基于DLL的GOTHIC 8.0程序氢气复合器模拟方法研究

非能动氢气复合器已广泛应用于核电厂氢气威胁的缓解和消除。本文通过对GOTHIC 8.0程序进行二次开发,采用外部动态链接库（DLL）编译、调用的方式,精确模拟了非能动氢气复合器的实际消氢能力,进而将采用该方法计算得到的消氢结果分别与公式计算、MAAP5程序算例计算结果进行比较,结果符合度高,验证了该方法的合理性。本文提供的模拟方法不仅为安全壳氢气风险缓解分析提供了新方法,也为GOTHIC程序开发提供了新思路。     

GOTHIC程序在海洋核动力平台反应堆舱安全分析领域的适用性评估

针对海洋核动力平台反应堆舱热工水力分析程序缺乏的现状,以一回路失水事故（LOCA）下反应堆舱压力响应为评价基准,基于安全壳现象识别与排序表（PIRT）分析方法,通过开展LOCA下反应堆舱热工水力现象识别、现象分级研究,建立了反应堆舱PIRT。通过开展GOTHIC程序模型验证矩阵与PIRT的匹配性分析,确认GOTHIC程序在海洋核动力平台反应堆舱热工水力分析领域的适用性。本文分析方法对其他安全分析程序在核电等领域的跨领域适用性评估具有一定参考价值。      

Three-gas mixture plume inducing mixing and stratification in a multi-compartment containment

A project named SETH (SESAR Thermal Hydraulics (SESAR: Senior Group of Experts on Nuclear Safety Research)) has been performed under the auspices of 15 OECD countries, with the aim of creating an experimental database suitable to assess the 3D code capabilities in analyzing physical key phenomena relevant for LWR safety analysis. This paper presents the results of OECD/SETH Project Test 25 performed in the PANDA facility (located at PSI in Switzerland). Test 25 consists of two phases for duration of approximately 2 h for each phase. During Phase 1 a steam-helium mixture (helium is used to simulate hydrogen) is released in a containment compartment initially filled with only air. This phase simulates the hydrogen released in a postulated accident due to the fuel cladding metal-water (MW) reaction. During Phase 2 only steam is released in the containment and the scenario that occurs after the MW reaction is exhausted is addressed. Temperature and gas concentration measurements obtained at several locations in the containment compartments allow the recognition of the complex stratification pattern evolution during the test period. The analysis of Test 25 carried out with the GOTHIC code is also presented in this paper. The analysis shows that the prediction of three-gas (hydrogen, steam, and air) stratification pattern in a multi-compartment geometry for scenarios characterized by the evolution of the density difference between and inside compartments and in particular of the hydrogen accumulation in a dead-end volume is a challenging task also for a code having 3D capabilities.     

Combustion of stratified hydrogen-air mixtures in the 10.7 m combustion test facility cylinder

This paper presents preliminary results from hydrogen concentration gradient combustion experiments in a 10.7 m³ cylinder. These gradients, also referred to as stratified mixtures, were formed from dry mixtures of hydrogen and air at atmospheric temperature. Combustion pressures, burn fractions and flame speeds in concentration gradients were compared with combustion of well-mixed gases containing equivalent amounts of hydrogen. The studied variables included the quantity of hydrogen in the vessel, the steepness of the concentration gradient, the igniter location, and the initial concentration of hydrogen at the bottom of the vessel.Gradients of hydrogen and air with average concentrations of hydrogen below the downward propagation limit produced significantly greater combustion pressures when ignited at the top of the vessel than well-mixed gases with the same quantity of hydrogen. This was the result of considerably higher burn fractions in the gradients than in the well-mixed gas tests. Above the downward propagation limit, gradients of hydrogen ignited at the top of the vessel produced nearly the same combustion pressures as under well-mixed conditions; both gradients and well-mixed gases had high burn fractions. Much higher flame speeds were observed in the gradients than the well-mixed gases.Gradients and well-mixed gases containing up to 14% hydrogen ignited at the bottom of the vessel produced nearly the same combustion pressures. Above 14%, hydrogen, gradients produced lower combustion pressures than well-mixed gases having the same quantity of hydrogen. This can be attributed to lower burn fractions of fuel from the gradients compared with well-mixed gases with similar quantities of hydrogen. When ignited at the bottom of the vessel, 90%, of a gradient's gases remained unburned until several seconds after ignition. The remaining gases were then consumed at a very fast rate.     

