Waste acceptance criteria for waste package destined for trench-type disposal facilities for waste generated from Research, Industrial and Medical Facilities; No harmful void

Nakata, Hisakazu; Takao, Hajime*; Chijimatsu, Masakazu*; Noma, Yasutaka*; Amazawa, Hiroya; Sakai, Akihiro

JAEA-Technology 2018-014, 43 Pages, 2019/03

JAEA-Technology-2018-014.pdf:5.91MB

Japan Atomic Energy Agency plans to install disposal facilities for radioactive waste arising from research institutes. One relevant technical standard by the safety regulation is that the disposal facility shall be performance so as not to be left with harmful voids after backfilling with soil. Additionally no harmful void needs to exist in the waste packed in metal containers. The harmful void is supposed to result in the collapse of the disposal facility after structural materials of the container deteriorate and then become a state that can not retain the structure on its own. That leads to have an adverse impact on the facility such that the shape of cover soil deforms the way in which stagnant water is likely to occure. For which reason, a waste acceptance criteria relating to the quantity of voidage in a waste package needs to be defined quantitatively, which is preliminary less than 20% in a volum ratio based on this study.

Waste Technical Standards Working Group annual report 2016

Waste Technical Standards Working Group

JAEA-Review 2017-017, 112 Pages, 2017/11

JAEA-Review-2017-017.pdf:2.87MB

In Japan Atomic Energy Agency, JAEA, a Waste Technical Standards Working Group has established since FY2015. The Working Group is composed of the members from waste management sections in each site in JAEA and from Radioactive Waste Management and Disposal Project Department. In this Working Group, we discussed quality management on conditioning waste packages, methodologies to evaluate the radioactivity concentration and measures for dismantling waste. This annual report summarizes the results of discussion in FY2016.

Waste acceptance criteria for waste packages destined for near surface disposal containing radioactive waste from research, industrial and medical facilities

Okada, Shota; Izumo, Sari; Nakata, Hisakazu; Tsuji, Tomoyuki; Sakai, Akihiro; Amazawa, Hiroya

JAEA-Technology 2016-023, 129 Pages, 2016/11

JAEA-Technology-2016-023.pdf:8.95MB

Waste packages must meet the technical requirements. This is because JAEA has been preparing an operating procedure manual for quality control of radioactive waste disposal to be applied to the processing of the waste packages. Raw wastes generated by JAEA are segregated and stored by a method specified in the manual. The composition of raw wastes was characterized on the basis of records of the segregation process. Simulated waste packages were produced by placing the waste materials in a 200 liter drum, which was then filled with mortar, followed by curing in a controlled manner. The static load test was conducted to measure deformation and strain performance of the simulated waste package. Compression apparatuses which can imitate loading conditions in pit-type and trench-type facility that are planned by JAEA were used. Based on the test result, waste packages produced in accordance with the manual met the technical requirement under the condition.

Waste acceptance criteria for waste packages destined for near surface disposal containing radioactive waste from research, industrial and medical facilities

Nakata, Hisakazu; Sakai, Akihiro; Okada, Shota; Izumo, Sari; Tsuji, Tomoyuki; Kurosawa, Ryohei; Amazawa, Hiroya

JAEA-Technology 2016-001, 112 Pages, 2016/03

JAEA-Technology-2016-001.pdf:16.71MB

The waste packages must meet the technical requirements that radioactive waste shall be solidified in a container by a method determined by the Nuclear Regulation Authority to prevent from radiation hazards. JAEA has been preparing operating procedure manual on quality control for radioactive waste disposal in order to promote the manufacturing the waste package. This report presents that simulant waste packages were produced by placing wastes in a 200 liter drum, which was then filled with mortar of a novel mix proportion, followed by curing in a controlled manner. Determination of the presence of harmful voidage and raw waste immobility were performed by direct measurement and visual inspection of a vertical cross section of the waste packages respectively.

