PIUS - status and perspectives

Today's environmental concerns show that nuclear energy is an important option for meeting future increases in global energy demand. Significant nuclear expansion will probably require new reactor designs in which safety is ensured by simple, convincing means. PIUS represents such a reactor design. It is a re-configured 600 MWe PWR, in which the primary safety goal, protection of the reactor core integrity, is entrusted to built-in, self-protective, passive features, without reliance on any monitoring, detection or actuation system, nor operator action. Its basic design features a core that is openly connected, in a natural circulation loop, to a large pool of borated water. The pool is enclosed in a prestressed concrete pressure vessel provided with redundant leakage barriers. The reactor coolant pumps are operated in such a way that there is hydraulic balance in the openings between the primary coolant loop and the pool. Thereby, the hot, low boron content primary loop water is kept separated from the pool water, in spite of the always open natural circulation loop. In severe transients this balance is disturbed, and pool water ingress occurs, shutting down the reactor, or restricting the power to a safe level. The decay heat is transferred to the pool by the natural circulation loop, and a passive pool cooling system, utilizing natural circulation and natural draft cooling towers, prevents boiling of the pool water, even in a station blackout situation. Transient analyses have shown that this passive long-term RHR function will be available in all accident situations, even after double-ended cold leg breaks. Such breaks result in a temporary pressurization of the reactor containment, but the releases of radioactivity will be extremely small and the doses at the fence boundary very low. Cost estimates indicate that PIUS will be quite competitive, and evaluation studies are now under way in several countries.     

自然循环铅基快堆一回路系统驱动力的关键影响因素研究

为深入研究影响自然循环铅基快堆一回路系统驱动力的关键因素,以自然循环铅基快堆SNCLFR-10为研究对象构建描述反应堆一回路自然循环稳态运行模型;从理论上量化分析冷/热池的热量传递、热源和热阱温度非线性分布、反应堆压力容器壁散热3种因素对自然循环能力的影响,并开展了相关数值模拟验证。结果表明,数值模拟结果与本研究理论计算值吻合较好;3种自然循环能力影响机制耦合作用将降低SNCLFR-10系统自然循环能力,导致自然循环流量与功率之间不再满足理论所得的1/3次方关系。      

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.     

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.     

基于响应面拟合方法中国铅基研究实验堆非能动余热排出系统可靠性分析

非能动系统已广泛地应用于新一代堆的设计中,其可靠性分析成为新型反应堆概率安全评价(Probabilistic Safety Analysis,PSA)的重要内容。本文提出一种用于非能动系统可靠性分析的响应面拟合方法,并应用于中国铅基研究实验堆反应堆容器空气冷却系统(Reactor Vessel Air Cooling System,RVACS)的可靠性分析。采用流体计算软件Fluent模拟RVACS系统的输入输出作为求解响应面性能函数的输入样本,利用最小二乘法和bootstrap方法估计响应面性能函数的系数,以响应面模型代替Fluent模型分析RVACS系统的非能动失效概率。分析表明,在所有能动余热排除系统不可用的情况下,RVACS四组并联排热管中的两组也能够可靠地导出反应堆余热。RVACS系统可靠性高。     

中国先进研究堆停堆冷却措施及其主要特点

停堆后冷却问题是中国先进研究堆(CARR)重要的安全问题之一.冷却措施的实施对CARR的安全和建设投资有较重要的影响.CARR采用停堆初期的强迫循环及停堆后期全堆芯自然循环相结合的策略实现正常停堆和事故停堆后的堆芯冷却.停堆冷却的过程具体分为主泵大质量惯性飞轮惰转强迫冷却、应急堆芯冷却系统强迫冷却、自然循环功能部件动作实现全堆芯自然循环3个阶段.3个阶段既相互衔接又相互独立,每个阶段各有特点.停堆冷却策略的实施证明,CARR停堆冷却过程是可靠、有效、合理的,符合先进研究堆的发展趋势.     

