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JAEA Reports

The 3rd technological meeting of Tokai reprocessing plant

Maki, Akira; ; Taguchi, Katsuya; ; Shimizu, Ryo; Shoji, Kenji;

JNC-TN8410 2001-012, 185 Pages, 2001/04

JNC-TN8410-2001-012.pdf:9.61MB

"The third technological meeting of Tokai Reprocessing plant (TRP)" was held in JNFL Rokkasyo site on March 14$$^{th}$$, 2001. The technical meetings have been held in the past two times. The first one was about the present status and future plan of the TRP and second one was about safety evaluation work on the TRP. At this time, the meeting focussed on the corrosion experrience, in-service inspection technology and future maintenance plan. The report contains the proceedings, transparancies and questionnaires of the meeting are contained.

JAEA Reports

Report of radiation exposure control on the 12th periodic inspection at experimental fast reactor JOYO

; Kano, Yutaka; ; Shindo, Katsutoshi

JNC-TN9410 2000-001, 20 Pages, 1999/12

JNC-TN9410-2000-001.pdf:1.84MB

The 12th periodic inspection had been executed at the experimental fast reactor JOYO from February 24,1998 to June 28,1999. This inspection had been extended about three months because it was addtion to the work for the safety countermeasure. The result of collective dose equivalent was 263.92 man*mSv, whereas, the expected collective dose equivalent was about 407 man*mSv in the whole period of this inspection. It was confirmed that this inspection was carried out with the suitable radiation protection programmes. In this report, provided in 12th periodic inspection, were described with taking the results of the past periodic inspections into consideration.

JAEA Reports

Report of radiation exposure control on the 11th periodic inspection at experimental fast reactor JOYO; Reported by radiation control section

; ; ;

PNC-TN9410 97-094, 27 Pages, 1997/10

PNC-TN9410-97-094.pdf:0.85MB

The 11th periodic inspection had been executed at the experimental fast reactor JOYO from May 10,1995 to March 24,1997. Because the inspection had been extended several times, the time span of external exposure control was divided into two period. The result of collective dose equivalent in the previous term(from May 10,1995 to December 7,1996: about seventeen months) was 243.34 man*mSv, whereas, the expected collective dose equivalent was about 280man*mSv. The result of collective dose equivalent in the latter term (from December 8,1996 to March 24,1997: about three months) was 44.73 man*mSv, whereas, the expected collective dose equivalent was about 85man*mSv. The collective dose equivalent in the whole period of this inspection was 288.07 man*mSv. It was confirmed that this inspection was carried out with the suitable radiation protection programmes. In this report, the method for the control of external exposure and the reduction of external exposure, provided in 11th periodic inspection, were described with taking the results of the past periodic inspections into consideration.

JAEA Reports

Effects of the chemical decontamination on the component parts of the ATR fuel assembly

; ; ; ; ; ;

PNC-TN9410 96-235, 258 Pages, 1996/03

PNC-TN9410-96-235.pdf:41.18MB

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.

JAEA Reports

None

; ; ; Mikami, Satoshi; ; ; Ebana, Minoru

PNC-TN8520 96-001, 2536 Pages, 1996/03

PNC-TN8520-96-001.pdf:62.42MB

JAEA Reports

Integrity evaluations for the 2nd Fugen pressure tube surveillance test

; ; ; ; ; Shibahara, Itaru

PNC-TN9410 92-321, 30 Pages, 1992/10

PNC-TN9410-92-321.pdf:0.67MB

Integrity evaluations have been performed for the 2nd Fugen pressure tube test (8 years irradiation, 5.6 $$times$$ 10$$^{21}$$n/cm$$^{2}$$ (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.

