ML19332C168
| ML19332C168 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/16/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19332C161 | List: |
| References | |
| NUDOCS 8911220381 | |
| Download: ML19332C168 (5) | |
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^l UNITED STATES pjf NUCLEAR REGULATORY COMMISSION e
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SAFETY EVALUATION BY Tht 0FFICE OF fluCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.135 TO FACILITY OPERATlhG LICENSE NO. DPR-32
!.ND AMENDMEllT N0.135 TO FACILITY OPERATING LICENSE t;0. DPR-37 d
VIRGild A ELECTRIC AhD POWER COMPANY.
SURRY POWER STATION, UNIT NOS.1 AND 2 l
DOCKET h05. 50-280 AND 50-281 1.0 ' INTRODUCTION Pursuant to 10 CFR E0.90 and 50.91 Virginia Electric and Power Company
'(VEPCC) proposed to amend Facility Operating Lit.enses Nos. DPR-32 and DPR-37 for the Surry Power Station, Units 1 and 2.
By letter dated
--Noverrber 10, -1589, VEPC0 prcposed to revise. the pressurizer safety valves' (PSVs) setpoint toleranc'e of Technical Specification 3.1. A.3.c from one
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porcent to minus-(-) one percent ano plus (+) tive percent for the remainder of. Cycle 10 for Surry, Units 1 and 2 by replacing the current footnote.
These Technical Specification changes are required because of recent information indicating a~ potential shift in the. ressurizer safety valve shif t setpoint tolerance that may exceed the. one percent value currently required by the Technical Specifications.
This change will maintain the reactor ecolant system pressure below the 110 percent design limit specified in the Updated Final Safety Analysis Report (UFSAR).
2.0 DISCUSSION AND EVALUATICN The Surry Units 1 agd 2 PSVs cre installed downstream.of loop seals which are filled with'300 F water.
The lift setpoints of the PSVs on both units were set with steam.
In October 1989, the licenset was. informed by Westinghouse of a finding that the actual PSV-lift setpciat could shif t by 4. to 8 percent under environments different from that used to establish the setpoint.
Since Unit 2 was shut down on Cctober 13, 1989 to correct a itakage problem in the "B" PSV, th7 licensee decided to test the Unit 2 FSVs. When tested in a loop secl water environment, the results showed an increase of lift setpoint of +3.L tc +5 percent from the as-found setpoint established with steam. The licensee, therefore, performed a safety analysis whose results indicated thtt the reactor coolant system (RCS) pressure of the lineiting overpressurization events would remain below the accep-tance criterion of 2750 psia (110 percent design pressure) with lift pres-
'sures up to 5.4 percent above the setpoint pressure.
In addition, the licensee proposed compensatory measures to maintain operability of at least one power-operated relief valve (PORV) and the anticipatory reactor trip on 8911220381 891116 DR ADOCK 05000280 PDC
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~2-turbine' trip circuitry.
Based on the licensee's analysis and' proposed compensatory actions, NRC granted relief from the existing Technical Specification in the. form of discretionary enforcement until December ~ 1, 1989 (NRC-letter to VEPC0 dated 0ctober 27,1989).
The= lif t pressures of the Unit 2 PSVs were subsequently reset with loop-seal water to correspond to the actual installation envircnnent.
Howev er, during reactor coolant system (RCS) pressure testing prior to return to:
service on.hovember 6,.1989, the "C" PSV lifted prematurely at 2335 psig j
due to an apparent loss of, loop seal wcter.
In order to minimize the potential 'for challenges to the PSVs, which may result in failure of the valve to reseat, resulting in a small break loss of coolant accident, the licensee decided to reset the lift pressures'for the Unit 2 PSVs with steam, consistent with Unit 1.
Considering the fact that the actual PSV lift pressure under a loop seal environment may be 3.5 to 5 percent higher than the setting-established L
with steam, the licensee has performed a' safety analysis for the relevant L
UFSAR transients including loss of load / turbine trip, locked rotor, main feedline break, loss of normal feedwater and rod ejection.- In all cases the peak RCS pressure was found to be below the acceptance criterion of i
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'2750 psia even if the PSV lif t pressure are assumed to increase by 5.4 J
percent.
