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.     

Methodology for a numerical simulation of an insertion or a drop of the rod cluster control assembly in a PWR

In a pressurised water reactor, the rod cluster control assembly is a system which controls the neutronic activity of the core. It consists of long rods, connected by a spider fixture and a cylindrical system for the control drive mechanism. At its withdrawn position, the activity of the core is maximum, and at its completely inserted position, the activity of the core vanishes. In case of emergency, an effective way to shutdown the reactor is to let it drop under its own weight. An other way to verify the efficiency of the rod cluster control assembly is the insertion test. It consists in inserting the rod into its guides and in checking if the reaction friction force is not high enough to block the movement of the rod cluster control assembly.We present in this paper a methodology for a numerical simulation of an insertion or a drop of the rod cluster control assembly into its guides (discontinuous and continuous guides, guide thimble). A numerical model is elaborated in which many loads are taken into account: fluid load, gravity and friction force between the rod and the guide. The numerical results are compared to experimental measurements obtained from a full-scale structure. A good agreement between the calculated and the measured data is observed.The numerical model takes into account the possible deflection of the guide. It shows clearly that the friction force cannot be neglected when the guide is bowed. So one can locate a faulty guiding system by examining the reaction force during the insertion test. Then, the numerical model can help the decider to make his choice among different rod cluster/fuel assembly components.     

高温气冷堆控制棒缓冲器的分析与试验验证

高温气冷堆控制棒缓冲器是缓解极端事故工况下驱动线断裂产生控制棒跌落冲击的塑性变形吸能设备。利用经典圆柱壳轴压吸能模型设计出一种满足实际工况的薄壁壳结构,采用J-C本构模型,利用ABAQUS/explicit对该薄壁壳结构及控制棒建立有限元模型来模拟碰撞过程。基于分析结果,设计并进行了控制棒跌落试验。试验结果表明,全尺寸缓冲器受等重试验棒冲击后发生稳定渐进屈曲,有效缓冲了控制棒跌落冲击,其底部石墨结构未发生破坏,同时分析结果表明试验模型能包络代表实际反应堆中的跌落情况。     

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.     

Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 °C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 °C, respectively, at full power operation and at the scram from the operation.     

Analysis of RA-3 experiments for the validation of a neutronic model

A 3D neutronic model for the RA-3 reactor was developed on the basis of previous experience and validated with selected experimental data. Control rod calibrations were reproduced in N94 and N136 cores. The calculated values are shown to be dependent on relative position of the rods and the procedure that gives the best estimation of the rod value is the one performed following the experimental method compensating small rod insertions with small extractions. Rod worth calculations differ from the measured values in less than 2%. Rod-drop experiments were used to evaluate rod effectivities. The experimental results showed discrepancies between estimators derived from the point reactor model, and from spatial modal kinetics. Discrepancies are also observed when using different detectors. Even when using the spatial modal kinetics approach, the estimators obtained from different detectors disagree when one of them is located near to the rod, but differences are considerably reduced with respect to point reactor model because in this case only the delayed evolution is considered. We can say that all estimators give fairly similar results when the detector field of view is not influenced by the local perturbation introduced by the falling rod. This indicates the existence of spatial effects which are not completely accounted for in the spatial modal kinetics approach. Also, the importance of verifying the form function behaviour during the delayed evolution. The rod drop experiments were simulated using the improved quasi-static model and static evaluations. The rate of overestimation static/dynamic is constant in both core configurations and varies between 18% and 23% for the analyzed rods. The dynamic model allows comparing also thermal flux ratios at the detector positions. The neutronic model is considered reliable for design and fuel management analysis. The estimations of criticality, control rod calibrations and excess reactivity are satisfactory and simple models representing the core components without any kind of correction factors have been used to achieve these results.     

