ML20199G165

TS Change Request 98-08-0 to Licenses NPF-39 & NPF-85, Proposing Rev to TS Section 3/4.4.2 & TS Bases Sections B 3/4.4.2,B 3/4.5.1 & B 3/4.5.2.Supporting TSs & Proprietary & non-proprietary Trs,Encl.Proprietary Info Withheld

ML20199G165
Person / Time
Site:	Limerick
Issue date:	01/12/1999
From:	Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20136B419	List: ML20199G165 ML20199G202 ML20199G237
References
NUDOCS 9901220205
Download: ML20199G165 (26)

Text

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I Ctauen support oeputment l A 10 CFR 50.90 v

PECO NUCLEAR mco u, c_ ,

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I January 12,1999 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF 85 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Limerick Genera mg anon, Units 1 and 2 Technical Specificdcr., Change Request No. 98-08-0 Increase Safety Relief Valve Setpoint Tolerance From 11% to 13%

Dear Sir / Madam:

PECO Energy Company is submitting Technical Specifications (TS) Change Request No.

98-08-0, in accordance with 10CFR50.90, requesting an amendment to the TS (Appendix A) for Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS),

Units 1 and 2. This proposed change involves revising TS Section 3/4.4.2, " Safety / Relief Valves," and TS Bases Sections B 3/4.4.2 and B 3/4.5.1 and B 3/4.5.2, to increase the allowable as-found main steam Safety Relief Valve (SRV) code safety function lift setpoint tolerance from 11% to i3%. This change will also require the as-left SRV code safety function lift setting to be set within i1% of the specified nominallift setpoint prior to reinstallation in the plant. In support of this proposed TS change, the required number of OPERABLE SRVs in Operational Conditions (OPCONs) 1,2, and 3 will be changed from 11 to 12. The number of SRVs in each lift pressure .

grouping will remain the same. This proposed TS change does not alter the SRV nominal lift I.

setpoints or the SRV lift setpoint test frequency currently specified by TS Section 3/4.4.2. The l' propo' sed change does not change the SRV testing commitment specified in LGS Updated Final f Safety Analysis Report (UFSAR) Chapter 5.2.2.10, " Inspection and Testing."

The proposed TS changes are consistent with guidance specified in Boiling Water Reactor Owners' Group (BWROG) document NEDC-31753P, "BWROG In-Service Pressure Relief w i I

Technical Specification Revision Licensing Topical Report," which was developed to support the use of a 13% safety lift setpoint tolerance for SRVs. NEDC-31753P has been reviewed and approved by NRC as documented in its Safety Evaluation Report (SER) issued by letter dated March 8,1993. The NRC determined that it is acceptable for licensees to submit TS amendment requests to revise the SRV code safety function lift setpoint tolerance to 13%,

provided that the setpoints for those SRVs tested are restored to i1% prior to reinstallation.

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J:nuiry 12,1999 Paga 2 In addition, General Electric (GE) performed plant specific analyses and evaluations in support of this proposed TS change request, the results of which have been documented in NEDC-32645P, " Limerick Generating Station Units 1 and 2, SRV Setpoint Tolerance Relaxation Licensing Report," Revision 2, dated December 1998. A copy of NEDC-32645P is provided in Enclosure 1. This document contains information of a proprietary nature to GE. Therefore, in accordance with the requirements of 10CFR2.790(a)(1)(i)(4) we are requesting that this document be withheld from public disclosure since it contains information pertinent to trade secrets and commercial or financial information considered to be privileged or confidential. In accordance with the requirements of 10CFR2.790(b)(1) the required affidavit supporting this request is contained in Enclosure 2 of this letter.10CFR2.790(b)(1)(ii) stipulates that "...the information sought to be withheld shall be incorporated, as far as possible, into a separate l

paper " Therefore, Enclosure 3 to this letter contains a non-propnetary version of NEDC-32645P.

The proposed TS changes have been endorsed by the industry and the NRC. The proposed TS changes are similar to other changes previously approved by the NRC (e.g., Fermi). In addition, approval of this proposed TS change will help reduce the number of non-safety significant Licensee Event Reports (LERs) reported at LGS, Units 1 and 2.

Information supporting this TS Change Request is contained in Attachment 1 to this letter, and the proposed TS pages (including marked-up pages) showing the proposed changes to the LGS, Units 1 and 2, TS are contained in Attachment 2. This information is being submitted under affirmation, and the required affidavit is contained in Attachment 3.

We request that, if approved, the amendments to the LGS, Units 1 and 2, TS be issued by May 5,1999, and become effective within 30 days of issuance in order to support startup activities associated with the upcoming LGS, Unit 2, refueling outage (2R05).

If you have any questions, please do not hesitate to contact us.

Very truly yours, i fi

/ Garrett D. Edwards

/ Director- Licensing Attachments / Enclosures cc: H. J. Miller, Administrator, Region I, USNRC (w/ attachments / enclosure)

A. L. Burritt, USNRC Senior Resident inspector, LGS (w/ attachments / enclosure)

R. R. Janati, PA Bureau of Radiological Protection (w/ attachments / enclosure)

Janucry 12,1999 Page 3 '

l bec: G. R. Rainey - 63C-3 w/o attachments / enclosures j J. D. von Suskil - LGS, SMB1-1

• l J. J. Hagan - 62C-3 "

l. M. P. Gallagher - LGS, GML5-1 l J. P. Grimes - LGS, SSB3-1 '"

T. A. Moore - LGS, SSB2-4 "

H. J. Ryan - LGS, SSB2-3 . w/ attachments, enclosures K. A. Hudson - 63B-3 l B. D. Dothanczyk - LGS, SSB2-4 l D. P. Helker - 62A-1 PA DEP BRP inspector - LGS, SSB2-4 Commitment Coordinator- 62A-1 Correspondence Control Desk - 61B-5 DAC - 61B-5 l

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General Electric (GE) Affidavit .

