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JAEA Reports

Report on research and development of nuclear forensics technologies

Okubo, Ayako; Kimura, Yoshiki; Shinohara, Nobuo; Toda, Nobufumi; Funatake, Yoshio; Watahiki, Masaru; Sakurai, Satoshi; Kuno, Yusuke

JAEA-Technology 2015-001, 185 Pages, 2015/03

JAEA-Technology-2015-001.pdf:56.65MB

Nuclear forensics is the analysis of intercepted illicit nuclear or radioactive material and any associated material to provide evidence for nuclear attribution by determining origin, history, transit routes and purpose involving such material. Nuclear forensics activity includes sampling of the illicit material, analysis of the samples and evaluation of the attribution by comparing the analyzed data with database or numerical simulation. Because the nuclear forensics technologies specify the origin of the nuclear materials used illegal dealings or nuclear terrorism, it becomes possible to identify and indict offenders, hence to enhance deterrent effect against such terrorism. Worldwide network on nuclear forensics can contribute to strengthen global nuclear security regime. In this paper, the results of research and development of fundamental nuclear forensics technologies performed in Japan Atomic Energy Agency during the fiscal term of 2011-2013 were reported.

JAEA Reports

Development of measurement technique of the equilibrium vapor pressure of simulated fission products

Takai, Toshihide; Nakajima, Kunihisa; Furukawa, Tomohiro

JAEA-Technology 2015-002, 20 Pages, 2015/03

JAEA-Technology-2015-002.pdf:3.46MB

To improve the evaluation technique of source term, the measurement technique of the equilibrium vapor pressure using a high temperature mass spectrometer is required to expand the thermodynamic database of the simulated FPs. Existing test apparatus was adapted for this purpose. A mass spectrometer capable of measuring a wide mass number range and glove box for handling simulated FPs were installed for analyzing heavy FPs and preventing deterioration of simulated FPs in an air atmosphere, respectively. Function verification using standard sample and precision investigation using simulated FP sample were carried out. The oxygen dissociation pressure and standard enthalpy of formation of RuO$$_{2}$$(s) were evaluated, and it was confirmed these evaluated values were agreed with the calculated value from existing thermodynamic data and evaluation value written in the literature. Consequently, it was proven that high precision thermodynamic data was able to obtain by using this apparatus.

JAEA Reports

Preliminary examination about the seal leak using the photocoagulation resin (Joint research)

Ooka, Makoto; Maekawa, Yasunari; Tomizuka, Chiaki; Murakami, Tomoyuki*; Katagiri, Genichi*; Ozaki, Hiroshi*; Kawamura, Hiroshi

JAEA-Technology 2015-003, 31 Pages, 2015/03

JAEA-Technology-2015-003.pdf:3.95MB

An action for the decommissioning of the Fukushima Daiichi Nuclear Power Station (Tokyo Electric Power Company) is pushed forward now. For fuel debris Remove, it is necessary to fill the Primary Containment Vessel (PCV) with water. Because a coolant leaks out from the PCV, it becomes the most important problem to seal leak the coolant. Nuclear Plant Decommissioning Safety Research Establishment has examined the method of seal leak using the photocoagulation resin. However, originally the photocoagulation resin is used as coating or the painting, and the applicability to seal leak water is unknown. This report describes the result that examined the applicability to seal leak using photocoagulation resin.

JAEA Reports

A Study on laser welding of ferritic/martensitic steel (PNC-FMS) for fast reactor fuel assemblies

Kono, Fumiaki; Sogame, Motomu; Yamada, Tomonori; Shobu, Takahisa; Naganuma, Masayuki; Ozawa, Takayuki; Muramatsu, Toshiharu

JAEA-Technology 2015-004, 57 Pages, 2015/03

JAEA-Technology-2015-004.pdf:20.87MB

Laser welding of ferritic/martensitic steel (PNC-FMS) sheets with different thicknesses (2 mm and 5 mm) was examined to investigate the weldability between the inner and outer duct in fast reactor fuel assemblies with inner duct structure (FAIDUS); the objective of the inner duct is to avoid the re-criticality in case of the core melting accident. Laser-spot and melt-run welding was performed at various laser powers, welding times and velocities to find out the appropriate welding conditions with few defects and enough penetration depth. As for the spot welding, furthermore, slow cooling rate or pulsed laser irradiation could reduce the crack and porosity in the welded zone. The strain of the welded zone almost disappeared and the hardness was comparable with that of the base metal by applying post welding heat treatment at 690 $$^{circ}$$C for 103 min. In addition, the shear strength of welded joints was confirmed to be sufficiently higher than the provisional allowance shear stress. These results indicate that laser welding would be probably applied to the PNC-FMS inner and outer ducts.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Observation techniques development in a reactor vessel of the fast reactor, 2

Okuda, Eiji; Sasaki, Jun; Suzuki, Nobuhiro; Takamatsu, Misao; Nagai, Akinori

JAEA-Technology 2015-005, 36 Pages, 2015/03

JAEA-Technology-2015-005.pdf:44.42MB

In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. In order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. The IVO equipment for the Upper Core Structure (UCS) fitting area was specifically developed in the experimental fast reactor "Joyo". And the IVO was successfully completed as shown below. (1) Improvement of picture quality and resolution. The IVO of UCS fitting area with the gap of 5mm in minimum was achieved using the IVO equipment with video-scope under the actual reactor environment. The picture quality and resolution could be improved comparing with the radiation resistant fiberscope which was used in past IVO. (2) Prevention of video-scope hypofunction by high temperature / radiation dose. Since video-scope is inferior in thermal and radiation resistance, the IVO equipment was designed to be able to withdraw and insert video-scopes with cooling gas. This measure could achieve the observation in short radiation time with available temperature under the actual reactor environment. The IVO equipment for UCS fitting area provided useful information on UCS replacement. In addition, the experience provided valuable insights into further improvements for IVO techniques in SFRs.

JAEA Reports

Improvement in oil seal performance of gas compressor in HTTR, 2

Nemoto, Takahiro; Kaneshiro, Noriyuki*; Sekita, Kenji; Furusawa, Takayuki; Kuroha, Misao; Kawakami, Satoru; Kondo, Masaaki

JAEA-Technology 2015-006, 36 Pages, 2015/03

JAEA-Technology-2015-006.pdf:16.77MB

The High-Temperature engineering Test Reactor (HTTR) has been developed for establishing and upgrading the technical basis of HTGR.HTTR facilities have their structures, systems and a lot of components including reciprocating gas compressors, commonly used to extract and/or discharge reactor coolant helium gas contained in primary/secondary coolant systems. From the fact of the operational experiences of these compressors, seal-oil leakage has been frequently observed, although rod-seal mechanisms with complicated structures are equipped and improved for preventing coolant helium gas. So, we tried to change the rod-seal materials which might be a primary reason of frequent seal-oil leakage, that resulted in decreasing a mass and frequently of seal-oil leakage. It is confirmed that it is important to select adequate materials of rod seal for sliding speed of the piston of the compressor to prevent seal-oil leakage. Additionally, the procedure to estimate seal-oil leakage for each compressor is discussed. This report describes the results of investigation for improvement on seal-oil leak tightness of the compressors in HTTR facilities.

JAEA Reports

The Uranium waste fluid processing examination by liquid and liquid extraction method using the emulsion flow method

Kanda, Nobuhiro; Daiten, Masaki; Endo, Yuji; Yoshida, Hideaki; Mita, Yutaka; Naganawa, Hirochika; Nagano, Tetsushi; Yanase, Nobuyuki

JAEA-Technology 2015-007, 43 Pages, 2015/03

JAEA-Technology-2015-007.pdf:5.33MB

The centrifuge which has the subtlety information concerning the nuclear nonproliferation used for uranium enrichment technical development exists in the uranium enrichment facilities of Ningyo-toge Environmental Engineering Center, Japan Atomic Energy Agency. This centrifugal is performing separation processing of the radioactive material adhering to the surface of parts by wet decontamination of ultrasonic cleaning by dilute sulfuric acid and water, etc. By removing the uranium contained in waste fluid, generated sludge reduces activity concentration. And the possibility of reduction of sludge processing is examined. For this reason, from the 2007 fiscal year, Nuclear Science and Engineering Directorate and cooperation are aimed at, and development of the extraction separation technology of the "uranium" by the emulsion flow method is furthered. The test equipment using the developed emulsion flow method was tested. And dilute sulfuric acid and water were used for the examination as actual waste fluid. The result checked whether the various performances in Basic test carried out in Nuclear Science and Engineering Directorate would be obtained.

