Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme

This paper presents the results and the main lessons learnt from the phase 3 of BEMUSE, an international benchmark activity sponsored by the Committee on the Safety of Nuclear Installations [CSNI: Committee on the Safety of Nuclear Installations (NEA, OECD), 2007. BEMUSE Phase III Report. NEA/CSNI R(2007) 4, October 2007] of the OECD/NEA. The phase 3 of BEMUSE aimed at performing Uncertainty and Sensitivity Analyses of thermal-hydraulic codes used for the calculation of LOFT L2-5 experiment, which simulated a Large-Break Loss-of-Coolant-Accident (LB-LOCA). Eleven participants coming from ten organisations and eight countries took part in this benchmark.In the first section of this paper, the context of BEMUSE is described as well as the methods used by the participants. In the second section, the results of the benchmark are presented. The majority of the participants find uncertainty bands which envelop the experimental data fairly well, however the width of these bands is much diverged. A synthesis of the sensitivity analysis results has been made and is expected to provide a useful basis for further uncertainty analysis dealing with LB-LOCA. Finally, recommendations are given both for uncertainty and sensitivity analysis.     

Methods for the evaluation and synthesis of multiple sources of information applied to nuclear computer codes

This work is devoted to methods used to evaluate and synthesize information given by multiple sources about a variable which true value is not precisely known. We first recall probabilistic and possibilistic approaches to solve the problem. Each approach offers a formal setting to evaluate, synthesize and analyze information coming from multiple sources. They are then applied to the results of uncertainty studies performed in the framework of BEMUSE project.     

钴调节棒更换后RFSP-IST程序通量计算不确定性分析

在中核核电运行管理有限公司秦山第三核电厂（简称秦山三核）调节棒组件变更设计的物理分析中,用于堆芯计算程序RFSP-IST的钴调节棒增量截面由DRAGON产生,它的方法模型与秦山三核安全分析报告RFSAR（2007版）所采用的超栅元计算程序MULTICELL不完全相同,因此有必要对调节棒组件变更前后RFSP-IST程序通量计算不确定性进行分析。基于秦山三核1、2号机组的相关历史运行数据,采用95/95单边上限不确定性分析方法,对调节棒组件变更前后RFSP-IST程序通量计算不确定性进行分析。数值计算结果表明,调节棒组件变更设计及超栅元增量截面计算程序变更未对RFSP IST程序通量计算不确定性产生影响。     

Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.     

Verification of the TECh'-M-97 program on the basis of experimental results obtained for a large leak on a large-scale stand with nuclear heating

Post-test calculations of the LOFT L2-5 experiment were performed as part of the international program BEMUSE using the TECh'-M-97 thermohydraulic computer code. The computational results were used to determine the parameters characterizing the quality of the the nodalization scheme developed and the stationary state obtained. The sensitivity of the computational results to a change in the basic initial data was estimated. Good agreement was obtained between the computational and experimental results on the temperature regime of the fuel elements during an accident. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 444–453, December, 2005.     

Simulation-based design process for the verification of ITER remote handling systems

The work behind this paper takes place in the EFDA's European Goal Oriented Training programme on Remote Handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. One of the projects of this programme focuses on the verification and validation (V&V) of ITER RH system requirements using digital mock-ups (DMU). The purpose of this project is to study and develop efficient approach of using DMUs in the V&V process of ITER RH system design utilizing a System Engineering (SE) framework. Complex engineering systems such as ITER facilities lead to substantial rise of cost while manufacturing the full-scale prototype. In the V&V process for ITER RH equipment, physical tests are a requirement to ensure the compliance of the system according to the required operation. Therefore it is essential to virtually verify the developed system before starting the prototype manufacturing phase. This paper gives an overview of the current trends in using digital mock-up within product design processes. It suggests a simulation-based process design centralized around a simulation lifecycle management system. The purpose of this paper is to describe possible improvements in the formalization of the ITER RH design process and V&V processes, in order to increase their cost efficiency and reliability.     

