Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Yan, X.; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro
Journal of Nuclear Science and Technology, 55(11), p.1275 - 1290, 2018/11
To reduce environmental burden and thread of nuclear proliferation, multi-recycling fuel cycle with High Temperature Gas-cooled Reactor (HTGR) has been investigated. Those problems are solved by incinerating TRans Uranium (TRU) nuclides, which is composed of plutonium and Minor Actinoide (MA), and there is concept to realize TRU incineration by multi-recycling with Fast Breeder Reactor (FBR). In this study, multi-recycling is realized even with thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium by reprocessing and natural uranium are enriched and mixed with recovered TRU by reprocessing and partitioning to fabricate fresh fuels. The fuel cycle was designed for a Gas Turbine High Temperature Reactor (GTHTR300), whose thermal power is 600 MW, including conceptual design of uranium enrichment facility. Reprocessing is assumed as existing Plutonium Uranium Redox EXtraction (PUREX) with four-group partitioning technology. As a result, it was found that the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for High Level Waste (HLW) can be reduced by 99.7% compared with GTHTR300 using existing reprocessing and disposal technology. Suppress plutonium is not generated from this cycle. Moreover, incineration of TRU from Light Water Reactor (LWR) cycle can be performed in this cycle.
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 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.
Arai, Yasuo; Pillon, S.*
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles (CD-ROM), 9 Pages, 2004/06
no abstracts in English
Ishitsuka, Etsuo; Nakamichi, Masaru*; Uchida, Munenori*; Kawamura, Hiroshi; Kaminaga, Katsuo; Tsuboi, Kazuaki; Kusunoki, Hidehiko
JAERI-Conf 2004-006, p.262 - 264, 2004/03
no abstracts in English
Okubo, Tsutomu; Iwamura, Takamichi; Takeda, Renzo*; Moriya, Kumiaki*; Yamauchi, Toyoaki*; Aritomi, Masanori*
Nippon Kikai Gakkai 2003-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.245 - 246, 2003/08
A design study on a 300MWe class small Reduced-Moderation Water Reactor (RMWR) has been performed, based on the experienced LWR technology. The core can be cooled by the natural circulation and can achieve a conversion ratio of 1.01, a negative void reactivity coefficient, a core average burn-up of 65 GWd/t and a cycle length of 25 months. The system has been simplified as much as possible by introducing the passive safety components, in order to reduce the construction cost per electric power output overcoming “the scale demerit" for a small reactor comparing with the large one. The results show a 1.35 times higher cost than for the ABWR case, but suggest the possible lower cost when the effects such as the mass production are taken into account.
Okubo, Tsutomu; Iwamura, Takamichi; Takeda, Renzo*; Yamauchi, Toyoaki*; Okada, Hiroyuki*
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/00
A water-cooled reactor concept named Reduced-Moderation Water Reactor is under development for effective fuel utilization through plutonium multiple recycling based on the water-cooled reactor technology. The reactor aims at achievement of a high conversion ratio more than 1.0 with MOX fuel. Especially, the core performances during the Pu multiple recycling have been investigated for the advanced fuel reprocessing schemes with low decontamination factors than the current PUREX process, and are shown to achieve the conversion ratio more than 1.0 and the negative void reactivity coefficient.
Isotope News, (583), p.20 - 24, 2002/11
no abstracts in English
Arie, Kazuo*; Abe, Tomoyuki*; Arai, Yasuo
Nippon Genshiryoku Gakkai-Shi, 44(8), p.593 - 599, 2002/08
no abstracts in English
Kawamura, Hiroshi; Tsuchiya, Kunihiko
FZKA-6720, p.151 - 160, 2002/06
no abstracts in English
Okubo, Tsutomu; Iwamura, Takamichi; Yamamoto, Kazuhiko*; Okada, Hiroyuki*
Nippon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.571 - 574, 2002/00
Based on the experienced light water reactor technology, conceptual design studies on advanced water-cooled reactors have been performed. They are named “Reduced-Moderation Water Reactor" (RMWR) with the high conversion ratio more than 1.0 and the negative void reactivity coefficients. Several concepts have been successfully established for them based on the neutronics calculations. Based on these concepts, detailed investigations on such as plutonium multiple recycling and control rod planning have been performed as well as improvement on core performances. Through these detailed core design investigation, the feasibility of those designs has been confirmed step by step.