JAEA Reports

Development of denitration technique for MA-type bituminized waste product by aqueous leaching method

Irisawa, Keita; Komatsuzaki, Toshio; Kawato, Yoshimi; Sakakibara, Tetsuro; Nakazawa, Osamu; Meguro, Yoshihiro

JAEA-Technology 2015-008, 28 Pages, 2015/03

JAEA-Technology-2015-008.pdf:13.63MB

In JAEA, 16,671 drums of intermediate-radioactive bituminized waste products (BWPs) have been stored in asphalt solidification storages. As a way of reduction of uncertainty in assessment of disposal of the BWPs, a processing technique of separation of nitrate salts from the BWP by means of an aqueous leaching method was studied. As elemental techniques for the denitration process, (1) crushing techniques of a BWP and (2) denitration techniques for the crushed BWP by the aqueous leaching method were investigated. In order to promote leaching amounts of nitrates, the BWP was crushed, and the grain size distribution was investigated by sieving. Moreover, leaching behaviors of nitrate, nitrite and elements as radionuclides including in the BWP were investigated.

JAEA Reports

Study on the evaluation methods to determine the radioactivity concentration in low-level radioactive wastes generated from JPDR facilities, 2

Tsuji, Tomoyuki; Sakai, Akihiro; Izumo, Sari; Amazawa, Hiroya

JAEA-Technology 2015-009, 46 Pages, 2015/06

JAEA-Technology-2015-009.pdf:1.45MB

It is necessary to establish practical evaluation methods to determine radioactivity concentrations of the important nuclides for safety assessment on disposal of radioactive wastes in order to dispose of low-level radioactive wastes generated from various nuclear facilities in JAEA. In this report, it has been studied that the practical evaluation methods are applied for the important nuclides (H-3, C-14, Cl-36, Ni-59, Co-60, Ni-63, Sr-90, Mo-93, Nb-94, Tc-99, Ag-108m, Cs-137, Eu-152, Eu-154, Ho-166m, $$alpha$$ nuclides) of radioactive wastes generated from JPDR facilities. As a result, it has been found that the appropriate methods to determine radioactivity concentrations such as the scaling factor method (Ni-63, Nb-94), the mean activity concentration method (H-3, C-14, Cl-36 and so on) and the theoretical method (Ni-59) can be applied and Co-60, Ag-108m and Cs-137 will be evaluated by $$gamma$$ measurements from outside of the waste package.

JAEA Reports

Developments of high-performance moderator vessel for JRR-3 cold neutron source

Arai, Masaji; Tamura, Itaru; Hazawa, Tomoya

JAEA-Technology 2015-010, 52 Pages, 2015/05

JAEA-Technology-2015-010.pdf:7.11MB

In the Department of Research Reactor and Tandem Accelerator, developments of high-performance CNS moderator vessel that can produce cold neutron intensity about two times higher compared to the existing vessel have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the design and manufacture of new vessel. In the present study, design strength evaluation, mockup test, simulation for thermo-fluid dynamics of the liquid hydrogen and strength evaluation of the different-material-bonding were studied. By these evaluation results, we verified that the developed new vessel can be applied to CNS moderator vessel of JRR-3.

JAEA Reports

Results of borehole investigation in -500m access/research gallery-north (13MI38$$sim$$13MI44 Boreholes)

Hasegawa, Takashi; Kawamoto, Koji; Yamada, Nobuto; Onuki, Kenji; Omori, Kazuaki; Takeuchi, Ryuji; Iwatsuki, Teruki; Sato, Toshinori

JAEA-Technology 2015-011, 135 Pages, 2015/07

JAEA-Technology-2015-011.pdf:28.63MB
JAEA-Technology-2015-011-appendix(CD-ROM).zip:566.32MB

The geological, hydraulic and geochemical data such as rock mass classification, groundwater inflow points and the volume, water pressure, and hydraulic conductivity were obtained from boreholes (13MI38$$sim$$13MI44) in the -500m Access/Research Gallery-North of Mizunami Underground Research laboratory (MIU). In addition to data acquisition, monitoring systems were installed to observe hydrochemical changes in the groundwater, and rock strain during and after the groundwater recovery experiment.

JAEA Reports

Validation of system analysis code with HTTR thermal load fluctuation test data (non-nuclear heating) and evaluation of reactor temperature behavior during upsets in hydrogen production plant

Honda, Yuki; Sato, Hiroyuki; Nakagawa, Shigeaki; Takada, Shoji; Tochio, Daisuke; Sakaba, Nariaki; Sawa, Kazuhiro

JAEA-Technology 2015-012, 17 Pages, 2015/06

JAEA-Technology-2015-012.pdf:11.38MB

Japan Atomic Energy Agency (JAEA) proposed a draft safety requirement, which consists of the requirements for constructing a H$$_{2}$$ plant under conventional chemical plant regulations as well as the requirements for collocation of a nuclear facility and a H$$_{2}$$ plant. One of the key requirements is to maintain reactor normal operation condition during every possible condition in the H$$_{2}$$ plant. In order to show that the requirement can be reasonably achieved, a system analysis code is validated with the HTTR experimental data obtained in January 2015. The validated code is applied for the evaluation of a postulated abnormal event in H$$_{2}$$ plant to be connected to the HTTR. The results showed that the evaluation items such as reactor power and reactor outlet coolant temperature do not exceed evaluation criteria. As a conclusion, a feasibility of H$$_{2}$$ plant construction under non-nuclear regulations is validated by showing that the stable reactor operation can be achieved against temperature transients induced by abnormal conditions in the H$$_{2}$$ plant.

JAEA Reports

Development of waste management system

Mitsuda, Motoyuki; Sasaki, Toshiki

JAEA-Technology 2015-013, 29 Pages, 2015/06

JAEA-Technology-2015-013.pdf:0.96MB

For implementation of disposal of the radioactive waste generated from Japan Atomic Energy Agency, Waste Management System which manages all of the waste data has been developed. We surveyed the kinds of data needed for the waste management at each site, and we set the standard waste management data items. We developed conceptual design for the waste management system and established the system for major sites, Nuclear Science Research Institute, Ningyo-toge Environmental Engineering Center, Fugen Decommissioning Engineering Center, Oarai Research and Development Center, Nuclear Fuel Cycle Engineering Laboratories. For other small sites, we accumulate waste data to the common waste storage database. Therefore, we have developed the system which manages the quality assurance waste data depending on waste treatment situation in JAEA.

JAEA Reports

Sensitivity of radioactive nuclide leaching rate under the sub-surface disposal

Tsuji, Tomoyuki; Nakamura, Yasuo; Nakatani, Takayoshi

JAEA-Technology 2015-014, 34 Pages, 2015/06

[The article has been found to have a problem about reliability of the corrosion data acquisition, and thus it is unavailable to download the full text in accordance with authors' intentions to retract the report.] In order to dispose of radioactive wastes for sub-surface disposal, JAEA has studied the safety assessment for likely scenario and less-likely scenario. Radioactive nuclide leaching rate under the sub-surface disposal is important parameter in the safety assessment because radioactive nuclides in activated metal wastes are released with its corrosion. In this report, sensitivity of radioactive nuclide leaching rate is studied for the safety assessment. As the result, it is confirmed that Cl-36 which is dominant for the safety assessment in groundwater scenario is sensitive to radioactive nuclide leaching rate, but Nb-94 which is dominant in tunnel excavation scenario is not sensitive to radioactive nuclide leaching rate but to distribution coefficients in engineered barrier.

JAEA Reports

Study on the evaluation method to determine the radioactivity concentration waste generated from post-irradiation examination facilities

Hoshino, Yuzuru; Sakamoto, Yoshiaki; Muroi, Masayuki*; Mukai, Satoru*

JAEA-Technology 2015-015, 96 Pages, 2015/07

JAEA-Technology-2015-015.pdf:20.34MB

In order to dispose of the radioactive waste which generates from post-irradiation examination (PIE) facilities, the common evaluation method of radioactivity in wastes from PIE should be established by the actual data such as radioactivity values and the theoretical calculation. In this study, the radioactivity concentrations of 17 nuclides (H-3, C-14, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-154, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-244) in combustible wastes stored in NUCLEAR DEVELOPMENT CORPORATION were measured from 3 samples and the radioactivity was calculated by ORIGEN-2 based on initial contents and operation record of the spent fuel. From the comparison of the obtained data by the radiological measurement with the calculated values, the subject to be solved for establishment of the radioactivity evaluation method for PIE was extracted.

JAEA Reports

Method and result for calculation of radioactivity concentration of radionuclide corresponding to dose criterion for near surface disposal of radioactive waste generated from research, medical, and industrial facilities

Okada, Shota; Kurosawa, Ryohei; Sakai, Akihiro; Nakata, Hisakazu; Amazawa, Hiroya

JAEA-Technology 2015-016, 44 Pages, 2015/07

JAEA-Technology-2015-016.pdf:5.8MB

In this report, we calculated radioactivity concentration of radionuclides potentially contained in low level radioactive waste (LLW) generated from research, medical, and industrial facilities corresponding to dose criterion (10 $$mu$$Sv/y) for near surface disposal. 220 kinds of nuclides whose half-life are more than 30 days were selected. Radioactivity concentrations corresponding to dose criterion of 40 nuclides among 220 ones were calculated by using the representative model because the concentrations of 40 nuclides had not been calculated yet. Skyshine dose from each of 19 nuclides, whose radioactivity concentration were invalid values that are larger than the specific radioactivity of nuclides, during operation of disposal facility was calculated. These radioactivity concentrations can be used as criteria of categorization of LLW between trench type and concrete vault type disposal and of preliminary selection of important nuclides of these disposals in the generic conditions.

