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Yoshida, Hiroyuki; Nagatake, Taku; Takase, Kazuyuki; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 10 Pages, 2013/07
Yamashita, Susumu; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 8 Pages, 2013/07
In accidents of the Fukushima Daiichi Nuclear Power Plants, by stop of the emergency core cooling system, fuel rods were overheated due to the radioactive decay heat and the oxidization of fuel cladding. It is considered that the core degradation occurred because fuels, control rods, and other components in a reactor vessel was melted and relocated. In order to estimate progress of the degradation phenomena in the reactor core, a numerical simulation code that can precisely evaluate the melting phenomena is required. Therefore a numerical simulation method for predicting the melting core behavior including solidification and relocation based on the three-dimensional multi-phase thermal-hydraulic simulation code has been developed in JAEA. From the present numerical results, it was confirmed that relocation of molten debris in the BWR lower plenum can be simulated by the currently developed code including effects of melting and solidification of debris.
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 10 Pages, 2013/07
This paper describes experimental analyses using SIMMER-III/IV, which are two/three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of the SIMMER-III/IV code validation. Two topics of key phenomena in core disruptive accidents were presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment.
Cheng, S.; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Nakamura, Yuya*; Takeda, Shohei*; Nishi, Shimpei*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 8 Pages, 2013/07
Irwanto, D.; Satou, Akira; Takeda, Takeshi; Nakamura, Hideo
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07
Takeda, Takeshi; Nakamura, Hideo
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 10 Pages, 2013/07
Wakasugi, Keiichiro; Nakajima, Kunihiko*; Shimemoto, Hidenori; Shibata, Masahiro; Yamaguchi, Masaaki
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 9 Pages, 2013/07
In Japan, uplift and erosion scenarios must be analysed since this natural phenomena would be inevitable at most sites in Japan. It's increasingly important to enhance the confidence of the assessment for the uplift and erosion scenarios, as no assessment cut-off times have yet been defined. In this context, this study carried out bounding analysis to find out parameter conditions to satisfy hypothetical dose criteria. The results show that there are no cases that satisfy 10 microSv/y. However, all cases are below 300 microSv/y. The discussion also implies that to accelerate the release from the EBS for minimising the dose in later phase is inadequate and ineffective at all, due to multiple barriers and multiple safety functions. In principal, the influence of uplift and erosion should be reduced by appropriate site selection and design as much as possible to ensure the sufficient nuclides decay while the repository is staying at the deep underground.
Ueta, Shohei; Inoi, Hiroyuki; Mizutani, Yoshitaka; Ohashi, Hirofumi; Iwatsuki, Jin; Sakaba, Nariaki; Sawa, Kazuhiro
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 4 Pages, 2013/07
Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.
Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07
Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under tube failure accident in a steam generator of sodium cooled fast reactors. The system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM codes calculates multicomponent multiphase flow involving sodium-water chemical reaction. In this study, a numerical model for chemical reaction about production of a sodium monoxide and its transport process were constructed. We also developed the numerical models of the TACT code for evaluation of shell-side flow around an adjacent tube, heat transfer from the fluid to the tube and occurrence of tube failure. In our analysis system, thermal hydraulic behavior of water inside the tube is evaluated by the RELAP5 code. The original heat transfer correlations were corrected for the rapidly heated tube in the present work.
Liu, W.; Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Nagase, Fumihisa
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 5 Pages, 2013/07
In the Fukushima Daiichi Nuclear Power Plant accident, seawater was injected into the reactors to cool down nuclear fuels. So far, core cooling with seawater has never been assumed and the effect of seawater on heat transfer in core is not clear. Then, effects of seawater on thermal-hydraulic behavior must be investigated to understand the phenomena occurred in the accident and to evaluate current state of the reactor cores. In this series of research work, the effects of seawater on thermal-hydraulic behavior before and after degradation of the cores will be researched experimentally. And experimental results will be incorporated to numerical simulation codes to evaluate effects of seawater on the Fukushima Daiichi Nuclear Power Plant accident. In the experimental research part, we have a plan performing two heat transfer experiments to evaluate thermal hydraulic performance of sea water and effects of salt precipitation. A preliminary experiment is performed to obtain basic information required for the experiments with salt precipitation. In this paper, outline of the research plan is explained and results of preliminary experiment is shown.
Tagawa, Akihiro; Tezuka, Masashi; Terakura, Yoshihiro*; Naito, Masayuki*; Miyajima, Kenji*
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07
It is one of the most urgent issues to remediate the nuclear power plants contaminated by radioactive materials discharged following the accident at the TEPCO's Fukushima Daiichi NPS. Concrete walls of nuclear power plants in Japan are coated an epoxy resin coating for easily performing decontamination. We experimented a cutting test in Fugen Decommissioning Engineering Center using maximum 280 MPa pressure and 30 L/min water quantity, ultra-high pressure water jet system and 40 m
/min air quantity vacuum system. We are conducting a study of decontamination technology in environmental pollution using this decontamination system. This decontamination system has achieved a decontamination factor 10 to 100. Thus, we have confirmed the change in a cutting ability by changing parameters. Parameters are the water pressure, the water quantity and air quantity. The impact force for water jet is a function that contains the test parameters. We have considered it using this function. The results of the test showed that there is a correlation between the impact force for water jet and a cutting capability. This decontamination technology can decontaminate radioactive material of the surface adhesion contamination and reduce the amount of waste generated for a thin cutting. In addition, we have experimented that the water can be recycled by chemical precipitation. After we experimented flocculation test using aluminum sulfate and zeolite flocculant, we have confirmed that it can clean water up to the level of suspended solids 5 mg/L, in turbid water using zeolite flocculant. This suspended solids concentration can be passed to the water processing system in nuclear power plant. From the test results, we found that ultra-high pressure water jet decontamination technology has a possibility that it can be used for decontamination of Fukushima Daiichi Nuclear Power Plant.
Horiguchi, Naoki; Yoshida, Hiroyuki; Uesawa, Shinichiro*; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07
Ote, Naosuke*; Koizumi, Yasuo*; Kamide, Hideki; Ohno, Shuji; Ito, Kei
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 7 Pages, 2013/07
A sodium-cooled fast breeder reactor is now at the developing stage in Japan. One concern for safety is cover gas entrainment into the sodium coolant. The gas entrainment rate into liquid by the vortex formed on the free surface was examined experimentally. Liquid flowed into a cylindrical vessel from a wall tangentially. Swirl flow was formed in the vessel, and then liquid drained from the bottom outlet of the vessel. A hollow vortex was formed on the free surface in the test vessel. Air was entrained under the free surface of the vortex and carried away from the bottom of the vessel. The flow state of the gas entrainment was visually observed by using a high speed video camera. The gas entrainment rate into liquid was measured. In the previous study, test fluid was water. Kerosene and 20 cSt silicone oil were newly introduced as the test fluid to examine the effect of the physical properties on the gas entrainment phenomena.
