ML20093D425
| ML20093D425 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/02/1995 |
| From: | Benes G COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20093D429 | List: |
| References | |
| NUDOCS 9510130253 | |
| Download: ML20093D425 (18) | |
Text
Q>mmonwealth lilium Q>mpany Hm Opm Plxe,
ihmners Grove !!.60515 October 2,1995 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
LaSalle County Nuclear Power Station Unit 1 Request for Technical Specification Amendment Facility Operating License NPF-11 Safety / Relief Valve Opening Setpoint Tolerance Change NRC Docket No. 50-373 Pursuant to 10 CFR 50.90, Commonwealth Edison (Comed) proposes to amend Appendix A, Technical Specifications, of Facility Operating License NPF-11, and requests that the Nuclear Regulatory Commission (NRC) grant an amendment to Section 3.4.2 to revise the safety / relief valve (SRV) safety function lift setting tolerances from +1%,-3% to 3% and include as-left SRV safety function lift setting tolerances of 1 %.
This proposed amendment closely resembles the amendment approved for Unit i
2 on April 25,1995 (TAC No. M91926) for the same SRV safety function lift setting tolerance change.
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 3
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 tallpipes. 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 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 as reference 2 in Attachment A.
Prior to the earlier Unit 2 amendment, Comed was in the process of preparing an amendment package requesting the *3% setpoint tolerance Increase in conjunction with a reduction in the number of installed SRVs for Units 1 and 2. However, the setpoint tolerance increase alone is being requested to ensure consistency between the Unit 1 and 2 Technical Specifications. A future amendment for removal of several SRVs is still planned. Due to the original plan of a joint submittal, 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 alone, as described in the GE report.
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Octob:r 2,1995 Comed requests that this amendment be categorized as a Cost Benefit i
Licensing Action (CBLA). The approval of this amendment has the potential to save approximately $117,000 per refueling outage. These savings may be realized since this amendment will possibly avoid the testing of additional groups of SRVs during outages where the first groups' test results require additional SRV samples. These savings are based on past data of failures which indicates that with out this amendment being approved, Comed, on average, will have to test approximately an -
additional 2.6 SRVs every refueling outage. From past data it is also estimated that it costs approximately $45,000 for each SRV to be removed for testing, sent out to be tested, tested, and sent back to LaSalle and reinstalled. Additionally, worker dose will l
also be reduced, and the possibility of outage extensions due to SRV replacements will decrease.
This proposed amendment request is subdivided as follows:
1.
Attachment A gives a description and safety analysis of the proposed changes in this amendment.
2.
Attachment B includes a summary of the proposed changes and the marked-up Technical Specifications pages for LaSalle Unit 1 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 involved.
4.
Attachment D provides an Environmental Assessment Applicability Review per 10 CFR 51.21.
5.
Attachment E is the General Electric SRV Setpoint Tolerance Relaxation Analysis for LaSalle Units 1 and 2.
6.
Attachment F is a GE letter describing applicability of the Attachment E analysis to Unit 1.
7.
Attachment G is a withholding affidavit for the GE SRV Setpoint Tolerance Relaxation Analysis Report.
8.
Attachment H is the Sargent and Lundy analysis supporting the increased main steam piping loads.
9.
Attachment I is a non-proprietary version of the GE Analysis presented in Attachment E.
This request for a Technical Specification Amendment has been revie'wed and approved by Comed Senior Management, as well as On-Site and Off-Site Review in accordance with Commonwealth Edison procedures.
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USNRC (3)'
' Octobtr 2,1995 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 analysis.is enclosed as Attachment G to support the withholding of this report from public disclosure, in addition, a non-proprietary version (Attachment I) of the GE report is being transmitted.
Comed requests that NRC review of this proposed amendment be completed by Unit 1 Cycle 7 shutdown, which is scheduled to occur January 25,1996. This will j
allow for SRV testing during the outage to proceed with the appropriate acceptance l
criteria identified. Additionally, it is requested that this amendment request be i
j approved with an implementation time of 30 days.
i 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 furnished by other Commonwealth Edison
' employees, contractor employees, and consultants. Such information has been l
reviewed in accordance with company practice, and I believe it to be reliable, j
Commonwealth Edison is notifying the State of Illinois of this application for i
amendment by transmitting a copy of this letter and its attachments to the designated state official.
