ML17311A512
| ML17311A512 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 12/19/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17311A511 | List: |
| References | |
| NUDOCS 9412280095 | |
| Download: ML17311A512 (12) | |
Text
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)ty*y0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY TH OFFICE OF NUC EAR REACTOR R GULATION RE ATED TO AMENDMENT NO.
TO FACILITY OP RATING ICENSE NPF-4 R ZONA PUBLIC S RVIC COMPANY T A PALO VERDE NUCLEAR GENERATING STATION UNIT 1
DOCKET NO'.
STN 50-528
1.0 INTRODUCTION
By letter dated November 22, 1994, the Arizona Public Service Company (APS or the licensee) submitted a request for changes to the Technical Specifications (TS) for the Palo Verde Nuclear Generating Station (PVNGS), Unit 1 (Appendix A to Facility Operating License No. NPF-41).
The Arizona Public Service Company submitted this request on behalf of itself, the Salt River Project Agricultural Improvement and Power District, Southern California Edison
- Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Water and
- Power, and Southern California Public Power Authority.
The proposed amendment would add a note to TS Table 3.7-2, The note would allow continuous operation of Unit 1 during Cycle 5 at 100X maximum steady state power with one NSSV inoperable per steam generator.
2.0 DISCUSSION TS Table 3.7-2 specifies maximum allowable steady state power level versus the number of inoperable main steam safety valves (NSSVs) on any steam generator.
The purpose of this TS is to ensure plant operation with sufficient pressure-relieving capacity to maintain the reactor coolant system (RCS) pressure below the staff acceptance criteria (120X of design pressure for large feedwater line breaks and control element assembly (CEA) ejection and llOX of design pressure for all other design basis overpressurization events) and maintain the peak main steam system pressure below 110X of design pressure during all design basis overpressurization events.
TS Table 3.7-2 currently limits Unit 1 to a maximum allowable steady state power level of 98.2X of rated thermal power because one HSSV in each steam generator is inoperable.
The licensee proposed to amend this TS by adding a note to TS Table 3.7-2.
The note will allow continuous operation of Unit 1 during Cycle 5 at 100X maximum steady state power with one NSSV inoperable per steam generator.
The
'icensee provided the results of an evaluation to support the proposed TS change.
The licensee uses a negative moderator temperature coefficient (HTC) of -1.0E-4 +/'F in its evaluation to offset the reduced, capacity of the 9'4i2280095 94i2i9 PDR ADOCK 05000528 P
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HSSVs.
This assumed negative HTC is bounded by the recent actual measured MTC of -2.039E-4 4o/'F at Unit l.
- 3. 0 EVALUATION The adequacy of the pressure relieving capacity.and the restriction specified in TS Table 3.7-2 is supported by the results of analyses documented in the updated final safety analysis report (UFSAR).
The current safety analyses for limiting pressure transients assume an HTC value of 0.0 Zp/'F at full rated power.
This is the maximum upper limit of HTC specified in the core operating limits report for Unit 1, Cycle 5.
However, the licensee states that the actual HTC at this point in the cycle for Unit 1, is less adverse than that assumed in the UFSAR.
The actual measured HTC for Unit 1, Cycle 5 at 292 effective full power day (EFPD) and 98.2X power (2/3 cycle HTC measured on October 20, 1994) was -2.039E-4 Zp/'F.
The predicted HTC at the end of the cycle is approximately -2.75E-4 dp/'F.
When the MTC becomes more negative, the temperature feedback is more negative and therefore has the effect of slowing down a heatup transient.
As a result, a lower transient peak RCS pressure is generated.
Because the actual HTC value in PVNGS, Unit 1, Cycle 5 is more negative. than that assumed in the existing analyzed pressure transients, the licensee has performed an evaluation which demonstrates that PVNGS, Unit 1, could operate at full power during the rest of Cycle 5 with one inoperable HSSV per steam generator and meet the acceptance criteria regarding allowable peak primary and secondary system pressure.
The licensee has reviewed the UFSAR to determine the design basis events (DBEs) that could be adversely affected'y the proposed TS change.
Based on a
review of these DBEs, the licensee has identified the loss of condenser vacuum (LOCV) and feedwater line break (FLB) as the most limiting transient and accident which require a detailed evaluation.
The CEA ejection event was also evaluated to provide an additional comparison of pressure response.
The reevaluation of the above events consisted of a sensitivity.study that quantified the benefit of the more negative HTC.
The sensitivity study involved running a base case which assumes the plant operates at full power with all 10 HSSVs per steam generator operable and an MTC of 0.0 Zp/'F.
A second case was run which assumes the plant operates at full power with one inoperable HSSV per steam generator and an HTC of -1.0E-4 Zp /'F.
The pressure transient results of the two cases were compared.
