A thermal-fluid assessment of a cooled-vessel concept for a VHTR

Flow distribution and thermal analyses of a conceptual design of a cooled vessel for a very high temperature reactor (VHTR), which has a forced vessel cooling with an internal coolant path through a permanent side reflector, have been performed. A computational fluid dynamics (CFD) code was employed to investigate flow distributions at inlet and upper plenums of the proposed cooled-vessel concept. Thermal-fluid analyses of the cooled vessel during a normal operation were carried out by using the CFD code with the boundary conditions provided by the GAMMA system analysis code. The transient analyses during postulated accidents were conducted by the GAMMA code itself. According to the results, the flow deviation at the riser holes due to a change of the inlet flow path to the core inlet is about ±20% which results in about a 3-7% core flow deviation from the average value depending on the upper plenum height. The pressure drops in the inlet and upper plenums are estimated to be from 13 to 25 kPa with a change of the upper plenum height. A cooling flow of more than 4 kg/s is sufficient to maintain the RPV temperature within the required limit during a normal operation. Transient analysis reveals that the reactor vessel is exposed to a temperature above its limit of 371 °C but this duration is shorter than the allowable time for a creep region with a sufficient safety margin. The results suggest that the cooled-vessel concept considered in this paper has the potential to be used for a VHTR but further and more detailed studies are required to realize the proposed concept.     

Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents

The high temperature engineering test reactor (HTTR) being constructed by the Japan Atomic Energy Research Institute is a graphite-moderated, helium-cooled reactor with an outlet gas temperature of 950 °C.Two independent vessel cooling systems (VCSs) of the HTTR cool the reactor core indirectly during depressurized and pressurized accidents so that no forced direct cooling of the reactor core is necessary. Each VCS consists of a water cooling loop and cooling panels around the reactor pressure vessel (RPV). The cooling panels, kept below 90 °C, cool the RPV by radiation and natural convection and remove the decay heat from the reactor core during these accidents.This paper describes the design details and safety roles of the VCSs of the HTTR during depressurized and pressurized accidents. Safety analyses prove that the indirect core cooling by the VCSs and the inherent safety features of the reactor core prevent a temperature increase of the reactor fuel and fission product release from the reactor core during these conditions. Furthermore, it is confirmed that even if VCS failure is assumed during these accidents, the reactor core and RPV can remain in a safe state.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

熔融物堆内滞留条件下压力容器变形

熔融物堆内滞留(In-Vessel Retention,IVR)已经成为第三代反应堆一项关键的严重事故缓解策略,而压力容器外部冷却(External Reactor Vessel Cooling,ERVC)技术则是保证IVR得以成功实施的关键。当发生堆芯熔化时,高温熔融物对压力容器(Reactor Pressure Vessel,RPV)下封头的热冲击会导致RPV壁面和由其构成的外部冷却通道的形状发生变化,使局部传热恶化,进而造成IVR的失效。因此,有必要对IVR条件下RPV壁面的变形进行研究。本文利用有限元软件ANSYS对RPV进行了几何建模、温度场分析和力学场分析。结果表明,在RPV外部实现冷却、内部实现泄压的前提下,壁面变形为13.85-18.75 mm。在1 MPa内压的作用下,高温蠕变会使壁面变形随时间增大,但其增量有限。热膨胀是造成壁面变形的主要因素。     

The HTTR reactor pressure vessel and its integrity against a PTS event

Before manufacturing the real steel to be used in the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) the vessel manufacturer and materials supplier made a sample steel by the same procedure as for the real steel (2.25Cr-1 Mo) and conducted many tests to obtain material strength data for its base and weld metals. The test results showed that the sample steel satisfied the HTTR design requirements. Vessel cooling panels are set on the inner surface of the biological shielding concrete around the RPV, and are circulated with cooling water at 0.5 MPa and 40°C to cool the shielding concrete during normal operation of the reactor. By supposing that the cooling panel breakes and the water discharges to the RPV outer surface heated at 400°C, the stress distribution generated in the vessel wall by a pressurized thermal shock (PTS) event can be calculated using a finite element method code. This paper describes some of the results obtained from the material testing of the sample steel and the estimated result using the scheme developed for a light water reactor pressure vessel, to clarify the integrity of the HTTR-RPV under a PTS event.     

