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Journal Articles

Core thermal-hydraulic analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*

Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 10 Pages, 2021/08

To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs, therefore, a thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics code (RV-CFD model) is necessarily required. In this study, the application of the subchannel analysis method for subassemblies to the RV-CFD model was attempted to reduce the calculation costs. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1. As the result, the behavior of cold sodium into the simulated core was well grasped and the calculated sodium temperature in the core had good agreement with the experimental result. The applicability of the RV-CFD model was confirmed.

Journal Articles

Development of neutronics and thermal-hydraulics coupled analysis method on platform for design optimization in fast reactor

Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 25, 4 Pages, 2020/06

With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.

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

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

Development of numerical estimation method for thermal hydraulics in reactor vessel of sodium-cooled fast reactor under decay heat removal system operation conditions; Preliminary thermal hydraulics simulation for simulated reactor vessel in sodium experimental apparatus PLANDTL-2

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

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08

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. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition 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 an appropriate numerical models for the direct heat exchanger (DHX).

Journal Articles

Study on applicability of fast reactor plant dynamics analysis code to core thermal hydraulics under natural circulation decay heat removal conditions

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04

A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.

Journal Articles

Validation of plant dynamics analysis code for fast reactor core thermal hydraulics under natural circulation conditions

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/06

Under natural circulation decay heat removal conditions, three characteristic phenomena; flow redistribution in the core as well as in the fuel subassemblies, inter-subassembly heat transfer and gap flow between wrapper tubes of fuel subassemblies are important for assessing the temperature distribution in the core. In order to improve the prediction accuracy, a whole core model which can consider these three phenomena has been incorporated into the plant dynamics analysis code Super-COPD. In this study, analyses of two kinds of sodium experiments were performed to validate Super-COPD with the whole core model, which were focusing on inter-subassembly heat transfer phenomena.

Journal Articles

Ongoing validation of sodium fire analysis code system for SFR safety evaluation

Ohno, Shuji; Hamase, Erina; Kamide, Hideki

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 8 Pages, 2013/04

This paper describes outline and current status of ongoing verification and validation activitiess of sodium fire analysis code system for fast reactor plant. The simulation accuracy of sodium droplet burning model is assessed and shown. The effectiveness of three-dimensional gas thermal-hydraulic analysis is also investigated.

Journal Articles

Investigation of dominant factors for evaluation of sodium leak and fire accident consequences by sensitivity analyses

Ohno, Shuji; Hamase, Erina; Kamide, Hideki

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 5 Pages, 2012/12

Sensitivity analyses of sodium leak and fire are performed to identify the dominant factors for the accident consequences. The analyses with a multi-cell zone model code SPHINCS treat sodium spray and pool simultaneous combustion and heat-mass transfer behaviors in a large-scale two-cell geometry. Atmospheric gas pressure increase and temperature increase of floor steel plate below the sodium pool are analyzed as the figures of merit to be directly focused on. The analyses clarify the important and dominant factors of the phenomena in the accident sequence quantitatively, resulting in the acquirement of the knowledge to conduct the appropriate code validation activity and to discuss the uncertainty in the safety evaluation results.

Oral presentation

Verification and validation of sodium fire analysis codes for fast reactors, 3; Assessment of dominant factors by sensitivity analyses

Ohno, Shuji; Hamase, Erina; Kamide, Hideki

no journal, , 

We conducted sensitivity analyses on the thermal effect of sodium leak and fire accident as one of the verification and validation activities for sodium fire analysis codes. The analyses enable to quantify the dominant physical/chemical factors in the sodium fire behavior.

