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.     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

Heat removal characteristics of a concrete cask by a simplified test model

The present study is concerned with the characteristic cooling flow in the annular gap of a concrete cask used to store spent nuclear fuel. The concrete cask is cooled by a natural convection flow of air passing through an internal annular gap between the outside of a canister and the inside of the concrete vessel. The heat transfer coefficient and friction loss coefficient of such a flow could not be fully estimated even if we used existing handbooks, because the airflow has unique features. Simulation experiments using a simplified model for the cooling path have been conducted to estimate the heat transfer coefficient and friction loss coefficient. It was found by this study that the heat transfer coefficient well agreed with the prediction by an empirical formula applied to the free convection, and the friction loss coefficient was 2–2.5 times the value of an isothermal flow.     

压力管式超临界水堆极限事故下“无堆芯熔化”概念评估

在自主研发的事故分析程序SCTRAN的基础上,开发并验证了二维导热模型和辐射换热模型,并将改进后的SCTRAN应用于加拿大压力管式超临界水堆在失水事故（LOCA）叠加丧失紧急堆芯冷却系统（LOECC）事故中的堆芯安全评估,并对燃料棒到慢化剂之间的传热效率以及关键的影响因素进行了评估。计算结果表明,在LOCA叠加LOECC工况下,燃料棒到燃料通道的辐射换热和燃料棒到蒸汽的自然对流换热能够有效导出反应堆的衰变余热,最高功率的燃料组件内、外圈燃料棒的最高包壳温度分别为1278℃和1192℃,均低于不锈钢包壳的熔化温度,因此整个事故过程中不会发生堆芯熔化。      

Inter-subassembly heat transfer of sodium cooled fast reactors: Validation of the NETFLOW code

This paper describes the prediction of temperature at the exit of subassemblies of a sodium cooled fast reactor using the NETFLOW code. Until present time, this plant dynamics calculation code is expected as a tool of nuclear education, and has been validated using data obtained at facilities or reactors cooled with water or sodium. A natural circulation test was conducted in the experimental fast reactor ‘Joyo’ with a 100 MW irradiation core. Also a turbine trip test was conducted in the prototype fast breeder reactor ‘Monju’. These tests were chosen to validate a model to calculate inter-subassembly heat transfer consisting of heat conduction and heat transfer by inter-wrapper flow. Based on the calculation for the natural circulation test in primary and secondary loops of ‘Joyo’, the model to calculate the heat transfer in radial direction of the inter-subassemblies simulated reasonable sodium temperature behaviors at the exit of subassemblies. Good agreement was also obtained in prediction of temperatures at the exit of the ‘Monju’ subassemblies. Through these validations, it was shown that the one-dimensional plant dynamics code NETFLOW could trace temperatures at the exit of the subassemblies of fast reactors with the inter-subassembly heat transfer model.     

Study of heat transfer by using DEM–CFD method in a randomly packed pebble-bed reactor

The pebble-bed reactor is one of the most promising designs for the nuclear energy industry. In this paper,a discrete element method-computational fluid dynamics(DEM-CFD) approach that includes thermal conduction, radiation, and natural convection mechanisms was proposed to simulate the thermal-fluid phenomena after the failure of forced circulation cooling system in a pebble-bed core. The whole large-scale packed bed was created using the DEM technique, and the calculated radial porosity of the bed was validated with empirical correlations reported by researchers. To reduce computational costs, a segment of the bed was extracted, which served as a good representative of the large-scale packed bed for CFD calculation. The temperature distributions simulated with two different fluids in this DEM-CFD approach were in good agreement with SANA experimental data. The influence of the natural convection mechanism on heat transfer must be taken into account for coolants with strong convective capacity. The proposed DEM-CFD methodology offers a computationally efficient and widely applied method for understanding the heat transfer process in a pebble-bed core. The method can also be easily extended to assess the passive safety features of newly designed fluoride-salt-cooled pebble-bed reactors.     

Sensitivity study on depressurized LOFC accidents with failure of RCCS in a modular gas-cooled reactor

A modular gas-cooled reactor design with a thermal output of 600 MWt and a core exit temperature of 950 °C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multicomponent problems was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600 °C at about 50 h at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to be devised based on the comprehensions on the heat transfer characteristics.     

Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel).This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes.In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 °C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures, etc.     

