measured on the basis of the one-point model, the first-harmonic λ-mode eigenvalue separation and the reactivity coupling coefficient were inferred. The present values of these parameters agreed with the results obtained by the reactor noise measurements and the diffusion calculations.

The experimental results show that the magnitudes of the spatial kinetic phenomena including the dependence of the rod reactivity worth on the detector position, the reactivity interaction effect between control rods and the transient flux tilts induced by the rod drop, which have been significantly observed in large thermal and fast power reactors, are inversely proportional to the eigenvalue separation. Applying the two-mode model, the inherent reactivity worths of control rods were also inferred from the space-dependent ones.     

CNP650压水堆不调硼负荷跟踪可行性研究

海南昌江核电厂等CNP650压水堆采用Mode-A控制模式,该模式采用黑体控制棒,有很好的基负荷运行能力,但负荷跟踪能力相对较差。而对一些具有小电网的国家或地区,负荷跟踪运行能力具有一定的市场需求。不调硼负荷跟踪通过棒控系统自动完成,大大减轻了操纵员负担;负荷跟踪过程基本不需要频繁地调硼操作,允许简化化学和容积控制系统设计,减少了废液处理成本。为此,在CNP650压水堆上进行了不调硼负荷跟踪研究。负荷跟踪过程主要有两个控制任务:一是反应性补偿;二是功率分布控制。根据不调硼负荷跟踪的控制任务,重新进行了控制棒的设计、分组和布置,设置两套独立控制的控制棒组(功率补偿棒组和轴向偏移控制棒组),分别用于堆芯反应性控制和轴向功率分布控制,以实现不调硼负荷跟踪。使用SCIENCE程序包进行典型的12h~3h~6h~3h、100%—50%—100%功率水平的日负荷循环计算来进行不调硼负荷跟踪分析。计算步骤为:进行三维堆芯模型计算;根据三维堆芯模型建立一维堆芯模型;在一维模型基础上,进行模拟计算。完成了海南昌江核电厂平衡循环寿期末典型的日负荷循环不调硼运行分析,模拟计算结果表明在CNP650压水堆上不调硼负荷跟踪运行模式是可行的。     

Research on Segment Design of Control Rod in Sodium-cooled Fast Reactor

In sodium-cooled fast reactors, control rods are commonly used to compensate for the excess reactivity and shut down the reactor. The traditional sodium-cooled fast reactor design consists of the safety rod, shim rod and regulating rod. The 10B enrichment of the shim rods is relatively higher, which unavoidably increases the burnup, the heat generation and the power peak factor of the fuel assemblies around the shim rods. To solve this issue, the segment design of control rods was proposed. Compared with traditional design, the new design can significantly reduce the heat generation by about 30 percent and burnup of control rods by about 50 percent, as well as improve the power peak factor of the fuel assemblies around the shim rods. The replacement cycle of the control rods can be extended by time.     

钠冷快堆控制棒分段设计研究

在钠冷快堆中,反应堆运行时的反应性补偿和停堆安全主要由控制棒来实现。当前的钠冷快堆设计中,一般含有安全棒、补偿棒和调节棒。其中,补偿棒中10B的富集度较高,使补偿棒的燃耗较高,且发热量较大,并造成周围燃料组件功率峰因子偏大。本文提出一种分段设计方案,可用于改进上述缺点。该方案相比于传统方案,控制棒发热减小约30%,控制棒燃耗减小50%,并能有效改善周围燃料组件的功率峰因子,控制棒更换周期可提升1倍。     

控制棒尖齿效应中非均匀不连续因子处理技术

To handle the control rod cusping effect in pressurized water reactor (PWR) fuel management calculation, the variational nodal method (VNM) in the fuel management calculation code system NECP-Bamboo has been extended to tread the heterogeneous cross section distribution by expanding the volumetric cross sections into piece-wise polynomials in the early work. However, the partially inserted control rods also introduces heterogeneous discontinuity factor (DF) on nodal interface. Thus, in this paper, an ultimate solution is proposed to fully handle this problem. Firstly, the surface integral in the VNM is modified to contain the discontinuity of neutron flux, incorporating a continuous discontinuity factor in that term. Secondly, the surface DF is expanded into the sum of pieces-wise orthogonal polynomials to construct the nodal response matrixes. Comparing with current representative re-homogenization methods, the application numerical results of the BEAVRS benchmark problem demonstrate the effectiveness of the heterogeneous VNM with heterogeneous DF. It can eliminate the cusping effect by providing more accurate differential control rode worth curves and pin power distributions.     