Supporting Submittal of Proprietary Version of Licensing Report NEDC-32645P t

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General Electric Company AFFIDAVIT I, David J. Robare, being duly sworn, depose and state as follows:

(1) I am a Technical Account Manager, Technical Projects, General Electric Company

("GE") and have been delegated the function of reviewing the information described i in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.  !

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(2) The information sought to be withheld is contained in the GE proprietary report GE- l NEDC-32645P, Limerick Generating station Units 1 and 2 SRV Setpoint Tolerance Relaxation Licensing Report, Class III (GE Company Proprietary Information),

dated December 1998. The proprietary information is delineated by bars marked in the margin adjacent to the specific material. l 1

(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 ,

Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 l USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and i 2.790(d)(1) for " trade secrets and commercial or financial information obtained from l a person and privileged or confidential" (Exemption 4). The material for which l 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 Energy Project v. Nuclear Regulatory Commission.

975F2d871 (DC Cir.1992), and Public Citizen Heallh Research Group v. FDA.

704F2dl280 (DC Cir.1983).

'(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting 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, installatien, assurance of quality, or licensing of a similar product; 12'13 93RTil Afridavit Page 1

. _ . _ . . ~ . ._ _ _ - . _ _ . _ . _ _ _ . _ _ _ . . _ _ . _ . _ -. _

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I L c. Information which reveals cost or price information, production capacities, l L budget levels, or commercial strategies of General Electric, its customers, or its j suppliers; l

d. Information which reveals aspects of past, present, or future General Electric -

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. customer-funded development plans and programs, of potential commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

L The information sought to be withheld is considered to be proprietary for the reasons l set forth in both paragraphs (4)a. and (4)b., above.

(5) The informatien 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. l The information sought to be withheld has, to the best of my knowledge and belief, I consistently been held in confidence by GE, no public disclosure has been made, and l 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 i 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, anJ licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because i ' it contains detailed results of analytical models, methods and processes, including compu;er codes which GE has developed, obtained NRC approval of, and applied to perform evaluations of the loss-of-coolant accidents and transients for the BWR.

l W D ".sRTI' Af0 davit Page 2

The development and approval of the BWR loss-of-coolant accident and transient analysis computer codes used in this analysis was achieved at a significant cost, on l

- the order of several million dollars, to GE. 1 The development of the evaluation process along with the interpretation and l application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

I (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original l development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a j substantial investment of time and money by GE. l l l l The precise value of the expertise to devise an evaluation process and apply the l l correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost ifits competitors are able to use the results of l the GE experience to normalize or verify their own process or if they are able to i

claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide l competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequa! return on its large investment in developing these very valuable analytical tools.

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STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

David J. Robare, being duly sworn, deposes end says:

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.

T6 Executed at San Jose, California, this // day of ACEMBER 1998.

YN&w David J. Robare General Electric Company Subscribed and sworn before me thisEc ((' SdayN#cifof 1998.

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.42 6 Notary Public, State of California "N >

Cc. Tem; e itaggy d Nokry PubNc-CoMornks (

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1213 93RTil Affidavit Page 4 l l

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ATTACHMENT 1 Limerick Generating Station Units 1 and 2 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 Technical Specifications Change Request No. 98-08-0

" Increase Safety Relief Valve Setpoint Tolerance From i1% to 13%"

Supporting Information for Change - 16 Pages

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l Dock:t Nos. 50-352/50-353 Attachm nt 1 January 12,1999 Page 1 of 16 l l

l Limerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 98-08-0 l Addition of Special Test Exception for Inservice Leak and Hydrostatic Testing i

l Subject l PECO Energy Company, licensee under Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, requests that the Technical Specifications (TS) contained in I Appendix A to the Operating Licenses be amended as proposed herein to permit changes in the main I steam Safety Relief Valve (SRV) setpoint tolerance. This proposed TS Change Request involves revising TS Section 3/4.4.2," Safety / Relief Valves," and associated TS Bases Sections B 3/4.4.2 and B 3/4.5.1 and B 3/4.5.2, to increase the allowable as-found safety / relief valve (SRV) code safety function lift setpoint tolerance from 11% to 3%. This change will also require the as-left SRV code safety function lift setting to be set within 11% of the specified nominallift setpoint prior to reinstallation in the plant.

These proposed changes do not alter the SRV nominal lift setpoints or the SRV lift setpoint test frequency currently specified by TS Section 3/4.4.2. The proposed changes do not change the SRV testing commitment specified in the LGS, Units 1 and 2. Updated Final Safety Analysis Report (UFSAR) Chapter l 5.2.2.10. " Inspection and Testing." l The proposed changes to the LGS, Units 1 and 2, TS are shown by vertical bars in the margins, as applicable, on the affected TS pages, and are contained in Attachment 2. Marked-up pages indicating the changes are also contained in Attachment 2.

We request that,if approved, the TS changes proposed herein be issued by May 5,1999, and become j effective within 30 days of issuance in order to support the upcoming LGS, Unit 2, refueling cutage 1 (2R05). l l

This TS Change Request provides a discussion and description of the proposed TS changes, a safety I assessment of the proposed TS changes, information supporting a finding of No Significant Hazards l Consideration, and information supporting an Environmental Assessment.

Discussion and Description of the Proposed Chanaes This proposed TS Change Request involves revising TS Section 3/4.4.2, and associated TS Bases  !

Sections B 3/4.4.2 and B 3/4.5.1 and B 3/4.5.2, to increase the allowable as-found SRV code safety l function lift setpoint tolerance from i1% to 3%. The proposed TS change will also require the as-left SRV code safety function lift setting to be set within 11% of the specified nominal lift setpoint prior to reinstallation in the plant. In support of the increase in the allowable as-found SRV code safety function lift setpoint tolerance, the required number of OPERABLE SRVs in Operational Conditions (OPCONs) 1,2, and 3 will be changed from 11 to 12. The number of SRVs in each lift pressure grouping will remain the same. This proposed change does not alter the SRV nominal lift setpoints or the SRV lift setpoint test l frequency currently specified by TS Section 3/4.4.2. The proposed changes do not change the SRV testing commitment specified in UFSAR Chapter 5.2.2.10. With the exception of increasing the Reactor Core Isolation Cooling (RCIC) turbine / pump maximum rated speed (controller setting change), which is required to support this proposed change, no other physical plant changes are being made.

