非能动氢复合器性能验证试验方法研究

消氢启停阈值和消氢速率是非能动氢复合器的关键性能参数。本文设计了一种直观方便的非能动氢复合器性能验证试验方法:将非能动氢复合器放于密闭容器中，并通入氢气，只要氢复合器启动消氢反应且整条消氢过程曲线在给定值直线A以下，则验证了启动阈值不大于给定值A；只要消氢过程曲线最终的水平段在给定值直线B以下，则验证了停止阈值不大于给定值B；只要氢复合器达到稳定消氢状态，通入容器的氢气质量流量即为消氢速率。本文设计并搭建了试验装置，采用非能动氢复合器样机PARQX-15进行消氢性能验证试验，成功验证了消氢启动阈值<2%（体积浓度，下同），停止阈值<0.5%，消氢速率大于536 g/h，证明了试验方法的实用性和有效性。      

Analysis of hydrogen depletion using a scaled passive autocatalytic recombiner

Hydrogen depletion tests of a scaled passive autocatalytic recombiner (PAR) were performed in the Surtsey test vessel Germany) at Sandia National Laboratories (SNL). The experiments were used to determine the hydrogen depletion rate of a PAR in the presence of steam and also to evaluate the effect of scale (number of cartridges) on the PAR performance at both low and high hydrogen concentrations.     

Scaling up experiments of honeycomb catalysts for oxidation of hydrogen and methane gases

For a large volume air cleanup system in nuclear fusion facility honeycomb type oxidizing catalyst offers more useful advantage in terms of their low pressure drop than the conventional particle packed bed. Scaling up examination of 35 times large volume Pd catalyst deposit on metal honeycomb Al–Cr-steel alloy had been performed comparing with fundamental small size honeycomb. The catalyst was heated up to over 400 °C with stepwise by combined the inner and external heaters under various flow rate of processing gas containing various concentrations of hydrogen and methane. Then oxidizing reaction rates and the catalytic activation energies were evaluated and thermal property of catalyst bed was estimated considering exothermic oxidizing reaction in the process gas. As a result it was revealed that the scaling up honeycomb catalyst bed would be practically useful for the air cleanup system.     

Ensuring the long-term functionality of passive auto-catalytic recombiners under operational containment atmosphere conditions—An interdisciplinary investigation

In order to sustain the structural integrity of the containment and other safety relevant components i.e. to avoid a detonation of the hydrogen-air mixture generated during a severe accident in light water reactors, passive auto-catalytic recombiners (PAR) are used for hydrogen removal in many European nuclear power plants (NPP). In 1999, the German NPP Emsland (KKE) was equipped with 58 PAR of AREVA design as an internal accident management measure for a beyond-design accident. Since that time the recombiners are in a stand-by state. As the catalyst elements are exposed to various airborne substances during normal plant operation their function is controlled periodically by testing selected catalyst sheets in a specially designed device. Under the conservative test conditions during this procedure some catalyst sheets showed a delayed responding behavior. First internal analysis gave indication of a beginning fouling on the catalytic surface.The aim of a precautionary investigation performed in cooperation between KKE, Forschungszentrum Juelich and RWTH Aachen University was to characterize the composition of the fouling and to correlate it with potential sources within the containment.In the framework of the investigation the reports of the periodic inspections were analyzed and appropriate sample sheets were selected from the installation. These samples were subjected to a comprehensive chemical surface analysis in order to identify effects like thermal sintering, poisoning or a blocking of the catalytic surface (Baerns, M., 2004. Basic Principles in Applied Catalysis, Springer Verlag). Along with the chemical analysis the catalytic activity of the samples was assessed in several test series in order to correlate the chemically quantified deposition on the catalyst samples with the characteristics of the start-up and the steady-state performance of the recombination reaction. In a final step, possible sources of the fouling were analyzed with regard to their possible contribution to the phenomena. According to the results achieved, measures have been implemented at KKE in order to optimize procedures and to enhance the performance of the PARs.     