Evaluation of Coupled Thermo-Hydro-Mechanical Phenomena in the Near Field for Geological Disposal of High-Level Radioactive waste

Chijimatsu, Masakazu*; Fujita, Tomoo; Sugita, Yutaka; Taniguchi, Wataru

JNC TN8400 2000-008, 339 Pages, 2000/01

JNC-TN8400-2000-008.pdf:30.96MB

Geological disposal of high-level radioactive waste (HLW) in Japan is based on a multibarrier system composed of engineered and natural barriers. The engineered barriers are composed of vitrified waste confined within a canister, overpack and buffer material. Highly compacted bentonite clay is considered one of the most promising candidate buffer material mainly because of its low hydraulic conductivity and high adsorption capacity of radionuclides. In a repository for HLW, complex thermal, hydraulic and mechanical (T-H-M) phenomena will take place, involving the interactive processes between radioactive decay heat from the vitrified waste, infiltration of ground water and stress generation due to the earth pressure, the thermal loading and the swelling pressure of the buffer material. In order to evaluate the performance of the buffer material, the coupled T-H-M behaviors within the compacted bentonite have to be modelled. Before establishing a fully coupled T-H-M model, the mechanism of each single Phenomenon or partially coupled phenomena should be identified. Furthermore, in order to evaluate the coupled T-H-M phenomena, the analysis model was developed physically and numerically and the adequacy and the applicability was tested though the engineered scale laboratory test and in-situ test. In this report, the investigative results for the development of coupled T-H-M model were described. This report consists of eight chapters. In Chapter l, the necessity of coupled T-H-M model in the geological disposal project of the high-level radioactive waste was described. In Chapter 2, the laboratory test results of the rock sample and the buffer material for the coupled T-H-M analysis were shown. The rock samples were obtained from the in-situ experimental site at Kamaishi mine. As the buffer material, bentonite clay (Kunigel V1 and Kunigel OT-9607) and bentonite-sand mixture were used. In Chapter 3, in-situ tests to obtain the rock property were shown. As ...

Application of plasma-induction-hybrid melter to the research on volume reduction and stabilization of low level radioactive solid waste

Hirabayashi, Takakuni; Kanazawa, Katsuo; Fujiki, Kazuo; *; *

Proc. of Int. Conf. on Incineration & Thermal Treatment Technologies (IT3 Conference), p.261 - 264, 1998/00

no abstracts in English

Development of partial models of comprehensive safety assessment model for shallow land disposal of low-level radioactive wastes, II; Simulation model for the release of radionuclides from an engineered barrier(ENBAR)

Matsuzuru, Hideo; Kurosawa, Naohiro; *

JAERI-M 87-124, 43 Pages, 1987/08

JAERI-M-87-124.pdf:1.11MB

no abstracts in English

Construction of Glass Database, 1; Acquisition of required thermodynamic properties for the construction of computational phase diagrams of vitrified wastes

Amamoto, Ippei; Oyama, Koichi; Nagano, Yuichi*; Jantzen, T.*; Hack, K.*; Fukayama, Daigen*

no journal, , 

As the vitrified study of the high-level radioactive waste is usually carried out under the high-temperature circumstance, it spends a lot of time and effort. The actual experiments and/or measurement, therefore, should be undertaken rationally after ascertaining the behaviors of target materials by the theoretical calculation, if possible. From such point of view, the construction of phase diagrams is considered after obtaining necessary thermodynamic properties from existing phase diagrams by CALPHAD method and/or published data. In this paper, several phase diagrams are presented such as the borosilicate glass which is currently used as the vitrified medium for the HLW, the iron-phosphate glass which will be potential vitrified medium for various wastes. Some phase diagrams are also prepared for the vitrified wastes which was loaded fission products such as molybdenum, palladium, etc..

Waste acceptance criteria for near surface disposal project of low-level radioactive wastes generated in research, industrial and medical facilities

Nakata, Hisakazu

no journal, , 

Low level radioactive wastes are generated from the research and development of the nuclear energy, medical and industrial use of radioisotope in Japan. Almost all the wastes are still in storage without conditioning and/or treatment suitable for final disposal. The total amount of the waste is estimated approximately 590,000 drums. In order to promote the disposal project for the waste generated from both JAEA and others, JAEA has become its implementing organization in 2008. Since then, JAEA carried out preliminary design of near surface disposal facilities under generic site conditions and estimated the total project cost to accommodate the waste quantity to be required; 750,000 drums. In addition, site selection plans and technical studies in designing disposal facilities are also being discussed. In this report, it explains technical studies relating to Waste Acceptance Criteria for supporting appropriate waste conditioning and its planned implementation by the generators.