Safety measures in the melting facilities of The Advanced Volume Reduction Facilities; Document collection of discussion meetings related to melting facilities

Iketani, Shotaro; Yokobori, Tomohiko; Ishikawa, Joji; Yasuhara, Toshiyuki*; Kozawa, Toshiyuki*; Takaizumi, Hirohide*; Momma, Takeshi*; Kurosawa, Shingo*; Iseda, Hirokatsu; Kishimoto, Katsumi; et al.

JAEA-Review 2018-016, 46 Pages, 2018/12

JAEA-Review-2018-016.pdf:12.79MB

Japan Atomic Energy Agency (JAEA) adopts melting process for the waste processing and packaging method of radioactive miscellaneous solid waste in NSRI because melting process is effective in radioactivity evaluation, volume reduction, and stabilization treatment. Metal melting processing facilities, Incinerator, and Nonmetal melting processing facilities (hereinafter referred to as melting process facilities) have taken lots of safety measures, including measures for preventing the recurrence of the fire accidents. To exchange opinions and discuss the validity of these measures and so on with internal personnel and external experts, "Discussions on Melting Process Facilities" was held. As a document collection, this paper summarizes presentation materials of discussion meetings. Presentation materials describe "Outline of AVRF", "Safety measures in the melting facilities in WVRF", "Operation management of the melting facilities in WVRF", "Comparison of the past accident cases between facilities in and outside Japan and WVRF", and "Introduction of past accident cases and safety measures in other facilities from each committee".

Pretreatment works for disposal of radioactive wastes produced by research activities, 1

Ishihara, Keisuke; Yokota, Akira; Kanazawa, Shingo; Iketani, Shotaro; Sudo, Tomoyuki; Myodo, Masato; Irie, Hirobumi; Kato, Mitsugu; Iseda, Hirokatsu; Kishimoto, Katsumi; et al.

JAEA-Technology 2016-024, 108 Pages, 2016/12

JAEA-Technology-2016-024.pdf:29.74MB

Radioactive isotope, nuclear fuel material and radiation generators are utilized in research institutes, universities, hospitals, private enterprises, etc. As a result, various low-level radioactive wastes (hereinafter referred to as non-nuclear radioactive wastes) are produced. Disposal site for non-nuclear radioactive wastes have not been settled yet and those wastes are stored in storage facilities of each operator for a long period. The Advanced Volume Reduction Facilities (AVRF) are built to produce waste packages so that they satisfy requirements for shallow underground disposal. In the AVRF, low-level beta-gamma solid radioactive wastes produced in the Nuclear Science Research Institute are mainly treated. To produce waste packages meeting requirements for disposal safely and efficiently, it is necessary to cut large radioactive wastes into pieces of suitable size and segregate those depending on their types of material. This report summarizes activities of pretreatment to dispose of non-nuclear radioactive wastes in the AVRF.

Validation of decay heat evaluation method based on FPGS cord for fast reactor spent MOX fuels

Usami, Shin; Kishimoto, Yasufumi; Taninaka, Hiroshi; Maeda, Shigetaka

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3263 - 3274, 2016/05

The present paper describes the validation of the new decay heat evaluation method using FPGS90 code with both the updated nuclear data library and the rational extent of uncertainty, by comparing the results of the decay heat measurement of the spent fuel subassemblies in Joyo MK-II core and by comparing with the calculation results of ORIGEN2.2 code. The calculated values of decay heat (C) by FPGS90 based on the JENDL-4.0 library were coincident with the measured ones (E) within the calculation uncertainties, and the C/E ranged from 1.01 to 0.93. FPGS90 evaluated the decay heat almost 3% larger than ORIGEN2.2, and it improved the C/E in comparison with the ORIGEN2.2 code. Furthermore, The C/E by FPGS90 based on the JENDL-4.0 library was improved than that based on the JENDL-3.2 library, and the contribution of the revision of reaction cross section library to the improvement was dominant rather than that of the decay data and fission yield data libraries.

Effects of ultra-intense laser driven proton beam on the hydriding property of palladium

Abe, Hiroshi; Orimo, Satoshi; Kishimoto, Masahiko*; Aone, Shigeo*; Uchida, Hirohisa*; Daido, Hiroyuki; Oshima, Takeshi

Nuclear Instruments and Methods in Physics Research B, 307, p.218 - 220, 2013/07

https://doi.org/10.1016/j.nimb.2013.01.024

We investigated the structure changes of a hydrogen storage alloy by ion irradiations, and its absorption property in order to obtain basic data and to elucidate relevant mechanisms of hydrogen absorption by the influence of the irradiation. In previous studies, the induction of vacancies in a hydrogen absorption alloy was found to be effective to increase in the hydrogen absorption rate. As well known, the rate of hydrogen absorption strongly depends upon the surface state of a hydrogen storage alloy because the dissociation of hydrogen molecules or water molecules needs electron change with the surface in the H gas or electrochemical reaction process. In this study, ion irradiations were made at a room temperature using the laser driven proton beam method, at Kansai Photon Science Institute, Japan Atomic Energy Agency. The beam treatment has several unique properties such as short pulse duration, high peak current, low transverse emittance, and wide energy range from KeV to MeV. The irradiation was used to modify the alloy surface using this equipment. From obtained results, the initial hydrogen absorption rate was found improved by the laser driven proton beam rather more effectively than a mono-energetic proton beam. Discussion is made on the correlation among proton irradiation (laser driven proton or mono-energetic proton) and the initial hydrogen absorption rate of the alloy. We argue about the usefulness of an energy spread beam.

