A program for a high-temperature design analysis and defect assessment has been developed for an elevated temperature evaluation according to the RCC-MRx for Generation IV and fusion reactor systems. The program, called ‘HITEP_RCC-MRx,’ consists of three modules: ‘HITEP_RCC-DBA,’ which computerizes the design-by-analysis (DBA) for class 1 components such as the pressure vessel and heat exchangers according to RB-3200 procedures, ‘HITEP_RCC-PIPE,’ which computerizes the design-by-rule (DBR) analysis for class 1 piping according to RB-3600 procedures and ‘HITEP_RCC-A16,’ which computerizes high-temperature defect assessment according to the A16 procedures. It is a web-based program, and thus can operate on a smartphone as well as on a personal computer once it is connected to the URL. The program has been verified with a number of relevant example problems on DBA, Pipe, and A16. It was shown from the verification works that HITEP_RCC-MRx with the three modules conducts a design evaluation and a defect assessment in an efficient and reliable way.

Current and future nuclear technologies such as fission and fusion reactor systems depend on well characterized structural materials, underpinned by reliable material models. ANSTO is pursuing research into the nuclear power cycle on many fronts, including: modelling and measurement of weld residual stresses; simulation of radiation damage by molecular dynamics modelling; assessment of radiation, cycling and aging effects in power plant structural materials; and characterization of materials for the next generation of nuclear power plants. Several examples of past and current research activities are used to highlight the potential of ANSTO facilities, techniques and capabilities available for collaborative research in the nuclear space.

During the last years, a particular attention has been paid to the consideration and evaluation of environmental effects in the design of mechanical components, particularly for the consequences of Pressure Water Reactors environment on fatigue damages. In the frame of RCC-MRx design code, developed for Sodium Fast Reactors (SFR), Research Reactors (RR) and Fusion Reactors (FR-ITER), the specific features are creep damage (covered by the design rules responding to the needs of fast breeder reactors and high temperatures) and irradiated material (covered by the design rules responding to the needs of research reactors developed for the JHR project). The purpose of this document is to detail the way to consider an environmental effect such irradiation in the material design rules. It covers: - the background for these rules, linked with the changes in materials behavior, - their content, in term of damage impact, - and the challenges to be met for their development to other applications.

This paper describes and explains the contents of bolt and bolted connection design rules included in the present 2012 edition of the French RCC-MRx code ([1], applicable to nuclear installation components). The aim of this work is to describe the rules and their technical and historical background, owing to the widespread fields of application in a NPP and to the relative complexity of methods; bolted connections have often a major safety-related role. The domain of the RCC-MRx covers high temperature sodium reactors, experimental nuclear facilities (Jules Horowitz Reactor, research reactor under construction in France), and fusion reactor components (ITER Vacuum Vessel). The first major application is at present for ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration). The bolt connection usual domain varies in a wide range, from very complex configurations, like high temperature, pressure retaining under cyclic load, to simpler flanges for low temperature piping. In order to fulfill such different needs in design, several sets of rules are included at present in RCC-MRx code, issued from the historical development of previous RCC-MR and RCC-MX codes, from which the “RCC-MRx” version is derived. In the following, the existing four configurations of bolted connections are identified, the respective rules are summarized, and, finally, a short comparison with other nuclear codes and industrial standards is provided. Available rules to-date concern respectively: 1) Preloaded bolts assuring leaktightness (type “B1”) 2) Preloaded bolts not assuring leaktightness (“B2”) 3) Non-preloaded bolts (“B3”) 4) Flange bolts

In 2012, AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires) will publish the fifth edition of the RCC-MR code, named RCC-MRx 2012. This RCC-MRx Code is the result of the merger of the RCC-MX 2008 developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007 which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER. RCC-MRx, developed especially for Sodium Fast Reactors (SFR), Research Reactors (RR) and Fusion Reactors (FR-ITER) can also be used for components of other types of nuclear facilities (except PWR). It has been consider for instance in the frame of the CEN-Workshop (CEN-WS-MRx) in order to develop, on its basis, the European code for the design and fabrication of mechanical equipments for ESNII innovative nuclear installations. The main objective of the RCC-MRx is to capitalize the technical feedback of projects such as SUPERPHENIX, JHR, but also to meet the needs of MYRRHA, PFBR and ASTRID projects and to prepare the design and construction of ALFRED and ALLEGRO. This paper presents the technical evolutions in the 2012 edition and the AFCEN organization dedicated to work in an international frame.

The ITER vacuum vessel (VV) is a double walled torus structure and one of the most critical components in the fusion reactor. The design and fabrication of the VV as nuclear equipment shall be consisted with the RCC-MR code based on French fast breeder reactor. The VV is a heavy welded structure with 60 mm thick shells, 40 mm ribs and flexible housing of 275 mm diameter. The welding distortion should be controlled since the total welding length is over 1500 m. To satisfy the design requirement, the electron beam welding (EBW) and narrow gap gas tungsten arc welding (GTAW) techniques are to be applied and developed through the fabrication of mock-ups. The fabrication design has been developed to manufacture the main vessel and port structures in accordance with the RCC-MR code. All fabrication sequences including welding methods are also established to meet the demanding tolerance and inspection requirement by HHI as a supplier.

For tritium supply to the fusion reactor of ITER (The Way to New Energy) [1], tritium need to be transported from tritium production sites, mainly the CANDU type reactor sites to the tritium plant building of ITER. Korea Atomic Energy Research Institute (KAERI) was commissioned the work of developing the transport package for tritium by ITER Organization and the first stage of the development has been just finished. The developed package was designed to carry 70 g of tritium and classified as a type B(U) package, which should comply with the requirements stipulated in IAEA Safety Standard Series [2]. The package is composed of a storage vessel, a containment vessel, an overpack and an aluminum liner which is a unique feature of the package. The aluminum liner between the storage vessel and the containment vessel is for containment control under the repetitive use of the package. The package has enough pressure resistance for 5 year in-site storage and the structural and thermal integrity under the hypothetical accident conditions has been demonstrated through a series of analyses.

