ML20087H785
| ML20087H785 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/31/1995 |
| From: | Benes G COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20046D708 | List: |
| References | |
| NUDOCS 9505040413 | |
| Download: ML20087H785 (15) | |
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March 31,1995 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
LaSalle County Nuclear Power Station Unit 2 Request for Emergency Technical Specification Amendment Facility Operating License NPF-18 Emergency Safety / Relief Valve Opening Setpoint Tolerance Change NRC Docket No. 50-374 Pursuant to 10 CFR 50.91 (a) (5), Commonwealth Edison (Comed) proposes to amend Appendix A, Technical Specifications, of Facility Operating License NPF-18, and requests that the Nuclear Regulatory Commission (NRC) grant an Emergency amendment to Section 3.4.2 to revise the safety / relief valve (SRV) safety function lift setting tolerances from +1%, -3% to i3% and include as-left SRV safety function lift setting tolerances of 11%. The amendment is needed by April 22,1995.
On March 18,1995, with Unit 2 in the Cold Shutdown condition for the L2R06 refueling outage, the results of offsite SRV safety mode lift setting tests concluded that two of the six SRVs which were tested, opened at pressures outside of the allowed tolerances.
This resulted in three more valves being sent offsite for testing per Technical Specification Section 4.0.5.b. On March 25,1995, the results of those tests showed that two of the additional three valves tested, also opened at pressures outside of the allowable tolerances. All of the four valves that failed, lifted at pressures greater than 1% above the nominal safety function lift settings. A summary of the valve testing data is enclosed with the safety analysis, as well as historical SRV testing data which demonstrates these four failures fall within the expected statistical distribution for as-found opening setpoints. )
l Based on these test results, which require Comed to test the remaining nine SRVs, !
Comed is requesting this emergency change be approved prior to Unit 2 Cycle 7 startup, >
which is currently scheduled for April 22,1995. This will allow Unit 2 startup without testing the remaining nine SRVs. This avoids a 4-5 day delay in Unit 2 startup, and approximately 11 person-rem in exposure. Comed requests this amendment be approved to allow startup without testing the final nine SRVs.
This proposal will have a minimum impact on safety. The change has been justified by an ASME Overpressurization analysis, an ATWS analysis, reviews of high pressure system performance, and an evaluation of LOCA containment response, all of which have been performed by General Electric (GE). In addition, Sargent and Lundy, who has the current design authority for piping loadings at LaSalle, has performed analyses justifying the increased loadings on main steam piping, including the SRV tailpipes. Comed has also evaluated the impact of this change on Minimum Critical Power Ratio, the Loss-of-Coolant Accident analysis of record, and Emergency Procedure Guidelines. The combination of 95o5o4o413 950331 4 & k DR ADOCK o50o vi
4 USNRC ( 2) March 31,1995 l
these above analyses constitutes an evaluation consistent with the BWR Owner's Group (BWROG) Licensing Topical Report for SRV setpoint tolerance relaxation, which is listed j as reference 2 in Attachment A. I I
Prior to these valves being tested, Comed was in the process of preparing an !
amendment package requesting the i3% setpoint tolerance increase in conjunction with a reduction in the number of installed SRVs for Units 1 and 2. Therefore, the attached GE analyses were performed with both the setpoint tolerance increase and a reduced number of SRVs. The analyses are conservative for application of the setpoint tolerance increase l alone, as described in the GE report.
