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Journal Articles

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

Journal Articles

Overview of Japanese development of accident tolerant FeCrAl-ODS fuel claddings for BWRs

Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 7 Pages, 2017/09

This paper will show the overview of current status of development of accident tolerant FeCrAl-ODS fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. This program is being carried out to create the technical basis for the practical use of the accident tolerant fuels and the other components in LWRs through multifaceted activities. In the development of FeCrAl-ODS fuel claddings both the experimental and the analytical studies have been performed. The acquisition and accumulation of key material properties of FeCrAl-ODS fuel claddings were conducted by using bar, sheet and tube shaped FeCrAl-ODS materials fabricated in this program to support the evaluations in the analytical studies. A neutron irradiation test was also started in the ORNL High Flux Isotope Reactor (HFIR) to examine the effect of neutron irradiation on the mechanical properties.

Journal Articles

FEMAXI-7 prediction of the behavior of BWR-type accident tolerant fuel rod with FeCrAl-ODS steel cladding in normal condition

Yamaji, Akifumi*; Yamasaki, Daiki*; Okada, Tomoya*; Sakamoto, Kan*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

Features of the accident tolerant fuel performance were evaluated with FEMAXI-7 when the current Zircaloy(Zry) cladding is replaced with FeCrAl-ODS steel cladding (a type of oxide dispersion strengthened steel being developed under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan) for BWR 9$$times$$9 type fuel rod. In particular, influences of the creep strain rate and thickness of the ODS cladding on the fuel temperature, fission gas release rate (FGR) and pellet-cladding mechanical interaction (PCMI) are investigated.

Journal Articles

Evaluation of large 3600 MWth sodium-cooled fast reactor OECD neutronic benchmarks

Buiron, L.*; Rimpault, G*; Fontaine, B.*; Kim, T. K.*; Stauff, N. E.*; Taiwo, T. A.*; Yamaji, Akifumi*; Gulliford, J.*; Fridmann, E.*; Pataki, I.*; et al.

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09

Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD, an international collaboration is ongoing on the neutronic analyses of several Generation-IV Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes the results obtained by participants from institutions of different countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for the large core numerical benchmarks. These results have been obtained using different calculation methods and analysis tools to estimate the core reactivity and isotopic composition evolution, neutronic feedbacks and power distribution. For the different core concepts analyzed, a satisfactory agreement was obtained between participants despite the different calculation schemes used. A good agreement was generally obtained when comparing compositions after burnup, the delayed neutron fraction, the Doppler coefficient, and the sodium void worth. However, some noticeable discrepancies between the k-effective values were observed and are explained in this paper. These are mostly due to the different neutronic libraries employed (JEFF3.1, ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations methods.

Journal Articles

Summary of OECD/NEA/NSC expert group on integral experiments for minor actinide management

Okajima, Shigeaki; Fougeras, P.*; Gil, C.-S.*; Glinatsis, G.*; Gulliford, J.*; Iwamoto, Osamu; Jacqmin, R.*; Khomyakov, Y.*; Kochetkov, A.*; Kormilitsyn, M. V.*; et al.

NEA/NSC/DOC(2013)3, p.265 - 278, 2013/04

The Expert Group on Integral Experiments for Minor Actinide Management (EG on IEMAM) was established under OECD/NEA/NSC. The objectives are to review integral experiments for validating MA nuclear data, to recommend additional integral experiments and to propose an international framework to facilitate them from view points of the MA management. The paper summarized the discussion results in the EG on IEMAM as follows: (1) Requirement of nuclear data for MA management, (2) Reviewing existing integral data and identifying specification of missing experimental work to be required, (3) Identifying the bottlenecks and considering possible solutions to them and (4) Proposal of action program for international cooperation.

