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Journal Articles

Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04

The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

Journal Articles

Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

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

The subchannel analysis code NASCA was applied to critical power prediction of 37-rod tight-lattice bundle experiments which JAERI has been carrying out to confirm the thermal-hydraulic feasibility of the RMWR. The NASCA can yield good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy of critical power deteriorated in case of the gap width of 1.0 mm. Predicted BT positions agree with the experimental results. Models in the code will be improved to consider the effect of the gap width based on further studies in the future.

Journal Articles

Critical power prediction for tight lattice rod bundles

Liu, W.; Onuki, Akira; Tamai, Hidesada; Akimoto, Hajime

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

In this research, the newest version of critical power correlation for tight-lattice rod bundles is proposed by using 7-rod and 37-rod bundle data derived in Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux - critical quality type. For low mass velocity region, it is written in critical quality - annular flow length type. The correlation is verified by JAERI data and Bettis Atomic Power Laboratory data. It is confirmed the correlation is able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The correlation is further implemented into TRAC code to analyze flow decrease and power increase transients. It is confirmed transient BT can be predicted within the accuracy of the implemented critical power correlation.

JAEA Reports

Mechanical characteristics evaluation of fuel cladding tube for reduced-moderation water reactor, 1 (Contract research)

Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi

JAERI-Research 2005-005, 23 Pages, 2005/03

JAERI-Research-2005-005.pdf:1.65MB

Fuel elements used in The Reduced-Moderation Water Reactor (RMWR) have the lamellar structure consisting of MOX pellets and UO$$_{2}$$ blankets in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod. Therefore, it is important to evaluate the local deformation behavior due to the transient temperature distribution. To estimate the thermal deformation behavior, the temperature and stress distribution of the fuel cladding tube assumed in the designed reactor were analyzed. Moreover, basic physical properties and mechanical properties for analyzing the deformation behavior were obtained by experiment using fuel cladding tubes made of candidate alloys. In addition, the appropriate experimental conditions for realizing the practical thermal deformation behavior of the fuel cladding tube was selected by adjusting the testing temperature distribution based on data obtained with thermal analysis.

Journal Articles

Critical power correlation for tight-lattice rod bundles

Liu, W.; Kureta, Masatoshi; Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 42(1), p.40 - 49, 2005/01

 Times Cited Count:7 Percentile:45(Nuclear Science & Technology)

In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For low mass velocity region ($$<$$ 300 kg/m$$^{2}$$s), the correlation is written in critical quality - annular flow length type. For high mass velocity region ($$>$$ 300 kg/m$$^{2}$$s), it is written in local critical heat flux - critical quality type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/m$$^{2}$$s and pressure from 2 to 11 MPa.

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large scale numerical simulation, 1; Development of a direct analysis procedure on two-phase flow with an advanced interface tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09

When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.

JAEA Reports

Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.

JAERI-Research 2004-008, 383 Pages, 2004/06

JAERI-Research-2004-008.pdf:21.49MB

The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.

Journal Articles

Critical power in 7-rod tight lattice bundle

Liu, W.; Kureta, Masatoshi; Akimoto, Hajime

JSME International Journal, Series B, 47(2), p.299 - 305, 2004/05

Experimental research on critical power in tight lattice bundle that simulates the Reduced-Moderation Water Reactor (RMWR) has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3 mm equilateral triangular pitch. Each rod is 13 mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under ${it G}$ = 100-1400 kg/m$$^{2}$$s, ${it P}$$$_{ex}$$ = 2-8.5 MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.

JAEA Reports

Development of testing techniques to evaluate thermal deformation behavior of fuel cladding tubes (Contract research)

Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi

JAERI-Tech 2004-035, 18 Pages, 2004/03

JAERI-Tech-2004-035.pdf:0.81MB

Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO$$_{2}$$ blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO$$_{2}$$ blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.