The proposed amendment is needed and could not be avoided due to the short time frame which existed between Comed's notification on March 25 of the second sample test failures and the scheduled startup date for Unit 2 Cycle 7 operation. After the first two failures, Comed and the valve testing agent worked on expedited schedules to ensure the '
next three valve testings were completed as soon as possible. On March 25, when Comed l was notified two of the three subsequently tested valves failed, work was begun to remove the remaining nine valves. On March 27,1995 it was decided to expedite the setpoint tolerance Technical Specification change for Unit 2 only, based on previous industry ,
precedent to increase SRV tolerances to i3%, which has been approved by the Nuclear )
Regulatory Commission. Work to remove the nine additional valves was halted. l Since instituting several corrective actions stemming from previous SRV failures, LaSalle has had success in SRV testing. During the Unit 2 L2R05 refueling outage, a I second SRV valve test sample had to be initiated as required by Technical Specification l Section 4.0.5.b due to a failure in the first sample of tested SRVs. This second sample did not have a failure resulting in further testing. During the Unit 1 L1R06 refueling outage,13 SRVs were as-found tested without a failure. Also, entering L2R06, none of the 18 SRVs installed on Unit 2 were leaking. SRV leakage has been attributed to many of the past setpoint failures. Therefore, after the first two failures identified on March 18, Comed did not have reason to believe the second sample would also result in failures. Therefore, the need for this emergency change could not be avoided. This recent notification of the testing results, combined with the time necessary to develop this request will limit the time for the normal public comment period. This condition was not created by the failure to make a timely application for a Technical Specification amendment.
Without this Technical Specification Amendment, the removal, testing and replacement of nine additional SRVs will delay LaSalle Unit 2 Cycle 7 startup by 4-5 days and will require an additional 11 person-rem of radiation exposure.
l
4-USNRC- ~ ( 3) March 31,1995 ;
l This proposed emergency amendment request is subdivided as follows: l
- 1. Attachment A gives a description and safety analysis of the proposed l changes in this amendment. ,
1
- 2. Attachment B includes a summary of the proposed changes and the marked- ,
up Technical Specifications pages for LaSalle Unit 2 with the requested changes indicated. ,
- 3. Attachment C describes Comed's evaluation performed in accordance with 10CFR50.92(c), which confirms that no significant hazard consideration is l involved. ;
- 4. Attachment D provides an Environmental Assessment Applicability Review ,
per 10 CFR 51.21.
l
- 5. Attachment E is the General Electric SRV Setpoint Tolerance Relaxation j Analysis for LaSalle Units 1 and 2, which is being applied now for Unit 2 only. !
Relaxation Analysis Report.
1
- 7. Attachment G is the Sargent and Lundy analysis supporting the increased j main steam piping loads. ;
I
- 8. Attachment H is a non-proprietary version of the GE Analysis presented in Attachment E.
This request for an Emergenc, Technical Specification Amendment has been reviewed and approved by Comed Senior Management, as well as On-Site and Off-Site Review in accordance with Commonwealth Edison procedures.
The attached General Electric SRV Setpoint Tolerance Relaxation Analysis Report contains information proprietary to General Electric Company. In accordance with the requirements of 10CFR 2.790(b), an affidavit for this letter is enclosed as Attachment F to support the wdhholding of this report from public disclosure. In addition, a non-proprietary version (Attachment H) of the GE report is being transmitted.
To the best of my knowledge and belief, the statements contained above are true and correct. In some respect these statements are not based on my personal knowledge, but obtained information fumished by other Commonwealth Edison employees, contractor
USNRC ( 4) March 31,1995 employees, and consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
Commonwealth Edison is notifying the State of Illinois of this application for amendment by transmitting a copy of this letter and its attachments to the designated state ,
official.
Please direct any questions you may have concerning this submittal to this office. i Very truly yours, GB W Gary G. Benes Nuclear Licensing Administrator Subscribed and Sworn to before me on this T/ 5 # day of
'-W/v<A 1995. gcfg
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(."Not6ry Public A
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$ r kt nois somsm7 Attachments: ,
l A. Description and Safety Analysis of the Proposed Changes !