JAEA Reports

Fuel and core design studies on metal fuel sodium-cooled fast reactor (4), (5) and (6); Joint research report for JFY2009 - 2012

Uematsu, Mari Mariannu; Sugino, Kazuteru; Kawashima, Katsuyuki; Okano, Yasushi; Yamaji, Akifumi; Naganuma, Masayuki; Oki, Shigeo; Okubo, Tsutomu; Ota, Hirokazu*; Ogata, Takanari*; et al.

JAEA-Research 2012-041, 126 Pages, 2013/02

JAEA-Research-2012-041.pdf:16.49MB

The characteristics of sodium-cooled metal fuel core compared to MOX fuel core are given by its higher heavy metal density and superior neutron economy. By taking advantage of these characteristics and allowing flexibility in metal fuel specification and core design conditions as sodium void reactivity and bundle pressure drop, core design with high burnup, high breeding ratio and low fuel inventory features will be achievable. On ground of the major achievements in metal fuels utilization as driver fuels in sodium fast reactors in U.S., the metal fuel core concept is selected as a possible alternative of MOX fuel core concept in FaCT project. This report describes the following items as a result of the joint study on "Reactor core and fuel design of metal fuel core of sodium-cooled fast reactor" conducted by JAEA and CRIEPI during 4 years from fiscal year 2009 to 2012.

Journal Articles

Design study to increase plutonium conversion ratio of HC-FLWR core

Yamaji, Akifumi; Nakano, Yoshihiro; Uchikawa, Sadao; Okubo, Tsutomu

Nuclear Technology, 179(3), p.309 - 322, 2012/09

 Times Cited Count:5 Percentile:48.35(Nuclear Science & Technology)

HC-FLWR effectively utilizes the uranium (U) and the plutonium (Pu) resources by achieving a fissile Pu conversion ratio of 0.84 without a significant technical gap from the current BWR technology. In this study, a new core design concept for HC-FLWR has been developed to achieve the conversion ratio of 0.95. The concept of the FLWR/MIX fuel assembly, which had been originally proposed for tight fuel bundle, was used to raise the conversion ratio without deteriorating the core void reactivity characteristics. For a semi-tight fuel rod lattice with rod clearance of 0.20 to 0.25 cm, the design ranges of the conversion ratio and the average discharge burnup are 0.91 to 0.94 and 53 to 49 GWd/t, respectively. The conversion ratio can be raised to 0.97 by increasing the $$^{235}$$U enrichment from 4.9 to 6.0 wt%. Two representative core designs and one alternative design option have been obtained. Hence, the flexibility of HC-FLWR concept to achieve the conversion ratio of 0.84 to 0.95 has been revealed.

JAEA Reports

FEMAXI-6 code verifications for predicting FLWR MOX fuel rod behaviors

Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu

JAEA-Research 2010-029, 54 Pages, 2010/09

JAEA-Research-2010-029.pdf:3.07MB

This study has evaluated uncertainties in FEMAXI-6 calculations and clarified key models and parameters for predicting LWR MOX fuel rod behavior. Irradiation data obtained from the Halden reactor experiments (IFA-597.4 rod-10, rod-11, and IFA-514 rod-1) were used. The maximum discharge burnup was about 40 GWd/tMOX (IFA-514 rod-1). The results showed that uncertainties in fission gas release calculations were particularly high, and contributions of pellet relocation, densification and swelling models on pellet temperature were also evaluated. The basic fission gas release mechanism of MOX fuels should be the same as that of UO$$_{2}$$ fuels, but the parameters in the model need to be revised for MOX fuels. More experimental data are needed. However, frequent reactor shutdowns and restarts may cause pellet relocation changes which need to be considered in the evaluations.