JAEA Reports

Development on nuclear design model for detailed design of clean burn HTGR

Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo

JAEA-Technology 2015-017, 61 Pages, 2015/07

JAEA-Technology-2015-017.pdf:1.97MB

A development on nuclear design model for detailed design of Clean Burn HTGR had been performed. In the previous study, the fuel composition assumed in the study of Deep Burn proposed by GA in U.S. was employed. In the present study, that is estimated to reflect the situation of the nuclear fuel cycle in Japan. And, the evaluation method is refined from the viewpoint of traceability and a guarantee of quality. The deployment of control rod columns is investigated. Moreover, the Er loading is also investigated to obtain negative temperature coefficients in all range of operation. This model is developed for MVP code, which solves neutron transportation by Monte Carlo method. Validation of burn-up chain is also performed to adopt very high burn-up calculation. The core design is performed by COREBN code, which solves neutron diffusion equation by deterministic approach. Thus, the convertor which converts the cross section of MVP to COREBN is also developed.

JAEA Reports

Implementation design of the remote control equipment and device development facility

Facility Management Department

JAEA-Technology 2015-018, 218 Pages, 2015/08

JAEA-Technology-2015-018-01.pdf:42.32MB
JAEA-Technology-2015-018-02.pdf:50.72MB
JAEA-Technology-2015-018-03.pdf:69.47MB
JAEA-Technology-2015-018-04.pdf:54.62MB
JAEA-Technology-2015-018-05.pdf:54.57MB
JAEA-Technology-2015-018-06.pdf:27.62MB

In Nuclear Plant Decommissioning Safety Research Establishment of Sector of Fukushima Research and Development of JAEA, the remote control equipment and device development facility has been constructed for R&D of decommissioning of TEPCO'S Fukushima Daiichi Nuclear Power Station. This facility consists of the test building for various demonstration tests and the research management building for equipment of worker training and user's workroom. The demonstration test area for the technique to repair a water leakage at the PCV, development and demonstration test area for the remote controlled devices and the shareable area, are prepared in the test building, assuming that the two types of tests are performed in parallel. Furthermore, the accessory building, consists of researcher's rooms and workshops for development/demonstration tests of disaster response robots, is prepared in the test building. The research management building consists of worker training rooms, user's rooms, office rooms and multipurpose area. In the multipurpose room, it is also possible to hold international conferences. This report summarized the result of implementation design of the remote control equipment and device development facility. In addition, this facility started construction in September 2014.

JAEA Reports

Criticality safety evaluation for the direct disposal of used nuclear fuel; preparation of data for burnup credit evaluation (Contract research)

Yamamoto, Kento*; Akie, Hiroshi; Suyama, Kenya; Hosoyamada, Ryuji*

JAEA-Technology 2015-019, 110 Pages, 2015/10

JAEA-Technology-2015-019.pdf:3.67MB

In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. The recent development of higher-enrichment fuel has enhanced the benefit of the application of Burnup Credit. In the present study, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study.

JAEA Reports

Transportation of a large amount of sodium between two facilities

Imamura, Hiroaki; Hayakawa, Masato; Handa, Takuya*; Namiki, Katsuo*

JAEA-Technology 2015-020, 85 Pages, 2015/08

JAEA-Technology-2015-020.pdf:24.15MB

Sodium using for past experiments was maintained at the sodium technology development facility in Oarai Research and Development Center. But the oldest dump tank of the sodium was fabricated over forty years ago. It is necessary to keep it safer and more stable condition. Therefore transportation of the sodium from the facility to a new facility was planned not only to keep it safer and more stable condition but also to reuse it for other sodium experiments. Such a large amount of sodium has never been transported between two facilities in Japan. Pipe transportation between two facilities was selected by considering safety, operating efficiency and cost. The transportation equipment with over 200 meters long pipes was constructed and sodium transportation was operated. Total amount of the transported sodium was about 270 m$$^{3}$$ and the flow rate was 12m$$^{3}$$/h on average.

JAEA Reports

Irradiation test with silicon ingot for NTD-Si irradiation technology

Takemoto, Noriyuki; Romanova, N.*; Kimura, Nobuaki; Gizatulin, S.*; Saito, Takashi; Martyushov, A.*; Nakipov, D.*; Tsuchiya, Kunihiko; Chakrov, P.*

JAEA-Technology 2015-021, 32 Pages, 2015/08

JAEA-Technology-2015-021.pdf:3.15MB

Silicon semiconductor production by neutron transmutation doping (NTD) method using the JMTR has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency in order to expand the industry use. As a part of investigations, irradiation test with a silicon ingot was planned using WWR-K in Institute of Nuclear Physics, Republic of Kazakhstan. A device rotating the ingot made with the silicon was fabricated and was installed in the WWR-K for the irradiation test. And that, a preliminary irradiation test was carried out using neutron fluence monitors to evaluate the neutronic irradiation field. Based on the result, two silicon ingots were irradiated as scheduled, and the resistivity of each irradiated silicon ingot was measured to confirm the applicability of high-quality silicon semiconductor by the NTD method (NTD-Si) to its commercial production.

JAEA Reports

Measurement experiment of oxygen concentration in liquid lead-bismuth eutectic; Fabrication of oxygen sensor and measurement under static condition

Sugawara, Takanori; Yamaguchi, Kazushi

JAEA-Technology 2015-022, 21 Pages, 2015/08

JAEA-Technology-2015-022.pdf:9.18MB

The oxygen sensors to measure the oxygen concentration in liquid LBE (lead-bismuth eutectic) were fabricated for future use in LBE-cooled ADS (accelerator-driven system) or LBE test loops. Two types of oxygen sensors were fabricated and used for the measurement under the oxygen saturated condition. Through the measurement experiment, it was confirmed that the electromotive force (EMF) from Pt-type sensor was reliable under 350$$^{circ}$$C to 450$$^{circ}$$C LBE temperature. The Pt-type sensor will be the first candidate for the use in LBE test loops.

JAEA Reports

Conceptual design for relocation of the underground monitoring systems to ground surface

Toya, Naruhisa*; Ogawa, Ken*; Iwatsuki, Teruki; Onuki, Kenji

JAEA-Technology 2015-023, 35 Pages, 2015/09

JAEA-Technology-2015-023.pdf:9.44MB

One of the major subjects of the ongoing geoscientific research program is the Mizunami Underground Research Laboratory (MIU) Project in the Tono area, central Japan is accumulation of knowledge about a recovery of the geological environment during and after the facility closure. Then it's necessary to plan the observation system which can use after the backfill of research tunnels. The main purpose of this report is contribution to the detailed design for relocation of the underground monitoring systems to ground surface. We discussed the restriction and requirement for the underground monitoring systems which can use after the backfill. Furthermore, we made the conceptual design for relocation of the current underground monitoring systems to ground surface.

JAEA Reports

Environmental sampling from the hot cell in the Tokai Reprocessing Plant for safeguards

Yamamoto, Masahiko; Mori, Eito; Surugaya, Naoki

JAEA-Technology 2015-024, 19 Pages, 2015/09

JAEA-Technology-2015-024.pdf:19.11MB

Environmental Sampling from the hot cell in the Operating Testing Laboratory (OTL) of the Tokai Reprocessing Plant is implemented as an inspection of International Atomic Energy Agency (IAEA) for the first time. The specified sampling place requested from IAEA inspector is a device inside the hot cell. Since it is expected to be highly radioactive, the dose rates of samples and inside the hot cell are evaluated in advance. Also, the threshold dose rates of samples are determined. Subsequently, the sampling procedure describing radiation protection has been prepared. The environmental sampling from OTL hot cell is safely performed in accordance with the procedure and the radioactivity of sample is measured. The samples are categorized as Excepted Package according to the transport regulation and are transported to Safeguards Analytical Services of IAEA.

JAEA Reports

Performance confirmation operation of water environment control facility

Magome, Hirokatsu; Okada, Yuji; Tomita, Kenji; Iida, Kazuhiro; Ando, Hitoshi; Yonekawa, Akihisa; Ueda, Haruyasu; Hanawa, Hiroshi; Kanno, Masaru; Sakuta, Yoshiyuki

JAEA-Technology 2015-025, 100 Pages, 2015/09

JAEA-Technology-2015-025.pdf:78.32MB

In Japan Atomic Energy Agency, in order to solve the problem in the long-term operation of a light water reactor, preparation which does the irradiation experiment of light-water reactor fuel and material was advanced. JMTR stopped after the 165th operation cycle in August 2006, and is advancing renewal of the irradiation facility towards re-operation. The material irradiation test facility was installed from 2008 fiscal year to 2012 fiscal year in JMTR. The material irradiation test facility is used for IASCC study, and that consists of mainly three equipments. This report is described performance operating test of the water environmental control facilities for IASCC study carried out 2013 fiscal year.

