ML20196J168
| ML20196J168 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/08/1998 |
| From: | Nunn D SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20196J132 | List: |
| References | |
| NUDOCS 9812090275 | |
| Download: ML20196J168 (10) | |
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i UNITED STATES OF AMERICA
. . NUCLEAR REGULATORY COMMISSION l
Application of SOUTHERN CALIFORNIA )
EDIS0N COMPANY, ET AL. for a Class 103 ) Docket No. 50-361 License to Acquire, Possess, and Use )
a Utilization Facility as Part of ) Amendment Application Unit No. 2 of the San Onofre Nuclear ) No. 183 l
Generating Station )
SOUTHERN CALIFORNIA EDISON COMPANY, ET AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 183.
l This amendment application consists of Proposed Change Number (PCN) 493 to Facility Operating License No. NPF-10. PCN 493 is a request to revise the Unit 2 Technical Specification (TS) 3.4.10, " Pressurizer Safety Valves,"
consistent with certain Updated Final Safety Analysis Report Chapter 15 events which were recently rear'alyzed. The proposed change increases pressurizer safety valve as-found setpoint tolerances.
Subscribed on this V day of [f mI M # 1998 Respectfully submitted, )
SOUTHERN CALIFORNIA EDIS0N COMPANY S
By --
unn DwightE.[
Vice Presi dent State of California ,
County of Am';a $lnDlkpLI On h before me, bb[ld6 ULNN b personally appeared hl ll/ldersonallyknowntometobethe person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument.
W u' d \lW V WITNESS m hand and offico seal. j ^ ^ ^ ^ Commission
^gmIsbc'd ^f f ; #1196482 l No' cry PutWic - Contamia l Signature NW
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L j w d$"3,zo2 9812090275 981200 PDR ADOCK 05000361
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UNITED STATES OF AMERICA ,
NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA )
EDISON COMPANY, ET AL. for a Class 103 ) Docket No. 50-362 License to Acquire, Possess, and Use )
a Utilization Facility as Part of ) Amendment Application Unit No. 3 of the San Onofre Nuclear ) No. 169 Generating Station l
)
SOUTHERN CALIFORNIA EDISON COMPANY, ET AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 169.
This amendment application consists of Proposed Change Number (PCN) 493 to Facility Operating License No. NPF-10. PCN 493 is a request to revise the Unit 3 Technical Specification (TS) 3.4.10, " Pressurizer Safety Valves,"
consistent with certain Updated Final Safety Analysis Report Chapter 15 events which were recently reanalyzed. The proposed change increases pressurizer ;
safety valve as-found setpoint tolerances. ;
Subscribed on this d day of A 0 Ph d V ( 1998 Respectfully submitted, SOUTHERNCALIFORNIAEDI{0NCOMPANY By __-
DwightE.j unn '
State of Californi
) Vice Presicent pyY County of_0 range- M Y On 9 beforeme,l t I N h personally appeared Id -
W tdersonally known to me to be the person whose name is subscribed to the within instrument and acknowledged to me that he executed the same in his authorized capacity, and that by his signature on the instrument the person, or the entity upon behalf of which the person acted, executed the instrument.
ofibit4 {n M -
WITNESS y hand and effise seal.
.O ,
f) j MANE sANCHEZ Commission #119Ma2 I
/ ) No' cry Puo6c - Corfamio [
Signatur
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DESCRIPTION OF PROPOSED CilANGE NPF-lb/15-493 PRESSURIZER SAFETY VALVE SETPOINTS SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 This proposed change to Technical Specification (TS) 3.4.10, " Pressurizer Safety Valves" is a request to incorporate increased as-found setpoint test values for Pressurizer Safety Valves (PSVs).
EXISTING TECIINICAL SPECIFICATIONS AND BASES Unit 2: See Attachment A Unit 3: See Attachment B PROPOSED TECHNICAL SPECIFICATIONS AND BASES CIIANGES -(Strike out for deletions and highlight for additions. Bases changes are provided for information only.)
