Abstract
Post test calculations for two tests with steep or flat radial core power distribution using the Slab Core Test Facility (SCTF) were performed to assess the TRAC-PF1/MOD1 code for the thermal-hydraulic behaviors in pressure vessel during reflood phase of a PWR-LOCA. The predictive capability for the two-dimensional thermal-hydraulic behavior in pressure vessel was also assessed in this report. The TRAC code predicted transients of clad surface temperature well including radial distribution of clad surface temperature caused by different bundle power. However, the TRAC predictions showed poor agreement on void fractions in the core and in the upper plenum. For the radial distribution of void fraction, the TRAC code predicted a peculiar distribution in the core which was not observed in the SCTF tests, and predicted a flatter distribution in the upper plenum compared to measured results. Recommendation was made for the future improvement on hydraulic and heat transfer models based on this assessment study. (author).
Ohnuki, Akira;
Akimoto, Hajime;
Murao, Yoshio
[1]
- Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Citation Formats
Ohnuki, Akira, Akimoto, Hajime, and Murao, Yoshio.
Assessment of TRAC-PF1/MOD1 code for thermal-hydraulic behavior in pressure vessel during reflood in SCTF test with a radial power distribution.
Japan: N. p.,
1993.
Web.
Ohnuki, Akira, Akimoto, Hajime, & Murao, Yoshio.
Assessment of TRAC-PF1/MOD1 code for thermal-hydraulic behavior in pressure vessel during reflood in SCTF test with a radial power distribution.
Japan.
Ohnuki, Akira, Akimoto, Hajime, and Murao, Yoshio.
1993.
"Assessment of TRAC-PF1/MOD1 code for thermal-hydraulic behavior in pressure vessel during reflood in SCTF test with a radial power distribution."
Japan.
@misc{etde_10120813,
title = {Assessment of TRAC-PF1/MOD1 code for thermal-hydraulic behavior in pressure vessel during reflood in SCTF test with a radial power distribution}
author = {Ohnuki, Akira, Akimoto, Hajime, and Murao, Yoshio}
abstractNote = {Post test calculations for two tests with steep or flat radial core power distribution using the Slab Core Test Facility (SCTF) were performed to assess the TRAC-PF1/MOD1 code for the thermal-hydraulic behaviors in pressure vessel during reflood phase of a PWR-LOCA. The predictive capability for the two-dimensional thermal-hydraulic behavior in pressure vessel was also assessed in this report. The TRAC code predicted transients of clad surface temperature well including radial distribution of clad surface temperature caused by different bundle power. However, the TRAC predictions showed poor agreement on void fractions in the core and in the upper plenum. For the radial distribution of void fraction, the TRAC code predicted a peculiar distribution in the core which was not observed in the SCTF tests, and predicted a flatter distribution in the upper plenum compared to measured results. Recommendation was made for the future improvement on hydraulic and heat transfer models based on this assessment study. (author).}
place = {Japan}
year = {1993}
month = {Jul}
}
title = {Assessment of TRAC-PF1/MOD1 code for thermal-hydraulic behavior in pressure vessel during reflood in SCTF test with a radial power distribution}
author = {Ohnuki, Akira, Akimoto, Hajime, and Murao, Yoshio}
abstractNote = {Post test calculations for two tests with steep or flat radial core power distribution using the Slab Core Test Facility (SCTF) were performed to assess the TRAC-PF1/MOD1 code for the thermal-hydraulic behaviors in pressure vessel during reflood phase of a PWR-LOCA. The predictive capability for the two-dimensional thermal-hydraulic behavior in pressure vessel was also assessed in this report. The TRAC code predicted transients of clad surface temperature well including radial distribution of clad surface temperature caused by different bundle power. However, the TRAC predictions showed poor agreement on void fractions in the core and in the upper plenum. For the radial distribution of void fraction, the TRAC code predicted a peculiar distribution in the core which was not observed in the SCTF tests, and predicted a flatter distribution in the upper plenum compared to measured results. Recommendation was made for the future improvement on hydraulic and heat transfer models based on this assessment study. (author).}
place = {Japan}
year = {1993}
month = {Jul}
}