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Nishino, Saki; Okada, Jumpei; Watanabe, Kazuki; Furuuchi, Yuta; Yokota, Satoru; Yada, Yuji; Kusaka, Shota; Morokado, Shiori; Nakamura, Yoshinobu
JAEA-Technology 2023-011, 39 Pages, 2023/06
JAEA-Technology-2023-011.pdf:2.51MB
Tokai Reprocessing Plant (TRP) which shifted to decommissioning phase in 2014 had nuclear fuel materials such as the spent fuel sheared powder, the diluted plutonium solution and the uranium solution in a part of the reprocessing main equipment because TRP intended to resume reprocessing operations when it suspended the operations in 2007. Therefore, we have planned to remove these nuclear materials in sequence as Flush-out before beginning the decommissioning, and conducted removal of the spent fuel sheared powder as the first stage. The spent fuel sheared powder that had accumulated in the cell of the Main Plant (MP) as a result of the spent fuel shearing process was recovered from the cell floor, the shearing machine and the distributor between April 2016 and April 2017 as part of maintenance. Removing the recovered spent fuel sheared powder was conducted between June 2022 and September 2022. In this work, the recovered powder was dissolved in nitric acid at the dissolver in a small amount in order to remove it safely and early, and the dissolved solution was sent to the highly radioactive waste storage tanks without separating uranium and plutonium. Then, the dissolved solution transfer route was rinsed with nitric acid and water. Although about 15 years had passed since previous process operations, the removing work was successfully completed without any equipment failure because of the organization of a system that combines veterans experienced the operation with young workers, careful equipment inspections, and worker education and training. Removing this powder was conducted after revising the decommissioning project and obtaining approval from the Nuclear Regulation Authority owing to operating a part of process equipment.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Shimada, Taro; Oshima, Soichiro; Ishigami, Tsutomu; Yanagihara, Satoshi
Proceedings of 10th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM '05) (CD-ROM), 8 Pages, 2005/09
In order to review an operator's "Decommissioning Plan" applied to the regulatory body for approval accurately and quickly, Public Dose Assessment Code for Decommissioning of Nuclear Reactors (DecDose) was developed. It evaluates public exposure doses for each year during dismantling activities in accordance with the "Decommissioning Plan". DecDose takes account of various exposure pathways such as cloudshine, surface ground deposition, ingestion of seafood, and direct and skyshine radiation based on the quantity of radionuclides discharged to the environment, and containers stored in the facility in consideration of cutting and storage of components and structures. Example evaluations with DecDose have shown that it is a useful tool for assessing public dose during decommissioning of nuclear reactors.
; ; ; Omori, Eiichi
JNC TN8410 2001-021, 33 Pages, 2001/09
JNC-TN8410-2001-021.pdf:4.37MB
Carbon-14 released from the nuclear facilities is an important radionuclide for the safety assessment, because it tends to accumulate in environment through food chain and has as a significant impact to personal dose. Carbon-14 has been monitored routinely as one of the main gaseous radionuclides exhausted from the Tokai Reprocessing Plant (TRP) since OCtober of 1991. Furthermore, behavior of carbon-14 in TRP has been investigated through the reprocessing operation and the literature survey. This report describes the result of investigation about the behavior of carbon-14 in TRP as followings. (1)Only a very small amount of carbon-14 in the fuel was liberated into the shear off-gas and most of it was liberated into the dissolver of-gass. Part of the carbon-14 was trapped at the caustic scrubber installed in the of-gas treatment process, and untrapped carbon-14 was released into the environment from the main stack. Amount of carbon-14 released from the main stack was about 4.1
6.5GBq every ton of uranium reprocessed. (2)Carbon-14 trapped at the caustic scrubbers installed in the dissolver off-gas and in the vessel off-gas treatment process is transferred to the low active waste vessel. Amount of carbon-14 transferred to the low active waste vessel was about 5.4 9.6GBq every ton of uranium reprocessed. (3)The total amount of carbon-14 input to TRP was summed up to about 11.915.5 GBq every ton of uranium reprocessed considering the released amount from the main stack and the trapped amount in the off-gas treatment devices. The amount of nitrogen impurity in the initial fuel was calculated about 1522ppm of uranium metal based on the measured carbon-14. (4)The solution in the low active waste vesselis concentrated at the evaporator.Most of the carbon-14 in the solution was transferred into concentrated solution. (5)Tokai vitrification Demonstration Facility (TVF) started to operate in 1994. Since then, carbon-14 has been measured in the ...
