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Takeda, Takeshi
JAEA-Data/Code 2023-012, 75 Pages, 2023/10
JAEA-Data-Code-2023-012.pdf:4.45MB
An experiment denoted as TR-LF-15 was conducted on June 11, 2014 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment TR-LF-15 simulated accident management (AM) actions during a station blackout transient with TMLB' scenario with pump seal loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). This scenario is featured by loss of auxiliary feedwater functions. The pump seal LOCA was simulated by a 0.1% cold leg break. The test assumptions included total failure of both high pressure injection system and low pressure injection system of emergency core cooling system (ECCS). Also, it was presumed non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. When steam generator (SG) secondary-side collapsed liquid level dropped to a certain low liquid level, the primary pressure turned to rise. After the SG secondary-side became voided, the safety valve of a pressurizer cyclically opened which led to loss of primary coolant. Core uncovery thus took place owing to core boil-off at high pressure. When an increase of 10 K was confirmed in cladding surface temperature of simulated fuel rods, SG secondary-side depressurization was started as the first AM action. At that time, the safety valves in both SGs were fully opened. Primary depressurization was initiated by completely opening the pressurizer safety valve as the second AM action with some delay after the first AM action onset. When the SG secondary-side pressure lowered to 1.0 MPa following the first AM action, water was injected into the secondary-side of both SGs via feedwater lines with low-head pumps as the third AM action. A reduction in the primary pressure was accelerated because the heat removal from the SG secondary-side system resumed shortly after the third AM action initiation.
Takeda, Takeshi
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
JAEA-Data-Code-2023-007.pdf:3.24MB
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.
Kurisaka, Kenichi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
This study aims to understand a time trend of the occurrence rate of steam generator (SG) tube leak in the existing sodium-cooled fast reactors (SFRs) based on the observed data. The target on SFRs in the present paper is Phenix in France and BN600 in Russia. From the open literature review, we investigated the number of tube-to-tube plate weld, the number of tube-to-tube weld, heat transfer area of tube base metal, operating time of SGs, dates when SG tube leak occurred, leaked location, corrective action after tube leak such as replacement of leaked module. Based on these observed data, time to leak is estimated and then time trend of the occurrence rate of SG tube leak for each of the above-mentioned parts was quantitatively analyzed by the hazard plotting method. As a result, the rate of leak at tube-to-tube weld in Phenix shows increase with time due to probable cause of cyclic thermal stress in a short term. As for a long-term trend, the rate of tube leak in both Phenix and BN600 SGs indicated decrease with time probably thanks to improvement in welding and in SG operating condition and to removal of initial failure.
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-013, 20 Pages, 2022/01
JAEA-Research-2021-013.pdf:2.35MB
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. An idea has been proposed to implement a steam condenser as an accident countermeasure. This measure is expected to prevent nitric acid steam diffusing in facility building and to increase gaseous Ru trapping ratio into condensed water. A simulation study has been carried out with a hypothetical typical facility building to analyze the efficiency of steam condenser. In this study, SCHERN computer code simulates chemical behaviors of Ru in nitrogen oxide, nitric acid and water mixed vapor based on the conditions obtained from simulation with thermal-hydraulic computer code MELCOR. The effectiveness of steam condenser has been analyzed quantitively in preventing mixed vapor diffusion and gaseous Ru trapping effect. Some issues to be solved in analytical model has been also clarified in this study.
Takeda, Takeshi
JAEA-Data/Code 2021-006, 61 Pages, 2021/04
JAEA-Data-Code-2021-006.pdf:2.78MB
An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.
Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12
Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.
Kobayashi, Takuya; Kawamura, Hideyuki; Kamidaira, Yuki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 62(11), p.635 - 639, 2020/11
It is important to predict the dispersion of radioactive materials released into the ocean due to nuclear accidents in the surrounding ocean of the east Asian countries. The Japan Atomic Energy Agency developed a Short-Term Emergency Assessment system of Marine Environmental Radioactivity (STEAMER) based on an oceanic dispersion model. STEAMER quickly predicts the oceanic dispersion of radioactive materials in the surrounding ocean of the east Asian countries using the online prediction data of oceanic condition. We validated the predictability of the oceanic dispersion and demonstrated the improvement of the predictability using an ensemble prediction method. Moreover, we developed a high resolution model in the coastal region using a Regional Ocean Modeling System (ROMS).
C studied by laser heatingPham, V. H.; Nagae, Yuji; Kurata, Masaki; Bottomley, D.; Furumoto, Kenichiro*
Journal of Nuclear Materials, 529, p.151939_1 - 151939_8, 2020/02
Times Cited Count：14 Percentile：87.35(Materials Science, Multidisciplinary)Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Furumoto, Kenichiro*; Sato, Hisaki*; Ishibashi, Ryo*; Yamashita, Shinichiro
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09
Kamidaira, Yuki; Kawamura, Hideyuki; Kobayashi, Takuya; Uchiyama, Yusuke*
Journal of Nuclear Science and Technology, 56(8), p.752 - 763, 2019/08
Times Cited Count：7 Percentile：61.18(Nuclear Science & Technology)Oceanic regional downscaling capability was implemented into Short-Term Emergency Assessment system of Marine Environmental Radioactivity (STEAMER) developed by Japan Atomic Energy Agency to enable us to predict more realistically the oceanic dispersion of radionuclides at higher spatiotemporal resolutions for broader applications. The system consisted of a double-nested oceanic downscaling circulation model with tidal forcing and an oceanic radionuclide dispersion model. This system was used to comparatively examine downscaling and tidal effects on the dispersion of radionuclides hypothetically released from the Fukushima Daiichi Nuclear Power Plant in the colder season. The simulated dissolved 
Cs distribution was different from that obtained using coarser-resolution models because downscaling enhanced both horizontal and vertical mixing. The suppression of horizontal mixing and the promotion of vertical mixing by tidal forcing synergistically reduced offshore Cs transport. In addition, the submesoscale effects strengthened the three-dimensional Cs fluctuations by 
10 times, while the tidal effects promoted slightly increased the intensity of three-dimensional Cs fluctuations by approximately 3%. This indicated that the submesoscale effects substantially surpassed tidal forcing in oceanic mixing in the coastal margin off Fukushima in the colder season.
