Refine your search:     
Report No.
 - 
Search Results: Records 1-16 displayed on this page of 16
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Research program for ageing of nuclear power plant based characteristics of Fukui area

Sakakibara, Yasuhide; Isomura, Kazutoshi; Yamashita, Takuya; Watashi, Katsumi; Doi, Motoo; Okusa, Kyoichi; Tagawa, Akihiro; Hirahara, Kenji

Nihon Hozen Gakkai Dai-3-Kai Gakujutsu Koenkai Yoshishu, p.283 - 286, 2006/06

This study was performed to enrich the contents of measures for an ageing of the nuclear power plants at Fukui area, where infrastructures of research works, for example, institutes, universities etc. are intensively existed., according to the Road-Map established by the Atomic Energy Society of Japan at 2005.

JAEA Reports

Development of Advanced Methodology for Defect Assessment in FBR Power Plants (Phase II)

Meshii, Toshiyuki*; Watashi, Katsumi; Doi, Motoo; Hashimoto, Takashi

JNC TY4400 2004-001, 90 Pages, 2004/03

JNC-TY4400-2004-001.pdf:10.67MB

Based on the results of the Phase I of the titled research, we started Phase II scheduled to reach conclusion (develop basic methods necessary for FBR post-construction code) in three years. The following results were obtained. (1) Development of assessment procedure for multiple defects due to creep damage: We first confirmed that multiple surface cracks in axial and circumferential direction inside a cylinder can be evaluated independently by investigating multiple flaws found in a cylinder due to cyclic thermal shock. Then we developed a method to evaluate the stress intensity factors (SIFs) of the multiple circumferential cracks in a finite legth cylinder under axisymmetric loads. (2) Improving creep-fatigue crack grown (C-FCG) assessment pfodedure: We reviewed the French code A16 for flaw assessment and compared it with the JNC proposal. Since the fundamenfal philosophy for both was to evaluate the C-FCG by estimating elastic-plastic fracture mechanics parameter from the SIF, we developed a "3D crack finite element analysis system that can specify the target error in SIF." (3) Crack propagation assessment under thermal stresses (fatigue crack grown (FCG) resistace for small load): To improve accuracy of FCG assesment for components in FBR power plants (designed to minimize thermal stresses) under thermal cycles, we obtained near threshold FCG data for S55C, SUS304, HT60, SS400, 2.25Cr-1Mo, SUS316, SUS321, T91, Inconel718 by Kmax = constant test method. The results showed that FCG curves in the JSME post construction code (which is an extrapolation of the curves in ASME PVP code sec. XI) are valid in general. However, precise review of S55C, HT60's data suggested that the JSME FCG, evaluation curve may not have enough safety margin. In addition, we proposed a method to predict the decrease in the threshold SIF range DeltaKth due to high Kmax and showed its validity.

JAEA Reports

Development of advanced methodology for defect assessment in FBR power plants (Phase II); 2002 Annual Report

Meshii, Toshiyuki*; Watashi, Katsumi; Doi, Motoo; Hashimoto, Takashi

JNC TY4400 2003-002, 63 Pages, 2003/03

JNC-TY4400-2003-002.pdf:3.95MB

no abstracts in English

JAEA Reports

None

*; Fusaeda, Shigeki*; Kataoka, Shinichi*; Nemoto, Kazuhiko*; Yanagisawa, Ichiro*; Fukui, Hiroshi*; Doi, Motoo*

PNC TJ1216 97-007, 77 Pages, 1997/03

PNC-TJ1216-97-007.pdf:2.93MB

None

JAEA Reports

None

*; *; Kataoka, Shinichi*; Nemoto, Kazuhiko*; Yanagisawa, Ichiro*; Fukui, Hiroshi*; Doi, Motoo*

PNC TJ1216 97-006, 202 Pages, 1997/03

PNC-TJ1216-97-006.pdf:10.03MB

None

JAEA Reports

None

Mukai, Satoru*; Kitao, Hideo*; Tachikawa, Hirokazu*; Fusaeda, Shigeki*; Yanagisawa, Ichiro*; Doi, Motoo*; *