B. Marked-Up Technical Specification Pages C. Evaluation of Significant Hazards Considerations D. Environmental Assessment Applicability Review E. General Electric SRV Setpoint Relaxation Analysis l F. Withholding Affidavit for GE Analysis G. Sargent and Lundy Main Steam Piping Loading Analysis H. Non-Proprietary Version of GE Analysis j cc: J. B. Martin - Regional Administrator, Region ill P. G. Brochman - Senior Resident inspector, LaSalle County Station i W. D. Reckley - Project Manager, NRR Office of Nuclear Facility Safety -IDNS
, E ATTACHMENT A l 1
DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES l
l l
l DESCRIPTION OF THE PROPOSED CHANGE This Emergency Technical Specification Amendment request proposes to amend )
Appendix A, Technical Specifications, of Facility Operating License NPF 18, Section 3.4.2 to revise the Unit 2 safety / relief valve (SRV) safety function lift setting .
tolerances from +1%, -3% to i3%. As-left SRV safety function lift setting tolerances I will be specified to be within *1%.
DESCRIPTION OF THE CURRENT OPERATING LICENSE / TECHNICAL SPECIFICATION REQUIREMENT The La'.3alle Unit 2 Technical Specifications (Section 3/4.4.2) provide required safety valve' unction pressure lift settings for the 18 SRVs. The current allowed setpoint tolesance for these lift settings is +1 %, -3%.
BASES FOR THE CURRENT OF ERATING LICENSE / TECHNICAL SPECIFICATION REQUIREMENT There are two purposes of the SRV safety function lift setpoint tolerances. First, the SRVs prevent vessel overpressurization. The analysis of the ASME Overpressurization Event, which requires peak vessel pressures to remain below the upset transient limit of the ASME Code (1375 psig), relies on the safety mode setpoints of the available SRVs. Previous reload analyses for this event assumed the tolerance on this setpoint was +1%, -3%.
Second, by requiring the valves to perform within strict tolerances, reliable performance of the safety valves is assured. Some deviations in safety valve lift setpoints are expected and do not indicate decreased valve reliability. It is expected that some setpoint drift and data scatter will occur, resulting in a normal distribution i centered around the design setpoint. j l
DESCRIPTION OF THE NEED FOR AMENDING THE TECHNICAL SPECIFICATIONS ;
AND BASIS FOR EMERGENCY On March 18,1995, with Unit 2 in the Cold Shutdown condition for the L2R06 refueling outage, the results of offsite SRV safety mode lift setting tests concluded that two of the six SRVs which were tested, opened at pressures outside of the allowed tolerances. This resulted in three more valves being sent offsite for testing per Technical Specification Section 4.0.5.b. On March 25,1995, the results of those 4
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ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES tests showed that two of the additional three valves tested, also opened at pressures outside of the allowable tolerances All of the four valves that failed, lifted at pressures greater than 1% above the nominal safety function lift settings. A summary of the historical valve testing data through the current Unit 2 refueling outage is shown in Figure 1. Each test value shown in this graph represents the worst of the first four tests for a given valve during one testing cycle. Data demonstrate the historical as-found opening setpoints for LaSalle Station follow an expected statistical distribution.
In addition, industry precedent in interpreting ASME Section XI, which requires that SRVs be as-found tested, is that this as-found test consists of a single lift to determine if the setpoint is within tolerance. Any subsequent setpoint measurements are not used in determining as-found conditions, but are used for information or as-left certification. This position is supported by ANSI /ASME OM-1987 Part 1 and NUREG-1482 (draft). However, LaSalle's safety and relief valve testing procedure states that a valve shall be lift setpoint tested three times. If any one of the three tests are outside the tolerance, the valve shall be considered an as-found failure.
Of the six valves in the initial sample, whose results were obtained March 18, two were considered as-found failures per the LaSalle testing procedure. However, only one failed on the initial lift. The other valve was within tolerance on the first lift, but failed the second and third lifts (the lift pressure on the third lift was 3.3% above the valve setpoint). Of the three valves tested subsequently, two valves were considered failures by LaSalle's procedure, but only one failed during the initial setpoint test. Thus, due to LaSalle's conservative testing procedure, four SRVs were considered failures even though, by cument industry standards, only two SRVs would be considered failures. None of the first lift failures were above a +3%
criterion.