; Murata, Eiichi*; Sawahata, Yoshikazu*; Saito, Akira*
JNC TN8430 2001-002, 43 Pages, 2001/02
JNC-TN8430-2001-002.pdf:1.98MB
Japan Nuclear Cycle Development Institute (JNC) is designing the Low level radioactive Waste Treatment Facility (LWTF) in the Tokai Reprocessing Plant (TRP). The low level liquid waste generated the TRP is separated salt (NaNO
, etc) and radionuclide in liquid treatment process of LWTF. The process can get higher volume reduction than previous bituminization. Based on the engineering tests equal to the liquid treatment process of LWTF, the validity of operational condition in LWTF is evaluated. As the results, it is confirmed that all operational condition in the processes which is Iodine immobilization, Pre-filter filtration, Pre-treatment, Coprecipitation and Ultrafiltration are available.
Kojima, Takuji
Hoshasen To Sangyo, (89), p.4 - 7, 2001/01
This paper describes the present status of dosimetry as one of the useful technique for the process/quality control in radiation application and radiation research/testing. It introduces (a)activities in dose standardization of International Organization for Standardization(ISO) and International Atomic Energy Agency(IAEA) on publication of standards and organization of workshops, (b)recent efforts for consistency check in dose evaluation of electrons having energies above 4MeV, and (c)development status of dosimeter systems relevant to new requirements in dosimetry. It also comments on importance of planning for succession of high-dose dosimetry technology and education/training of dosimetry workers/researchers.
to obtain the high U solution; *; Sakurai, Koji*; ; Nomura, Kazunori; *
JNC TN8400 2000-032, 98 Pages, 2000/12
JNC-TN8400-2000-032.pdf:1.94MB
Concerning the preparation of high U solution for the crystallization process and the application of UO powder dissolution to that, the effects of final U concentration, dissolution temperature, nitric acid concentration and powder size on the dissolution of UO powder in the nitric acid where the final U concentration was 800g/L were investigated. The experimental results showed that the solubility of UO decreased with the increase of final UO concentration and powder size, and with the decrease of dissolution temperature and nitric acid concentration. It was also confirmed that in the condition where the final U concentration was sufficiently lower than the solubility of U, UO dissolution behavior in the high U solution could be estimated with the equation based on the fragmentation model which we had already reported. Based on these experimental results, the dissolution behavior of irradiated MOX fuel in high U solution was estimated and the possibility of supplying high U solution to the crystallization process was discussed. In the preparation of high U solution for the crystallization process, it was estimated that the present dissolution process (dissolution for fuel pieces of about 3cm long) needed a lot of time to obtain a high dissolution yield, but it was shorted drastically by the pulverization of fuel pieces. The burst of off-gas at the early in the dissolution of fuel powder seems to be avoidable with setting the appropriate dissolution condition, and it is important to optimize the dissolution condition with considering the capacity of off-gas treatment process.
Kajiyama, Tadashi;
JNC TN8410 2000-015, 7 Pages, 2000/10
JNC-TN8410-2000-015.pdf:0.09MB
Some falsification has been detected in the results of quality control data relating to the diameter of samples of pellets produced in the BNFL's MOX Demonstration Facility (MDF) on September 1999. This document is the outlines of inspection procedure for the MONJU fuel pellet in plutonium fuel center of JNC.