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

Earthquake vibrations cause large forces and stresses that can significantly increase the scram time required for safe shutdown of a nuclear reactor. The horizontal deflections of the reactor system components cause impact between the control rods and their guide tubes and ducts. The resulting frictional forces, in addition to other operational forces, delay the travel time of the control rods. To obtain seismic responses of the various reactor system components (for which a linear response spectrum analysis is considered inadequate) and to predict the control rod drop time, a non-linear seismic time history analysis is required. Nonlinearities occur due to the clearances or gaps between various components. When the relative motion of adjacent components is large enough to close the gaps, impact takes place with large impact accelerations and forces.This paper presents the analysis and results for a liquid metal fast rector system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% co-efficient of restitution.The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a 10 sec safe shutdown eathquake (SSE) acceleration-time history at 0.005 sec intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then these were used by the second program for the scram time determination.The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about four times longer than that calculated without the eathquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions.     

Development of a control rod depletion methodology for the Westinghouse NEXUS system

The NEXUS project is an effort to merge and modernize the methods employed in Westinghouse PWR and BWR steady-state reactor physics codes. The NEXUS system relies on a once-through nodal cross-section generation methodology with an innovative and efficient technique for pin power recovery. The pin power methodology overcomes a well-known limitation of existing methodologies, namely the incapacity to properly account for heterogeneity changes due to the depletion environment. The so-called control rod history problem where control rods are repeatedly inserted and withdrawn during core depletion is a good example of such a case. In addition to the control rod history impact on pin power distributions, the insertion of control rods during extended periods leads to significant control rod depletion that affects the reactivity worth of the control rods which in turn can have a significant impact on pin powers. The importance of accurately predicting pin powers, combined with the need to adequately estimate the reactivity worth and nuclear end of life of control rods in BWRs and in generation III+ PWRs, has motivated the development of a novel control rod depletion model. This methodology and its numerical qualification, initially for PWR application only, is the topic of this paper. The focus is on describing the salient features of the model and on illustrating its performance by means of numerical experiments. It is shown that together with the NEXUS pin power recovery model, the control rod depletion methodology accurately predicts the reactivity feedback from repeated control rod insertions in a PWR core.     

控制棒落棒时间分析软件CRAC的研发与验证

紧急停堆的落棒时间对反应堆安全至关重要。为适应华龙一号堆型的新型燃料组件设计,中国核动力研究设计院研制出一款落棒时间分析软件CRAC。采用一维流体力学公式结合经验机械阻力模型的方法,构建出CRAC软件理论框架,通过软件开发标准流程完成设计编码,并利用落棒试验数据开展了CRAC软件的验证。结果表明软件计算精度与保守性能满足华龙一号堆型安全停堆时间准则分析的需求。     

HTR-10控制棒系统的试验与调试

10MW高温气冷实验堆共有10套控制棒组件及其驱动机构，用于高温堆启动，功率运行和停闭过程中补偿和调节所需的反应性变化，并保证足够的停堆裕度，为确保达到工程设计的要求，对全部控制棒组件及驱动机构，都在实验室进行了热态试验，在高温堆上进行了安装后调试，以及首次临界前的测试，各项数据表明，所有驱动机构运行良好，控制棒的提棒，落棒运行功能正常，位置保持功能正常，棒位显示准确。     

A fast-running core prediction model based on neural networks for load-following operations in a soluble boron-free reactor

A fast prediction model for load-following operations in a soluble boron-free reactor has been proposed, which can predict the core status when three or more control rod groups are moved at a time. This prediction model consists of two multilayer feedforward neural network models to retrieve the axial offset and the reactivity, and compensation models to compensate for the reactivity and axial offset arising from the xenon transient. The neural network training data were generated by taking various overlaps among the control rod groups into consideration for training the neural network models, and the accuracy of the constructed neural network models was verified. Validation results of predicting load following operations for a soluble boron-free reactor show that this model has a good capability to predict the positions of the control rods for sustaining the criticality of a core during load-following operations to ensure that the tolerable axial offset band is not exceeded and it can provide enough corresponding time for the operators to take the necessary actions to prevent a deviation from the tolerable operating band.     

