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Journal Articles

CFD analysis of natural circulation in LBE-cooled accelerator-driven system

Sugawara, Takanori; Watanabe, Nao; Ono, Ayako; Nishihara, Kenji; Ichihara, Kyoko*; Hanzawa, Kohei*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 10 Pages, 2022/03

Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides (MAs) included in high level wastes discharged from nuclear power plants. The ADS is a lead-bismuth cooled tank-type reactor with 800 MW thermal power. It is supposed that the ADS is safer than conventional critical reactors because it is operated in a subcritical state. The previous study performed the transient analyses for the typical ADS accidents such as unprotected loss of flow or beam overpower. It was shown that all calculation cases except loss of heat sink (LOHS) satisfied the no-damage criteria. To avoid the damage by LOHS, the ADS equips Direct Reactor Auxiliary Cooling System (DRACS) to remove the decay heat. The most important points of a DRACS operation are its reliability and to ensure the flowrate in a natural circulation state. This study aims to perform the CFD analysis of the natural circulation to clarify the flowrate in the ADS reactor vessel.

Journal Articles

Scaling-up capabilities of TRACE integral reactor nodalization against natural circulation phenomena in small modular reactors

Mascari, F.*; Bersano, A.*; Woods, B. G.*; Reyes, J. N.*; Welter, K.*; Nakamura, Hideo; D'Auria, F.*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

JAEA Reports

Data report of ROSA/LSTF experiment SB-SL-01; Main steam line break accident

Takeda, Takeshi

JAEA-Data/Code 2020-019, 58 Pages, 2021/01


An experiment denoted as SB-SL-01 was conducted on March 27, 1990 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment SB-SL-01 simulated a main steam line break (MSLB) accident in a pressurized water reactor (PWR). The test assumptions were made such as auxiliary feedwater (AFW) injection into secondary-side of both steam generators (SGs) and coolant injection from high pressure injection (HPI) system of emergency core cooling system into cold legs in both loops. The MSLB led to a fast depressurization of broken SG, which caused a decrease in the broken SG secondary-side wide-range liquid level. The broken SG secondary-side wide-range liquid level recovered because of the AFW injection into the broken SG secondary-side. The primary pressure temporarily decreased a little just after the MSLB, and increased up to 16.1 MPa following the closure of the SG main steam isolation valves. Coolant was manually injected from the HPI system into cold legs in both loops a few minutes after the primary pressure reduced to below 10 MPa. The primary pressure raised due to the HPI coolant injection, but was kept at less than 16.2 MPa by fully opening a power-operated relief valve of pressurizer. The core was filled with subcooled liquid through the experiment. Thermal stratification was seen in intact loop cold leg during the HPI coolant injection owing to the flow stagnation. On the other hand, significant natural circulation prevailed in broken loop. When the continuous core cooling was ensured by the successive coolant injection from the HPI system, the experiment was terminated. The experimental data obtained would be useful to consider recovery actions and procedures in the multiple fault accident with the MSLB of PWR. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-SL-01.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Effects of temperature fluctuation on PIV measurement of natural circulation flow field

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*

Proceedings of 14th International Symposium on Advanced Science and Technology in Experimental Mechanics (14th ISEM'19) (USB Flash Drive), 4 Pages, 2019/11

The particle image velocimetry (PIV) was measured in scaled-model water experiments simulating a natural circulation flow field in a sodium-cooled fast reactor vessel. The temperature fluctuation in the natural circulation flow field causes the distribution of the refractive index. Thus, the temperature fluctuation affects the uncertainty of the velocity in the PIV measurement. In this study, the authors evaluated the effects of the temperature fluctuation on the PIV measurement in the natural circulation flow field.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.

Journal Articles

Numerical analysis of EBR-II shutdown heat removal test-17 using 1D plant dynamic analysis code coupled with 3D CFD code

Doda, Norihiro; Hiyama, Tomoyuki; Tanaka, Masaaki; Ohshima, Hiroyuki; Thomas, J.*; Vilim, R. B.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

In sodium-cooled fast reactors, a natural circulation is expected to remove the core decay heat when the plant gets into a station blackout. From a perspective of reactor safety, the core hot spot temperature arising in the natural circulation should be evaluated accurately. To this end, Japan Atomic Energy Agency is trying to couple a 1-D plant dynamics analysis code Super-COPD and a 3-D CFD code AQUA to solve the thermal-hydraulic field in the whole plant under natural circulation condition. As a validation study, the coupled code was applied to an analysis of EBR-II shutdown heat removal test. The obtained numerical results reasonably agreed with the measured data, which demonstrated the validity of the coupled code.

Journal Articles

Validation and applicability of reactor core modeling in a plant dynamics code during station blackout

Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04

Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.

Journal Articles

An Experimental study on natural circulation decay heat removal system for a loop type fast reactor

Ono, Ayako; Kamide, Hideki; Kobayashi, Jun; Doda, Norihiro; Watanabe, Osamu*

Journal of Nuclear Science and Technology, 53(9), p.1385 - 1396, 2016/09

 Times Cited Count:6 Percentile:45.9(Nuclear Science & Technology)

Decay heat removal by natural circulation is a significant passive safety measure of a fast reactor against station blackout. The decay heat removal system (DHRS) of the loop type sodium fast reactor being designed in Japan comprises a direct reactor auxiliary cooling system and primary reactor auxiliary cooling system (PRACS). The thermal hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant. The experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, the experiments varying the pressure loss coefficients of the loop as the experimental parameters showed robustness of the PRACS.