The following conclusions were reached from the licensee's sensitivity study:
(1) the peak RCS pressures for all three events analyzed in the second case are less than that in the base
- case, (2) the peak secondary system pressure for the FLB accident analyzed in the second case is less than that in the base
- case, and (3) the peak secondary system pressure for the LOCV and CEA ejection accidents analyzed in the second case is slightly higher than that in the base case.
However, the results of the second case are still within,the maximum allowable limit per the acceptance criteria.
The licensee also performed a reanalysis of a postulated small-break LOCA accident and concluded that the proposed TS change will not affect the results of the analysis in UFSAR.
The staff
!5 reviewed the proposed changes to TS and the supporting analysis and concluded that the proposed TS change is acceptable.
4.0 EXIGENT CIRCUMSTANCES
The Commission's regulations in 10 CFR 50.91 contain provisions for issuance of amendments with less than a 30-day comment period if either emergency or exigent circumstances are determined to exist.
Emergency situations involve those cases in which failure to act in a timely way results in the derating or shutdown of a nuclear power plant or prevents either resumption of operation or increase in power output up to the plant's licensed power level.
Under emergency circumstances, the Commission may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment.
In such a
situation, the Commission publishes a notice of issuance under 10 CFR 2. 106, providing for opportunity for a hearing and for public comment after issuance.
The processsing of an amendment under exigent circumstances usually applies to those cases in which the licensee and Commission must act promptly, but failure to act promptly does not involve a plant shutdown, derating, or delay in startup.
For both emergency and exigent circumstances, the licensee is required to explain the reason for the condition and why it could not be avoided.
This requirement is intended to prevent the abuse of the special provisions of 10 CFR 50.91(a)(6).
Under exigent circumstances, the Commission notifies the public in one of two ways:
by issuing a Federal
~Re ister notice providing notice of an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comment; or by us'ing local media to provide reasonable notice to the public in the area surrounding a
l.icensee's facility and providing special instructions for providing comment.
For this amendment
- request, the Commission employed the first approach with a Federal
~Re ister notice publised on December 2,
1994 (59 FR 61907),
which presented the staff's proposed no significant hazards consideration determination and requested public comment within 15 days of the date of publication of the notice.
Palo Verde Nuclear Generating Station (PVNGS) Unit 1 is currently operating at a reduced power level as a result of two inoperable HSSVs (one in each SG).
Per Table 3.7-1 of LCO 3.7. 1. 1, PVNGS Unit 1 is limited to a maximum Allowable Steady State Power Level of 98.2X of rated thermal power due to one inoperable MSSV in each SG.
The original basis for this restriction is to aver t challenges to the integrity of the RCS and secondary system pressure boundaries during the most severe anticipated overpressurization events by ensuring there is sufficient MSSV relieving capacity.
Consistent with TS 3. 1. 1.3, "HODERATOR TEMPERATURE COEFFICIENT," the current safety analyses for the most severe overpressurization events use the most adverse Moderator Temperature Coefficient (HTC) at 100K RATED THERMAL POWER.
Specifically, the safety 'analyses for the limiting pressure transients assume an HTC value of 0.0 Zdc/k/
F (i.e., 0.0 +/'F) at full rated power.
For the present time-in-cycle for PVNGS Unit 1, the actual HTC is less adverse than
0 4i
that assumed in the safety analyses.
The HTC for Unit 1, Cycle 5 measured at 43 EFPD and 85X power was -0.978 E-4 dp/'F (at 100X power it would be slightly more negative),
and 292 EFPD and 98.2X power (2/3 cycle HTC measured on October 20, 1994) was -2.039 E-4 Zp/ F.
The predicted end of cycle 'HTC, based on these actual HTCs, is approximately -2.75 E-4 Zp/'F.
As HTC becomes more negative, there is more negative temperature
- feedback, less power mismatch primary, to secondary, and a lower pressure peak.
Since the actual MTC value in PVNGS Unit 1 is less adverse than that assumed in the safety analyses, the reanalysis demonstrates that there is sufficient relieving capacity, without the one inoperable HSSV in each SG, to preclude exceeding the secondary or RCS pressure limit during the most, severe anticipated operational transient initiated from 102X rated thermal power.
- Hence, the current restriction on power is not justified and a return to 100X rated thermal power for the remainder of Unit 1, Cycle 5 is technically supported.
As such, APS is requesting approval of a one time amendment to the Unit 1 TS to allow full power operation with one HSSV inoperable per SG for the remainder of Unit 1, Cycle 5 (the next Unit 1 refueling, Cycle 6, is scheduled for April 1995).
This amendment is being requested on a exigent basis because the current condition (one inoperable MSSV per SG), is limiting Unit 1 to 98.2X power until the next refueling outage.