GOTHIC局部氢气风险三维分析研究

目前的氢气风险分析中,主要采用一体化严重事故分析程序进行分析计算。日本福岛事故后,对氢气风险分析提出了更高的要求。为了实现对集总参数程序的有益补充,本文开展了GOTHIC程序氢气风险三维分析的研究。利用GOTHIC建立了局部氢气风险三维分析模型,在模型验证的基础之上,对典型严重事故序列下的氢气风险进行三维分析研究。研究表明:安全壳上部空间气流混合较好,氢气分层并不是非常明显;对于核电厂压力容器直接注射(DVI)管道破口所在的非能动堆芯冷却系统隔间B(PXS-B),由于破口以下部分区域被水淹没,破口以上区域的氢气浓度较高,但氢气风险较小。     

GOTHIC局部氢气风险三维分析研究

Integrated severe accident code is used to analyze the hydrogen risk in current safety assessment. After Fukushima accident, higher requirements are placed on hydrogen risk analysis. In order to supplement the lumped parameter analysis, three dimension hydrogen risk analysis method using GOTHIC is studied. Local three dimension hydrogen risk model is constructed by GOTHIC. Based on model validation, typical severe accident cases are chosen to analyze the hydrogen distribution. The results show that, hydrogen and other gas are mixed well in the upper compartment of the containment, and hydrogen stratification phenomenon is not obvious. For DVI rupture accident in PXS-B, the lower area of the break is flooded, and the hydrogen concentration for the upper area of the break is large, however, the hydrogen risk is little.     

应用GOTHIC8.0程序模拟非能动安全壳冷却系统冷凝和蒸发现象的适用性研究

在非能动安全壳冷却系统（PCS）设计基准事故的排热过程中,安全壳内壁面蒸汽冷凝现象和安全壳外壁面水膜蒸发现象是两种非常关键的排热途径。本文应用GOTHIC8.0程序模拟了安全壳内壁面蒸汽冷凝和安全壳外壁面水膜蒸发传热过程,并通过蒸汽冷凝试验和水膜蒸发试验数据,对GOTHIC程序的模拟结果进行了分析和评价。研究结果表明：GOTHIC程序的蒸汽冷凝模型可较好地模拟蒸汽冷凝传热现象;水膜蒸发模型明显低估了水膜蒸发换热量,这对设计基准事故安全壳完整性分析是非常保守的,建议对GOTHIC程序进行适当开发,更好地模拟水膜蒸发换热过程。     

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.     

应用GOTHIC8.0程序分析AP1000核电厂PCS传热传质过程

研究应用GOTHIC8.0程序分析AP1000核电厂非能动安全壳冷却系统(PCS)传热传质过程,通过理论计算和程序分析两种方式对分析结果进行比较和评价。研究结果表明:GOTHIC8.0程序的DLM-FM模型适用于模拟安全壳内蒸汽在安全壳内壁面的冷凝传热传质过程,Film模型适用于模拟安全壳外水膜的蒸发传热传质过程。GOTHIC8.0程序可用于分析AP1000核电厂PCS传热传质过程,为AP1000核电厂在设计基准事故(DBA)下安全壳响应分析提供了另一种可行的工具。     

Conceptual design and analysis of a semi-passive containment cooling system for a large concrete containment

An internal evaporator-only (IEO) concept has been developed as a semi-passive containment cooling system for a large dry concrete containment. The function of this system is to keep the containment integrity by maintaining the internal pressure not to exceed ultimate design pressure, i.e. 0.83 MPa (120 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. The ability of the concept to protect the containment was evaluated for the design basis accident (DBA) large break loss of coolant accident (LB LOCA) and severe accident scenarios (LB LOCA without Emergency Core Cooling System (ECCS) and containment spray flow, 100% zirconium oxidation and complete hydrogen combustion). All were modeled using the GOTHIC computer code. It was concluded that a practical system requiring four IEO loops could be utilized to meet design criteria for severe accident scenarios.     

Dissolution of uranium metal without hydride formation or hydrogen gas generation

This study shows that metallic uranium will cleanly dissolve in carbonate-peroxide solution without generation of hydrogen gas or uranium hydride. Metallic uranium shot, 0.5–1 mm diameter, was reacted with ammonium carbonate–hydrogen peroxide solutions ranging in concentration from 0.13 M to 1.0 M carbonate and 0.50 M to 2.0 M peroxide. The dissolution rate was calculated from the reduction in bead mass, and independently by uranium analysis of the solution. The calculated dissolution rate ranged from about 4 × 10⁻³ to 8 × 10⁻³ mm/h, dependent primarily on the peroxide concentration. Hydrogen analysis of the etched beads showed that no detectable hydrogen was introduced into the uranium metal by the etching process.     

Laser Rayleigh measurement of mixing processes and control of hydrogen combustion using quenching meshes

Two issues concerning hydrogen combustion under a severe accident scenario are addressed: (1) a laser Rayleigh scattering technique to investigate hydrogen mixing processes; and (2) the installation of metallic meshes between compartments to control and isolate hydrogen combustion within a single compartment. The Rayleigh scattering techniques are tested to determine hydrogen/air mixing processes locally and temporally as a non-intrusive probing method. To simulate mixing processes, helium is injected into a chamber filled with n-butane. Results show that helium concentration can be successfully monitored with sufficiently fast responses. Isolation and control of hydrogen burning is simulated by installing metallic meshes between compartments. Hydrogen is injected into one compartment and subsequently transported to the second compartment. Two sets of experiments are conducted with and without installing metallic meshes between the compartments. With the mixture ignited near the second compartment outlet, hydrogen combustion can be successfully contained within the second compartment with meshes, while flame propagates to the first compartment when meshes are not installed. These results demonstrate that hydrogen combustion can be controlled and isolated by installing meshes locally such that unwanted rapid pressure rise in a containment can be prevented. It also suggests the applicability of meshes for equipment survivability and protection from flame propagation by enclosing equipments with properly designed meshes.     