辐照对银铟镉控制棒价值的影响分析

为分析银铟镉（Ag-In-Cd）控制棒内各核素经反应堆中子辐照后的消耗情况以及核素消耗对控制棒价值的影响，本研究采用蒙特卡罗程序模拟了Ag-In-Cd控制棒内主要核素在反应堆运行期间的燃耗，并结合控制棒宏观中子吸收截面和控制棒内的中子注量率水平变化，分析了辐照前后控制棒价值的变化。研究结果表明，控制棒中113Cd随着辐照时间增加而加速消耗，107Ag、109Ag和115In消耗速率相对较慢；控制棒总的宏观中子吸收截面在辐照后降低，但是107Ag、109Ag和115In的中子吸收截面明显地增加；辐照后控制棒内的中子注量率增大，控制棒总中子吸收率无明显变化，即控制棒价值无明显变化。      

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。     

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。     

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.     

Integrity assessment of the high temperature engineering test reactor (HTTR) control rod at very high temperatures

The high temperature engineering test reactor (HTTR) is the first high temperature gas-cooled reactor (HTGR) in Japan with a reactor outlet coolant temperature of 950°C at high temperature test operation. The HTTR contains 16 pairs of control rods for which Alloy 800H is chosen of the metallic parts. Because the maximum temperature of the control rods reaches about 900°C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Under the guideline, temperature and stress analysis was conducted, and it is confirmed that the target life of the control rods of 5 years can be achieved.     

可动线圈控制棒电磁驱动线验证试验

可动线圈控制棒电磁驱动线是新研制的一种反应堆用驱动线,靠电磁力驱动控制棒实现反应堆的开堆、停堆和功率调节。为突破其研制过程中的关键技术,进行了一系列的验证试验。通过原理性能试验,验证了设备研制的理论基础,实现了可行性的突破;通过行程试验和寿命考验,验证了设备的稳定性、可靠性,并得到了设备的运行特性参数;通过抗震试验,验证了设备在极端条件下的安全功能。性能试验得到的试验数据,可为该驱动线的安装、调试、安全运行提供依据。     

Testing of Lwr Fuel Rods to Support Criticality Safety Analysis of Transport Accident Conditions

Abstract

For the transport of low enriched materials, criticality safety may be emonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where light water reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data are needed. This requires a method by which the response of LWR fuel under accident impact conditions can be approximated or bounded. In 2000, British Nuclear Fuels and Areva Cogema Logistics jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. ACL were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the transport regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and ACL would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasistatic loading on fuel samples. Pressurised water reactor (PWR) fuel rods loaded with uranium pellets were dropped vertically from 9 m onto a rigid target and this was repeated on boiling water reactor (BWR) fuel rods; similar tests on empty fuel rods were also conducted. Quasistatic tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high-burn-up fuel rods of both PWR and BWR types. These were believed to be original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500°C during loading. All specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data were gained from this test programme.     

ABSTRACTS OF JAPANESE ARTICLES

A new approximate method for calculating the effectiveness of multiple control rods fully inserted in a reactor is described. This method is appicable to a bundle of many control rods, regardless of the number of rods, as well as to an array of a limited number of few rods.

Using either the sink model or the well model, a reactor equation of kernel form is obtained. The reactor equation is a two-dimensional diffusion equation with two-group diffusion kernel. In order to facilitate numerical computation of the eigenvalue, the integral equation is reduced to a set of linear homogeneous equations, by dividing the reactor into a large number of unit cells containing at most one control rod.

This method has been programmed for the IBM 7090, the code being given the designation ELC. The iterative procedure used converges much faster than the standard accelerated finite-difference programs. Using the ELC code, the effectiveness of an array of four control rods fully inserted in a cylindrical reactor was calculated. The results are in good agreement with those found by the Scaletter-Nordheim method. In the case of a large number of control rods, there is no alternative method to be compared with.