ML20027B585
| ML20027B585 | |
| Person / Time | |
|---|---|
| Issue date: | 09/07/1982 |
| From: | Keane M Office of Nuclear Reactor Regulation |
| To: | Sheron B Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20027A678 | List:
|
| References | |
| FOIA-82-543 NUDOCS 8209220434 | |
| Download: ML20027B585 (3) | |
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NOTE T0:
Brian W. Sheron, Chief, Reactor Systems Branch, DSI FROM:
Mary Ellen Keane, Section A, RSB, DSI Walt Jensen, Section C, RSB, DSI SU5 JECT:
COMPARISON OF WESTINGHOUSE FEED AND BLEED ANALYSIS AND SEMISCALE S-SR-2 W'estinghouse performed a plant calculation for conditions similar to those of the Semiscale test (IP-3 in WCAP-9744).
The Westinghouse calculation kept the core covered while the test resulted in core heat-up.
Analysis Test W (IP-3)
SS. (Trojan / Zion)
Full power:
3025 MWt based on 3411MWt PORV Cap:
139 lb/hr/MWt based on 129.2 lb/hr/MWt*
SI Shutoff:
1470 psi 1500 psi PORVs Ogen:
1500 see after loss of FW = '25%
Start of Test Conditions at PORV Openina Power:
2%
2%
Pressure:
2200 psi 2250 psi Temp:
566 F 533 F SG Water:
5 ft. (4 min, before dryout) empty
==
Conclusion:==
No core uncovering Core heat up @ 20-30 min.
The following remarks concern difference between the test and the calculation.
1)
The initial conditions in the test and the calculation appear to be roughly equivalent.
The major difference is that in the W analysis, the PORV opened when there was still 6 ft. left in the SG (four minutes before SG dryout).
Semiscale began with empty Steam Gererators.
The W analysis had an initial fluid temperature of 566 F while Semiscale had a temperature of 533 F.
It is not clear whether this difference is significant.
W concluded that if the operator waited until the SG was empty, the core would eventually be uncovered.
2)
The Westinghouse analysis was for Indian Point 3 which has only low head HPI whereas the Semiscale test was based on Trojan and Zion which have safety grade charging which was considered inoperable.
The Westinghouse report pointed out that piants with non safety grade _ charging are at lower power (3025MWt) and have a larger PORV capacity.
Semiscale is based on 3411MWt plant.
- We understand semiscale had an additional 20% PORV capacity over the reference plant.
This should aX.e_ptoridedMdjlionald pressurization and increased f,10 / MO 9l~3 f j
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- B. Sharon SEP 0 7 19B2 3)
The two tests differed.in the quality of the PORV discharge.
The two RCS pressure transients are similar..For the Westinghouse analysis the PORY dis-charges subcooled or two-phase fluid for the first 370 seconds of the transient.
The Semiscale preliminary test data (Figure 9 of North to Tiller letter August 6,1982) indicates that the subcooled or two phase period lasts for 1000 seconds.
The difference may be due to atypicalities in Semiscale pressurizer.
4)
Although Semiscale was stated to be based on flow from non-degraded ECCS other than inoperable charging, the total ECCS flow out of Semiscale uns less than that of one.ECCS pump for IP-3 based on a scaling factor of 2MW/3411MW.
(see attached curve) 5)
Semiscale used a constant power level of 2% of initial power throughout the test. This was about the decay heat ratio as used by Westinghouse at the time of PORV opening (1500 sec) but at the time of, semiscale core heat up
(+1500 sec) a value of 1.65 % should have been used. This is a mismatch iri" decay heat ratio of 21%.
==
Conclusion:==
At the time of core uncovery, 1500 sec after PORV opening, Semiscale had a 37% higher core power (initial power mismatch plus power decay nismatch).
At that time, the ECCS flow rate was less than the W rate by,approximately 5%.
These conditions probably account for the liifferences between Semiscale and the Westi.nghouse analys'is.
hAA Mary Ellen Keane, Section A, RSS V
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MEf'. ORA! DUM FOR:
Darrell Eisenhut, Director, Divis en of Licensing
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M76 FROM:
Ro rJ !
son, Director, Division of Systems g
SUSJECT:
50ARD NOTIFICAT10% CONCERNING RECENT SEMISCALE TEST RESULTS
SUMMARY
The purpose of this memorandum is to request that you notify all PUR licensing boards of the results of a recent Semiscale " feed and bleed" test.