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Please direct any questions you may have concerning this submittal to this office.
j Sincerely, l
l i
6/3M j
Garv G. Benes l
Nuclear Licensing Administrator
]:! OFFICIAL SEAL:::::::: -::::::::::::=:=
Subscribed and Sworn to before me j
on this.h/
day of h
MARY JO YACK h(NOTARY PUsuc, ST ATE
&<M 1995.
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Notar/ Pud llo' ()
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USNRC (4)
October 2,1995 l
i Attachments:
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 F.
General Electric Letter Supporting Unit 1 Analysis Applicability
.G.
Withholding Affidavit for GE Analys s H.
Sargent and Lundy Main Steam Piping Loading Analysis 1.
Non-Proprietary Version of GE Analysis cc:
H.- J. Miller - Regional Administrator, Region lll P. G. Brochman - Senior Resident inspector, LaSalle County Station R. M. Latta - Project Manager, NRR Office of Nuclear Facility Safety - lDNS l
1 i
1 1
4 i
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l ATTACHMENT G
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4 WITHHOLDING AFFIDAVIT r
FOR l
4 GENERAL ELECTRIC SRV SETPOINT TOLERANCE 4
RELAXATION ANALYSIS REPORT i
1 1
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General Electric Company i
I, George B. Stramback, being duly sworn, depose and state as follows:
4 (1) I am Project Manager, Licensing Services, General Electric Company ("GE") and have been delegated the function of resiewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
2 (2) The information sought to be withheld is contained in the GE proprietary report GE-NE-B 13-01760, Safety Review for La Salle County Station Units I arul 2, Safety / Relief Valves Reduction arul Setpoint Tolerance Relaxation Aruslyses, Class III (GE Proprietary Information), dated March 1995. The proprietary information is delineated by bars marked in the margin adjacent to the specific material.
(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of i
Information Act ("FOIA"), 5 USC Sec. 552(bX4), and the Trade Secrets Act,18
]
USC Sec. 1905, and NRC regulations 10 CFR 9.17(aX4), 2.790(aX4), and 2.790(dXI) for " trade secrets and commercial or financial information obtained from j.
a person and privileged or con 6dential" (Exemption 4). The material for which i-exemption from disclosure is here sought is all " confidential commercial information",
and some portions also qualify under the narrower definition of" trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerny Proiect v. Nuclear Regulatory Commission.
975F2d871 (DC Cir.1992), and Public Citizen Health Research Grouo v. FDA.
704F2dl280 (DC Cir.1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
Information that discloses a process, method, or apparatus, including supporting a.
j data and analyses, where prevention ofits use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other' companies; b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; C:\\userdata\\winwordiproj\\LaSaAff' doc Affidavit Page I
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1 Information which reveals cost or price information, production capacities, c.
budget levels, or commercial strategies of General Electric, its customers, or its suppliers; d.
Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric; i
e.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.
(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held.
The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confident:e by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatcrv provisions or proprietary agreements.
(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical methods and processes, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of the safety relief valves for the BWR.
i C$userdata\\winword\\projbSaAff. doc AfDdavit Page 2
The development and approva :f :ne BWR computer codes used in this analysis was achieved at a significant cost, cc :n= order of several million dollars, to GE.
The development of the evwanuon process along with the interpretation and application of the analytical re:u!:s is derived from the extensive experience database that constitutes a major GE asm.
(9) Public disclosure of the inftrmzation sought to be withheld is likely to cause substantial harm to GE's com;e:invve position and foreclose or reduce the availability ofprofit-making opportunities Thne information is part of GE's comprehensive BWR safety and technology base, a:d.ms commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analyd:al methodology and includes development of the expertise to determine and ap;iy thne appropriate evaluation process. In addition, the technology base inciudes the va:uue derived from providing analyses done with NRC-approved methods.
The research, development, enginceering, anitytical and NRC review costs comprise a substantial investment of time and anoney by GE.
The precise value of the exprusae to devise an evaluation process and apply the correct analytical methodology is cziifficult to quantify, but it clearly is substantial.
GE's competitive advantage wil bee lost ifits competitors are able to use the results of the GE experience to normaf2e our verif their own process or if they are able to f
claim an equivalent understandng rny demonstrating that they can arrive at the same or similar conclusions.