Basic study on surface chemical combination between beryllium metal and hydrogen isotope gas, 2

Ito, Masayasu; Kitagishi, Shigeru; Hanawa, Yoshio; Tsuchiya, Kunihiko; Hatano, Yuji*; Matsuyama, Masao*; Nagasaka, Takuya*; Hishinuma, Yoshimitsu*

Annual Report of National Institute for Fusion Science; April 2011 - March 2012, P. 535, 2012/12

http://www.nifs.ac.jp/report/annrep12/pdf/535.pdf

Beryllium has been utilized as a moderator and/or reflector in a number of material testing reactors. Beryllium is also supposed to be widely used in fusion reactors as neutron multiplier and protective walls of plasma facing components. It is important to perform the characterization of the different grade beryllium such as the productivity, mechanical and chemical properties and the interaction under water and/or gas environment. In this study, three kinds of beryllium (S-200F, S-65H, I-220H) were prepared, and corrosion test and surface analysis of these beryllium samples were carried out for life time expansion under pure water. As a result, the surface change of each Be sample was observed by the corrosion test and influenced by the content of BeO and the grain size.

Performance evaluation of friction welded joint of zircaloy-2 to type 316L stainless steel

Kitagishi, Shigeru; Saito, Takashi; Kikuchi, Taiji; Endo, Yasuichi; Tsuchiya, Kunihiko

Nihon Kikai Gakkai Rombunshu, A, 78(788), p.564 - 570, 2012/04

https://doi.org/10.1299/kikaia.78.564

The development of the friction welded joint between Zircaloy-2 (Zry-2) and Type 316L stainless steel (SUS316L) has been carried out for the fabrication of fuel irradiation capsules in Japan Materials Testing Reactor (JMTR). The welding condition was determined from the fabrication test results of total loss and tensile strength. The joint interface was characterized by metallographic observation, elemental analysis and micro-hardness test. It was found that heat-affected zones at the joint interface were different between the center and outside positions. The tensile strength of the joint was equal to that of base Zry-2 metals at 300C. And, the joint performance was good from the results of helium leak test and hydraulic pressure burst test. From the results of these tests, it was obvious that the Zry-2/SUS316L friction welded joints were promised to use for the structural material of irradiation capsules for fuel irradiation tests.

Conceptual design study of JSFR, 2; Reactor system

Eto, Masao*; Kamishima, Yoshio*; Okamura, Shigeki*; Watanabe, Osamu*; Oyama, Kazuhiro*; Negishi, Kazuo; Kotake, Shoji*; Sakamoto, Yoshihiko; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

In the JSFR design, the diameter of the Reactor Vessel (RV) shall be minimized and the reactor internal structures shall be simplified for reduction in construction cost. The reduction in the RV diameter is achieved by adopting an advanced refueling system and the hot RV with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the RV is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system.

Basic study on surface chemical combination between beryllium metal and hydrogen isotope gas

Tsuchiya, Kunihiko; Kitagishi, Shigeru; Ito, Masayasu; Hanawa, Yoshio; Hatano, Yuji*; Matsuyama, Masao*; Nagasaka, Takuya*; Hishinuma, Yoshimitsu*

Annual Report of National Institute for Fusion Science; April 2010 - March 2011, P. 545, 2011/11

http://www.nifs.ac.jp/report/annrep11/index.html

no abstracts in English

Conceptual design study for the demonstration reactor of JSFR, 4; Structural design of reactor vessel

Kawasaki, Nobuchika; Okamura, Shigeki*; Sawa, Naoki*; Sakamoto, Yoshihiko; Negishi, Kazuo

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10

Japan Sodium-Cooled Fast Reactor adopts an compacted hot reactor vessel concept. From the point of structural designs to ensure both seismic design and elevated temperature design is important. In this study, based on a common conservative seismic loading condition considered with the Niigata-ken Chuetsu-oki Earthquake, seismic evaluations were carried out, the thicknesses of reactor vessels of 750 MWe and 500 MWe output plants were determined. For both plants 50 mm was selected as the thickness, and ensured buckling evaluation margins were more than 2.4. From the point of seismic design, the difference of plant output was negligible. With the condition of 50 mm thickness of reactor vessel, thermal integrities were evaluated. For three plant start-up conditions which were 2.2, 3.2, and 4.3 days, thermal ratcheting and creep-fatigue damage were evaluated. As a result plant start-up period needed more than 3.2 days for both 750 MWe and 500 MWe output plants. Caused thermal stress were the nearly same for both plants, therefore from the point of thermal design, the difference of plant output was negligible.

A Study on the technology for reducing cement-type materials used for tunnel supports at high-level radioactive waste disposal sites (Joint research)

Hayashi, Katsuhiko; Noguchi, Akira; Kishi, Hirokazu; Kabayashi, Yasushi*; Nakama, Shigeo; Fujita, Tomoo; Naito, Morimasa; Tada, Hiroyuki*; Kumasaka, Hiroo*; Goke, Mitsuo*; et al.

JAEA-Research 2010-057, 101 Pages, 2011/03

JAEA-Research-2010-057.pdf:7.47MB

Cement-type materials that are used for supports or grouting at high-level radioactive waste disposal facilities leach into the groundwater and create a highly alkaline environment. Of concern in highly alkaline environments are the alteration of bentonite used as buffers or backfill materials, and of surrounding rock mass, and the increased uncertainty regarding the provision of performance of the disposal system over a long period of time. In this study, to reduce the quantity of cement-type materials that cause highly alkaline environments, technical feasibility of the support structure including the materials which considered the long-term performance of the HLW disposal system are discussed by using knowledge and technology accumulated in JAEA and Shimizu Construction. Moreover, based on the results, the problems remained in the application to the future HLW disposal institution are summarized.