The proposed TS changes have been endorsed by the industry and the NRC. The proposed TS changes are similar to other changes previously approved by the NRC (e.g., Fermi). In addition, approval of this proposed TS change will help reduce the number of non-safety significant Licensee Event Reports (LERs) reported at LGS, Units 1 and 2.

l Dock:t Nos. 50-352/50-353 Attachm:nt 1 i January 12,1999 Page 2 of 16 l

Safety Assessment l

Backaround The 1% allowable as-found SRV code safety function lift setpoint tolerance currently specified in TS j l Section 3/4.4.2 for LGS, Units 1 and 2, is based on the acceptance criteria originally defined by the  ;

l American Society of Mechanical Engineers (ASME), Section lil, NB-7000. ASME Section til is mainly  :

I used for the design of new nuclear plants and new components and structures installed in nuclear plants.  !

l Existing plant equipment is normally within the jurisdiction of ASME,Section XI, requirements. ASME Section XI. Subsection IWV-3500 endorses ANSI /ASME OM-1-1981," Requirements for Inservice l 1

Performance Testing of Nuclear Power Plant Pressure Relief Devices," for the testing of SRVs.

j ANSI /ASME OM-1-1981 specifies a 3% acceptance criteria.

The use of the 11% allowable as-found SRV code safety lift setpoint tolerance in plant TS is generic in the  ;

industry. As a result, the Boiling Water Reactor Owners' Group (BWROG) developed NEDC-31753P, l "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report," to I support the use of the 13% SRV code safety lift setpoint tolerance consistent with that specified in ASME Section XI requirements.

NEDC-31753P was reviewed and approved by the NRC as documented in a Safety Evaluation Report (SER) issued by letter dated March 8,1993. The NRC determined that it was acceptable for licensees to submit TS amendment requests to revise the SRV code safety function lift setpoint tolerance to 13%,

provided that the setpoints for those SRVs tested are restored to 1% prior to reinstallation. The NRC also indicated in its SER that licensees planning to implement TS changes to increase the SRV setpoint tolerances should provide the following plant specific analyses:

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1. Transient analysis, using NRC approved methods, of abnormal (anticipated) operational occurrences (AOOs) as described in NEDC-31753P utilizing a 13% setpoint tolerance for the safety mode of the SRVs. (Note: LGS UFSAR designates these events as abnormal operational transients (AOTs.)

2. Analysis of the design basis overaressure event using the i3% tolerance limit for the SRV setpoints to confirm that the vessel pressure coes not exceed ASME pressure vessel code upset limits.

3. Plant specific analyses described in items 1 and 2 should assure that the number of SRVs included in l the analyses corresponds to the number of valves required to be operable in the Technical  ;

Specifications.

4. Re-evaluation of the performance of high pressure systems (pump capacity, discharge pressure, etc.),

motor-operated valves, and vessel instrumentation and associated piping considering the 3%

tolerance limit.

5. Evaluation of the 3% tolerance on any plant specific alternate operating modes (e.g., increased core flow, extended operating domain, etc.).

6. Evaluation of the effects of the 13% tolerance limit on the containment response during loss-of-coolant accidents (LOCA) and the hydrodynamic loads on the SRV discharge lines and containment.

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Dock:t Nos. 50-352/50-353 Attachm:nt 1 January 12,1999 Page 3 of 16 l Analysis in support of this proposed TS Change Request, plant specific analyses and evaluations, as described above, were preformed by General Electric (GE) and documented in NEDC-32645P," Limerick Generating Station Units 1 and 2, SRV Setpoint Tolerance Relaxation Licensing Report," Revision 2, dated December 1998 or in PECO Energy generated documents. A copy of NEDC-32645P is included in Enclosure 1.

(Note: This document is deemed to be of a proprietary nature to GE and is requested to be withheld from public disclosure). The following discussion provides a summary of the results of these analyses.

1. Analysis of Anticipated Operational Transients (AOTs)

An evaluation of the effects of pressurization transients on the fuel thermallimits was performed to determine whether the increase in SRV code safety function lift setpoint tolerance from 11% to 3% would be acceptable. The Minimum Critical Power Ratio (MCPR)is the most significant thermallimit parameter for this evaluation. A review of the transient analysis identified the Feedwater Controller Failure (FWCF) with Failure of the Bypass System and Load Rejection with Bypass Failure (LRNBP) as being limiting.

Both events assumed the End-of-Cycle (EOC) recirculation pump trip (RPT) was Out-of-Service (OOS).

Both events have comparable MCPR, heat fluxes, and neutron fluxes. The LRNBP with EOC RPT OOS event was used to determine the impact of the SRV code safety function lift setpoint tolerance change on thermallimits. The results are based on an analysis for LGS, Unit 2, Cycle 4. However, the conclusions are generic for both LGS, Units 1 and 2.

For the LRNDP with EOC RPT OOS event, peak neutron flux occurs at 0.77 seconds, peak surface heat flux occurs at 0.9G seconds and MCPR occurs at 1.22 seconds. With uniform SRV drift of +i% above nominal, the first SRV actuation occurs at 1.43 seconds. Therefore, increasing the setpoint to +3% will not affect the calculated peak noutron flux, peak surface heat flux, or the MCPR. Hence, for SRV setpoints greater than the nominal setpoint, increasing the SRV code safety function lift setpoint tolerance to +3%

will not have an impact on fuel thermal limits. With any number of SRVs OOS, the conclusion still applies since the peak pressure occurs after the MCPR occurrence.

