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.     

Demonstration of Control Rod Holding Stability of the Self Actuated Shutdown System in Joyo for Enhancement of Fast Reactor Inherent Safety

Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large-scale liquid metal cooled fast breeder reactor (LMFBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor Joyo MK-III. The rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation.     

Performance evaluation of an advanced integral reactor against an anticipated transient without scram

Performance evaluation of KAERI’s advanced integral reactor against an anticipated transient without scram has been carried out with the transients and setpoint simulation/small and medium reactor code, by considering a decrease in the heat removal by the secondary system, a loss of offsite power and an inadvertent control rod withdrawal event as an initiating event. In a decrease in the heat transfer by the secondary system and a loss of offsite power, the reactor coolant system pressures can be maintained below 110% of the design pressure during the transition period due to the effect of the large negative moderator temperature coefficient. On the other hand, in an inadvertent control rod withdrawal event, the pressure of the reactor coolant system increases up to the ASME service level C stress limit due to a high reactivity insertion into a reactor core by the adoption of a boron free core concept. Therefore, a hardware installation against an anticipated transient without scram is essential to mitigate the consequences resulting from an inadvertent control rod withdrawal event. A diverse protection system, which is an independent and diverse reactor shutdown system that is initiated by the signals of a high core power or a high pressurizer pressure, is adopted in the advanced integral reactor. According to the reassessment results by considering the diverse protection system for a reactor shutdown, the diverse protection system is helpful in mitigating the consequences of an anticipated transient without scram.     

控制棒驱动机构备件数量优化方法

控制棒驱动机构是反应堆控制和保护系统的伺服机构,是执行反应堆功率调节、紧急停堆的重要核安全设备。控制棒驱动机构成本较高,如何合理确定其备件数量对于提高反应堆的可运行性具有重要意义。本文针对控制棒驱动机构,在系统连续运转时间不小于换料周期的约束条件下,提出了一种确定控制棒驱动机构备件数量的优化方法--分组备件数量优化方法,给出了总费用最少的各子系统的备件配置方案。通过随机模拟计算对分组备件数量优化方法与常规算法进行比较,结果表明：该方法常优于常规算法,在保证控制棒驱动机构可用性的前提下,优化备件数量配置可降低成本。该优化方法同样适用于其他设备的备件分析,对工程中设备备件的分析与研究具有指导意义。     

10MWt固态燃料熔盐堆控制棒失控抽出事故分析

钍基熔盐堆核能系统项目是中科院先导科技专项之一,其战略性目标是研发第四代熔盐冷却裂变反应堆核能系统。基于10 MWt固态燃料熔盐堆的系统设计,开发了适用于球床式反应堆系统的安全分析软件,并以高温气冷堆为对象对程序计算结果的准确性进行了验证。基于该软件程序,对固态燃料球床堆(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF)控制棒失控抽出事故进行了分析计算,研究了不同停堆限值及各停堆信号对事故的影响。计算结果表明,超功率停堆限值越高,出口温度限值越大,信号延迟时间越长,反应堆停堆越晚,堆芯功率和燃料最高温度越高。在TMSR-SF控制棒失控抽出事故下,燃料最高温度不超过860°C,远低于1 600°C的熔化温度限值。     

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。     

兆瓦级空间气冷堆系统启停堆瞬态分析

本文针对兆瓦级高温气冷堆布雷顿循环系统,采用Fortran语言开发系统分析程序TASS,包括堆芯、透平-发电机-压气机、回热器、冷却器和热管式辐射散热器等模型。通过设计值与程序计算值对比对TASS进行验证,并利用TASS对系统启动、停堆瞬态工况进行数值模拟。结果显示,通过分两阶段、阶梯式引入正反应性和提高涡轮机械的转轴速度,堆芯流量和功率匹配良好,系统可在3.5 h内完成启动过程,达到反应堆功率3 406 kW、流量14.2 kg/s的稳态运行。系统停堆过程中,反应堆可依靠自身的非能动余热排出能力,确保芯块和包壳温度与熔点间存在较大安全裕量,实现安全停堆。     

Passive cooldown performance of a 65 MW integral reactor

Advanced integral-type pressurized water reactor with a maximum thermal power of 65 MW is under development at the Korea Atomic Energy Research Institute (KAERI). This 65 MW integral reactor incorporates a number of innovative design features. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the TASS/SMR code and a sensitivity analysis has been performed to evaluate a passive cooldown capability for various system conditions such as natural and forced circulation conditions for the reactor coolant system or the passive residual heat removal system, and number of active PRHRS trains. The reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation and the 65 MW integral reactor can cool the coolant to the SCS entry condition in the primary system for all the possible operational conditions.     

Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value.     

Control rod worth calculations using deterministic and stochastic methods

Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refuelling, etc.) as well as several operational needs, such as calibration of the control rods, e.g. in case that reactivity insertion experiments are planned. The control rod worth can be assessed either experimentally or theoretically, mainly through the utilization of neutronic codes. In the present work two different theoretical approaches, i.e. a deterministic and a stochastic one are used for the estimation of the integral and the differential worth of two control rods utilized in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system SCALE (modules NITAWL/XSDRNPM) and CITATION is used, while the stochastic one is made using the Monte Carlo code TRIPOLI. Both approaches follow the procedure of reactivity insertion steps and their results are tested against measurements conducted in the reactor. The goal of this work is to examine the capability of a deterministic code system to reliably simulate the worth of a control rod, based also on comparisons with the detailed Monte Carlo simulation, while various options are tested with respect to the deterministic results’ reliability.     