Unit 2: See Attachment C Unit 3: See Attachment D PROPOSED TECHNICAL SPECIFICATIONS - (Changes incorporated)
Unit 2: See Attachment E Unit 3: See Attachment F UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) CIIAPTER 15 SAFETY ANALYSIS -(Strike out for deletions and highlight for additions. Provided for information only.)
Unit 2 and 3: See Attachment G DESCRIPTION OF CIIANGE:
This proposed change is a request to revise the as-found PSV setpoint tolerance for PSV operability in both the Technical Specification 3.4.10 Limiting Condition for Operation (LCO) and Surveillance Requirement (SR) 3.4.10.1. The proposed as-found setpoints for PSV operability are requested to be 2500 psia + 3% and - 2%. Technical Specification 3.4.10 currently requires the setting and as-found PSV setpoint tolerance to be
- 1% of the specified value.
This PCN is contingent upon the approval of PCN-470, which is a request to reduce the allowable pressurizer water volume for pressurizer operability, submitted for NRC approval by Southern California Edison (SCE) on December 19,1997. The basis of PCN-470 is the change in
- Technical Specification 3.4.9 maximum allowable value to 57% to accommodate up to a 4.0%
pressurizer level Total Loop Uncertainty (TLU). The reanalysis basis of PCN-470 assumed an l analytical value of 2.0% for the PSV tolerance, based on a nominal set pressure of 2500 psia.
The reanalysis basis of PCN-493 increases the maximum as found setpoint tolerance from +2%
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to +3%. The Anticipated Transient Without Scram (ATWS)/ Diverse Scram System (DSS) has been analyzed to establish a revised ATWS/ DSS pressurizer pressure trip setpoint which will be implemented upon approval of this Technical Specification change. The revised ATWS/ DSS pressurizer pressure trip setpoint will maintain San Onofre Units 2 and 3 compliance with the ATWS rule 10CFR50.62 by ensuring that the ATWS/ DSS Pressurizer pressure trip occurs between the Plant Protection System (PPS) High Pressurizer Pressure Trip (HPPT) and the PSV opening setpoint, including tolerances. The DSS setpoint remains below the PSV as found setpoint to ensure that the ATWS/ DSS plant trip signal occurs prior to the PSV opening.
The proposed change also provides the summary results of the reanalysis of Updated Final Safety Analysis Report (UFSAR) Chapter 15 events affected by the PCN-493 changes in PSV setpoint tolerance.
DISCUSSION A. Background During refueling outages, to facilitate required setpoint testing, the PSVs are removed and tested.
As previously reported in LERs 2-95-010,3-96-002,2-97-003, and 3-97-003, when the valves were tested their as-found setpoints were found to be outside the Technical Specification (TS) required setpoint band of 2500 psia 1%.
No unusual conditions were found when the valves were disas< cabled, inspected, reassembled, and reset to the proper setpoint. SCE personnel observed the testing and resetting of the PSV setpoints, which were conducted properly by the testing facility. Therefore, SCE believes the cause of the PSVs being outside the required setpoint band to be setpoint drift.
To minimize TS violations caused by setpoint drift, SCE has performed the necessary analyses to support expansion of the TS tolerances, as recommended in the enclosure to NRC memorandum
" Task Interface Agreement: Interpretation of Reporting Requirements - 93TIA006 (TAC No.
M86339) to S. L. Collins (Region IV) from E. G. Adensam (N'RC Regions IV and V Director) dated November 2,1993. The enclosure to this 11/2/93 NRC memorandum is titled,
" Assessment of Various Reporting Requirements for Applicability to Primary or Secodary Safety Valves Found Outside Technical Specification Acceptable Tolerance Band." The reanalyses performed conservatively support the expansion of the PSV as found setpoint tolerance to + 3%
and - 2%.
UFSAR Chapter 15 events that are sensitive to pressurizer primary safety valve set point tolerances were reanalyzed to accommodate the proposed PSV as found setpoint tolerance of
+ 3% and - 2%. All Chapter 15 events that result in Pressurizer insurges have been evaluated.
