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Nuclear Engineering and Design | 1994

Fatigue crack growth behavior of weld heat-affected zone of type 304 stainless steel in high temperature water

Masao Itatani; Juichi Fukakura; Masayuki Asano; Masaaki Kikuchi; Noriyuki Chujo

Abstract The fatigue crack growth behavior of the weld heat-affected zone (HAZ) of type 304 stainless steel in high temperature water which simulates the boiling-water reactor environment was investigated to clarify the effects of welding residual stress, cyclic frequency f and thermal aging on crack growth rate. A lower crack growth rate of the HAZ than of the base metal was observed in both the high temperature water and the ambient air caused by the compressive residual stress. The crack closure point was measured in the high temperature water. The effect of the welding residual stress on the crack growth rate of the HAZ can be evaluated separately from the environmental effect through the crack closure behavior. The high temperature water increased the crack growth rate at a cyclic frequency of 0.0167 Hz but did not affect it much at 3 and 5 Hz. The crack growth behavior of the thermally aged HAZ at 400 °C for 1800 h was almost the same as that of the unaged material tested at 0.0167 and 5 Hz in the high temperature water.


Nuclear Engineering and Design | 1997

Integrity assessment of the high temperature engineering test reactor (HTTR) control rod at very high temperatures

Yukio Tachibana; Shusaku Shiozawa; Juichi Fukakura; F. Matsumoto; T. Araki

The high temperature engineering test reactor (HTTR) is the first high temperature gas-cooled reactor (HTGR) in Japan with a reactor outlet coolant temperature of 950°C at high temperature test operation. The HTTR contains 16 pairs of control rods for which Alloy 800H is chosen of the metallic parts. Because the maximum temperature of the control rods reaches about 900°C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Under the guideline, temperature and stress analysis was conducted, and it is confirmed that the target life of the control rods of 5 years can be achieved.


Nuclear Engineering and Design | 1993

Effect of thermal aging on fracture toughness of RPV steel

Juichi Fukakura; Masayuki Asano; Masaaki Kikuchi; Masaaki Ishikawa

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness JIC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.


Nuclear Engineering and Design | 1991

Ductile fracture analysis of carbon steel pipe with a circumferential through-wall crack

Masayuki Asano; Juichi Fukakura; Hideo Kashiwaya; Masahiro Saito

Abstract It is necessary to make clear the pipe fracture conditions based on elastic-plastic fracture mechanics to assess the leak before break situation of carbon steel pipes for LWR plants. The aim of the present work is to discuss the effects of pipe size, initial crack length and fracture toughness on the estimated fracture load and mode of carbon steel pipes with a circumferential through-wall crack. As an analytical method, the R6-Rev.3 approach was applied to the pipe fracture analyses considering its simplicity in the application. The results indicate that the net-section collapse attainment becomes difficult with increasing diameter and decreasing thickness of carbon steel pipes. The degree of the net-section collapse attainment decreases with increasing crack length up to some critical size and then increases. The predicted fracture load is more sensitive to the materials J– R curve than to the elastic-plastic fracture toughness J IC . And a simple limit load analysis based on the yield stress is appropriate to evaluate the fracture load, as long as a proper margin was included.


Nuclear Engineering and Design | 1986

Proportional extrapolation techniques for determining stress intensity factors

Yoshiyasu Itoh; Juichi Fukakura; Hideo Kashiwaya

Abstract Proportional extrapolation techniques are proposed to compute simply and accurately the stress intensity factor by use of the boundary element method (BEM). They are based on the procedures that the effects of boundary division near crack tip on stresses and displacements are corrected by comparing with a standard problem and the corrected results are only accurate in the limit as r → 0 ( r = distance from crack tip). Comparisons of a few crack problems are made between results using the proposed techniques and those obtained by previously recommended methods. They are seen to be less sensitive than any other techniques regarding human work and accurate results are obtained even in the case of coarse boundary division.


Archive | 1992

Near-Threshold Fatigue Crack Growth of Austenitic Stainless Steels at Liquid Helium Temperature

Kenichi Suzuki; Juichi Fukakura; Hideo Kashiwaya

When designing superconducting magnets for nuclear fusion equipment or superconducting generators, consideration must be made of the difficulties of in-service inspections arising from structural complications. Through paying proper consideration, it is necessary to evaluate the remaining service life by means of fatigue crack growth analyses based on preliminarily assumed initial flaws.


