Discussion and validation of a simplified analysis against buckling

The Liquid Metal Fast Breeder Reactor poses special problems in the design and construction of its important components. Its low pressures permit utilization of less expensive, thin cross-sections. But the high temperatures result in serious thermal stress and buckling problems. This paper describes the buckling design rules for the French Fast Reactor design for Class I and II components.The paper contains a simplified analysis method, offers experimental validation, and a comparison with the ASME Section III code.     

A survey of the french creep-fatigue design rules for LMFBR

The paper provides a survey of the creep-fatigue design rules for the LMFBR in France. These rules are the ones currently implemented in French component manufacturing. The background of each item is discussed and the trends for improvements currently investigated are described. The cree-fatigue rules apply to elastic analysis only.     

Structural analysis and design against failure of fusion power plant components with special reference to pulsed operation

This study presents the theory of structural analysis and design code rules and constraints as applied to predict lifetime and prevent failure of components and is tailored to deal with the specific design requirements of fusion components. Example calculations are performed for a proposed ITER-like first wall using finite element analysis codes and an assessment of the structural behaviour is also included. The results of the calculations are used to arrive at a proposed design methodology, which accounts for the structural mechanisms likely to occur when operating a tokamak in pulsed mode, as proposed for ITER.     

Sizing rules for the pressure design of pump casings

The paper presents the theoretical and experimental studies undertaken by AFCEN (French Society for Design and Construction Rules for nuclear island components) and SCP (French Union of Pump Manufacturers) in order to develop sizing rules for the pressure design of radially split pump casings.Shape factors used in simplified design formulas have been developed for different configurations using results of finite element analyses and experimental extensometric analyses.     

Progress of IRSN R&D on ITER Safety Assessment

The French “Institut de Radioprotection et de S?reté Nucléaire” (IRSN), in support to the French “Autorité de S?reté Nucléaire”, is analysing the safety of ITER fusion installation on the basis of the ITER operator’s safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.     

Reactor cooling systems thermal-hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

The French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe PWRs. This PSA-2 study is relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, a wide-ranging series of comparisons with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe-accident scenarios. The present paper details 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.     

Analysis of the Phebus FPTO containment thermal hydraulics with the Jericho and Trio-VF codes

This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPTO test performed on 2 December 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting severe accidents in pressurized-water reactors. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPTO experimental protocol. Then, comparisons of the experiment and Jericho calculations are analyzed. Because the Jericho code assumes a well-mixed atmosphere, some additional three-dimensional calculations have been carried out in order to gain further insight into the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment.     

基于FR-Sdaso的法国凤凰堆寿期末自然循环实验分析

系统分析程序是开展反应堆安全分析的重要工具之一,也可用于开展系统瞬态实验过程的分析。法国凤凰堆（Phenix）在停运之前开展的自然循环实验是钠冷快堆领域非常重要的系统瞬态实验,为研究钠冷快堆的瞬态特点提供了很好的参考。为分析此实验过程,利用自主研发的系统分析程序FR-Sdaso对凤凰堆进行建模,对其自然循环实验开展计算分析,并将主要参数的计算值与实验值进行了对比分析。结果表明,FR-Sdaso可较好地模拟此实验的瞬态过程,可用于开展钠冷快堆此类瞬态的安全分析。     

On behaviour of fuel elements subject to combined cyclic thermomechanical loads

This paper presents detailed finite element formulations on the kinematic hardening rule of plasticity included in an existing thermoelastoplastic stress analysis code primarily designed to predict the thermomechanical behaviour of nuclear reactor fuel elements. The kinematic hardening rule is considered to be important for structures subject to repeated (or cyclic) loads. The start-up/operation/shut-down and various power excursions in a reactor all can be classified as cyclic loadings. In addition to the shifting of material yield surfaces as usually handled by the kinematic hardening rule, the thermal effect and temperature-dependent material properties have also been included in the present work for the first time.A case study related to an in-reactor experiment on a single fuel element indicated that significantly higher cumulative sheath residual strains after two load cycles was obtained by the present scheme than those calculated by the usual isotropic hardening rule. This observation may alert fuel modellers to select proper hardening rules in their analyses.     

