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JAEA Reports

Criticality safety evaluation for the direct disposal of used nuclear fuel; preparation of data for burnup credit evaluation (Contract research)

Yamamoto, Kento*; Akie, Hiroshi; Suyama, Kenya; Hosoyamada, Ryuji*

JAEA-Technology 2015-019, 110 Pages, 2015/10

JAEA-Technology-2015-019.pdf:3.67MB

In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. The recent development of higher-enrichment fuel has enhanced the benefit of the application of Burnup Credit. In the present study, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study.

Journal Articles

Comparative study of plutonium and minor actinide transmutation scenario

Nishihara, Kenji; Iwamura, Takamichi*; Akie, Hiroshi; Nakano, Yoshihiro; Van Rooijen, W.*; Shimazu, Yoichiro*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.388 - 395, 2015/09

The present study focuses on transmutation of Pu and minor actinide in Japanese case without utilizing Pu as resource. Pu can be transmuted by two groups of technology: conventional ones without reprocessing of spent fuel from transmuter and advanced ones with reprocessing. Necessary number of transmuters, inventory reduction of actinide and impact on repository are revealed by nuclear material balance analysis. As a whole advanced technology performs better in transmutation efficiency, although required number of transmuters is larger.

Journal Articles

A Study on the criticality safety for the direct disposal of used nuclear fuel in Japan; Application of burnup credit to the criticality safety evaluation for the disposal canister

Yamamoto, Kento; Akie, Hiroshi; Suyama, Kenya

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.228 - 237, 2015/09

Japan has recently started to study the technical issues for the direct disposal of the used nuclear fuel (UNF) to prepare various disposal options. The criticality safety is important for the direct disposal because of the presence of certain amount of the fissile nuclides in UNF. This paper gives the outline of the research to be addressed in this field and the relevant studies in Japan. Especially, it presents the first result of the criticality safety evaluation for a disposal canister model adopting burnup credit. The uncertainties of effective neutron multiplication factor ($$k_{eff}$$) caused by the depletion calculation errors as well as the effect of the axial burnup profile and the horizontal burnup gradient on $$k_{eff}$$ were also evaluated. It was found that the $$k_{eff}$$ including these uncertainties and conservatism was below 0.95 for the representative used PWR fuel when the fuel assemblies and the disposal canister were assumed to keep intact.

Journal Articles

Utilization of rock-like oxide fuel in the phase-out scenario

Nishihara, Kenji; Akie, Hiroshi; Shirasu, Noriko; Iwamura, Takamichi*

Journal of Nuclear Science and Technology, 51(2), p.150 - 165, 2014/02

 Times Cited Count:2 Percentile:73.24(Nuclear Science & Technology)

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by material balance analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the nuclear material balance analysis (NMB) code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels and an ability to estimate waste and repository. Three scenario-groups of once-through, Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. By construction of two full MOX- or ROX- reactors, Pu amount is reduced to about a half and isotopic vector of Pu is deteriorated as nuclear weapon especially in terms of spontaneous fission neutron. Effects by ROX are more significant than MOX in both amount and vector. Repository footprint and potential radio-toxicity is not reduced by MOX and ROX because heat and toxicity of MOX and ROX spent fuel is considerably high.

Journal Articles

Evaluation of neutron economical effect of new cladding materials in light water reactors

Oizumi, Akito; Akie, Hiroshi; Iwamoto, Nobuyuki; Kugo, Teruhiko

Journal of Nuclear Science and Technology, 51(1), p.77 - 90, 2014/01

 Times Cited Count:1 Percentile:85.45(Nuclear Science & Technology)

Journal Articles

Simple formula to evaluate helium production amount in fast reactor MA-containing MOX fuel and its accuracy

Akie, Hiroshi; Sato, Isamu; Suzuki, Motoe; Serizawa, Hiroyuki; Arai, Yasuo

Journal of Nuclear Science and Technology, 50(1), p.107 - 121, 2013/01

 Times Cited Count:1 Percentile:84.29(Nuclear Science & Technology)

A simple formula is developed for the evaluation of the helium production amount in the fast reactor fuel. For the subroutine use in the existing fuel behavior analysis code, the formula is designed putting emphasis on simplicity and quickness rather than accuracy. The accuracy of the formula is confirmed by comparing with the detailed calculation with SWAT code, and also with the post irradiation examination (PIE) results of the fuel pin irradiated at the experimental fast reactor JOYO. As a result, the formula is found to evaluate the helium amount with the difference of less than about 10% from the detailed calculation and from the PIE results. Based on these results, the formula is installed in the fuel behavior analysis code for the simulation of helium behavior in fast reactor fuels.

Journal Articles

Fundamental research on behavior of helium in MA-bearing oxide fuel

Arai, Yasuo; Serizawa, Hiroyuki; Nakajima, Kunihisa; Takano, Masahide; Sato, Isamu; Katsuyama, Kozo; Akie, Hiroshi; Suzuki, Motoe; Shirasu, Noriko; Haga, Yoshinori; et al.