反应堆非能动氢气复合器严重事故下消氢性能研究

非能动氢气复合器（PAR）是核电厂主要的消氢措施。在严重事故下,某些裂变反应产生的特殊杂质(如I₂、CsI、CO等)可能对催化剂产生有害的影响,为保证PAR消氢性能的可靠性,中毒机理研究十分必要。本文根据相关试验研究结果,从铂、钯金属原子分布结构、催化反应性面积两个方面对催化剂中毒效应进行机理分析,研究得出催化剂中金属原子的分布结构、进入PAR的实际毒物浓度等均是影响催化剂中毒效应的关键因素。     

Quantitative Evaluation of Heat Transfer Enhancement due to Radial Power Distribution during Reflood Phase of PWR-LOCA

Experimental studies using Slab Core Test Facility (SCTF) and Cylindrical Core Test Facility (CCTF) indicated that the degree of heat transfer enhancement due to the radial power distribution during the reflood phase of a PWR-LOCA was governed mainly by the radial power ratio itself and less dependent on the shape of radial power distribution within the maximum power ratio of 1.36. The experimental condition covering the wide ranges of the reflood phase and the scale of core radius from 1/4.6 to 1/1 had little effect on the two- dimensional heat transfer behavior. The heat transfer coefficient under nonuniform radial power distribution was expressed as a sum of the heat transfer coefficient obtained under a complete mixing condition and an additional value given by an empirical correlation based on the SCTF results. The temperature rise at the peak power rod calculated with this expression tended to be lower than that calculated with the complete mixing model used in a reflood analysis code REFLA. That is, the complete mixing model was proved to give a conservative result under a nonuniform radial power distribution condition in a full size core.     

An Overview of INEL Fusion Safety R&D Facilities

The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.     

Studies on innovative hydrogen recombiners as safety devices in the containments of light water reactors

In order to prevent the containment and other safety relevant components from incurring serious damage caused by a detonation of the hydrogen/air-mixture generated during a severe accident in light water reactors (LWR) passive autocatalytic recombiners (PAR) are used for hydrogen removal in an increasing number of European plants. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces generating steam and heat.

Experimental investigations at several research facilities indicate that existing PAR systems bear the risk of igniting the gaseous mixture due to an overheating of the catalyst elements caused by strong reaction heat generation. Innovative devices could overcome existing limitations making use of the knowledge deduced from experiments performed at the REKO facilities at Forschungszentrum Juelich (FZJ).

The paper analyses the mechanisms of the thermal behaviour of catalytic plate-type recombiners and presents experimental results on existing and innovative devices for hydrogen removal introducing the modular recombiner concept.     

Numerical simulation using CFD software of countercurrent gas-liquid flow in a PWR hot leg under reflux condensation

In order to improve the countercurrent flow model of a transient analysis code, countercurrent air-water tests were previously conducted using a 1/15 scale model of the PWR hot leg and numerical simulations of the tests were carried out using the two-fluid model implemented in the CFD software FLUENT 6.3.26. The predicted flow patterns and CCFL characteristics agreed well with the experimental data. However, the validation of the interfacial drag correlation used in the two-fluid model was still insufficient, especially regarding the applicability to actual PWR conditions. In this study, we measured water levels and wave heights in the 1/15 scale setup to understand the characteristics of the interfacial drag, and we considered a relationship between the wave height and the interfacial drag coefficient. Numerical simulations to examine the effects of cell size and interfacial drag correlations on numerical predictions were conducted under PWR plant conditions. Wave heights strongly related with the water level and interfacial drag coefficient, which indicates that the interfacial drag force mainly consists of form drag. The cell size affected the gas velocity at the onset of flooding in the process of increasing gas flow rate. The gas volumetric fluxes at CCFL predicted using fine cells were higher than those using normal cells. On the other hand, the cell size did not have a significant influence on the process of decreasing gas flow rate. The predictions for the PWR condition using a reference set of interfacial drag correlations agreed well with the Upper Plenum Test Facility data of the PWR scale experiment in the region of medium gas volumetric fluxes. The reference interfacial drag correlations employed in this study can be applied to the PWR conditions.     

Downcomer and tie plate countercurrent flow in the Upper Plenum Test Facility (UPTF)

The limitation of vertical steam-water countercurrent flow, called flooding, is important for the operation of Emergency Core Cooling (ECC) Systems in Nuclear Reactors. The ECC water injection flooding behavior is scale dependent and the reactor size behavior cannot be extrapolated from small scale data.A new flooding correlation is presented, based on the classic flooding equation where the effects of gravity, interphase momentum exchange, and instability of the gas/liquid interface are considered. Development of a new correlation was necessary in order to correlate the reactor scale downcomer and upper tie plate countercurrent flow data gained in the Upper Plenum Test Facility (UPTF). The new correlation is an extension of the well-known Wallis-type and Kutateladze-type correlations. Each of the three correlations is valid for the experimental facilities on a certain scale. The range of applicability for each of the three correlations is defined for the case of downcomer and tie plate countercurrent flow.     