Journal Articles

The Impact of americium target in-core loading on reactivity characteristics and ULOF response of sodium-cooled MOX FBR

Yamaji, Akifumi; Kawashima, Katsuyuki; Oki, Shigeo; Mizuno, Tomoyasu; Okubo, Tsutomu

Nuclear Technology, 171(2), p.153 - 160, 2010/08

 Times Cited Count:4 Percentile:62.36(Nuclear Science & Technology)

The homogeneous MA loading core with 3wt% MAs is used as a reference design to evaluate the impact of the americium target in-core loading (20wt% MAs) on reactivity characteristics and ULOF response of sodium-cooled MOX-FBR. The Am target loading method of this study can flatten core radial reactivity worth distributions and effectively reduce reactivity insertion into the core during ULOF. As the result, the core power increase speed during ULOF is reduced. The maximum fuel temperature of the target region does not become particularly high compared with that of the inner core and it is much lower than the melting point. It is promising from the viewpoints of the reactivity characteristics and ULOF response.

Journal Articles

FEMAXI-6 code verification with MOX fuels irradiated in Halden reactor

Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu

Journal of Nuclear Science and Technology, 46(12), p.1152 - 1161, 2009/12

The advanced reactor concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed and being studied to achieve effective and flexible utilization of the uranium and plutonium resources based on the well-developed light water reactor (LWR) technology. In order to design and evaluate the FLWR fuel rod behavior, the uncertainties in the FEMAXI-6 calculations and the key models and parameters for predicting LWR MOX fuel rod behavior need to be evaluated. In this study, the Test-Fuel-Data-Base (TFDB) obtained from the Halden reactor experiments (IFA-597.4 rod-10, rod-11, and IFA-514 rod-1) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX.

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

Design study of nuclear power systems for deep space explorers, 2; Electricity supply capabilities of solid cores

Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9366_1 - 9366_8, 2009/05

This study has been carried out in series with the other study, "Criticality of Low Enriched Uranium Fueled Core" to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of three different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The three systems are the solid thermal conduction system (STC), core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and cover down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept.

Journal Articles

Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behavior in FLWR design

Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9092_1 - 9092_9, 2009/05

The concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed and being studied at JAEA to achieve effective and flexible utilization of the uranium and plutonium resources based on the well-developed LWR technology. FLWR is a BWR type concept and it is planned to be introduced by two stages. In the first stage, the MOX fuels are irradiated in a similar condition to that of the current BWR but with a harder neutron spectrum. The core average discharge burnup is about 45 GWd/tHM. In the second stage, the neutron spectrum is further hardened to achieve a multiple recycling of plutonium with higher burnups. In order to design and evaluate the integrities of FLWR fuel rods, the uncertainties in FEMAXI-6 calculations and models for predicting LWR MOX fuel behavior need to be evaluated. As an introduction to the evaluation process, the Test-Fuel-Data-Base (TFDB) obtained from the Halden reactor experiments (IFA-514) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX. Based on the present investigation, the following models were found to be particularly important. Namely the FGR, pellet densification, swelling, and relocation models. These models of FEMAXI-6 have been developed and the parameters have been optimized based on the past UO$$_{2}$$ irradiation test data. For predicting MOX fuel behavior, the FGR model has a relatively large uncertainty and causes a large uncertainty in the FGR calculations. On the other hand, the uncertainties in the other models are within the range expected by the property variations of typical UO$$_{2}$$ fuels. Hence, the densification, swelling, and the relocation models of FEMAXI-6 can be applied to MOX fuel analyses provided the corresponding MOX property variations are taken into account in the input parameters of these models.

Journal Articles

Rationalization of the fuel integrity and transient criteria for the super LWR

Yamaji, Akifumi*; Oka, Yoshiaki*; Ishiwatari, Yuki*; Liu, J.*; Koshizuka, Seiichi*; Suzuki, Motoe

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 7 Pages, 2005/05

Ensuring the fuel integrities is one of the most fundamental parts in the High Temperature Supercritical-Pressure Light Water Reactor. Most abnormal transient events of SCLWR-H last for a short period of time and the fuel rods are replaced after being irradiated in the core. In this study, the fuel integrity criteria are rationalized based on the fact that the fuel rod mechanical failures can be represented by the strain of the fuel rod cladding. A new fuel rod is designed with a Stainless Steel cladding. It is internally pressurized to reduce the stress on the cladding and also to increase the gap conductance between the pellet and the cladding. The fuel integrities both at normal operation and abnormal transient conditions are evaluated using the fuel analysis code FEMAXI-6 of JAERI.