Oral presentation

Verification and validation of sodium fire analysis codes for fast reactors, 4; Analysis of combustion test with large sodium leakage rate in 2 cells geometry

Hamase, Erina; Ohno, Shuji

no journal, , 

Verification and validation for sodium fire analysis codes have been the focus of our recent research aiming to apply these codes to a safety assessment for a sodium-cooled fast reactor. In the present study, the combustion experiment with large sodium leakage rate in 2 cells geometry was simulated by using the sodium fire analysis code, SPHINCS. Two cells geometry consists of the combustion room and the adjacent room which have the opening between them. Also, the adjacent room has another opening through outside. It was found that the propagation of gas pressure from the combustion room to the adjacent room had generally good agreement between the experiment and SPHINCS analysis. It has been also confirmed that it could be possible to adopt the spray droplet combustion model to the simulation of combustion experiment in which the sodium jet hit the obstacle and the ceiling in the combustion room.

Oral presentation

Verification and validation of sodium fire analysis codes for fast reactors, 5; Evaluation of multi-dimensional analysis effectiveness with AQUA-SF

Ohno, Shuji; Oki, Hiroshi*; Hamase, Erina; Kamide, Hideki

no journal, , 

Verification and validation activities have been under progress for a sodium fire analysis code system to be utilized in fast reactor plant design and safety evaluation. In this work, the authors performed parametric analyses of sodium spray fire using a fire analysis code AQUA-SF, demonstrating the effectiveness of multi-dimensional thermal-hydraulic analysis which enables the realistic evaluation of fire consequence by the consideration of local oxygen deficiency near the sodium burning region.

Oral presentation

Current status of sodium fire analysis code system development and validation

Ohno, Shuji; Hamase, Erina

no journal, , 

Validation study of numerical analysis tools has been underway in the Japan Atomic Energy Agency in order to establish a safety evaluation method for sodium leak and fire accident postulated in sodium-cooled fast reactor plants. This paper describes briefly the current status of sodium fire analysis codes development and validation activities, emphasizing that the code validation is performed with focusing on important phenomena and dominant factors for accident consequence. It is also demonstrated from a numerical experiment that one of the important evaluating issues of large-scale sodium spray leak and fire would be predicted more realistically and precisely by a three-dimensional sodium fire analysis code since it would enable the estimation of local effect; i.e. the oxygen deficiency behavior in the spray fire region.

Oral presentation

Verification and validation of sodium fire analysis codes for fast reactors, 6; Analysis of combustion test with sodium droplet splashing on the wall

Hamase, Erina; Ohno, Shuji

no journal, , 

In the present study, the fire experiment with large sodium leakage rate in 2 cells geometry was simulated by using the sodium fire analysis code, SPHINCS. Two cells geometry consists of the fire room and the adjacent room which have the opening between them. In the fire room, the sodium jet was directed vertically upwards and it was dispersed with fire after impacting on the obstacle, ceiling and side walls. In order to analyze this complicated phenomenon, the pathway of splashing sodium droplets was obtained in the numerical calculation and the motion of sodium droplets was arranged in chronological sequence. It was found that the behavior of gas pressure had generally good agreement between the experiment and SPHINCS analysis, and by performing SPHINCS calculation the gas pressure in the fire room is significantly increased due to gas expansion with temperature increasing and then is reduced due to the propagation of gas pressure from the fire room to the adjacent room.

Oral presentation

Development of estimation technology for availability of measure for failure of containment vessel in sodium cooled fast reactor, 2; Development of detailed numerical simulation methods for sodium fire

Aoyagi, Mitsuhiro; Hamase, Erina; Ohno, Shuji; Yamada, Takahiro*; Watanabe, Tadashi*; Uno, Masayoshi*

no journal, , 

We have been conducting an R&D project to develop estimation technology for availability of measure for failure of containment vessel in sodium cooled fast reactors. The aim of this study is development of detailed numerical simulation methods for sodium fire consequences considering multi-dimensional thermal-hydraulic behavior.