Nuclear Engineering and Design | 1992

Fatigue strength of dissimilar welded joints for the main vessel of an LMFBR

Kazuo Kuwabara; Yukio Takahashi; Seiichi Kawaguchi; Yoshio Fukuda; Juichi Fukakura

Abstract The employment of welded joints composed of dissimilar metals is one simple and inexpensive way to connect a main vessel made of austenitic stainless steel and a roof slab constructed of ferritic steel in the design of liquid metal fast reactors. Since dissimilar-metal welded joints have not been used for such large structures so far in Japan, the structural integrity of this type of joint should be carefully examined for such a design option to be selected. Here various kinds of tests were conducted for eleven types of welded joints of 50 mm thickness to obtain this fundamental strength characteristics. Type 304 stainless steel was used as one of the parent metals in all the joints. They differ from each other in regard to the type of ferritic steel, welding metal and welding procedure. Low-cycle fatigue tests were conducted for round-bar specimens made from these welded joints at room temperature. Fatigue crack-propagation tests were also conducted for some of the joints. Tests after manufacturing a large-scale shell model were also conducted. The results of these tests demonstrated that the present manufacturing technique can, produce welded joints of high quality and reliability. A trial calculation for actual design conditions showed the existence of large margins against fatigue failure or fatigue crack-propagation of a significant amount.


Archive | 1987

Cryogenic Fatigue Design of Austenitic Stainless Steels for Superconducting Magnet Applications

Juichi Fukakura; Kenichi Suzuki; Hideo Kashiwaya

The 4 K fatigue life curves on the mechanical failure of domestic structural materials for superconducting magnet applications, such as type 304L and type 316L stainless steels, are obtained under axial strain control. These curves are compared with ones at 300 K and 77 K, and with the literature data on foreign materials. Then, the paper focuses on the permeability increase with cyclic straining, due to martensitic transformation. The 4 K fatigue life curves on the permeability limit of type 304L and type 316L, which are the metastable austenitic stainless steels, are obtained. The fatigue tests on the permeability at 300 K and 77 K are also conducted for comparison. Finally, a cryogenic fatigue design criterion to prevent the mechanical failure and the permeability increase of type 304L and type 316L is proposed referring to the approach in the AS ME Boiler and Pressure Vessel Code Section III.


Archive | 1986

On a Conventional Stress Intensity Factor Calculation Technique of Surface Cracks

Masayuki Asano; Kenichi Suzuki; Juichi Fukakura; Hideo Kashiwaya

This paper describes one conventional stress intensity factor calculation procedure of surface cracks by Boundary Integral Equation Method. Using both stress and displacement type singular elements, the procedure utilizes displacement and traction solutions to give accurate stress intensity factor in the analysis with coarse mesh data. The numerical results show that the present approach improves the stability of tractions near crack tip and gives accurate stress intensity factor solutions in a sense of engineering evaluation.


Transactions of the Japan Society of Mechanical Engineers. A | 1985

Size effect and crack propagation of turbine-generator rotor steel under torsional loading.

Kohsoku Nagata; Juichi Fukakura; Tadao Mori; Toshiyuki Aiba

The fatigue strength of large and small-scale smooth specimens of a turbine-generator rotor steel was examined under torsional loading. Shear strain behaviour and axial crack growth of Mode II on smooth specimens and notched specimens were investigated in relation to size effect of the rotor steel. Elastic-plastic shear stress-strain characteristics at notch root were analyzied by finite element method. The main conclusions are as follows; (1) Size effect factor defined by the ratio of fatigue limits between 100 and 10 mm diameters was 0.78 and agreed the value of JSME recomendation (2) Shear stress-strain of notch root under large scale yielding condition can be estimated by expanded Neubers rule, using inelastic nominal stress-strain relationship. (3) Using the modified parameter √(EJIIa), Mode II axial crack growth rate of smooth and notched specimens, in both elastic and plastic shear strain fields, was successfully expressed by linear regression lines.

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