For a decrease in SRV opening pressure of -3% below the nominal setpoint, the analysis shows that the drift.will cause earlier actuation of the SRVs. An earlier actuation of SRVs would reduce the rate of vessel pressurization and, therefore, decrease the rate of void collapse if SRV actuation occurs at or before the time of the MCPR, the decreased rate of pressurization and void collapse will produce a lower peak neutron flux, a lower peak surface heat flux, and therefore a smaller change in MCPR. Therefore, lowering the SRV code safety function lift setpoint tolerance to -3% will either not affect the MCPR or will result in a less limiting MCPR. Hence, for SRV setpoints lower than the nominal setpoint, decreasing the SRV code safety function lift setpoint tolerance to -3% will not have an impact on fuel thermal limits. The number of SRVs OOS does not affect the timing of the SRV actuation and, hence does not effect thermal limits.

For the FWCF with EOC RPT OOS event, the minimum MCPR also occurs prior to the SRV actuation due to the high water level turbine trip and the anticipatory SCRAM associated with the turbine trip. Therefore, as in the case of the LRNBP with EOC RPT OOS event, the 3% SRV code safety function lift setpoint tolerance will not have any impact on the thermal limit foi the FWCF with EOC RPT OOS event. With any l number of SRVs OOS, the conclusion still applies since the peak pressure occurs after the MCPR occurrence.

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l Dock t Nos. 50-352/50-353 Attachm:nt 1 January 12,1999 Page 4 of 16

2. Reactor Vessel Pressure Overpressure Prutection The ASME Pressure Vessel Code requires that the peak vessel pressure remains less than 110% of the vessel design pressure. The design pressure of the LGS reactor vesselis 1250 psig.110% of the design pressure is 1375 psig. TS Safety Limit 2.1.3, " Reactor Coolant System Pressure," requires that the reactor vessel steam dome pressure not exceed 1325 psig. The limiting overpressure event is the Main Steamline isolation Valve (MSIV) Closure with Flux Scram event. This event assumes the failure of the MSIVs not fully open Scram. The reactor is scrammed by the high neutron flux Scram caused by the vessel pressurization and the resultant collapse of moderator voids within the reactor core.

GE has performed an analysis (Enclosure 1) of this event assuming three (3) SRVs with the lowest nominal lift setpoint were OOS. The analysis assumed 102% core thermal power,110% core flow, and conservative end of Unit 2, Cycle 4, nuclear dynamic parameters. The setpoint of each of the SRVs was assumed to be 3% above its nominal setpoint. The NRC approved GE thermal hydraulic and nuclear kinetics coupled transient code (ODYN) was used to obtain the system response and peak calculated vessel pressure. The peak calculated reactor vessel bottom head pressure is 1348 psig which is below the ASME limit of 1375 psig. However, the peak calculated steam dome pressure is 1329 psig which exceeds the TS Safety Limit of 1325 psig. The LGS, Unit 1, results are expected to be the same due to i similarity of the units.

The MSIV Closure with Flux SCRAM event was re-analyzed (Enclosure 1) with the assumption that two SRVs with the lowest nominal lift setpoint were OOS. All other assumptions mentioned above remained the same. The peak calculated reactor vessel bottom head pressure is 1338 psig which is below the ASME limit of 1375 psig and the peak calculated steam dome pressure is 1318 psig which is below the Technical Specification Safety Limit of 1325 psig. Therefore, LGS, Unit 2, satisfies the ASME Code limit and the TS Safety Limit with a SRV code safety function lift setpoint tolerance of 13% using 12 of the 14 SRVs. The Unit 1 results are expected to be the same due to similarity of the units.

3. Number of SRVs As previously discussed in item 1 above, the number of SRVs OOS does not affect the fuel thermallimits since the timing of the actuation of an SRV is not affected. As discussed in item 2 above, the allowable number of SRVs to be OOS will be reduced from three (3) to two (2) in order to maintain the peak calculated reactor vessel bottom head pressure below the ASME Code limit and the peak calculated steam dome pressure below the TS Safety Limit. TS Section 3/4.4.2, and associated TS Bases Section B 3/4.4.2, will be revised to reflect this change.

4. High Pressure System Performance The impact of the increased setpoint tolerance on the safety functions of the Reactor Core Isolation Cooling (RCIC) system, the High Pressure Coolant injection (HPCI) system, and the Standby Liquid Control (SLC) system was evaluated. The impact on the LGS, Units 1 and 2, Generic Letter 89-10 Motor.

Operated Valve (MOV) Program was also evaluated. The most significant impact is the increased reactor vessel pressure specified for system operation. The RCIC, HPCI, and SLC systems are currently evaluated to provide injection into the reactor pressure vessel at up to 1182 psig (lowest SRV nominal setpoint of 1170 psig +1%). System performance was reevaluated (Enclosure 1) as a result of this proposed TS change for injection et 1205 psig (lowest SRV nominal setpoint of 1170 psig +3%).

Docket Nos. 50-352/50-353 Attachment 1 January 12,1999 Page 5 of 16 RCIC System The current design basis for the RCIC system is that it must be capable of injecting design rated flow into the reactor vessel at a maximum reactor vessel pressure equal to the lowest SRV nominal setpoint plus the allowable setpoint tolerance. The RCIC system analysis (Enclosure 1) requires that the RCIC turbine / pump maximum rated speed be increased from 4575 rpm to 4625 rpm in order for the RCIC system to have the capability to inject its design flow rate of 600 gpm into the reactor vessel at a pressure of 1205 psig (lowest SRV nominal setpoint of 1170 psig +3% setpoint tolerance). The increase in the turbine / pump speed is within the performance capabilities of the RCIC system. The time increase to reach the higher maximum rated speed is expected to add less than 0.2 seconds to the overall response time of the RCIC system (Enclosure 1). The RCIC system is designed to reach rated flow within 30 seconds; however, the response time used in event analyses is 55 seconds. Therefore, this increase in response time is negligible. The increased turbine / pump rated speed reduces the overspeed trip margin from 123% of rated speed plus a 2% margin for trip setting to 122.1% of rated speed plus a 2% margin for trip setting. The RCIC turbine / pump overspeed setpoint will not be changed; therefore, the potential for a RCIC turbine / pump failure (missile hazard or system overpressurization) is not increased. This reduction in the margin to the overspeed trip is acceptable because of the currently demonstrated minimal amount of speed overshoot during startups. The installation of plant Modification P00210,"RCIC System Startup Transient Improvement," partially opens the steam admission valve for approximately 7 seconds to allow the turbine to reach idle speed to prevent a turbine overspeed trip. The increased turbine / pump rated speed can be achieved with the currently specified steam flowrate; therefore, a change to the RCIC steamline break instrumentation is not required. The operating pressure and temperature of the RCIC system components directly impacted by this change, including valves, piping, and instrumentation, do not exceed design values. The impact of the increased RCIC turbine / pump maximum rated speed on the RCIC system motor operated valves have been evaluated using the guidance of the LGS Generic Letter 89-10 MOV Program and determined to be acceptable. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not have an adverse effect on the operation of the RCIC system.