控制棒尖齿效应中非均匀不连续因子处理技术

为了在压水堆（PWR）中限制控制棒尖齿效应,燃料管理计算程序系统NECP-Bamboo中的变分节块法（VNM）在之前的工作中通过将截面用分片多项式展开,具备了处理非均匀截面的能力。但半插的控制棒同时在节块表面带来了非均匀的不连续因子（DF）。本文提出了一个可以完全解决这个问题的方案。首先,通过包含非均匀不连续因子的表面积分将非均匀不连续因子显式地表达在变分节块法的泛函中,然后用分片多项式展开不连续因子,使其出现在响应矩阵的构建中。与现有的再均匀化方法相比,BEAVRS基准题的数值结果表明,含非均匀不连续因子的非均匀变分节块法可以消除控制棒尖齿效应,同时获得更准确的功率分布。      

300MW压水堆核电厂堆芯反应性控制组件的设计和研究

总结了我国３００ＭＷ压水堆核电厂堆芯反应性控制组件设计的基本经验。针对控制棒的主要失效模式，讨论了关键的技术问题，对于首次使用的的硼硅酸盐玻璃可燃毒物，着重研究了抗强辐照性能，以于次级和初级中子源棒，分别阐述了重要的内压问题和有关的安全性能要求.?     

Pin Power Reconstruction Methods of the Few-Group BWR Core Simulator NEREUS

An improved pin power reconstruction method has been incorporated in the few-group nodal BWR core simulator NEREUS, which is based on the analytic polynomial nodal method. With the analytic polynomial nodal method, accurate node surface fluxes are available, which are used later to reconstruct pin powers. The intranodal homogeneous thermal flux is corrected using a semi-empirical proportional relation between surface transition components of the homogeneous and heterogeneous fluxes. This correction method is effective for BWR calculations, especially for controlled assemblies or mixed loading of off-set assemblies. A unified model accounting for effects of spectral histories, caused by spectral interactions between fuel assemblies and the control blade insertion, was also developed for intranodal burnup correction. The change in pin powers due to the control blade history can be predicted well, without laborious assembly depletion calculations with control blade insertion. R-factors used in critical power ratio calculations are also reconstructed from the pin powers. The NEREUS pin power reconstruction method was verified against heterogeneous multi-assembly depletion calculations.     

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.     

Feasibility Study for Evaluation of Control Rod Worth in Pressurized Water Reactors using Neutron Count Rate during a Control Rod Drop Testing

Applicability of the modified Neutron Source Multiplication (NSM) method with extraction of the fundamental mode to subcriticality measurement has been proposed. Following the feasibility verification in the previous study based on numerical analyses, its applicability has been proven in a more realistic situation; in a withdrawal sequence of control rod banks during the PWR startup. Subcriticalities with various control rod insertion configurations were estimated based on the modified NSM method. The subcriticality could be evaluated with a good accuracy even with the mockup experiment where any special treatments for accurate measurement were not taken into account and furthermore the insensitivity of measured signals by reactivity changes and their large fluctuations were seen.

Based on this fact, we further investigated a feasibility to use neutron count rate data obtained during the control rod drop testing, which is carried out before the reactor physics tests at hot zero power condition. When it is proven that these data could be used for the estimation of each control rod worth, the following reactor physics tests could be performed with the advanced knowledge of each control rod worth and procedures for detailed control rod worth measurement could be simplified or eliminated from the reactor physics tests.     

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。     

启明星Ⅱ号安全控制部件的实验验证

启明星Ⅱ号零功率装置（启明星Ⅱ号）所设计的安全控制部件有安全棒和调节棒,这些控制部件是反应堆安全运行的关键。本文采用逆动态反应性计测量的方法对所选定的控制部件的反应性价值进行了实验测量,并与理论计算结果进行了比较。结果表明,安全控制部件的反应性价值的实验测量结果与理论计算结果的相对偏差为4.46%,二者吻合较好。安全棒系统经力学分析评定,结果表明不会出现卡棒现象,能实现快速停闭反应堆的目的。安全棒系统、调节棒系统的机械性能经堆上反复实验验证,各系统性能稳定可靠,重复性好。