; Terano, Toshihiro; ; ; Okubo, Toshiyuki
JNC TN9410 2000-004, 30 Pages, 2000/03
JNC-TN9410-2000-004.pdf:0.86MB
The Operation and Maintenance Support Systems for JOYO are being developed, with the aim of ensuring the stable and safe operation of JOYO and improving operational reliability of future FBR plants. Plant Operation Management Expert Tool named JOYPET had been developed as one of the Operation and Maintenance Support Systems, which helps plant operation management. The following functions were developed and applied. (1)Papers management (Plant status management) function for maintenance activities (2)Isolation management support function for plant operation (3)Automatically drawing function of plant operation schedule (4)Isolation judgment function for plant operation By use this system, the plant management of JOYO was able to improved reliability and reduced manpower.
;
JNC TN9400 2000-034, 48 Pages, 2000/03
JNC-TN9400-2000-034.pdf:1.56MB
The study and the development to put FBR (Fast Breeder Reactor) to practical use have been doing. So many kinds of technologies are investigated to construct nuclear fuel recycle received to the society. The most important aim of reprocessing has been to extract U and Pu from spent fuels effectively, but, now, the demands for reprocessing are many kinds on nuclear fuel recycle system's construction. These need to be accepted sufficiently. The system that consists of electrolysis, extraction, with molten salt and melting metal, volatilization and condensation using the difference of vapor pressure is suggested, because, differently from LWR (Light Water Reactor), FBR can use the low decontamination factor's fuel. When the engineering scale plant is designed, the dry reprocessing has unsolved problems(ex. process flow) because of less demonstrative scale plants of the dry reprocessing than ones of the wet reprocessing. So the analysis and the estimation of mass balance that is most fundamental in the dry reprocessing system's design need to keep up with the system's alteration (to add new processes etc.) flexibly. This study aim is to develop the mass balance estimation code of dry reprocessing that satisfies the demand mentioned above.
Kon, Tetsuro
JNC TN8440 2000-004, 93 Pages, 2000/03
TCMD (Tokai Construction Maintenance Division) Intranet System Integration has started in 1995. The first active activities between 1995 and 1996 has reported in a previous PNC report (PNC PN8440 97-032 TCMD Intranet System Integration Part 1). This is the second active activity's report between 1997 and 1999. Main content of this report is as follows. TCMD Intranet System Integration has completed by TCMD LAN system construction with high-speed LAN equipments and WINS services. TCMD group scheduler and conference rooms appointment system has developed by the operation of Exchange Server 5.5. The plant construction management data base system has developed - ACCESS 97/SQL Server 6.5 version. The plant construction management data base system has also developed---Internet Explorer / Active Server Pages version. The formal TCMD homepage service has begun in l998.
*; Arai, Tsuyoshi*; Kumagai, Mikio*
JNC TJ9400 2000-002, 80 Pages, 2000/02
JNC-TJ9400-2000-002.pdf:4.67MB
In order to develop an economically efficient wet separation process other than solvent extraction for reprocessing spent FBR-fuel (MOX fuel), we have investigated the possibility of an advanced ion exchange process. Based on the fundamental research results, we proposed an advanced ion exchange process considering the characteristics of FBR-fuel cycle. The separation system consists of a main separation process using a novel anion exchanger which has a rapid kinetics and two extraction chromatography processes for minor actinides isolation using novel impregnation adsorbents with high selectivity. The chemical flow sheet, mass balance chart, list of main equipment and installation layout of each equipment were estimated and designed for the process in a reprocessing plant with the capacity of 200 tHM/y FBR-fuel. The process was pfeliminarny evalualed from the aspects of economy performance, recovery of potentially useable resources, minimization of environmental risk and proliferation-resistance by comparing with the advanced PUREX process. Furthermore, the subjects which are important for the practical application of the process are also listed.
*; *; Tanai, Kenji
JNC TN8400 99-050, 94 Pages, 1999/11
This paper reports on the evaluations of operational activities for a High Level Radioactive Wastes Disposal Facility, from initial acceptance of vitrified waste at a surface facility to emplacement engineered barriers in underground facilities. The purpose of this analysis is to confirm the technical feasibility of geological disposal. First, the basic design and repository system requirements are identified. Second, operational activities in surface facilities, access facilities and underground facilities are described. The required procedures and equipment, suitable for specific emplacement concepts and configurations for engineered barrier systems are discussed for specific examples. Countermeasures for potential adverse events or conditions are based on extensive civil engineering and mining experiences in Japan and abroad. The time schedule is also evaluated on the basis of these concepts. In addition, the concept of stationary and mobile radiation control areas is studied based on experiences and practice in current nuclear facilities. Finally future research and development items are summarized.
