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Aoyagi, Mitsuhiro; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 14 Pages, 2014/12
A numerical model for freezing and blockage formation of solid-liquid flow in the SIMMER code was validated in order to improve the accuracy in evaluating fuel discharge behavior in the core disruptive accident of FBR. The THEFIS experiment which investigated fuel discharge behavior was chosen as reference data in this study. The numerical conditions were set according to the experimental system. Although the experimental result was well simulated by using the existing numerical model of SIMMER, the melt flow was suppressed excessively in some cases. Overestimation of flow resistance by the solid particles in the numerical model was discovered though the comparison between the numerical model and the physical phenomenon in the experiment. The numerical model caused the excessive melt flow suppression. Therefore, we improved the numerical model to adapt to the actual phenomenon. Then, it was confirmed the improved numerical model brought more appropriate numerical results.
Ito, Kei; Ohno, Shuji; Koizumi, Yasuo*; Kawamura, Takumi*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12
Nishimura, Masahiro; Fukano, Yoshitaka
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 11 Pages, 2014/12
Deterministic evaluation of localflow blockage (LB) on the basis of state-of-the-art knowledge was performed using ASFRE code. In order to evaluate the effect of the realistic accidental condition, nominal power and flow rate were used for the analyses. Moreover the realistic blockage feature in the subassembly was newly adopted on the basis of existing experimental data which means LB in hound's-tooth pattern at a cross-section was assumed for the fuel subassemblies of wire spacer type. As the result, it was founds that the temperature increase in the downstream of LB was smaller than that in the past safety licensing because the flow pass is available around the blockage. And it was concluded that LB never lead to the large core damage from the evaluation results even if the blockage conditions beyond design criteria are assumed.
Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12
Probabilistic Risk Assessment (PRA) for external events has been recognized as an important safety assessment method after the TEPCO's Fukushima Daiichi Nuclear Power Station accident. The PRA should be performed not only for earthquake and tsunami which are especially key events in Japan, but also the PRA methodology should be developed for the other external hazards (e.g. tornado). In this study, the methodology was developed for Sodium-cooled Fast Reactors paying attention to that the ambient air is their final heat sink for removing decay heat under accident conditions. First, tornado hazard curve was estimated by using data recorded in Japan. Second, important structures and components for decay heat removal were identified and an event tree resulting in core damage was developed in terms of wind load and missiles (i.e. steel pipes, boards and cars) caused by a tornado. Main damage cause for important structures and components in tornado is the missiles and the tornado missiles that can reach those components and structures placed on high elevations were identified, and the failure probabilities of the components and structures against the tornado missiles were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and a probability of failure caused by the missile impacts. Finally, the event tree was quantified. As a result, the core damage frequency was enough lower than 1E-10/ry.
Takamatsu, Kuniyoshi; Hu, R.*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (
C). The RCCS uses novel shape so that the heat released from the RPV can be removed efficiently with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design greatly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting owing to overheating the fuels.
Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
This paper describes mainly snow probabilistic risk assessment (PRA) methodology development in addition to the project overview. In snow hazard category, the accident sequence was evaluated by producing event trees which consist of several headings representing the loss of decay heat removal. Snow removal action and manual operation of the air cooler dampers were introduced into the event tree as accident managements. The snow PRA showed less than 10
/reactor-year of core damage frequency.
Aizawa, Kosuke; Kobayashi, Jun; Onojima, Takamitsu; Tanaka, Masaaki; Ohno, Shuji; Kamide, Hideki; Nagasawa, Kazuyoshi*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12
Thermal striping phenomenon is one of the most important issues in an advanced loop type sodium cooled reactor JSFR. Temperature fluctuation caused by mixing of fluids at different temperature from the control rod channels and the core fuel assemblies may touch the Core Instrument Plane (CIP) at bottom of the Upper Internal Structure (UIS) and cause high cycle thermal fatigue there. In JAEA, the 1/3-scaled Five Jets Water Test (FIWAT) has been performed in order to investigate thermal striping phenomena around the CIP. In the FIWAT, the test section was simulating a control rod channel, adjacent four fuel subassemblies and a part of the CIP. The flow rate ratio and the absolute velocity of hot jets as the reference experimental condition were equal to that of the JSFR and a third of JSFR, respectively. In the experiment, it was shown that the fluid temperature fluctuation characteristics around the structure depended on the flow rate ratio. The temperature fluctuation which showed sudden decrease and recovery like a spike form was intermittently observed in the fluid near the structure. The amplitude of such spike-like temperature fluctuation in the fluid was much mitigated on the structure surface.