A new cross-section generation model in the FAST code system and its application to gas- and sodium-cooled Generation IV fast reactors

The cross-section generation scheme employed in the 3D spatial kinetics PARCS code included in the FAST code system being used and developed at the Paul Scherrer Institute (PSI) is currently based on region-wise macroscopic cross-sections for reference conditions and their first-order derivatives with respect to the state variables. Since for some transients, feedback effects may likely not in this way be precisely approximated in their interrelations, this standard method was recently complemented for (hex, z) geometry by a more rigorous cross-section representation scheme. The main idea behind the new approach within the FAST code system is that of preparing sets of microscopic cross-sections for a studied design. The full library consisting of such isotopic tables is generated based on using the ECCO cell code of the code system ERANOS developed by the French Atomic Energy Commission (CEA) for a suitable range of temperatures and background cross-sections (σ₀) on a common grid. In the paper, this σ₀-model is described. In addition, within a detailed verification study needed due to its complexity, it is extensively compared to the standard method for both steady-state and transient conditions. Thereby use is made of current Generation IV fast-spectrum concepts. Reactivity and power evolution indicate overall good agreement of the two methods. Such a good consistence in various transient situations for systems characterized by different neutron spectra gives a large degree of confidence in the correct implementation and suitability of the microscopic cross-section methodology. Therefore, besides for the safety analysis of advanced fast-spectrum core concepts in general, it is foreseen, within the FAST code system, to use the σ₀-model for the assessment of uncertainty propagations in transient calculations in conjunction with the available ERANOS isotopic covariance matrices. The new development makes it also in principle possible to use the PARCS code as a reactor kinetics solver for severe accident analysis, allowing through the use of microscopic cross-sections to account for relocation of the core material during the accident.     

适用于微机的核蒸汽发生器热工水力分析程序—SGTH—2

本程序用于计算核蒸汽发生器的热工水力分布参数以及一次侧流动压降、二次侧自然循环和稳态特性,将本程序的计算结果与法国对同型号蒸汽发生器的实测数据以及用 ATHOS 程序的相应计算结果进行比较,表明主要热工水力参数能令人满意地吻合。     

The TONUS CFD code for hydrogen risk analysis: Physical models, numerical schemes and validation matrix

The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code, called the TONUS code, which incorporates both lumped-parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present the governing equations, numerical strategy and schemes used for the CFD approach.Several benchmark exercises based on experimental results obtained on large-scale facilities, such as MISTRA, TOSQAN and RUT, are presented. They have been used as verification and assessment procedures for modelling and numerical approaches of the code. Specific emphasis is given to the sensitivity analysis of the computed results with respect to numerical and physical parameters. The powerful Design-Of-Experiments technique for sensitivity analysis is successfully applied to the ISP47 MISTRA test case.The TONUS CFD code is presently used to support the hydrogen risk assessment for the European Pressurized Reactor (EPR) plant and to investigate the impact of the two-room concept on hydrogen distribution in the EPR containment.     

Practical aspects of the development of a code of rules for assuring the serviceability of the elements of nuclear power facilities

The 2009 edition of the Federal Law On Technical Regulation, indicating the particulars of technical regulation, makes it possible to implement more fully the provisions of the Federal Law On the Use of Atomic Energy for assuring the required production quality for objects using atomic energy, confirming that the production and associated manufacturing conform to modern standards, including international technologies and procedures. This possibility is realized for equipment and pipelines by means of a domestic code of rules which assure the integrity of the elements of nuclear facilities. A code of rules must preserve the succession of more than 60 years of domestic experience with nuclear power and industry and become a platform for operational adoption of the results of scientific research, and modern industrial technologies.     

Generation Method of Educational Materials Using Qualitative Reasoning

Central Research Institute of Electric Power Industry has developed a nuclear power plant educational system in which educational materials for several events are included. The system effectively teaches operators by tailoring the event explanations to their knowledge levels of understanding. The preparation of the educational materials, however, is laborious and this becomes one of the problems in the practical use of the system. Discussed in the present paper is a basic explanation generation method using qualitative reasoning. This has been developed to solve the problem.