A PANDA integral test on the effect of light gas on a Passive Containment Cooling System (PCCS)

As part of the Euratom project TEMPEST (Testing and Enhanced Modelling of Passive Evolutionary Systems Technology for Containment Cooling), a series of five tests was performed in the PANDA facility to experimentally investigate the distribution of hydrogen inside the containment and its impact on the performance of the Passive Containment Cooling System (PCCS) designed for the Economic Simplified Boiling Water Reactor (ESBWR). In a postulated severe accident a large amount of hydrogen could be released in the Reactor Pressure Vessel (RPV) as a consequence of the cladding Metal-Water (M-W) reaction and discharged together with steam to the Drywell (DW) compartment. The retention of hydrogen in the DW, instead to be vented in the Wetwell (WW), has a positive effect toward the mitigation of the system pressure build-up. Hydrogen retention in the DW is a consequence of the stratification phenomena driven by the steam-hydrogen density difference. The paper presents the experimental results of the integral Test T1.2 performed in the PANDA facility. Helium was used to simulate hydrogen and the specific PANDA facility configuration included a dead-end volume, allowing for retaining a portion of the released helium in the DW compartment. The results from Test T1.2 showed that the containment end pressure is mainly determined by the redistribution of non-condensable gas inside the containment system and the temporary deterioration of the PCCS performance during the helium release phase plays a minor role.     

Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.     

RELAP5-3D modeling of PWR steam generator condensation experiments at the Oregon State University APEX facility

This report describes modeling using RELAP5-3D of a series of six steam generator U-tube steam condensation (without non-condensable gas) tests conducted at the Oregon State University Advanced Plant Experiment Test Facility from 2005 through 2007. These tests were designed to evaluate steam condensation rates in a scaled pressurized water reactor steam generator at various primary and secondary side pressures and inlet steam mass flow rates. Comparisons between the experimental data and the RELAP5-3D model results are made to quantify the effectiveness of RELAP5-3D in handling steam condensation in U-tube steam generators. RELAP5-3D tends to over predict the condensation rate and heat transfer coefficient when compared against the experimental data when the code uses the laminar Nusselt correlation to determine the heat transfer coefficient. When RELAP5-3D results are used with the Shah correlation the comparison between the heat transfer coefficients is much improved.     

A detailed chemistry model for transient hydrogen and carbon monoxide catalytic recombination on parallel flat Pt surfaces implemented in an integral code

A detailed chemistry model has been adapted and developed for surface chemistry, heat and mass transfer between H₂/CO/air/steam/CO₂ mixtures and vertical parallel Pt-coated surfaces. This model is based onto a simplified Deutschmann reaction scheme for methane surface combustion and the analysis by Elenbaas for buoyancy-induced heat transfer between parallel plates. Mass transfer is treated by the heat and mass transfer analogy. The proposed model is able to simulate the H₂/CO recombination phenomena characteristic of parallel-plate Passive Autocatalytic Recombiners (PARs), which have been proposed and implemented as a promising hydrogen-control strategy in the safety of nuclear power stations or other industries. The transient model is able to approach the warm-up phase of the PAR and its shut-down as well as the dynamic changes within the surrounding atmosphere. The model has been implemented within the MELCOR code and assessed against results of the Battelle Model Containment tests of the Zx series. Results show accurate predictions and a better performance than traditional methods in integral codes, i.e. empirical correlations, which are also much case-specific. Influence of CO present in the mixture on the PAR performance is also addressed in this paper.     

Analysis of hydrogen isotopes absorption between liquid lithium and yttrium under dynamic conditions

The use of D–Li stripping reaction to generate multiple neutron is expected in the International Fusion Materials Irradiated Facility (IFMIF). Tritium generated by its side reaction needs to be recovered from liquid Li for safety. Y hot trap is expected as an effective purification system to recover tritium from the liquid Li loop. In IFMIF, since liquid Li circulates in a loop, it is important to trace tritium behavior in Y and liquid Li under dynamic conditions. In this study, the authors investigate the effectiveness of tritium recovery from liquid Li under dynamic conditions of Y hot trap. Hydrogen absorption in Li instead of tritium is investigated under dynamically fluidized conditions by stirring liquid Li with an arbitrary rotational rate. The experimental results prove that the hydrogen absorption rate from liquid Li to Y is independent of the rotational rate. This means that the rate-determining step is hydrogen diffusion in Y.     

