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Koike, Mitsutaka; Colema, C. E.*; Causey, A. R.*; Ells, C. E.*; Hosbon, R. R.*
Zirconium in the Nuclear Industry; 11th International Symposium (ASTM STP 1295), p.469 - 491, 1998/00
None
; ; ; ; ; ;
PNC TN9410 96-235, 258 Pages, 1996/03
The chemical decontamination technique has been developed in order to remove the crud adhering to the surface of the components constructing the primary coolant system, as a part of the measure to decrease the exposure in the annual inspection. The technique has been already applied to the prototype reactor "Fugen", in the core of which the fuel assemblies were not loaded. The chemical decontamination, for the core in which the fuel assemblies are loaded, has been planned for the purpose of improving the utilization factor. It is necessary to confirm, through the test before putting the plan into practice, that the decontamination reagent does not exert a bad influence upon the components constructing the fuel assembly. This report describes the test results which have been carried out so as to investigate the influence of the reagent on the components constructing the fuel assembly. The outline of the results is as follows: (1)The susceptibility to stress corrosion cracking of the chemical decontamination treatment and the residual decontamination reagent on the components constructing the fuel assembly is low enough. (2)The chemical decontamination treatment and the residual decontamination reagent do not exert a bad influence upon the integrity of the fuel assembly concerning the fuel rod holding function of the spacer and the characteristics of the fretting wear caused on the fuel claddings.
; ;
Nihon Genshiryoku Gakkai-Shi, 38(5), p.382 - 392, 1996/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)None
; Mizuno, Junichi; *; Kawajiri, Michio*
Nihon Genshiryoku Gakkai-Shi, 37(6), p.526 - 534, 1995/06
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)None
; ; ; ;
PNC TN9410 94-052, 251 Pages, 1994/01
The structure of the pressure tube rolled joint portion for the ATR Demonstration Reactor is somewhat changed from that for Fugen, in order reduce the residual stress around the portion. Therefore, Constant Temperature Endurance Test and Thermal Cycle Endurance Test have been conducted under the reactor operating conditions except irradiation to examine the rolled joint integrity. (1)Constant Temperature Endurance Test. In fiscal 1988, Constant Temperature Endurance Test have been performed in the Component Test Loop for 2,033 hours under the reactor operating conditions (pressure 75kg/cm
, temperature 280
C) for the JP-3 specimen (total testing period 4,033 hours) and the JP-4, JP-5 specimens (total testing period 9,533 hours). After the endurance test it was found by the helium leak test that the rolled joint tightness was maintained enough. Therefore, it was confirmed that the reduction of the residual atress at the rolled joint portion which occurred at the initial stage of the operation did not affect the rolled joint tightness. (2)Thermal Cycle Endurance Test. The helium leak test and the ultrasonic flaw detection test were performed before and after Thermal Cycle Endurance Test (the unmber of thermal cycle increaced from 60 times to 140 times in fisical 1988), of which results showed the integrity of the rolled joint tightness and no crack propagation near the rolled joint portion. Therefore, it was confirmed that 140 times thermal cycle which was the design value for the reactor life of 30 years did not affect the rolled joint tightness and the initiation and propagation of a crack even for the specimen with 200 ppm hydrogen.
Mochizuki, Hiroyasu; Koike, Mitsutaka; Sakai, Takaaki
Nuclear Engineering and Design, 144(2), p.293 - 303, 1993/10
Times Cited Count：12 Percentile：74.44(Nuclear Science & Technology)None
Koike, Mitsutaka; ; Nagamatsu, Kenji; Shibahara, Itaru
10th International Symposium on Zirconium in the Nuclear Industry, 0 Pages, 1993/00
None
Koike, Mitsutaka; Onose, Shoji; Nagamatsu, Kenji; Kawajiri, Michio*
JSME International Journal, Series B, 36, p.464 - 470, 1993/00
None
; ; ; ; ; Shibahara, Itaru
PNC TN9410 92-321, 30 Pages, 1992/10
Integrity evaluations have been performed for the 2nd Fugen pressure tube test (8 years irradiation, 5.6 
10
n/cm (E
1Mev)). Test items mainly consist of tensile test, bending test, corrosion test and hydrogen analysis. It has become clear using these data that the pressure tube material has maintained its integrity during the irradiation by the integrity assessment on both tensile and fracture toughness properties. Besides, both thickness loss by corrosion and absorbed hydrogen content were lower than those of design values.
Mochizuki, Hiroyasu; Koike, Mitsutaka; Sakai, Takaaki
International Conference on Design and Safety of Advanced Nuclear Power Plants (ANP '92), 0 Pages, 1992/00
None
*; *; Fukumura, Nobuo*; *; *; *; *
PNC TN1410 91-063, 239 Pages, 1991/08
no abstracts in English
Experiments for domestically-made H.T.Zr-2.5wt% Nb pressure tube material*
PNC TN9410 90-033, 85 Pages, 1990/01
Elastic-plastic fracture toughness value J and J-R curve were obtained by compact tests at room temperaturer, accoding to the JSME standard for circumferential specimens sampled from H.T.Zr-2.5wt%Nb pressure tube which was domestically made. Judgement conditions of JSME standard for J tests were fulfilled for the present test results. For the ATR pressure tube material, the method to obtain plane-strain type elastic-plastic fracture toughness value J has been established.
*; *; *
PNC TN9410 89-102, 34 Pages, 1989/06
Pressure tubes of Advanced Thermal Reactor (boiling-light-water-cooled heavy-water-moderated pressure-tube-type reactor) in Japan are made of Heat Treated Zr-2.5wt%Nb alloy and the both ends are mechanically joined with stainless steel extension tubes. Sharp artificial cracks were introduced in the rolled jointed pressure tube specimen and the cracks were propagated and penetrated the tube wall by fatigue and DHC in high-temperature high-pressure water loop. From the results, it was shown that the LBB phenomena were valid for the rolled jointed pressure tube under the reactor operating conditions and that the critical crack length was more than 50mm. Moreover, calculations were performed about the leak rate using critical flow data.
*; *
PNC TN9410 88-044, 41 Pages, 1988/05
Erosion experiments of stainless steel were performed under water-vapour two-phase flow conditions. the experiments were performed by using Component Test Loop and erosion losses were estimated by weight differences of specimens between before and after tests. Quantitative erosion-corrosion equations both under two-phase flows and under single-phase flows were proposed by semi-theoretical considerations with the present results. also, the quantitative erosion/corrosion equation was originally made for all metals.
*; *
PNC TN9410 87-105, 47 Pages, 1987/08
The 165MWe prototype HWR Fugen has been in commercial operation since March 1979. The material of the pressure tube is heat treated Zr-2.5 wt%Nb alloy and the pressure tubes in the Fugen have been irradiated with the maximum fast neutron flux of about 3 10
n/m
sec. The pressure tubes have been inspected periodically according to the pressure tube monitoring program. In March 1984, inside diameter measurements on a small number of the pressure tubes were performed by using the pressure tube monitoring device adopting an ultrasonic wave method, and the diametrical irradiation creep and growth strain has been assessed. In February 1987, tube length measurements were performed and these data are to be used as the standard value for the estimation of the axial irradiation creep and growth strain. Besides, small diameter specimens pressurized by helium gas have begun being irradiated in the Fugen since April 1987.