Implementation of the renormalization group (RNG) k–ε turbulence model in GOTHIC/6.lb: solution methods and assessment

In GOTHIC, the standard k–ε model is used to model turbulent flow. In an attempt to enhance the turbulence modelling capabilities of the latest 6.1b version of the code for simulation of mixing driven by highly buoyant discharges, we implemented the renormalization group (RNG). This model, which is only implemented in the “gas” phase, was tested with different simple test-problems and its predictions were compared to the corresponding ones obtained when the standard k–ε model was used. As expected, the RNG model is less dissipative than the standard k–ε model. However, comparison of code predictions with experimental measurements show that the predictions of the RNG model are better only near the source. Finally, we show how in the framework of this latest code version, one can implement higher-order non-linear turbulence models.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 × 8 BWR fuel assembly are low up to 60 GWd/t.     

Air oxidation of Zircaloy-4 in the 600-1000 °C temperature range: Modeling for ASTEC code application

Progress in the treatment of air oxidation of zirconium in severe accident (SA) codes are required for a reliable analysis of severe accidents involving air ingress. Air oxidation of zirconium can actually lead to accelerated core degradation and increased fission product release, especially for the highly-radiotoxic ruthenium. This paper presents a model to simulate air oxidation kinetics of Zircaloy-4 in the 600-1000 °C temperature range. It is based on available experimental data, including separate-effect experiments performed at IRSN and at Forschungszentrum Karlsruhe. The kinetic transition, named “breakaway”, from a diffusion-controlled regime to an accelerated oxidation is taken into account in the modeling via a critical mass gain parameter. The progressive propagation of the locally initiated breakaway is modeled by a linear increase in oxidation rate with time. Finally, when breakaway propagation is completed, the oxidation rate stabilizes and the kinetics is modeled by a linear law. This new modeling is integrated in the severe accident code ASTEC, jointly developed by IRSN and GRS. Model predictions and experimental data from thermogravimetric results show good agreement for different air flow rates and for slow temperature transient conditions.     

An evaluation of containment inerting and air dilution systems as methods for post-accident hydrogen control in BWRS

Hydrogen control is important in post-accident situations because of possibilities for containment rupture due to hydrogen deflagration or detonation. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control failure rate (air dilution: 8.3 × 10⁻²−1.1 × 10⁻²; nitrogen inerting: 1.3 × 10⁻²−2 × 10⁻³). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify ‘unidentified’ leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations-related disadvantages.     

水蒸气对氢气燃烧影响的数值研究

核电站严重事故下,氢气的燃烧风险是影响安全壳完整性的重要因素,而水蒸气的存在对氢气、空气混合气体的燃烧会产生重要的影响。本文采用CAST3M软件,对局部小空间内氢气的燃烧特性以及水蒸气的影响进行研究。首先对THAI装置的典型实验工况进行模拟,表明了相关燃烧模型的可用性。然后将高度为6 m、直径为2.2 m的圆柱空间作为燃烧域,对其分别计算了8%、10%、12%氢气浓度下的燃烧,并与添加25%水蒸气的相应工况进行了对比。通过对燃烧域的温度、压力以及火焰传播速度的分析,表明添加水蒸气后燃烧产生的最大压力下降,火焰的最大温度下降,火焰传播的速度下降。研究表明,水蒸气的存在对氢气的燃烧具有抑制作用,能有效降低氢气燃烧产生的后果。     

Development and steady state level experimental validation of TASS/SMR core heat transfer model for the integral reactor SMART

An advanced integral pressurized water reactor (PWR) of a small size (330 MWt) is being developed by the Korea Atomic Energy Research Institute (KAERI). The purposes of the reactor are a sea water desalination and an electricity generation. To enhance its safety, many advanced design concepts are introduced such as a passive residual removal system and a low power density core. For the safety validation of the designed reactor, a system analysis code named TASS/SMR, was developed. TASS/SMR code uses a one dimensional node/path modeling for the thermal hydraulic calculation and point kinetics for the core power calculation. The code also has specific models for the developed integral reactor, such as a helical tube heat transfer model and a passive residual heat transfer model. One of the important models for the safety or performance calculation is the core heat transfer model. The core heat transfer model of TASS/SMR was developed to meet the requirements of the 10 CFR 50 appendix K EM model as well as the realistic models. The developed model was validated with experimental data. The results show that the model predicts the heat transfer phenomena in the reactor core with a reasonable conservatism.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.