The use of the COMPACT code for calculation of long term environmental effects as a result of high temperature discharges into confined areas

COMPACT is a computer code used to provide long term thermo-hydraulic safety analysis both for buildings associated with light water nuclear reactors and other similar structures. This paper discusses the main features of the code and the validation programme, which has been put in place to establish its ability to predict accurately the environmental conditions within the containment of a reactor. In particular the experience gained in modelling the hydrogen tests performed in the HDR reactor building is described.     

基于多孔介质模型的快堆蒸汽发生器热工水力特性数值研究

蒸汽发生器（SG）作为钠冷快堆一次侧钠与二次侧水的热交换器,其可靠程度直接影响反应堆能否安全运行,因此对SG的一次侧热工水力特性的研究具有重要意义。本研究采用多孔介质模型,对快堆蒸汽发生器一次侧流场进行分析。通过对支撑板模型的计算,获得多孔介质控制方程的阻力源项。一次侧向二次侧的释热量通过系统程序Relap5计算,确定多孔介质控制方程的能量源项。通过用户自定义程序将动量源项与能量源项编译至FLUENT求解器中。通过FLUENT求解器求解控制方程,获得SG一次侧流场、压力场、温度场等信息。并通过对比模拟结果与设计值,验证了计算的准确性。      

LBE-water interaction in LIFUS 5 facility under different operating conditions

In the frame of the IP-EUROTRANS Project, an experimental program, focused on studying the LBE/water interaction has been performed using the LIFUS 5 facility available at ENEA-Brasimone. The physical effects and the possible consequences of this interaction have been evaluated over a wide range of different conditions. Besides the experimental activities, a numerical simulation activity has been performed with SIMMER code in order to better investigate the thermo-hydraulic phenomena involved in the interaction and to confirm the capabilities of the code to simulate this kind of phenomena. The experimental and the calculated results in terms of pressure and temperature evolutions in the system show a good agreement.     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

基于ECC安注热混合试验的LOCUST 1.2分析与验证

在失水事故（LOCA）工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统（ECCS-XJTU）试验台架进行的堆芯应急冷却（ECC）安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明：试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。     

Strength evaluation of a steam distribution device in the Ignalina NPP accident localisation system

The Ignalina NPP has a pressure suppression type of confinement, which is referred to as the accident localization system (ALS). The ALS prevents the release of the radioactive material from the NPP to the environment during a loss-of-coolant accident (LOCA). Ten water pools are located in the two ALS towers (five pools in each tower), which separate the dry well from the wet well. These water pools condense the accident-generated steam and prevent high overpressures in the compartments.The steam distribution device (SDD), with the vertical vent pipes (nozzles) that are inserted under the water of the condensing pools, connects the dry well and the wet well. In case of an accident, these components must be capable of withstanding the dynamic loads generated by a LOCA for successful pressure suppression function.This paper presents the transient analysis of the SDD and their connections to the vertical steam corridors following a LOCA. A thermo-hydraulic analysis of the SDD was performed using the state-of-the-art COCOSYS code to determine pressure and temperature histories resulting from a LOCA. The finite element code NEPTUNE was used to evaluate the structural integrity of the SDD and its supporting reinforced concrete wall. Results show that, although portions of the SDD undergo plastic response and the outside surface of the vertical steam corridor reinforced concrete wall cracks, the structural integrity of the SDD and wall are maintained during a LOCA.     

FARST — A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:

1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.

2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.

3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.

4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.

The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR).The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant.     

静止和摇摆条件下浮动式核电站一二回路PRHRS特性分析

浮动式核电站长期在海洋环境中运行,各系统都会受到海洋运动条件的影响。非能动余热排出系统（PRHRS）可在核电站发生全厂断电事故的情况下带出堆芯衰变余热,防止堆芯熔化,是重要的反应堆辅助系统。本文以一种采用海水作为最终热阱的浮动式核电站作为研究对象,分别设计了一回路和二回路PRHRS,开展了静止和摇摆条件下反应堆系统发生全厂断电事故的计算,对两种PRHRS在静止和摇摆条件下的运行特性进行了分析。研究表明,静止条件二回路PRHRS具有更强的带热能力,摇摆条件下一回路PRHRS的带热能力更加稳定。     