Therefore, the TS change to allow the PSV setpoint tolerance increase to 5 percent would not result in the RCS' pressure exceeding 110 percent of design pressure.
Since the "C" PSV on t' nit 2 lifted at a pressure about 6 percent lower I
than the set pressure,- contrary to the maximum of 5 percent shift found during the vahe testing earlier, the licensee was requested to examine causes of the apparent discrepancy, in addition to indicating a RCS pressure control accuracy) of 2.5 percent, the licensee attributed the j
discrepancy as due to (1 the slower pressurization rates in the RCS
't pressure test relative to the rapid pressurization rate-in the valve
. setting testing, and (2) the leakage of a steam / water mixture through the
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l valve seat resulting in uneven heating of the dissimilar material of the
. valve seat and body which is postulated to result in a earlier lif ting.
This explanation may have merit;' however, the staff is unable to make a determination that the actual PSV lif t setting Will be within ~5 percent of the valve setting.
However, considering the fact that (1) earlier j
analysis showed that, even without PSVs, the maximum RCS pressure would remain below 2750 psia with operability of one PORV and the reactor trip on turbine trip circuitry, and (2) the licensee indicated that menures will be taken to ensure operability of at least one PORV and the anticipatory i
reactor trip on turbine trip, there is reasonable assurance that the 110 percent design pressure criterion will not be exceeded even if the actual PSV setpoint increased by more than 5 percenc.
We therefore conclude that the TS change request for the remainder of Cycle 10 is acceptable.
- However, because of 'the uncertainty in the actual PSV lift pressure, we require that the licensee maintain the measures discussed above as compensatory VEPCO has counitted to continue to work with the NRC, industry measures.
and Owners Group to determine and expedite a satisfactory resolution to this generic issue in order to support the end of Cycle 10 application of this Technical Specification change.
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3.0 SUMMARf '
L The staff has reviewed the licensee's request for en emergency TS change h
to increase the PSV lift setpoint tolerance from +1 percent to +5 percent t
for the remainder of Cycle 10 operation for both Surry Units 1 and 2.
Based on-the licensee's safety analysis and its intended measures to -
ensure operability.of at least one PORV and the reactor trip on turbine trip circuitry, we hcve found the TS change request acceptable.
The. staff is currentb evaluating the PSV setting problem on a generic i
basis. The outcome of the staff generic evaluation for a long-ter'n so'ution will also apply to Surry Units I and 2.
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<4.0 EMERGENCY CIRCUMSTANCES In its Noven.ber 10, 1989 letter, VEPC0 requested thM t es bc treated on an emergency basis because, unless apprc would be required to shut down upon expiration of i ;
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. fnr o-ment period on December 1,1959 and Surry Unit 2 i 6
.o restart, currently-scheduled for November 23,19P i
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inforcation of a generic nature, on a shift in the ns f
pressurizer safety valves due to setpoint testing
- slogy, a
potential that the setpoint tolerance of the curren.
.;eral' t
. Unit I safety valves may exceed the 1 percent value required c
.nt i
Technical Specifications.
On October 19,1989, VEPC0 requested and was granted a discretionary enforcement to permit continued operation and to further evaluate this generic issue.
This-discretionary enforecment will t
expire on December 1, 1989. As previously stated, on November 6,1989
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during RCS pressure testing a Urit 2 PSV lifted prematurely at 2335 psig.
As a result of this premature lifting of the PSV, VEPC0 elected to have all v.
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three of the Unit 2 PSVs tested and reset using steam.
Subsequently,
l based on additional data obtained from testing of the Surry Unit 2 safety valves and re-analysis, VEPC0 submitted the subject proposed amendment dated November 10, 1989 stating that the proposed change would not result in reactor coolant system pressure exceeding the 110 percent design limit i
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specified in the UFSAR. Moreover, VEPC0 stated that additional measures would be taken by monitoring the operability of the pcwer operated relief valves and the anticipatory reactor trip on turbine trip circuitry. Thus, unless these amendments are promptly authorized, Unit I would be required l'
_to shut down on December 1,1989 and restart of Unit 2 would be delayed beyond the current schedule:! date
' November 23, 1989.