The resulting impacted events are presented here. They are Loss of Condenser Vacuum with and without the Concurrent Single Failure of an Active Component (UFSAR Sections 15.2.2.3 and 15.2.1.3), the Chemical and Volume Control System (CVCS) Malfunction with and without the I
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Concurrent Single Failure of an Active Component (UFSAR Sections 15.5.2.1 and 15.5.1.1), and Feedwater System Pipe Breaks (UFSAR Section 15.2.3.1).
B. Analyses The reanalysis of these events was performed using the approved CESEC-III computer code.
(Topical Repcrt CE-CES-78 Rev 0.-P, CESEC User's Manual, ABB/CE Computer Services, May 1987). The analytical value assumed for the PSV tolerance was + 3.0% and - 2.0%, based on a nominal set pressure of 2500 psia. The reanalysis also included the analytical value of 61%
for the pressurizer level (57% Technical Specification value and up to 4.0% TLU), which is the basis of the previously submitted PCN-470.
The plant high pressurizer pressure trip setpoint is 2375 psia. In previous analyses the setpoints were 2437 psia for non-environmental events and 2450 psia for environmental events. To accommodate the larger PSV tolerances, the total loop uncertainty analysis was reperformed. As a result of the reanalysis, a lower total loop uncertainty was identified which allowed a change in the analysis setpoints to 2410 psia for non-environmental events and 2434 psia for environmental events. There has been no change to the actual" Plant" trip setpoint.
The results of the analysis are summarized as follows:
- 1. UFSAR Section 15.2.1.3 Loss of Condenser Vacuum - Summary of Analysis The Loss of Condenser Vacuum (LOCV) Event is classified as a moderate frequency incident.
LOCV may occur due to failure of the main condenser evacuation system to remove noncondensible gases, or excessive leakage of air through a turbine gland packing.
The turbine generator trip that occurs due to a LOCV would normally generate an immediate reactor trip signal from the turbine stop valves (through unitized actuator pressure monitors).
If credit is not taken for reactor trip on turbine trip, reactor trip would occur as a result of high-pressurizer pressure. The turbine bypass valves are unavailable following a loss of condenser vacuum due to the actuation of the condenser vacuum interlock on the turbine generator trip. The feedwater pumps would trip on low suction pressure soon after turbine trip. It is conservatively assumed that feedwater flow is terminated immediately after turbine trip. The pressure increases in the primary and secondary systems following reactor trip are limited by the pressurizer and steam generator safety valves. Following turbine trip, offsite power is available to provide AC power to the auxiliaries. The operator action to mitigate the event is assumed to occur 30 minutes after the initiation of the event. The consequences of this event are bounded by the LOCV with single failure.
- 2. UFSAR Section 15.2.1.3 Loss of Condenser Vacuum with a Concurrent Single Failure of an Active Component - Summary of Analysis
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. 1 The Loss of Condenser Vacuum with a Concurrent Single Failure of an Active Component (LOCV yr) Event is classified as an infrequent incident. LOCV may occur due to failure of the main condenser evacuation system to remove noncondensible gases or excessive leakage of air through a turbine gland packing. The most adverse single failure following LOCVsr is the failure of the pressurizer level measurement channel. This failure is assumed to produce a false low level signal, resulting in the activation of both standby charging pumps and closing I of the letdown control valve to its minimum flow area.
1 The system and the reactor trip which operate following LOCV rs are the same as those l described in item 1 for LOCV. The dynamic behavior of the Reactor Coolant System (RCS) following LOCV rs is similar to that of LOCV. This transient produces the maximum RCS pressure for this frequency classification, and therefore results in the most severe challenge, however, the maximum RCS pressure and the secondary pressure both remain below the I acceptance criteria of 110% of design pressure for an incident of this frequency, and no water will be released through the pressurizer safety valves for the maximum RCS pressure case.