Based on these test results, which require Comed to test the remaining nine SRVs per Technical Specification 4.0.5, Comed is requesting this emergency change be approved prior to Unit 2 Cycle 7 (L2C7) startup, which is currently scheduled for April 22,1995. This will allow Unit 2 startup without testing the remaining nine SRVs. This avoids a 4-5 day delay in startup, and approximately 11 person-rem in exposure. Comed requests th:s amendment be approved to allow startup without testing the final nine SRVs.
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l ATTACHMENT A ,
l DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES l
l i The proposed amendment is needed and could not be avoided due to the short time ,
frame which existed between Comed's initial notification on March 25 of the second i sample test failures and the scheduled startup date for Unit 2 Cycle 7 operation. .j After the first two failures, Comed and the valve testing agent worked on expedited i schedules to ensure the next three valve testings were completed as soon as l possible. On March 25, when Comed was notified two of the three subsequently I tested valves failed, work was begun to remove the remaining nine valves. On !
March 27,1995 it was decided to expedite the setpoint tolerance Technical Specification change for Unit 2, only based on previous industry precedent to increase SRV tolerances to i3%, which has been approved by the Nuclear Regulatory Commission. Work to remove a nine additional valves was halted.
Since instituting several corrective actions vemming from previous SRV setpoint I tolerance failures, LaSalle has had succeu in SRV testing. During the fifth Unit 2 l refueling outage (L2R05), a second sample had to be initiated due to a failure in the j first sample of tested SRVs. This second sample did not have a failure resulting in further testing. During the sixth Unit i refueling outage (L1R06),13 SRVs were as- i found tested without a failure. Also, entering the current Unit 2 refueling outage :
(L2R06), none of the 18 SRVs installed on Unit 2 were leaking. SRV leakage has l been attributed to many of the past setpoint failures. Therefore, after the first two failures identified on March 18, Comed did not have reason to believe the second sample would also result in failures. Therefore, the need for this emergency change could not be avoided. This recent notification of the testing results, combined with l the time necessary to develop this request will limit the time for the normal public l comment period. However, as described above, this condition was not created by -
the failure to make a timely application for a Technical Specification amendment.
Without this Technical Specification Amendment, the removal, testing and replacement of nine additional SRVs will delay LaSalle Unit 2 Cycle 7 startup by 4-5 days and will require an additional 11 person-rem of radiation exposure.
DESCRIPTION OF THE AMENDED TECHNICAL SPECIFICATION REQUIREMENT The requested change would increase the SRV safety mode opening setpoint tolerances from +1%, -3% to i3%. The as-left opening setpoints would be required to be *1%. Therefore, all tested valves must meet the *1% criterion prior to being A-3 l
l ATTACHMENT A j DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES !
reinstalled. Additionally, any new valves would also be required to meet the t1%
critanon prior to being installed for use. This will ensure tested and new valves will continue to meet the high expectations for allowed initial setpoint deviations. ;
BASES FOR THE AMENDED TECHNICAL SPECIFICATION REQUIREMENT I Because the changed setpoint tolerances increase the possible opening pressures l of the SRVs, several analyses and evaluations have been performed by GE, Sargent and Lundy (S&L), and Comed to verify adequate safety and design margins l are maintained. The combination of these analyses constitutes an evaluation i consistent with the BWR Owner's Group (BWROG) Licensing Topical Report for l SRV setpoint tolerance relaxation given as reference 2. Discussions of the evaluations performed are included below. i l
- 1. ASME Overpressurization Event l
General Electric has performed an analysis (Attachment E) of the ASME l Overpressurization Event (MSIVF) assuming SRV opening setpoint tolerances of 1 i3%. Since this analysis was performed under the initial assumption that several SRVs would also be removed (part of the originally planned changes), the number of SRVs considered available for the analysis was 10 of the 18 currently installed (17 required operable). Although no SRVs will be removed for L2C7, and no SRV removals are being requested in this amendment, this analysis is bounding since the current SRV configuration with the proposed setpoint tolerance changes would provide a lower peak vessel pressure than that obtained with the GE assumed SRV removals.
