Low activation materials applicable to the IFMIF accelerator

Sugimoto, Masayoshi; Takeuchi, Hiroshi

Journal of Nuclear Materials, 329-333(Part1), p.198 - 201, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.328

IFMIF is an accelerator-based intense neutron source for developing fusion reactor materials, and a stable and continuous operation is a primary importance. The radioactivity produced by the beam loss prolongs the cooling time before maintenance and results in a potential limit to the availability. The materials selection is important for minimizing the total radioactivity produced by ion beam and neutrons, after about one day cooling time in advance of hands-on maintenance. For the major parts, Cu for cavity and Al for beam duct, the beam loss needs to be reduced below 5nA/m. Such a low loss can be realized by suppressing the beam halo and by using the scrapers made from high-Z material, Ta, to cut the tail of beam. The insertion points of the scrapers are determined from the beam size evolution along the accelerator and the available room for radiation shielding around them. The reduction of the radioactivity after the whole lifetime of facility is another issue and the possibility to use low activation materials coating for cavity and beam duct is presented.

Fusion blanket construction

Konishi, Satoshi*; Kimura, Akihiko*; Akiba, Masato; Nakamura, Hiroo; Nagasaka, Takuya*; Muroga, Takeo*; Hasegawa, Akira*; Matsui, Hideki*

Nihon Genshiryoku Gakkai-Shi, 46(5), p.311 - 322, 2004/05

http://ci.nii.ac.jp/naid/10012348483

no abstracts in English

Swelling behavior of F82H steel irradiated by triple/dual ion beams

Wakai, Eiichi; Kikuchi, Kenji; Yamamoto, Shunya; Aruga, Takeo; Ando, Masami; Tanigawa, Hiroyasu; Taguchi, Tomitsugu; Sawai, Tomotsugu; Oka, Keiichiro*; Onuki, Somei*

Journal of Nuclear Materials, 318, p.267 - 273, 2003/05

https://doi.org/10.1016/S0022-3115(03)00122-3

no abstracts in English

Status of activities on the lithium target in the key element technology phase in IFMIF

Nakamura, Hiroo; Burgazzi, L.*; Cevolani, S.*; Dell'Ocro, G.*; Fazio, C.*; Giusti, D.*; Horiike, Hiroshi*; Ida, Mizuho*; Kakui, Hideo*; Loginov, N.*; et al.

Journal of Nuclear Materials, 307-311(2), p.1675 - 1679, 2002/12

https://doi.org/10.1016/S0022-3115(02)01108-X

no abstracts in English

Annual report on the services for chemical analysis and scientific glassblowing in the fiscal year 1999

Ito, Mitsuo; Obara, Kazuhiro; Toida, Yukio*; Suzuki, Daisuke; Gunji, Katsubumi*; Kato, Kaneharu; Watanabe, Kazuo

JAERI-Review 2000-032, 58 Pages, 2001/01

JAERI-Review-2000-032.pdf:2.68MB

no abstracts in English

Development of a small specimen test machine to evaluate irradiation embrittlement of fusion reactor materials

Ishii, Toshimitsu; Omi, Masao; Saito, Junichi; Hoshiya, Taiji; Ooka, Norikazu; Jitsukawa, Shiro; Eto, Motokuni

Journal of Nuclear Materials, 283-287(Part.2), p.1023 - 1027, 2000/12

https://doi.org/10.1016/S0022-3115(00)00093-3

no abstracts in English

Research on development of high-purity iron-based alloys; Manufacture, analysis of small amount of element and property tests

; *; ; ; Aoto, Kazumi;

JNC TN9400 2000-059, 43 Pages, 2000/05

JNC-TN9400-2000-059.pdf:2.08MB

The purpose of this study is to understand the material properties of manufacturable high-purity iron and high-purity iron-based alloy in present technology and to get an applicable prospect for the structural and functional material of the frontier fast reactor. Then the about 10kg high-purity iron and iron-based alloy were melted using a cold-crucible induction melting furnace under the ultra-high vacuum. Subsequent to that, the compatibility between the melted material and the high-temperature sodium environment which is a special feature of the fast reactor and tensile property at room and elevated temperatures were investigated using the melted materials. Also, the creep test using the high-purity 50%Cr-Fe alloy at 550C in air in order to understand the high temperature creep property. ln addition, the material properties such as thermal expansion coefficient, specific heat and electrical resistance were measured and to evaluate the outlook for the structural material for the fast reactor. The following results were obtained based on the property test and evaluation. (1)lt was possible to melt the about 10kg high-purity ingot and high-purity 50%Cr-Fe alloy ingot using a cold-crucible induction melting furnace under the ultra-high vacuum. (2)The tensile tests of the high-purity 50%Cr-Fe alloy were performed at room and elevated temperatures in order to understand the deformation behavior. From the experimental results, it was clear that the high-purity 50%Cr-Fe alloy possesses high strength and good ductility at elevated temperatures. (3)The physical properties (the thermal expansion coefficient and specific heat etc.) were measured using the high-purity 50%Cr-Fe alloy. lt was clear that the thermal expansion coefficient of high-purity 50%Cr-Fe alloy was smaller than that of SUS304. (4)From the corrosion test in liquid sodium, the ordinary-purity iron showed the weight loss after corrosion test. However the high-purity iron showed ...

