The inherent safety characteristics of an unconventional LMR design—The Trench Reactor

This paper presents the transient behavior during off-normal operation of an unconventional liquid metal reactor design, called the Trench Reactor. Under the postulated accident conditions, this reactor design responds in an inherently safe manner to loss of heat sink accidents, loss of flow accidents, overcooling accidents and transient overpower accidents with 25 cents of reactivity insertion. The characteristics that cause such inherently save behavior are the properties of the materials and the configuration of the reactor primary system, even without any activated safety devices.     

Best Estimated and Conservative Analyses for a Feedwater Pipe Break Accident of an Integral Type Reactor

An integral type reactor, which is an innovative design to achieve a high degree of safety, is currently being developed at the Korea Atomic Energy Research Institute. A feedwater pipe break accident is one of the most important accidents regarding the safety of an integral type reactor. A best estimated calculation, a conservative calculation, and a parameter study for a feedwater pipe break have been carried out. The sensitivity analysis in this paper performed is to establish the parameters which greatly affect the feedwater pipe break accident. A power level, an initial system pressure, a moderator reactivity coefficient and a break size are the major parameters which maximize a system pressure. The important function that must operate following a feedwater pipe break accident is an opening of the pilot operated safety relief valves, and an initiation of the passive residual heat removal system. The integral reactor safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters.     

10 MW高温气冷实验堆事故分析的结果与对策

1 0MW高温气冷实验堆 (HTR 1 0 )的事故分析表明 ,在设计基准事故和严重事故条件下 ,HTR 1 0的堆芯燃料元件的最高温度和反应堆冷却剂系统的压力都低于规定的安全限值 ,燃料元件和冷却剂系统压力边界都能保持其完整性 ,不会造成裂变产物大量向外释放。根据事故分析结果并参照国外高温气冷堆安全运行的管理实践经验 ,针对HTR 1 0所提出的一系列事故对策有效地保证了HTR 1 0在较高的安全水平上进行设计、建造、运行及管理等 ,能够确保HTR 1 0、人员、社会以及环境的安全     

Thermal-hydraulic behavior of the NUR nuclear research reactor during a fast loss of flow transient

Nuclear safety analysis remains of crucial importance for both the design and the operation of nuclear reactors. Safety analysis usually entails the simulation of several selected postulated accidents, which can be divided into two main categories, namely reactivity insertion accident (RIA) and loss of flow accident (LOFA). In this paper, thermal-hydraulic simulations of fast LOFA accident were carried out on the new core configuration of the material test research reactor NUR. For this purpose, the nuclear reactor analysis PARET code was used to determine the reactor performance by calculating the reactor power, the reactivity and the temperatures of different components (fuel, clad and coolant) as a function of time. It was observed that during the transient the maximum clad temperature remained well below the critical temperature limit of 110 °C, and the maximum coolant temperature did not exceed the onset of nucleate boiling point of 120 °C. It is concluded that the reactor can be operated at full power level with sufficient safety margins with regard to such kind of transients.     

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10^?5 and 1×10^?6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.     

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT

This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria).

The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor.

The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior.

3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.     

超临界二氧化碳模块化微型堆瞬态安全分析

超临界二氧化碳反应堆是一种极具潜力的新堆型,目前正处于概念设计阶段。本文以韩国科学技术院（KAIST）设计的超临界二氧化碳模块化微型堆（MMR）为研究对象,对一回路系统主要部件进行建模,并利用FORTRAN语言开发了适用于超临界二氧化碳反应堆的瞬态安全分析程序TRA_SCR。基于该程序,对KAIST MMR进行了稳态计算分析,验证了程序的正确性。同时,对部分无保护失流事故和无保护反应性引入事故进行了瞬态计算,获得了关键热工水力参数的瞬态特性。计算结果表明该反应堆系统具有较强的固有负反馈特性,且在所计算的事故中,包壳、燃料和冷却剂温度均未超出安全限值,表明了系统在上述事故下的安全性。但在上述无保护失流事故中,堆芯冷却剂出口温度接近安全限值,表明在该事故工况下,反应堆出口温度是制约系统安全性能的关键因素。     

Development of critical heat flux correlation for In-vessel retention

ABSTRACT

In-vessel retention (IVR) is a strategy for severe accident management in which the lower head of the reactor vessel is submerged in a water-flooded reactor cavity. Critical heat flux (CHF) data for IVR are important for estimating cooling capacity of the reactor vessel. The existing CHF data for IVR which were obtained for the specific geometries and thermal-hydraulic conditions of actual plants are difficult to be applied to plants with other specifications. Hence, the purpose of this study is to develop CHF correlations applicable to various pressurized water reactor plants in a wide range of thermal outputs based on newly obtained CHF data. A rectangular test section with a cross-section of 150 mm × 150 mm and length of 600 mm was used for simulating a cooling channel. The thermal-hydraulic conditions expected in actual plants were studied, and the results were used in the experiment. The effects of parameters such as pressure, mass flux, thermodynamic quality, and angle on CHF were investigated . Based on these results, we developed a CHF correlation formula that can be applied to a wider range than previously, up to a maximum heat flux of 3000 kW/m², and that predicts CHF with an error of ± 10%.     