Room-Temperature Reactor Packed with Hydrophobic Catalysts for the Oxidation of Hydrogen Isotopes Released in a Nuclear Facility

A room-temperature reactor packed with hydrophobic catalysts for the oxidation of hydrogen isotopes released in a nuclear facility will contribute to nuclear safety. The inorganic-based hydrophobic Pt catalyst named H1P has been developed especially for efficient oxidation over a wide concentration range of hydrogen isotopes at room temperature, even in the presence of saturated water vapor. The overall reaction rate constant for hydrogen oxidation with the H1P catalyst in a flow-through system using a tritium tracer was determined as a function of space velocity, hydrogen concentration in carriers, temperature of the catalyst, and water vapor concentration in carriers. The overall reaction rate constant for the H1P catalyst in the range near room temperature was considerably larger than that for the traditionally applied Pt/Al₂O₃ catalyst. Moreover, the decrease in reaction rate for H1P in the presence of saturated water vapor was slight compared with the reaction rate in the absence of water vapor due to the excellent hydrophobic performance of H1P. Oxidation reaction on the catalyst surface is the rate-controlling step in the range near room temperature and the rate-controlling step is shifted to diffusion in a catalyst substratum above 313K due to its fine porosity. The overall reaction rate constant in the range near room temperature was dependent on the space velocity and hydrogen concentration in carriers. The overall reaction rate constants in the range of 1;000=T greater than 3.2 correlated to k _overall[s^?1] = 5.59 × 10⁷ × SV[h^?1] × exp (?67.7 [kJ/mol]/R_gT), where the space velocity range was from 600 to 7,200 h^?1.     

The real-time functional test facility for advanced instrumentationand control in nuclear power plants

Testing and validation of the functions and performance of the digital instrumentation and control (I&C) system should be done prior to installation in nuclear power plants. The objective of the I&C Functional Test Facility (FTF) is to test and validate the functions of developed digital control and various monitoring systems. The FTF provides the simulated testing environment as an experimental test bed. The FTF software consists of a mathematical modeling program which simulates a three-loop 993 MWe pressurized water reactor and a supervisory program that comprises all the instructions necessary to run the FTF. The hardware equipment provides an interface between a host computer and a simple test panel or the developed target systems to be tested. The graphical user interface supports an easy and friendly interface between the FTF and users. It is implemented through a Picasso-3 graphic tool developed by the Halden Reactor Project. The FTF is applied to an advanced I&C system prototype, such as an automatic start-up intelligent control system, dynamic alarm system, accident identification system, and intelligent logic tracking system, to test its algorithm or performance. The results of the test show good operational performance of the FTF in normal and transient conditions     

自然循环静态特性实验研究

介绍了在自然循环整体实验装置(NCIF)上完成的自然循环静态特性实验结果。通过对3种泵路(泵及其出口节流件)阻力系数条件下的自然循环静态特性实验,获得了一回路平均温度维持不变的自然循环工况下,系统各热工水力参数随加热功率的变化规律;并对影响稳态自然循环流动的主要因素进行了分析。     

压水堆冷管段 2% 小破口失水事故实验研究

在高压综合实验装置（ＨＰＩＴＦ）上进行了压水堆冷管段２％小破口失水事故实验（ＮＳＢ－６），破口方向为冷管段底部，破口面积为２％。实验再现了核电厂发生小破口失水事故时的热工水力学现象，实验结果与ＲＥＬＡＰ５／ＭＯＤ２系统分析程序的计算结果作了比较，验证了该程序对小破口失水事故的分析能力。     

大型干式安全壳消氢系统的初步设计

以岭澳核电站为分析对象,利用MELCOR和TONUS(CEA)程序进行分析计算,给出了初步的消氢系统设计方案,对不同核电站的消氢系统设计方案进行了对比和讨论.结果表明:安全壳内安装33个FR750型或者17个左右的FR1500型氢气复合器可以满足氢气控制要求.