秦山核电二期工程冷停堆放射性峰值效应对化容系统净化设备的屏蔽设计影响

通过计算,验证在秦山核电二期工程辐射屏蔽设计中,采用0．25％的燃料元件破损率源项以及法国核电厂经验反馈数据来设计化学和容积控制系统(RCV)主要净化设备(过滤器和除盐器)的屏蔽,是否能够满足冷停堆后,由于氧化运行引起的更多放射性物质积累形成的高放射性峰值所造成的外照射辐射影响的要求,为新建同类型核电厂的辐射防护设计积累经验。     

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.     

Comparison of JSFR design with EDF requirements for future SFR

A comparison of the Japan sodium-cooled fast reactor (JSFR) design with the future French sodium-cooled fast reactor (SFR) concept has been done based on the requirements of Electricité de France (EDF), the investor-operator of the future French SFR, and the French safety baseline, under the framework of an EDF and Japan Atomic Energy Agency (JAEA) bilateral agreement of research and development cooperation in future SFRs..     

EPR accident scenarios and provisions

For the EPR an improved defence-in-depth concept is applied. In an evolutionary way, accident control is developed from existing French and German PWR designs, thereby achieving a high safety level quantified by probabilistic safety assessment. Independent of that, severe accidents are considered in the design. By a robust containment and severe accident mitigation measures, the need for offsite emergency response actions (population evacuation or relocation) is restricted to the immediate plant vicinity. This paper gives an overview of the features introduced for, and the analyses correlated to, the dedicated primary depressurization, melt–coolant interaction, melt stabilization, hydrogen control, and containment heat removal.     

Fresh Fuel Shipment: the FCC Concept

Abstract

For more than 25 years, Framatome, the foremost world supplier of PWR fuel, has been delivering fuel elements directly or through Fragema, from the factories of its subsidiary FBFC to the various French and foreign sites. Since 1995, the French Safety Authority has undertaken a process of reviewing the concepts of the radioactive material shipping containers, with the purpose of ensuring that their safety files are consistent with the regulations in force. For now the safety analyses on the former packages, known as the RCC type, have shown the limits of this design. Framatome has committed itself to this process by performing a global review of the RCC containers transporting uranium-bearing fresh fuel assemblies, which has led to upgrading of the design of these former containers to an up-to-date model: known as the FCC type. The design of the new FCC container is presented from the mechanical strength and criticality standpoints, together with the regulatory tests which validated the basic hypotheses of the FCC design.     

Descriptive characteristics of different types of test for irradiation embrittlement

Within the IAEA coordinated programme on optimizing of reactor pressure vessel surveillance programmes and their analysis, phase 3, a specially tailored radiation sensitive correlation monitor material, a Japanese steel plate with code designation JRQ, a French forging material (FFA) and a Japanese forging material (JFL) were selected for the investigations to be carried out in Finland.Based on the evaluation of the experimental results it was demonstrated that dynamic fracture toughness transition shift is equivalent to the Charpy-V shift, but the static fracture toughness transition shift may be considerably larger than the dyanamic shift. Thus, Charpy-V is not suitable for estimating the static fracture toughness transition shift.These findings have a strong impact upon the design of future surveillance programmes.     

The fundamental role of fluid dynamic analyses in the design of the solid EU Test Blanket Module

In the frame of the activities of the European Test Blanket Module Consortium of Associates, the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), the so-called solid TBM, is developed in Karlsruhe Institute of Technology (KIT). In the EU experimental strategy, a series of 4 different HCPB TBMs will be connected to the dedicated equatorial port n.16 during the ITER lifetime. The ITER TBM program has to provide DEMO relevant experimental data for the main functions of the blanket modules of a future fusion reactor.The preliminary thermo-mechanical design assessment of the TBM box (based on transient, steady state and accidental analyses) has been presented. All along the design assessment phase the fluid dynamic analyses play a fundamental role for the TBM sub-components, the Breeder Units (BUs) and the manifolds (MF) stages. This paper highlights the methodology implemented for the Computational Fluid Dynamic (CFD) analyses in the TBM design life cycle, and presents the results and the impact on the overall performance evaluation of the HCPB TBM. The following models are presented in detail: the CFD model of the TBM First Wall and its application to a reduced scale First Wall, the 3D CFD model of the BUs, and the thermo fluid dynamic modelling of the manifold systems.     