In accordance with 10 CFR 50.91(a)(5), VEPC0 has explained that it could not have avoided this emergency situacion since this generic concern was only recently identified.
Thus, the NRC staff does not believe that VEPC0 has abused the emergen~
.v' W ns in this instance.
According ly,
the Commissica has determ0
'4 are emergency circumstances warranting prompt approv' g
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5.0 TINAL No SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission's regulctions in 10 CFR 50.92 state that the Commission may take a final determination that a license amenoment involves no significant.
hazards considerations if Operation of the faulity, in accordance with the proposed changes would not:
1.
Involve a significaat increase in the probability-or consequences of any accident previously evaluated; or i
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated; or-3.
Involve a significant reduction in a margin of safety.
1his amendment has been evaluated against the standards in'10 CFR 50.92.
It does not involve a significant hazards consideration because the changes would not:
- 1. - Involve a significant-increase in the probability of occurrence or consequences of any accident or malfunction of equipment which is important to safety and which has been evaluated in the UFSAR. The proposed change effectively recognizes.the potential shif t in lift setpoint due to testing methodology. As such, the.setpoint shift being positive, the probability of a safety valve challenge:may be reduced.
The consequences of such a challenge are unaffected as the UFSAR aralysis remains bounding within the proposed setpoint tolerance.
-In addition, the Units 1 and 2 valve setpoint shift is expected to be in the same range as the Unit 2 valve test results (+3.5 percent to
+5 percent) and therefore no increase in the consequences of any accident or malfunction of equiptrent important to safety is tapected.
2.
Create the possibility of a new cr different type of accident from those previously evaluated in the safety analysis report. No modifi-cations are being made to the pressurizer safety valves' for either unit at this time.
Potential installation of temporary strap-on
' temperature instrumentation has no operational impact on valve perfor-
.mance. Capping.of loop seal ara 1ns is being performed only to ensure that the loop seals are not lost due to leakage through the drains anc hence has no impact on the intended design of the safety valves.
With the setpoint change expected to be in the same range as the Unit 2 valve' test results, there is no new or different kind of accident or accident precursors expected. Tt.c acaitional measures being imple-aented are only being used to further ensure that the system pressure will remain below'2750 psia (110 percent of design pressure) during any analyzed transient or operating condition.
3.
Involve a significant reduction in the margin of safety. Plant operations are not being changed.
Although accident analysis assump-tions have been modified to assume an initial 5.4 percent shift in i
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- i pressurizer safety valve lif t pressure, there is no reduction in the margin of safety since the 110 percent cesign pressure is not exceeded in any accident evaluated in the UFSAR.
For valve setpoint tolerance consistent with setpoint shift experienced during testing, the accident analysis remains bounding.
Accordingly, the Comission has determined that this amendment involves no significant hazards considerations, f
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6.0 STATE CON _SULTATION L
'In accordance with the Commission's regul.stions, the Connonwealth of v
Virginia representative was contacted and had no corsnents regarding issuance of this amendment.
7.0 ENVIRONMEhTAL CONSIDERATION This anendment charces a requirement with respect to the instal 16 tion or F
use of a fecility component located within the restricted area as oef;ned in 10 CFR Part 20. The staff has determined that these amendments involve-r.o significant increase in the amounts, and no significant change in tile 1
types, of any effluents that may be released offs 1te, and that there is no significant increase in inoividual or cumulative occupational raciation exposure.
The=Comission has made a final.no significant hazards consid-eration-finding with respect to this anendment. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 551.22(c)(9).
Pursuant to 10 CFR 551.22(b) no environmental impact statenient cr environmental essessment need be preparea in connection with
,a.
j' the issuence of the anenaments.
i-8.0- CONCLUSION L
L We have concluced, based.on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of t.e public l
will not be endangereo by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Coirmission's regulations ano thel issuance of the amenoments will rot be inimical to the connon defense and security or to the health'and safety of the public.
L 0ated:
November 16, 1989 Principal Contributog:
Y. HS11
- 8. Buckley
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