Simulation of the VVER-1000 pump start-up experiment in the OECD/DOE/CEA V1000CT benchmark by the FLICA4/CRONOS2 coupled code system

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat à l'Energie Atomique (CEA), France, a coupled 3-D thermal–hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark [Ivanov, B., Ivanov, K., Groudev, P., Pavlova, M., Hadjiev, V., 2002. VVER-1000 Coolant Transient Benchmark (V1000-CT). Phase 1 – Final Specifications, NEA/NSC/DOC] is to assess computer codes used in the analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The CEA presented results for the V1000CT-1 Exercise 2 using a coupling of FLICA4 [Toumi, I., Gallo, D., Bergeron, A., Royer, E., Caruge, D., 2000. FLICA4: a three dimensional two-phase flow computer code with advanced numerical methods for nuclear applications. Nuclear Engineering and Design 200, 139–155] and CRONOS2 [Akherraz, B., Baudron, A.M., Lautard, J.J., Magnaud, C., Moreau, F., Schneider, D., Gonzales, M., 2004. Manuel de Référence CRONOS 2.6. Technical Report SERMA/LENR/RT/04-3433/A, CEA] via the coupling tool ISAS [Toumi, I., et al., 1995. Specifications of the general software architecture for code integration in ISAS. Euratom Fusion Technology, ITER task S81TT-01/1]. The FLICA4/CRONOS2 VVER-1000 model is based on the data available in the benchmark specifications. This paper summarizes the FLICA4/CRONOS2 model build-up with the associated sensitivity studies and presents the CEA results for V1000CT-1 Exercise 2 as well as a comparison with experimental results at hot power steady state (HP SS).     

CFD applications in the Pebble Bed Modular Reactor Project: A decade of progress

Of all the systems and components that have to be designed for a nuclear plant, the Reactor Unit is the most significant since it is at the very heart of the plant. At Pebble Bed Modular Reactor (Pty) Ltd. (PBMR), the design of the Reactor Unit is conducted with the aid of extensive analysis work. Due to the rapid computational improvements, the analysis capabilities have had to evolve rather significantly over the last decade. This paper evaluates the evolution of RU Computational Fluid Dynamics (CFD) analysis in particular and presents a historical timeline of the analyses conducted at PBMR. The influence of advances in the hardware and software applications on the evolution of the analysis capabilities is also discussed. When evaluating the evolution of analysis, it is important to look not only at the advances in mesh generation and the representation of the geometry, but also at the improvements regarding the physics that were included in the models. The discussion evaluates the improvements from the pre-conceptual analyses, the concept design, the basic design and finally, the detail design. It is however important to note that the focus of this research was on establishing a methodology for the integrated CFD analysis of High Temperature Reactors. It is recognized however that results from this research can currently only be used to investigate and understand trends and behaviors rather than absolute values. It was therefore required to also launch an extensive V&V program of which the focus was to verify the approach and validate the methodology that was established. The final aim was therefore to combine the research into the methodology with that of V&V in an effort to determine uncertainty bands which would enable the researcher to supply absolute results with an uncertainty value attached.     

18F-FDG放射化学纯度分析方法的建立及方法不确定度的评定

建立了¹⁸F-FDG放射化学纯度的分析方法。以丙酮-水（体积比为50∶20）为展开剂,用2%KAlSO₄溶液处理的Whatman No.1纸为固定相, ¹⁸F^-的R_f＝0,¹⁸F-FDG的R_f＝0.78,方法的相对标准偏差小于1%。本方法简便、快速,适用于¹⁸F-FDG注射液的放射化学纯度分析。对放射化学纯度分析方法的不确定度进行了初步评定,本方法的扩展不确定度u＝0.199 1,u由合成不确定度u_c＝0.095 00及包含因子k＝2.096而得。     

Development of design guidance for low velocity impacts on concrete floors

It has been recognised in the UK, that where designers or safety assessors have to consider the effects of dropped loads on various targets, there is uncertainty associated with the fact that these are low velocity impacts. This uncertainty arises for a variety of load/target combinations, but in this paper we concern ourselves with loads which are massive and with reinforced concrete floor targets.The major uncertainty for dropped load assessments, lies in the fact that the designer does not have access, in general, to relevant data. Instead, it is usual, for design purposes, to make use of empirical formulae which are intended for high velocity missiles. The design process involves using the empirical formulae not only for the local damage which they are intended to address, but also to define a load-time history or impulse for the purposes of determining slab response.This paper indicates the range of events that need to be considered, the manner in which the designer approaches the task of making his assessment of the target and the limits of applicability of the available empirical methods. The programme now in progress in the UK is outlined and some of the results emerging are presented.     

