Study on the passive safe technology for the prevention of air ingress during the primary-pipe rupture accident of HTGR

A primary-pipe rupture accident is one of the design-basis accidents of a high-temperature gas-cooled reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This study is to investigate the air ingress phenomena and to develop the passive safe technology for the prevention of air ingress and of graphite corrosion. This paper describes the method for the prevention of air ingress into the reactor during the primary-pipe rupture accident. It is found that a safe cooling rate of the reactor core exists for the prevention of air ingress. The experimental results show that the natural circulation flow of air during the accident can be controlled by the method of helium gas injection into the reactor pressure vessel.     

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.     

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.     

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.     

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。     

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.     

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.     

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。     

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.     

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.     

Estimation of graphite density and mechanical strength variation of VHTR during air-ingress accident

This study focuses on predicting the changes in graphite density and mechanical strength in VHTR during the air-ingress accident via thermal hydraulic system analysis code. A simple graphite burn-off model was developed based on the similarities between a parallel electrical circuit and graphite oxidation. The developed model along with other comprehensive graphite oxidation models were integrated into the VHTR system analysis code, GAMMA. GT-MHR 600 MW t reactor was selected as a reference reactor. Based on the calculation, the main oxidation process was observed 5.5 days after the accident when followed by natural convection. The core maximum temperature reached 1430 °C, but never exceeded the maximum temperature criteria, 1600 °C. However, the oxidation process did significantly decrease the density of bottom reflector, making it vulnerable to mechanical stress. The stress on the bottom reflector is greatly increased because of the reduction of loaded surface area with graphite oxidation. The calculation proceeded until 11 days after the accident, resulting in an observed 4.5% decrease in density and a 25% reduction of mechanical strength.     

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。     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

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.     

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.     

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.     

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。     

反应堆用石墨材料的氧化

在反应堆运行过程中，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将导致石墨氧化，进而影响反应堆的正常运行和安全。文章主要对近期有关反应堆用石墨的氧化机理、氧化对石墨性能的影响、事故工况下的安全评估以及预防石墨氧化的措施等进行综述，并在此基础上指出，在辐照和氧化共同作用下的石墨材料性能变化是今后有关反应堆石墨研究的一个主要方面。     

Safety Analysis Results for Cryostat Ingress Accidents in ITER

Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.     

Increasing the power of the high temperature reactor module

To alleviate the economic problems of the modular pebble bed high temperature reactor, its design was modified in such a way that the power output was increased from 200 to 350 MWth. The core geometry was changed from cylindrical to annular, and the pressure vessel diameter was increased to 6.7 m. Control rods are placed in both the outer reflector and the graphite central column. In a safety analysis, loss of heat sink, loss of coolant and water ingress accident were examined. Reactor shutdown and decay heat removal take place passively, and the maximum fuel temperature stays theoreticallybelow 1600 °C, implying full retention of the fission products in the fuel elements. The central column has a diminishing effect on the positive reactivity effect of water ingress. A cost analysis shows that the specific investment costs of a four-module plant would decrease by 26% and the electricity generating costs would reduce by 19%.