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.     

Studies on diffusion and natural convection of two-component gases

A primary-pipe rupture accident is one of the design-based accidents of a high-temperature engineering test reactor (HTTR), which is being developed at JAERI. When the primary pipe ruptures, air is expected to enter into the reactor core from the breach by molecular diffusion and natural convection. In order to investigate the process of air ingress during the early stage of the primary-pipe rupture accident, experimental and analytical studies are performed on the conjugate phenomenon of the transient molecular diffusion and natural convection of a two-component gas mixture in two test sections, a reverse-U-shaped tube and a test model simulating simply the reactor. One-dimensional basic equations for continuity and momentum conservation are numerically solved to obtain a concentration change of gas species and an initiation time of a natural circulation of pure nitrogen in the reverse-U-shaped tube. Moreover, a modified numerical solution is proposed to reduce the computing time. A one-dimensional flow net work model is employed to calculate the transport process of air in the test model simulating the reactor. The calculated results agree well with the experimental ones on the concentration change of gas species and the initiation time of the natural circulation of pure nitrogen or pure air.     

Consequences of delayed air ingress following a depressurization accident in a high temperature reactor

Air ingress is a specific event in a high temperature reactor (HTR). The potential threat posed by air ingress lies in the chemical reaction of oxygen with hot graphite at a temperature above 500 °C leading to reaction heat and graphite corrosion. In order to assess the consequence of air ingress into the reactor, it is postulated that breaks are present above and below the reactor core and that unobstructed ingress of air through them is possible. It is obvious that the air ingress incident has to be preceded by a depressurization accident. For this hypothetical scenario the maximum possible air flow rate through the core resulting solely from the pressure losses in the core is estimated as a function of the break cross-sections exposed above and below the core.In this paper, the thermal behavior of an HTR with prismatic fuel (operating inlet/outlet temperatures 450/850 °C) during air ingress accident conditions has been investigated. In particular, maximum temperatures and burn-off of the fuel and bottom graphite reflector for various air flow rates for the postulated hypothetical scenario have been analyzed. It also indicates the limiting time at which the graphite layer of fuel will be completely burnt-off and the fuel compacts exposed. In addition, the consequences of delayed air ingress after a core heat up following depressurization have been investigated.This paper, thus, throws light on the safety aspects of the new generation HTRs with prismatic fuels (e.g. NGNP/ANTARES) conceived for power generation and process heat application.     

Study on air ingress during an early stage of a primary-pipe rupture accident of a high-temperature gas-cooled reactor

A primary-pipe rupture accident is one of the design-based accidents of the HTTR. As the first step of our final goal of predicting the multicomponent gas flow in a reactor during the early stages of the accident, the present paper aims at studying experimentally and analytically, the basic features of air ingress and gas transportation by transient molecular diffusion and the transient natural convection of a two-component gas mixture.The present paper comprises two main parts. The first part deals with analytical and experimental studies on N₂ ingress (corresponding to air ingress) and gas transportation by molecular diffusion and the one-dimensional natural convection of an He-N₂ two-component gas mixture in a reverse-U-shaped tube. Analytical and experimental results are discussed on the N₂ mole fraction change with time after the simulated pipe rupture and on the initiation time of the natural circulation of pure N₂.The second part deals with a preliminary simulation test of air ingress during the early stages of the accident. The test is performed with a very simple model of the reactor. The experimental results are discussed on the change in mole fraction of air with time and on the initiation time of the natural circulation of pure air.     

HTR-PM两根一回路连接管断裂的进气事故分析

进气事故是模块式高温气冷堆关注的超设计基准事故之一,石墨氧化腐蚀反应可能导致反射层结构强度减弱、燃料元件完整性和包容裂变产物能力被破坏,以及产生可燃气体等较严重后果。进气事故的分析研究对进一步掌握高温气冷堆的事故特性以及提高反应堆的安全设计具有重要意义。本文基于200MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步设计,假设与一回路压力边界上、下相连的燃料元件进料管和卸料管同时发生断裂,从而形成烟囱效应并导致空气进入堆芯,利用高温气冷堆专用系统分析软件TINTE对自然循环建立及后续的进气腐蚀过程进行了研究,分析了自然循环流量、堆内石墨腐蚀速率、舱室氧气消耗量、燃料元件温度等关键参数的变化。结果表明,即使考虑腐蚀反应的不均匀性,事故后约60h时才会出现首个燃料包覆颗粒裸露现象,燃料元件最高温度峰值低于1620℃的设计限值,保持完好的燃料包覆颗粒仍具有包容放射性裂变产物的能力。同时,如果在相应的时间内采取措施切断进气源,使石墨腐蚀反应不能继续发展,将不会对反应堆的安全造成严重的影响。     