The first wall (FW) of ITER blanket includes beryllium (Be) armor tiles joined to CuCrZr heat sink with stainless steel cooling tube and backing plate in order to improve plasma performance and reduce thermal stress. Therefore dissimilar materials joints are indispensable for fabricating the high heat flux components. Since these joints must withstand thermal and mechanical loads caused by the plasma and electromagnetic force, it is important to evaluate the strength and thermal fatigue life of dissimilar materials joints. When the dissimilar materials joints are subjected by external force and thermal loading, the stress of the joint may indicate singularity at the interface edge. Since the stress singularity may lower the strength of joints, the singularity is evaluated numerically for the various materials combinations and joint configuration to be used in high heat flux components of fusion reactors in this investigation. Moreover, tensile test and elasto-plastic FEM analysis are performed to investigate the fracture behavior of Be/Cu alloy and stainless steel/ Cu alloy obtained the FW mock-up. The results reveal two singular solutions of type r p j −1 for a half-plane bonded to a quarter-plane joint and the singularity is larger than that of a bonded quarter-planes joint. From the viewpoint of stress singularity, the configuration of bonded quarter-planes joint is better than the half-plane bonded to a quarter-plane joint. The singularity for W/Cu alloy combination is large compared to other combination of materials. Especially the singularity of stainless steel/ Cu alloy is very small. Tensile specimen of Be/CuCrZr joint fractured at the bonding interface due to the stress singularity. For the stainless steel/ Cu alloy, however, the specimens fractured at the Cu alloy region apart from the interface.

Put abstract text here. This paper summarizes manufacturing technologies of the water-cooled-solid-breeder (WCSB) blanket module for a fusion reactor using reduced activation ferritic/martensitic steel (RAFM). Although RAFM is very similar to commercial 9 Cr heat resistant steel, RAFM in the blanket is to be used as thin wall structure. Moreover, it is necessary to employ some new manufacturing technologies for the components such as hot-isostatic-pressing (HIP) and fiber-laser-welding (FLW). Some full-scale mock-ups of the blanket have been developed using conventional and newly developed method. The mock-ups have been developed in industrial scale, and the mock-ups demonstrated integrity in the service condition of the blanket without non-nuclear environment. The mock-ups demonstrated their soundness under the service condition of the blanket.

Superconducting magnets are structures which have an important role in Tokamak-type fusion reactor plants. They are huge and complicated structures exposed to very low temperature, 4K and the methods for keeping their integrity need to be newly developed. To maintain their structural integrity during the plant operation, a procedure for structural design was developed as a part of JSME Construction Standard for Superconducting Magnet. General structures and requirements of this procedure basically follow those of class 1 and class 2 components in light water reactor plants as specified in Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, and include the evaluation of primary stress, secondary stress and fatigue damage. However, various new aspects have been incorporated considering the features of superconducting magnet structures. They can be summarized as follows: (i) A new procedure to determine allowable stress intensity value was employed to take advantage of the excellent property of newly developed austenitic stainless steels. (ii) Allowable stress system was simplified considering that only austenitic stainless steels and a nickel-based alloy are planned to be used. (iii) A design fatigue curve at 4K was developed for austenitic stainless steels. (iv) In addition to the conventional fatigue assessment based on design fatigue curves, guidelines for fatigue assessment based on crack growth prediction were added as a non-mandatory appendix to provide a tool of assurance for welded joints which are difficult to evaluate nondestructively during the service.

The objective of this paper is to describe and evaluate mechanical tests of the First Wall (FW) panel Attachment System (AS) for the blanket system of the fusion reactor ITER according to the 2001 design. The tests were performed in SˇKODA VY´ZKUM, Ltd. with FE simulation support from Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering. The goal of the tests was to check the stiffness, strength limit, and fatigue behavior of the bolted joint under loads simulating conditions during off-normal plasma operations in the reactor. The FW panels are attached to a bottom thick shield block by means of ten special studs located on a shaped key-way on a shield block surface. A special device for a long-lasting test of stud tensile pre-load relaxation over of 30 000 temperature cycles between 100° and 200°C was developed. Two methods were used to determine the real stud pre-load force drop during such temperature cycling. An optimum procedure for pre-loading of the AS studs has been developed. Four panel mock-ups (1080×250×50 mm) and one massive shield block having all the features of the real AS were fabricated from 316L stainless steel at VI´TKOVICE Research and Development, Ltd. The panel screwed to the shield with stud preload from 45 to 100 kN and was then loaded alternatively by 2500 cycles of radial moment (±24.5 kN·m), poloidal (longitudinal to panel axis) force (±108 kN), or poloidal moment (±53 kN·m) at room temperature. The stud bending stresses, stud pre-load relaxation, cyclic deformation leading to undesirable radial gap opening at the key-way and a possible plastic deformations of AS were studied. Additional FE simulations were used for better interpretation of measurements. The experimental results have shown that thermal cycling leads to a stud pre-load drop from 100 to 60 kN, whereas the dynamic cycling itself does not cause an additional loss of the pre-load. The individual loads applied do not cause a loss of radial contact in the keyway or a damage of AS even under a low stud’s pre-load of only 54 kN. A small radial gap in the key way was observed only under maximum poloidal moment with an extremely low stud pre-load of 45 kN.