JAEA Reports

None

; ; ; ; ;

PNC-TN8470 92-006, 224 Pages, 1992/09

PNC-TN8470-92-006.pdf:4.31MB

no abstracts in English

JAEA Reports

Mockup test apparatus for the inspection system of steam generator tubes; Design and Manufacturing

; ; ; ;

PNC-TN9410 92-254, 76 Pages, 1992/07

PNC-TN9410-92-254.pdf:2.02MB

A verification test of the inspection system of Monju steam generator(SG) tubes will be performed in near future. Mockup Test Apparatus for the inspection system of SG tubes was manufactured and installed at Mechatronics Application Reserch Facility (MARF) in OEC. The test apparatus has the same specification, which is prepared for verification test, as Monju plant; for instance, which are dimension and material of tubes, and workability for the inspection equipment. About one hundred and forty SG tubes are radially arranged in tube sheets in Monju SG, however, three tubes, inner, center and outer one, are sellected in this test apparatus for testing of inspection system, It was verified that the test apparatus was manufactured with the same accuracy and dimension as Monju. System verification test is planned using this test apparatus.

JAEA Reports

Measurement and evaluation of radioactive corrosion product behavior in primary sodium circuits of JOYO (II)

; Chatani, Keiji; ; ;

PNC-TN9410 92-224, 81 Pages, 1992/07

PNC-TN9410-92-224.pdf:1.87MB

The radioactive corrosion product (CP) deposition density and gamma dose rate have been measured along the primary sodium circuits in Experimental Fast Reactor "JOYO" during every annual inspection and the CP behavior analysis code "PSYCHE" has been verified with measurement data in order to contribute the reduction of exposure dose of plant personal. The deposition density is measured by using a pure germanium detector system and determined by multiplying count rates by conversion factor. Gamma dose rate is measured with CaSO$$_{4}$$ thermoluminescence dosimeters (TLD). This report presents measurement results during the 9th annual inspection and the evaluation results for all data measured so far. The results on this study are summarized as follows: (1)Major CP nuclides deposited along the primary sodium circuits are $$^{54}$$Mn and $$^{60}$$Co. $$^{54}$$Mn is most dominant isotopes. Amounts of deposited $$^{54}$$Mn is about twenty times as much as those of $$^{60}$$Co. (2)$$^{54}$$Mn is deposited mainly on the cold leg pipings between the outlet of the intermediate heat exchanger (IHX) and the inlet of the reactor vessel. $$^{60}$$Co is deposited mainly on the hot leg pipings between the outlet of the reactor vessel and the inlet of IHX. (3)The buildup of $$^{54}$$Mn is saturated at 4$$sim$$4.5 EFPY. The averaged dose rate of the pipings is saturated at about 1.5mSV/h. The dose rates of IHX and primary sodium pump are about 1.5 mSv/h and 2.1 mSv/h, respectively. The dose rate distributions around IHX and primary sodium pump show the peaks at the stagnant part of the flow and at the turbulence part. (4)Calculation by "PSYCHE" and measurement are compared. Calculation-to-measurement ratio is 1.2 for the CP deposition density and 1.5 for the dose rate. It can be said that the features of the CP behavior in the primary circuit of "JOYO" is made clear. The more effort will be required for the evaluation of CP behavior for subassemblies such as outer reflectors, clearness of ...

JAEA Reports

Experimental fast reactor "JOYO" operation test; Operation history of auxiliarry core cooling system

*; *; *; *

PNC-TN941 83-08, 51 Pages, 1983/03

PNC-TN941-83-08.pdf:1.34MB

Experimental Fast Reactor "JOYO" completed its 75MWt power operation with the Mark-I core (breeding core) in Dec., 1981. The Auxiliary Core Cooling System (ACCS) has been operated satisfactorily since the first sodium charge in Jan., 1977. This paper describes the operation history until the end of Mark-I Operation. (1)Accumulated operation time of Primary Auxiliary Cooling System with circulating pump on during outages for annual inspection or others was only 530 hours, while the rest being occupied by the counterflow from main circulating punps. Auxiliary circulating pump experienced the automatical start only when the reactor scrammed and reactor sodium level lowered because of the failure of overflow make-up pump in July, 1981. (2)Secondary auxiliary cooling system had been operated approximately 39,140 hours in full flow rate, meanwhile circulating pump failed 4 times because of power loss.

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