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Evaluation report on SCTF-III test S3-3, S3-4 and S3-5. Counter Current Flow Limitation Phenomena in full radius core

Technical Report:

Abstract

In order to investigate the Counter Current Flow Limitation (CCFL) phenomena in full radius core, CCFL simulation tests (S3-3, S3-4 and S3-5) were performed using the Slab Core Test Facility (SCTF) Core-III. In these tests, the pressure in the pressure vessel was maintained at 0.3 MPa. Steam upflow was established in the core by injecting steam into the cold leg and water was injected into the upper plenum. Intact loops and pressure-vessel-side broken loop were closed to establish the steam upflow in the core. The break-through occurred in a localized region, while in the other region steam flowed up. Thus, the break-through occurrence was not uniform in full radius core. The radial break-through location was dependent mainly on the water subcooling distribution in the upper plenum. The break-through area decreased with decrease in steam injection rate. For typical PWR case, the ratio of the break-through area was approximately 20%. The break-through rate increased with the increase in the injected water subcooling and with the decrease in steam injection flow, similarly to small scale model tests. However, the quantitative relation between the break-through rate and the steam upflow rate was different from small scale model test. Two types of the relation  More>>
Authors:
Iguchi, Tadashi; Iwamura, Takamichi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Murao, Yoshio; [1]  Sakaki, Isao; Adachi, Hiromichi
  1. Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Publication Date:
Oct 01, 1991
Product Type:
Technical Report
Report Number:
JAERI-M-91-172
Reference Number:
SCA: 220900; 210200; PA: JPN-91:012206; SN: 92000659240
Resource Relation:
Other Information: PBD: Oct 1991
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; PWR TYPE REACTORS; REWETTING; REACTOR SAFETY; REACTOR EXPERIMENTAL FACILITIES; ECCS; TWO-PHASE FLOW; LOSS OF COOLANT; HYDRAULICS; EVALUATION; MEDIUM PRESSURE; COUNTERFLOW SYSTEMS; TIME DEPENDENCE; 220900; 210200; POWER REACTORS, NONBREEDING, LIGHT-WATER MODERATED, NONBOILING WATER COOLED
OSTI ID:
10117114
Research Organizations:
Japan Atomic Energy Research Inst., Tokyo (Japan)
Country of Origin:
Japan
Language:
English
Other Identifying Numbers:
Other: ON: DE92768128; TRN: JP9112206
Availability:
OSTI; NTIS (US Sales Only); INIS
Submitting Site:
JPN
Size:
164 p.
Announcement Date:
Jun 30, 2005

Technical Report:

Citation Formats

Iguchi, Tadashi, Iwamura, Takamichi, Akimoto, Hajime, Okubo, Tsutomu, Ohnuki, Akira, Murao, Yoshio, Sakaki, Isao, and Adachi, Hiromichi. Evaluation report on SCTF-III test S3-3, S3-4 and S3-5. Counter Current Flow Limitation Phenomena in full radius core. Japan: N. p., 1991. Web.
Iguchi, Tadashi, Iwamura, Takamichi, Akimoto, Hajime, Okubo, Tsutomu, Ohnuki, Akira, Murao, Yoshio, Sakaki, Isao, & Adachi, Hiromichi. Evaluation report on SCTF-III test S3-3, S3-4 and S3-5. Counter Current Flow Limitation Phenomena in full radius core. Japan.
Iguchi, Tadashi, Iwamura, Takamichi, Akimoto, Hajime, Okubo, Tsutomu, Ohnuki, Akira, Murao, Yoshio, Sakaki, Isao, and Adachi, Hiromichi. 1991. "Evaluation report on SCTF-III test S3-3, S3-4 and S3-5. Counter Current Flow Limitation Phenomena in full radius core." Japan.
@misc{etde_10117114,
title = {Evaluation report on SCTF-III test S3-3, S3-4 and S3-5. Counter Current Flow Limitation Phenomena in full radius core}
author = {Iguchi, Tadashi, Iwamura, Takamichi, Akimoto, Hajime, Okubo, Tsutomu, Ohnuki, Akira, Murao, Yoshio, Sakaki, Isao, and Adachi, Hiromichi}
abstractNote = {In order to investigate the Counter Current Flow Limitation (CCFL) phenomena in full radius core, CCFL simulation tests (S3-3, S3-4 and S3-5) were performed using the Slab Core Test Facility (SCTF) Core-III. In these tests, the pressure in the pressure vessel was maintained at 0.3 MPa. Steam upflow was established in the core by injecting steam into the cold leg and water was injected into the upper plenum. Intact loops and pressure-vessel-side broken loop were closed to establish the steam upflow in the core. The break-through occurred in a localized region, while in the other region steam flowed up. Thus, the break-through occurrence was not uniform in full radius core. The radial break-through location was dependent mainly on the water subcooling distribution in the upper plenum. The break-through area decreased with decrease in steam injection rate. For typical PWR case, the ratio of the break-through area was approximately 20%. The break-through rate increased with the increase in the injected water subcooling and with the decrease in steam injection flow, similarly to small scale model tests. However, the quantitative relation between the break-through rate and the steam upflow rate was different from small scale model test. Two types of the relation were observed. When the steam upflow rate was high (corresponding to condition during early reflood phase), the break-through rate was nearly the same as one predicted with a typical one-dimensional CCFL correlation, indicating that the break-through behavior is almost one-dimensional even in a full radius core. However, when the steam upflow rate was decreased gradually (corresponding to condition during later reflood phase), the break-through rate increased suddenly at a certain steam upflow rate and could not be predicted with a typical one-dimensional CCFL correlation, indicating that the break-through behavior can switch to be nonuniform in full radius core. (J.P.N.).}
place = {Japan}
year = {1991}
month = {Oct}
}