The value of this information a GE would be lost if the information were disclosed to the public. Making such inf:rm=stion available to competitors without their having been required to undertake a cmmr expenditure of resources would unfairly provide competitors with a windfall, ano. deprive GE of the opportunity to exercise its competitive advantage to sei _ m adequate return on its large investment in developing these very valuable anaan,tical tools.
C:\\userdata\\winword\\proj\\LaSaAff. doc Affidavit Page 3
STATE OF CALIFORNIA
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COUNTY OF SANTA CLARA
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George B. Strambad, being duly sworn, deposes and says:
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That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.
Executed at San Jose, California, this A 9
- day of M
1995.
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Ge6rge B. Sframback 3
General Electric Company i
Subscribed and swom before me thisMbday of afc.k 1995.
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@ A T~ h Notary Public, State of CalifornA
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d ATTACHMENT A
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DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES i
DESCFilPTION OF THE PROPOSED CHANGE This Technical Specification Amendment request proposes to amend Aopendix A, Technical Specifications, of Facility Operating License NPF-11, Section 3.4.2 to revise the Unit 1 safety / relief valve (SRV) safety function lift setting tolerances from +1%, -3% to 13%. As-left SRV safety function lift setting tolerances will be specified to be within 1 %.
DESCRIPTION OF THE CURRENT OPERATING LICENSE / TECHNICAL SPECIFICATION REQUIREMENT The LaSalle Unit 1 Technical Specifications (Section 3/4.4.2) provide required d
safety valve function pressure lift settings for the 18 SRVs. The current allowed i
setpoint tolerance for these lift settings is +1%, -3%.
BASES FOR THE CURRENT OPERATING LICENSE / TECHNICAL SPECIFICATION REQUIREMENT 2
There are two purposes of the SRV safety function lift setpoint tolerances.
i 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%.
I Second, by requiring the valves to perform within strict tolerances, reliable j
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 centered around the design setpoint.
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DESCRIPTION OF THE NEED FOR AMENDING THE TECHNICAL SPECIFICATIONS j
This amendment is needed for Unit 1 to provide uniformity between the two units, as Unit 2 was approved for the increased tolerance limit on April 25, 1995. In addition, this amendment will possibly avoid the testing of additional groups of SRVs during outages where the first groups' test results require additional SRV samples. This avoids outage extensions and testing expenditures from being incurred, while still maintaining high expectations for valve performance.
DESCRIPTION OF THE AMENDED TECHNICAL SPECIFICATION REQUIREMENT The requested change would increase the SRV safety mode opening setpoint tolerances from +1%, -3% to 3%. The as-left opening setpoints would be required to be 1%. Therefore, all tested valves must meet the *1% criterion prior to being reinstalled. Additionally, any new valves would also be required to meet the 1% criterion 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 4
Because the changed setpoint tolerances increase the possible opening j
pressures of the SRVs, several analyses and evaluations have been performed by GE, Sargent and Lundy (S&L), and Comed to verity adequate safety and design margins are maintained. The combination of these analyses constitutes an evaluation consistent with the BWR Owner's Group (BWROG) Licensing Topical Report for SRV setpoint tolerance relaxation given as reference 2.
Discussions of the evaluations performed are included below.
1.
ASME Overpressurization Event General Electric has performed an analysis (Attachment E) of the ASME Overpressurization Event (MSIVF) assuming SRV opening setpoint tolerances of *3%. 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 L1C8, and no SRV removals are being requested in this amendment, this A2 i
i method 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.
i However, this analysis only serves as a generically bounding analysis, as this analysis is required to be performed each cycle. A specific analysis using the L1C8 neutronics parameters will be performed and verified to be acceptable j
through the 10CFR50.59 reload licensing process as usual, prior to L1C8 i
stariup. The additional conservatism of the removed valves provides extra assurance that the L1C8 reload specific analysis, with only the setpoint tolerance relaxation, will be found acceptable.
j The MSIVF analysis assumes that, upon closure of the MSIVs, the reactor is shutdown by the high neutron flux scram. SRVs in the safety mode are used to j
mitigate the pressure increase. Credit is not taken for the relief mode of i
operation or the direct MSIV position scram. The GE analysis assumes a 102% power /105% flow condition, which supports LaSalle's current ability to 1
operate in the extended operating domain and increased core flow regions.
The analysis was performed using ODYN, the NRC-approved GE thermal-hydraulic and nuclear kinetics coupled transient code. As explained above, the reduced number of SRVs assumed in this analysis adds conservatism to the l
results.