The MSlVF analysis assumes that, upon closure of the MSIVs, the reactor is shut down by the high neutron flux scram. SRVs in the safety mode are used to mitigate the pressure increase. Credit is not taken for the relief mode of operation or the direct MSIV position scram. The GE analysis assumes a 102% power /105% flow condition, which supports LaSalle's current ability to operate in the extended operating domain and increased core flow regions.
The MSIVF analysis performed in Attachment E serves as both a generically i bounding analysis (which is required to be verified by a specific analysis prior to i each reload) and a reload-specific analysis for L2C7, since the L2C7 nuclear j dynamic parameters were used for input into ODYN, the NRC-approved GE l thermal-hydraulic and nuclear kinetics coupled transient code used in this analysis.
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. i ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES ]
)
As explained above, the reduced number of SRVs assumed in this analysis adds !
conservatism to the results.
The peak vessel pressure calculated in the GE MSIVF analysis is 1341 psig. This !
provides sufficient margin to the ASME upset code limit of 1375 psig. Table 3-1 and Figure 3-1 of Attachment E give detailed results of the transient response for this )
event. Table 3-1 compares the new analysis with the i3% setpoint tolerance to the L2C7 reload MSIVF analysis which had been performed by GE in preparation for the ;
L2C7 reload using the +1%, -3% criterion. Upon approval of this setpoint tolerance <
increase, the new GE analysis provided in Attachment E will supersede the analysis previously performed with the +1%, -3% criterion, and will be included as the L2C7 MSlVF analysis of record in the reload licensing documents. l l
- 2. High Pressure System Performance l l
GE has performed evaluations (Attachment E) of the impact of the increased-setpoint tolerance on the safety functions of the High Pressure Core Spray System (HPCS), the Reactor Core isolation Cooling System (RCIC), and the Standby Liquid ,
Control System (SLCS). Results of their evaluations are summarized below.
A. High Pressure Core Spray System Evaluation The impact of the SRV setpoint tolerance relaxation on HPCS is to raise the maximum pressure for HPCS injection from the current design value of 1162 psig (lowest SRV setpoint,1150 psig+1%) to 1185 psig (1150 psig+3%).
The HPCS system is designed to deliver water to the reactor vessel at a rate equal to or greater than 516 gpm in the event of a LOCA, a loss of all feedwater flow due to a transient, or a reactor isolation with a failure of the RCIC system.
The GE evaluation shows that the system was designed to operate at pressures greater than those introduced by this SRV opening setpoint tolerance change, and that there are no changes needed to the HPCS system to maintain its current design function. A Comed engineering review of HPCS pump curves verified the pump would meet the minimum required flow rate of 516 gpm at 1185 psig reactor pressure.
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- ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES P
B. Reactor Core isolation Cooling System Evaluation As described above for HPCS, the impact on RCIC operation from the i proposed SRV opening setpoint tolerance change is to increase the maximum pressure for RCIC injection.
Per NUREG-0519 (Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2, March 1981), Section 5.4.1, and Technical Specification 4.7.3.b, the RCIC pump is required to develop flow greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is supplied to the turbine at 1000 +20, -80 psig.
Due to the increase in injection prsssure, and to maintain the 600 gpm design requirement for injection, the pump and turbine shaft speeds will be increased from 4530 rpm to 4580 rpm.' These speeds are well within the capability of the system, and increases up to 4650 rpm have been previously approved for similar SRV performance improvement and power uprate programs, as stated in the Attachment E GE Analysis. A detailed evaluation of the effects of the increased pressures and temperatures on the RCIC system is included in Attachment E. In addition, Comed has evaluated the operability of the Unit 2 RCIC Motor Operated (MOVs), Air Operated (AOVs),
and Manual Valves with the increased dP of 1185 psi. The MOV reviews used actual test data and guidelines developed to ensure compliance with Generic Letter 89-10 to evaluate available margins. Engineering calculations were performed to ensure the AOVs would continue to operate as required.