Investigation of the properties of high temperature resistance alloys used in the helium gas cooled high temperature reactor

Uwaba, Tomoyuki

JNC TN9420 2000-005, 28 Pages, 2000/03

JNC-TN9420-2000-005.pdf:0.94MB

In the first phase of the feasibility study, their basic objectives are presentating the feasible image and scenario of development of the FBR cycle system, which is composed of the fast reactor, spent fuel reprocessing and fuel manufacturing facility. In the development of the FBR system in this phase, various ideas of plants are to be studied, which include coolant types such as sodium, heavy metals, gases(CO, He), wator, and middle or small size of the reactor, and fuel types (MOX, metal and nitride). In this report, as a part of this study, materials used for the core of the helium gas cooled reactor and their integrity (corrosion, mechanical and irradiation property) under high temperature helium atmosphere were investigated from open literatures.

Design concepts for overpack

*; *; Tanai, Kenji

JNC TN8400 99-047, 54 Pages, 1999/11

JNC-TN8400-99-047.pdf:3.16MB

This paper reports on the design process for a carbon-steel overpack as a key component in the engineered barrier system of a deep geological repository described in the 2nd progress report. The results of the research and development regarding design requirements, configuration, manufacturing and inspection of overpack are also described. The concept of a composite overpack composed of two different materials is also considered. First, the design requirements for an overpack and presume environmental and design conditions for a repository are provided. For a candidate material of carbon steel overpack, forging material is selected considering enough experience of using this material in nuclear power boilers and other components. Second, loading conditions after emplacement in a repository are set and the pressure-resistant thickness of overpack is calculated. The corrosion thickness to achieve an assigned 1000 year life time and the required thickness to prevent radiolysis of ground water which might enhance corrosion rate are also determined. As aresult, the total required thickness of a carbon-steel overpack is conservatively estimated to 190 mm. This is a reduction of about 30% from the previous estimate provided in the 1st Progress Report. Additional items that must be considered in manufacturring and operating overpacks (i.e. sealing of vitrified waste, examination of main body and sealing welding, mechanism of handling) are evaluated on the basis of current technology, specific future data needs are identified. With respect to the concept of composite overpack (i.e., an outer vessel to provide corrosion-allowance or corrosion-resistant performance and an inner vessel to provide pressure-resistance), the differences in design concepts between the carbon-steel overpack and such composite overpacks are analyzed. Future data needs and analytical capabilities with respect to overpacks are also summarized.

Simulation of creep test on 316FR stainless steel in sodium environment at 550

Satmoko, A.*;

JNC TN9400 99-035, 37 Pages, 1999/04

JNC-TN9400-99-035.pdf:1.54MB

In sodium environment, materia1 316FR stainless steel risks to suffer from carburization. In this study, an analysis using a Fortran program is conducted to evaluate the carbon influence on the creep behavior of 316FR based on experimental results from uni-axial creep test that had been performed at temperature 550C in sodium environment simulating Fast Breeder Reactor condition. As performed in experiments, two parts are distinguished. At first, elastic-plastic behavior is used to simulate the fact that just before the beginning of creep test, specimen suffers from load or stress much higher than initial yield stress. In second part, creep condition occurs in which the applied load is kept constant. The plastic component should be included, since stresses increase due to section area reduction. For this reason, elastic-plastic-creep behavior is considered. Through time carbon penetration occurs and its concentration is evaluated empirically. This carburization phenomena are assumed to affect in increasing yield stress, decreasing creep strain rate, and increasing creep rupture strength of material. The model is capable of simulating creep test in sodium environment. Material near from surface risks to be carburized. Its material properties change leading to non-uniform distribution of stresses. Those layers of material suffer from stress concentration, and are subject to damage. By introducing a damage criteria, crack initialization can thus be predicted. And even, crack growth can be evaluated. For high stress levels, tensile strength criterion is more important than creep damage criterion. But in low stress levels, the latter gives more influence in fracture. Under high stress, time to rupture of a specimen in sodium environment is shorter than in air. But for stresses lower than 26 kgfmm, the time to rupture of creep in sodium environment is the same or little longer than in air. Quantitatively, the carburization effect at ...