Iwamura, Takamichi; Okubo, Tsutomu; Yonomoto, Taisuke; Takeda, Renzo*; Moriya, Kumiaki*; Kanno, Minoru*
Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 8 Pages, 2002/00
Research and developments of reduced-moderation water reactor (RMWR) have been performed. The RMWR can attain the favorable characteristics such as high burn-up, long operation cycle, multiple recycling of plutonium and effective utilization of uranium resources, based on the matured LWR technologies. MOX fuel assemblies in the tight-lattice fuel rod arrangement are used to reduce the moderation of neutron, and hence, to increase the conversion ratio. The conceptual design has been accomplished for the small 330MWe RMWR core with the discharge burn-up of 60GWd/t and the operation cycle of 24 months, under the natural circulation cooling of the core. A breeding ratio of 1.01 and the negative void reactivity coefficient are simultaneously realized in the design. In the plant system design, the passive safety features are intended to be utilized mainly to improve the economy. At present, a hybrid one under the combination of the passive and the active components, and a fully passive one are proposed. The former has been evaluated to reduce the cost for the reactor components.
; ; Shigetome, Yoshiaki
JNC-TN8200 2001-006, 19 Pages, 2001/12
Nishimura, Kazuhisa; Shoji, Shuichi*; *; Sato, Seiichi*; ;
JNC-TN8430 2001-005, 64 Pages, 2001/09
The external gelation process is one of the candidates of MOX particle fuel fabrication for advanced recycle system. It was necessary to perform preliminary fuel fabrication with uranium before starting MOX tests. As the result of the preliminary examination, Basics conditions of the fabrication were obtained: (1)Optimized uranyl nitrate solution and PVA solution, as raw materials were prepared. (2)The frequency of vibration and the amount of flow were obtained with optimized broth (mixture) in the vibration dropping process. (3)The influence of composition of broth and concentration of ammonia solution on gelation process was obtained. (4)Impurities after aging, washing and drying spHerical gel were surveyed, (5)The spherical gel were calcined to oxide particles and the particles were characterized by TG-DTA, therefore it is certain that outlook on the sintered particles as final products is very clear. On the top of that, there were no fatal technicalities of the external gelation process through material balance and a diameter dispersion of spherical gel and oxide particles. It is necessary to perform uranium examination to solve some new problems, for instant, surface crack of spherical gel. Although almost of all the preparations are completed and fabrication examination of MOX particles with vibration dropping equipment are ready for starting.
JNC-TN8400 2001-022, 60 Pages, 2001/03
A numerical simulation code for the TRUEX (Transuranium Extraction) process was developed. Concentration profiles of americium and europium were calculated for some experiments of the counter current extraction system those were carried out in CPF (Chemical Processing Facility) by using the code. Calculation profiles were in agreement with the experimental results. Operational conditions were also examinted for the americium recovery experiment by the TRUEX process carried out in the Plutonium Fuel Center. It was shown that lowering the concentration of nitric acid in the scrub solution and decreasing the flow rate of solvent and strip solution was effective for improving the performance of the stripping step and reducing the volume of the waste solution. In order to find the optimum conditions for various experiments, this simulation code was modified to calculate the concentration profiles of other metal elements such as zirconium and iron and the effect of oxalic acid on the extraction behavior of the metal elements. The calculated concentration profiles of americium and europium were varied by this modification. In the experiment at CPF, the calculations were carried out to obtain recovery ratio of americium in the product stream with the amount of oxalic acid added to the process. This calculation result showed that it was possible to improve the performance of decontamination of fission products by increasing oxalic acid concentration added to the process. The calculation was also carried out for finding the optimum conditions of oxalic acid concentration added to the europium recovery process.
