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.     

Scram reliability under seismic conditions at the Experimental Breeder Reactor II

A level 1 probabilistic risk assessment of the Experimental Breeder Reactor 11 has recently been completed. Seismic events are among the external initiating events included in the assessment. The analysis indicated that the reactor shutdown system had a high reliability of operation in response to internally initiated events. One of the major tasks within the seismic assessment concentrated on the ability to shut down the reactor under seismic conditions. A comprehensive analysis of the shutdown system, including the development of a finite element model of the reactor control rod drive system, has been used to estimate the system response when subjected to input seismic accelerations. The results indicate the control rod driven system has a high seismic capacity and that the overall reactor shutdown system is capable of maintaining its high reliability under seismic conditions. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure.     

The experimental breeder reactor II inherent shutdown and heat removal tests — results and analysis

A test program is being conducted to demonstrate that a power-producing liquid-metal reactor (LMR) can (1) passively remove shutdown heat by natural convection, (2) passively reduce power in response to a loss of reactor flow, and (3) passively reduce power in response to a loss of the balance-of-plant heat sink. Measurements and pretest predictions confirm that natural convection is a reliable, predictable method of shutdown heat removal and suggest that safety-related pumps or pony motors are not necessary for safe shutdown heat removal in an LMR. Measurements from tests in which reactor flow and heat rejection to the balance of plant were perturbed show that reactivity feedbacks can passively control power and temperature. Data from these tests form a basis for additional tests including a complete loss of flow without scram and a complete loss of heat sink without scram.     

控制棒水压驱动系统落棒减速性能实验研究

控制棒水压驱动系统是清华大学为低温核供热堆发明的新型的内置式控制棒驱动技术,控制棒水力减速部件是水压驱动系统的关键部件之一,通过其对控制棒落棒过程进行减速,在保证落棒时间的前提下,降低控制棒快速落棒过程的冲击力。分析了水力减速部件组成和工作原理,确定了水力减速箱侧壁开孔方案,完成了不同开孔方案工况下控制棒水压驱动系统冷态落棒减速性能实验,在实验结果的基础上对比和分析了不同方案下的落棒减速机理和落棒过程特征参数随开孔方案的变化规律。分析结果表明:随开孔面积的增大,落棒时间逐渐减小,落棒峰值速度逐渐增大。在开孔面积大于0.004 m~2时,随开孔面积的增大,落棒峰值速度增大过程趋于平缓,落棒稳定速度和落棒延迟时间变化不大,控制棒触碰碟簧速度缓慢增大。实验研究成果为控制棒水压驱动系统落棒减速部件的理论建模和设计优化提供了基础。     

空间核动力装置控制鼓系统试验样机快速复位时间测量方法研究

控制鼓系统是空间核动力装置上执行功率调节、紧急停堆的重要安全级设备,其能否正常运行直接关系到核动力装置的安全性。为测试控制鼓系统的快速复位时间,通过分析控制鼓系统的转动和传动过程,提出了快速复位零点判断、计算复位时间的方法。采用1∶1全尺寸控制鼓系统试验样机和综合测试平台,对快速复位时间进行了实测试验。试验结果表明,该测试方法是真实、有效、可靠的,可应用于控制鼓系统各阶段研发、使用过程中快速复位时间的测量。同时也验证了控制鼓系统的设计满足设计指标,机械快速复位时间小于1 s。     

Study on Scram Time Measurement Method of Control Drum System Experimental Prototype of Space Nuclear Power Device

The control drum system is important safety level device for power regulation and emergency shutdown of the space nuclear power device. The safety of the reactor directly depends on whether the control drum system can operate well or not. In order to test the scram time of the control drum system, a method of judging the zero position and calculating the scram time was proposed by analyzing the rotation and transmission process of the control drum system. The full-scale model of the control drum system experimental prototype and a comprehensive test platform were used to test the scram time. The test results show that the test method is real, effective and reliable, and can be applied to the measurement of scram time in the development and use of control drum system at all stages. At the same time, it is also verified that the design of the control drum system meets the design target that the mechanical scram time is less than 1 s.     

控制棒水压驱动系统落棒减速部件三维流场分析

控制棒水压驱动系统是清华大学为低温核供热堆研制的新型内置式控制棒驱动技术,控制棒水力减速部件是水压驱动系统的关键部件之一,在保证落棒时间的前提下,通过其对落棒过程进行减速,降低控制棒快速落棒过程的冲击力,避免控制棒十字翼的变形和损坏。本文分析了控制棒水压驱动系统落棒减速机理,利用CFD软件FLUENT对驱动系统水力减速箱流道进行了三维流场数值分析,并分析了对应不同落棒位置水力减速箱流道在不同边界条件下的流场分布特性。在流场分析结果的基础上计算得到了水力减速箱侧壁孔流道和底部缓冲腔流道流量系数随落棒位移的变化,将该结果与驱动系统落棒减速理论模型联立,获得了控制棒落棒位移曲线,理论计算结果同冷态落棒性能实验结果符合得很好,从而验证了流场分析结果的正确性,在此基础上分析了落棒过程减速箱内外差压和落棒速度与水力减速箱流量系数的关系,为控制棒水压驱动系统落棒减速部件的设计和优化提供了指导。     

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.     