JAEA Reports

Design and manufacture of Joyo upper core structure for replacement

Ota, Katsu; Ushiki, Hiroshi*; Maeda, Shigetaka; Kawahara, Hirotaka; Takamatsu, Misao; Kobayashi, Tetsuhiko; Kikuchi, Yuki; Tobita, Shigeharu; Nagai, Akinori

JAEA-Technology 2015-026, 180 Pages, 2015/11

JAEA-Technology-2015-026.pdf:79.87MB

In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of "MARICO-2" (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). The replacement of the UCS was conducted from May to December 2014. The design and manufacture of UCS was started from 2008, and the installation of UCS was completed successfully in November 21th 2014. The major results gained during the design and manufacture of UCS is as follows.

JAEA Reports

Information collection regarding geoscientific monitoring techniques during closure of underground facility in crystalline rock

Hosoya, Shinichi*; Yamashita, Tadashi*; Iwatsuki, Teruki; Saegusa, Hiromitsu; Onoe, Hironori; Ishibashi, Masayuki

JAEA-Technology 2015-027, 128 Pages, 2016/01

JAEA-Technology-2015-027.pdf:33.66MB

The study for development of drift backfilling technologies is one of the critical issues in the Mizunami Underground Research Laboratory (MIU) project, and its purposes are to develop closure methodology and technology, and long-term monitoring technology, and to evaluate resilience of geological environment. To achieve the purposes, previous information from the case example of underground facility constructed in crystalline rock in Europe has been collected. In particular, the boundary conditions for the closure, geological characteristics, technical specifications, and method of monitoring have been focused. The information on the international project regarding drift closure test and development of monitoring technologies has also been collected. In addition, interviews were conducted to specialists who have experiences involving planning, construction management, monitoring, and safety assessment for the closure. Based on the collected information, concept and point of attention, which are regarding drift closure testing, and planning, execution management and monitoring on the closure of MIU, have been specified.

JAEA Reports

Research and development on chemical reactors made of industrial structural materials and hydriodic acid concentration technique for thermochemical hydrogen production IS process

Kubo, Shinji; Iwatsuki, Jin; Takegami, Hiroaki; Kasahara, Seiji; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Onuki, Kaoru

JAEA-Technology 2015-028, 32 Pages, 2015/10

JAEA-Technology-2015-028.pdf:23.69MB

JAEA has been conducting a study on IS process for thermochemical hydrogen production in order to develop massive hydrogen production technology for hydrogen society. Integrity of the chemical reactors and concentration technology of hydrogen iodide in HIx solution were studied. In the former study, the chemical reactors were trial-fabricated using industrial materials. A test of 30 times of thermal cycle test under circulating condition of the Bunsen reaction solution showed integrity of the Bunsen reactor made of fluororesin lined steel. Also, 100 hours of reaction tests showed integrity of the sulfuric acid decomposer made of silicon carbide and of the hydrogen iodide decomposer made of Hastelloy C-276. In the latter study, concerning electro-electrodialysis using cation-exchange membrane, sulfuric acid in the anolyte had little influence on the concentration performance. These results suggest the purification system of HIx solution can be simplified. Based on the Nernst-Planck equation and the Smoluchowski equation, proton transport number, water permeance, and IR drop of the cation exchange membrane were formulated. The derived equations enable quantitative estimation for the performance indexes of Nafion membrane and, also, of ETFE-St membranes made by radiation-induced graft polymerization method.

JAEA Reports

Development of multi pass pumping module for thin disk laser material

Okada, Hajime; Maruyama, Momoko; Ochi, Yoshihiro; Nagashima, Keisuke

JAEA-Technology 2015-029, 29 Pages, 2015/11

JAEA-Technology-2015-029.pdf:11.21MB

We develop high repetitive and short pulse laser system using Yb doped laser ceramics for future laser application of generating high power tera-hertz wave for experimental proof of isotope separation and generating a Laser-Compton Scattering $$gamma$$-ray source for precise specific isotopes analysis. As these future laser applications needs very stable and high average power laser system, we develop thin disk laser system using Yb:YAG ceramics because of its large stored energy property and large thermal conductivity, which could avoid thermal induced effects and thermal fracture limitation due to large pumping power. A few hundred $$mu$$m thickness of thin disk laser material in laser amplifier needs multi pass optical components for enough absorption of pump power, the pumping profile and spot size on laser disk should be variable for required laser performance. We design multi pass pumping module for thin disk laser material which consists of commercial optical components and precise stages and also design doublet and triplet for various pumping profile and spot size, the laser oscillation power is evaluated with this pumping module.

JAEA Reports

Establishment of experimental system for $$^{99}$$Mo/$$^{99m}$$Tc production by neutron activation method

Ishida, Takuya; Shiina, Takayuki*; Ota, Akio*; Kimura, Akihiro; Nishikata, Kaori; Shibata, Akira; Tanase, Masakazu*; Kobayashi, Masaaki*; Sano, Tadafumi*; Fujihara, Yasuyuki*; et al.

JAEA-Technology 2015-030, 42 Pages, 2015/11

JAEA-Technology-2015-030.pdf:4.82MB

The research and development (R&D) on the production of $$^{99}$$Mo/$$^{99m}$$Tc by neutron activation method ((n, $$gamma$$) method) using JMTR has been carried out in the Neutron Irradiation and Testing Reactor Center. The specific radioactivity of $$^{99}$$Mo by (n, $$gamma$$) method is extremely low compared with that by fission method ((n,f) method), and as a result, the radioactive concentration of the obtained $$^{99m}$$Tc solution is also lowered. To solve the problem, we propose the solvent extraction with methyl ethyl ketone (MEK) for recovery of $$^{99m}$$Tc from $$^{99}$$Mo produced by (n, $$gamma$$) method. We have developed the $$^{99}$$Mo/$$^{99m}$$Tc separation/extraction/concentration devices and have carried out the performance tests for recovery of $$^{99m}$$Tc from $$^{99}$$Mo produced by (n, $$gamma$$) method. In this paper, in order to establish an experimental system for $$^{99}$$Mo/$$^{99m}$$Tc production, the R&D results of the system are summarized on the improvement of the devices for high-recovery rate of $$^{99m}$$Tc, on the dissolution of the pellets, which is the high-density molybdenum trioxide (MoO$$_{3}$$) pellets irradiated in Kyoto University Research Reactor (KUR), on the production of $$^{99m}$$Tc, and on the inspection of the recovered $$^{99m}$$Tc solutions.

JAEA Reports

Performance tests of radiation detectors for inspection of $$^{99}$$Mo/$$^{99m}$$Tc solution, 1

Suzuki, Yumi*; Nakano, Hiroko; Suzuki, Yoshitaka; Ishida, Takuya; Shibata, Akira; Kato, Yoshiaki; Kawamata, Kazuo; Tsuchiya, Kunihiko

JAEA-Technology 2015-031, 58 Pages, 2015/11

JAEA-Technology-2015-031.pdf:14.57MB

Technetium-99m ($$^{99m}$$Tc) is one of the most commonly used radioisotopes in the field of nuclear medicine. In the Japan Atomic Energy Agency (JAEA), the research and development (R&D) have been carried out for production of molybdenum-99 ($$^{99}$$Mo) by (n, $$gamma$$) method, a parent nuclide of $$^{99m}$$Tc, with the Japan Material Testing Reactor (JMTR). On the other hand, the new project as "Domestic Production of Medical Radioisotope (Technetium preparation) in Japan" was adopted in the Tsukuba International Strategic Zone on October, 2013 and the demonstration tests will be planned for the domestic production of $$^{99}$$Mo/$$^{99m}$$Tc with the JMTR. Thus, new facilities and analysis devices were equipped in the JMTR Hot Laboratory in 2014 as the part of this project. As the part of the analytical device equipment, the $$gamma$$-TLC analyzer and the radiation detector connected with the High Performance Liquid Chromatography (HPLC) were installed for quality inspection of the $$^{99}$$Mo/$$^{99m}$$Tc solution and the extracted $$^{99m}$$Tc solution in the JMTR Hot Laboratory. The performance tests of these devices such as detection sensitivity, resolution, linearity and selectivity of energy range were carried out with $$^{137}$$Cs and $$^{152}$$Eu as alternative radionuclides of $$^{99}$$Mo and $$^{99m}$$Tc, respectively. In the results, bright prospects were obtained concerning the quality inspection of the $$^{99}$$Mo/$$^{99m}$$Tc and $$^{99m}$$Tc solutions using these devices. This report describes the results of those performance tests.

