Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

Severe accident analysis for Korean OPR1000 with MELCOR 1.8.6 was performed by adapting a mitigation strategy under different entry conditions of Severe Accident Management Guidance (SAMG). The analysis was focused on the effectiveness of the mitigation strategy and its adverse effects. Four core exit temperatures (CETs) were selected as SAMG entry conditions, and Small Break Loss of Coolant Accident (SBLOCA), Station Blackout (SBO), and Total Loss of Feed Water (TLOFW) were selected as postulated scenarios that may propagate into severe accidents. In order to delay reactor pressure vessel (RPV) failure, entering the SAMG when the CET reached 923 K, 923 K, and 753 K resulted in the best results for SBLOCA, SBO, and TLOFW scenarios, respectively. This implies that using event-based diagnosis for severe accidents may be more beneficial than using symptom-based diagnosis. There is no significant difference among selected SAMG entry conditions in light of the operator's available action time before the RPV failure. Potential vulnerability of the RPV due to hydrogen generation was analyzed to investigate the foreseeable adverse effects that act against the accident mitigation strategies. For the SBLOCA cases, mitigation cases generated more hydrogen than the base case. However, the amount of hydrogen generated was similar between the base and mitigation cases for SBO and TLOFW. Hydrogen concentrations of containment were less than 5% before RPV failure for most cases.     

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.     

Effect of wall wettability condition on drift-flux parameters in lead–bismuth two-phase flow in circular and annular bubble columns

Lead–bismuth two-phase flow in a cylindrical vessel and annulus was experimentally investigated by varying the surface wettability of the vessel wall. The test section used in this study was a cylindrical stainless vessel with/without inner sleeve to change the hydraulic diameter. Volume-averaged void fraction was measured by varying the surface wettability of the test section, which was enhanced by using a soldering flux. Measured void fraction was compared with existing two-phase flow correlations and with one-dimensional theoretical simulations assuming one-dimensional drift-flux model. From experimental results, measured distribution parameters of the lead–bismuth two-phase flow are much larger than that of ordinary two-phase flow regardless of the surface wettability. In the present work, the one-dimensional analysis was carried out for the cylindrical vessel to reproduce the distribution parameter. From the simulation results, predicted value for the cylindrical vessel showed good agreement with experimental results. However, in annulus, the distribution parameters in annulus were underestimated by the present model. It was suggested that, in case of annulus, steeper void fraction profile might be formed near the inner surface for poor wettability condition.     

严重事故下安全壳卸压箱隔间氢气浓度场模拟

根据MELCOR程序对全厂断电诱发的严重事故下安全壳内各隔间的氢气浓度分布的计算结果,参考美国联邦法规关于氢气控制和风险分析的标准,分析安全壳内氢气的燃烧风险。结果表明：安全壳内平均氢气浓度不会导致整体性氢气燃烧,但存在局部燃烧的风险。通过CFD程序对氢气浓度较高的卸压箱隔间进行氢气释放和空间气体流动过程的模拟,得到更细致的卸压箱隔间内氢气浓度场分布,给出氢气聚集区域的准确位置,为采取严重事故缓解措施,设计氢复合器布置方案提供了参考依据。     

严重事故下正常余热排出系统向堆芯注水策略分析

选取导致堆芯熔化频率最高的始发严重事故--直接注入（DVI）管线断裂事故,以及典型高压熔堆事故--丧失主给水始发事故(LOFW),利用MAAP4程序,分析反应堆堆芯热工水力行为,并对正常余热排出系统（RNS）堆芯注水策略的有效性与负面效应进行评估。分析结果表明,在DVI管线断裂事故和LOFW严重事故序列中,利用RNS进行堆芯注水可有效终止堆芯熔化进程,维持堆芯长期冷却。但堆芯再淹没会产生更多的氢气,存在增加安全壳氢气燃烧风险的可能性。此外通过分析利用严重事故管理导则中辅助计算文件给出的堆芯最小流量实施堆芯注水策略,讨论注水流量对堆芯冷却的影响,结果表明,在实施堆芯注水策略时,建议在系统允许的情况下采用更高的流速进行堆芯冷却。     

Experimental investigation of weir instability in main vessel cooling system of 1/4 FBR model

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5 m³/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion.     

Elastic-plastic FEM-analysis of a nozzle corner crack and discussion of the results by some fracture mechanics concepts

This paper presents the results of an elastic-plastic threedimensional finite element analysis for a nozzle corner crack in a pressurized reactor test vessel. The calculations were performed by the finite element program ADINA incorporating von Mises' yield condition and isotropic hardening. The crack plane was taken parallel to the axis of the vessel and the crack front straight and perpendicular to the symmetry line of the nozzle corner in order to obtain the worst position for a nozzle corner crack. The calculations were performed up to that pressure level where general yield of the ligament in the nozzle corner section takes place. The results of the finite element analysis are compared with figures obtained from analytical procedures of elastic-plastic fracture mechanics.     