Testing and qualification of Control & Safety Rod and its drive mechanism of Fast Breeder Reactor

Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control & Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements.CSR & CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot & dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism.In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM & CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM & CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO₂ process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.     

Impacts and fluid-structure interaction in pressurized water reactor safety analysis

A numerical method is described for the analysis of coupled three-dimensional fluid-structure motion with impacts between structural parts at rigid or flexible supports with small clearances. The method is used for the analysis of the blowdown loadings and the response of internal structures in the vessel of a pressurized water reactor (PWR) in the hypothetical event of a sudden break of a coolant inlet pipe. The method is a generalization of the existing code FLUX which simulates the three-dimensional fluid-structure motion by means of an implicit time integration scheme. The additional supports with clearances are taken into account by applying support forces to the freely moving fluid-structure system. The forces are determined such that the kinematic constraints are enforced at each time step. Numerically, these forces are determined efficiently using a precomputed influence matrix which defines the dynamic displacement per time step at each support due to a unit force at each other support. According to the actually “active” supports the relevant influence matrix in constructed. Energy is conserved for rigid supports and for supports which are so flexible that the impact time is large in comparison to the time steps. Treatment of plastic supports is possible.An application of the new method is demonstrated by analysis of the core barrel motion in a PWR with and without impacts at the lower core barrel edge and at the upper flange. The results show the large effects of such impacts in changing the global motions. Large local impact forces and accelerations appear. The interaction with the fluid reduces these loads. By proper design of the supports the resultant stresses can be minimized. Thus the method can be used to demonstrate and enlarge nuclear reactor safety.     

Lead-cooled flexible conversion ratio fast reactor

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO₂ (S-CO₂) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO₂ PCS.     

钠冷快堆停堆保护方案研究

在借鉴中国实验快堆(CEFR)热工模型建模经验的基础上,利用Relap5程序建立霞浦示范快堆(CFR)的主要系统模型,并参考快堆安全分析中的预期瞬态无停堆保护(ATWS)的分析方法,对发生反应性意外引入事故时的安全裕度和停堆保护进行仿真研究。仿真结果表明,额定功率下发生反应性引入时,不会触发短周期的报警和停堆;当发生补偿棒失控提升5 s和10 s时的反应性意外引入事故,目前一回路保护参数整定值、信号测量延迟及安全棒落棒时间可以取其他值;当补偿棒失控提升15 s时,在目前的设计下,核功率和功率流量比信号能确保事故下的反应堆状态符合事故验收准则。当其他保护信号失效,堆芯出口钠温所触发的停堆保护若要实现同样的功能,则需保证反应堆在14.85 s之前进入深度次临界。     

机组临时停机停堆期间的化学监督与控制

目前的化学与放射化学程序和措施,都是针对正常功率运行和按部就班有计划的大修状态而设置,如遇到机组跳堆、跳机或冷停堆等紧急情况,则没有相应的应急预案或相关程序进行提前或有目的地干预。基于这种情况,电厂化学人员经过多年的实践和不断经验反馈,总结并编写了专门针对紧急停机停堆的化学监督与控制应急预案。通过停堆过程和停堆后的不同状态,启机过程的化学与放射化学监测,监督燃料包壳状态,控制一回路的剂量水平,以防止设备腐蚀。     

Analysis of impact induced accelerations on internal components of light water reactor packages

Abstract

When designing and then licensing a package for the transport of light water reactor fuel, it is normal practice to demonstrate impact performance by conducting drop tests at orientations determined to be the most severe. Usually, accelerometers are fitted to the package during the impact testing so that data may be applied in supporting stress analysis. In most cases the accelerometers are fitted to the external surfaces of the package while the data so obtained is frequently applied to the study of internal components. However, this approach is frequently challenged on the basis that internal accelerations could be different to those measured on the outside of the package, perhaps even higher! Accordingly, International Nuclear Services commissioned a theoretical study looking at a range of accelerations, as measured on the package body and compared these to accelerations on the fuel. This concluded that, with certain parameters acting, accelerations experienced by the fuel could indeed be higher than measured on the package. However, it was more likely that accelerations on the fuel would be lower and of longer duration. The present study demonstrated that there is no simple answer to this issue, nevertheless there is clearly potential for a package designer to minimise impact accelerations on the fuel by considering the fuel basket stiffness and internal clearances in conjunction with package impact characteristics.     

