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Journal Articles

Analysis of fast reactor fuel irradiation behavior in the MA recycle system

Ozawa, Takayuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07

In a recycle system for minor actinides (MAs) currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. MA content is expected to be $$sim$$5 wt.% in the future recycle system, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important behaviors would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. The MA-containing influences were evaluated with CEPTAR.V2, including fuel properties and analysis models to evaluate the MA-MOX fuel irradiation behavior, by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo. The irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed the influences of Am-containing.

Journal Articles

Oxide dispersion-strengthened/ferrite-martensite steels as core materials for Generation IV nuclear reactors

Ukai, Shigeharu*; Otsuka, Satoshi; Kaito, Takeji; de Carlan, Y.*; Ribis, J.*; Malaplate, J.*

Structural Materials for Generation IV Nuclear Reactors, p.357 - 414, 2017/00

 Times Cited Count:13 Percentile:3.45

Oxide dispersion strengthened (ODS) steels are the most promising candidate materials for fuel cladding of generation IV nuclear reactors. The progress and current status for development of ODS/FM(ferrite-martensite) steels conducted mainly in Japan and France are overviewed. The chemical compositions of ODS/FM steels are listed. Fabrication routes of cladding tube are mentioned for ferrite-type ODS steels using recrystallized process and martensite-type one using $$alpha$$-$$gamma$$ phase transformation. The optimized process is identical for both countries. Joining process between cladding and end-plug has been also developed by using the pressurized resistance upset welding method. The improvements brought by ODS/FM steels in high-temperature strength and irradiation resistance are verified.

Journal Articles

The Research of MOX fuels in Japan

Kato, Masato

Transactions of the American Nuclear Society, 114, p.987 - 988, 2016/06

In Japan, uranium and plutonium mixed oxide (MOX) has been developed as fuels of sodium-cooled fast reactors. The developing MOX fuels come in variety of O/M ratio, Pu content, minor actinide (MA) content and density. We have studied a science based fuel technology to evaluate fuel behaviors in fabrication process and irradiation condition of such various fuels. The technologies which are constructed based on experimental database can apply to mechanistic evaluation of fuel behaviors. To develop the science based fuel technology, many different varieties of basic properties have been investigated, and experimental database was constructed. And a mechanistic physical property model has been studied. The models contribute to describe various behaviors in fuel fabrication process and irradiation condition.

JAEA Reports

Summary of the dissolution experiments of the irradiated fast reactor fuels in CPF

; Koyama, Tomozo; Funasaka, Hideyuki

JNC-TN8400 2000-016, 188 Pages, 2000/03

JNC-TN8400-2000-016.pdf:3.6MB

We summarized the conditions and results of all dissolution experiments (bench scale experiments (dissolution of sheared fuel pins) and beaker scale experiments (dissolution of a few sheared fuels pieces) of the irradiated fast reactor fuels, which were carried out in the Chemical Processing Facility (CPF). The fabrication and irradiation conditions of the dissolved fuels were also put in order.

JAEA Reports

Study about the dissolution behavior of the irradiated fast reactor fuels in CPF

; Koyama, Tomozo; Funasaka, Hideyuki

JNC-TN8400 2000-014, 78 Pages, 2000/03

JNC-TN8400-2000-014.pdf:2.13MB

We investigated the factors which affected the dissolution of U and Pu to the nitric acid solution with the fragmentation model, which was based on the results of dissolution experiments for the irradiated fast reactor fuels in the Chemical Processing Facility(CPF). The equation that gave the fuel dissolution rate was estimated with the condition of fabrication (Pu ratio (Pu/(U+Pu))), irradiation (burn-up) and dissolution (nitric acid concentration, solution temperature and U+Pu concentration) by evaluating these effects quantitatively. We also investigated the effects of fuel volume ratio to the solution in the dissolver, burn-up and flouring ratio of the fuel on the f-value (the parameter which shows the diffusion and osmosis of nitric acid to the fuel) in the fragmentation model. It was confirmed that the fuel dissolution rate calculated with this equation had better agreement with the results of dissolution experiments for the irradiated fast reactor fuels in the CPF than that estimated with the surface area model. In addition, the efficiency of this equation was recognized for the dissolution of unirradiated U pellet and high Pu enriched MOX fuel. It was shown that the dissolution rate of the fuel slowed down at the condition of the high U-Pu concentration dissolution by the calculation of the dissolution behavior with this equation. The dissolution of the fuel can be improved by increasing the nitric acid concentration and temperature, but from the viewpoint of lowering the corrosion of the dissolver materials, it is desirable that the f-value is increased by optimizing the condition of shearing and stirring for the improvement of dissolution.

JAEA Reports

Development of sodium facilities for NSRR fast reactor fuel tests, 2; Sodium capsule

Yoshinaga, Makio; Nakamura, Takehiko; Yamazaki, Toshi*

JAERI-Tech 2000-017, p.59 - 0, 2000/03

JAERI-Tech-2000-017.pdf:2.31MB

no abstracts in English

JAEA Reports

A Study on the reprocessing of spent FBR-fuel by ion exchange

*; Arai, Tsuyoshi*; *

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.