The impact of design on the decay heat removal capabilities of a modular pebble bed HTR

The decay heat removal capabilities are an important safety feature of the modular pebble bed HTR. It is designed in a way that also during loss of cooling accidents the decay heat can be removed purely by passive means without exceeding predefined temperature limits for fuel and structures. Such a plant design, however, yields limitations on the power output. Thus, from the thermal hydraulic point of view a reactor with maximum power which still obeys the temperature limits of fuel and components, represents an optimal design of a modular pebble bed HTR.In this paper, design options for a modular pebble bed HTR are discussed with respect to their capabilities of decay heat removal.Both pressurized and depressurized loss of coolant accidents are investigated. Optimization of design features is considered with reference not only to the maximum fuel temperature during the accidents, but also to the temperature of structures, mainly that of the reactor pressure vessel. It is pointed out that annular cores can produce higher power without exceeding fuel temperature limits, especially during depressurization accidents. This is mainly due to geometrical effects. Heat storage effects of the inner column also have an influence on the maximum fuel temperature by increasing the time at which this temperature is reached. While a thermal insulation of the core and the reflector increases the fuel temperature, maximum temperature of the pressure vessel and the core barrel is decreased. Thus, carbon blocks represent an important element for optimization of the design.     

基于非能动系统功能可靠性的IVR-ERVC保温层几何优化与可靠性评估

针对CPR1000在严重事故条件下实施熔融物堆内滞留 压力容器外部冷却(IVR ERVC)方案的保温层几何参数优化设计需求,按设计参数及关键参量可能范围及分布,采用拉丁超立方抽样(LHS)确定输入参数组合,运用Relap5/Mod3程序进行不确定性传递计算。根据计算结果,进行参数对ERVC功能及行为的敏感性分析;基于提出的ERVC相关功能可靠性准则与统计分析,进行CPR1000一类非能动ERVC保温层设计参数名义值的初步选取。进一步在确定保温层结构参数基础上,进行ERVC功能可靠性分析,为CPR1000概率安全评价提供ERVC系统可靠性估计。     

Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of ‘overpressure-molten pool structure’ when the vessel failure started at the ‘hot’ layers of the vessel. It was shown in this study that the processes in the molten pools reach a quasistationary state at 2000…3000 s after molten pool formation. Numerical results in this paper illustrate that the large creep deformations of the vessel lower head can lead to an appearance of the gaps between the vessel surface and the molten pool crust. It is obvious that the joint thermal and structural analyses are needed for the accurate tracing of the initial bounds of the vessel and molten pool during simulations.     

压力容器-保温层流道变形条件下临界热流密度试验研究

Aiming at the deformation issue of flow gap between reactor pressure vessel (RPV) and insulation in external reactor vessel cooling (ERVC), the effects of insulation deformation on the critical heat flux (CHF) of the bottom head were investigated with FIMR test facility within the same flow rate range. The influence laws of different factors including angle of bottom head wall, flow rate and insulation deformation on CHF of RPV wall were analyzed. The safety margin of ERVC under deformation condition was finally obtained. The results show that the CHF of the bottom head will increase as the angle of the bottom head wall increases or the flow rate increases. Compared to the CHF of the bottom head in the prototype channel, the varying amplitude of CHF under deformation condition is less than 7%. In a word, the effect of insulation deformation on CHF is not significant. What is more, the location, where the safety margin is smallest, at the bottom head wall in the deformed gap is the same with that in the normal gap, while the smallest safety margin in the deformed gap is slightly improved.     