JAEA Reports

Neutronic study on seed-blanket type reduced-moderation water reactor fuel assembly

Shelley, A.; Kugo, Teruhiko; Shimada, Shoichiro*; Okubo, Tsutomu; Iwamura, Takamichi

JAERI-Research 2004-002, 47 Pages, 2004/03

JAERI-Research-2004-002.pdf:3.08MB

Neutronic study has been done for a PWR-type reduced-moderation water reactor with seed-blanket fuel assemblies to achieve a high conversion ratio, a negative void coefficient and a high burnup by using a MOX fuel. The results of the precise assembly burnup calculations show that the recommended numbers of seed and blanket layers are 15(S15) and 5(B5), respectively. By the optimization of axial configuration, the S15B5 assembly with the seed of 1000$$times$$2 mm high, internal blanket of 150 mm high and axial blanket of 400$$times$$2 mm high is recommended. In this configuration, the conversion ratio is 1.0 and the core average burnup is 38 GWd/t. The S15B5 assembly can attain the core average burnup of 45 GWd/t by decreasing the height of seed to 500$$times$$2 mm, however, the conversion ratio becomes 0.97. The void and fuel temperature coefficients are negative for both of the configurations. Effect of metal or T-MOX (PuO$$_{2}$$+ThO$$_{2}$$) fuel has been also investigated. Metal improves the conversion ratio but makes the void coefficient worse. T-MOX improves the void coefficient, but decreases the conversion ratio.

JAEA Reports

Summary of the 6th Workshop on the Reduced-Moderation Water Reactor; March 6, 2003, JAERI, Tokai

Nabeshima, Kunihiko; Nakatsuka, Toru; Ishikawa, Nobuyuki; Uchikawa, Sadao

JAERI-Conf 2003-020, 240 Pages, 2003/11

JAERI-Conf-2003-020.pdf:27.66MB

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since 1998 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The workshop began with five lectures on status of research and development on RMWRs in JAERI entitled "Status and Future Program of Research and Development on Reduced-Moderation Water Reactors", "Design of Small Reduced-Moderation Water Reactors", "Critical Experiments for Reduced-Moderation Water Reactors", "Critical Heat Flux Experiments in Tight Lattice Core" and "Development of High Performance Cladding". Then two lectures followed: "Status of Phase II of Feasibility Studies on Commercialized Fast Breeder Reactor System" by JNC and "Present Status of Study on Super-critical water Cooled Power Reactor" by Toshiba Corporation.

Journal Articles

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Shelley, A.; Shimada, Shoichiro*; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi

Nuclear Engineering and Design, 224(3), p.265 - 278, 2003/10

 Times Cited Count:14 Percentile:66.99(Nuclear Science & Technology)

Parametric studies have been done for a PWR-type reduced-moderation water reactor (RMWR) with seed-blanket fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup. It was found that 50 to 60% of seed in a seed-blanket assembly has higher conversion ratio. The number of seed-blanket layers is 20, in which the number of seed layers is 15 and blanket layers is 5. The fuel assembly with the height of seed of 1000mm$$times$$2, internal blanket of 150 mm and axial blanket of 400mm$$times$$2 is recommended. The conversion ratio is 1.0 and the average burnup in core region is 38.2 GWd/t. The enrichment of fissile Pu is 14.6 wt%. The void coefficient is +21.8 pcm/% void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account. It is also possible to use this fuel assembly for a high core averaged burnup of 45GWd/t, however, the height of seed must be 500mm$$times$$2 to improve the void coefficient. The conversion ratio is 0.97 and void coefficient is +20.8 pcm/%void.

Journal Articles

Designing of reduced-moderation water reactor and related development issues

Okubo, Tsutomu

Konsoryu, 17(3), p.228 - 235, 2003/09

The Reduced-Moderation Water Reactor is supposed to realize plutonium multiple recycling, and furthermore, plutonium breeding cycle, based on the well-established Light Water Reactor technologies. In the present paper, the overview of the design study is presented and the related R&D issues are introduced, especially focusing on the thermal hydraulic activities.