The staff has determined that the licensee and the Commission must act quickly to minimize the delay in returning the plant to operation at its licensed power level and that the licensee has not created this exigency.
Therefore, the staff is issuing this amendment on an exigent basis following a 15-day comment period as permitted by 10 CFR 50.91(a)(6).
The two inoperable MSSVs in PVNGS Unit 1 are currently gagged and are not physically able to open.
The valve gags are seismically evaluated and will not interfere with the operation of the remaining MSSVs.
As a prudent and precautionary
- measure, APS has reviewed the valve performance test results for the other 18 MSSVs currently installed in Unit 1 and the 40 MSSVs installed in PVNGS Units 2 and 3.
Specifically, valve performance during testing was reviewed.
Based on the review, no other MSSVs have been declared inoperable.
5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves. no significant hazards considerations if operation of the facility in accordance with the amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibil'ity of a new or a different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
Operation of the facility in accordance with the proposed amendment would not involves a significant increase in the probability or consequences of an
1,
accident previously evaluated.
The licensee stated that the primary pressure peaking events floss of condenser vacuum (LOCV), feedwater line break (FLB),
and control element assembly (CEA) ejection events]
were analyzed to provide a
comparison of pressure response using a base case with a moderator temperature coefficient (HTC) of 0.0 Zp/'F and ten operable main steam safety valves (HSSVs) per steam generator (SG) and a second case using an HTC of -1.0 E-4 dp/'F and nine operable HSSVs per SG.
The analyses performed confirmed that the existing safety analysis (i.e., the analysis of record) for PVNGS Unit 1, Cycle 5 will remain valid for 102X rated thermal power operation with one HSSV inoperable in each SG.
That is, the reactor coolant system (RCS) and secondary system design pressure limits will not be exceeded.
The analysis of the pressure peaking events was conservative and included the following:
(1) The actual HTC expected for full power operation for the remainder of PVNGS Unit 1, Cycle 5 is more negative, and thus more beneficial, than the
-1.0 E-4 4p/'F used in the reanalysis (actual HTC measured on October 20, 1994 was -2.039 E-4 Zp/'F).
Thus, the mitigating affect on peak system pressures would be expected to be even greater than those reported herein.
(2) The core parameters used in the reanalysis (other than MTC) are generic and selected in the most adverse direction.
Less adverse cycle specific or time-in-cycle specific values were not used in the reanalysis of PVNGS Unit 1, Cycle 5.
(3) The inoperable HSSVs are assumed to be in the first bank of HSSVs which have the lowest lift setpoint pressure (i.e.,
1303 psia).
In fact, one of the two HSSVs currently inoperable is from the third bank of HSSVs (with a higher lift setpoint of 1370 psia) and the other HSSV is in the first bank.
If the actual HSSV, lift setpoint pressures had been simulated, the results would be less adverse since there would be more relief capacity near the beginning'f the event to reduce the pressure peak.
Operation of the facility in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The licensee stated that the analyses performed demonstrates that the current licensing basis analyses results remain valid at 102X rated thermal, power with one HSSV inoperable in each SG and that all safety system settings will remain unchanged.
The Palo Verde TS currently allows operation at 98.2X Maximum Steady State Power Level (ACTION
- a. of Limiting Condition for Operation 3.7. 1. 1) with one inoperable HSSV per SG.
The analysis shows that for the current Unit 1 fuel cycle, operation at 102X Maximum Steady State Power Level with one inoperable HSSV per SG is acceptable.
Operation of the facility in accordance with the amendment will not involve a
significant reduction in a margin of safety.
The licensee stated that there is no reduction in the margin of safety since the analysis performed, crediting the remaining operable
- HSSVs, shows the results of the analysis of record remain, valid.
That is, the RCS and secondary system design pressure
Ik
limits will not be exceeded at 102X rated thermal power with one MSSV inoperable in each SG.
In addition, all other safety limits and safety system settings remain unchanged.
The actual MTC expected for full power operation for the remainder of PVNGS Unit 1, Cycle 5 is more negative, and thus more beneficial, than the -1.0 E-4 +/'F used in the reanalysis study (actual MTC measured on October 20, 1994 was -2.039 E-4 dp/'F).
Based upon the above considerations, the staff concludes.that the amendment meets the three criteria of 10 CFR 50.92.
Therefore, the staff has made a
final determinati'on that the proposed amendment
.does not involve a significant hazards consideration.
6.0 STATE CONSULTAT ON In accordance with the Commission's regulations, the Arizona State official was notified of the proposed issuance of the amendment.
The State official had no comments.
- 7. 0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.
The NRC staff has determined that the, amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released
- offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has made a final determination that the amendment involves no significant hazards considera-tion.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
- 8. 0 CONCLUSION The Commission has concluded, based on:the considerations discussed
- above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, '(2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
C. Liang December 19, 1994
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