- 3. UFSAR Section 15.2.3.1 Feedwater System Pipe Breaks - Summary of Analysis The feedwater system pipe break is classified as a limiting fault event. The initiating event is a break in a pipe in the main feedwater system. A rupture of a feed line will cause rapid reduction of the liquid inventory in the affected steam generator and therefore partial loss of the secondary heat sink. This leads to the heat up of the RCS and an increase in RCS pressure. A reactor trip could occur through either a Low Steam Generator Water Level '
Trip, a Low Steam Generator Pressure Trip, or a High Pressurizer Pressure Trip. Loss of non-emergency AC power was assumed at the time of reactor trip. Operator action to mitigate the event is assumed to occur 30 minutes after initiation of the event. Due to the increase in PSV setpoint tolerance from + 2% to + 3%, in combination with the reduction in the pressurizer high pressure trip setpoint, the peak RCS pressure will decrease from 2893.7 psia to 2832.2 psia. The RCS pressure will remain below the acceptance criteria of 120% of design pressure for an incident of this frequency, and no water will be released through the pressurizer safety valves for the maximum RCS pressure case. !
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- 4. UFSAR Section 15.5.1.1 Chemical and Volume Control System (CVCS) Malfunction Summary of Analysis The CVCS malfunction is classified as a moderate frequency event. The initiating malfunction is a failure of the pressurizer level control system which willinitiate all 3 charging pumps and isolate letdown. Various pressurizer level and pressure control actuations will terminate this event. A pressurizer high-level signal will trip the charging pumps and energize the pressurizer heaters resulting in an increase in pressurizer pressure. The high pressure trip setpoint has been reduced to 2410 psia (from 2437 psia) to compensate for the increase in PSV setpoint tolerance from + 2% to + 3%. The peak RCS pressure will increase from 2629 psia to 2644 psia, remaining below the acceptance criteria of 110% of design pressure, and no water will be released through the pressurizer safety valves for the maximum RCS pressure case.
- 5. UFSAR Section 15.5.2.1 Chemical and Volume Control System Malfunction with a Concurrent Single Failure of an Active Component - Summary af Analysis The CVCS malfunction with a single failure is classified as an infrequent event. The results are similar to those discussed in Item 4 above with the exception of the single failure. The worst case single failure postulated for this event is the loss of offsite power at the time of reactor trip. Various pressurizer level and pressure control actuations will terminate this event. A pressurizer high-level signal will trip the charging pumps and energize pressurizer heaters resultin;;in an increase in pressurizer pressure. The high pressure trip setpoint has been reduced to 2410 psia (from 2437 psia) to compensate for the increase in PSV setpoint tolerance from + 2% to + 3%. The peak RCS pressure will change from 2629 psia to 2625 psia, remaining below the acceptance criteria of 110% of design pressure, and no v'ater will be released through the pressurizer safety valves for the maximum RCS pressure case.
C. Plant Operation All of the UFSAR Chapter 15 events were evaluated to determine which events would be sensitive to the proposed change in pressurizer primary safety valve set point tolerance. The impacted events were reanalyzed to assure that they could accommodate the change in PSV as-found setpoint tolerance to + 3% and - 2%. Expanding the TS tolerances for the PSVs helps minimize possible future TS violations. As discussed in the description, this analysis is contingent upon the proposed maximum pressurizer operation level reduction to 57% (PCN-470). There will be no change in the operation of San Onofre Unit 2 or 3 as a result of this TS PCN-493 change.