An object-oriented approach to simulation of IRIS dynamic response

In this paper the development of an adequate modelling and simulation tool for Dynamics and Control tasks is presented. The key features of the developed simulator are: “Modularity” - the system model is built by connecting the models of its components, which are written independently of their boundary conditions; “Openness” - the code of each component model is clearly readable and close to the original equations and easily customised by the experienced user; “Efficiency” - the simulation code is fast; “Tool support” - the simulation tool is based on reliable, tested and well-documented software.To achieve these objectives, the Modelica language was used as a basis for the development of the simulator. The Modelica language is the result of recent advances in the field of object-oriented, multi-physics, dynamic system modelling. The language definition is open-source and it has already been successfully adopted in several industrial fields.The test bed for the application of the object-oriented approach has been the new generation, integral type, IRIS nuclear reactor. IRIS (International Reactor Innovative and Secure) is a pressurized light water cooled, small/medium power (335 MWe) reactor reactor, under development by an international consortium of nineteen organizations from ten countries. The preliminary design has been completed and the pre-application licensing process with the US-Nuclear Regulatory Commission (NRC) is underway.To provide the required capabilities for the analysis, specific models for the nuclear reactor components have been developed, to be applied for the dynamic simulation of the IRIS integral reactor, albeit keeping general validity for PWR plants. The following Modelica models have been written to satisfy the IRIS modelling requirements and are presented in this paper: point reactor kinetic, fuel heat transfer, control rods model, and a once-through type steam generator, thus obtaining a specific library of nuclear models and components. As far as other classical power generation plant components are concerned, the Thermo Power open library, developed at Politecnico di Milano as well, has been adopted and is briefly presented in the paper. Originally conceived for conventional, fossil-fired plants, the highly modular approach allowed to effectively reuse the models of the balance of plant systems, which have been connected to the models of the nuclear power generation process, to obtain a system simulator for the IRIS reactor.Finally, preliminary results of the code validation process and the reactor dynamics are presented.     

Current status of accident analysis for Korean HCCR TBS

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.     

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

The HPLWR (high performance light water reactor) is the European concept design for a SCWR (supercritical water reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction – mass flow rate or gravity – mass flow rate feedback of the system), or they can be coupled thermo-hydraulic–neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic–neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 s. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating conditions. A clear instability zone was found within the stability plane, which seems to be similar to that of a BWR. Experimental data on the stability of a supercritical loop is rare in open literature, and these data could serve as an important benchmark tool for existing codes and models.     

Simplified numerical model for defining Ledinegg flow instability margins in MTR research reactor

Establishment of safety margins and the corresponding operating condition limits will ensure achievement of a safe operation of nuclear installations. For this purpose, several critical phenomena have been analyzed theoretically and experimentally and a great number of models and correlations are made available. Among these critical issues the well-known flow instability has been intensively investigated by several authors especially for nuclear power plants' (NPPs) operating conditions. However, limited published work is available for research reactor operation conditions. In general, the Whittle and Forgan correlation is widely used to define the margin to static flow instabilities in narrow parallel heated channels for research reactors.In the framework of verification and assessment of the capabilities of the RELAP5/Mod 3 system code to determine the onset of flow instability in research reactor conditions, a simple model based on steady-state equations adjusted with drift-flux correlations has been developed. The program is used to draw the pressure drop characteristic curves and to establish the conditions of the Ledinegg instability in a uniformly heated channel subject to constant outlet pressure. The model is assessed by using experimental data from a thermal hydraulic test loop by Siman-Tov and numerical results from RELAP5/Mod 3. The model presents acceptable estimation of the target mass flow that would induce flow instability and the latter could be then used to establish a conservative margin to the Ledinegg instability.     

化学种态分析软件CHEMSPEC最新进展

我国高放废物地质处置研发工作已经步入建造地下实验室阶段。地下实验室建造安全评价和未来的处置库性能评价中均需要关键放射性核素在相应深部地质条件下的扩散和迁移参数,而关键核素的扩散和迁移参数与核素在相应水岩体系中的化学种态密切相关。为满足我国核设施退役治理工作的需要,尤其是我国高放废物地质处置相关安全评价的需要,北京大学核环境化学课题组于2008年开始编写具有完全著作权的化学种态分析软件CHEMSPEC。经过多次修改和完善,目前已经具备了较好的计算功能。本文介绍该软件在表面配合模型、数据库补充和程序优化方面的最新进展,以实例形式介绍该软件的新性能,以期为我国相关实验研究者使用该软件提供参考。     

Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions.     

