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Kato, Masato; Nakamichi, Shinya; Takano, Tatsuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.916 - 920, 2007/09
no abstracts in English
Murakami, Tatsutoshi; Suzuki, Kiichi; Aono, Shigenori
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.891 - 896, 2007/09
no abstracts in English
Ozawa, Takayuki
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.404 - 408, 2007/09
The probabilistic annular fuel design code "BORNFREE-CEPTAR" was developed for the reasonable design of annular fuels to be applied for fast reactors in future. In the probabilistic design method, the performance parameters, i.e. fuel center temperature, cladding temperature, cladding stress, etc., used to be evaluated with the Monte Carlo method under the irradiation behavior, and the quantitative design margin could be obtained. As the result of probabilistic evaluation with this code, the possibility of the improvement of reactor performance of the advanced fast reactor was quantitatively indicated.
Kitamura, Akihiro; Okada, Takashi; Kashiro, Kashio; Yoshino, Masanori*; Hirano, Hiroshi*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.531 - 536, 2007/09
We present our waste handling activities in glovebox dismantling facility, installed in Plutonium Fuel Production Facility, Nuclear Fuel Cycle Engineering Laboratories, JAEA. In this facility, we treat only one size gloveboxes (3m
3m1m), but for the future waste treatment, we segregate waste into material categories. We analyzed the data collected for the future decommissioning, waste treatment and waste disposal. We also present the improvements which are already made and will be made in the near future.
Morita, Yasuji; Kim, S.-Y.; Ikeda, Yasuhisa*; Nogami, Masanobu*; Nishimura, Kenji*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1508 - 1512, 2007/09
We have been developing an advanced reprocessing system for spent FBR fuels based on precipitation method using pyrrolidone derivatives. In previous investigation, N-cyclohexyl-2-pyrrolidone (NCP) is used as a precipitant and a process consisting of selective U precipitation step and U-Pu co-precipitation step was developed. In order to make the process more effective and more economical, we are now studying precipitation of U and Pu with other pyrrolidone derivatives. In the present study, precipitation behavior of Pu was examined using N-butyl-2-pyrrolidone (NBP) and N-propyl-2-pyrrolidone (NProP), which have lower hydrophobicity than NCP. The experiments with Pu(IV) or Pu(VI) solutiona and U(VI)-Pu(IV) solutions showed that Pu is less precipitated with NBP or NProP than with NCP. From these results, it is expected that NBP and NProP can be used as precipitants for the selective U precipitation step and make the step more selective and effective.
Ikeda, Yasuhisa*; Takao, Koichiro*; Harada, Masayuki*; Morita, Yasuji; Nogami, Masanobu*; Nishimura, Kenji*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1503 - 1507, 2007/09
We have developed a reprocessing process for spent FBR fuels based on the precipitation method using pyrrolidone derivatives. In previous investigation, N-cyclohexyl-2-pyrrolidone (NCP) is used as a precipitant and a process consisting of selective U precipitation step and U-Pu co-precipitation step was developed. In the present study, in order to examine the applicability of precipitants with lower hydrophobicity than NCP to the selective U precipitation step, we have carried out precipitation experiments of U(VI) by N-butyl-2-pyrrolidone (NBP) and N-propyl-2-pyrrolidone (NProP) and measured decontamination factors of some fission products.
Ojima, Hisao; Dojiri, Shigeru; Tanaka, Kazuhiko; Takeda, Seiichiro; Nomura, Shigeo
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.273 - 282, 2007/09
The Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency (JAEA) was established to take over activities of the Tokai Works of Japan Nuclear Cycle Development Institute (JNC). From 1959, several kinds of technologies (such as uranium refining, centrifuge for uranium enrichment, LWR spent fuel reprocessing and MOX fuel fabrication) have been accomplished. And also, R&Ds on the treatment and disposal of high level waste and the FBR fuel reprocessing have been carried out. Through such activities, control of environmental release of radioactive material and radiation exposure and management of nuclear materials have been done appropriately. The Laboratories will contribute to establish the closed cycle with R&Ds of the reprocessing technology during the transition period from LWR era to FBR era, improved MOX fuel fabrication technology, advanced FBR fuel reprocessing technology and high level waste disposal technology.