Qualitative equations describing a recirculation pumps trip were transformed into production rules. These were stored in the knowledge base of an event explanation generation system together with explanation sentences. When an operator selects a certain variable's time-interval in which he wants to know the reasons for a variable change, the inference engine searches for the rule which satisfies both the qualitative value and qualitative differential value concerned with this time-interval. Then the event explanation generation section provides explanations by combining the explanation sentences attached to the rules. This paper demonstrates that it is possible to apply qualitative reasoning to such complex reactor systems, and also that explanations can be generated using the simulation results from a transient analysis code.     

运用MC3D程序计算分析堆坑内堆芯熔融物与冷却剂反应

简要介绍了国际上用于计算分析反应堆堆芯熔融物与冷却剂反应（FCI）的主要程序及其差异。随后，介绍了法国FCI计算分析程序MC3D的特点、验证和使用情况，以及运用MC3D程序计算分析反应堆压力容器外FCI的过程和结果，并总结了在MC3D程序使用过程中遇到的问题和解决办法。     

APOLLO: Analysis of hexagonal critical experiments

Three groups of slightly enriched UO₂ fueled, hexagonal light-water moderated lattice critical experiments were analyzed with the APOLLO code at the Hungarian Academy of Sciences KFKI AEKI Reactor Analysis Laboratory, in collaboration with CEA. The work was a part of the NURESIM project where KFKI undertook to develop and qualify some calculation schemes for hexagonal problems using APOLLO2 version 2.7. In the first step non-perturbed asymptotic lattices characterized by material buckling, regular lattice perturbed by different content of gadolinium and reactivity coefficient measurements were chosen for simulation. The modeling approaches used in this analysis are discussed for selected cases. Generally the results performed by the French code using the JEF 2.2 based CEA93.V7 group library were compared to the measurements but in some cases inter-comparisons were performed to both Monte Carlo solutions as well as against results of the Hungarian lattice transport code MULTICELL based on ENDF-B/VI nuclear data. Even if the agreement between calculations and measurements differs for different codes using different nuclear data and methods the overall agreement among measured and calculated figures are good.     

On the state of LMFBR structural analysis and existing rules with regard to class 1 components of SNR 2

The design and construction of LMFBR components largely depend on thermal loading, normally coupled with the effects of cyclic operation. The structural behaviour must be assessed in relation to high temperature and time-dependent characteristic material data. This also leads to additional demands on analysis and evaluation procedures. Although a considerable effort has been made to cope with these requirements, the important phenomena of ratchetting, elastic follow-up, creep, creep-fatigue, time-dependent buckling and thermal striping continue to need to be studied. This has become evident during the construction of the recent LMFBRs. Some procedures are still conservative simplifications, others such as plastic strain concentration factors, creep-fatigue and the influence of multi-axial states on creep, relaxation and creep-fatigue need deeper and larger investigation.On the other hand non-elastic methods have been developed and used for final analyses and code assessment of important components and highly stressed regions as well as to assess equivalent linear model evaluation schemes. Thus, the rules have been completed to account for these requirements and supplemented by criteria formulated for each kind of non-linear calculations. These new code parts and the articles introduced for the requirements mentioned above cannot easily be grouped into a clear arrangement. The ways in which the diverse codes achieve this matter differ remarkably. Although the primary goal of improved and more sophisticated design rules is a more realistic code assessment, they should also be a tool for optimizations and, in accordance with constructional experience, lead finally to a code for LMFBR constructions similar to LWR codes. This paper presents comparative considerations of the statements formulated in the international codes and in the German licensing procedure with regard to the aforementioned problems.     

Qualification of the CFD code Trio_U for full scale reactor applications

The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).¹ The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.     

大型核电站全厂断电事故的失效模式分析

运用故障树分析方法，对核电站全厂断电事故进行分析。建造了全厂断电事故即厂用电力系统Ａ、Ｂ两列６．６ｋＶ交流应急母线ＬＨＡ和ＬＨＢ同时失电故障树。利用ＳＥＴＳ程序及法国标准９０万千瓦压水堆核电站２００堆年运行经验反馈的可靠性数据，对全厂断电故障树作了定性、定量分析，得到了全厂断电事故的发生概率。给出了全厂断电事故的主要失效模式及发生的概率和事件的重要度。