Uesawa, Shinichiro; Horiguchi, Naoki; Suzuki, Takayuki*; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
Takeda, Takeshi
JAEA-Data/Code 2018-003, 60 Pages, 2018/03
JAEA-Data-Code-2018-003.pdf:3.68MB
Experiment SB-PV-07 was conducted on June 9, 2005 using LSTF. Experiment simulated 1% pressure vessel top small-break LOCA in PWR under total failure of HPI system and nitrogen gas inflow to primary system from ACC tanks. Liquid level in upper-head was found to control break flow rate. Coolant was started to manually inject from HPI system into cold legs as first accident management (AM) action when maximum core exit temperature reached 623 K. Fuel rod surface temperature largely increased because of late and slow response of core exit temperature. SG secondary-side depressurization was initiated by fully opening relief valves as second AM action when primary pressure decreased to 4 MPa. However, second AM action was not effective on primary depressurization until SG secondary-side pressure decreased to primary pressure. Pressure difference became larger between primary and SG secondary sides after ACC tanks started to discharge nitrogen gas.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 54(10), p.1036 - 1045, 2017/10
Times Cited Count：5 Percentile：44.07(Nuclear Science & Technology)To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium. The effect of use of the unstructured mesh was also investigated by the two analyses using structured and unstructured mesh.
Hotta, Akitoshi*; Morita, Akinobu*; Kajimoto, Mitsuhiro*; Maruyama, Yu
Nihon Genshiryoku Gakkai Wabun Rombunshi, 16(3), p.139 - 152, 2017/09
Kobayashi, Takuya; Kawamura, Hideyuki; Fujii, Katsuji*; Kamidaira, Yuki
Journal of Nuclear Science and Technology, 54(5), p.609 - 616, 2017/05
Times Cited Count：10 Percentile：68.85(Nuclear Science & Technology)The Japan Atomic Energy Agency has, for many years, been developing a radionuclide dispersion model for the ocean, and has validated the model through application in many sea areas using oceanic flow fields calculated by the ocean model. The Fukushima Dai-ichi Nuclear Power Station accident caused marine pollution by artificial radioactive materials to the North Pacific, especially to coastal waters northeast of mainland Japan. In order to investigate the migration of radionuclides in the ocean caused by this severe accident, studies using marine dispersion simulations have been carried out by JAEA. Based on these as well as the previous studies, JAEA has developed the Short-Term Emergency Assessment system of Marine Environmental Radioactivity (STEAMER) to immediately predict the radionuclide concentration around Japan in case of a nuclear accident.
Kojima, Saori*; Uchibori, Akihiro; Takata, Takashi; Ohno, Shuji; Fukuda, Takeshi*; Yamaguchi, Akira*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 8 Pages, 2016/11
Analytical evaluation on a self-wastage phenomenon at heat transfer tubes in the steam generator of sodium cooled fast reactors has been performed by using the sodium-water reaction analysis code SERAPHIM. In this study, a fluid-structure thermal coupling model was developed and incorporated in the SERAPHIM code to evaluate heat transfer between the sodium-side reacting flow and the outer surface of the heat transfer tube. The effect of the fluid-structure thermal coupling model on the temperature field was demonstrated through the numerical analyses.
4 bundle at 2 MPaLiu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki
Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06
To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 4 bundle under 1.6 MPa (202 C), 2.1 MPa (215 C) and 2.6 MPa (226 C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.
Takeda, Takeshi
JAEA-Data/Code 2015-022, 58 Pages, 2016/01
JAEA-Data-Code-2015-022.pdf:3.31MB
The SB-HL-12 test simulated PWR 1% hot leg SBLOCA under assumptions of total failure of HPI system and non-condensable gas (nitrogen gas) inflow. SG depressurization by fully opening relief valves in both SGs as AM action was initiated immediately after maximum fuel rod surface temperature reached 600 K. After AM action due to first core uncovery by core boil-off, the primary pressure decreased, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before LSC induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after nitrogen gas inflow. Third core uncovery by core boil-off occurred during reflux condensation. The maximum fuel rod surface temperature exceeded 908 K.
Narukawa, Takafumi; Amaya, Masaki
Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01
Times Cited Count：7 Percentile：54.81(Nuclear Science & Technology)Jang, S.*; Takata, Takashi; Yamaguchi, Akira*; Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4275 - 4288, 2015/08
Numerical analysis of the self-wastage phenomenon was carried out using a multi-dimensional computational code called SERAPHIM. Several steps of numerical analysis were constructed to reproduce transient self-wastage phenomenon caused by Sodium Water Reaction (SWR). Numerical analysis of multiphase flow with chemical reaction near the initial crack is firstly performed. The wastage amount is evaluated based on hypothetical Arrhenius equation by using the temperature and molar concentration of sodium hydroxide. New analytical grid is created by exchanging the solid cells to fluid cells in the reaction based on the wastage amount evaluation. These series of procedure is repeated. The width and the shape of the enlarged crack showed good agreement with the experimental results.