Numerical simulation of the ultimate load capacity of a reactor containment building

From the 1980's onwards, specific requirements of standard guidelines related to failure analysis have made it necessary to extensively apply nonlinear numerical models for containment safety assessment. In the same period new structure types have been proposed which reveal high sensitivities to load cases reproducing accident conditions. This paper presents the results of preliminary non linear calculations on simplified models. The results point out how load conditions with different contribution of temperature loads respect to pressure can lead to large variations in failure mechanisms and limit load factors. This uncertainty is related to the poor definition of accident scenarios and stresses the need for seeking a general agreement on simulation criteria to be used in safety evaluations in order to obtain fully understandable and comparable results.     

Considerations of uncertainties in evaluating dynamic reliability by GO-FLOW methodology – example study of reliability monitor for PWR safety system in the risk-monitor system

The uncertainty analyses have been considered as a relevant topic since WASH-1400 and analysis was performed for identifying the risk measure, e.g. plant- and core-damage frequency or the frequency of a large early release of radioactivity in the probabilistic safety assessment (PSA) or probabilistic risk assessment. There are two main sources of uncertainty such as aleatory uncertainty and epistemic uncertainty (parameter uncertainty, model uncertainty and completeness uncertainty) for risk analysis in PSA or risk-monitor system. A sensitivity analysis is related field to uncertainty, which can provide information of the most effective on those inputs of PSA, which are mostly contributed to the uncertainty.

In this paper, uncertainty analysis (epistemic) has been conducted in the evaluation of dynamic reliability of safety-related subsystem for risk analysis. GO-FLOW methodology has been employed for the procedure of uncertainty analysis alternatively to Fault Tree Analysis and Even Tree because it is success-oriented system-analysis technique and comparatively easy to conduct the reliability analysis of the complex system. The method used sample data from Monte Carlo simulation to quantify uncertainty in terms of appropriate estimates for analysis results. Pressurized water reactor containment spray system has been taken as an example of safety-related subsystem. The results of this paper show that the uncertainty analysis is an important part for the practical evaluation of the system dynamic reliability and makes the reliability prediction more accurate compared with the result without the uncertainty analysis. The GO-FLOW methodology can be employed easily for uncertainty analysis with its advance functions.     

A probabilistic study of the effect of retardation factor uncertainty using a compartment model for radionuclide release into the biosphere

The effect of retardation factor uncertainty with reference to the results of a compartment model, representing radionuclide release into the biosphere from a disposal facility, is studied through the application of polynomial chaos theory. We review the derivation of the working equations and apply the polynomial chaos expansion to these equations. The uncertainty in the retardation factor typically covers several orders of magnitude and stems from uncertainties in the distribution coefficient, these large uncertainties make any quantitative analysis of radionuclide release highly problematic. In this paper, we assume that the uncertainty in the retardation factor is fully described by a log-uniform distribution. We compare results using polynomial chaos against a semi-analytical solution for a short decay chain and present numerical results from polynomial chaos applied to a longer generic decay chain.     

Main control system verification and validation of NPP digital I&C system based on engineering simulator

Full-scope digital instrumentation and controls system (I&C) technique is being introduced in Chinese new constructed Nuclear Power Plant (NPP), which mainly includes three parts: control system, reactor protection system and engineered safety feature actuation system. For example, SIEMENS TELEPERM XP and XS distributed control system (DCS) have been used in Ling Ao Phase II NPP, which is located in Guangdong province, China. This is the first NPP project in China that Chinese engineers are fully responsible for all the configuration of actual analog and logic diagram, although experience in NPP full-scope digital I&C is very limited. For the safety, it has to be made sure that configuration is right and control functions can be accomplished before the phase of real plant testing on reactor. Therefore, primary verification and validation (V&V) of I&C needs to be carried out. Except the common and basic way, i.e. checking the diagram configuration one by one according to original design, NPP engineering simulator is applied as another effective approach of V&V. For this purpose, a virtual NPP thermal-hydraulic model is established as a basis according to Ling Ao Phase II NPP design, and the NPP simulation tools can provide plant operation parameters to DCS, accept control signal from I&C and give response. During the test, one set of data acquisition equipments are used to build a connection between the engineering simulator (software) and SIEMENS DCS I/O cabinet (hardware). In this emulation, original diagram configuration in DCS and field hardware structures are kept unchanged. In this way, firstly judging whether there are some problems by observing the input and output of DCS without knowing the internal configuration. Then secondly, problems can be found and corrected by understanding and checking the exact and complex configuration in detail. At last, the correctness and functionality of the control system are verified. This method is also very convenient for expansion to other type digital I&C V&V. This paper is mainly focused on V&V of closed-loop control systems in full-scope DCS and several detailed reactor control (RRC) systems, including pressurizer pressure and water level control, steam generator water level control. The V&V works were carried out by applying engineering simulator. This paper describes the structure and function of the simulator, V&V procedure, results analysis and problems identified. Through the actual on-line virtual closed-loop testing on Ling Ao Phase II NPP project, many problems of DCS configuration were found and solved. And it proved that V&V based on engineering simulator enables significant time saving, improves economics and safety in the phase of engineering debugging.     

Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWRs. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados, as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had previously undergone a 2400-h ageing heat treatment at 400°C. The test preparation and execution, as well as the material characterization programme, were handled by MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200°C. For safety reasons, it took place at an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN m), the notch did not initiate. This paper presents the results of the experiment and the results of the fracture mechanics analysis, based on finite element calculations.     

CEFR首炉堆芯核设计计算不确定度分析

在中国实验快堆（CEFR）设计阶段，堆芯计算不确定度分析主要是基于在俄罗斯开展的零功率模拟实验获得的，相关不确定度的理论分析评价工作存在不足。本文采用统计抽样方法、确定论微扰方法及直接扰动方法，通过对不确定度来源进行计算分析，给出了堆芯核设计计算的主要结果参数，包括k_eff、控制棒价值、钠空泡效应及功率分布的不确定度定量评价。通过CEFR的分析工作，建立了核设计不确定度评价的方法流程，为后续中国示范快堆核设计的不确定度评价分析奠定了基础。     

An approach to numerical simulation and analysis of molten corium coolability in a boiling water reactor lower head

This paper discusses an approach for application of the computational fluid dynamics (CFD) method to support development and validation of computationally effective methods for safety analysis, on the example of molten corium coolability in a BWR lower head. The approach consists of five steps designed to ensure physical soundness of the effective method simulation results: (i) analysis and decomposition of a severe accident problem into a set of separate-effect phenomena, (ii) validation of the CFD models on relevant separate-effect experiments for the reactor prototypical ranges of governing parameters, (iii) development of effective models and closures on the base of physical insights gained from relevant experiments and CFD simulations, (iv) using data from the integral experiments and CFD simulations performed under reactor prototypic conditions for validation of the effective model with quantification of uncertainty in the prediction results and (v) application of the computationally effective model to simulate and analyze the severe accident transient under consideration, including sensitivity and uncertainty analysis. Implementation of the approach is illustrated on a so-called effective convectivity model for simulation of turbulent natural convection heat transfer and phase changes in a decay-heated corium pool. It is shown that detailed information obtained from the CFD simulations are instrumental to ensure the effective models capture safety-significant local phenomena, e.g. the enhanced downward heat flux in the vicinity of a cooled control rod guide tube.     

Research and development studies on plant and core seismic behaviour for a fast reactor

This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA in co-operation with ANSALDO and ISMES for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary to satisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactors is also pointed out.     

Surface Charging Controlling of the Chinese Space Station with Hollow Cathode Plasma Contactor

A highly charged manned spacecraft threatens the life of an astronaut and extravehicular activity, which can be effectively reduced by controlling the spacecraft surface charging.In this article, the controlling of surface charging on Chinese Space Station(CSS) is investigated,and a method to reduce the negative potential to the CSS is the emission electron with a hollow cathode plasma contactor. The analysis is obtained that the high voltage(HV) solar array of the CSS collecting electron current can reach 4.5 A, which can be eliminated by emitting an adequate electron current on the CSS. The theoretical analysis and experimental results are addressed,when the minimum xenon flow rate of the hollow cathode is 4.0 sccm, the emission electron current can neutralize the collected electron current, which ensures that the potential of the CSS can be controlled in a range of less than 21 V, satisfied with safety voltage. The results can provide a significant reference value to define a flow rate to the potential controlling programme for CSS.     

Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.