Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2014

Watanabe, Hitoshi; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Inoue, Kazumi; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; Kikuchi, Masaaki*; Sakauchi, Nobuyuki*; et al.

JAEA-Review 2015-030, 115 Pages, 2015/12

JAEA-Review-2015-030.pdf:25.28MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2014. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2012

Sumiya, Shuichi; Watanabe, Hitoshi; Miyagawa, Naoto; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Inoue, Kazumi; Yoshii, Hideki; Otani, Kazunori*; Hiyama, Yoshinori*; et al.

JAEA-Review 2013-041, 115 Pages, 2014/01

JAEA-Review-2013-041.pdf:19.01MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, and the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki prefecture), this report describes the effluent control results of liquid waste discharged from the JAEA's Nuclear Fuel Cycle Engineering Laboratories in the fiscal year 2012, from 1st April 2012 to 31st March 2013. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other facilities were much lower than the authorized limits of the above regulations.

Material Properties of High Cr-Mo Steel (III) Mechanical Properties of HCM12A(FBR) after Thermal Aging

Ando, Masanori; Kato, Shoichi; Yoshida, Eiichi; Suzuki, Takaichi*

JNC TN9400 2003-113, 49 Pages, 2004/02

JNC-TN9400-2003-113.pdf:2.85MB

In the FBR components,cyclic thermal loads are predominant and creep deformation must be taken into account.Therefore,the applicability of high chromium ferritic steel for structural material of the future advanced Fast Breeder Reactor is investigated in the feasibility study on commercialized FBR cycle systems,since both of thermal properties and high temperature strength of the steel are superior to those of conventional austenitic stainless steels. In this study, tensile, hardness,impact and relaxation tests are conducted in order to evaluate the basic mechanical properties of each HCM12A (2001-FBR). The material is aged at 600C for 3000h and 6000h. The aged materials are also tested as well as the as-received one. As a result, the following conclusions are obtained;

Material test data of SUS304 steel (III)

Odaka, Susumu; Kato, Shoichi; Kawakami, Tomohiro*; Suzuki, Takaichi*; Takamori, Yuji*

JNC TN9450 2003-002, 962 Pages, 2003/03

JNC-TN9450-2003-002.pdf:35.27MB

Material test of SUS304 steel, which was used for structure material of fast breeder reactor (FBR) has been preformed in New Technology Development Group. In this report, the result of test obtained up to this time was collected. Many valuable data that long time test more than 100000 hours including sodium environment test data in this report will be use for material examination of FBR in the future. Contents of the data sheet are as follows; (1) Material: SUS304 Steel, (2) Test environment: In air, in sodium, (3) Test temperature: Room temperature to 800C, (4) Test method: According to JIS and FBR metallic materials test manual (revised edition), (5) Number of data : Tensile tests 1,185 Creep tests 1,044 Fatigue tests 1,037 Creep-fatigue tests 263 Total 3,529 This report consists of the printouts from -the structural material data Processing system, SMAT-.

Radiation Protection against Tritium in the Fugen Nuclear Power Station

Matsushima, Akira; ; ; ; ; ;

Saikuru Kiko Giho, (11), p.81 - 91, 2001/06

https://rdreview.jaea.go.jp/gihou/pdf2/5948.pdf

None

Materials properties data sheet (No.B 04); Tensile properties data on FBR grade 316 (Weld Joint)

; ; *; *

PNC TN9450 96-001, 70 Pages, 1996/01

PNC-TN9450-96-001.pdf:1.35MB

In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the tensile properties of FBR grade 316 (Abbreviation 316FR), based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; (1)Material : 316FR (Weld Joint , Weld Metal) (2)Test temperature : R.T650C (3)Test method : According to JIS and FBR Metalic Materials Test Method (4)Number of data : 40 points

Material properties data sheet (No.B 01 R 01); Tensile properties data on FBR grade 316FR (base metal)

; ; Yoshida, Eiichi; *; *;

PNC TN9450 95-003, 98 Pages, 1995/02

PNC-TN9450-95-003.pdf:1.94MB

In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the tensile properties of FBR grade 316FR on air and sodium environment conditions. Contents of the data sheet are as follows; (1)Material : FBR grade 316FR (Base Metal) (a)B6 heat 25mm1,000㎜1,000mm (Plate) (b)B7 heat 50mm1,000mm1,000mm (Plate) (c)B8 heat 40mm1,000mm1,000mm (Plate) (d)B9 heat 25mm1,000mm1,000mm (Plate) (d)B11 heat 50mm1,000mm1,000mm (Plate) (2)Pre-test treatment: (a)Argon aged for 5000hr at 500, 550, 600C (B6, B7, B8 Heats) (b)Argon aged for 20000hr at 500, 550, 600C (B7, B8 Heats) (c)Sodium exposed for 5000hr at 500, 550, 600C (B6, B7, B8 Heats) (d)Sodium exposed for 20000hr at 500, 550, 600C (B6, B7, B8 Heats) (e)As-recieved (B6, B7, B8, B9, B11 Heats) (3)Test temperature : R.T.750C (4)Test method : Accoding to JIS and FBR Metallic Materials Test Methods (5)Number of data : 153 points (Including 64 points on the last report)

Materials properties data sheet (No. F02); Creep properties data on Mod.9Cr-1Mo steels (Base Metal)

; ; *; *; *; Yoshida, Eiichi;

PNC TN9450 91-010, 259 Pages, 1991/10

PNC-TN9450-91-010.pdf:4.55MB

In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the creep properties of Mod.9Cr-1Mo steels for steam generator, based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; Material : Mod.9Cr-1Mo steels (Base Metal) Plate 7 Heats (F2, F6, F7, F9, F10, NSC1, NSC2) Forging 8 Heats (F4, F5, F8, F11, VIM, ESR, F520, F550) Tube 1 Heats (F3) Test temperature : 450650C Test method : According to JIS and FBR Metallic Materials Test Method Test environment : In Air and in Sodium Number of deta : 314 points