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JAEA Reports

Heat transfer coefficients model for SIMMER-III and SIMMER-IV

Brear, D. J.*; Kondo, Satoru; Sogabe, Joji; Tobita, Yoshiharu*; Kamiyama, Kenji

JAEA-Research 2024-009, 134 Pages, 2024/10

JAEA-Research-2024-009.pdf:2.45MB

The SIMMER-III/SIMMER-IV computer codes are being used for liquid-metal fast reactor (LMFR) core disruptive accident (CDA) analysis. The sequence of events predicted in a CDA is often influenced by the heat exchanges between LMFR materials, which are controlled by heat transfer coefficients (HTCs) in the respective materials. The mass transfer processes of melting and freezing, and vaporization and condensation are also controlled by HTCs. The complexities in determining HTCs in a multi-component and multi-phase system are the number of HTCs to be defined at binary contact areas of a fluid with other fluids and structure surfaces, and the modes of heat transfer taking into account different flow topologies representing flow regimes with and without structure. As a result, dozens of HTCs are evaluated in each mesh cell for the heat and mass transfer calculations. This report describes the role of HTCs in SIMMER-III/SIMMER-IV, the heat transfer correlations implemented and the calculation of HTCs in all topologies in multi-component, multi-phase flows. A complete description of the physical basis of HTCs and available experimental correlations is contained in Appendices to this report. The major achievement of the code assessment program conducted in parallel with code development is summarized with respect to HTC modeling to demonstrate that the coding is reliable and that the model is applicable to various multi-phase problems with and without reactor materials.

JAEA Reports

Thermal design study of lead-bismuth cooled accelerator driven system, 1; Study on thermal hydraulic behavior under normal operation condition

Akimoto, Hajime; Sugawara, Takanori

JAEA-Data/Code 2016-008, 87 Pages, 2016/09

JAEA-Data-Code-2016-008.pdf:15.62MB

Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.

JAEA Reports

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03

JAEA-Data-Code-2014-031.pdf:37.23MB

A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

Journal Articles

CFD analysis

Takase, Kazuyuki; Misawa, Takeharu*

Supercritical-Pressure Light Water Cooled Reactors, p.301 - 319, 2014/12

https://doi.org/10.1007/978-4-431-55025-9_3

no abstracts in English

Journal Articles

Natural convection heat transfer of high temperature gas in an annulus between two vertical concentric cylinders

Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki

Heat Transfer-Asian Research, 34(5), p.293 - 308, 2005/07

https://doi.org/10.1002/htj.20070

Water cooling panels have been adopted as the vessel cooling system of the HTTR to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate the heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with thermal radiation heat transfer in an annulus between two vertical concentric cylinders with the inner cylinder heated and the outer cylinder cooled. In the present experiments, Rayleigh number based on the height of the annulus ranged from 2.010 $^{7}$ to 5.410 $^{7}$ for helium gas and from 1.210 $^{9}$ to 3.510 $^{9}$ for nitrogen gas. The numerical results were in good agreement with the experimental ones regarding the surface temperatures of the heating and cooling walls. As a result of the experiments and the numerical analyses, the heat transfer coefficient of natural convection coupled with thermal radiation was obtained.

JAEA Reports

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAERI-Research 2005-014, 170 Pages, 2005/06

JAERI-Research-2005-014.pdf:7.64MB

A small break LOCA (SBLOCA) experiment was conducted at the LSTF of ROSA-V program to study effects of accident management (AM) on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a PWR. The experiment, SB-PV-03, simulated ten instrument-tube break LOCA at the PWR vessel bottom equivalent to 0.2% cold leg break, total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and AM actions on secondary depressurization at -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes. It was clarified that the AM actions were effective on primary depressurization until AIS injection end at 1.6 MPa, but thereafter became less effective by the gas inflow, resulting in low pressure injection (LPI) delay and whole core heatup under continuous water discharge at the break. The report describes these phenomena including core heatup related with primary coolant mass and AM actions, primary-to-secondary heat transfer analysis and estimation of gas in the primary loops.

Journal Articles

Temperature transient analysis of gas circulator trip test using the HTTR

Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30%(9MW). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.

Journal Articles

Study on natural convection heat transfer of high temperature gas in a vertical annular space of a double coaxial cylinder

Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki

Nihon Kikai Gakkai Rombunshu, B, 70(694), p.1518 - 1525, 2004/06

https://doi.org/10.1299/kikaib.70.1518

no abstracts in English

Journal Articles

Development of plant dynamics analytical code named Conan-GTHTR for the Gas Turbine High Temperature Gas-cooled Reactor, 1; Code validation by Use of the experimental data of HTTR

Takamatsu, Kuniyoshi; Katanishi, Shoji; Nakagawa, Shigeaki; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.76 - 87, 2004/03

https://doi.org/10.3327/taesj2002.3.76

The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in Higt Temperature Gas-Cooled Reactor systems.

JAEA Reports