I 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.
l Table 3-1 and Figure 3-1 of Attachment E give detailed results of the transient j
response for this event. Table 3-1 compares the new analysis with the *3%
setpoint tolerance to the L2C7 reload MSIVF analysis which had been performed by GE in preparation for the L2C7 reload using the +1%, -3%
l criterion. This provides a comparison of the expected L1C8 results for MSIVF (which should be similar to the L2C7 reload analysis with the setpoint relaxation alone) and the generically bounding results which also include SRV removals.
2.
High Pressure System Performance GE has performed evaluations (Attachment E) of the impact of the increased a
setpoint tolerance on the safety functions of the High Pressure Core Spray i
System (HPCS), the Reactor Core Isolation Cooling System (RCIC), and the Standby Liquid Control System (SLCS). Results of their evaluations are summarized below.
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A.
High Pressure Core Spray System Evaluation j
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.
B.
Reactor Core Isolation Cooling System Evaluation As described above for HPCS, the impact on RCIC operation from the 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 pressure, and to maintain the 600 gpm design requirement for injection, the pump and turbine shaft speeds will i
be increased from 4530 rpm to 4580 rpm. These speeds are well within j
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
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4 addition, Comed has evaluated the operability of the Unit 1 RCIC Motor Operated (MOVs), Air Operated (AOVs), and Manual Valves with the 4
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 i
functions with the increased dP. The two manual valves (1E51-F362 and 1E51-F363) are currently required to be closed and are thus not impacted by the increased dP.
l Comed is installing a modification to the RCIC turbine steam admission i
valve, which addresses the concerns of GE SIL No. 377 discussed in Attachment E.
l 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 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, l
The GE analysis states that the HPCS and RCIC system interfaces are not affected by this change.
The above described evaluations, in conjunction with the Attachment E analysis and adjustment of the pump and turbine speeds ensures tho original design criteria for the RCIC system continue to be met.
C.
Standby Liquid Control System Evaluation The SLCS is not impacted by the proposed change to the safety made setpoint tolerance of the SRVs. The SLCS is designed to inject to a maximum pressure based on the relief setpoint pressure of the SRVa.
As no changes are being requested to the relief setpoint tolerances, the SLCS will continue to function as designed.
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3.
Containment Dynamic Loads GE has analyzed the impact of the proposed SRV setpoint tolerance change on l
containment response during a LOCA (Attachment E). GE concluded that during a Design Basis Accident LOCA there is no impact on containment pressure and temperature and the peak suppression pool temperature since the vessel depressurizes without any SRV actuation.
1 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 I
actuate to control 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 containment pressure and temperature response is not affected.
GE also evaluated submerged pool boundary and submerged structure loads in the Attachment 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 l
increase in the submerged structure load was also determined to be less than j
1%. This is much less than the margins used in the original LaSalle j
submerged structure load analysis (reference 7 of Attachment E). Therefore j
the proposed change will not impact the current design basis SRV l
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 H). The systems evaluated were the four main steamline header subsystems, which include the SRV discharge piping up to j
the wetwell penetration anchors. Also evaluated were the 18 wetwell discharge i
piping subsystems. These 22 piping subsystems are the only ones affected by the proposed SRV setpoint tolerance increase. S&L determined that, when l
l conservatisms in the initial design input were taken into account, all 22 subsystems had the available margin required to accommodate the increased loadings.
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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.
As with the ASME Overpressurization Event analysis, the assumptions made by GE include the removal of several SRVs. As stated before, this is a conservative configuration relative to operating with the current number of SRVs. This analysis also assumes the upper analytical safety function settings of the SRVs is increased 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 ComEi GE analyzed this event using the REDY computer code. Reference 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.
j 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 proposed 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 turned 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 A-7
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 L1C8 will be determined based upon results of the cycle-specific reload licensing analyses. The input parameters for these analyses take into account the increased SRV safety function tolerance. Results of these analyses will be documented in the L1C8 reload licensing reports through the normal reload 10CFR50.59 process.
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 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.
SCHEDULE Comed is requesting this change be approved prior to Unit 1 Cycle 7 shutdown, which is currently scheduled for January 25,1996. This will allow for SRV testing during the outage to proceed with the appropriato acceptance limit in place.
Additionally, it is requested that this amendment request be approved with an implementation time of 30 days.
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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, Februa:y 1990.
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