PNC TJ1216 96-003, 106 Pages, 1996/03

PNC-TJ1216-96-003.pdf:2.44MB

None

JAEA Reports

None

Mukai, Satoru*; *; Doi, Motoo*

PNC TJ1214 95-006, 34 Pages, 1995/03

PNC-TJ1214-95-006.pdf:0.53MB

None

JAEA Reports

None

Tachikawa, Hirokazu*; *; Doi, Motoo*

PNC TJ1214 95-002, 37 Pages, 1995/03

PNC-TJ1214-95-002.pdf:1.73MB

None

JAEA Reports

None

Tachikawa, Hirokazu*; *; Doi, Motoo*

PNC TJ1214 95-004, 187 Pages, 1995/02

PNC-TJ1214-95-004.pdf:8.35MB

None

JAEA Reports

JOYO special test report; Decay heat measurement by auxiliary cooling System

Doi, Motoo*; *; *; *

PNC TN941 81-78, 32 Pages, 1981/04

PNC-TN941-81-78.pdf:5.22MB

Decay heat measurement by auxiliary cooling system was conducted to measure the decay heat after 75 MWt 1st cycle operation of the Joyo experimental fast reactor. Auxiliary primary pump was started after plant condition reached warm stand-by, then main primary and secondary pumps were stopped after parallel operation with auxiliary loop and primary loops. Test results: (1)Decay heat was approximately 470 kW. It contains the heat which lost from the auxiliary piping in the reactor vessel. (2)Auxiliary cooling system had enough capacity to remove the decay heat at this test condition. (3)After main primary pumps were stopped, temperature difference between hot leg and cold leg of the main primary loops soon reached the limitation of this test. So this test was stopped and plant condition was reverted to warm stand-by.

JAEA Reports

Heat balance and thermal power calculations for the JOYO experimental fast reactor

*; *; Tamura, Seiji*; Doi, Motoo*; *; Yamamoto, Hisashi*

PNC TN941 80-168, 53 Pages, 1980/12

PNC-TN941-80-168.pdf:9.04MB

Heat balance measurements and calculations were performed for the JOYO experimental fast reactor. Some pertinent results of this tests are presented below. (1)The heat removal rates calcalated using the air flowrates and $$Delta$$Ts from the DHXs differed from heat removal rates obtained using measured primary sodium flowrates and $$Delta$$Ts. The heat removal rates determined from the DHX data are about 11% larger and about 4% larger than the rates obtained from the primary sodium data for the A Loop and B Loop of JOYO, respectively. The air flowrates and outlet temperatures from the DHX are measured on an 8$$times$$8 grids in the outlet area of the air cooler, using Pitot tubes and thermocouples. (2)The heat removal rates obtained using measured secondary sodium flowrates and DHX $$Delta$$Ts are almost same as and about 8% smaller than the rates obtained from the primary sodium data for the A Loop and B Loop of JOYO, respectively. (3)The core heat generation rate (including blanket effects, etc.) calculated using the measured individual subassembly outlet temperatures, and subassembly flowrates, is about 7% larger than the value calculated using the primary sodoum flowrate and the reactor $$Delta$$T. The reason for this discrepancy is conjectured to be due to the fact that the subassembly average outlet temperature is lower than the measured value, since the thermocouple is located in the center of the subassembly outlet channel. In both the A and B Loops of JOYO, the error between the heat removal rates from the measured DHX and the primary or secondary sodium flowrate and temperature data, and in the B Loop between the heat removal rates obtained from the primary sodium data and the secondary flowrate and DHX $$Delta$$T, are greater than the error band of the instrumentation. The eauses of this unbalance in the heat removal are being investigated. THis report is based on the meetings held in February and March 1980 in the Reactor Technology Section ...

JAEA Reports

JOYO 75MWt Start-up test report; PT-12 Heat transfer characteristics of IHX and DHX

Doi, Motoo*; Endo, Masayuki*; *; *; *; Wada, Hozumi*; Hirose, Tadashi*

PNC TN941 80-65, 269 Pages, 1980/05

PNC-TN941-80-65.pdf:48.51MB

The purpose of this test is to confirm that the heat transfer characteristics of the intermediate heat exchangers (IHX) and the dump heat exchangers (DHX) satisfy their design values. Each primary and secondary heat transport systems has two loops (A&B) and each secondary loop has two DHX's as terminal heat exchangers. The IHX/DHX heat transfer characteristics are measured under normal steady-state operating conditions (at 25, 40, 50, 65 and 75MWt), and also under special steady-state conditions (low sodium temperature, low secondary sodium flow rate) The major test results were : (1)The heat transfer coeffients of the IHX's and the DHX's measured in the reactor power range of 50 to 75 MWt closely matched predicted values. (2)The pressure drop in the DHX air flow, duct based on measured blower outlet pressure was less than the predicted value. This report presents the results of these plant tests.