These reviews concluded that all MOVs and AOVs listed in Section 4.2.4 of Attachment E will remain operable and capable of performing their intended safety functions with the increased dP. The two manual valves (2E51-F362 and 2E51-F363) are currently required to be closed and are thus not impacted by the increased dP.
Comed has installed a modification to the RCIC turbine steam admission valve, which addresses the concerns of GE SIL No. 377 discussed in Attachment E.
The RCIC turbine steam supply isolation flowrate setpoint, which is currently required to be less than 300% of steady state flow, provides isolation in the A-6
ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES ,
event of a leak. Because the steady state flow will increase, and no physical changes to this setpoint will be performed, isolation will now occur at less than 300% of steady state flow, which will continue to meet the above criterion.
The GE analysis states that the HPCS and RCIC system interfaces are not ;
affected by this change. i The above described evaluations, in conjunction with the Attachment E analysis and adjustment of the pump and turbine speeds ensures the original design criteria for the RCIC system continue to be met.
C. Standby Liquid Control System Evaluation I
The SLCS is not impacted by the proposed change to the safety mode !
setpoint tolerance of the SRVs. The SLCS is designed to inject to a !
maximum pressure based on the relief setpoint pressure of the SRVs. As no changes are being requested to the relief setpoint tolerances, the SLCS will continue to function as designed.
- 3. Containment Dynamic Loads l
GE has analyzed the impact of the proposed SRV setpoint tolerance change on ;
containment response during a LOCA (Attachment E). GE concluded that during l a Design Basis Accident LOCA there is no impact on containment pressure and I temperature and the peak suppression pool temperature since the vessel depressurizes without any SRV actuation.
For smaller breaks, the vessel remains at high pressure for a longer time period, ,
which can result in high drywell temperature conditions. When the vessel pressure !
reaches the SRVs relief mode opening setpoints, the SRVs will actuate to control 1 the vessel pressurization. Since the proposed amendment is only to the setpoint ;
tolerance of the safety function of the SRVs, there is no impact to containment temperature and pressure.
For LOCA hydrodynamic loads, such as pool swell, condensation oscillation and chugging are not impacted as pressure and temperature containment response is not affected.
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ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES GE also evaluated submerged pool boundary and submerged structure loads in the Attadiment E report. GE concluded that the increase to the SRV loads due to the proposed setpoint tolerance change would be less than 1%. This is acceptable based on NUREG-0802 (reference 8 of Attachment E). The increase in the submerged structure load was also determined to be less than 1%. This is much less than the margins used in the original LaSalle submerged structure load analysis (reference 7 of Attachment E). Therefore the proposed change will not impact the current design basis SRV hydrodynamic loads analyses results.
- 4. Main Steam Piping Loads Sargent and Lundy (S&L) evaluated the impact of the +3% tolerance on main steam piping loads (Attachment G). The systems evaluated were the four main steamline header subsystems, which include the SRV discharge piping up to the wetv/ ell penetration anchors. Also evaluated were the 18 wetwell discharge piping subsystems. These 22 piping subsystems are the only ones affected by the proposed SRV setpoint tolerance increase. S&L determined that, when conservatisms in the initial design input were taken into account, all 22 subsystems had the available margin required to accommodate the increased loadings.
- 5. ATWS Mitigation Capability GE has reanalyzed (Attachment E) the Main Steam isolation Valve Closure (MSIVC) ,
event under ATWS conditions to support the requested SRV setpoint tolerance relaxation. The MSIVC event is the limiting ATWS event in this analysis, the SLCS provides eventual shutdown of the plant, and the initial reduction in reactor power :
is by the ATWS Recirculation Pump Trip. The analysis assumes 100% power /105%
flow operating conditions. l As with the ASME Overpressurization Event analysis, the assumptions made by GE l include the removal of several SRVs. As stated before, this is a conservative j configuration relative to operatir'g with the current number of SRVs. This analysis I also assumes the upper analytical safety function settings of the SRVs is increased i to +3%. As an input assumption which serves only to add more conservatism to the analysis, GE also assumes the relief function settings are increased to +3%,
although no change to the relief function tolerances is being requested by Comed.