Journal Articles

A Study on the thermal-hydraulics in the damaged subassemblies under the operation of decay heat removal system

Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04

Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.

Journal Articles

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*

Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09

 Times Cited Count:11 Percentile:73.32(Nuclear Science & Technology)

A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.

Journal Articles

Analysis of natural circulation tests in the experimental fast reactor JOYO

Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08

Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.

Journal Articles

Safety requirements expected for the prototype fast breeder reactor "Monju"

Saito, Shinzo; Okamoto, Koji*; Kataoka, Isao*; Sugiyama, Kenichiro*; Muramatsu, Ken*; Ichimiya, Masakazu*; Kondo, Satoru; Yonomoto, Taisuke

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05

JAEA Reports

Analysis on non uniform flow in steam generator during steady state natural circulation cooling

Susyadi; Yonomoto, Taisuke

JAERI-Research 2005-011, 64 Pages, 2005/06


Steady-state natural circulation (NC) in the PWR was investigated focusing on non uniform flow among steam generator (SG) U-tubes observed in the ROSA/LSTF experiments. In the analysis using the RELAP5/MOD3 code, the SG behavior was analyzed using the partial SG model with one, five, or nine parallel flow paths in the primary side and boundary conditions based on the experiments. The results showed that simulations using the model with five or nine tubes were capable to capture important non uniform phenomena such as reverse flow, fill and dump and stagnant vertical stratification, and the stable SG outlet flow as observed in the experiments. Heat transfer rates to the secondary side were, however, underpredicted by up to 15%. Furthermore, difficulties were found in establishing the steady state condition especially for the low pressure analysis: only when the inlet flow rate was carefully imposed, stable NC behavior was obtained.

Journal Articles

Thermal-hydraulic research on future reactor systems in the ROSA program at JAERI

Yonomoto, Taisuke; Otsu, Iwao; Svetlov, S.*

Proceedings of 3rd Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-3), p.521 - 528, 2002/00

A research project is being conducted at the Japan Atomic Energy Research Institute on thermal hydraulics for the future reactor systems. The present paper provides the belief description of the project, followed by two recent topics: the natural circulation in the PWR loop and the condensation heat transfer for a passive cooling system. For the first topic, we discuss the importance of the modeling of the nonuniform flow behavior among SG U-tubes for the assessment of the long-term decay heat removal systems relying on the SG secondary side cooling. Such a system is planned to be used in APWR+, a Japanese next-generation PWR. The condensation heat transfer was investigated using the data obtained at the SPOT test facility in Russia. The results have shown that the measured heat transfer rates on the inner surface of the tube consisting of several bends and short straight sections can be predicted using the existing correlations with the accuracy of several percentage, although the correlations are based typically on the data taken using relatively long straight tube.

Journal Articles

Analysis of multi-dimensional boiling flow in secondary water pool of horizontal PCCS; Effect of pool size

Onuki, Akira; Nakamura, Hideo; Kawamura, Shinichi*; Saishu, Sadanori*

Nihon Kikai Gakkai Netsu Kogaku Koenkai Koen Rombunshu, p.31 - 32, 2001/11

A passive containment cooling system (PCCS) is under planning to use in a next-generation-type BWR for long-term cooling by condensing steam using horizontal heat exchangers. Heat transfer behavior in a secondary water pool is one of important phenomena governing heat removal performance of the PCCS. Boiling and condensation can be supposed under high heat flux regions and the characteristics of the two-phase natural circulation should be evaluated. This study investigated effects of pool size on the characteristics by multi-dimensional two-fluid model code ACE-3D. It was found from the analyses that the pool size gives no significant influences for the characteristics in tube bundle under local-boiling mode.

Journal Articles

ROSA/LSTF experiments on low-pressure natural circulation heat removal for next-generation PWRs

Yonomoto, Taisuke; Otsu, Iwao

Proceedings of 12th Pacific Basin Nuclear Conference (PBNC 2000), Vol.1, p.317 - 329, 2000/00

no abstracts in English

Journal Articles

Safety analysis of a submersible compact reactor SCR for under-sea research vessel

Yoritsune, Tsutomu; Kusunoki, Tsuyoshi; Odano, Naoteru; Ishida, Toshihisa

Proceedings of the 4th JSME-KSME Thermal Engineering Conference, p.1_31 - 1_36, 2000/00

no abstracts in English

JAEA Reports

Journal Articles

Nonuniform flow behavior among SG U-tubes occurringduring a low-pressure two-phase natural circulation phase in a PWR

Yonomoto, Taisuke; Anoda, Yoshinari

Nihon Kikai Gakkai Dai-6-Kai Doryoku, Enerugi Gijutsu Shimpojiumu '98 Koen Rombunshu, p.198 - 203, 1998/00

no abstracts in English

40 (Records 1-20 displayed on this page)