FHR在全厂断电事故下对RVACS散热能力的要求

氟盐冷却高温堆(Fluoride salt-cooled High-temperature Reactor,FHR)是一种采用包覆颗粒燃料、高温熔融氟盐冷却剂的先进反应堆。部分FHR概念采用了反应堆容器辅助冷却系统(Reactor Vessel Auxiliary Cooling System,RVACS)导出事故下的堆芯余热。RVACS通过导热、对流换热、辐射换热等非能动过程,在事故发生时将堆芯余热排出至大气中。本文采用中国科学院上海应用物理研究所设计的10 MW FHR作为基准,利用RELAP5-MS程序,对其在全厂断电事故下的瞬态过程进行了模拟,验证了RVACS的余热导出能力。本文进一步研究了高反应堆功率情况下的全厂断电事故的瞬态过程,探讨了不同反应堆功率的FHR对RVACS散热能力的要求。     

EBR-Ⅱ余热排出实验及非能动余热排出系统性能分析

本文基于SAC-CFR事故分析程序,在国际原子能机构联合研究项目（IAEA CRP）框架下,对美国EBR-Ⅱ快堆余热排出实验（SHRT-17、SHRT-45R）进行了分析,计算了事故余热排出系统（DRACS）的响应、衰变热功率、关键部件的冷却剂温度、一回路的质量流量等关键参数。将计算参数与实验数据进行了对比,对程序的有效性进行了验证。计算结果表明,在SHRT-17工况下,随DRACS风门的打开,每台事故热交换器可带走330 406.4 W的堆芯余热,DRACS具有长期带走衰变热的能力。     

DHR-200池式低温供热堆SBO-ATWS事故及自然循环能力分析

DHR-200池式低温供热堆(简称DHR-200池式堆)设计有自然循环瓣阀,为检验其安全性,选取典型的全厂断电叠加紧急停堆系统失效(SBO-ATWS)事故,使用RELAP5程序对其热工水力参数瞬态特性及其自然循环能力进行分析。结果表明,DHR-200池式堆具有很好的负温度反应性反馈效应,即SBO-ATWS事故后,由于燃料和冷却剂温度升高,引入负反应性,可使反应堆实现热停堆;事故后,通过非能动方式开启自然循环瓣阀,可建立稳定的自然循环,将堆芯衰变热导出至堆水池内,验证了DHR-200池式堆的固有安全性。     

池式钠冷快堆事故余热排出系统一回路仿真研究

池式钠冷快堆事故余热排出系统采用了非能动工作原理,依靠液态钠及空气的自然对流排出堆芯余热。为研究事故工况下余热排出系统一回路的换热能力,基于FORTRAN语言,建立堆芯单通道及盒间流模型,采用全隐二阶迎风差分格式及改进的欧拉法离散求解,对事故余热排出系统一回路系统进行数值模拟,并对全厂断电事故进行仿真计算验证。结果表明:该程序能较好地反映事故余热排出系统瞬态变化过程,并可达到超实时仿真。     

Containment for the low temperature district nuclear-heating reactor

An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happened during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety.     

Natural circulation studies in a lead bismuth eutectic loop

Lead Bismuth Eutectic (LBE) is increasingly getting more attraction as the coolant for advanced reactor systems. It is also the primary coolant of the Compact High Temperature Reactor (CHTR), being designed at BARC. A loop has been set up for thermal hydraulics, instrument development and material related studies relevant to CHTR. Steady state natural circulation experimental studies were carried out for different power levels. Transient studies for start-up of natural circulation in the loop, loss of heat sink and step power change have also been carried out. An 1D code named LeBENC has been developed at BARC to simulate the natural circulation characteristics in closed loops. The salient features of the code include ability to handle non-uniform diameter components, axial thermal conduction in fluid and heat losses from the piping to the environment. This paper deals with the experimental studies carried out in the loop. Detailed validation of the LeBENC code with the experimental data is also discussed in the paper.     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

Analysis of SBO-ATWS Accident and Natural Circulation Capacity of DHR-200 Pool Type Low Temperature Heating Reactor