BWR steam dryer for extended power uprate

A new steam dryer for extended power uprated conditions, based on a design with proven performance record, has been designed by Westinghouse with performance characteristics meeting all utility requirements. Extensive use of computational fluid dynamics (CFDs) has been made to design the new dryer intended for plants of the BWR 3000-type. In addition to numerical simulations, the analyses rely on previous knowledge, acquired from experimental work on steam flow characteristics in the BWR 3000 with its asymmetrically placed steam line nozzles. The new design is expected to both decrease vibration levels in the steam lines and to solve a water-level measurement problem. An extensive experimental verification of the new design is currently in progress.     

Modified Quasi-Steady-State Method to Evaluate the Mean Power Profiles of Nuclear Excursions in Fissile Solution

A modified quasi-steady-state method has been developed in order to evaluate the mean power during a nuclear excursion in fissile solution. The conventional method used the critical equation based on the one-group theory in order to calculate the reactivity. However, the one-group approximation reduces the calculation accuracy, and the geometrical buckling used in the critical equation is not applicable to complex geometries. Thus, we have modified the method to use the reactivity feedback coefficients, which are widely used in the calculation of one-point reactor kinetics. Although the modified method requires an external calculation to obtain the feedback coefficients, it is applicable to complex geometries and provides more accurate results than does the one-group approximation when the proper coefficients are given.

Moreover, a new method to calculate the boiling power has been developed. In this method, the power corresponding to the void fraction that compensated for the inserted reactivity along with the temperature feedback was calculated using the relationship, which was derived using the French SILENE experimental data.

Experimental analyses have been conducted to validate the new method for the French CRAC and Japanese TRACY experiments. The analytical results showed close agreement with the experimental results.     

RCC-M (rules for design and construction of nuclear components)

The paper deals with a presentation of the design rules included in the French RCC-M code applicable to mechanical components of PWR nuclear islands and published by the French Society for Design and Construction rules for Nuclear Island Components (AFCEN). Particular attention is paid to the major principles which constitute the background of the rules of the code and to recent developments included in the code.     

Validation of the CATHARE2 code against experimental data from Brayton-cycle plants

In recent years the Commissariat à l’Energie Atomique (CEA) has commissioned a wide range of feasibility studies of future-advanced nuclear reactors, in particular gas-cooled reactors (GCR). The thermohydraulic behaviour of these systems is a key issue for, among other things, the design of the core, the assessment of thermal stresses, and the design of decay heat removal systems. These studies therefore require efficient and reliable simulation tools capable of modelling the whole reactor, including the core, the core vessel, piping, heat exchangers and turbo-machinery. CATHARE2 is a thermal-hydraulic 1D reference safety code developed and extensively validated for the French pressurized water reactors. It has been recently adapted to deal also with gas-cooled reactor applications. In order to validate CATHARE2 for these new applications, CEA has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops.In the short-term perspective, CATHARE2 is being validated against existing experimental data. And in particular from the German power plants Oberhausen I and II. These facilities have both been operated by the German utility Energie Versorgung Oberhausen (E.V.O.) and their power conversion systems resemble to the high-temperature reactor concepts: Oberhausen I is a 13.75-MWe Brayton-cycle air turbine plant, and Oberhausen II is a 50-MWe Brayton-cycle helium turbine plant. The paper presents these two plants, the adopted CATHARE2 modelling and a comparison between experimental data and code results for both steady state and transient cases.     

Design Assessment for Transport and Storage Casks

Abstract

The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress–strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been investigated experimentally. By consideration of these dynamic values in strength analyses of casks at impact a good agreement between experimental and calculational results has been achieved.     