JAEA Reports

Calculation of decay heat by new ORIGEN libraries for high temperature engineering test reactor

Simanullang, I. L.*; Honda, Yuki; Fukaya, Yuji; Goto, Minoru; Shimazaki, Yosuke; Fujimoto, Nozomu*; Takada, Shoji

JAEA-Technology 2015-032, 26 Pages, 2016/01

JAEA-Technology-2015-032.pdf:2.07MB

Decay heat of the High Temperature Engineering Test Reactor had been evaluated by the Shure Equation and/or ORIGEN code based on the LWR's data. However, to evaluate more accurately, a suitable method must be considered because of the differences neutron spectrums from the LWRs. Therefore, the decay heat and the generated nuclides for the neutron spectrums of the core with different graphite moderator amount were calculated by the ORIGEN2 code. As a result, it is clear that the calculated decay heats are similar value with LWRs for about one year after the reactor shutdown, and that the significant differences are observed on the longer period affected by the generated nuclides such as $$^{90}$$Y, $$^{134}$$Cs, $$^{144}$$Pr, $$^{106}$$Rh, $$^{241}$$Am etc. It is also clear that the dose is affected by $$^{241}$$Pu on the initial stage after the reactor shutdown.

JAEA Reports

Study on the applicability of the fiber-optic crack detection sensor as a safety technology (Joint research)

Hiraiwa, Kenichi*; Hirai, Kazuhide*; Sano, Tadashi*; Osawa, Hideaki; Sato, Toshinori; Aoyagi, Yoshiaki; Fujita, Tomoo; Aoyagi, Kazuhei; Inagaki, Daisuke*

JAEA-Technology 2015-033, 50 Pages, 2015/11

JAEA-Technology-2015-033.pdf:12.6MB

Japan Atomic Energy Agency (hereinafter referred to as JAEA) has been conducted a geoscientific research and development project at the Mizunami Underground Research Laboratory and the Horonobe Underground Research Laboratory in order to construct scientific and technological basis for geological disposal. As a collaborative research between JAEA and Tokyo Sokki Kenkyujo Co., Ltd. (hereinafter referred to as TML), we focused on the fiber-optic crack detection sensor developed by TML as a method to detect cracks in the support system that may affect the stability of rock cavern during the operation. To verify long-term safety performance of the sensor for decades, "Evaluation test of long-term durability of fiber-optic crack detection sensor and the support system" at the Mizunami Underground Research Laboratory and "Performance evaluation test of fiber-optic crack detection sensor for in-situ crack detection" at the Horonobe Underground Laboratory Research Laboratory were conducted. As the result, we understand that fiber-optic crack detection sensor is applicable measurement method to promptly detect the cracks in the support system.

JAEA Reports

Progress of construction work excavation of shafts and research galleries

Construction Department; Tono Geoscience Center, Sector of Decommissioning and Radioactive Waste Management; Horonobe Underground Research Center, Sector of Decommissioning and Radioactive Waste Management

JAEA-Technology 2015-034, 411 Pages, 2016/03

JAEA-Technology-2015-034.pdf:103.07MB

This report presents the results of shaft and gallery excavation performed focusing on crystalline rock (Mizunami) and sedimentary rock (Horonobe) from the point of view of construction technologies applied and the information obtained at respective construction stages, which is required for construction designing Facility construction in general, its goal is to build the facilities and the general and accrual designs are made based on the specific construction plan, while the construction of shafts and research galleries is being conducted based on the research plan. This construction is performed in the deep underground where significant uncertainties exit, for instance, it is difficult to obtain the precision information from preliminary investigation, construction work is inextricability liked to the stepwise research, and this very long-term construction period is likely to be receiving restrictions concerning environmental and social interests. Therefore, there are a number of conditions can not to be predicted at the initial design stage. Timely and appropriate actions will be taken to deal with these particular conditions, such as changing on design due to the revision of the research and construction plan while conducting excavation construction. In the series of construction activities: from input (i.e. construction conditions) to completion of the construction under the particular conditions, we summarize the experiences obtained at respective construction stages as the important information to transfer the technology to the similar construction in the future. This report describes the general consideration and summary of chapters at the beginning, and introduces the construction activities performed at each rock series.

JAEA Reports

Development of a cutting technique of core structural materials and fuel debris; Applicability test of the plasma jet cutting technique

Shoji, Tsugio; Fukui, Yasutaka; Ueda, Takiho

JAEA-Technology 2015-035, 70 Pages, 2016/01

JAEA-Technology-2015-035.pdf:8.07MB

The plasma jet cutting technology (Max output current is 250A) is developed for the dismantling of nuclear facilities in Oarai Research and Development Center. The plasma jet cutting technology is applicable to take out the debris. The plasma jet torch (Max output current is 600A) was produced for this application. This torch is available for the cutting of thick core internal materials in water. The ability of taking out debris and core internal material has been confirmed.

JAEA Reports

Current status of a decommissioning project in the Enrichment Engineering Facilities; Results in the first-half of the fiscal year of 2014

Matsumoto, Takashi; Morimoto, Yasuyuki; Takahashi, Nobuo; Takata, Masaharu; Yoshida, Hideaki; Nakashima, Shinichi; Ishimori, Yuu

JAEA-Technology 2015-036, 60 Pages, 2016/01

JAEA-Technology-2015-036.pdf:9.15MB

The Enrichment Engineering Facilities of the Ningyo-toge Environmental Engineering Center was constructed in order to establish the technical basis of the uranium enrichment plant in Japan. Uranium enrichment tests, using natural and reprocessed uranium, were carried out from 1979 to 1990 at two types of plants in the facilities. UF$$_{6}$$ handling equipment and Supplemental equipment in these plants are intended to be dismantled by 2019 in order to make places for future projects, for example, inventory investigation, precipitation treatment, etc. This report shows the basic plan of this decommissioning project and presents the current state of dismantling in the first-half of the fiscal year of 2014, with indicating its schedule, procedure, situation, results, and so on. The dismantled materials generated amounted to 37 mesh containers and 199 drums, and the secondary waste generated amounted to 271.4 kg.

JAEA Reports

Deuterium Critical Assembly (DCA) Decommissioning work in 2013

Morita, Kenji; Morimoto, Makoto; Hisada, Masaki; Fukui, Yasutaka

JAEA-Technology 2015-037, 28 Pages, 2016/01

JAEA-Technology-2015-037.pdf:8.44MB

Deuterium Critical Assembly (DCA) achieved first critically in 1969 and used for research and development program of Advanced Thermal Reactor. To achieved the aim of facility, DCA decommissioning work started in 2002. Decommissioning schedule consists of 4 stages. The third stage, which is the main work (To dismantle and remove reactor vessel and main equipment), was started in 2008 and will be finished at 2023. This report describes DCA decommissioning work and data (Ability of cutting tools and Man-hours) in 2013.

JAEA Reports

Development of dismantling technology for nuclear fuel facility; Discussion of dismantling method for Old Waste Treatment Facility for JOYO

Morita, Kenji; Morimoto, Makoto; Hisada, Masaki; Fukui, Yasutaka

JAEA-Technology 2015-038, 30 Pages, 2016/02

JAEA-Technology-2015-038.pdf:14.65MB

The Old Waste Treatment Facility for JOYO (Old JWTF) has been operated to treat radioactive liquid waste from the experimental fast reactor JOYO and post irradiation examination facilities. Operation of Old JWTF stopped in 1995, and dismantling & decontamination method has discussed. As a response to discussion results of remote and dismantling method in high dose environment on 2013, its concept examination was discussed on 2014. Results are follows. As a cutting tool for Old JWTF equipment, wire saw is selected from cutting ability (speed and thickness of objects). Discussed the component technology of wire saw remote operation system (handling, monitoring, collection method of secondary waste, else).

JAEA Reports

Study on engineering technologies in the Mizunami Underground Research Laboratory (FY 2014); Development of design and construction planning and countermeasure technologies (Contract research)

Kobayashi, Shinji*; Niimi, Katsuyuki*; Tsuji, Masakuni*; Yamada, Toshiko*; Aoyagi, Yoshiaki; Sato, Toshinori; Mikake, Shinichiro; Osawa, Hideaki

JAEA-Technology 2015-039, 170 Pages, 2016/02

JAEA-Technology-2015-039.pdf:37.73MB

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) plan consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies regarding restoration or reversal and mitigating of the excavation effect. To develop design and construction planning technologies, and countermeasure technology, the analysis of measured data during earthquake and seismic movement characteristics at deep underground, and the examination of grouting method were carried out. For the characteristics of earthquake ground motion, measurement data obtained by seismometers installed in the Mizunami Underground Laboratory were analyzed, and the comprehensive assessment of the relationship between the measurement data and the geological condition at each depth was performed. As for "Study on grouting method at deep underground ", post grouting was carried out and evaluated based on the Construction plan in FY2013. Furthermore, target of the future R&D was proposed.

JAEA Reports

Application of FORNAX-A

Aihara, Jun; Ueta, Shohei; Nishihara, Tetsuo

JAEA-Technology 2015-040, 32 Pages, 2016/02

JAEA-Technology-2015-040.pdf:0.83MB

Original FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation what calculations become possible with minor changed FORNAX-A.