Okamura, Nobuo; Takeuchi, Masayuki; Ogino, Hideki; Kase, Takeshi; Koizumi, Tsutomu
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1070 - 1075, 2007/09
no abstracts in English
Aihara, Jun; Ueta, Shohei; Mozumi, Yasuhiro; Sato, Hiroyuki; Motohashi, Yoshinobu*; Sawa, Kazuhiro
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.416 - 422, 2007/09
In high temperature gas-cooled reactors (HTGRs), coated particles are used as fuels. For upgrading HTGR technologies, present SiC coating layer which is used as the 3rd layer could be replaced with ZrC coating layer which have much higher temperature stability in addition to higher resistance to chemical attack by fission product palladium than the SiC coating layer. The ZrC layer could deform plastically at high temperatures. Therefore, the Japan Atomic Energy Agency modified an existing pressure vessel failure fraction calculation code to treat the plastic deformation of the 3rd layer in order to predict failure fraction of ZrC coated particle under irradiation. Finite element method is employed to calculate the stress in each coating layer. The pressure vessel failure fraction of the coated fuel particles under normal operating condition of GTHTR300C is calculated by the modified code. The failure fraction is evaluated as low as 3.510
.
Sagayama, Yutaka
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.251 - 258, 2007/09
JAEA launched a new FR Cycle Technology Development (FaCT) Project in cooperation with the Japanese electric utilities. The FaCT project is based on the conclusion of the Feasibility Study on Commercialized FR Cycle Systems (FS), which carried out in last seven years. In the FS, the combination of the sodium-cooled FR with oxide fuel, the advanced aqueous reprocessing and the simplified pelletizing fuel fabrication was selected as the main concept which should be developed principally. A conceptual design study of the main concept and R&D of innovative technologies are implemented toward an important milestone at 2015. The development targets, which were set up at the beginning stage of FS, were revised for the FaCT project based on the results of FS and change in Japanese society environment and in the world situation. International collaboration is promoted to pursue fast reactor cycle technology which deserves the global standard and its efficient development.
Sugino, Kazuteru
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.653 - 661, 2007/09
For a validation of MAs nuclear data and improvement on the prediction accuracy of MAs transmutation properties in fast reactor cores, the MAs sample irradiation tests data of Joyo were utilized. Result of their analyses showed good agreement with experimental value, which indicates that the MAs cross sections in JENDL-3.3 are almost satisfactory for an application to fast reactor cores. Further, the present study clarified that the utilization of those data with cross section adjustment technique has the potential to reduce the uncertainty of MAs transmutation properties in fast reactor cores to less than half.
Sato, Hiroyuki; Kubo, Shinji; Sakaba, Nariaki; Ohashi, Hirofumi; Sano, Naoki; Nishihara, Tetsuo; Kunitomi, Kazuhiko
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.812 - 819, 2007/09
The objective of this study is to confirm the availability of proposed mitigation methodology against thermal load increase events initiated by the thermochemical water splitting IS process hydrogen production system coupling with the HTTR. JAEA has been performing the development of dynamic simulation code which can evaluate complex phenomena in the HTTR-IS system all at one once to achieve the requirement. The notable feature of the developed code is the APHX module which enables to estimate the IS process thermal load variation considering phase change and chemical reaction behavior assumed in the APHX. In this paper, two cases of dynamic calculation for the thermal load increase events were performed using the newly developed APHX module. The results of the analytical studies clearly show the availability of the developed model for dynamic simulation of the HTTR-IS system and the thermal load increase mitigation methodology.
Iwamura, Takamichi; Okubo, Tsutomu; Uchikawa, Sadao
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1718 - 1724, 2007/09
An advanced LWR concept of FLWR for TRU recycling has been investigated. The design study has shown the promising results for the feasibility of the concept, in conjunction with the investigated results obtained from the related R&D's for some key issues of FLWR development. In order to establish a robust nuclear energy supply system for the future, an appropriate combination of both the LWR and the FBR technologies, i.e. FLWR and Na-FBR, is considered to be preferable and realistic. This type of preferable combination is proposed in this paper.
Tanaka, Nobuyuki; Yoshida, Mitsunori; Okuda, Hiroyuki; Sato, Hiroyuki; Kubo, Shinji; Onuki, Kaoru
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.833 - 836, 2007/09
Breakdown of the cell voltage in the electro-electrodialysis process for concentrating HIx solution (HI-H
O-I mixture) was preliminarily examined in an effort to clarify the optimal operation condition as well as to optimize the cell design for the application to the thermochemical water-splitting IS process for large-scale hydrogen production. Basic data such as electric resistance of HIx solution, overvoltage of the iodine-iodide ion redox reaction at graphite electrode, and the membrane voltage drop, were measured using HIx solution with composition of interest. Also, a methodology for estimating the cell voltage was discussed. The calculated cell voltage agreed well with the experimental one indicating the validity of the procedure adopted.
