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Journal Articles

Validation study of finite element thermal-hydraulics analysis code SPIRAL to a large-scale wire-wrapped fuel assembly at low flow rate condition

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10

A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.

Journal Articles

Methodology development for transient flow distribution analysis in high temperature gas-cooled reactor

Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.

Journal Articles

Development of a flow network calculation code (FNCC) for high temperature gas-cooled reactors (HTGRs)

Aoki, Takeshi; Isaka, Kazuyoshi; Sato, Hiroyuki; Ohashi, Hirofumi

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

The flow distribution analysis performed in the HTGR design has to take into account the interaction thermal and radiation deformations of the graphite structure, and the gaps between the graphite structures forming unintended flow. In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The validation of FNCC was performed for the one-column flow distribution test. The analytical results using FNCC showed good agreement with the experimental results. It is concluded that FNCC was validate for the analysis of distributions of flowrate and pressure for the flow network model including the unintended flow paths in prismatic-type HTGRs.

Journal Articles

Numerical simulation of laser welding different kinds of materials using a thermohydraulics computational science numerical simulation code SPLICE

Muramatsu, Toshiharu; Sato, Yuji; Kamei, Naomitsu; Aoyagi, Yuji*; Shobu, Takahisa

Nippon Kikai Gakkai Dai-13-Kai Seisan Kako, Kosaku Kikai Bumon Koenkai Koen Rombunshu (No.19-307) (Internet), p.157 - 160, 2019/10

no abstracts in English

Journal Articles

Prospects based on T-H roadmap through communication

Nakamura, Hideo

Nippon Genshiryoku Gakkai-Shi, 61(4), p.270 - 272, 2019/04

no abstracts in English

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*

Nippon 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

Observation of dry patch behavior on copper heat transfer surface and measurement of surface temperature distribution in nucleate pool boiling

Uesawa, Shinichiro; Ono, Ayako; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Dai-55-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 8 Pages, 2018/05

no abstracts in English

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

Nippon 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

Failure evaluation analysis of reactor pressure vessel lower head in BWR due to severe accident

Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Osaka, Masahiko

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

To investigate the inhomogeneous temperature and stress distribution by geometrical complex of BWR lower head, the detailed 3D model of RPV lower head with control rod guide tubes and shroud supports are constructed and the 3D thermal hydraulic analysis of simulated molten pool and creep deformation analysis of lower head are performed using ANSYS Fluent / Mechanical finite element code. It is found that failure for BWR lower head might be caused by combination between melting failure in inner surface of lower head and creep failure in outer surface of lower head.

Journal Articles

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Nuclear Engineering and Design, 312, p.30 - 41, 2017/02

 Times Cited Count:6 Percentile:35.22(Nuclear Science & Technology)

In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

Journal Articles

Development of V2UP (V&V plus uncertainty quantification and prediction) procedure for high cycle thermal fatigue in fast reactor; Framework for V&V and numerical prediction

Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki

Nuclear Engineering and Design, 299, p.174 - 183, 2016/04

 Times Cited Count:3 Percentile:59.87(Nuclear Science & Technology)

A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) was made by referring to the existing guidelines on V&V and the methodologies of the safety assessment (CSAU, ISTIR, EMDAP). The V2UP consisted of five components as follows: (1) phenomena analysis with the Phenomena Identification and Ranking Table (PIRT) method, (2) implementation of the V&V, (3) design and rearrangement of experiments for the V&V, (4) uncertainty quantification in each problem and integration of uncertainties and (5) numerical prediction (estimation) for the target issue. Although the complete application of the procedure has not been performed at this moment, a flow chart of the V2UP procedure was described in this paper with recent results of the examinations.

Journal Articles

Critical heat flux prediction for subcooled flow boiling in annulus

Liu, W.

Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.391 - 392, 2015/06

Subcooled flow boiling is a boiling that begins and develops even though the mean enthalpy of liquid phase is lower than saturation. This forced convective boiling is one of the most efficient ways for the removal of high heat flux. It is widely used in the high heat flux components such as nuclear reactor cores, accelerator targets and fusion reactor components. The thermal outputs of these systems are restricted by Critical Heat Flux(CHF). Because of the importance of the CHF, lots of researches, including both experimental and mechanistic modelling, have been performed. However, the CHF prediction for rod bundles still remains unsolved. As the first step for the CHF prediction in rod bundles, in this paper, we tried to predict the CHF in annulus, which is the most basic flow geometry simplified from a rod bundle. We performed the CHF prediction by using liquid sublayer dryout model, combined with Nouri single phase velocity distribution correlation for annulus. The results show that the CHF in annulus can be predicted in an accuracy of about $$pm$$20%.

Journal Articles

Numerical visualization of herted liquid film flow behavior around a spacer in a narrow channel

Kume, Etsuo; Kitamura, Tatsuaki*; Takase, Kazuyuki; Ose, Yasuo*

Kashika Joho Gakkai-Shi, 25(Suppl.2), p.369 - 370, 2005/10

no abstracts in English

JAEA Reports

Energy confinement and transport of H-mode plasmas in tokamak

Urano, Hajime

JAERI-Research 2004-027, 131 Pages, 2005/02

JAERI-Research-2004-027.pdf:5.69MB

no abstracts in English

Journal Articles

Effect of cladding surface pre-oxidation on rod coolability under reactivity initiated accident conditions

Sugiyama, Tomoyuki; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11

 Times Cited Count:8 Percentile:48.23(Nuclear Science & Technology)

The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1$$mu$$m and 10$$mu$$m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.

Journal Articles

Divertor power handling in a low aspect ratio tokamak reactor

Sakurai, Shinji; Tobita, Kenji; Nishio, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 80(11), p.955 - 958, 2004/11

no abstracts in English

JAEA Reports

Thermal-hydraulic analyses of poisoned cold moderator vessel, 1; Study on Poison Plate Layout

Sato, Hiroshi; Aso, Tomokazu; Kogawa, Hiroyuki; Teshigawara, Makoto; Hino, Ryutaro

JAERI-Tech 2004-018, 23 Pages, 2004/03

JAERI-Tech-2004-018.pdf:2.42MB

The Japan Atomic Energy Research Institute is constructing a mega-watt class spallation neutron source in cooperation with the High Energy Accelerator Research Organization. A cold moderator using liquid hydrogen is one of the key components in the system, which directly affects the neutronic performance both in intensity and pulse time structure. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the recirculation and stagnant regions which would cause hot spots. A cold moderator with a poison plate (poisoned decoupled moderator) has a high possibility to generate the stagnant region on and near the poison plate. Thermal-hydraulic analyses were carried out with proposed inner structure of the poisoned cold moderator. The stagnant and recirculation regions could be reduced by making a gap between the poison plate end and the vessel bottom surface, and the local temperature rise also could be kept under the required design value.

Journal Articles

Multiphase flow problems on thermofluid safety for fusion reactors

Takase, Kazuyuki

Konsoryu, 17(3), p.267 - 276, 2003/09

no abstracts in English

Journal Articles

Critical power correlation for axially uniformly heated tight-lattice bundles

Kureta, Masatoshi; Akimoto, Hajime

Nuclear Technology, 143(1), p.89 - 100, 2003/07

 Times Cited Count:10 Percentile:41.13(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Critical power in axially uniformly-heated tight-lattice rod bundles

Kureta, Masatoshi; Akimoto, Hajime

Nippon Kikai Gakkai Rombunshu, B, 69(682), p.1469 - 1476, 2003/06

no abstracts in English

260 (Records 1-20 displayed on this page)