JAEA Reports

Study of assessing aqueous reprocessing process for the pipeless reprocessing plant

*; *; *; *; *

JNC-TJ9400 2000-001, 112 Pages, 2000/02

JNC-TJ9400-2000-001.pdf:6.67MB

The purpose of this study is to investigate the possibility of new reprocessing process for the purpose of introducing pipeless plant concept, where aqueous separation methods other than solvent extraction method are adopted in order to develop more economical FBR fuel (MOX fuel) reprocessing process. At it's first stage, literature survey on precipitation method, crystallization method and ion-exchange method was performed. Based on the results, following processes were candidated for pipeless reprocessing plant. (1)The process adopting crystallization method and peroxide precipitation method (2)The process adopting oxalate precipitation method (3)The process under mild aqueous conditions (crystallization method and precipitation method) (4)The process adopting crystallization method and ion-exchange method (5)The process adopting crystallization method and solvent extraction method The processes (1)$$sim$$(5) were compared with each others in terms of competitiveness to the conventional reference process, and merits and demerits were evaluated from the viewpoint of applicability to pipeless reprocessing plant, safety, economy, Efficiencies in consumption of Resources, non-proliferation, and, Operation and Maintenance. As a result, (1)The process adopting crystallization method and peroxide precipitation method was selected as the most reasonable process to pipeless plant. Preliminary criticality safety analyses, main process chemical flowsheet, main equipment list and layout of mobile vessels and stations were reported for the (1) process.

Journal Articles

Experimental research on nitride fuel cycle in JAERI

Arai, Yasuo; Iwai, Takashi; Nakajima, Kunihisa; Shirai, Osamu; Suzuki, Yasufumi

Proc. of the Int. Conf. on Future Nuclear Systems (GLOBAL'99)(CD-ROM), 8 Pages, 1999/00

no abstracts in English

JAEA Reports

Designstudy on advanced nuclear fuel recycle system; Conceptual design study of recycle system using molten salt

; ; Moro, Satoshi; ; ; ;

JNC-TN9400 98-003, 422 Pages, 1998/10

JNC-TN9400-98-003.pdf:21.36MB

Advanced recycle system engineering group of OEC has being carried out a design study of the advanced nuclear fuel recycle system using molten salt (electro-metallurgical process). This system is aiming for improvements of fuel cycle economy and reduction of environmental burden (MA recycles, Mimmum of radioactive waste disposal), and also improvement of safety and nuclear non-proliferation. This report describes results of the design study that has been continued since December 1996. (1)A design concept of the advanced nuclear fuel recycle system, that is a module type recycle system of pyrochemical reprocessing and fuel re-fabrication was studied. The module system has advantage in balance of Pu recycle where modules are constructed in coincidence with the construction plan of nuclear power plants, and also has flexibility for technology progress. A demonstration system, minimum size of the above module, was studied. This system has capacity of 10 tHM/y and is able to demonstrate recycle technology of MOX fuel, metal fuel and nitride fuel. (2)Each process of the system, which are pyrochemical electrorefining system, cathode processor, de-cladding system, waste disposal system, etc., were studied. In this study, capacity of an electrorefiner was discussed, and vitrification experiment of molten salt using lead-boric acid glass was conducted. (3)A hot cell system and material handling system of the demonstration system was studied. A robot driven by linear motor was studied for the handling system, and an arrangement plan of the cell system was made. Criticality analysis in the cell system and investigation of material accountancy system of the recycle plant were also made. This design study will be continued in coincidence with design study of reactor and fuel, aiming to establish the concept of FBR recycle system.

Journal Articles

Fission gas release of uranium-plutonium mixed nitride and carbide fuels

Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo;

IAEA-TECDOC-970, 0, p.137 - 153, 1997/10

no abstracts in English

Journal Articles

Performance of uranium-plutonium mixed carbide fuel under irradiation

Suzuki, Yasufumi; Arai, Yasuo; Iwai, Takashi; Nakajima, Kunihisa

Proc. of Int. Conf. on Future Nuclear Systems (Global'97), 1, p.522 - 527, 1997/00

no abstracts in English

JAEA Reports

Fabrication of uranium-plutonium mixed carbide and nitride fuel pins for the irradiation test in JOYO

Arai, Yasuo; Iwai, Takashi; ; Okamoto, Yoshihiro; Shiozawa, Kenichi;

JAERI-Research 96-009, 17 Pages, 1996/02

JAERI-Research-96-009.pdf:1.09MB

no abstracts in English

JAEA Reports

None

; ; ;

PNC-TN8420 93-011, 40 Pages, 1993/07

PNC-TN8420-93-011.pdf:2.39MB

None

JAEA Reports

Application of a ferritic steel for advanced FBR fuel cladding; Material and welding tests

Iwai, Takashi; ;

JAERI-M 91-077, 44 Pages, 1991/05

JAERI-M-91-077.pdf:2.65MB

no abstracts in English

Journal Articles

Characteristics and irradiation behaviors of U-Pu mixed oxide, carbide, nitride and metal fuels for FBRs

Handa, Nuneo; ; Iwai, Takashi

Nippon Genshiryoku Gakkai-Shi, 31(8), p.886 - 893, 1989/08

 Times Cited Count:2 Percentile:65.64(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Determination of oxygen in uranium carbide

; ; ;

Nippon Genshiryoku Gakkai-Shi, 21(9), p.738 - 743, 1979/00

 Times Cited Count:0

no abstracts in English

Journal Articles

Some adaptations of dry techniques to the aqueous reprocessing

Nippon Genshiryoku Gakkai-Shi, 18(4), p.202 - 207, 1976/04

no abstracts in English

JAEA Reports

27 (Records 1-20 displayed on this page)