JAEA Reports

JOYO 75MWt Start-up test report; Calbration of Chromel/Alumel thermocouples used for primary and secondary cooling Systems

*; Doi, Motoo*; *

PNC TN941 79-240, 36 Pages, 1979/12

PNC-TN941-79-240.pdf:0.92MB

Thermocouples were used to measure the temperature of the sodium in the primary and secondary cooling systems of the JOYO Experimental Fast Reactor during the 75 MW start-up test. The temperature measurement systmes (which consist of the themocouple, the compensating lead wires, and the digital recorder were calibrated by an electrical furnace usinf the melting temperature of a pure metal as a reference. New thermocouples were fabricated for the 75 MW start-up test, and were installed in the reactor vessel and the IHX inlet and outlet lines. Prior to the 75 MW start-up test, calibrations indicated differences from -0.3$$^{circ}$$C to +0.6$$^{circ}$$C from the standard reference of the freezing point of pure zinc (419.6$$^{circ}$$C). After the start-up tests, it was found that two thermocouples drifted in the positive direciton, two thermocouples drifted in the negative direction, and the remaining three thermocouples did not drift. The thermocouples used in the DHX inlet and outlet sodium lines were in use for over three years, except for the one in the A loop DHX outlet which was changed to a new one prior to the 75 MW start-up test. All of the old thermocouples used in the DHXs drifted in a positive direction from +3.9$$^{circ}$$C to +6.0$$^{circ}$$C at the pure zinc reference temperature (based on pre-test calibrations). Reasons for the drift of the chromel/alumel thermocouples were investigated in the areas of a measurement, a metalic material and a atomic energy. These investigations established the definite occurance of the drift of the thermocouples. These calibration results indicate that chromel/alumel thermocouples are not suitable for long-term use where accurate measurements are required; and if they must be used, periodic calibration is necessary.

JAEA Reports

"JOYO" Start-up test report; PT-42 Decay heat removal by auxiliary cooling system

Hirose, Tadashi*; Endo, Masayuki*; Nanashima, Takeshi*; Doi, Motoo*; Enomoto, Toshihiko*; Suzuki, Yukio*; Sekiguchi, Yoshiyuki*; Yamamoto, Hisashi*

PNC TN941 79-91, 81 Pages, 1978/12

PNC-TN941-79-91.pdf:4.66MB

This system is used to remove reactor decay heat in cases where the main cooling system is inoperable for unexpected reasons, a lower than normal sodium level exists in the R/V or during in-service inspection in the R/V. The purpose of this test is to verify that the design heat removal rate (2.6MWt) can be achieved by the Auxiliary cooling system. With the sodium level lowered below main cooling system outlet nozzles and the coolant temperature (DHX outlet temperature) controlled at 250 $$^{circ}$$C, the reactor power was increased first to approx. 1MW (1.16 MWt actual) and then to 2 MW (2.16 MWt actual) to provide the "decay heat". At both steps, steady-state conditions were verified and test data were recorded, from which the heat removal rate at design conditions was calculated. (Testing was terminated after the second step to maintain the calculated distortion of the partially-filled R/V within prescribed limits.) Test Results : At the second test condition (reactor power = 2.16 MWt) the R/V inlet Na temperature of 267 $$^{circ}$$C corresponded to a 72% open DHX inlet vane setting. Extraporating this to the design condition (R/V inlet temperature = 370$$^{circ}$$C), a 100% DHX vane opening would permit the removal of 3.1 MWt decay heat.

Oral presentation

Study on Risk-Based Maintenance approach to FBR plant

Doi, Motoo; Tsukimori, Kazuyuki; Watashi, Katsumi

no journal, , 

no abstracts in English

Oral presentation

Vibration and deformation monitoring for high temperature structures by heat-resistant FBG

Shimada, Yukihiro; Masuzumi, Takashi; Nishimura, Akihiko; Doi, Motoo*; Tsukimori, Kazuyuki

no journal, , 

We performed irreversible lattice processing into the quartz board using the technique of non-heat processing which used femtosecond laser. The heat-resistant evaluation was performed for the purpose of development of heat-resistant FBG for always supervising heat transfer system vibration and modification of a high temperature structure. As a result, the intensity ratio of diffraction light did not change but after rising temperature up to 950 $$^{circ}$$C showed that the system of heat-resistant FBG was effective as a local vibration of high temperature piping, and the distorted detection technique.

16 (Records 1-16 displayed on this page)
  • 1