ML20210V146
| ML20210V146 | |
| Person / Time | |
|---|---|
| Issue date: | 02/11/1987 |
| From: | NRC |
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| Shared Package | |
| ML20210V110 | List: |
| References | |
| NUDOCS 8702180699 | |
| Download: ML20210V146 (7) | |
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SAFETY EVALUATION SUPPLEMENT RELATING.T0 VERIFICATION OF ANALYSIS MtTH005 FOR SMALL BREAK LOCAs TMI ACTION ITEM II.K.3.30-FOR COPBUSTION ENGINEERING PLANT 5 POST TEST ANALYSIS OF SEMISCALE TEST 5-UT-08 CEN-203-P, PEVISION 1-P, SUPPLEMENTS 3 AND 4
1.0 INTRODUCTION
In its safety-evaluation (Reference 1), the NRC staff found TMI Action Item II.K.3.30 to be resolved for all licensed Combustion Engineering (CE) plants with the condition that it be shown that the computer program CEFLASH-4AS could acceptably calculate the results of Semiscale Test S-UT-08, a small-break loss-of-coolant accident (SBLOCA). During this test, the water level in the simulated reactor vessel dropped rapidly prior to loop seal clearing. The rapid drop in the reactor vessel we'.er level was attributed to liquid holdup in the steam generator, i.e., the liquid briefly accumulated in the U-tubes of the intact loop steam generator.
To satisfy the condition in the staff's safety evaluation, the Combustion Engineering Owners Group (CEOG) performed an analysis of the S-UT-08 test.
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Because of the small scale of the Semiscale facility, unique phenomena which occured during the test and conservatisms that are an integral part of the CEFLASH-4AS licensing model, the CEFLASH-4AS model could not be used directly to calculate results that would agree with the experimen-
-tal data. A "best estimate" (BE) version of CEFLASH-4AS was devised for the analysis. Since the BE version would not be used for licensing, it,also had to be shown that the licensing " evaluation model" (EM) version of CEFLASH-4AS contained modeling features which would predict the drop in
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a water level in the reactor vessel prior to loop seal clearing in a SBLOCA in a full-scale plant (i.e., the " core level depression" or the " core uncovery spike").
The procedure that the CEOG followed for doing this was as follows:
1.
The EM version of CEFLASH-4AS was modified to a best estimate (BE) program for calculating the S-UT-08 test results, and the test results were calculated. The results of these calculations are given and discussed in Section 6 of Reference 2.
2.
The BE models of components that determine the steam generator liquid holdup and the core level depressien in S-UT-08 were replaced by their EM counterparts or eliminated if they were not part of the EM version of CEFLASH-4AS. This BE/EM version, as it was called, was used to calculate the results of the S-UT-08 test. The j
results of these calculations are given and discussed in l
Section 7 of Reference 2.
3.
The factors that caused the core uncovery spike in S-UT-08 were examined and the relative magnitudes of the effects of each of the factors were determined semi-quantitatively.
The results of these efforts are discussed in pages 4 to 23 of Reference 3.
4.
Those items of the EM that were not used in the BE/EM analysis were reviewed to assure that they do not affect the ability of the EM version of CE FLASH-4AS to conser-vatively account for the core uncovery spike in a full-scale plant. The results of this review are discussed in pages 24 to 30 of Reference 3.
2.0 EVALUATION In general, the BE analysis results are in excellent agreement with the experimental values. However, in the 50 to 180 second time interval, there are differences between the calculated and S-UT-8 liquid levels in the steam generator and pump suction leg in the intact loop. After this time interval, there is excellent agreement in the liquid levels; so the minimum liquid level in the simulated reactor vessel is accurately cal-culated by the BE program. On an overall basis, the staff finds that the BE analysis acceptably ;.redicted the S-UT-08 experimental values, especially the core level response.
The agreement between BE/EM results and the S-UT-08 data was also generally i
good. However, the BE/EM analysis did not conservatively calculate the minimum liquid level in the simulated reactor vessel during the experiment.
The staff was concerned that this non-conservatism might be present in the EM program and requested the CE0G to evaluate its cause.