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i s.
l ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES I
GE analyzed this event using the REDY computer code. Ref6rance 10 of Attachment E is the topical report approved by the NRC for use of this code in analyzing ATWS events.
The results of the GE analysis show that the peak vessel pressure is 1457 psig.
This is well within the ASME emergency criterion of 1500 psig. Because this analysis assumed several SRVs removed, the peak vessel pressure with 18 SRVs (17 required operable) will be less than the pressure given above.
- 6. Minimum Critical Power Ratio impact Comed has evaluated the impact of the increased setpoint tolerance on Minimum Critical Power Ratio (MCPR) and concluded there is no adverse safety impact due to the prnpa=i change During the limiting reload licensing events for LaSalle, the MCPR occurs before the actuation of the lowest SRV setpoint.
SRV actuation is also required for FSAR events which are not reload licensing events. These transients are not included in reload analysis and are not limiting MCPR events because the core pressurization rates are slower and the power transient is tumed very quickly by scram reactivity. The transient ACPRs for these events are significantly lower than the ACPRs for the reload licensing events. In addition, the FSAR pressurization events and the reload licensing events (with the exception of the MSIVF event analyzed in Attachment E) credit relief valve operation, so the increased tolerances on the safety function opening setpoints for the SRVs will not impact the results of these analyses.
The thermal limits for L2C7, which are determined based upon results of the cycle-specific reload licensing analyses, are not impacted by the safety function setpoint tolerance increase. This will also be documented in the L2C7 reload licensing reports.
- 7. LOCA impact in Reference 1, the LaSalle LOCA analysis of record, the Automatic Depressurization System (ADS) is assumed to operate along with other ECCS functions to mitigate the consequences of LOCAs. In the analysis,6 of 7 ADS valves are assumed to operate. The ADS is used to depressurize the vessel, which allows low pressure ECCS functions to operate. As the increase in safety function A-9 e
= - _ _ _
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ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES setpoint tolerance for the SRVs does not impact the ADS setpoint, there is no impact on the LOCA analysis for the requested change.
- 8. Emergency Procedure Guidelines No credit is taken in any of the Emergency Procedure Guidelines for the safety setpoints of the SRVs. Therefore, no modifications to these guidelines are ;
necessary with the proposed changes.
l SCHEDULE I
Due to the previously mentioned SRV test results obtained on March 25,1995, !
Comed is requesting this emergency change be approved prior to Unit 2 Cycle 7 l startup, which is currently scheduled for April 22,1995. This will allow Unit 2 startup i without testing the remaining nine SRVs. This avoids a 4-5 day delay in startup and I approximately 11 person-rem in exposure.
REFERENCES
- 1. GE document, NEDC-32258P, "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-Of-Coolant Accident Analysis", October,1993.
- 2. General Electric Company, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report", NEDC-31753P, February 1990.
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LaSalle Station SRV Setpoint Performance Frequency Distribution 12 . .
10 - - - - - - -
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C
- s
. . . x R : : . m G
- 1 .
4 . _ _ _ _ . . . _ - . . ~ _ _ . . _ . - - - _ .
g 2= = :- -
0 l :::::: - b : x ! ;- : % :::::
-10 -8 -6 -4 -2 0 2 4 6 8 10 SRV Drift (% of Setpoint)
- of SRV Pop Tests -- + - Tech Spec Low Limit -3%
q -- + - Tech Spec Hi limit, + 1% - * - Proposed Tech Spec Hi Limit, +3%
_ _ _ _ - - _ . . . - - _ _ _ _ _ _ - - _ _ -_- -______-_- _____ _ -_ _ - _ __-__ _ _ ___- _ _-_ - -_ -_ - __