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12

High amount of He is generated in MA-bearing fuel during irradiation and storage periods compared with that in U or U-Pu fuel. Laboratory scale experiments, post irradiation examinations and modeling study were carried out in order to understand the He behavior in MA-bearing oxide fuel. Diffusion characteristics of He in single-crystal UO$$_{2}$$ were investigated by the Knudsen effusion mass spectrometry. Effects of the He accumulation on lattice and bulk expansion of oxide pellets were examined by use of alpha-decay of $$^{244}$$Cm. Post irradiation examinations of 0.5%Am-MOX fuel irradiated at a fast test reactor JOYO were carried out, concentrating on the He behavior in the fuel pellets. A model describing the He behavior in MA-MOX fuel was constructed based on the principle processes, such as generation, diffusion, equilibrium and release to outer gaseous phase. By use of the model as a subroutine of a conventional fuel behavior analysis code, the He behavior in MA-MOX fuel for fast reactors was simulated.

Journal Articles

Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/05

JAEA Reports

Research on high conversion type FLWR (HC-FLWR) core

Nakano, Yoshihiro; Fukaya, Yuji; Akie, Hiroshi; Ishikawa, Nobuyuki; Okubo, Tsutomu; Uchikawa, Sadao

JAEA-Research 2009-061, 92 Pages, 2010/03

JAEA-Research-2009-061.pdf:9.5MB

A series of research on a high conversion type innovative water reactor for flexible fuel cycle (FLWR) has been conducted. This FLWR is a boiling water reactor (BWR) with a tight triangular fuel rod lattice and the uranium plutonium mixed oxide (MOX) fuel. FLWR is designed for two types of cores to be developed in succession. The preceding core is a high conversion type FLWR (HC-FLWR) and the other core is Reduced Moderation Water Reactor (RMWR) of which the conversion ratio is more than 1.0. Three design studies and a senario study on HC-FLWR are presented in this report. The first design study is for a representative core. The second one is for a transition core from HC-FLWR to RMWR. In the transition core, both assemblies for HC-FLWR and RMWR exist. The third one is for a core to recycle minor actinides (MAs). Regarding to the scenario study, based on design results of the representative core, effective plutonium utilization in future LWR was considered within general framework.

Journal Articles

Design study of nuclear power systems for deep space explorers, 1; Criticality of low enriched uranium fueled core

Kugo, Teruhiko; Akie, Hiroshi; Yamaji, Akifumi; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9371_1 - 9371_8, 2009/05

Combining a nuclear reactor with thermoelectric converters is expected to be one of promising options to supply a propulsion power for deep space explorers. One of the key features of the concept is to use low enriched uranium fuels from the viewpoint of nuclear non-proliferation. Fuels of uranium oxide, nitride and metal were examined. Zirconium and yttrium hydrides, beryllium, zirconium beryllide and graphite were considered as moderators. Reflectors of beryllium, beryllium oxide, zirconium beryllide and graphite were taken into consideration. A criticality survey of the core was performed by changing the ratio of the fuel, moderator and structure, and the reflector thickness. As a result from the viewpoint of a smaller mass of reactor, it is better to use thermal spectrum cores than fast ones, and the metal hydride moderators than beryllium or graphite. For example, a combination of uranium nitride, yttrium hydride and beryllium reflector achieves a reactor mass of as low as 500kg.

Journal Articles

Neutronic characteristics of FLWR in the transition phase changing from high conversion core to breeder core

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9304_1 - 9304_9, 2009/05

Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation water reactor which can realize Pu breeding and multiple recycling. And for the introduction stage, a high conversion (HC) type FLWR is also proposed to keep technical continuity from current LWRs. When the HC type core is shifted to the breeder (BR) core, there exist both types of fuel assemblies in the same core configuration. The power distribution in the HC + BR assemblies mixed core configuration is studied, because there might appear a power peaking in the adjacent region between HC and BR assemblies due to the difference in neutron spectrum. As a result, though a power peaking can be very large in the adjacent regions between the assemblies, the power distribution can be effectively flattened by considering a rod-wise fuel enrichment distribution and by optimizing the fuel assembly loading pattern. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties.

JAEA Reports

Core design of high conversion type FLWR

Nakano, Yoshihiro; Akie, Hiroshi; Okumura, Keisuke; Okubo, Tsutomu; Uchikawa, Sadao

JAEA-Research 2008-006, 37 Pages, 2008/03

JAEA-Research-2008-006.pdf:35.13MB

A core design of a high conversion type innovative water reactor for flexible fuel cycle (HC-FLWR) with thermal output of 3926 MW has been constructed. The design study of HC-FLWR consists of two steps of analyses. The first step was preliminary parametric survey calculations and the second step was more detailed calculations with a nuclear and thermal-hydraulic coupled calculation code MOSRA. Through the 1-D core burnup calculations, the following design values were obtained. The Puf enrichment of MOX fuel is 11%. The heights of upper blanket, MOX and lower blanket regions are 5 cm, 85 cm and 5 cm, respectively. With these values, 3-D core burnup calculations were performed. In this analysis, effects of the fuel loading pattern were also investigated. Finally, a neutronics design of HC-FLWR core with a negative void reactivity coefficient, a conversion ratio of 0.84 and a discharged burnup of 56 GWd/t was obtained.