JAEA Reports

Vacuum operation test for optical window protective shutters of JT-60SA diagnostic port-plugs

Kokusen, Shigeharu; Yoshida, Maiko; Tojo, Hiroshi

JAEA-Technology 2015-041, 26 Pages, 2016/02

JAEA-Technology-2015-041.pdf:8.03MB

Shutters will be installed in front of vacuum windows of some diagnostics in JT-60SA to avoid deposition of impurity onto the vacuum window during wall cleaning and plasma conditioning. Two types of shutter systems had been designed. One has a light shutter (0.5 kg) with rotary motion in vacuum. Another has a heavy shutter (3 kg) with vertical motion in vacuum. For the both types, malfunction due to increase of friction coefficients of the sliding parts in vacuum is concerned. In the present paper, Durability, abrasion and friction coefficient of the sliding parts has been investigated and the shutter designs have been validated. The light shutter with rotary motion successfully rotated required times (4000 times), and the design was validated. On the other hand, for the heavy shutter with vertical motion, the test could not be carried out due to the large friction of the sliding parts. This result suggests that we should change the material of the sliding parts to reduce the friction or change the design from the sliding structure to other ones such as a structure using pulleys and bearings. In the test of the heavy shutter with vertical motion, the friction coefficient increased from 1.3 to 4.5 in the vacuum chamber and from 0.4 to 2.5 in the air.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Development of cover gas recycling system with precise pressure control

Ushiki, Hiroshi*; Okuda, Eiji; Suzuki, Nobuhiro; Takamatsu, Misao; Nagai, Akinori

JAEA-Technology 2015-042, 37 Pages, 2016/02

JAEA-Technology-2015-042.pdf:16.51MB

The reactor vessel of a sodium-cooled fast reactor (SFR) is filled with sodium coolant and cover gas (argon gas). In case of a cover gas boundary open (ie., in-vessel repair), installation of a temporary cover gas boundary and controlling the cover gas pressure slightly positive are required to prevent the cover gas release and the contamination of impurities, and during upper core structure (UCS) replacement in the experimental SFR Joyo from March to December 2014, a vinyl bag was installed as a part of the temporary cover gas boundary. However, because it has inferior thermal resistance, supply a cooling gas too much was required to maintain proper temperature for two months. On the basis of this requirement, a cover gas recycling system with precise pressure control was developed and adopted for UCS replacement. The system has a good pressure controllability and recyclability. The successful results of this system contributed to the certain promotion of UCS replacement. In addition, the insights and the experience gathered in this development are expected to improve the in-vessel repair techniques in sodium-cooled fast reactors.

JAEA Reports

Development of evaluation procedure of vapor species transition behavior; Investigation of applicable measurement technology for estimation of chemical form and physical parameters, and validity verification

Takai, Toshihide; Sato, Isamu*; Yamashita, Shinichiro; Furukawa, Tomohiro

JAEA-Technology 2015-043, 56 Pages, 2016/02

JAEA-Technology-2015-043.pdf:23.14MB

Fundamental research on FP-chemistry for fission product release behaviors under severe accident was carried out for reinforcement of source term evaluation, and implementation of the 1F decommissioning R&D project. There were subjects to clarified (1) FP chemistry behavior between vapor species release and aerosol formation and (2) physical parameters which would be affect subsequent aerosol's chemical behavior, for improvement of FP transport model. Applicability of measuring/analyzing techniques presently used was studied for evaluating foregoing properties. And the validity was verified by trial measurements. In conclusion, Raman spectrometry and high temperature X-ray diffraction were hopeful to determine FP-chemical form against vapor/aerosol species and aerosol species, respectively. Combination use of cascade impactor and scanning type electron microscope with energy-dispersive X-ray spectrometry was hopeful to determine physical parameters of aerosol.

JAEA Reports

Study on release behavior of radioiodine from fuel solution under criticality accident condition

Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2015-044, 20 Pages, 2016/03

JAEA-Technology-2015-044.pdf:1.26MB

In order to estimate public dose under a criticality accident in fuel solution of a fuel reprocessing plant, release behavior of radioiodine from the fuel solution to atmosphere is very important. In this report, time evolution of $$^{133}$$I concentration in gas phase of TRACY core tank was measured until the concentration in the solution decreased. Furthermore, cumulative release ratio (CRR) and release rate (RR) from the solution to the atmosphere of radioiodine were evaluated by applying previously-reported evaluation model. As a result, for the case of short transient criticality, RR of $$^{133}$$I became maximum at 1 hour later from the ending and almost constant after 8 hour later. Furthermore, relationship of each elapsed time between total fission number and release rate of $$^{133}$$I could be derived. On the other hand, for the case of long criticality excursion, such as JCO criticality accident, the CRR and RR of radioiodine increased monotonously with time.

JAEA Reports

The Selection of the cutting technologies for dismantling the FUGEN reactor

Nakamura, Yasuyuki; Iwai, Hiroki; Mizui, Hiroyuki; Sano, Kazuya

JAEA-Technology 2015-045, 137 Pages, 2016/03

JAEA-Technology-2015-045.pdf:27.77MB

FUGEN is 9 m outer-diameter and 7m height, and characterized by its tube-cluster construction that contains 224 fuel channels arranging both the pressure and the calandria tubes coaxially in each channel. And the periphery part of the core has the laminated structure composed of up to 150 mm thickness of carbon steel for radiation shielding. The structure of the reactor, which is made of various materials such as stainless steel, carbon steel, zirconium alloy and aluminum. The reactor is planning to be dismantled under water in order to shield the radiation ray around the core and prevent airborne dust generated by the cutting, the temporary pool structure and the remote-operated dismantling machines needs to be installed on the top of reactor. In consideration of above the structure of Fugen reactor, the cutting method was selected for dismantling the reactor core in order to shorten the dismantling term and reduce the secondary waste.

JAEA Reports

The Development of the basic dismantling procedure of the reactor of FUGEN

Iwai, Hiroki; Nakamura, Yasuyuki; Mizui, Hiroyuki; Sano, Kazuya

JAEA-Technology 2015-046, 110 Pages, 2016/03

JAEA-Technology-2015-046.pdf:85.22MB

Advanced Thermal Reactor (ATR) FUGEN is a proto-type heavy water moderated, boiling light water cooled, pressure tube-type reactor with the thermal power of 557 MW and the electrical power of 165 MW. The reactor of FUGEN is classified into the core region and the shielding region. The core region is highly activated owing to the long term operation, and characterized by its tube-cluster construction that contains 224 fuel channels arranging both the pressure and the calandria tubes coaxially in each channel closely. And the shielding region surrounding the core region has the laminated structure composed of up to 150 mm thickness of carbon steel. The reactor is planning to be dismantled under water remotely in order to shield the radiation around the core and prevent airborne dust generated by the cutting, and firing of zirconium material. This paper reports on the result of development of the basic dismantling procedure of the reactor of FUGEN.

JAEA Reports

Applicability test of plasma cutting technology for dismantling of the core internals of Fukushima Daiichi NPS

Tezuka, Masashi; Nakamura, Yasuyuki; Iwai, Hiroki; Sano, Kazuya

JAEA-Technology 2015-047, 114 Pages, 2016/03

JAEA-Technology-2015-047.pdf:46.17MB

It was reported that Fukushima Daiichi Nuclear Power Plant had been lost the function of cooling the reactor by the Tohoku Earthquake. It is assumed that the original shapes of the internal core are not kept and the inside of the reactor makes so narrow in the space, however the fuel debris and the molten internal core will have to be removed for the decommissioning of 1F. We concerned the suppression of dross by optimization of cutting conditions, in using some moderated test pieces. And we can improve the cutting capability by heating the objects in advance. Moreover, it's possible that plasma arc cutting can cut off the mixed material the fuel debris and the molten internal core by using the cooperation cutting technique both the plasma arc and the plasma jet cutting. From these results, we have got the prospect that plasma cutting method can apply the removal of the fuel debris and the molten internal core.