The CE0G determined that the primary cause for the non-conservative core liquid level prediction was "rewet steam," i.e., steam that is produced when water in the hot legs flows back into the reactor vessel and drops on the fuel rods. When the S-UT-08 model of this "rewet steam"
. production from the BE program was put into the BE/EM pregram, the minimum liquid level in the simulated reactor vessel was conservatively calculated.
The phenomena of "rewet steam" is a Semiscale specific phenomena.
In Semiscale, the simulated reactor vessel is only 3 inches in diameter.
During the S-UT-08 test, the water flowing back from the hot legs spread fairly uniformly over all of the 25 simulated fuel rods ard produced a significant amount of steam. However, since the diameter of a reactor vessel in a plant is about 12 feet instead of 3 inches, the water flowing back from the hot legs in a plant would only contact. the fuel elements in the portions of the core periphery directly beneath them. This is such a small percentage of the total number of fuel elements that "rewet steam" production is not expected to be a significant phenomenon in a SBLOCA in a plant. Hence, the staff finds that it does not have to be modeled in the EM version of CEFLASH-4AS.
Based upon the analysis discussed above, the staff finds that the BE/EM analysis acceptably predicts the S-UT-08 experimental values. Thus, the staff concludes that the EM version of CEFLASH-4AS contains modeling f: tures widch would conservatively predict the " core uncovery spike" in a plant.
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The CEOG identified those items of the EM which were not used in the BE/EM analysis and classified them into three categories:
1.
The conservatisms required by Appendix V of 10 CFR 50.
2.
The boundary conditions that were not compatible with those of the S-UT-08 test.
3.
The component models (e.g., reactor kinetics model) that were not applicable to S-UT-08.
While these items are significant portions of the EM, the CE0G concluded that these items are not directly related to the phenomena which leads to the core level depression. Thus, the CEOG concluded that BE/EM analysis is sufficient for demonstrating that the EM version of CE FLASH-4AS will conservatively calculate the core level depression. The staff has reviewed the CE0G assessment and concurs with their conclusions.
3.0 CONCLUSION
The CE0G submitted CEN-203, Revision 1-P, Supplement 3 and Revision 1-P, Supplement 4 in response to NRC's concern that the CEFLASH-4AS computer program might not be able to calculate the initial rapid drop in water level experienced in the simulated reactor vessel in the Semiscale Test S-UT-08. The staff has reviewed these submittals.
It finds that the
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CEFLASH-4AS program can acceptably calculate this test. Therefore, as stated in NRC's safety evaluation (Reference 1) the requirements to per-form plant specific analyses, per Action Item II.K.3.31, will no longer be required. The staff finds topical report CEN-203-P, Revision-1-P, Supplements 3 and 4 acceptable for referencing by licensees of CE plants for resolving TMI Action Item II.K.3.30.
4.0 REFERENCES
1.
NRC Safety Evaluation Report, "TMI Action Item II.K.3.30 for Combustion Engineering Plants," dated May 23, 1985.
2.
Combustion Engineering Report, " Post-Test Analysis of Semiscale Test
'S-UT-08;" CEN-203, Revision 1-P, Supplement 3; dated December 1985.
3.
Combustion Engineering Report, " Response to NRC Request for Additional l
Information for Verification of Analysis Methods for.Small Break LOCA's,"
CEN-203, Revision 1-P, Supplement 4, dated November 1986.
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Dr. J. K. Casper 2-February 11, 1987 resubmit their respective documentation, or stitnit justification for the continued effective crplicability of tiie topical report without revision of their respective documentation.
Sincerely, Ortsaanisa nea g a
L % %ehtteld Dennis M. Crutchfield, Assistant Director Divisinn of PWR Licensing-B Office of Nuclear Reactor Regulation Safety Evaluation cc: See next page DISTRIBUTION Central File CRossi NRCPDR REnch PPAS RSB File WMinners ELantz RDosnak RJones BSheron CThomas AThadani Glainas WHedces CBerlinger RDiggs, LFMS GDick Division R/F FMiraglia CCrutchfield f-0FC :RSB:DPLB-P(p:SL RSB:DPLB:BC 3: AD NAME :E 7:cm RJones D
2g( /87 2/// /87 DATE :1/11/87 j
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