Oral presentation

Evaluation of uncertainties in FEMAXI-6 calculations with MOX fuels irradiated in Halden reactor

Yamaji, Akifumi; Suzuki, Motoe; Okubo, Tsutomu

no journal, , 

FEMAXI-6 is considered to be a powerful tool for developing an advanced LWR concept that utilizes MOX fuel rods, such as FLWR. In order to use this code, verifications with LWR MOX fuel irradiation data is necessary. In this study, the MOX fuel test data obtained from Halden reactor is used to verify FEMAXI-6. Up to about 40 GWd/t, the pellet temperature calculations through the gap conduction calculations are important to predict the fuel rod behavior. Especially, the FGR, pellet relocation, densification, and swelling are influential. These models in FEMAXI-6 have been developed based on UO2 irradiation experience. However, the results show that these models can also be applied to LWR MOX fuel analysis except the FGR model.

Oral presentation

Research and development of FLWR

Yamaji, Akifumi

no journal, , 

An advanced LWR concept (FLWR) is being developed based on the well experienced LWR technologies for ensuring sustainable energy supply in the future. For developing the core concept, a light water fuel rod analysis code, FEMAXI-6 has been verified with MOX fuel rods irradiated at Halden. The results showed that uncertainties in fission gas release calculations were particularly high. The basic fission gas release mechanism of MOX fuels should be the same as that of UO$$_{2}$$ fuels, but the parameters in the model need to be revised for MOX fuels. More experimental data are needed. However, frequent reactor shutdowns and restarts may cause pellet relocation changes and give difficulties in code verifications.

Oral presentation

R&D for introducing advanced fuels contributing to safety improvement of current LWRs, 2; FeCrAl-ODS steels for BWR fuel claddings

Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Ioka, Ikuo; Yamashita, Shinichiro; Kaji, Yoshiyuki

no journal, , 

no abstracts in English

Oral presentation

R&D for introducing advanced fuels contributing to safety improvement of current LWRs, 1; Overview of project

Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Shirasu, Noriko; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; et al.

no journal, , 

Since the accident at Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of Light Water Reactors (LWRs) became a topic of serious discussion, and the research and development (R&D) for improving the safety LWRs has been activated in many countries. In Japan, the R&D project on accident tolerant fuel and other components (ATFs) of LWRs, which is sponsored and organized by the Ministry of Economy, Trade and Industry (METI), has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys in LWRs.

Oral presentation

R&D for introducing advanced fuels contributing to safety improvement of current LWRs, 1; Overview of project

Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Shirasu, Noriko; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Watanabe, Seiichi*; Murakami, Nozomu*; Sato, Hisaki*; et al.

no journal, , 

Since the accident at Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of Light Water Reactors (LWRs) became a topic of serious discussion, and the research and development for improving the safety LWRs has been activated in many countries. This project is one of the projects that entrusted from the Ministry of Economy, Trade and Industry Agency for Natural Resources and Energy. Using a domestic human source, an analysis tool, know-how and experience, the basic technique to apply an accident tolerant fuel material to an existing LWRs is developed in the project. It is started from October, 2015 and is carrying out this project continuously.

Oral presentation

R&D for introducing advanced fuels contributing to safety improvement of current LWRs FeCrAl, 2; FeCrAl-ODS steels for BWR fuel claddings

Sakamoto, Kan*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro; Fukahori, Tokio

no journal, , 

Using the acquired material data by 2016, the feasibility of the reactor core, accident progress relaxation effect at an accident and a severe accident were confirmed by various analysis when Zircaloy replaced to FeCrAl-ODS steel. The results including the post irradiation examination in fiscal year 2017 will be introduced by this presentation.

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