Simulation and analysis of transient cooldown natural convection experiments

The characteristics of natural convection heat transfer during transient cooldown in 3-dimensional fluid layers and hemispherical cavity have been investigated by means of a finite-difference numerical method. It was found that the turbulent structure and heat transfer characteristics of the unstably-stratified upper wall region are similar for internal heating (IH) case and transient cooldown (TCD) case, except near the cooled bottom wall of fluid layer, and for a small region (φ≤15°) near the bottom of the hemispherical cavity. For most of the surface area of the hemispherical cavity, there is excellent agreement between the heat fluxes calculated for the IH and the TCD cases. Calculations also showed that pseudosteady-state natural convection (PSSNC) is a better model for simulation of volumetric energy sources.     

Calculating decay heat removal rates via natural circulation along variable heat flux reactor fuel plates

A recently developed integral technique is applied to natural convection cooling along test reactor fuel plates. The technique is demonstrated for water and air flow. In the case of air flow, the process is characterized by a large temperature rise along the fuel channel, thereby rendering the commonly applied Boussinesq approximation invalid. This case is a heat transfer problem of particular interest in accident analyses such as determining the level of decay heat dissipation possible, without exceeding the melting temperature of the fuel, subsequent to a hypothetical loss of primary coolant.     

Hydrodynamics of degraded core cooling

The literature on hydrodynamics of forced convection cooling of particle beds is reviewed and used to assess the characteristics of in-situ cooling of a degraded LWR reactor core under conditions representative of severe accidents. It is found that the pressure head required, for a given rate of liquid water flow, through a totally degraded core is one to two orders of magnitude higher than the case of intact core geometry. To remove decay heat of up to 1–2% of the reactor normal power, even with conservative assumptions, the pressure head is within the capability of the main reactor pumps. However, particles with very small diameter (less than 100 μm) will potentially be swept out by the flow. The sensitivity of the hydraulic characteristics to the allowed coolant temperature rise across the fragmented core is also investigated.     

Experimental study on the natural circulation of a passive residual heat removal system for an integral reactor following a safety related event

Natural circulation characteristics of an integral type reactor during the operation of a passive residual heat removal system (PRHRS) following a safety related event has been experimentally investigated by using the VISTA facility. A PRHRS actuation trip signal is generated by a high power trip signal following a steam flow increasing event. The experimental results show that the single-phase coolant flows steadily in the primary loop by a natural convection process and that it effectively removes the decay heat from the core through a steam generator during the PRHRS operation. The heat transfers through the PRHRS heat exchanger and the emergency cooldown tank (ECT) are sufficient enough to enable a two-phase natural circulation of the coolant in the PRHRS loop.     

Computation of heat transfer through a horizontally arranged HTR-fiber insulation

The heat transfer through a horizontally arranged ceramic fiber insulation operating at high pressure and temperature is studied theoretically. The contributions of heat conduction, natural convection and thermal radiation are considered. With an increasing value of the product of the Rayleigh and Darcy numbers, (Ra, Da), the natural convection is amplified, thus diminishing the efficiency of the thermal insulation. For high temperatures the thermal radiation is no longer negligible. The results of the computer code agree well with experiments available up to a temperature of 670 K and up to a pressure of 40 bar in helium or air. Finally, the heat transfer of a fiber insulation operating under high temperature reactor conditions is predicted, demonstrating that natural convection is nearly suppressed when an insulant bulk density of _b = 165 kg/m³ is provided.     

直流蒸汽发生器传热管破裂事故分析

通过对直流蒸汽发生器传热管破裂（SGTR）事故的分析,可看出RELAP5瞬态分析程序能较好地模拟一体化反应堆在SGTR事故后的事件响应序列及主要热工水力现象,例如环路的不对称效应、主回路的自然循环等。一体化反应堆在发生SGTR事故后,可通过一系列安全与保护系统的动作得到有效缓解,并最终能应用非能动余热排出系统（PRHRS）的自然循环导出堆芯余热,使反应堆处于安全状态。同时,受事故影响蒸汽发生器压力在PRHRS投入运行后会快速升高,最终与一回路压力相平衡,此后,破口处的泄漏也会终止。此外,本文还研究了破口处临界流量及其积分流量结果不确定性的影响因素,其中主要考虑了采用不同的临界流模型和破口建模方式等两个方面。     