压力容器-保温层流道变形条件下临界热流密度试验研究

针对压力容器外部冷却（ERVC）应用中的压力容器-保温层流道（RPV-保温层流道）变形问题,利用提高临界热通量影响因素（FIMR）的试验装置,在相同流量范围开展了变形条件下壁面临界热流密度（CHF）的试验研究,分析了流道变形和流量变化对压力容器（RPV）下封头壁面CHF的影响规律,获得了流道变形情况下ERVC的安全裕度。结果表明:随着RPV下封头角度升高,循环流量增加,下封头壁面CHF增大;与原型流道相比,变形流道下封头壁面CHF的变化幅度小于7%,流道变化的影响并不显著;变形流道中,下封头壁面安全裕量最小的位置与原型流道相同,其安全裕量略有提高。      

针对REPEC加热实验的RELAP5程序模拟与分析

为研究压力容器外部流道的冷却能力及流动传热过程,在反应堆压力容器外部冷却(REPEC, Reactor Pressure vessel External Cooling)实验台架前期加热实验的基础上,采用RELAP5程序对实验工况进行模拟和对比。模拟结果与实验数据一致性较好。随加热热流、进出口面积的增加,系统内自然循环流量也增加;入口欠热度对自然循环流量的影响不是很明显;近饱和沸腾条件下,系统出现明显的两相不稳定流动。     

Structural integrity assessment of the reactor pressure vessel under the external reactor vessel cooling condition

Since the suggestion of external reactor vessel cooling (ERVC), the effects of melting and cooling on the response of structural integrity of the reactor pressure vessel (RPV) under core melting accident conditions have been investigated. To investigate the initial behavior of RPV lower head and the effects of analysis conditions on the structural integrity of RPV, the transient analysis is utilized considering the transient state. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The temperature and stress analyses are performed for core melting accident by using ABAQUS. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. In addition, the results by transient analysis are compared with those by steady state analysis and the effects of analysis conditions on structural integrity are reviewed.     

大功率先进压水堆压力容器外部冷却能力研究

目前压力容器外部冷却（ERVC）作为严重事故管理策略中压力容器内熔融物滞留（IVR）的一部分已得到了广泛应用。本文采用RELAP5系统安全分析程序定性研究一些流动参数和边界条件（如进出口面积、冷却水的入口温度、下封头处的加热功率、下封头处流道的间隙尺寸及注水高度等）对大功率先进压水堆压力容器外部冷却的自然循环能力产生的效应,它为结构的设计和系统的瞬态响应行为提供了一定的分析依据。     

Air ingress phenomena in a depressurization accident of the very-high-temperature reactor

The inherent properties of the very-high-temperature reactor (VHTR) facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However, it is still not clear if the VHTR can maintain a passively safe function during the primary-pipe rupture accident, or what would be a design criterion to guarantee the VHTR with the high degree of passively safe performances during the accident. The primary-pipe rupture accident is one of the most common of accidents related to the basic design regarding the VHTR, which has a potential to cause the destruction of the reactor core by oxidizing in-core graphite structures and to release fission products by oxidizing graphite fuel elements. It is a guillotine type rupture of the double coaxial pipe at the nozzle part connecting to the side or bottom of the reactor pressure vessel, which is a peculiar accident for the VHTR. If a primary pipe ruptures, air will be entered into the reactor if there is air in the reactor containment or confinement vessels. This study is to investigate the air ingress phenomena and to develop the passively safe technology for the prevention of air ingress and of graphite corrosion. The present paper describes the influences of a localized natural circulation in parallel channels onto the air ingress process during the primary-pipe rupture accident of the VHTR.     

Fracture mechanical evaluation of an in-vessel melt retention scenario

This paper presents methods to compute J-integral values for cracks in two- and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.     

