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Doda, Norihiro; Igawa, Kenichi*; Iwasaki, Takashi*; Murakami, Satoshi*; Tanaka, Masaaki
Nuclear Engineering and Design, 410, p.112377_1 - 112377_15, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To enhance the safety of sodium-cooled fast reactors, the decay heat in the core must be removed by natural circulation even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, sodium flow is driven by buoyancy, and flow velocity and temperature distribution influence each other. Thus, it is difficult to evaluate the core hot spot temperature by deterministically considering the uncertainties affecting flow and heat. In this study, a statistical evaluation method is developed for the core hot spot temperature by using Monte Carlo sampling methods. The applicability of the core hotspot evaluation method was confirmed in three representative events during natural circulation decay heat removal operations in loop-type sodium-cooled fast reactors.
CLiu, X.*; Morita, Koji*; Yamano, Hidemasa
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.47 - 51, 2019/09
Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In this study, we performed 1st step of validation analysis using a fast reactor safety analysis code, SIMMER-III, with the developed model based on a new series of experiments, where a BC pellet was immersed into a molten stainless steel (SS) pool. The simulation results showed the general behavior of eutectic material formation measured in the experiments reasonably. The eutectic reaction consumes solid BC and liquid SS, and then the liquid eutectic composition is produced at the early stage of reaction due to the high temperature of molten SS. Movement of the eutectic material in the molten pool leads to the redistribution of boron element. Molten SS pool then freezes to solid SS and movement of eutectic material is stopped by surrounding solid SS. Boron concentration in the pool was measured after molten SS freezes into a solid. Simulation results indicate that boron tends to accumulate in the upper part of the molten pool. This is attributed to the buoyancy force acting on lighter boron in the molten SS pool. A parametric study was also conducted by changing the initial temperature of BC pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.
Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki
Mechanical Engineering Journal (Internet), 4(3), p.16-00495_1 - 16-00495_11, 2017/06
This paper intends to propose a practical solution to protect the HTR from severe oxidation against air ingress accidents without reliance on subsystems. Firstly, a change is made to the center reflector structure to minimize temperature difference during the accident condition in order to reduce buoyancy-driven natural circulation in the reactor. Secondly, a modified structure of the upper reflector is suggested to prevent massive air ingress against a rupture in standpipes. As a preliminary study, a numerical analysis is performed for a typical prismatic-type HTGR. The results showed that amount of air ingress into the reactor can be significantly reduced with practical changes to local structure in the reactor.
Tsujimoto, Kazufumi; Tazawa, Yujiro; Oigawa, Hiroyuki; Sasa, Toshinobu; Takano, Hideki
JAERI-Tech 2003-085, 158 Pages, 2003/11
JAERI-Tech-2003-085.pdf:7.79MB
A safety analysis was performed for the Transmutation Physics Experiment Facility which was to research and develop the reactor physics aspects of the nuclear transmutation technology using the accelerator driven subcritical system. Design policies were evaluated for design of each equipment and system which had important role from view point of safety. Classification of safety class for reactor building, system, and equipment was also reconsidered. Based on the results of safety design policy, acceptance criteria for safety evaluation were reestablished and preliminary analysis were performed. Public exposure by the accident for site appropriateness assessment was evaluated based on revised guidelines in safety evaluation contained in the 1990 Recommendations of ICRP. A recritical event was analyzed by utilizing the newest knowledge for core disruptive accident and calculation code as the beyond design basis accident. The analytical results showed that the isolation capability of the container buildings was ensured against the recritical accident.
Sakurai, Kiyoshi; Nojiri, Ichiro*
JAERI-Conf 2003-019, p.855 - 857, 2003/10
This paper provides overview of sub-criticality safety analysis seminar (July 2000-July 2003, JAERI, total 40 engineers from universities, research institutes and enterprises) for nuclear fuel cycle facility with the Monte Carlo method in Japan. MCNP-4C2 system (MS-DOS version) was installed in each note-type personal computer. Fundamental theory of reactor physics and Monte Carlo simulation including MCNP-4C manual was lectured. Effective neutron multiplication factor and neutron spectrum were calculated for JCO deposit tank, JNC uranium solution storage tank, JNC plutonium solution storage tank and JAERI TCA core. In the seminar, methodology of safety management for nuclear fuel cycle facility was discussed in order to prevent criticality accident.