JAEA Reports

Development of active control technology for impurities in coolant helium using High Temperature Engineering Test Reactor (HTTR)

Hamamoto, Shimpei; Nemoto, Takahiro; Sekita, Kenji; Saito, Kenji

JAEA-Technology 2015-048, 62 Pages, 2016/03

JAEA-Technology-2015-048.pdf:2.58MB

The decarburization may take place depending on the chemical impurity composition in helium gas used as the primary coolant in High-Temperature Gas-cooled Reactors, and will significantly reduce the strength of the alloy. The ability to remove impurities by a helium purification system was designed according to the predicted generation rate of impurities so as to make the coolant become the carburizing atmosphere. It has been confirmed that the coolant becomes the carburizing atmosphere during the operation period of the High Temperature engineering Test Reactor (HTTR). However, it is necessary to consider changes of generation rates of impurities since lifetime of commercial reactor is longer than the life of the HTTR. To avoid the influence of the change of generation rate, the control of removal efficiency of impurity in the helium purification system was considered in this study. To reform the decarburizing into the carburizing atmosphere, it is effective to increase the H$$_{2}$$ and CO concentration in the coolant helium. By controlling the efficiency of the Cooper Oxide Trap (CuOT), it is possible to increase the H$$_{2}$$ and CO concentrations. Therefore, an experiment was carried out by injecting the gas mixture of H$$_{2}$$ and CO into the existing purification system of HTTR to investigate the dependencies of temperature and impurity concentration on the removal efficiency of CuOT. The experimental results are described as the following, (1) By adjusting the temperature of helium at the CuOT within a range from 110$$^{circ}$$C to 50$$^{circ}$$C, it is possible to reduce the removal efficiency of H$$_{2}$$ sufficiently. (2) Temperature change of helium gas in the CuOT is sufficiently reduced by the cooler located at the downstream of the CuOT, which does not affect the primary cooling system of HTTR. As the results, the applicability of removal efficiency control of CuOT was verified to improve the decarburizing atmosphere for the actual HTGR system.

JAEA Reports

Performance test of ex-core high temperature and high pressure water loop test equipment (Contract research)

Nakano, Hiroko; Uehara, Toshiaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nakamura, Jinichi; Matsui, Yoshinori; Tsuchiya, Kunihiko

JAEA-Technology 2015-049, 61 Pages, 2016/03

JAEA-Technology-2015-049.pdf:14.7MB

In Japan Atomic Energy Agency, we started a research and development so as to monitor the Nuclear Plant Facilities situations during a severe accident, such as a radiation-resistant monitoring camera under a severe accident, a radiation resistant in-water transmission system for conveying the information in-core and a heat-resistant signal cable. As part of advance in a heat-resistant signal cable, we maintained to ex-core high-temperature and pressure water loop test equipment which can be simulated conditions of BWRs and PWRs for evaluation reliability and property of construction sheath materials. This equipment consists of Autoclave, water conditioning tank, water pump, high-pressure metering pump, preheater, heat exchanger and pure water purification equipment. This report describes the basic design and the results of performance tests of construction machinery and tools of ex-core high-temperature and pressure water loop test equipment.

JAEA Reports

A Study on optimization of dismantling scenarios in Fugen decommissioning project; Preliminary evaluation of dismantling scenario of the heat exchangers, etc.

Koda, Yuya; Tezuka, Masashi; Yanagihara, Satoshi*

JAEA-Technology 2015-050, 74 Pages, 2016/03

JAEA-Technology-2015-050.pdf:3.43MB

The implementation of the decommissioning work is accompanied by long-term period and considerable expense, so it is important that we make the most optimized work scenario in consideration of safety or the work and effectiveness. For this reason, we are studying selection method of the optimal work scenarios as a management index of the manpower and dose etc., in dismantling work for Fugen. In this report, results of a study shows the method of selecting the best scenarios for the heat exchangers of the reactor coolant purification system by evaluating execution multiple work scenarios, as well as evaluating the manpower and dose, etc., moreover by setting the importance of each evaluation item.

JAEA Reports

Fabrication and test results of testing equipment for remote-handling of MA fuel, 1; Testing equipment for fuel cooling

Nishihara, Kenji; Tazawa, Yujiro; Inoue, Akira; Sugawara, Takanori; Tsujimoto, Kazufumi; Sasa, Toshinobu; Obayashi, Hironari; Yamaguchi, Kazushi; Kikuchi, Masashi*

JAEA-Technology 2015-051, 47 Pages, 2016/03

JAEA-Technology-2015-051.pdf:3.6MB

This report summarizes fabrication and test results of a testing equipment for fuel cooling that is a component of the testing equipment for remote-handling of highly-radioactive MA fuels in the transmutation physics experimental facility (TEF-P) planned in the J-PARC. Evaluation formula of pressure drop and temperature increase used in the design of TEF-P was validated by the test, and, feasibility of cooling concept was confirmed.

JAEA Reports

Fabrication and test results of testing equipment for remote-handling of MA fuel, 2; Evaluation of heat transfer parameter crossing rectangular lattice matrix

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Inoue, Akira; Tsujimoto, Kazufumi

JAEA-Technology 2015-052, 34 Pages, 2016/03

JAEA-Technology-2015-052.pdf:5.02MB

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. So it is necessary to verify reliability and accuracy of heat transfer effect used in this area. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.

JAEA Reports

Work and safety managements for on-site installation, commissioning, tests by EU of quench protection circuits for JT-60SA

Yamauchi, Kunihito; Okano, Jun; Shimada, Katsuhiro; Omori, Yoshikazu; Terakado, Tsunehisa; Matsukawa, Makoto; Koide, Yoshihiko; Kobayashi, Kazuhiro; Ikeda, Yoshitaka; Fukumoto, Masahiro; et al.

JAEA-Technology 2015-053, 36 Pages, 2016/03

JAEA-Technology-2015-053.pdf:8.33MB

The superconducting Satellite Tokamak machine "JT-60SA" under construction in Naka Fusion Institute is an international collaborative project between Japan (JA) and Europe (EU). The contributions for this project are based on the supply of components, and thus European manufacturer shall conduct the installation, commissioning and tests on Naka site. This means that Japan Atomic Energy Agency (JAEA) had a quite difficult issue to manage the works by European workers and their safety although there is no direct contract. This report describes the approaches for the work and safety managements, which were agreed with EU after the tough negotiation, and then the completed on-site works for Quench Protection Circuits (QPC) as the first experience for EU in JT-60SA project. With the help of these approaches by JAEA, the EU works for QPC were successfully completed with no accident, and a great achievement was made for both EU and JA.

JAEA Reports

Preparation of Zr-91 stable isotope standard solution for determination of Zr-93 by mass spectrometry

Konda, Miki; Asai, Shiho; Hanzawa, Yukiko; Magara, Masaaki

JAEA-Technology 2015-054, 22 Pages, 2016/03

JAEA-Technology-2015-054.pdf:3.44MB

Isotope dilution mass spectrometry (IDMS) with ICP-MS is reliable method for determination of Zr-93, which is one of the long-lived fission products found in spent nuclear fuel and high-level radioactive wastes. In order to use an isotope standard solution of zirconium as the spike for IDMS, dissolving a commercially available solid isotope standard is indispensable. Prior to the dissolution of the Zr-91 isotope standard, solubility of metal zirconium in a mixture of HNO$$_3$$ and HF was evaluated using zirconium metal chips. Then, 2 mg of the Zr-91 isotope standard was dissolved with 0.2 mL of 1 M HNO$$_3$$-3 v/v% HF mixed solution, followed by adjusting the concentration of Zr-91 to approximately 1,000 $$mu$$g/g. IDMS, in which a natural isotopic abundance standard solution of zirconium was used as the spike, was employed for the determination of the concentration of Zr-91 in the prepared Zr-91 isotope standard solution. The concentration of Zr-91 in the prepared Zr-91 isotope standard solution was (9.6$$pm$$1.0) $$times$$ 10$$^2$$ $$mu$$g/g, which is in good agreement with the predicted concentration. This indicates that the Zr-91 metal isotope standard was completely dissolved with sufficient chemical stability. Additionally, no impurities were detected in the prepared Zr-91 isotope standard solution. These positive results denote that the Zr-91 isotope standard solution with the preferable quality for IDMS of Zr-93 can be obtained by the proposed dissolution procedures.

JAEA Reports

Applicability test of abrasive water jet cutting technology for dismantling of the core internals of Fukushima Daiichi NPS

Nakamura, Yasuyuki; Iwai, Hiroki; Tezuka, Masashi; Sano, Kazuya

JAEA-Technology 2015-055, 89 Pages, 2016/03

JAEA-Technology-2015-055.pdf:17.54MB

It was reported that Fukushima Daiichi Nuclear Power Station (1F) had lost the cooling function of the reactor by the Tohoku Earthquake. It is assumed that the core internals became narrow and complicated debris structure mixed with the molten fuel. In consideration of the above situations, the AWJ cutting method, which has features of the long work distance and little heat effect for a material, has been developed for the removal of the molten core internals through cutting tests for 3 years since FY 2012. And it was confirmed that AWJ cutting method is useful for the removal of the core internals etc. The results in FY 2012 were reported in "R&D of the fuel debris removal technologies by abrasive water jet cutting technology (JAEA-Technology 2013-041)" and this report summarizes the results of FY 2012, 2013 and 2014 in this report. It was confirmed the possibility to apply the removal work of the fuel debris and the core internals.

JAEA Reports

Replacement of power supply for reactor control rod magnet of JRR-3

Kurumada, Osamu; Ikekame, Yoshinori; Ouchi, Satoshi; Sato, Masayuki; Kamiishi, Eigo; Wada, Shigeru

JAEA-Technology 2015-056, 35 Pages, 2016/03

JAEA-Technology-2015-056.pdf:29.49MB

The power supply for reactor control rod magnet of JRR-3 has been utilized for generating electromagnetic power of control rod coil and that was using more than 25 years. The power supply was required for provide to stabilize DC current. Therefore, we adopted series regulator method. Although, the power supply generate a high heat. Then, we decided to create switching regulator method in order to improve the aging and heat generation of the series regulator method. This paper reports the replacement of switching regulator method.