HPCI System The current design basis for the HPCI system is that it must be capable of injecting design rated flow into the reactor vessel at a maximum reactor vessel pressure equal to the lowest SRV nominal setpoint plus the allowable setpoint tolerance. The HPCI system analysis (Enclosure 1) indicates that the HPCI turbine / pump maximum rated speed must be increased in order for the HPCI system to be capable of injecting 5600 gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint of 1170 psig +3%

tolerance). However, the HPCI system analysis also indicates that if the HPCI turbine / pump speed is increased the HPCI discharge piping pressure would exceed currently specified design values during a system overpressurization event (discharge line isolation). As concluded in the HPCI System analysis (Enclosure 1), summarized below, the current HPCI system capabilities are acceptable and no chenges to the HPCI system are required. The HPCI system is currently capable of providing 5600 gpm at a reactor pressure of 1182 psig (lowest SRV nominal setpoint of 1170 psig +1% tolerance) and 5400 gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint of 1170 psig +3% tolerance) (Enclosure 1).

The HPCI system serves both an Emergency Core Cooling System (ECCS) function and a non-ECCS function. The ECCS function is to provide adequate injection of coolant into the reactor vessel during a small break Loss-of-Coolant Accident (LOCA) until the reactor is depressurized and injection is provided by low pressure ECCS. The non-ECCS function is to serve as a backup to the RCIC System by providing adequate injection of coolant to maintain reactor vessel water level during abnormal operational transients and special events. The small break LOCA is the limiting event for required HPCI System flow.

Dock:t Nos. 50-352/50-353 Attachment 1 January 12,1999 Page 6 of 16 The HPCI system flow input parameter to the current LGS SAFER /GESTR-LOCA analysis is 5400 gpm over an operating reactor vessel range of 200 to 1141 psig. However, for the full spectrum of breaks analyzed, acceptable results were obtained without crediting any injection from the HPCI system. The spectrum of breaks included various sizes of line breaks and equipment failures, including the failure of an Automatic Depressurization System (ADS) valve. Therefore, the reduction in the HPCI system flow to less than 5600 gpm at reactor vessel pressure between 1182 psig and 1205 psig will not impact the current LOCA analysis. TS Surveillance Requirement (SR) 4.5.1.b.3 requires that the HPCI system be capable of developing 5600 ppm against a test line pressure which corresponds to a reactor vessel pressure of 1040 psig plus head and line losses. This SR has been evaluated as be!ng adequate to ensure that the HPCI system is capable of meeting design requirements. Since the ECCS-LOCA analysis is not impacted by this proposed TS change, no revision to TS SR 4.5.1.b.3 is required. However, the Bases for TS Section 3/4.5.1, *ECCS-Operating and Shutdown, will be revised to clarify the HPCI system capability at a reactor vessel pressure range of 1182 to 1205 psig.

As described in LGS UFSAR Chapter 6.3, the HPCI system is capable of maintaining reactor vesse! water level above the reactor core and preventing an ADS actuation for line breaks of less than 1 inch nominal diameter will be maintained at the reduced flow. The HPCI system flow achievable over the reactor vessel pressure operating range of 200 psig to 1205 psig is more than adequate to makeup the reactor vessel water inventory lost due to a break of a 1 inch diameter line.

For other events that utilize the HPCI system in a non-ECCS capacity, analyses (Enclosure 1) determined that 5000 gpm is adequate for events that result in a reactor vessel pressure between 1182 psig and 1205 psig. The HPCI system is more than capable of providing 5000 gpm at a reactor vessel pressure of 1205 psig (lowest SRV nominal setpoint +3% tolerance) with no change to the system. The operating pressure and temperature of the HPCI system components directly impacted by thic change, including valves, piping, and instrumentation, do not exceed design values. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not have an adverse effect on the operation of the HPCI system.

The design basis of the HPCI system will be revised to specify the required HPCI system flow rate is 5600 gpm for a reactor vessel pressure range or 200 psig to 1182 psig and 5000 gpm for a reactor pressure range of 1182 psig to 1205 psig.

SLC System The SLC system capability of shutting down the reactor during a postulated event in which all or some of the control rods cannot be inserted or during a postulated /snticipated Transient Without Scram (ATWS) event is not impacted by this proposed change. The current design basis for the SLC system is that it must be capable of injecting the system design rated flow into the reactor vessel using a single pump at a maximum reactor vessel pressure equal to the lowest SRV nominal setpoint plus the allowable setpoint tolerance. The SLC system must also be capable of meeting the requirements of 10CFR50.62,

" Requirements far Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants," for a postulated ATWS event. During a postulated event in which all or some of the control rods cannot be inserted (non-ATWS), the SLC system has been evaluated to be capable of performing its design basis function (Enclosure 1). As discussed later, the ATWS event is a low probability event for which the use of nominal system operating parameters for event anaiysis has been accepted. Since the nominal SRV setpoints are not being changed, this proposed change does not affect the capability of the SLC system to mitigate the consequences of an ATWS event.

Dock t Nos. 50-352/50-353 Attachment 1 January 12,1999 Page 7 of 16 Motor-Operated Valves (MOVs)

The impact on MOVs due to the potential for increased reactor vessel and system pressure as a result in the increase in SRV setpoint tolerance has been evaluated in accordance with the LGS Generic Letter 89-10 MOV Program and has been determined to be acceptable. The LGS Generic Letter 89-10 MOV Program currently uses SRV nominal setpoints for differential pressure determinations for valves in which reactor pressure at the SRV setpoint is limiting. Use of nominal SRV setpoints is consistent with current industry practice.

