Present status of development of high chromium steel for Japanese FBR components

Wakai, Takashi; Aoto, Kazumi; Sukekawa, Masayuki*; Date, Shingo*; Shibamoto, Hiroshi

Nuclear Engineering and Design, 238(2), p.399 - 407, 2008/02

https://doi.org/10.1016/j.nucengdes.2006.09.023

This paper presents the establishment of the provisional specifications and material strength standard of the high chromium (Cr) steels for Fast Breeder Reactor (FBR) components. For the improvement of toughness and ductility of the steels, a series of mechanical tests and metallurgical examinations are performed for several kinds of high Cr steels controlled the balance of tungsten (W) and molybdenum (Mo). In addition, the effects of heat treatment conditions on material properties are also investigated. Based on these results, it is revealed that W should be diminished to achieve better ductility and toughness and that it is difficult to improve the long term properties by changing heat treatment conditions. Then the provisional specifications of the high Cr steel for FBR components are given and the provisional material strength standard is proposed for the specifications of the steel. The standard is utilized in the study on the FBR plant design.

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

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi*; Inoue, Kazuhiko*; Kasahara, Naoto

Nuclear Engineering and Design, 238(2), p.389 - 398, 2008/02

https://doi.org/10.1016/j.nucengdes.2006.09.021

In this paper, the inelastic analysis procedures 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). 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.

Extension of applicable area of thermal stress charts; Thermal stress of plate subjected to heat transfer on both surfaces

Furuhashi, Ichiro*; Kasahara, Naoto; Shibamoto, Hiroshi

Nihon Kikai Gakkai Rombunshu, B, 72(721), p.2083 - 2090, 2006/09

https://doi.org/10.1299/kikaib.72.2083

Thermal stresses in plate structures subjected to fluid temperature change on both surfaces were analyzed. Through analyses study, new design charts for temperature and stress were developed. Design charts for plate, where one surface is adiabatic, were conventionally used. By using new ones, applicable area of design charts can be greatly extended. New stress design charts were normalized by steady-state stress at fixed back surface temperature in order to reduce reading errors. Maximum stress design charts for step or ramp change of fluid temperature were made. Stress reduction by transferring step change to ramp change can be read directly from the design charts.

Development of thermal transient stress charts for screening evaluation of thermal loads

Furuhashi, Ichiro*; Kasahara, Naoto; Shibamoto, Hiroshi

JAEA-Research 2006-026, 178 Pages, 2006/03

JAEA-Research-2006-026.pdf:12.78MB

Thermal transient stress charts were developed for screening evaluation of thermal loads. Summay of obtaned results are as follows. (1) Thermal stress was theoretically analyzed on the plate subjected to thermal transient on both surfaces, and the design charts were proposed for evaluation of thermal transient stress. Compared with conventional design charts for the plate under single surface heat transfer, their applicable area is further extended. (2) Developed design charts can predict temperature and stresses responses to step or ramp change of fluid temperature. Utilizing these charts, surface temperature, average temperature in thickness, surface stress, bending stress and peak stress at arbitrary time can be obtained. (3) Non-dimensional temperature and stress were introduced, and reading errors can be reduced compared with the conventional ones. (4) Design charts were also proposed on the maximum thermal stresses and their arising times. It was revealed that the maximum thermal stresses never exceed 2 times of steady-state stress under the fixed back surface temperature. (5)Green functions of transient temperature and thermal stresses were developed. Temperature and thermal stresses can be predicted within 1.4% error. These charts will contribute to the screening evaluation of thermal loads with their locations, and will be employed for sensitive analyses for design and understanding of thermal stress mechanisms.

Measurement of thermal ratcheting strain on the structures by the laser speckle method

Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto

Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 7 Pages, 2006/00

Prevention of thermal ratcheting is an important problem for high temperature components of fast breeder reactors that are subjected to cyclic thermal loads. To clarify ratcheting behaviors, structural model tests were planned. Strain measurement is important for understanding the thermal ratcheting phenomenon, however the conventional measurement by strain gauge is difficult at high temperature. Then, Laser speckle strain measurement system using the dual-beam set-up was developed to apply to high temperature structural model tests. This system was applied to the thermal ratcheting tests, which demonstrated the actual operative conditions of reactor vessels. Through comparison with uniaxial test results obtained by extensometers, the laser speckle method was verified. Measured data of structural model tests were utilized to certify the guidelines of inelastic analysis for design, which provide prediction method of strain in components of fast reactor.

DEVELOPMENT OF GUIDELINE FOR THERMAL LOAD MODELING

Shibamoto, Hiroshi; Nagashima, Hedeaki; Inoue, Kazuhiko; Kasahara, Naoto; Jimbo, Masakazu*; Furuhashi, Ichiro*

Proceedings of 2005 ASME International Mechanical Engineering Congress and Exposition (CD-ROM), 0 Pages, 2005/08

Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.