Thermal and structural analysis of light-water reactor vessel in-place annealing

A major concern for a localized annealing of a reactor pressure vessel is the dimensional stability both during and after the annealing cycle. The effects of residual stresses in the vessel after such a localized heat treatment also need to be assessed. A two-dimensional, axisymmetric finite element study has been conducted using a typical vessel design, and the results indicate that there is no problem with the vessel itself. There is, however, a problem with the attached primary piping in that during the heating operation the nozzle region sees a thermal gradient which produces a small bending, rotation which can plastically deform the piping. Further analytical studies were conducted using an expanded heating zone in an attempt to reduce this axial gradient. Even though the overall temperature gradient was minimized, the actual gradient acting at the nozzle was only slightly reduced and plastic bending can still occur. This temperature gradient problem needs further analytical study but can be resolved with proper heating conditions which are well within the industrial heat treating state-of-the-art for temperature control and monitoring.     

Investigation on Slit Jet through Upper Internal Structure (UIS) in Highly Compact Vessel of Sodium-Cooled Fast Reactor

An advanced loop-type sodium-cooled fast reactor has been developed by the Japan Atomic Energy Agency. The upper internal structure (UIS) above the core is a key component where control rod guide tubes are housed. A radial slit is set in the UIS to simplify the fuel-handling system and to reduce the reactor vessel diameter. A high-velocity upward flow is formed in the UIS slit. This slit jet influences thermal hydraulic issues in the reactor vessel. A water experiment was carried out to understand the flow field in the UIS, which is composed of the control rod guide tubes and several horizontal perforated plates with a slit. A refractive index matching method was applied to visualize the flow in such a complex geometry. Velocity measurement using particle image velocimetry showed that the velocity in the UIS slit was accelerated by the multiple slits and kept at a high value at the mid-height of the reactor upper plenum. A numerical simulation was carried out for this complex geometry of the UIS to obtain an adequate simulation method. A comparison between the experimental and analytical velocity profiles showed that the numerical simulation is highly applicable.     

Analysis of steam and hydrogen distributions with PAR mitigation in NPP containments

The 3-D-field code, GASFLOW is a joint development of Forschungszentrum Karlsruhe and Los Alamos National Laboratory for the simulation of steam/hydrogen distribution and combustion in complex nuclear reactor containment geometries. GASFLOW gives a solution of the compressible 3-D Navier–Stokes equations and has been validated by analysing experiments that simulate the relevant aspects and integral sequences of such accidents. The 3-D GASFLOW simulations cover significant problem times and define a new state-of-the art in containment simulations that goes beyond the current simulation technique with lumped-parameter models. The newly released and validated version, GASFLOW 2.1 has been applied in mechanistic 3-D analyzes of steam/hydrogen distributions under severe accident conditions with mitigation involving a large number of catalytic recombiners at various locations in two types of PWR containments of German design. This contribution describes the developed 3-D containment models, the applied concept of recombiner positioning, and it discusses the calculated results in relation to the applied source term, which was the same in both containments. The investigated scenario was a hypothetical core melt accident beyond the design limit from a large-break loss of coolant accident (LOCA) at a low release location for steam and hydrogen from a rupture of the surge line to the pressurizer (surge-line LOCA). It covers the in-vessel phase only with 7000 s problem time. The contribution identifies the principal mechanisms that determine the hydrogen mixing in these two containments, and it shows generic differences to similar simulations performed with lumped-parameter codes that represent the containment by control volumes interconnected through 1-D flow paths. The analyzed mitigation concept with catalytic recombiners of the Siemens and NIS type is an effective measure to prevent the formation of burnable mixtures during the ongoing slow deinertization process after the hydrogen release and has recently been applied in backfitting the operational German Konvoi-type PWR plants with passive autocatalytic recombiners (PAR).     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.     

Gravity-driven countercurrent two-phase flow during filling of an unvented vessel

Countercurrent two-phase flow associated with filling of sealed vessels via gravity-driven liquid injection through inclined channels was experimentally studied and analytically modeled. Experiments were performed using transparent tubular test sections connected at one end to the bottom of a large, open water tank, and at the other end to an unvented tank. The test section parameters (including the channel diameter (1.27–2.54cm), length (30.5–122 cm), angle of inclination with respect to horizontal plane (0–30°), and the empty volume in the sealed vessel) were systematically varied. Flow regimes in the test section were recorded and transient flow rates were measured during the experiments. Oscillatory, and intermittent stratified slug, were dominant flow regimes in most tests. The quasi-steady liquid superficial velocity in the test section was sensitive to the test section dimensions, and varied in the range 0.04–0.95 m s⁻¹. These flow regimes were mechanisally modeled. The models are shown to satisfactory predict the measured hydrodynamic parameters.     