Sugawara, Takanori; Nishihara, Kenji; Tsujimoto, Kazufumi; Iwanaga, Kohei; Kurata, Yuji; Sasa, Toshinobu; Oigawa, Hiroyuki
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.998 - 1007, 2007/09
no abstracts in English
Oigawa, Hiroyuki; Nishihara, Kenji; Yokoo, Takeshi*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.434 - 442, 2007/09
In Japan, the partitioning and transmutation (PT) technology is being studied and developed to reduce the burden of the high-level radioactive waste (HLW) management. To demonstrate clearly the benefit of the PT technology on the waste management of future nuclear fuel cycles, the repository area necessitated to dispose of the HLW was discussed quantitatively for spent fuels from UO-LWR, MOX-LWR and MOX-FBR. Four options of separation process were assumed in the analysis: (1) Conventional PUREX reprocessing, (2) Transmutation of minor actinide (MA), (3) Partitioning of FP, and (4) PT for both MA and FP. The results showed that MA transmutation would be necessary to keep the emplacement area for MOX fuel at the same level as that for UO fuel. The adoption of PT for both MA and FP was effective to further reduce the repository area independently on the fuel type, the reactor type and the cooling period.
Kaito, Takeji; Otsuka, Satoshi; Inoue, Masaki
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.37 - 42, 2007/09
High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective candidate materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future commercialized SFRs. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and the Fast Flux Test Facility (FFTF). The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to improve the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025.
Kawaguchi, Koichi; Namekawa, Takashi
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.290 - 295, 2007/09
The JAEA has developed advanced FBR cycle system since 1999 as the Feasibility Study (FS). Several combination of fuel and reactor type, reprocessing method and fuel fabrication method were studied. As the result of FS, the combination of oxide fuel, sodium cooling reactor, advanced aqueous reprocessing system and simplified pelletizing fuel fabrication system is chosen as the most promissing fuel cycle system. In the Fast Reactor Cycle technology (FaCT), six development issues for simplified pelletising technology were selected. TRU fuel handling technology, which is heat removal from nuclear fuel material, is one of these issues. Accumulation of decay heat of MA which is contained in TRU fuel cause oxidation of fuel powder, fuel pellet and cladding tube. Authors designed concept of powder hoppper, O/M adjusting furnace and fuel assembling equipment with heat removal function, and evaluated temperature distribution using thermal hydraulics analysis technique. As a result, it is shown that it is possible to cool fuel materials with specific heat generation up to 20 W/kgHM enough, though more detailed study is required for comprehensive equipments.
Koyama, Shinichi; Ozawa, Masaki; Okada, Ken*; Kurosawa, Kiyoko*; Suzuki, Tatsuya*; Fujii, Yasuhiko*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1530 - 1536, 2007/09
Simplified separation process was proposed based on ion-exchange technique. HCl, HNO
and MeOH were used as an eluent. To develop an engineering scale concept, it is indispensable to establish the condition for safety operation. Corrosion test of structural materials in the HCl was performed by using some metals. In this experiment, it was proved that the Ta, Zr, Nb and hastelloy have good endurance to HCl solution. Research of thermal hazard of pyridine-type ion-exchange resin, MeOH and HNO media system was studied in the view point of fire and explosion safety. There is no hazardous reaction between IER/MeOH, HNO media system. In the case of more than 150
C, we should pay attention to the exothermic reaction at dried condition NO-IER or IER/HNO media system.
Ueno, Fumiyoshi; Kato, Chiaki; Motooka, Takafumi; Ichikawa, Shiro*; Yamamoto, Masahiro
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1389 - 1393, 2007/09
Authors were aimed for development of life evaluation method of components and clarification of corrosion mechanism of the components in nuclear reprocessing plant. Corrosion behavior of heat exchanger tubes in the reduced pressure evaporator made by ultra-low carbon type 304ULC stainless steel was studied. A simplified mock-up test apparatus was used for corrosion test with long-term test duration. Following results were obtained. The corrosion rates were increased from beginning of the test to more than 25,000 hours and then corrosion rate was reached to constant. From the measurement results of intergranular penetration depths, it was thought that intergranular corrosion was progressed on entire grain boundary around a grain and then the grain dropped out to the solution.