JAEA Reports

Development of BDI behavior evaluation method in the fast reactor fuel assembly; Improvement of out-of-pile bundle compression test technology

Higashiuchi, Atsushi; Ishimi, Akihiro; Katsuyama, Kozo; Uwaba, Tomoyuki; Ichikawa, Shoichi

JAEA-Technology 2015-057, 72 Pages, 2016/03

JAEA-Technology-2015-057.pdf:36.91MB

Bundle-duct interaction (BDI) in fast reactors (FRs) is one of the limiting factors for burnup. To study the high performance fuel for FR fuel, it is important to establish the method to predict accurately the BDI behavior for the fuel assembly of large-diameter fuel pins. Therefore, it was adopted a new method that the bundle compression test apparatus is placed outside the cell, the bundle specimen is put in the airtight container for contamination prevention, and the bundle specimen is carried in the cell for internal observation by X-ray CT examination apparatus. From the result of this test, it was confirmed that the new method of out-of-pile bundle compression test is carried out as it was before. The results of this test are available to study integrity assessment of fast reactor fuel, validation of the BDI analysis code and substantiation of the safety design guidelines of fast reactor. In addition, it is possible to reflect in the BDI behavior evaluation for "ASTRID".

JAEA Reports

Development of field emission SEM to observe high radioactive irradiated fuels

Isozaki, Misaki; Sasaki, Shinji; Maeda, Koji; Katsuyama, Kozo

JAEA-Technology 2015-058, 28 Pages, 2016/03

JAEA-Technology-2015-058.pdf:23.51MB

During irradiation in the fast reactor "JOYO", the changes of fuel structures with the formation of central void occur in the uranium-plutonium mixed oxide fuels (MOX fuels) because of radial temperature gradient. The changes of element (U, Pu, and so on) distributions along radial direction proceed from these changes. Therefore, it is important to study the changes of fuel structures of the minute area in fuel pellet and the changes of element distribution behavior for development of fast reactor fuels. In order to make detailed observations of microstructure and elemental analyses of fuel samples, a field emission scanning electron microscope (FE-SEM) equipped with a wavelength-dispersive X-ray spectrometer (WDS) and an energy-dispersive X-ray spectrometer (EDS) were installed in Fuel Monitoring Facility (FMF). The samples of this FE-SEM are very high radioactivity because the samples contain the nuclear fuel elements (U, Pu, etc.), the fission products (Cs, Rh, etc.) and activation product (Co, Mn etc.). Owing to this, it is necessary to prevent leakage of radioactive materials (particularly, U, Pu is need tight accountancy in law) and to protect operators from radiation. In this installation of FE-SEM, it is selected JSM-7001F (made by JEOL) for base model. The notable modified points were as follows. (1) To protect operators from radiation, lead shields was installed around FE-SEM. (2) To prevent leakage of radioactive materials, the instrument was attached to a remote control air-tight sample transfer unit between a shielded hot cell and the FE-SEM and the instrument was fixing rigid structure without vibration damper. (3) The design and manufacture the lead shields with consideration of instrument maintainability. This paper was described the summary of FE-SEM, the notable modified points, the ways of FE-SEM installation, the result of performance test.

JAEA Reports

Corroborative tests for Oarai Waste Reduction Treatment Facility using the in-can type high frequency induction heating method

Sakauchi, Hitoshi; Sato, Isamu*; Donomae, Yasushi; Kitamura, Ryoichi

JAEA-Technology 2015-059, 352 Pages, 2016/03

JAEA-Technology-2015-059.pdf:51.53MB

OWTF (Oarai Waste Reduction Treatment Facility) is constructed for volume reduction processing and stabilization treatment of $$alpha$$ solid waste, which was generated from hot facilities in Oarai Research and Develop Center of Japan Atomic Energy Agency, using in-can type high frequency induction heating by remote control. This report describes corroborative tests, in which incinerating and melting performance for OWTF is confirmed with a full-scale testing furnace. We have been carrying out the tests of incinerating and melting treatment with some kinds of simulated wastes, such as enclosure form of radioactive wastes, material and articles.

JAEA Reports

Study on decreasing trend of air dose rates observed in Fukushima Prefecture and neighboring prefectures

Ando, Masaki; Kanno, Takashi; Saito, Kimiaki

JAEA-Technology 2015-060, 40 Pages, 2016/03

JAEA-Technology-2015-060.pdf:50.1MB

The ratios of air dose rate averaged in prefecture-wise measured by car borne surveys in wide area has been performed by with the use of the KURAMA and KURAMA-II systems were investigated. The changes in air dose rate from the first (June 2011) car borne survey to the fourth (September 2012) and seventh (November 2013) car borne surveys in Fukushima Prefecture was similar to those in Tochigi Prefecture, and the ratio in Miyagi Prefecture showed quicker decay than those in the other prefectures in the groups of less than 0.5$$mu$$Sv/h. Distribution maps of the ratios of air dose rate obtained in 100m mesh wise in Fukushima, Tochigi and Miyagi prefectures showed that the ratios (i.e. decreasing rates) were depending on the area or road. Further, decreasing in the air dose rate was quicker than the physical decay in the big cities where population is big and the traffic is huge.

JAEA Reports

Development of ITER CODAC compatible gyrotron local control system and its operation

Oshima, Katsumi; Oda, Yasuhisa; Takahashi, Koji; Terakado, Masayuki; Ikeda, Ryosuke; Hayashi, Kazuo*; Moriyama, Shinichi; Kajiwara, Ken; Sakamoto, Keishi

JAEA-Technology 2015-061, 65 Pages, 2016/03

JAEA-Technology-2015-061.pdf:24.28MB

In JAEA, an ITER relevant control system for ITER gyrotron was developed according to Plant Control Design Handbook. This control system was developed based on ITER CODAC Core System and implemented state machine control of gyrotron operation system, sequential timing control of gyrotron oscillation startup, and data acquisition. The operation of ITER 170 GHz gyrotron was demonstrated with ITER relevant power supply configuration. This system is utilized for gyrotron operation test for ITER procurement. This report describes the architecture of gyrotron operation system, its basic and detailed design, and recent operation results.

JAEA Reports

Impact assessment of the forest fires on Oarai Research and Development Center Waste Treatment Facility

Shimomura, Yusuke; Hanari, Akira*; Sato, Isamu*; Kitamura, Ryoichi

JAEA-Technology 2015-062, 47 Pages, 2016/03

JAEA-Technology-2015-062.pdf:1.85MB

In response to new standards for regulating waste management facilities, it was carried out impact assessment of forest fires on the waste management facilities existed in Oarai Research and Development Center of Japan Atomic Energy Agency. At first, a fire spread scenario of forest fires was assumed. The intensity of forest fires was evaluated from field surveys, forest fire evaluation models and so on. As models of forest fire intensity evaluation, Rothermel Model and Canadian Forest Fire Behavior Prediction (FBP) System were used. Impact assessment of radiant heat to the facilities was carried out, and temperature change of outer walls for the assumed forest fires was estimated. The outer wall temperature of facilities was estimated around 160$$^{circ}$$C at the maximum, it was revealed that it doesn't reach allowable temperature limit. Consequently, it doesn't influence the strength of concrete. In addition, a probability of fire breach was estimated to be about 20%. This report illustrates an example of evaluation of forest fires for the new regulatory standards through impact assessment of the forest fires on the waste management facilities.

JAEA Reports

Environmental remediation for the upstream of Yotsugi Mill Tailings Pond, Ningyo-toge Uranium Mine

Saito, Hiroshi; Sato, Yasushi*; Sakamoto, Atsushi*; Torikai, Kazuyoshi; Fukushima, Shigeru; Sakao, Ryota; Taki, Tomihiro

JAEA-Technology 2015-063, 119 Pages, 2016/03

JAEA-Technology-2015-063.pdf:86.2MB

Ningyo-toge Environmental Engineering Center has been conducting environmental remediation of the Ningyo-toge Uranium Mine, after decades of mine-related activities were terminated. Its purposes are to take measures to ensure safety and radiation protection from the exposure pathways to humans in future, and to prevent the occurrence of mining pollution. As part of the remediation, upstream part of the Yotsugi Mill Tailings Pond, the highest prioritized facility, has been remediated to fiscal year 2012. Multi-layered capping has been constructed using natural material, after specifications and whole procedure being examined in terms of long-term stability, radiation protection, economics, etc. Monitoring has been carried out to confirm the effectiveness of the capping, in terms of settlement, dose and radon exhalation rates, etc. Monitoring of drainage volume of penetrated rainwater is planned. Accumulated data will be examined and used for remediation of downstream part of the Pond.

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