Modelling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

The objective of this paper is to study the heat and mass trasnfer processes related to core melt discharge from a reactor vessel in a light water reactor severe accident. The phenomenology modelled includes the convection in, and heat transfer from, the melt pool in contact with the vessel lower head wall, the fluid dynamics and heat transfer of the melt flow in the growing discharge hole and multi-dimensional heat conduction in the ablating lower head wall. A research programme is underway at the Royal Institute of Technology (Kungliga Tekniska Högskolan, KTH) to (1) identify the dominant heat and mass transfer processes determining the characteristics of the lower head ablation process: (2) develop and validate efficient analytical/computational models for these processes; (3) apply models to assess the character of the melt discharge process in a reactor-scale situation; (4) determine the sensitivity of the melt discharge to structural differences and variations in the in-vessel melt progression scenarios. The paper also presents a comparison with recent results of vessel hole ablation experiments conducted at KTH with a melt simulant.     

Thermal hydraulic challenges of Gas Cooled Fast Reactors with passive safety features

Transient response of a Gas Cooled Fast Reactor (GFR) coupled to a recompression supercritical CO₂ (S-CO₂) power conversion system (PCS) in a direct cycle to a Loss of Coolant Accident (LOCA) and a Loss of Generator Load Accident is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post-LOCA events when system pressure is lost and when reliance on passive decay heat removal (DHR) is emphasized. Although it is possible to design a S-CO₂ cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal that provide flow in well defined regimes with low uncertainty, and can be easily overdesigned to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.     

Heat transfer from a partially uncovered reactor core

This paper documents a model which has been developed for predicting the temperature distribution along a “flow channel” of a pressurized water reactor during simulated, uncovered core conditions. In the model, heat conduction along the fuel element, convection from the surface to the coolant, radiation exchange between the clad surface and steam, and surface exchange between adjacent fuel rods are considered. Variations of the thermophysical properties of the fuel road and of the coolant with temperature are accounted for, but oxidation of Zircaloy is not modeled. Extensive sensitivity studies on the effects of heat generation in the core, steam velocity, pressure level, uncovered core height, presence of hydrogen gas in the coolant, power skew, clad emissivity, and convective heat transfer correlations have been examined. The results show that the importance of radiation in comparison with convection increases with an increase in the fuel rod temperature, pressure, and clad emissivity.     

大功率先进压水堆IVR有效性评价中熔池换热研究

熔融物堆内滞留-压力容器外部冷却（IVR-ERVC）是一种重要的核电厂严重事故缓解措施。当前针对IVR有效性评价的方法主要是基于集总参数模型对下封头熔池换热进行分析。在大功率先进压水堆熔池集总参数法计算中,堆芯重量变大、压力容器尺寸增加会导致熔池自然对流换热中的瑞利数Ra ′增大。通过使用集总参数分析程序,对比研究熔池氧化层各换热模型的适用范围,计算大功率先进压水堆高瑞利数条件下稳态熔池的自然对流换热,模拟两层稳态熔池模型中压力容器外壁面的热流密度分布,对其进行选定严重事故序列下的IVR-ERVC有效性评价,并对堆内构件设计提出建议。     

Heavy-gas injection in the Generation IV gas-cooled fast reactor for improved decay heat removal under depressurized conditions

Decay heat removal is a key safety and design issue for the Generation IV gas (helium)-cooled fast reactor. This paper investigates the natural convection capability of the dedicated DHR loops under depressurized conditions while injecting a heavy gas into the system. Investigated is a loss-of-coolant accident using the TRACE code. The goal of the study is to improve fuel/cladding temperature behavior during LOCA transients with the enhancement of passive safety by operation in natural convection only, while accepting 10 bar back-up pressure in the guard containment. The paper investigates the cooling capabilities of different heavy gases. Furthermore, different injection locations and mass flow rates have been tested, in order to address possible core-overcooling problems resulting from rapid depressurization of the gas reservoir. It has been shown that, among the gases investigated, CO₂ is the best choice from the thermal-hydraulics viewpoint, being able to cool the core satisfactorily for a broad range of injection rates. N₂ can be envisaged as an alternative solution in case of chemical problems with CO₂. Supplementary studies carried out for the CO₂ and N₂ injection cases include that of the sensitivity to the number of available DHR loops and to the LOCA break-size. The effect of the resulting neutron spectrum changes on the shutdown-reactivity margin has also been investigated.