自然循环反应堆NHR-200Ⅱ失水事故初步研究

破口事故是压水堆最为关注的一类重要事故,其失水量与事故后果严重程度密切相关。NHR-200Ⅱ是由清华大学核能与新能源技术研究院经过多年研究和不断改进,设计的一种全功率自然循环低温供热反应堆,其设计中采用了多种先进的非能动和固有安全设计。本研究针对NHR-200Ⅱ反应堆,选取后果最为严重的控制棒引水管断裂且无法隔离事故,利用系统热工瞬态分析程序对事故过程进行了模拟和分析。结果表明,即使在最严重的破口失水事故下,NHR-200Ⅱ主回路中剩余的冷却剂始终能覆盖反应堆堆芯,并有效通过非能动余热载出系统带走堆芯热量,从而保证反应堆堆芯不会因裸露造成烧毁,这表明NHR-200Ⅱ具有很好的安全特性。     

Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor (JPSR)

A probabilistic safety assessment (PSA) technique was applied to the design of JAERI Passive Safety Reactor (JPSR). A PSA was performed to clarify safety features and identify vulnerabilities of the original design. Based on the PSA results and considering thermal-hydraulic analyses and experiments, the JPSR design was improved to enhance plant safety. The improved design was re-evaluated with the PSA. Initiating events selected in this study were: large-break LOCA, medium- and small-break LOCAs, SGTR, main steam line break, loss of offsite power, loss of feed water, and other transients. Fault tree analyses were used to evaluate the system unavailabilities. The total core damage frequency due to internal events was estimated to be less than 10^?7/RY. The contribution of high frequency non-LOCA events could be significantly reduced by the design modification. The dominant initiating event was the small break LOCA and the dominant sequence was the failure of residual heat removal system. The present study indicated that the improved JPSR design has sufficient safety margin and the PSA methodology is very effective to improve reactor safety systems in a conceptual design phase.     

Neutronic and thermal hydraulic design of the graphite moderated helium-cooled high flux reactor

A new design concept for a high flux reactor was investigated, where a graphite moderated helium-cooled reactor with pebble fuel elements containing (²³⁵U, ²³⁸U)O₂ TRISO coated particles was taken as its design base. The reactor consists of an annular pebble bed core, a central heavy water region, and inner, outer, top, and bottom graphite reflectors. The maximum thermal neutron flux in its central heavy water region as high as 10¹⁵ cm⁻² s⁻¹ with thermal neutron spectral purity of more than two orders of magnitude and a large usable volume of more than 1,000 liters can be achieved by (1) diluting the fissile material in the core and (2) optimizing the core-reflector configuration. The in-core thermal-hydraulic analysis was done to obtain adequate information about the coolant flow pattern and pressure distribution, core temperature profile, as well as other safety aspects of the design. To protect the reactor during off-normal or accident events, a large margin of temperature difference between the operating fuel temperature and the fuel limit temperature is confirmed by lowering the coolant inlet and core rise temperatures.     

Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO₂ rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.     

基于MOOSE框架的五方程两相流分析程序开发

基于多物理场耦合框架MOOSE,采用五方程两相流模型开发了模块化程序ZEBRA,实现了高阶时间、空间离散格式两相流动传热问题的求解。采用Bartolomei开展的垂直圆管过冷沸腾实验对ZEBRA进行验证,在不同热流密度、质量流密度、压力工况下,将程序计算值与实验值进行了数值验证和计算分析。结果表明：ZEBRA中五方程模型预测值与实验值符合良好,沸腾起始点和空泡份额的预测合理,表明ZEBRA初步具备了处理两相流问题的能力。     

Effect of Delays in Afterheat Removal on Consequences of Massive Air Ingress Accidents in High-Temperature Gas Cooled Reactors

Massive ingress of air into the core of a high-temperature gas cooled reactor is among the accidents with a low occurrence frequency, but there are still gaps in understanding with respect to its consequences. In the present paper, massive air ingress combined with a delayed start of the afterheat removal system is investigated and compared to air ingress, accidents with normal operation of the afterheat removal procedure. A computer programme REACT/THERMIX used for these accident analyses is described. For a high-temperature gas cooled reactor with a pebble bed core, it is shown that massive air ingress has no real safety endangering consequences even if the operation of the afterheat removal system is delayed by 6 h.     

