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.

Research on the fundamental process of thermal-hydraulic behaviors in severe accident; Heat transfer on the liquid-liquid interface between molten core pool and coolant, JAERI's nuclear research promotion program, H10-027-6 (Contract research)

Mishima, Kaichiro*; Saito, Yasushi*

JAERI-Tech 2002-014, 83 Pages, 2002/03

JAERI-Tech-2002-014.pdf:6.83MB

no abstracts in English

Numerical investigation of heat transfer enhancement phenomenon during the reflood phase of PWR-LOCA

Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 36(11), p.1021 - 1029, 1999/11

https://doi.org/10.3327/jnst.36.1021

no abstracts in English

Study of film boiling collapse behavior during vapor explosion

; Abe, Yutaka*; ; ; Yamano, N.; Sugimoto, Jun

JAERI-Research 96-032, 152 Pages, 1996/06

JAERI-Research-96-032.pdf:4.05MB

no abstracts in English

Effect of fuel assembly configuration and fuel rod configuration on thermal-hydraulic behavior in core during reflood phase of PWR-LOCA

Onuki, Akira; ; Iguchi, Tadashi; Murao, Yoshio

JAERI-Research 94-012, 59 Pages, 1994/08

JAERI-Research-94-012.pdf:1.75MB

no abstracts in English

Assessment of TRAC-PF1/MOD1 and TRAC-PF1/MOD2 codes for thermal-hydraulic behavior in pressure vessel during reflood in SCTF test with an inclined radial power distribution

Onuki, Akira; ; Murao, Yoshio

JAERI-M 93-138, 55 Pages, 1993/07

JAERI-M-93-138.pdf:1.31MB

no abstracts in English

Applicability of core thermal-hydraulic models in REFLA code to 1717 type fuel assembly of PWR

Onuki, Akira; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(3), p.187 - 202, 1993/03

https://doi.org/10.3327/jnst.30.187

no abstracts in English

The Effects of coolant/fuel ratio on fuel behavior under reactivity initiated accident conditions; Flow shroud parameter test in the NSRR experiments

Tanzawa, Sadamitsu; ;

JAERI-M 91-215, 40 Pages, 1992/01

JAERI-M-91-215.pdf:2.03MB

no abstracts in English

Effect of liquid flow rate on film boiling heat transfer during reflood in rod bundle

Onuki, Akira; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 27(6), p.535 - 546, 1990/06

https://doi.org/10.3327/jnst.27.535

no abstracts in English

Heat transfer enhancement in SCTF tests

; ; ; ; ; Abe, Yutaka; Murao, Yoshio

NUREG/CP-0082 Vol.4, p.429 - 444, 1987/00

no abstracts in English