Tanai, Kenji; Iwasa, Kengo; Hasegawa, Hiroshi; Miura, K.*; Okutsu, Kazuo*; Kobayashi, Masaaki*
JNC TN8400 99-046, 177 Pages, 1999/11
For the construction of underground facilities comprising access tunnels, connecting tunnels, main tunnels and disposal tunnels, a large number of tunnels will be excavated in deep rock formations. These excavations will extend hundreds kilometers in total length. Therefore, special attention must be paid, to transporting large volume of debris, ventilation, emergency escape routes in case of accident, and other factors. In addition, special attention must also paid to potential accidents which might in underground excavations, including unstable facing phenomena (such as collapse and swelling of facing at weak layer sections), spring water flow resulting collapse of rock, gas eruption, and rock burst. While considering these factors to be emphasized during the construction of geological disposal facilities, the investigation reviewed the existing working methods on individual construction technologies of access tunnels, main tunnels, connecting tunnels, disposal tunnels, and disposal pit, based on the recognition that the present state deals with a wide range of geological environments, and conducted investigation about the construction methods for each tunnel on the basis current technologies, and described the outline of these methods. Furthermore, for the items to be particulaly emphasized on site characterization koko and siting data such as ground pressure and spring water, the investigation reviewed the current countermeasure works, and made survey on the phenomena appeared during actual tunnel construction works and their countermeasres, and carried out a study on effectiveness of these countermeasures. This constructing of disposal site in deep geological formations is basically possible by applying, or confirming, current excavation technologies for tunnels and underground facilities, A systematic construction system combines separate technologies relating to construction, (excavation technology, tunnel support work method, etc.). Such systems ...
Sugita, Yutaka; Fujita, Tomoo; Tanai, Kenji; Hasegawa, Hiroshi; Furuichi, Mitsuaki*; Okutsu, Kazuo*; Miura, K.*
JNC TN8400 99-039, 58 Pages, 1999/11
Regarding disposal techniques of high-level radioactive waste (HLW), the HLW is vitrified and then stored for cooling for a period of 30 to 50 years. After cooling, the HLW is isolated in the deep underground. The concept of geological disposal is based on the requirement to enclose the HLW in the deep underground for the long-term durability of the human's environmental safety. Backfilling of a repository is a unique activity on the geological disposal. If underground tunnels excavated to construct the repository are left, they may have significant influences on the barrier performance of an entire repository, such as: the mechanical stability of a tunnel may be damaged by rock stresses and a tunnel may provide a fast pathway for ground water flow. Therefore, the underground facilities are expected to be backfilled with a backfilling material after emplacement of the HLW and a buffer material. The material for the backfilling of the underground facilities is backfilling material. In this report, bentonite-aggregate mixture is considered, as one of the candidate materials for the backfilling material. Aggregate imitates the muck that is generated during construction phase of the underground facilities. The combination of backfilling, plugging and grouting is considered in some underground situations. Plug is composed of concrete material or clay-based one. Grouting material is concrete material or clay-based one, too. In this report, the concept of the backfilling, mechanical and hydrological characteristics of the bentonite-aggregate mixture, the function, work methods and a schedule of the backfilling materials, plugging and grouting are considered, and items of quality control for the bentonite-aggregate mixture, concrete material and grouting are listed.
Tachimori, Shoichi
JAERI-Review 97-018, 37 Pages, 1998/01
JAERI-Review-97-018.pdf:2.05MB
no abstracts in English
; *; Ikeda, Hisashi ; Kaminaga, Kazuhiro; ; ; Kuno, Yusuke
PNC TN8410 96-266, 67 Pages, 1996/05
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