Journal Articles

Conceptual design study on high conversion type core of FLWR

Nakano, Yoshihiro; Akie, Hiroshi; Okubo, Tsutomu; Uchikawa, Sadao

Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 9 Pages, 2007/05

no abstracts in English

Journal Articles

Conceptual design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

Okubo, Tsutomu; Uchikawa, Sadao; Nakano, Yoshihiro; Akie, Hiroshi; Kobayashi, Noboru; Fukaya, Yuji

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04

For the future sustainable energy supply based on the matured LWR technologies, a concept of FLWR has been studied in JAEA. The concept utilizes the tight-lattice core loaded with the MOX fuel, and consists of two steps. The first step realizes a high conversion type one (HC-FLWR) to keep the smooth technical continuity from the LWR technologies. The second is the RMWR concept, which realizes a high conversion ratio over 1.0 for Pu multiple recycling. The key point is that the two core concepts utilize the same size fuel assemblies, and hence, the former can proceed to the latter in the same reactor system based flexibly on the future fuel cycle circumstances. In the present paper, investigation results on the FLWR conceptual design are presented. The design of the HC-FLWR core has been recently improved, and detailed core properties have been evaluated by the neutronics and thermal-hydraulics coupled calculations. The core can achieve the average burn-up around 55GWd/t.

Journal Articles

Conceptual design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) and its recycle characteristics

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Takeda, Renzo*; Nakano, Yoshihiro; Onuki, Akira; Iwamura, Takamichi

Journal of Nuclear Science and Technology, 44(3), p.277 - 284, 2007/03

 Times Cited Count:24 Percentile:13.14(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

Iwamura, Takamichi; Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Nakatsuka, Toru

Nuclear Engineering and Design, 236(14-16), p.1599 - 1605, 2006/08

 Times Cited Count:17 Percentile:19.14(Nuclear Science & Technology)

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming Mixed Oxide (MOX)-LWR technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Agency (JAEA). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming MOX-LWR technologies without significant technical gaps. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-developed LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the future fuel cycle circumstances during the reactor operation period around 60 years. Investigation on the core for both the parts of the FLWR concepts has been performed, including the core conceptual design, the core characteristics under Pu multiple recycling, the thermal hydraulic investigation in the tight-lattice core, and so forth. Up to the present, promising results have been obtained.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle(FLWR)

Okubo, Tsutomu; Uchikawa, Sadao; Kugo, Teruhiko; Akie, Hiroshi; Iwamura, Takamichi

Proceedings of 3rd Asian Specialist Meeting on Future Small-sized LWR Development, p.9_1 - 9_12, 2005/11

A concept of Innovative Water Reactor for Flexible Fuel Cycle(FLWR) has been investigated in JAEA in order to ensure sustainable energy supply in the future based on the well-developed LWR technologies. The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two steps. In the first step, FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second step represents the RMWR core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the LWR technologies. The key point is that the core concepts in both steps utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology for the MOX spent fuel. Detailed investigation have been performed on the core design, in conjunction with the other related studies such as on the thermal hydraulics in the tight-lattice core including the experimental activities, and the results obtained so far have shown that the proposed concept is feasible and promising. For commercial realization of the FLWRs in 2030s, a 400MWe class small reactor is proposed to be constructed in 2010s as a leading demonstration plant.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 2; Recycle characteristics

Okubo, Tsutomu; Uchikawa, Sadao; Kugo, Teruhiko; Akie, Hiroshi; Takeda, Renzo*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

In order to ensure sustainable energy supply in the future based on the commercialized LWR technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in JAERI. Results on the FLWR recycling characteristics under possible various reprocessing schemes are presented in the present paper. The results show the recycling is possible a few times at most as long as the fissile Pu content stays over 60%, even in the high conversion type core with the conversion ratio around 0.9, under the simplified PUREX reprocessing, with relatively high average decontamination factor. For breeding core, the results have indicated that even under the reprocessing with relatively low DFs and with whole MA, the recycling is also feasible, suggesting all MAs from the core can be possibly recycled itself, although the core performances are a little degraded depending on MA and FP contents.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 1; Conceptual design

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Onuki, Akira; Iwamura, Takamichi

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

no abstracts in English

JAEA Reports

Summary report of the 7th Reduced-Moderation Water Reactor Workshop; March 5, 2004, JAERI, Tokai

Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

JAERI-Conf 2005-009, 153 Pages, 2005/08

JAERI-Conf-2005-009.pdf:14.7MB

As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI. The workshop on RMWRs is aiming at information exchange between JAERI and other organizations, and has been held every year since 1998. The program of the 7th workshop was composed of 5 lectures and an overall discussion time. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture as well as of the discussion time. In addition in Appendix, there are included presentation handouts of each lecture.

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