HTR-10进气事故下堆芯石墨腐蚀分析

热气导管的双端断裂事故是１０ＭＷ高温气冷堆（ＨＴＲ－１０）的假想极限事故,该事故喷放阶段结束后,在气体扩散和浮升力的作用下,堆舱中的空气通过破口进入堆芯,并在堆芯流道和堆舱组成的回路中慢慢形成自然对流,从而引起进气事故。为了分析堆芯石墨的腐蚀情况,本文首先对ＨＴＲ－１０堆芯结构作了简化处理,然后计算了堆体简化流道内气体自然对流的质量流量、固相和气相的温度、石墨的腐蚀率、石墨的腐蚀总量以及燃料元件经腐蚀后的裸露率。这些计算结果表明,即使在该极限事故下,ＨＴＲ－１０仍有很好的安全特性。     

高温气冷堆用石墨材料的氧化性能研究

高温气冷堆均选用石墨材料作为结构材料和慢化剂.在反应堆的运行过程中,由于冷却剂中含有的氧化性气体杂质以及可能发生的进水事故和进气事故,会发生石墨材料的氧化,进而影响反应堆的正常运行和安全.本文主要对近期反应堆用石墨材料的氧化研究进行综合评述,并在此基础上,指出今后需要进一步研究的内容.     

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.     

Fission product behaviour and graphite corrosion under accident conditions in the HTR

This paper primarily gives an overview of methods and data in source term estimations for the HTR with pebble bed core. For medium size HTRs the risk dominating accidents are tied to core heat-up events, where a significant portion of the fission product inventory may be released from the coated fuel particles. Here the research mainly is focused on temperature-induced coated particle failure and the interaction of metallic fission products with the core graphite. For small HTRs, with their limitation of maximum temperatures below coated particle failure limits, core heat-up accidents virtually play no role with respect to source terms. Here the risk is dominated by accidents like water ingress or rapid depressurization which may lead to a partial release of fission products accumulated on primary circuit surfaces like the steam reformer. Deposition of fission products and remobilization under the conditions mentioned above are predominant research areas. It can be expected that the ongoing and planned improvements of models and data base, in particular for the medium size HTR, will result in a further reduction of the already low source terms.A principal possibility for core degradation and hence destruction of fission product barriers is graphite corrosion caused by massive air ingress. The research effort in this field as well as for graphite corrosion during water ingress accidents is described in Part B of this paper. From the viewpoint of risk for this type of accident no significant contribution to that of present reactor concepts was found.     

底反射层氧化对模块式高温气冷堆进气事故的影响分析

进气事故是模块式高温气冷堆(HTR-PM)事故分析中重点考虑的一种事故类型。核级石墨在高温气冷堆中被广泛用作反射层材料、结构材料和慢化材料等。在进气事故中,燃料元件基体石墨发生氧化反应增加了燃料颗粒裸露和放射性释放的风险,底反射层发生氧化反应降低了石墨材料的机械性能,可能破坏堆芯底部结构的完整性。本文利用高温气冷堆专用系统分析程序TINTE,分别选取两种不同氧化速率的石墨材料作为底反射层材料,以热气导管双端断裂的进气事故为例,分析不同材料对进气事故的影响。在保证底反射层完整性的前提下,底反射层采用高氧化速率的材料时,能明显降低燃料颗粒裸露和放射性释放的风险。     

Air ingress benchmarking with computational fluid dynamics analysis

The air ingress accident is a complicated accident scenario that may limit the deployment of high-temperature gas reactors. The complexity of this accident scenario is compounded by multiple physical phenomena that are involved in the air ingress event. These include diffusion, natural circulation, and complex chemical reactions with graphite and oxygen. In an attempt to better understand the phenomenon, the FLUENT-6 computational fluid dynamics code was used to assess two air ingress experiments. The first was the Japanese series of tests performed in the early 1990s by Takeda and Hishida. These separate effects tests were conducted to understand and model a multi-component experiment in which all three processes were included with the introduction of air in a heated graphite column. MIT used the FLUENT code to benchmark these series of tests with quite good results. These tests are generically applicable to prismatic reactors and the lower reflector regions of pebble-bed reactors. The second series of tests were performed at the NACOK facility for pebble bed reactors as reported by Kuhlmann [Kuhlmann, M.B., 1999. Experiments to investigate flow transfer and graphite corrosion in case of air ingress accidents in a high-temperature reactor]. These tests were aimed at understanding natural circulation of pebble bed reactors by simulating hot and cold legs of these reactors. The FLUENT code was also successfully used to simulate these tests. The results of these benchmarks and the findings will be presented.     