Numerical analysis of hydrogen risk mitigation measures for support of ITER licensing

In the ITER wet bypass scenario, water leakage, air ingress and hot dust (Be, W, and C) in the vacuum vessel could generate combustible hydrogen-air-steam mixture. Hydrogen combustion may threaten the integrity of the ITER VV and lead to radioactivity release. To prevent hydrogen energetic combustion, nitrogen injection system in VV and hydrogen recombination system in the pressure suppression tank (ST) were proposed. The main objectives of this analysis are to study the distribution of hydrogen-air-steam mixtures in the ITER sub-volumes, to investigate the feasibility of the nitrogen injection system to fully inert the atmosphere in the VV and to evaluate the capability and efficiency of the hydrogen recombination system to remove hydrogen in the ST. 3D computational fluid dynamics (CFD) code GASFLOW was used to calculate the evolution of the mixtures and to evaluate the hydrogen combustion risks in the ITER sub-volumes. The results indicate that the proposed hydrogen risk mitigation systems will generally prevent the risks of hydrogen detonation and fast deflagration. However, the atmosphere in ITER sub-volumes cannot be completely inerted at the early stage of the scenario. Slow deflagrations could still generate quasi-static pressures above 1 bar in the VV. The structural impact of the thermal and pressure loads generated by hydrogen combustions will be investigated in future studies.     

Containment venting as a mitigation technique for BWR MARK I plant ATWS

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it.Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure. For both cases the operators are assumed to take action to intentionally restrict injected flow such that fuel in the upper part of the core would be steam cooled. Resulting fuel temperatures are estimated with an off-line calculation and found to be acceptable.     

严重事故氢气燃爆缓解措施的初步研究

轻水堆核电站发生严重事故时,氢气的大体积氢燃爆可能会严重威胁安全壳的完整性.氢气点火器与氢气复合器是2种严重事故下的氢气燃爆缓解设备.本文分别研究了3种氢气燃爆缓解措施,包括仅采用氢气点火器、仅采用氢气复合器和采用氢气复合器结合点火器.结果表明,采用氢气复合器结合点火器的方式可以安全、持续、有效地降低大体积氢燃爆带来的风险.     

Permeation Analysis of Tritium Through the Vessel Wall of a Titanium Hydride Storage Container

Abstract

A preliminary design for a stainless steel vessel for the long-term storage of hydrogen isotopes has been proposed. The immobilised hydrogen, as a titanium hydride, could be stored in a stainless steel vessel for this application. The vessel, as a primary package, is designed to form titanium hydride and to contain the hydrogen isotopes and helium-3 produced from the decay of tritium. In order to predict the possibility of contamination and the deterioration of the mechanical properties, a numerical diffusion analysis calculation of the hydrogen isotopes and helium inside the stainless steel vessel was carried out. Numerical results showed that a negligible amount of tritium would be released by permeation through a 0.7 cm thick vessel wall at normal conditions over the entire period of the storage. When the vessel is heated up to a temperature of 600^°C for the routine conditions of activation or exothermic hydriding, tritium loss or contamination would be of little concern. However, if the vessel were exposed to fire conditions with a temperature of 800^°C, permeation of hydrogen through the vessel wall would result in a serious increase in the amount of tritium escaping, in a very short time.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

Demonstrative testing of honeycomb passive autocatalytic recombiner for nuclear power plant

Passive autocatalytic recombiners (PAR) are widely being used as hydrogen control device in the current and advanced light water reactors (ALWRs). The PARs lend themselves to very effective means of circumventing buildup of combustible or detonable hydrogen gas mixtures in the reactor containment. Korea Nuclear Technology Inc. has recently developed a new PAR system with high porous catalyst material in the shape of honeycomb. The honeycomb PAR catalyst has a design characteristic of improved hydrogen removal performance by increasing the surface area and enhancing the flow rate through the catalyst at the same time, without increasing PAR size compared to the conventional PARs. The experimental study was focused on the development of the hydrogen depletion rate correlation of the honeycomb PAR. Two different sizes of PARs, KPAR-40 and KPAR-T2, have been employed in the tailor-made Integral Test Facility and Performance Test Facility. Multiple tests were conducted in various conditions of pressure, temperature, and hydrogen concentration. The hydrogen depletion rate correlation and the PAR performance constant were determined from the experimental results, which can be applied to the honeycomb PAR system. Also determined was the scale effect due to the PAR size, i.e., the number of catalysts in a PAR.     

Organic decomposition in supercritical water by an aid of ruthenium (iv) oxide as a catalyst-exploitation of biomass resources for hydrogen production-

The present paper deals with the production of hydrogen from biomass in supercritical water in the presence of ruthenium(IV) dioxide, RuO₂, as a catalyst. Experiments were carried out to gasify cellulose and lignin, the mixtures of cellulose and lignin for the simulation of biomass, and finally actual biomass such as pulp, waste paper and paper sludge. Each sample was loaded in a reaction vessel under argon atmosphere at 450°C and around 44 MPa for 120 min. The gas produced was quantitatively analyzed by an on-line gas chromatography. It was found that RuO₂ enhanced the decomposition ratios for cellulose and pulp, but not for lignin and lignin containing compounds. Obviously, lignin deactivated the catalytic effect of RuO₂. The hydrogen production ratios in gasses produced were 15.0% from cellulose, 14.1% from pulp, 21.0% from the mixture of cellulose and lignin, 16% from waste paper and 27% from and paper sludge, respectively. We have concluded that the present method would be feasible for the production of hydrogen.