5. Altemate Operating Modes Evaluation The attemate operating modes, including Maximum Extended Load Line Limit (MELLL), increased Core Flow, and Single Loop Operation (SLO) were considered in determining the most restrictive analytical conditions (i.e., most limiting operating mode) for performing the analysis associated with the proposed TS change (Enclosure 1).

6. Containment Response During LOCA and the Hydrodynamic Loads on SRV Discharge Lines and Containment Containment Response during LOCA The most limiting event in terms of peak containment pressure and temperature, and peak suppression pool temperature is the design basis accident LOCA. Increasing the SRV code safety function lift setpoint tolerance from 11% to 13% and reducing the allowable number of SRVs OOS from three (3) to two (2) will have no effect on this event since reactor vessel depressurization occurs without any SRV actuation.

Small steam or liquid line breaks can result in high drywell temperature conditions which can last for relatively long periods of time since the reactor vessel remains pressurized for a longer period than for the design basis LOCA. For small line breaks with SRV actuation, the peak drywell temperature occurs relatively late in the event following many SRV actuation cycles. The containment temperature is primarily dependent on the total reactor vessel water inventory loss. The reactor vessel water inventory loss is dependent on the reactor decay heat which is unaffected by the SRV setpoint tolerance relaxation.

Therefore, the increase in the SRV code safety function lift setpoint tolerance from 1% to 13% will not have an impact on the peak drywell temperature for small steam or liquid line breaks. Similarly, the number of SRVs OOS does not affect the overall inventory loss, and therefore, will not have an impact on the peak drywell temperature for small steam or liquid breaks.

Hydrodynamic Loads on SRV Discharge Lines and Containment Pressure and thrust loads are exerted on the SRV discharge piping and quencher during SRV actuation.

Additionally, the expulsion of water and air into the suppression pool through the quencher results in pressure loads on the submerged portion of the containment wall and drag loads on submerged structures. These loads have the potential to be affected by the increase in SRV flow rate and reactor vessel pressure due to the potentially higher opening setpoints.

The standard method used to calculate steam flow though a SRV at various reactor pressure uses nominal SRV setpoint with a correction factor applied. Contained within this correction factor there is 5%

conservatism added for manufacturing uncertainties which would include setpoint tolerance. Since the current nominal SRV setpoint values are not being changed and the setpoint tolerance is less than 5%,

this proposed change will have no impact on the SRV hydrodynamic loads on the suppression pool wall or submerged structures.

Dock;t N:s. 50-352/50-353 Attachm:nt 1 January 12,1999 Page 8 of 16 The SRV quencher loads are based on actual test data and generic quencher loads based on a reactor vessel pressure of 1276 psig. Applying the maximum setpoint tolerance of +3% to the highest nominal SRV setpoint of 1190 psig yields a maximum reactor vessel pressure of 1226 psig which is bounded by the design reactor vessel pressure for quencher loads of 1276 psig.

Conclusions of NRC Specified Evaluations As previously discussed in items 1 through 6 above, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% and reducing the allowable number of SRVs OOS from three (3) to two (2)is considered acceptable based on the plant specific analyses, es required by the NRC in its SER accepting NEDC-31753P, dated March 8,1993. In addition to the requirements specified in the NRC's SER, the following items were also considered.

ECCS/LOCA Performance Evaluation The LGS, Units 1 and 2, LOCA analysis was reviewed to determine the effect of the increase in the allowable as-found SRV code safety function setpoint tolerance from 11% to 13% and the reduction in the allowable number of SRVs OOS from three (3) to two (2) on the ECCS performance. The ECCS is designed to provide adequate core cooling during a postulated LOCA by limiting the fuel Peak Cladding Temperature (PCT) to below the requirements of 10CFR50.46 (i.e., to less than 2200 F). A change in the SRV opening pressure can only affect the pipe break events for which SRV actuation occurs. The limiting break LOCA, the small break LOCA, and the main steamline break outside containment events were evaluated to determine the effects of the increased SRV setpoint tolerance and the allowable number of SRV OOS.

The limiting break LOCA for LGS, Unit 1 and 2, is the design basis accident recirculation line break. No SRV actuations occur during this event because the reactor vessel depressurizes rapidly through the break. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not affect the limiting break LOCA event ECCS performance analysis.

For the postulated small break LOCA, an increase in the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% will result in either a slightly sooner or later SRV actuation and either a decrease in initial SRV flow rate or increase in initial SRV flow rate, depending on whether the actual SRV opening pressure is lower or higher than nominal. The net effect of these changes however, is that at the time of ADS initiation, the change in total reactor vessel water inventory loss due to the change in the allowable as-found SRV code safety function lift setpoint tolerance is negligible. This is because the reactor vessel water inventory loss is primarily dependent on the reactor decay heat which is unaffected by the setpoint tolerance change. Similarly, the number of SRVs OOS does not affect the overall reactor vessel water inventory loss or the ECCS performance. Therefore, increasing the allowable as-found SRV code safety fun.ction lift setpoint tolerance from 1% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not affect the small break LOCA event ECCS performance analysis. The small break LOCA will remain non-limiting.

During the main steamline break outside containment event, the reactor vessel is completely isolated due to the closure of the main steam isolation valves (MSIVs). This terminates the break flow and reactor vessel pressure increases rapidly due to decay heat. Reactor system pressure is relieved by multiple SRV actuations, which slowly deplete reactor vessel water inventory. If the HPCI system is not available (the worst case single failure for this event), core cooling is achieved by ADS actuation to reduce reactor

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Dock:t Nos. 50-352/50-353 Attachmtnt 1 l - January 12,1999 Page 9 of 16

' vessel pressure and injection of low pressure ECCS. Similar to the small line break, the overall reactor

' vessel water inventory loss is dependent on the reactor decay heat which is not affected by the setpoint tolerance change or the number of SRVs OOS. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 3% and a reduction in the allowable number of SRVs OOS from tnroe (3) to two (2) will not affect the main steamline break outside containment event ECCS performance analysis. The main steamline break outside containment will remain non-limiting.

i L Station Blackout i

During the Station Blackout (SBO) event the reactor vessel is isolated due to the closure of the MSIVs.

