Analysis of Semiscale Mod-1 integral test with asymmetrical break (Test S-29-1)
Description
Selected experimental data obtained from Semiscale Mod-1 cold leg break Test S-29-1 and results obtained from analytical codes are analyzed. This test was the first integral blowdown reflood test conducted with the Mod-1 system and was a special test designed specifically to evaluate the sensitivity of the early Mod-1 core thermal response (0 to 5 sec after rupture) to the magnitude and direction of the core flow. To achieve this specific objective in Test S-29-1, the vessel side break area was reduced to approximately one-half the scaled break area associated with a 200 percent cold leg break test. The reduction in break area significantly reduced the core flow reversal that took place immediately after rupture and resulted in periods of positive core flow in the early portion of the test. The results obtained from this test are compared with results obtained from a 200 percent cold leg break test and the effect of core flow on early core thermal response is evaluated. Since Test S-29-1 was the first integral blowdown reflood test conducted with the Mod-1 system, data are also presented through the reflood stage of the test and the results are analyzed. The test data and the core thermal response calculated with the RELAP4 code are also compared
Availability note (English)
MF available from INIS under the Report Number; Available from NTIS., PC A03/MF A01.Files
8328201.pdf
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Additional details
Additional titles
- Augmented title (English)
- PWR
Publishing Information
- Imprint Pagination
- 44 p.
- Report number
- TREE-NUREG--1076
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 8328201
- Subject category
- S61: RADIATION PROTECTION AND DOSIMETRY; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- BLOWDOWN; COMPUTER CODES; FLOW RATE; HEAT TRANSFER; HYDRAULICS; LIQUID FLOW; LOSS OF COOLANT; PRESSURE GRADIENTS; PWR TYPE REACTORS; REACTOR CORES; REACTOR SAFETY; SPATIAL DISTRIBUTION
- Descriptors DEC
- ACCIDENTS; DISTRIBUTION; ENERGY TRANSFER; FLUID FLOW; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; WATER COOLED REACTORS; WATER MODERATED REACTORS