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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.
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.
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.
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.
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.
Sato, Takumi; Iwai, Takashi; Arai, Yasuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.1090 - 1098, 2007/09
The electrolysis of burnup-simulated uranium nitride, UN, containing representative solid fission product elements (Mo, Pd, Nd) was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl from the view point of application of pyrochemical reprocessing to nitride fuel cycle. It was found from cyclic voltammetry and anodic polarization curve measurement that anodic dissolution of UN began at about -0.75 V vs. Ag/AgCl reference electrode in all samples. After the electrolysis at the constant anodic potential of -0.65 
-0.60 V vs. Ag/AgCl, most of UN was dissolved into LiCl-KCl as UCl at the anode, and U was recovered in the liquid Cd cathode in all samples. Further, Nd was dissolved into LiCl-KCl as NdCl, while Mo and Pd were not dissolved but remained at the anode.
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.
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
.
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.
Amamoto, Ippei; Kofuji, Hirohide; Myochin, Munetaka; Terai, Takayuki*
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.748 - 757, 2007/09
FP such as alkali metals, alkaline earth metals and rare-earth elements are apt to remain in the eutectic medium used in pyroreprocessing even after treatment at the pyrocontactor step. It is desirable to have the spent electrolyte purified for recycling which in turn, could lead to the reduction of HLW. This study is carried out to evaluate the feasibility of the electrolyte recycling process by the phosphate conversion technique. First of all, a reference block flow diagram, which consists of three steps, was designed based on known developmental results from literature. Subsequently, evaluation was undertaken by comparison with conventional relevant experimental and theoretical analysis results after gathering the essential basic data for thermodynamic calculation. The obtained computational value was then reflected to establish the preliminary conceptual flow diagram which would facilitate the next discussion and experiment for the realization of this process.
Inoue, Tadashi*; Koyama, Tadafumi*; Myochin, Munetaka; Arai, Yasuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.728 - 737, 2007/09
no abstracts in English
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
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.