Preliminary evaluation of aircraft impact on a near term nuclear power plant

For the assessment of the safety and durability of a nuclear power plant (NPP), the containment building behaviour shall be evaluated, under various service and extreme conditions, both natural or produced by natural accident or vicious man activities, like September 2001 jet aircraft crashes.The aim of this paper is to preliminary evaluate the effects and consequences of the energy transmitted to the outer containment walls (according to the international safety and design code guidelines, as NRC or IAEA ones) due to a military or civil aircraft impact into a nuclear plant, considered as a ‘beyond design basis’ event.To perform reliable analysis of such a large-scale structure and determine the structural effects of the propagation of this types of impulsive loads (response of containment structure), a realistic but still feasible numerical model with suitable materials characteristics were used by means of which relevant physical phenomena are reflected. Moreover a sensitivity analysis has also been carried out considering the effects of different containment wall thickness and reinforced/prestressed concrete features. The obtained results were analysed to check the NPP containment strength margins.     

Numerical simulations on natural convective heat transfer and active cooling of IFMIF Test Cell

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. The Test Cell (TC) accommodates the Test Modules and the lithium target assembly. Due to the nuclear heat generation, all the Test Modules inside the TC will be actively cooled. Other components like supporting structures, pipelines, cables etc., will be passively cooled by natural convection. The heat will be removed from the steel liners surrounding the TC by active water cooling. This paper concerns the thermo-hydraulic simulations of the Test Cell using Ansys-CFX. The current simulation model includes the natural convection inside the TC, several forced convective water flows in the pipelines attached on the steel liners and the helium-cooled HFTM (High Flux Test Module). The simulations provide the only means for validating the design before the construction and operation.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Experimental and theoretical study of steam condensation induced water hammer phenomena

We investigate the steam condensation induced water hammer (CIWH) phenomena and present experimental and theoretical results. The experiments were performed in the PMK-2 facility, which is a full-pressure thermo-hydraulic model of the primary loop of the VVER-440/312 type nuclear power plant and located in the Atomic Energy Research Institute Budapest, Hungary.The present experimental setup is capable to measure CIWH phenomena in a wide range of steam pressure, cold water temperature and mass flow rate at a high level of accuracy. On the theoretical side CIWH is studied and analyzed with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one-dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. Our study clearly shows that Relap5 and Cathare which are used in the nuclear industry to simulate nuclear power plant accidents cannot resolve the narrow pressure peaks created during a CIWH event. Only WAHA3 can model CIWH properly. Experimentally measured and theoretically calculated pressure peaks are in good agreement, however simulations always show additional pressure peaks. As a new feature in this study we present calculations without additional unphysical reflections caused by boundary conditions.     

Steady state and linear stability analysis of a supercritical water natural circulation loop

Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN using supercritical water properties has been developed to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been qualitatively assessed with published results and has been extensively used for studying the effect of diameter, height, heater inlet temperature, pressure and local loss coefficients on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). The present paper describes the linear stability analysis model and the results obtained in detail.     

An experimental study on evaporative heat transfer coefficient and applications for passive cooling of AP600 steel containment

After TMI and Chernobyl accidents, many efforts have been made to enhance the nuclear safety with passive features. Among such passive features, the passive containment cooling system (PCCS) has been suggested by Westinghouse in the AP600 plant. The containment with PCCS is a dual containment, and consists of a stainless steel vessel and a concrete wall. In the gap between these structures, air and water can counter-currently pass and cool the steel surface. This paper experimentally investigates evaporative heat and mass transfer at the surface of a falling water film with counter-current air flow in a vertical duct with one-side heated plate. Experiments included various conditions of mass flow rate of film and air. Experimental results show the strong effects of water temperature and air mass flow rate, but little effect of the water flow rate. Also, simple analyses based on heat and mass transfer analogy were performed to evaluate the experimental results. With experimental data, a new correlation on evaporative mass transfer coefficient was developed, and with the correlation, the containment pressure and temperature was calculated for the design basis accident of AP600 by the use of CONTEMPT4/MOD5 code implementation.