真实表面材料及其老化效应对反应堆压力容器ERVC-CHF影响的试验研究

通过反应堆压力容器外部冷却（ERVC）实现熔融物堆内滞留（IVR）技术是核电厂严重事故缓解的重要措施之一。在本文的研究中,建立了二维切片式、全尺寸的试验台架FIRM,开展严重事故条件下反应堆压力容器ERVC-临界热流密度（CHF）试验研究。试验采用去离子水作为试验工质,获得了反应堆压力容器下封头ERVC过程的CHF限值。研究了真实表面材料对CHF的影响及其影响机理,讨论了在去离子水下表面材料SA508 Gr3. Cl.1钢的老化效应。本试验研究对于认识反应堆压力容器IVR-ERVC条件下的CHF行为、提高反应堆压力容器安全性有重要意义。     

Global rupture of a nuclear reactor lower head with elevated temperature due to severe accidents

This paper is concerned with the global rupture of a reactor pressure vessel (RPV) with elevated temperature due to severe accidents in order to check if the RPV wall can retain the high-elevated pressure. The global rupture of an RPV is simulated by finite element limit analysis for the collapse load and mode to secure the safety criteria of a nuclear reactor under severe accident conditions. Finite element limit analysis is a systematic tool dealing with upper bounding and minimization technique to calculate the collapse load and mode. The finite element code (CALF, computer analysis of lower head failure) developed provides the temperature elevation in the lower head of a nuclear reactor under severe accident conditions as well as the collapse load and mode. The thermal analysis has to deal with heat transfer from the debris pool to the RPV wall and the top of the pool. The temperature distribution in such a system depends sensitively on the initial temperature of the debris pool and the thermal properties of a gap between the debris crust and the RPV wall. For accurate calculation, the thermal properties of a gap have to be determined in consideration of the gap size and conditions.     

Conceptual design modifications of the cooling system of MNSR reactor to increase its maximum continuous operation time

To increase the maximum daily operation time of Miniature Neutron Source Reactor (MNSR) reactor several conceptual thermal hydraulic design modifications have been investigated aiming at the improvement of reactor cooling conditions to limit the increase of average core temperature. For this purpose an integrated full-scale thermal hydraulic-neutronics model using the advanced code ATHLET has been developed, tested and verified. The selected design modifications rely upon introducing auxiliary cooling systems operating in four different modes to cool pool water or reactor water using heat exchanger located either inside or outside of reactor pool. The simulation results show that the increase of continuous reactor operation time varies between 1 and 8 additional operation hours. The optimal results are achieved for the second and the fourth options that use external heat exchanger. The second option enables the extending of continuous operation time up to 10 h and the fourth up to 15 h, both at nominal reactor power and under the assumption of initial excess reactivity corresponding to the fresh reactor core. The analysis included the evaluation of xenon poisoning effect on the increase of operation time. It has been shown that its remarkable effect starts after the first 3 operation hours and increases continuously after that. For the best cooling options, where the average core temperature is being fixed at certain value resulting in complete elimination of reactivity feedback of cooling temperature, xenon effect becomes the exclusive limiting effect during the later operation phase. The analysis discuss also general aspects of technical realization for the different cooling options in relation with the specific features of MNSR and the preliminary engineering safety measures and operational radiological protection that have to be taken. The performed analysis and the achieved results during this work would make valuable contribution for updating the Safety Analysis Report (SAR) of MNSR.     

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.     

示范快堆堆坑通风冷却三维数值模拟

为详细研究示范快堆堆坑内空气流动状态和温度分布情况,检验现行堆坑通风系统布置合理性与冷却效果,本文利用CFD软件对正常运行工况下的示范快堆堆坑空气流域进行三维数值模拟。结果表明,通风系统冷却效果满足设计要求,堆坑混凝土内壁最高温度为50.7 ℃,但堆坑内部流场复杂,温度分布的不均匀性较高,通风系统进出口排布方式需进一步优化。计算结果为主容器及贯穿件支承热工计算提供了更为准确的边界条件,为示范快堆一回路设计提供参考。