ML20215J540

Draft Safety Evaluation Re Safe Emergency Shutdowns.Facility Operation at 82% Power Acceptable

ML20215J540
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/22/1987
From:	Office of Nuclear Reactor Regulation
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ML20215J514	List: ML20215J512 ML20215J540
References
NUDOCS 8706250033
Download: ML20215J540 (4)

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GDRAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR RE RELA 1,3G,TDSAFEEMERGENCYShl)TDOWNS' PUBLIC5ERVICECOMPANYOFCOLORADO

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1.0 INTRODUCTION

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., d In Fort St. Vrain License Event Report- (LER) #86-026,' dated October.17,1986, the Public Service Company of Colorado (PSC) reported that the. Safe Shutdown Cooling System for removing the decay heat follbwing the. postulated " Design  ;

i Basis Earthquake" or " Maximum Tornado" accidents was inadequate. PSC state this LER that.it one of these t'wo accidents were to occur while the reacto; operating at 105% power, and if, as postulated in.Section 10.3.9 of the FSAR, .{

the. functions of all'non-seismic, non-Category I components were lost and.the primary helium coolant flow was assumed interrupted for 90 minutes to' allow for

' manual realignments, the safe shutdown cooling system would be unable to keep j the fuel temperature below the 2900* F limit. FurtherthisLERstates.that'th analysis for the removal of decay heat by the Safe Shutdown Cooling System, 'l "did not consider firewater pump capacity nor the associated steam generator inlet or. discharge piping configurations".

For the corrective action in the LER, the PSC committed to reanalyzing this .

Safe Shutdown Cooling System and providing an acceptable method to remove the

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decay lieat and cool the ple'nt without fuel failure, j In the two-loop, Fort St. Vrain plant each loop has six steam generator modules which have parallel secondary coolant flow paths. .Each steam.

.generatorhastwosections,i.e.,an' economizer-evaporator-superheater(EES)

, -section and a reheater section.

The reheater sections of the steam generators are much smaller than the EES sections;'so' their use se'emed logical for removing the smaller decay heat -

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2 load. :DRAFI cross sectional flow area is relatively hig

, so.their the firewater pumps have only enough Theflow ca consequence of this is that

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sections, rather than all six as previously assumedpacity to that this partial flooding would not provid .

PSC's reanalysis showed

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concluded Cooling. that the reheater sections should not bee enough used for the Safe Shutdown

'Instead, atmosphere.

PSC proposed,to usecthe'EES se ti available vent lines was too high; so ne )

installed. resistance in the \

temperature could not be kepta the' Cooling accident scenario.

below fuel the 2 imit for the Safe Shutdown In addition PSC has to comply with Appendix R 'i shutdown method that uses only equipment qualified to following fires in non-congested cable areas urvive and operate PSC' had analyses done to determine from what

' shutdowns could be accomplished for all of-these accid! \

analyses, which PSC submitted to the NRC, showed that e accident scenario,. the fuel temperature can be kept b l,-_

during emergency shutdowns after long-term . e ow the 2900* F limit /

82 percent for the Appendix R shutdown scenarioperation at po Shutdown Cooling scenario. o and 87.5 percent for the Safe

! 2.0 EVAtt!ATION y _

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The.NRC had the Oak Ridge National Laboratory The technical evaluation (TER) report'on evaluate thi all of these submittals.

Three parts of. the TER pertain to this Safety Es evaluation is ,

part, which is on possible structural'and metalluvaluation (SE).- Th 1 generators, is the subject'of a separate safety ev lrgical failures i

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a uation.

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The first of the three parts is the evaluation of the calculations of the maximum fuel temperature that will be obtained after these postulated accidents. This evaluation was made by using the Oak Ridge developed ORECA computer program to independently calculate these temperatures. As can be seen l

in the TER the ORECA calculations show that 82% is a conservative power level for a limiting fuel temperature of 2900* F.

The second of the three parts in the TER that pertain to this SE is the evaluation of the ability of the existing systems to supply sufficient water flow to both the helium circulator pelton wheel drives and the EES sections of the steam generators during these emergency cooldowns. The final conclusion i I

of this lengthy review is that for these scenarios, "there is substantial margin in the existing cooling systems to provide for a safe shutdown". This conclusion is contingent on several items, two of which are:

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1. There are operating procedures for these accidents and that the operators I have been trained to follow them.

2. PSC should perform an explicit analysis to demonstrate that the original Class I firewater flow path can accomodate a single active failure in i the new Class I firewater flow path when the required EES pre-cooling j times and the long term cooling are accounted for.  !

i Another contingency in the TER findings is for the NRC to perform an audit or do confirmatory analyses of the PSC flow calculations. However, the staff l believes that with the satisfactory agreement between PSC's calculated results l and the results of the firewater flow test, which are reported in Reference 2 and mentioned on page 7 of the TER, no confirmatory analyses are required.

l The remaining contingency in the TER findings and conclusions is concerned j with a passive failure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of cooling. However, PSC's calculations show that aftc.r 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of cooling adequate flow can be obtained from a redundant flow path. Based upon these calculations, the staff finds that a j

passive failure can be accomodated, 1 I

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i The third of the three parts in the TER thet pertain to this SE'is the  ;

evaluation of the possibility of water hammer that would prevent these l emergency cooldowns. From these qualitative statements the staff finds that j J

it is unlikely that water hammer will preclude a safe shutdown.

3.0 CONCLUSION

S The staff finds that after PSC has:

. l (1) provided operating procedures for the postulated " Design Basis Earthquake", " Maximum Tornado", and " Appendix R Fire" accidents and has trained its operators to follow them; and (2) performed an explicit analysis that shows that the original Class I  ;

firewater flow path can accommodate a single active failure in the new Class I firewater flow path when the required EES pre-cooling times and long term cooling are accounted for; the Fort St. Vrain reactor can be shutdown after prolonged operation at 82 percent of the licensed power without having the fuel temperature exceed the 2900' F limit. Thus the staff finds that after these two confirmations have been made operation at 82 percent power is acceptable.

4.0 REFERENCES

1. SECY-77-439 dated August 17, 1979.

2. Letter from H. L. Brey, Public Service Company of Colorado, to J. A.

Calvo, USNRC, dated May 4, 1987.

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