DHR-200 Pool Type Low Temperature Heating Reactor (DHR-200) was designed with natural circulation flap valve. In order to examine the safety of the DHR-200, the RELAP5 code was used to analyze the transient thermal-hydraulic characteristics and the natural circulation capacity under the station blackout anticipated transient without scram (SBO-ATWS). The results show that DHR-200 has enough negative temperature reactivity feedback effect. With the rising of the temperatures of the fuel and the coolant, finally the reactor can be shut down by the effect of the negative temperature reactivity feedback effect. After the accident, the natural circulation flap valve will be opened by passive means to establish a stable natural circulation, and then the residual heat of the core can be removed to pool of the reactor. Therefore, it is demonstrated that the DHR-200 has good inherent safety features.     

Core power limits for a lead-bismuth natural circulation actinide burner reactor

The Idaho National Engineering and Environmental Laboratory and Massachusetts Institute of Technology are investigating the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The design being considered here is a pool type reactor that burns actinides and utilizes natural circulation of the primary coolant, a conventional steam power conversion cycle, and a passive decay heat removal system. Thermal-hydraulic evaluations of the actinide burner reactor were performed to determine allowable core power ratings that maintain cladding temperatures below corrosion-established temperature limits during normal operation and following a loss-of-feedwater transient. An economic evaluation was performed to optimize various design parameters by minimizing capital cost. The transient power limit was initially much more restrictive than the steady-state limit. However, enhancements to the reactor vessel auxiliary cooling system for transient decay heat removal resulted in an increased power limit of 1040 MWt, which was close to the steady-state limit. An economic evaluation was performed to estimate the capital cost of the reactor and its sensitivity to the transient power limit. For the 1040 MWt power level, the capital cost estimate was 49 mills per kWhe based on 1999 dollars.     

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.     

Inherently safe pool-type reactor as a generator of low-grade heat for district heating, air conditioning and salt water desalination

The society has a heavy demand for low-grade heat to satisfy its various needs. Different factors govern the expediency of applying nuclear reactors for these purposes. The required capacity of heat sources varies in a very wide range. In a majority of cases heat sources have to be located in the immediate vicinity of the users, therefore, nuclear reactors to be used for heat generation must feature enhanced safety. Pool-type reactors can be successfully used for producing low-grade heat. Owing to their design they feature a very high safety level. The absence of positive pressure excludes the possibility of a sudden rupture of reactor tank (vessel) or a fast loss of coolant. The availability of a large amount of water in the tank ensures long-term accumulation of residual heat. The adopted integral layout of equipment, as well as natural circulation of primary coolant improve reactor reliability and safety even further. Negative temperature coefficients of reactivity provide for reactor self-protection against reactivity accidents. Pool-type reactors can be used in newly established heat supply systems and can be built in the operation systems as well, which allows to reduce fossil fuel consumption by 80–90% depending on local conditions. Pool-type reactor heat can be used for desalinating salt water and for cooling water in absorption refrigerating machines with subsequent utilization of cold water for air conditioning, cooling of special premises, and the like. Pool-type reactors can also generate electric power to their in-house needs as well as household power requirements of a neighboring town.     

Thermal-hydraulic analyses of transients in an actinide-burner reactor cooled by forced convection of lead–bismuth

The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.     

Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling. The MTRTHA model consists of five interactively coupled submodels for: (a) coolant, (b) fuel plate, (c) chimney, lower plenum, suction box and cold leg, (d) flap valve and (e) natural circulation flow. The model divides the active core into a specified axial regions and the fuel plate into a specified radial zones, then a nodal calculation is performed for both average and hot channels with a chopped cosine shaped heat generation flux. The reactor simulation under loss of off-site power is performed for two cases namely: two-flap valves open and one flap-valve fails to open. The simulation is performed under a hypothetical case of loss of off-site power. Unfortunately, the flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. In most cases, the flow inversion phenomenon is accompanied by boiling which is undesirable phenomenon in this type of reactors as it could affect the fuel-clad integrity. The model results for the flow inversion phenomenon prediction are analyzed and a solution of the problem is suggested.     

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m³/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.