- uring a recent test in the Semiscale facility
- in which the " feed and cieed" mode cf core cooling" was being tested, uncovery of the core simulator occurred, causing the test to be prematurely terminated to prevent core simulator overheating. The relevancy of this result is that core simulator uncovery was not expected to occur.
.o SACKGROUND Recent licensing proceedings (in particular T!:I-1 restart hearing) have focused on the ability of PWRs to remove decay heat using " feed and bleed" cooling in the event of loss of all feedwater.
Although neither the staff-nor the licensees or applicants have ever relied upon feed and bleed in order to meet the Comission's regulations, and although the staff has never c'onclude.d that all plants with installed HPI and safety-relief systems can successfully " feed and bleed," we believe that there is an inherent margin of safety attributable to a feed and bleed capability.
CONTACT:
M. Keane X28957
-Semiscale is a test facility approximately 1/1500th volume-scaled to a typical Westinghouse 4-loop PWR.
"" Feed and Sleed" refers to a mode of core ccolino in which all feedstater (main and auxiliary) is r.ot available, and decay heat removal is accomplished.by adding coo'. ant. inver. tory with the HPI system, and removing decay heat energy through the safety or relief valves.
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Darrell Eisenhut -
SEMISCALE RESULTS AND RELIArd 3 0 ;332 The Semiscale test simul'ated a loss of all feedwater kith a complete dryout of the steam generator secondary side.
The scaled PORY was opened as the recommended action to depressurize the system to below the HPI pump shutoff head to allow the HPI flow to restore primary coolant-inventory.
Prior to achieving an equilibrium th2rmal hydraulic condition for core ' cooling, the core simulator rods began to heat up 4
excessively.
This caused the test to be prematurely terminated to ~
protect the core simulator rods.
The relevance of this result is that core simulator uncovery was not
, expected to occur.
Pretest predictions were not perform?d for this particular test, and it is not known if any new phenomena occurred that.
were not capable of being predicted by current analysis computer codes..
(The expectation that no core simulator uncovery would occur was based on engineering judgment and not on detailed calculations.)
Thus, the applicability of these results to the feed and bleed capability of large PURs is unknown.
Further information is presented in the RES memorandu'a fro,EsssetttoSpeiscoveringthistopigwhichisattached.
A relaped test has been run in LOFT, which is approximately 1/60th in volume compared to a typical Westinghouse 4-loop plant.
In this~ test, the 69RV was latched open and the system depressurized to below the HPI shutoff head (the HPI was not allowed to inject for other testing purposes). There was no indication of core uncovery.
I Westinghouse has also. performed an analysis that indicates that with low-head HPI, core uncovery would occur it feed and bleed is not i
initiated before. the steam generators have dried out. An analysis of.a l
PWR at the,Semiscale tes.t conditions is part of the resolution plan.
t FESOLUTION PLAN l
To fully understand the relevance of the test, the following resolution plan will be pursued by the Office of Nuclear Regulatory Research:
1.
Analysis of the Semiscale test, including study of the data and l
RELAP 5 computer code calculation of the experiment.
l 2.
Analysis of the atypicality of Semiscale as compared to the PWR for this type of operation.
3.
Analysis of a PUR for the same conditions that existed during the Semiscale test with the RELAP5 code.
CONCLUSIONS i
cased on our assessment of the results to date and on the criteria of l.
Office Letter' Number 19, we do not believe that a board notification is wa rra nted.
However, due to the interest in feed and bleed cooling in rccent licensing proceedings, we believe it is in the best interest of the regulatory process to inform the licensing boards of this recent a
(
e Darrell Eisenhut _
AUG 3 01932 test result.
We do not believe that sufficient information is available yet to draw any conclusion from the results.
We also do not believe that these results adversely impact our present staft position regarding reliance on feed and bleed cooling.
We intend to pursue resolution of the issue with RES.
b'e expect this resolution by approximately September 30, 1982 and we will inform the boards of our conclusions at that time.
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cp Roget J. lattson, i rector l
Division of Systems Integration
Enclosure:
As Stated cc:
H. Denton E. Case S. Hanauer G. Knighton P., Minogue, RES D. Ross, RES
W. Hodoes G. Mazetis
- TT. Lyon R. Barrett M. Keane T. Novak i
l G. Lainas W. Jensen
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