Safety evaluation on the depressurization accident in the gas turbine high temperature reactor (GTHTR300)

Japan Atomic Energy Agency has been developing a gas turbine high temperature reactor (GTHTR300) with electric power of approximately 300 MW. One of the unique safety design concepts of this system is that events with frequency of occurrence of higher than 10⁻⁸/reactor-year are evaluated as design basis events in order to show that the frequency of large amount of FP release is less than 10⁻⁸/reactor-year. According to this concept, a depressurization accident by a large break of helium piping is postulated as a design basis event. This accident is one of the most serious accidents in the high-temperature gas-cooled reactors from the viewpoint of loss of coolability. The safety evaluation on the accident was conducted based on the actual design of the system. The short-term and long-term behaviors of fuel temperature after occurrence of the accident, internal pressure of the reactor building, oxidation behavior of fuels and graphite structures were evaluated and exposure dose of general public was also estimated using the results of evaluation of fuel temperature and fuel failure by oxidation. All of the evaluation results meet the safety criteria and feasibility of the GTHTR300 was shown by this study.     

NHR-200Ⅱ型低温供热堆安全特性

为提高低温供热堆的经济性,实现其供电、供气、海水淡化以及供暖的多用途目标,其主要热工参数须大幅提升。因此,提出一种新的低温供热堆堆型NHR-200Ⅱ,相比于NHR-200,其热工参数须大幅提升,同时又必须保持低温供热堆系统简化、固有安全性好的特性。为达到这一目标,沿用了低温供热堆一体化、全功率自然循环、自稳压以及非能动安全系统的设计特点,通过挖掘潜力、合理匹配系统参数来提高效率。对两种设计扩展工况的分析表明非能动安全系统的设计是有效的,反应堆堆芯不会发生裸露。本研究成果也可为其他小型水冷堆的设计提供借鉴。     

船用堆控制棒失控抽出事故时堆芯安全特性研究

采用船用堆三维动态安全分析仿真软件对发生控制棒失控抽出事故时堆芯安全特性进行了仿真分析,研究了反应堆分别处于高、低功率运行工况下1组或1束控制棒以不同的速率失控抽出时堆芯燃料芯块中心最高温度、最小烧毁比和冷却剂出口温度等参数的变化规律,并进行了比较,得出了一些有益的结论,对于考察反应堆安全状况和事故发生后反应堆操纵人员制定安全措施具有重要的指导意义。     

Development of a reduced order model for severe accident analysis codes by singular value decomposition aiming probabilistic safety margin analysis

ABSTRACT

Severe accident codes (e.g. MAAP, RELAP, and MELCORE) model various physical phenomena during severe accidents. Many analyses using these codes for safety margin evaluation are impractical due to large computational costs. Surrogate models have an advantage of quickly reproducing multiple results with a low computational cost. In this study, we apply the singular value decomposition to the time-series results of a severe accident code to develop a reduced order modeling (ROM). Using the ROM, the probabilistic safety margin analysis for the station blackout with a total loss of feedwater capabilities at a boiling water reactor is carried out. The dominant parameters to the accident progression are assumed to be the down-time and the recovery-time of the reactor core isolation cooling system, and decay heat. To reduce the number of RELAP5/SCDAPSIM analyses while maintaining the prediction accuracy of ROM, we develop a data sampling method based on adaptive sampling, which selects the new sampling data based on the dissimilarity with the existing training data for ROM. Our ROM can rapidly reproduce the time-series results and can estimate core conditions. By reproducing multiple results by ROM, a time-dependent core damage probability distribution is calculated instead of the direct use of RELAP5/SCDAPSIM.     

Steady State and Transient Analysis of Novel Design Helium Cooled Ceramic Blanket (HCCB) System of China Fusion Engineering Test Reactor (CFETR)

An upgraded form of China fusion engineering test reactor (CFETR) was investigated for the safety performance. In the current study, modification of the designs were presented with relative tolerance. The steady state were calculated for the new design using Relap5 code. Two accidents were simulated i.e., in-vessel and In-box loss of coolant accident. These accidents were simulated in helium cooled ceramic blanket (HCCB) system for the purpose to investigate the safety measures of the CFETR. It is utmost important to ensure the safety performance of the reactor. In this research, sudden break at blanket system was assumed and calculated different parameters including temperature, pressure and coolant fluxes to observe the differences in pattern during the accident under limited time domain. The research is very important because the design of HCCB is new and there is a need to conduct steady state and transient state of the reactor in order to make sure and authenticate the design and to safer the reactor.     

微型中子源反应堆热工水力安全分析

微型中子源反应堆（简称微堆）是一种典型的罐池式反应堆,采用自然对流循环冷却。为研究微堆的安全性,对其额定功率运行以及事故工况下的瞬态热工水力特性进行了模拟。针对额定功率运行工况,采用CFD软件进行瞬态热工水力三维数值模拟,同时采用RELAP5程序进行一维计算,二者计算结果相符,表明了计算结果的正确性及额定功率工况的安全性。采用RELAP5程序对反应性引入事故进行了计算,计算结果进一步印证了微堆的自稳特性和固有安全性。