NO SIGNIFICANT IIAZARDS CONSIDERATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (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. A discussion of these standards as they relate to this amendment request follows:
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1 (1) Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated? ,
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Resoonse: No
- All Updated Final Safety Analysis Report (UFSAR) Chapter 15 events have been 3 evaluated to determine the impact of the increases in as found Pressurizer Safety Valve j- - (PSV) tolerance from +1% and -l% to + 3% and - 2%. The events that result in ;
challenging the opening of the PSVs are Loss of Condenser Vacuum With and Without i Single Failure, Loss of Normal Feedwater Flow, Feedwater System Pipe Breaks, Total Loss of Reactor Coolant System (RCS) Flow, Uncontrolled Control Element Assembly i (CEA) Withdrawal, CEA Ejection, Chemical and Volume Control System (CVCS)
Malfunction With and Without Single Failure, Inadvenent Emergency Core Cooling System (ECCS) Actuation With and Without Single Failure, and Inadvertent Opening of a i PSV. Of these, the limiting events are the Loss of Condenser Vacuum (LOCV), Loss of ;
- Condenser Vacuum with a Concurrent Single Failure of an Active Component (LOCVsf), '
CVCS Malfunction, CVCS Malfunction With a Concurrent Single Failure of an Active Component, and Feedwater System Pipe Breaks. These limiting events have been reanalyzed for the wider PSV tolerance. For all the reanalyzed events it is assumed that
- plant operation is maintained at a maximum pressurizer level of 57%. For the CVCS
- Malfunction With and Without Single Failure Events and the Inadvertent ECCS Actuation With and Without Single Failure Events, it is also assumed that the operator can respond ;
within 15 minutes to mitigate the event. l The change in as found PSV tolerance from -1% to -2% results in the. earlier opening of '
the PSV's for the analyzed events. To compensate for this earlier opening of the PSVs the j high pressurizer pressure trip analysis setpoint was reduced from 2437 psia (non-harsh 1 environment) and 2450 (harsh environment) to 2410 psia (non-harsh environment) and i 2434 (harsh environment). These setpoint changes insure that the high pressurizer pressure trip is actuated sufficiently early before the opening of the PSV's such that no liquid is released through the PSV's. Therefore, the change to the PSV negative tolerance does not result in a significant increase in the probability or consequences of any previously evaluated accident.
The change in PSV as found tolerance from +1% to +3% results in the later opening of the PSV's for the analyzed events. The PSV actuation to mitigate the consequences of the analyzed accidents are thus delayed. However, the lowering of the high piessurizer pressure trip setpoint, as discussed above, mitigates the increase in peak primary pressure and assures that no liquid is released through the PSV's. Therefore, this change to the PSV positive tolerance does not result in a significant increase in the probability or consequences of any previously analyzed design basis event.
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There are no other changes to the plant equipment or operation which could create an
, increase in the probability or consequences of any event previously evaluated.
l Therefore, operation in accordance with this proposed change will not involve a significant increase in the probability or consequences of any previously evaluated accident. 1 l
(2) Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
l Response: No I 1
l Operation in accordance with this proposed change will not involve any change to plant
! equipment or operation which could create a new or different kind of accident. The as-left PSV tolerance will continue to remain at +/- 1%. The change in as-found tolerance of the PSV's to -2% and +3% will not introduce the possibility of a new or different kind of l accident because evaluation of the design basis events shows that no water is expected to 1 l be released through the PSV's.
There are no other changes to the plant equipment or operation which could create the possibility of a new or different k.ind of accident from any accident previously evaluated.
l Therefore, this proposed change will not create the possibility of a new or different kind of l accident from any accident previously evaluated.
(3) Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No l
l Operation in accordance with this proposed change will not change the manner in which l safety limits, limiting safety settings, or limiting conditions for operation are determined.
l The acceptance criteria for all of the events reanalyzed include an appropriate margin of l safety. l l
There are no changes to the acceptance criteria nor are the acceptance criteria exceeded for these events assuming plant operation at a maximum pressurizer level of 57% and l
operator response time of 15 minutes for the CVCS Malfunction With and Without Single Failure Events and the inadvertent ECCS Actuation With and Without Single Failure Events.
Therefore, this proposed change does not involve a significant reduction in a margin of safety.
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Based on the responses to these three criteria, Southern California Edison (SCE) has concluded ;
that the proposed amendment involves no significant hazards consideration.
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ENVIRONMENTAL CONSIDERATION '
t SCE has determined that the proposed amendment involves no changes in the amount or type of j efiluent that may be released offsite, and results in no increase in individual or cumulative ;
occupational radiation exposure. As described above, the proposed amendment involves no i significant hazards consideration and, as such, meets the eligibility criteria for categorical ;
exclusion set forth in 10 CFR 51.22(c)(9). l i
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