Study on Advanced Structural Design for Commercialized Fast Breeder Reactors

Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai; Inoue, Kazuhiko*; Shibamoto, Hiroshi*; Tanaka, Yoshihiko*

JNC TY9400 2004-025, 984 Pages, 2004/08

JNC-TY9400-2004-025.pdf:159.61MB

Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company(JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R&D activities for the feasibility studies on commerdalized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study, This report contains the results of the research on the structural design technology. The research scope was identified as (1) FDS(FBR Design Standard), (2) Standardization of new material, and (3)System Based Code for Integrity, and the results of this year's studies are summarized as follows. (1)FDS (FBR Design Standard) * As for failure criteria, ratcheting-fatigue tests were continued. Applicability of rational settling method on creep design regime was evaluated and evaluation method of primary stress was studied. * As for a guideline on inelastic analysis for design, development of conservative detail modle (CRIEPI model for design) is underway. Loading history effect was evaluated through analysis. Conservative evaluation method of creep-fatigue damage coped with inelastic analysis was also developed. Aiming for verification of the guidline, structure model test simulated sodium surface level of reactor vessel is continuing. Policy and items of the guideline were studied. * As for a guideline on thermal loads modeling for design, provisions of the guideline on rational settling method of thermal striping loads were discussed. Screening method to grasp severe thermal load and parts in higher stress was developed. (2)Standardization of new material * As for candidate 12-chromium stainless steel (added tungsten, non-added tungsten), that is expected to improve strength of components of commercialized fast reactor, short and medium-term strength tests (including long-term aged test piece), ob

Development of the guideline on inelastic analysis for design

Tanaka, Yoshihiko; Shibamoto, Hiroshi; Inoue, Kazuhiko; kasahara, Naoto; Ando, Masanori; Ito, Kei

ASME PVP-Vol.472, p.53-60, p.53 - 60, 2004/07

The guideline on inelastic analysis for design, one of the key items of Fast Reactor Design Standard(FDS), is being developed.The basic policies of this guideline are as follows:(a) to emphasis conservative analysis output rather than nominal value representing actual behavior, (b) to clarify the applicable area for assurance of conservative results. With such concepts, it would be possible that the guideline provides useful explanations on the manner of analysis and estimation in the form of concrete examples of design as well as general rules (somehow vague). As the first step of the guideline development, the following five issues to be solved were extracted:1) applicable area, 2) selection of constitutive equation, 3) modeling method of the load history, 4) ratchet strain and creep fatigue damage evaluation methods by inelastic analysis and 5) example design problems to check users' analysis quality and to complement the general rules. In parallel, inelastic analyses with the promising constitutive equations were applied by way of trial to obtain rough presumption on their effects on structural design of the components. As a result,all inelastic analyses provided smaller cumulative strains and equivalent strain ranges than the existing design method based on elastic analysis,suggesting advantage of introducing them into actual design.

Development of thermal stress screening method; Application of Green function method

Furuhashi, Ichiro*; Shibamoto, Hiroshi; Kasahara, Naoto

JNC TN9400 2004-013, 118 Pages, 2003/11

JNC-TN9400-2004-013.pdf:5.2MB

This work was achieved for the development of the screening method of thermal transient stresses in FBR components. We proposed an approximation method for evaluations of thermal stress under variable heat transfer coefficients (non-linear problems) using the Green functions of thermal stresses with constant heat transfer coefficlents (linear problems). Detailed thermal stress analyses provided Green functions for a skirt structure and a tube-sheet of intermediate Heat Exchanger. The upper bound Green functions were obtained by the analyses using those upper bound heat transfer coefficients. The medium and the lower bound Green functions were got by the analyses of those under medium and the lower bound heat transfer coefficients. Conventional evaluations utilized the upper bound Green functions. On the other hand, we proposed a new evaluation method by using the upper bound, medium and the lower bound Green functions. The comparison of above results gave the results as follows. The conventional evaluations were conservative and appropriate for the cases under one fluid thermal transient structure such as the skirt. The conventional evaluations were generally conservative for the complicated structures under two or more fluids thermal transients such as the tube-sheet. But the danger locations could exists for the complicated structures under two or more fluids transients, namely the conventional evaluations were non-conservative. The proposed evaluations gave good estimations for these complicated structures. Though above results, we have made the basic documents of the screening method of thermal transient stresses using the conventional method and the new method.

Study on Advanced Structural Design for Commercialized Breeder Reactors

Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai; Sagayama, Yutaka*; Dozaki, Koji*; Shibamoto, Hiroshi*; Tanaka, Yoshihiko*

JNC TY9400 2002-025, 889 Pages, 2003/01

JNC-TY9400-2002-025.pdf:26.72MB

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