混合能谱超临界水冷堆反应性引入事故安全分析

本文研究了混合能谱超临界水冷堆(SCWR-M)在发生控制棒失控提升事故和弹棒事故这两类反应性引入事故后的反应堆系统响应。首先利用修改的可用于超临界条件下的系统程序RELAP5对混合能谱超临界水冷堆进行系统建模,并计算分析在功率运行工况下事故过程中功率、流量及包壳温度等重要参数的变化趋势,最后对反应性参数如控制棒价值、控制棒抽出速率和负反馈系数进行了参数效应分析。结果表明,在设计工况下混合能谱超临界水冷堆系统可有效地将衰变热导出堆芯,保证了燃料棒的完整性。另外,反应性参数对控制棒失控提升事故的安全性影响不大,但对弹棒事故的包壳峰值温度影响很大,过于保守的反应性参数估计会使安全裕量大为减小。     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

Neutronics analysis of megawatt-class gas-cooled space nuclear reactor design

ABSTRACT

Neutronics analysis was conducted for a proposed megawatt-class gas cooled space nuclear reactor design. The reactor design has a high operating temperature of up to 1500 K. Annular UO₂ fuel rods were used to reduce the central temperature of the fuel. The thermal power is 2.3 MWt and is converted into electric power by a direct Brayton cycle. The control rods were arranged in different configurations and were analyzed in order to evaluate the influence on the reactor design in different scenarios. The calculation results reveal that the control rods arrangements have influences on the begin-of-life (BOL) excess reactivity and the shutdown reactivity. The distribution of control rods affects the neutron economy and leakage in the fuel region, consequently affecting the reactivity. It is also known that the reactivity in flooding scenarios are not sensitive to different control rod arrangements. Meanwhile, according to calculation results, the proposed reactor design has enough shutdown reactivity margin which will allow for flexible control strategy. Further analysis is still needed for more detailed and accurate parameters of the reactor design.     

反应堆控制棒驱动机构模拟负载系统的研制

介绍了反应堆控制棒驱动机构(CRDM)模拟负载装置的设计原理和方法,研制出了一种新型的模拟负载系统,用来模拟反应堆棒控系统对控制棒的控制过程。对设计的模拟负载系统进行了功能性试验和性能参数测试,并与实际运行系统进行比较后,发现该系统达到了各项功能控制要求,且性能稳定可靠,模拟负载的电磁线圈散热性能与负载特性良好,各项性能指标达到了设计要求。     

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。     

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.     

泳池式反应堆自启停控制技术研究

反应堆实现自动启停,可以有效减轻运行人员工作强度,减少误操作,提高反应堆启动运行的安全可靠性。本文基于对典型泳池式反应堆的工艺特点以及启动操作的分析,对泳池式反应堆自启停系统的控制范围、层次结构、断点、典型控制逻辑进行研究,并搭建泳池式反应堆自启停的仿真测试系统。该自启停系统能够实现泳池式反应堆的自动启停,启停过程无人工操作,降低人员误操作可能性。     

安全棒系统试验样机冷、热态性能试验研究

安全棒系统是空间核反应堆的关键设备之一,它具有结构紧凑、传动精度高、与反应堆容器连接接口多、工作温度高等特点。通过采用全尺寸的安全棒系统试验样机,确定了冷、热态性能试验方案,设计了专用的试验装置开展冷、热态性能试验。试验结果表明,安全棒系统试验样机运行正常,性能达到设计要求,为试验样机的抗震试验提供了条件,也为安全棒系统后续设计及试验装置的改进提供了参考依据。     

Design study of the demonstration FBR in Japan

The objective of the plant design study Phase 2, conducted by the Japan Atomic Power Company since 1997 for 3 years, is to accomplished a plant overall concept of the Demonstrative FBR (DFBR) that has economical potential toward commercialization and offers high reliability to plant operators not to cause a long unexpected shutdown resulting from a trouble, i.e., sodium leakage or fires. This has been successfully achieved by establishment of a plant overall design of 672 MWe consisting of the reactor system with drastically simplified internals, the compact and double walled coolant boundaries, the well rationalized fuel handling system, the BOP systems introducing up-to-date LWR equipment, and the compact reactor building.

The plant construction cost has been estimated based on the quantity of materials to be about 130 % on the bases of a 1000 MW LWR, which is well contented with the requirement.

The DFBR plant concept, having economical potential toward commercialization, safety and reliability, has been established in the plant design study Phase 2.     

Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes

A very complex type of power instability occurring in boiling water reactor (BWR) consists of out-of-phase regional oscillations, in which normally subcritical neutronic modes are excited by thermal-hydraulic feedback mechanisms. The out-of-phase mode of oscillation is a very challenging type of instability and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, simulations of out-of-phase instabilities in a BWR obtained by assuming an hypothetical continuous control rod bank withdrawal are being presented. The RELAP5/Mod3.3 thermal-hydraulic system code coupled with the PARCS/2.4 3D neutron kinetic code has been used to simulate the instability phenomenon. Data from a real BWR nuclear power plant (NPP) have been used as reference conditions and reactor parameters. Simulated neutronic power signals from local power range monitors (LPRM) have been used to detect and study the local power oscillations. The decay ratio (DR) and the natural frequency (NF) of the power oscillations (typical parameters used to evaluate the instabilities) have been used in the analysis. The results are discussed also making use of two-dimensional plots depicting relative core power distribution during the transient, in order to clearly illustrate the out-of-phase behavior.