Research on factor analysis and technical process for achieving denuclearization, 3; Dismantlement and verification of nuclear reactor

Nakatani, Takayoshi; Shimizu, Ryo; Tazaki, Makiko; Kimura, Takashi; Hori, Masato

Dai-43-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2022/11

Denuclearization is the verification of nuclear development and the freezing, disabling, decommissioning, and verification of nuclear weapons, nuclear materials that can be used for nuclear weapons, and their manufacturing facilities, equipment, equipment, and materials. In this study, the technical process of measures and verification methods to effectively and efficiently achieve denuclearization at nuclear installations is examined and considered from a technical point of view.

Derivation of clearance levels for solid materials in Japan

Sakai, Akihiro; Okoshi, Minoru

Radiation Risk Assessment Workshop Proceedings, p.175 - 186, 2003/00

To establish the clearance levels, the Nuclear Safety Commission (NSC) has been discussing the clearance levels since May 1997. The NSC derived the unconditional clearance levels for the solid materials, namely concrete and metal, arising from the operation and dismantling of nuclear reactors and post irradiation examination (PIE) facilities. Two destinations of the cleared materials, namely disposal and recycle/reuse, were considered. Deterministic calculation models were established to assess individual doses resulting from 73 exposure pathways, and realistic parameter values were selected considering the Japanese natural and social conditions. The clearance levels for 21 radionuclides of nuclear reactors and for 49 of PIE facilities were derived as radioactivity concentration equivalent to the individual doses of 10 Sv/y. Most of calculated clearance levels were nearly the same as those shown in IAEA-TECDOC-855. Some, however, were different. It is considered that the major reasons depend on differences of fixed scenarios and of selected values of parameters.

Annual report on the environmental radiation monitoring around Tokai reprocessing plant FY 1999

; Shinohara, Kunihiko; ; ; ; Takeyasu, Masanori;

JNC TN8440 2000-007, 141 Pages, 2000/06

JNC-TN8440-2000-007.pdf:3.02MB

Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed since 1975, based on "Safety Regulations for the Tokai Reprocessing Plant, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitants due to the radioactivity discharged from the plant during April 1999 to March 2000. Appendices present comprehensive information, such as monitoring program, monitoring results, meteorological data and annual discharges from the plant.

Development of an on-site plant analyzer (1); Development of a GUI for building plant models for analyzes and retrieval of real-time plant data

JNC TN4400 2000-002, 33 Pages, 2000/06

JNC-TN4400-2000-002.pdf:5.22MB

An on-site plant analyzer can provide analysis support in evaluating plant dynamic characteristics when unplanned events occur in a nuclear power station. The plant analyzer contains a plant-dynamics analysis code, which efficiently and quickly analyzes the plant dynamic characteristics. Elements being developed for the on-site plant analyzer include utilities to build plant models for performing analyses and to retrieve plant operation data. The addition of these elements to the analysis code supports the plant-dynamics analysis works in MONJU, in particular, to assist the analyses of start up tests. The system contains the FBR plant-dynamics analysis code "Super-COPD", which is based on the "COPD" code that was used in the safety licensing of MONJU. One feature of the system is that all operations, e.g., assembling plant models for analysis, are prepared using a GUI (Graphical user Interface). In addition to this feature, the system is able to retrieve directly on- and off-line plant data from MIDAS, the Monju Integrated Data Acquisition System. These plant data are used to supply time-dependent boundary conditions for the plant analysis models. For this report, two case studies were performed. First, the analysis result of a turbine trip test at 40% power operation using the full plant model is described. For the second, performance of the IHX model was evaluated using retrieved plant data for boundary conditions. With the development of this system, improvement in the efficiency of analyses of MONJU start-up tests is expected.

Use results of MOX fuel in ATR Fugen nuclear power station

Ijima, Takashi; ; Matsumoto, Mitsuo; *

JNC TN3410 2000-002, 93 Pages, 2000/01

JNC-TN3410-2000-002.pdf:2.54MB

Fugen Nuclear Power Station ("Fugen") is a prototype Advanced Thermal Reactor (ATR), it has been demonstrated the plutonium utilization by loading many Mixed Oxide Fuels (MOX) since the reactor start up March 1979, and no fuel defect had been occurred, The MOX fuel assemblies has the high reliability and has been loaded more than 700 fuel assemblies. This is the largest in the world as a thermal neutron reactor. However, "Fugen" is planning to stop its operation in the year 2003, because the role of the Fugen almost finished. Therefore, we are going to summarize the ATR project including the Plutonium utilization experience. This paper is summarized as part of the experience.

Clearance levels for solid materials arising from nuclear reactors in Japan

Nakanishi*; *; Okoshi, Minoru; Takahashi, Tomoyuki*

Proceedings of 2nd International Symposium on Release of Radioactive Material from Regulatory Control, p.55 - 63, 1999/00

no abstracts in English

A Study on aerosol particle behaviors in dismantling nuclear facilities

; Tachibana, Mitsuo; Yanagihara, Satoshi

Proc. of SPECTRUM'98, p.227 - 232, 1998/00

no abstracts in English

ATR Data base 「Fugen」 design/ R&D /characteristics [Thermal/Hydraulic]

PNC TJ1409 97-021, 27 Pages, 1997/03

PNC-TJ1409-97-021.pdf:0.59MB

1 Reflection of R&D Results, Design and Operation Experiences All knowledge obtain in the project, such as R&D results, design, operation experiences and so on, are tbe reflected to following items. 1) Improvement of safety and reliability in plant eration 2) Design modification of the plant 3) Design of the next plant 2 Basic Sndpoint for Data Base Composition "Design/R&D/Plant Performance Data Base" will be mposed ad shown in the following, so that the Data Base could be utilized for reflecng them to the above items and for improving the above items, effectively and efficitly. (1) United Data Bases of Design and R&D "Design/R&D/Plant Performance Data Ba" will be composed by uniting design data base and R&D data base, considering that t R&D of the project is mainly made in order to establish the design engineering and chnical basis, such as design policy, design criteria, design conditions, alloable dign limits, design verification, etc. (2) Addition of Initial Plant Performa

Seismic design of reactors in NUCEF

*; *; *; *; ; Sakuraba, Koichi; Izawa, Naoki; Takeshita, Isao

JAERI-Tech 97-010, 276 Pages, 1997/03

JAERI-Tech-97-010.pdf:8.45MB

no abstracts in English

Development of reactor decommissioning technologies; Completion of the JPDR decommissioning program

Tanaka, Mitsugu

Nihon Kikai Gakkai-Shi, 100(938), 93 Pages, 1997/01

http://ci.nii.ac.jp/naid/110002438187

no abstracts in English