Tatematsu, Kenji; Tanaka, Yoji*; Sato, Osamu
JAERI-Research 2001-014, 25 Pages, 2001/03
no abstracts in English
JNC-TN8400 2000-029, 54 Pages, 2000/10
This report describes the study done by the author as a postdoctoral research associate at Japan Nuclear Cycle Development Institute. This report is divided into three parts: construction of a relativistic band calculation formalism based on the density functional theory, using this method, investigation of the electrical properties for ferromagnetic UGe and antiferromagnetic UO. (1)A relativistic band calculation (RBC) method. Band calculations for the s, p, and d electric structure have been developed well in the practical application and theoretical study. But band calculation method treating magnetic 5f electrons as actinide compounds are complicated and needed relativistic approach, so it is behind with the study of the 5f system. In this study we construct the relativistic band calculationformalism valid for magnetic 5f electrons. (2)Electric properties of UGe. The actinide compounds UGe is ferromagnetic, so the theoretical analysis is not well yet. The electric structure and fermi surface of UGe are analyzed using the RBC. The theoretical results show that UGe is heavy electron with the 5f character and are agreement with experimental one. (3)Electric structure of nuclear fuel UO. It is important to understand the mechanism of the thermal conductivity of nuclear fuel as antiferromagnetic UO. The UO band calculation reflecting the thermal properties, into account of relativistic effect, have not done yes. So using the RBC the detailed electric structure of UO are obtained.
JNC-TN8400 2000-028, 70 Pages, 2000/10
This report describes the study done by the author as a postdoctoral research associate at Japan Nuclear Cycle Development Institute. This report is divided into two parts: improvements in accuracy in determination of thermal neutron capture cross sections, and improvements in accuracy of photo-nuclear absorption cross section measurements using the HHS. (1)In the measurements of thermal neutron capture cross sections using an activation method, accuracies of the final results attained are limited by (1) accuracy of -ray peak detection efficiencies, and (2) accuracies of -ray emission probabilities. In this study; to determine thermal neutron capture cross sections more accurately, the following researches have been done using a newly developed three-dimensional coincidence measurement system: (1)accurate determination of -ray standard sources using a - coincidence method, for precise calibration of -ray peak detection efficiency, and (2) development of a - coincidence measurement system using a plastic scintillation detector as a -ray detector, for the determination of -ray emission probabilities of short-lived nuclides, and measurement of -ray emission probabilities of Tc nuclide using the coincidence system. (2)To transform radioactive nuclides with small thermal neutron capture cross sections, use of photonuclear absorption reaction has been suggested. In order to transform these nuclides efficiently using the reaction, one has to know detailed behavior of the photo-absorption cross sections. In this study, a Monte-Carlo simulation code has been used to create a standard set of -ray response functions of the high-resolution high-energy spectrometer (HHS), to enable reliable analyses of the data obtained by the spectrometer.
JNC-TN1440 2000-007, 115 Pages, 2000/08
no abstracts in English
JNC-TN9400 2000-054, 84 Pages, 2000/04
This report describes accomplishment of the study on the quality of vipac (vibro-packed) oxide fuel obtained by pyrochemical processing (molten salt electrolytic processing). This study is intended to contribute to the design study of the pyro-reprocessing-vipac fuel recycling system of oxide fuel. In this study, vibro-packing experiment has been conducted using granular U0 obtained by molten salt electrolytic processing (cold experiment). The oxide pyro process developed by Research lnstitute of Atomic Reactors (RIAR) is the method in which the sintered oxide is electrically deposited on the cathode at approximately 600C. 0xide granules for vipac fuel are obtained by crushing the oxide deposited on the cathode. This process is also developed as recycle process because it is capable of FP separation. Also in Japan, this process is studied as one of the new FBR fuel recycling systems. ln this study, we made an effort to clarify the mechanisms of vibro-packing of the electrically obtained granules, which influence on the effective parameters of vibro-packing density and fuel particles size distribution in the fuel cladding in case of non-sphere particles of the granules. As a result of the study, smear density of 75% and almost uniform distribution of U0 particles have been taken in the experiment, and much knowledge for the improvement of the vibro-packing quality has been found. And the possibility of the smear density over 80% and the uniform distribution of U0 particles has been suggested in this study.