An experimental validation of the guideline for inelastic design analysis through structural model tests

In this paper, the inelastic analysis procedures recommended to use in the advanced elevated temperature structural design guide under development in Japan for the improved design of future fast breeder reactors were validated through the structural model tests and the evaluation of the experimental results by the inelastic analyses. First, a thermal fatigue test of a 316FR hollow cylinder with two longitudinal weldments was conducted under the condition of combined constant axial load and cyclic movement of axial temperature distribution, which simulated the loading condition near the free surface of coolant sodium in the main vessel of fast breeder reactors (FBRs). In the experiments, longitudinal and radial ratcheting deformation were measured and crack initiation life was also examined. Second, the inelastic analyses were carried out in accordance with the recommended procedure by using the measured results of oscillating temperature distribution. Finally, the results of inelastic analyses were compared with the experimental results and it was validated that the recommended practice gave a conservative result for the deformation and a good estimation of strain range for the fatigue life evaluation.     

控制棒水力驱动系统的设计和研究

分析了２００ＭＷ核供热堆控制棒水力驱动系统的设计特点;系统中主要设备的设计特点及特性;旁路自调节结构的设计及其高温下的补偿作用以及系统温度特性的实验结果。经对实验结果的分析表明：ＨＤＳＣＲ和各设备的设计合理,运行可靠;各设备的设计不仅降低了设备的加工难度及安装难度,而且改善了系统的温度特性;系统安全能满足２００ＭＷ核供热堆对控制棒驱动机构的要求。     

ITER safety and licensing update

Safety files were submitted by the ITER Organization to the French nuclear safety authorities in March 2010 as a part of the licensing process. These included the preliminary safety report (RPrS) which presents the extensive safety analyses performed for ITER. The report has been the subject of examination by the authorities and their advisors, and discussions with them have been held on many topics. In the light of this process, this paper discusses some of the topics that remain prominent in the safety analysis of ITER. In particular, the provision of the two safety functions, confinement of radioactive material and limitation of exposure to radiation, is explained and some of the potential challenges to them are identified. Amongst these are the risks of fire and explosion, and external events such as earthquake and loss of all electric power. Provisions in the ITER design, together with the characteristics of fusion, ensure that a very good safety performance will be achieved.     

Application of fractional scaling analysis (FSA) to loss of coolant accidents (LOCA): Methodology development

A quantitative methodology is developed to

(a) scale time-dependent evolution processes involving an aggregate of interacting modules and processes (such as a NPP) and

(b) integrate and organize information and data of interest to NPP design and safety analyses.

The methodology is based on two concepts: fractional scaling and hierarchy. Fractional scaling is used to provide a synthesis of experimental data to generate quantitative criteria for assessing the effects of various design and operating parameters on thermo-hydraulic processes in a NPP. The synthesis via fractional scaling is carried out at three hierarchical levels: process, component and system. The methodology is demonstrated by applying it to a LOCA.The fractional scaling analysis (FSA) identifies dominant processes, ranks them quantitatively according to their importance and provides thereby an objective basis for establishing phenomena identification and ranking tables (PIRT) as well as a basis for conducting uncertainty analyses.The paper also discusses the benefits to be realized by applying the methodology to presently operating NPP as well as to future design of NPP.     

Engineering the development of optical fiber sensors for adverse environments

Optical fiber sensors and measurement systems must be engineered to meet tough environmental requirements necessary for applications outside the laboratory. No generalized computer-aided tools exist to help advance the development, design and use of these systems in the field. Computer-aided design tools currently being developed are described. Structural finite element analyses are coupled with opto-elastic analyses of all-fiber interferometers and serial microbend sensors for the distributed measurement of various physical quantities. The combined analyses are parameterized and implemented on personal computers or workstations for use as design and development tools to determine the performance of different sensor configurations in various environments. Potentially, these computer-aided tools could be used for failure diagnosis and redesigning of existing optical fiber sensors. Performances predicted by the computer simulations are verified with experimental data and numerical analyses from the literature. The long-term goal is to develop user-friendly software packages for sensor manufacturers and end-users.     

Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future

As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity.     

Progress of ITER vacuum vessel

Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R&D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.