Reactor vessel pressure increases rapidly due to decay heat. Reactor pressure is controlled by multiple SRV actuations. The reactor vessel is eventually depressurized using manual actuation of the SRVs or ADS and shutdown cooling is placed in service. Reactor vessel water inventory loss due to SRV actuation is made up by injection from either the RCIC or HPCI system. The RCIC and HPCI systems have been

evaluated as being capable of supplying adequate injection flow at the lowest lift setting SRV pressure L plus 3% tolerance. The heatup of the suppression pool is dependent on the reactor decay heat which is

!- not affected by this change. Therefore, increasing the allowable as-found SRV code safety function lift j- setpoint tolerance from 11% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not significantly affect the SBO event analysis.

Fire Safe Shutdown During the Fire Safe Shutdown (FSSD) event the reactor vessel is isolated due to the closure of the j MSIVs. Reactor vessel pressure increases rapidly due to decay heat. Reactor pressure is controlled by multiple SRV actuations. The reactor vessel is eventually depressurized using manual actuation of the SRVs or ADS and shutdown cooling is placed in service. Reactor vessel water inventory loss due to SRV actuation is made up by injection from either the RCIC system, HPCI system or a low pressure ECCS in conjunction with ADS depending on the shutdown method. The RCIC and HPCI systems have been evaluated as being capable of supplying adequate injection flow at the lowest lift setting SRV pressure plus 3% tolerance. The heatup of the suppression pool is dependent on the reactor decay heat which is not affected by this change. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% and a reduction in the allowable number of SRVs OOS from three (3)

- to two (2) will not significantly affect the FSSD event analysis.

High Energy Line Break The limiting High Energy Line Break (HELB)is the main steamline break outside containment. The impact on the main steamline break outside containment event was discussed earlier. For other HELBs, the maximum normal operating pressure is used in the analysis rather than SRV lift pressure due to the characteristics of the event. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not affect the HELB event analyses.

- High Pressure - Low Pressure Interface l The evaluation of high pressure -low pressure piping interfaces used the nominal setpoint of the lowest l SRV. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from i1% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not affect

. the High Pressure - Low Pressure interface evaluation.

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Dock *,t Nos. 50-352/50-353 Attachment 1 January 12,1999 Page 10 of 16 Control Rod Drive System The Control Rod Drive (CRD) system performs two (2) functions. The first function is the movement of the control rod blades to control reactor power during normal plant operations (e.g., startup, power, and shutdown). The second function is the rapid insertion of all control rods (Scram) due to abnormal plant conditions, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 1% to 3% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) does not change the normal reactor vessel operating pressure or the high reactor vessel pressure SCRAM setpoint; therefore, the proposed TS changes will not impact the CRD system capability of controlling reactor power during normal plant operations. Increasing the allowable as-found SRV lift setpoint tolerance will not change the reactor vessel pressure at which a Scram is initiated but does have the potential for increasing the reactor pressure the CRD system is subjected to at the end of the Scram process. The CRD system is designed such that reactor vessel pressure aids in the insertion of the control rod blades during the Scram process.

Therefore, the proposed change has the potential of improving the CRD system Scram performance.

Components of the CRD system are designed to the same or greater pressure / temperature requirements as the reactor vessel which have been evaluated as being acceptable for this proposed change.

Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 1% to 3% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not affect the CRD system.

Reactor Vessel Instrumentation System The components of the Reactor Vessel Instrumentation System are designed to the same or greater pressure / temperature requirements as the reactor pressure vessel which have been evaluated as being acceptable for this proposed change. Normal plant operating parameters, which are unchanged by this proposed change, are used as inputs for instrument calibration. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 1% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to (2) will not affect the Reactor Vessel Instrumentation System.

Embrgency Procedure Guidelines The Emergency Procedure Guidelines (i.e., TRIP /SAMP Procedures) use nominal, realistic, and best estimate plant parameters for determining action levels. The nominal setpoints of the SRVs are not ,

impacted by the proposed TS changes. Therefore, increasing the allowable as-found SRV code safety l function lift setpoint tolerance from 1% to 3% and a reduction in the allowable number of SRVs OOS '

from three (3) to two (2) will not affect the Emergency Procedure Guidelines.

Anticipated Transient without Scram (ATWS)

An ATWS is a beyond design basis event. The supporting Technical Evaluation Report (TER) issued with the NRC's SER for NEDC-31753P, discussed evaluations associated with beyond design basis events.

Specifically, regarding a discussbn of severe accident analyses, the TER concluded that "because the severe accident analyses gc beyond the design basis, it is not necessary to revisit them because of this change." Therefore, becauss ATWS is a beyond design basis event, the same logic applies, and a reanalysis of ATWS is nut necessary. The ATWS event is part of the licensing basis for LGS, Units 1 and

2. LGS, Units 1 and 2, incorporate many features and systems designed to mitigate the consequences of an ATWS. These features and systems meet the requirements of 10CFR50.62," Requirements for Reduction of Risk from Anticipated Transients Without SCRAM (ATWS) Events for Light-Water Cooled Nuclear Power Plants," and plant specific requirements.

l Dock:t Nos. 50-352/50-353 Attachm:nt 1 Janua y 12,1999 Page 11 of 16 l

l Although not specifically identified as a requirement in the NRC SER approving NEDC-31753P, an ATWS analysis (Enclosure 1) has been performed because of precedent with previous industry submittals. Due to the low probability of an ATWS event, nominal plant parameters including SRV setpoints are allowed to be used as specified in NUREG-0460," Anticipated Transients Without SCRAM for Light Water Reactors,"

NEDE-24222 " Assessment of BWR Mitigation of ATWS," and NEDC-31897P-1," Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," with the exception of the overpressure evaluation which uses SRV setpoint tolerance. Therefore, only the overpressure portion of the ABNS event was reevaluated. System (SLC, HPCI, and RCIC) response to an ATWS event was not reevaluated. The Pressure Regulator Failure-Open (PREGO) event under ATWS cor,ditions was reevaluated to support the proposed condition of 12 of 14 SRVs OPERABLE and the increase in the allowable as-found SRV code safety function lift setpoint tolerance from 1% to 3%. In this analysis (Enclosure 1), the ATWS Recirculation Pump Trip (RPT) causes the initial reduction in reactor power, and the Standby Liquid Control System (SLCS) eventually shuts down the plant. The analysis (Enclosure 1) assumes 100%

power and 81% core flow operating conditions which is limiting for this event. LGS, Unit 2, Cycle 5 data was used for dynamic void and Doppler reactivity, and is representative of both LGS Units.