Air-ingress analysis: Part 2—Computational fluid dynamic models

Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy (DOE), is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high-temperature reactors (VHTRs). Phenomena identification and ranking studies to date have ranked an air-ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air-ingress-related models and verification and validation data are a very high priority.Following a loss of coolant and system depressurization incident, air will enter the core of the high-temperature gas-cooled reactor through the break, possibly causing oxidation of the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of lower plenum graphite because of oxidation might lead to a reactor safety issue. Computational fluid dynamics models developed in this study will improve our understanding of this phenomenon and is used to mitigate air ingress.This paper presents three-dimensional (3D) computational fluid dynamic (CFD) results for the quantitative assessment of the air-ingress phenomena. The 3D CFD simulation results show that the air-ingress accident is not controlled by molecular diffusion but density gradient driven stratified flow when the double-ended-guillotine break is assumed in a horizontal pipe configuration. It concludes that the previous air-ingress scenarios based on the molecular diffusion might not be correct and should be extensively modified to include real phenomena. This paper also presents a preliminary two-dimensional (2D) CFD simulation for validating an air-ingress mitigation concept using helium injection at the lower plenum. This simulation shows that the helium replaces air by buoyancy force and effectively mitigates air-ingress into the core.     

Multi-component diffusion analysis and assessment of GAMMA code and improved RELAP5 code

A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature.New and safer nuclear reactors (Generation IV) are now in the early planning stages in many countries throughout the world. One of the reactor concepts being seriously considered is the VHTR. To achieve public acceptance, these reactor concepts must show an increased level of inherent safety over current reactor designs (i.e., a system must be designed to eliminate any concerns of large radiological releases outside the site boundary).A computer code developed from this study, gas multi-component mixture analysis (GAMMA) code, was assessed using a two-bulb experiment and in addition the molecular diffusion behavior in the prismatic-core gas-cooled reactor was investigated following the guillotine break of the main pipe between the reactor vessel and the power conversion unit. The RELAP5 code was improved for the VHTR air ingress analysis and was assessed using inverse U-tube and NACOK natural circulation data.     

Comprehensive Analysis of Mechanical and Thermohydraulic Processes in the Graphite Masonry of Channel Rectors during an Accidental Rupture of a Fuel Channel

A methodology and preliminary results of a computational analysis of the processes occurring in the graphite masonry in RBMK channel reactors during the rupture of a fuel channel as a result of accidental overheating are presented. The methodology for the computational analysis is implemented using the U_STACK code, simulating the thermohydraulic and mechanical processes occurring in masonry with varying geometry in continuous coupling with the processes in the circulation loop and accident containment system. The deformation and rupture of a pipe in the damaged channel and coolant efflux are calculated using the prescribed accident scenario. After channel rupture, the parameters of the medium, the displacement of graphite blocks, and the sagging of channel pipes in the entire volume of the core are calculated. As a result, the additional force loads on neighboring fuel channels in the rupture zone, the reactor case, and the top and bottom plates are estimated. The objective of the calculations performed with the U_STACK code are assessment of the integrity of the components of the reactor core and substantiation of the impossibility of multiple ruptures occurring in the reactor fuel channels.     

ASTEC V2.0 computational evaluation of source term and its chemical forms under accidental conditions during mid-loop operation

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO₄ is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I₂ + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I₂ persistence). Creation of gaseous RuO₄ is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram.     

高温气冷堆用石墨材料的腐蚀实验与软件模拟

在高温气冷堆进水进空气事故下,空气和水蒸气会与堆内的石墨材料发生化学腐蚀反应,从而可能影响反应堆的安全。为研究高温气冷堆内石墨材料的氧化腐蚀特性,本文利用气相色谱法实验测量了IG-110石墨在不同温度和不同气体组分配比情况下的腐蚀速率及腐蚀产物,并利用THERMIX/REACT软件对整个石墨腐蚀过程进行了模拟。研究结果表明：反应温度对石墨腐蚀的影响最为显著,腐蚀速率随着温度的升高而增大,同时随着温度升高,CO与CO₂的含量比也逐渐增大。通过与实验结果对比分析,验证了THERMIX/REACT软件用于高温气冷堆安全分析的可靠性。     