Oral presentation

Validation of plant dynamics analysis code super-COPD for fast reactor core thermal hydraulics under natural circulation with PLANDTL experiment

Hamase, Erina; Ono, Ayako; Doda, Norihiro; Nabeshima, Kunihiko; Ohshima, Hiroyuki

no journal, , 

In order to validate the one-dimensional thermal hydraulic analysis code, Super-COPD, the experiment under natural circulation (NC) conditions using the plant dynamics test loop (PLANDL) was analyzed. In the present analysis, all seven subassemblies in the PLANDTL core were modeled and each subassembly was divided into the center and peripheral region due to evaluate the radial heat transfer through the wrapper tube walls. Boundary conditions were as follows: the flow rate, the core inlet temperature and the heater power in each subassembly. As a result, the sodium temperature at the top of the heated region was in good agreement with the experiment. Furthermore, the scaling analysis was performed based on the top-down approach in the Hierarchical Two-Tiered Scaling. In the top-down approach, characteristic time ratios normalized the characteristic frequency of various transfer processes such as NC head, shape and friction loss, heat transfer and heat conduction etc. with the residence time in the core region were obtained by normalizing conservation equations of mass, momentum and energy. It was found that under NC, the largest characteristic time ratio was the one for NC head, or NC was the most important process in the core region. Then, the scale distortions of these characteristics time ratios between PLANDTL and MONJU were evaluated and thus the modeling performance of important process could be reconfirmed.

Oral presentation

Analysis of core thermal hydraulics under natural circulation condition with plant dynamics analysis code Super-COPD

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

no journal, , 

Flow re-distribution in a fuel subassembly is one of the most important phenomena under natural circulation in the sodium fast breeder reactor from the viewpoint of safety assessment. The core model of the plant dynamic code (Super-COPD) was validated through analyses using natural circulation experimental data obtained in the plant dynamics test loop (PLANDTL). The simulated sodium temperatures at the top and middle of the heated region in the subassembly showed good agreement with the experimental data. It was confirmed that the core model of Super-COPD can precisely reproduce the flow re-distribution in the subassembly.

Oral presentation

Analysis of core thermal hydraulics under natural circulation condition with plant dynamics analysis code Super-COPD, 2; Applicability evaluation of the inter-subassembly heat transfer model

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

no journal, , 

In order to validate the plant dynamic analysis code Super-COPD for inter-subassembly heat transfer in the core, the analyses that simulate the steady-state experiments with the plant dynamic test loop (PLANDTL) in the low flow conditions corresponding to natural circulation have been carried. The two test cases that have a significant temperature difference between adjacent subassemblies were simulated. In both cases, the simulated sodium temperature distributions in subassemblies had good agreement with experimental results. It was found that the Super-COPD could evaluate the inter-subassembly heat transfer in the core.

Oral presentation

Development of estimation technology for availability of measure for failure of containment vessel in sodium cooled fast reactor, 18; Summary of Development of Detailed Numerical Simulation Methods for Sodium Fire

Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Hamase, Erina; Uno, Masayoshi*

no journal, , 

An R&D project has been conducted to develop detailed numerical simulation methods for sodium fire consequences which is one of risks to put heat and pressure loads to containment vessel of sodium-cooled fast reactor. This contribution summarizes the R&D results involving development of spray combustion and heat transfer models, and a validation of the code through a sodium fire analysis of an upward spray experiment.

Oral presentation

Development of core thermal hydraulic evaluation methods under natural circulation decay heat removal in sodium-cooled fast reactors; Validation of subchannel CFD analysis using a porous body model

Hamase, Erina; Kikuchi, Norihiro; Doda, Norihiro; Tanaka, Masaaki; Imai, Yasutomo*

no journal, , 

In order to develop numerical analysis that can perform a core-plenum interaction in a commercialized sodium-cooled fast reactor, a subchannel analysis based on a porous body model in CFD code was initially applied to the 37-pin fuel assembly sodium test. As a result of CFD analysis for typical high and low flow conditions, the sodium temperature and the axial velocity at upper heated section in the subassembly were consistent with the ASFRE code that has been validated with experiments data within relative error 1.5%. It was found that a subchannel analysis based on a porous body model in CFD code could apply to the thermal hydraulics analysis in the fuel assembly.

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