Journal Articles

Subchannel analysis of CHF experiments for tight-lattice core

Nakatsuka, Toru; Tamai, Hidesada; Kureta, Masatoshi; Okubo, Tsutomu; Akimoto, Hajime; Iwamura, Takamichi

Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 6 Pages, 2003/09

It is important to evaluate thermal margin of the tight lattice core in the Reduced-Moderation Water reactor (RMWR). In the present study, to assess the applicability of subchannel analysis for tight lattice cores, tight lattice CHF experiments were analyzed with COBRA-TF code. For the axial uniform heated rod bundle, the code gives good prediction of critical power for mass velocity of around 500kg/(m$$^{2}$$s), while the code underestimates it for lower mass velocity and overestimates for higher mass velocity. The predicted BT position was outer channels and differed from the measured position. For the axially double-humped heated bundle, the code gives good prediction for mass velocity of around 200kg/(m$$^{2}$$s), and overestimates for higher mass velocity. It turned out that the two-phase multiplier of friction loss have a large influences on the flow distribution among the subchannels. To improve the calculation accuracy, it is required to predict precisely the flow distribution including the prediction of pressure distribution in a tight lattice bundle.

Journal Articles

Development of predictable technology for thermal/hydraulic performance of reduced-moderation water reactors, 1; Master plan

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime

Nihon Kikai Gakkai 2003-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.247 - 248, 2003/08

We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with power company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. This series presentation focuses on the feasibility study and shows the R&D plan using large-scale test facility and advanced numerical simulation technology.

JAEA Reports

CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu; Kureta, Masatoshi; Yoshida, Hiroyuki; Ishikawa, Nobuyuki; Sato, Takashi; Watanabe, Hironori; Okubo, Tsutomu; Iwamura, Takamichi; et al.

JAERI-Research 2002-018, 37 Pages, 2002/10

JAERI-Research-2002-018.pdf:2.62MB

no abstracts in English

JAEA Reports

Summary of the 4th Workshop on the Reduced-Moderation Water Reactor; March 2, 2001, JAERI, Tokai

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

JAERI-Conf 2001-013, 263 Pages, 2001/09

JAERI-Conf-2001-013.pdf:13.37MB

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The 4th workshop on the RMWRs was held on March 2, 2001 aimed at information exchange between JAERI and other organizations. This report includes the eight original papers presented at the workshop and summaries of the questions and answers for each lecture. Titles of the papers are“Recent Situation of Research on Reduced-Moderation Water Reactor“, “Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors“,“Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR“,“Micro Reactor Physics of MOX Fueled LWR“, “Fast Reactor Cooled by Supercritical Light Water “, “Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System“, “Integral Type Small PWR with Stand-alone Safety“ and “Utilization of Plutonium in Reduced-Moderation Water Reactors“.

Journal Articles

Advanced concept of Reduced-Moderation Water Reactor (RMWR) for plutonium multiple recycling

Okubo, Tsutomu; Takeda, Renzo*; Iwamura, Takamichi; Yamamoto, Kazuhiko*; Okada, Hiroyuki*

Proceedings of International Conference on Back-End of the Fuel Cycle: From Research to Solutions (GLOBAL 2001) (CD-ROM), 7 Pages, 2001/09

An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated and confirmed for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle.

Journal Articles

Conceptual designing of a reduced moderation pressurized water reactor by use of MVP and MVP-BURN

Kugo, Teruhiko

Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications, p.821 - 826, 2001/00

no abstracts in English

JAEA Reports

Research on reduced-moderation water reactor (RMWR)

Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro*; Usui, Shuji*; Shirakawa, Toshihisa*; Nakatsuka, Toru; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Wada, Shigeyuki*

JAERI-Research 99-058, p.61 - 0, 1999/11

JAERI-Research-99-058.pdf:3.3MB

no abstracts in English

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