Simulation of the chernobyl accident

An analysis of the April 26, 1986 accident at the Chernobyl-4 nuclear power plant in the Soviet Union is presented. The peak calculated core power during the accident was 550 000 MW_t. The analysis provides insights that further understanding of the plant behavior during the accident. The plant was modeled with the RELAP5/MOD2 computer code using information available in the open literature. RELAP5/MOD2 is an advanced computer code designed for best-estimate thermal-hydraulic analysis of transients in light water reactors. The Chernobyl-4 model included the reactor kinetics effects of fuel temperature, graphite temperature, core average void fraction, and automatic regulator control rod position. Preliminary calculations indicated the effects of recirculation pump coast down during performance of a test at the plant were not sufficient to initiate a reactor kinetics-driven power excursion. Another mechanism, or “trigger” is required. The accident simulation assumed the trigger was recirculation pump performance degradation caused by the onset of pump cavitation. Fuel disintegration caused by the power excursion probably led to rupture of pressure tubes. To further characterize the response of the Chernobyl-4 plant during severe accidents, simulations of an extended station blackout sequence with failure of all feedwater are also presented. For those simulations, RELAP5/MOD2 and SCDAP/MOD1 (an advanced best-estimate computer code for the prediction of reactor core behavior during a severe accident) were used. The simulations indicated that fuel rod melting was delayed significantly because the graphite acted as a heat sink.     

高温气冷堆蒸汽发生器换热管断裂事故进水量影响因素分析

以清华大学核能与新能源技术研究院设计的250 MW球床模块式高温气冷堆(HTR-PM)为例,对蒸汽发生器换热管断裂事故下影响一回路进水量的一些因素进行了分析.分析结果表明:除了断管位置、破口面积等对一回路进水量有直接影响外,进水量还与泄放管线直径、节流孔直径、泄放阀门选择、泄放系统动作设定等因素有关.合理地选择参数可有效排空蒸汽发生器内存留的水,避免一回路大量进水并减少一回路放射性物质向二次侧泄漏所造成的污染.     

Sensitivity studies of air ingress accidents in modular HTGRs

Postulated air ingress accidents, while of very low probability in a modular high-temperature gas-cooled reactor (HTGR), are of considerable interest to the plant designer, operator, and regulator because of the possibility that the core could sustain significant damage under some circumstances. Sensitivity analyses are described that cover a wide spectrum of conditions affecting outcomes of the postulated accident sequences, for both prismatic and pebble-bed core designs. The major factors affecting potential core damage are the size and location of primary system leaks, flow path resistances, the core temperature distribution, and the long-term availability of oxygen in the incoming gas from a confinement building. Typically, all the incoming oxygen entering the core area is consumed within the reactor vessel, so it is more a matter of where, not whether, oxidation occurs. An air ingress model with example scenarios and means for mitigating damage are described. Representative designs of modular HTGRs included here are a 400-MW(th) pebble-bed reactor (PBR), and a 600-MW(th) prismatic-core modular reactor (PMR) design such as the gas-turbine modular helium reactor (GT-MHR).     

A preliminary investigation of rapid depressurization phenomena following a sudden DLOFC in a VHTR

Air ingress has been identified as a potential threat for Very High Temperature gas-cooled Reactors (VHTR). Reactor components constructed of graphite will, at high temperatures, produce exothermic reactions in the presence of oxygen. The danger lies in the possibility of fuel element damage and core structural failure. Previous investigations of air ingress mechanisms have focused on thermal and molecular diffusion, density-driven stratified flow due to hydrodynamic instability, and natural convection. Not yet investigated is the possibility of a rapid flow reversal of helium coolant due to a Taylor (rarefaction) wave expansion after a hypothetical sudden Depressurized Loss of Forced Cooling (DLOFC) scenario in a VHTR. Conceivably, flow reversal of the helium coolant could entrain significant quantities of air into the reactor vessel. Our goal here is to simply demonstrate this natural phenomena of compressible flow that could possibly result in rapid air ingress into a VHTR. We start with a one-dimensional shock tube simulation to simply illustrate the development of a Taylor wave. The simulation is carried out far enough in time to allow the resulting reentrant flow to occur. Then, a simulation is performed of an idealized two-dimensional axisymmetric representation of the lower plenum of General Atomics GT-MHR subjected to a hypothetical catastrophic break of the hot duct. Results show the potential for significant and rapid air ingress into the reactor vessel in the case of a large break in the cooling system.