Takeda, Takeshi; Onuki, Akira*; Nishi, Hiroaki*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12
Shibamoto, Yasuteru; Sun, Haomin; Yonomoto, Taisuke
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12
Okano, Yasushi; Yamano, Hidemasa
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
The forest fire hazard assessment methodology, which is subject to an external hazard probabilistic risk assessment, is being developed for Sodium cooled Fast Reactor in order to quantitatively evaluate frequency and consequence of a forest fire that has a potential impact on a NPP and to obtain the core damage frequency. The new methodology consists of two parts; the first one is hazard frequency-consequence domain and the second one is Level 1 PRA. This paper focuses on forest fire propagation simulations in the first part of the methodology. The simulation is utilized to evaluate intensities of the challenges by a forest fire, and sensibility studies were performed on weather conditions. Forest fire propagation simulations were performed using FARSITE code, and the results show that the key outcome parameters depend much on wind speed and humidity but less on temperature. The fire arrival time to the site is shortened around 1/5 with changing wind speed condition from the recorded highest to the condition without wind. The time is prolonged around 3.4 times with the most humid to the recorded lowest conditions, although it is changed little when varying ambient temperature from recorded-highest to the lowest. A loss of offsite power due to fireline passage across through external power lines is a potential subsequent event, and the lag time of the fire arrival to the site after the loss of offsite power was evaluated.
Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12
Kawada, Kenichi; Takahashi, Katsuhiko*; Tobita, Yoshiharu
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12
Honda, Kei; Ohira, Hiroaki; Mori, Takero
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Thermal-hydraulics analyses of the reactor upper plenum had been performed in IAEA/Monju-CRP from 2008 to 2012. However, all of the participants got a temperature distribution which didn't agree the measured data on the thermocouple plug. In this study, we re-evaluated the inlet boundary conditions and performed another analysis. The calculated temperature distribution on the thermocouple plug had good agreement with the measured data. Thermocouples and flow guide tubes are attached over the subassembly outlets. The calculated temperature at the thermocouples agreed with the temperature of the boundary conditions. And the calculated temperature at the thermocouples had good agreement with the measured data. Therefore, the temperature at the thermocouples can be regarded as the temperature of the subassembly outlets. From these results, the inlet conditions are an appropriate ones.
Kikuchi, Shin; Seino, Hiroshi; Ohno, Shuji
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12
For countermeasure against sodium leak, structural concrete is protected by steel liner in a sodium-cooled fast reactor. However, if considering severe accidental condition such as breach of steel liner by intensive sodium leak, the reaction of concrete with liquid sodium potentially may occur. The sodium-concrete reaction (SCR) may result in significant damage of structural concrete elements, the release of hydrogen and exothermic heat. Thus it is important to understand mechanism of SCR in terms of soundness of reactor structure. However, finding on the reaction kinetics is quite limited due to the experimental difficulty. In this study, kinetics of Na
O-SiO reaction as subsequent reaction was focused. Based on the measured results by DSC equipment, kinetic parameters such as activation energy and frequency factor were obtained by the laws of chemical kinetics. XRD analysis was also performed to identify the reaction products and to discuss possible overall reactions.
Ohno, Shuji
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 9 Pages, 2014/12
This paper presents the typical characteristics of sodium combustion and subsequent reaction heat transfer behaviors observed and investigated in sodium columnar leak and fire experiment which was conducted in an enclosed steel vessel with large inner volume of about 100 m
. Especially the experiment was carried out with the main focus on the burning phenomenon within a limited spatial area in the case of large sodium leak rate as well as on the multi-dimensional thermal-hydraulics both near a sodium burning zone and in a whole region in the vessel. The investigated experimental results show us that the sodium combustion of columnar leak and its splashed droplets would lead to important oxygen deficiency behavior near the burning region, and be followed by the limitation or saturation of maximum sodium burning rate.
Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Suzuki, Toru; Tobita, Yoshiharu; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12
Onoda, Yuichi; Kurisaka, Kenichi; Sakai, Takaaki
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Tagami, Hirotaka; Tobita, Yoshiharu
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 7 Pages, 2014/12
In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the sodium plenum, a sodium experiment with visual observation was conducted using an X-ray imaging system. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Based on the experimental results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the sodium experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of molten core material. As a next step, sodium experiments using higher-density molten materials will be conducted to enrich the experimental knowledge. Besides, a new semi-empirical correlation will be developed to evaluate more appropriately the distance for fragmentation under CDA conditions.
Mizutani, Jun*; Ebara, Shinji*; Hashizume, Hidetoshi*; Yamano, Hidemasa
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
In this study, the influence of the inflow condition upon the flow separation and pressure fluctuation characteristic was evaluated by using a 1/7-scale mockup of the cold-leg piping of Japan Sodium-cooled Fast Reactor. The inflow condition to the 3rd elbow was changed from previous studies by varying the distance between the 2nd and 3rd elbows from 6.4D to 9.4D, where D is the pipe diameter. The visualization experiment showed that the flow separation appeared in the intrados of the 3rd elbow as was the case with 6.4D and separated regions became larger than that in the case of 6.4D. This is because a swirling flow observed at the inlet of the 3rd elbow became weaker than that of the case of 6.4D. The frequency analysis of pressure fluctuations showed that gentle but apparent peaks in the power spectral density (PSD) distributions of pressure fluctuations were observed at about 0.4 of the Strouhal number around the separated regions, and this peak value was half of that in the case of 6.4D. In addition, prominent peaks in the PSD distributions were observed at about 0.6 of the Strouhal number in the downstream of the reattachment point in the intrados. The peak value was approximately 3 times larger than that in the case of 6.4D. The results revealed the weakened swirling flow made the separated region larger in the downstream and the pressure fluctuation magnitude stronger.