间接氦气透平循环高温堆紧急停堆动态仿真

通过对10 MW高温气冷堆氦气透平发电装置(HTR-10GT)的堆芯、热交换器和透平压气机组等主要设备的数学建模和程序编制,初步建立起了一套模拟该装置瞬态特性的仿真程序.通过对该装置于5s时刻堆内引入0.1$阶跃正反应性引发的紧急停堆事故的瞬态模拟,初步验证了该装置紧急停堆预案设置的安全性和合理性,证明了旁路快开阀的设...     

Identification of damping mechanism of TRR-II reactor control rod during free fall insertion

In light water reactors, control rods are in general inserted into reactors by gravity. In order to achieve a rapid shutdown, it is required to insert control rods as fast as possible. On the other hand, a control rod with a fast falling velocity would impose a substantial impact to reactor structure as well as to the rod itself. Therefore, a damping force must come into effect, especially during the final stage of the free fall of the control rod. The purpose of this study is to develop a mathematical model and a numerical simulation to describe and identify the damping mechanism; and apply this model to the design of the control rod used in TRR-II reactor of the Institute of Nuclear Energy Research (INER) of Taiwan.The damping effect of a falling control rod comes from two factors: the viscous shear stress occurred in a narrow gap between the rod and an outer tube which confines the lateral movement of the rod, and the pressure force exerted on the rod by the compressed water under the rod. The viscous shear stress can be analyzed by assuming a couette flow between the rod and the outer tube similar to the viscous force occurred in rheology. In doing this, the flow rate in each flow path is closely related to the pressure gradient in the flow path and can be evaluated using an electrical circuit analogy. The results of the code prediction were compared to the experimental results as carried out by the INER. Finally, a parametric study was applied to estimate the effects of the various factors including gap thickness, size of the flow holes, and other geometric considerations on the rod falling velocity. The results of this study can serve some technical support during the stage of rod design and manufacture.     

Implications of the EBR-II inherent safety demonstration test

Extensive thermal-hydraulics testing at EBR-II culminated in the Inherent Safety Demonstration Test on April 3, 1986. This work may well lead to fundamental changes in the approach to the design and licensing of liquid-metal-cooled reactor (LMR) power plants. The EBR-II test program has thus far demonstrated (1) passive removal of decay heat by natural circulation, (2) passive reactor shutdown for a loss of flow without scram, and (3) passive reactor shutdown for a loss of heat sink without scram. Supporting analyses indicate that these characteristics can be incorporated into larger commercial LMRs and be used as the basis for a totally new passive control strategy. Analyses and tests are now in progress to show that LMRs with these characteristics and the passive control strategy are also inherently safe for unprotected overpower accidents.     

Safety characteristics of the multipurpose fast reactor (MPFR)

The safety characteristics of a long-life multipurpose nuclear reactor (MPFR) with self-sustained liquid metallic fuel and lead coolant, which is proposed to meet the requirements for the energy production in the future, were investigated. The application of liquid plutonium–uranium metallic alloys used as a nuclear fuel demonstrated high potential to reach excellent reactor shutdown characteristics against anticipated transients without scram such as unprotected loss-of-flow and unprotected transient overpower. The calculations indicated that the thermal expansion of liquid fuels would cause the negative reactivity insertion that would be larger in magnitude than any other thermally induced reactivity changes. This created the reactivity balance for the passive shutdown and power stabilization capabilities of the MPFR core. It was found that MPFR satisfies such design characteristics to be a potential candidate providing the replacement of fossil fuels by alternative energy sources in the next century.     

Investigating the execution of EOPs in LOTDFP + ATWS + LOCA scenarios for an ABWR using MAAP code

The effectiveness of the execution of emergency operation procedures (EOPs) for an advanced boiling water reactor (ABWR) during postulated accident conditions using MAAP 4 code is discussed in this paper. The simulation scenarios included the loss of turbine driven feedwater pump (LOTDRFP), the anticipated transients without scram (ATWS), and the loss of coolant accident (LOCA). Based on the comparisons of responses on different parameters for cases with and without EOP actions, we concluded that the EOPs could effectively mitigate the consequences of the accidents. In addition, the emergency depressurization (ED) timing and the times spent between executing the EOP steps were also considered. The simulation results clearly reveal that both the earlier execution of ED and the decrease of times spent between each EOP step could delay the boron injection and leave the operator ample time to take some other remedy actions for reactor safe shutdown.     

Self-Limiting Power Excursion Characteristics of Light Water Reactor, (I)

The Doppler limited power excursion characteristics of a light water reactor and the shutdown mechanism by scram were analyzed on the Hitachi Training Reactor (HTR). For the purpose of the pulse operation tests, modifications were applied to the HTR to provide pulsing capability; a pulse rod was added, together with a back up device for shutdown, and provision of three instrumented fuel assemblies, equipped with thermocouples; the Al-clad fuel rods were replaced by stainless steel clad rods.

About 100 runs of pulse operation tests were performed in fullest security with reactivity insertions ranging up to 1.0 % Δk/k, in which last case the peak power reached 38 MW, with a reactor period of 29 msec.