Nakano, Yoshihiro; Ishikawa, Nobuyuki; Nakatsuka, Toru; Iwamura, Takamichi
JAERI-Conf 2002-012, 219 Pages, 2002/12
JAERI-Conf-2002-012.pdf:17.4MB
no abstracts in English
Kobayashi, Kaoru*; Kaminaga, Masanori; Haga, Katsuhiro; Kinoshita, Hidetaka; Aso, Tomokazu; Hino, Ryutaro
JAERI-Review 2002-010, 52 Pages, 2002/05
JAERI-Review-2002-010.pdf:3.38MB
no abstracts in English
Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro
JAERI-Tech 2002-002, 22 Pages, 2002/02
JAERI-Tech-2002-002.pdf:5.37MB
The Neutron Scattering Facility will be utilized for advanced fields of Material and Life science using high intensity neutron generated by spallation reaction of 1MW pulsed proton beam and mercury target. Design of spallation mercury target system is in progress to obtain high neutron performance with high reliability and safety. The target system is using mercury and contains large amount of radioactive spallation products, therefore to establish the safety of the target system, transient behaviors of the system during anticipated events should be well understood. The safety protection system and an instrumentation system for detecting abnormal conditions must be designed based on the transient behaviors in order to terminate the transient events safely. Transient behaviors of the mercury system have been analyzed by using RELAP5 code. This report presents a test plan of mercury system transient phenomena during abnormal events using a mercury experimental loop and modification of the mercury experimental loop for abnormal transient simulation tests.
Gordon, C.*; Bartels, H.-W.*; Honda, Takuro; Iseli, M.*; Raeder, J.*; Topilski, L.*; Moshonas, K.*; Taylor, N.*
Fusion Engineering and Design, 54(3-4), p.397 - 403, 2001/04
Times Cited Count：2 Percentile：19.66(Nuclear Science & Technology)no abstracts in English
Suzuki, Toru; Tobita, Yoshiharu
JNC TN9400 2000-019, 35 Pages, 2000/03
JNC-TN9400-2000-019.pdf:1.79MB
The transition phase analysis code SIMMER-III has been developed to appropriately evaluate the core disruptive accident in a fast breeder reactor. The momentum exchange model used in the fluid dynamics portion of the code uses the conventional correlation based on ordinary flows such as air-water flows. lt has already been confirmed that this code can represent the experimental results of ordinary flows. However, more detailed research is needed to confirm that this code is applicable to two-phase flow with large liquid density, which would be formed in an actual molten core pool. In addition, since the shapes of bubbles affect their drag in the bubbly flow where the liquid and the gas form continuous and dispersed phases, respectively, it is necessary to take this effect of bubble shape into account to improve SIMMER-III's analytical precision. ln this study, using experimental results obtained through a joint research program with Kyoto university, the momentum exchange model of SIMMER-III is assessed with regard to the bubbly flow regime of two-phase flow with large liquid density, on which experimental data and information on bubble shapes had been lacking. This study suggests that the original SIMMER-III can appropriately represent the characteristics of bubbly flows containing ellipsoidal bubbles with relatively small gas flux. Moreover, this study shows that the precision of SIMMER-III for bubbly flows containing cap bubbles with relatively large gas flux is much improved by using Kataoka-lshii's correlation to determine the drag coefficient of bubbles in the momentum exchange model.
Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro
JAERI-Tech 2000-007, p.21 - 0, 2000/02
JAERI-Tech-2000-007.pdf:1.61MB
no abstracts in English
Ajima, Toshio*; Kurihara, Ryoichi; Seki, Yasushi
JAERI-Data/Code 99-040, 84 Pages, 1999/08
JAERI-Data-Code-99-040.pdf:2.75MB
no abstracts in English
Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*
JNC TN9400 2000-005, 57 Pages, 1999/05
JNC-TN9400-2000-005.pdf:2.92MB
An improved analytic equation-of-state (EOS) model using flexible thermodynamic functions is developed for a reactor safety analysis code, SIMMER-III. The present EOS model is designed to have adequate accuracy in describing thermodynamic properties of reactor-core materials over wide temperature and pressure ranges and to consistently satisfy basic thermodynamic relationships without deterioration of the computing efficiency. The fluid-dynamic algorithm for pressure iteration consistently coupled with the EOS model is also described in the present report. The EOS data of the basic core materials, uranium dioxide, mixed-oxide fuel, stainless steel, and sodium, are developed up to the critical point by compiling the most up-to-date and reliable sources using basic thermodynamic relationships. The thermodynamic consistency and accuracy of the evaluated EOS data are also discussed by comparison with the available sources.
Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*
JNC TN9400 2000-004, 38 Pages, 1999/05
JNC-TN9400-2000-004.pdf:1.11MB
An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms arc designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless stee1, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-corc materials.
Tabata, Toshio; Komukai, Bunsaku; Nagao, Yoshiharu; Shimakawa, Satoshi; Koike, Sumio; Takeda, Takashi; Fujiki, Kazuo
JAERI-Tech 99-021, 68 Pages, 1999/03
no abstracts in English
Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro
JAERI-Tech 99-017, 34 Pages, 1999/03
no abstracts in English
Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro
JAERI-Tech 98-061, 55 Pages, 1999/01
no abstracts in English
Zimin, V. G.; Asaka, Hideaki; Anoda, Yoshinari; *
9th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-9) (CD-ROM), p.16 - 0, 1999/00
no abstracts in English