The analysis (Enclosure 1) was performed using the one-dimensional ODYN computer code which has been approved by the NRC for this application. The results of the analysis, using the ODYN code, shows that the vessel pressure reaches a maximum of 1468 psig, which is within the reactor vessel overpressure criterion of 1500 psig for the ATWS event. Tho long-term effect on suppression pool temperature due to the increase in the allowable SRV setpoint tolerance and the reduction in the number of SRV allowed to be OOS is negligible, since there is little change in the total energy discharged to the suppression pool through the SRVs. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 3% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will not have a significant effect on the ATWS event analyses.

SRV Simmer Margin GE SIL 196, Supplement 3," Target Rock Safety / Relief Valve Simmer Margin," recommends that a simmer margin of 120 psi be maintained between the maximum normal operating pressure and the lowest SRV setpoint. Currently LGS, Units 1 and 2, have a simmer margin of 125 psi based on the nominal difference of the maximum normal operating pressure of 1045 psig and the lowest SRV setpoint of 1170 psig. If the current 1% setpoint tolerance is considered this simmer margin is 113.3 psi and if the proposed setpoint tolerance of 3% is considered, the minimum simmer margin is 89.9 psi. A reduction in simmer margin will not directly result in an increase of the probability of an inadvertent self actuation of a SRV. A reduction in simmer margin will reduce the seating force which may initiate leakage. However, this leakage is monitored and corrective actions can be implemented prior to progressing to the point of the potential of an inadvertent actuation. A potential reduction in the simmer margin to 89.9 psi has been evaluated by the SRV manufacturer and determined to be acceptable. This is applicable to both two (2) and three (3) stage SRVs. Therefore, increasing the allowable as-found SRV code safety function lift setpoint tolerance from 1% to 13% and a reduction in the allowable number of SRVs OOS from three (3) to two (2) will maintain adequate SRV simmer margin.

Information Supportina a Findina of No Slanificant Hazards Consideration We have concluded that the proposed changes to the Limerick Generating Station (LGS), Units 1 and 2, Technical Specifications (TS) to revise TS Section 3/4.4.2, and associated TS Bases Sections B 3/4.4.2 and B 3/4.5.1, to increase the main steam Safety Relief Valve (SRV) as-found safety function lift setpoint tolerance from 11% to 13%, does not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.

Dock t Nos. 50-352/50-353 Attachment 1 January 12,1999 Page 12 of 16

1. The crocosed Technical Soecifications (TS) chanaes do not involve a sianificant increase in the orobability or consecuences of an accident oreviousiv evaluated.

The proposed TS changes allow for an increase in the as-found main steam Safety Relief Valve (SRV) setpoint tolerance from 11% to 13%. The proposed changes also reduce the allowable number of SRVs to be out-of-service from three (3) to two (2). The proposed changes do not alter the SRV nominallift setpoints or SRV lift setpoint test frequency. The actuation of an SRV is the precursor to the inadvertent opening of a SRV transient, as discussed in Updated Final Safety Analysis Report (UFSAR) Chapter 15.1.4. Increasing the allowable as-found SRV code safety function lift setpoint tolerance from 11% to 3% does have the potential for the minimum SRV simmer margin to be reduced from 113.3 psig to 89.9 psig. A reduction in simmer margin will not directly result in an increase of the probability on an inadvertent self actuation of an SRV. A reduction in simmer margin will reduce the seating force which may initiate leakage. However, this leakage is monitored and corrective actions can be implemented prior to progressing to the point of the potential of an inadvertent actuation. This reduction in SRV simmer margin has been evaluated by the SRV manufacturer and determined to be acceptable; therefore, the probability of an inadvertent SRV actuation remains unchanged. Actuation of an SRV is not a precursor for any other event evaluated in the Safety Analysis Report (SAR).

The proposed TS changes have been evaluated on both a generic and plant specific basis. The NRC has approved the general approach of this change; however, implementation is contingent on several plant specific evaluations. The required plant spoeific analyses and evaluations included transient analysis of the anticipated operatinnal traasients (AOTs); analysis of the design basis overpressurization event; evaluation of the performance of high pressure systems, motor operated valves, and vesselinstrumentation and associated piping; and evaluation of the containment response during Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the SRV discharge lines and containment. In addition to the plant specific analyses and evaluations required by the NRC, the following items were also considered: ECCS/LOCA performance, SRV simmer margin, high pressure -low pressure interfaces, i.e., High Energy Line Break (HELB),

Station Blackout (SBO), and Fire Safe Shutdown (FSSD), and the short term pressurization phase of an ATWS event. These analyses and evaluations show that there is adequate margin to the design core thermallimits and reactor vessel pressure limits using the 13% SRV code safety function lift setpoint tolerance and two (2) SFtVs out-of service. The analyses and evaluations also show that the operation of the high pressura injection systems will not be adversely affected, that SRV discharge piping stresses will not be exceeded, and that the containment response during a LOCA will be acceptable.

Evaluations of the impact of the proposed change on the Equipment important to Safety have been performed and no adverse conditions were identified. The reactor pressure vessel and attached systems and piping havo been evaluated for the impact of this proposed TS change. A plant specific analysis has been performed which indicates that neith-er the American Society of Mechanical Engineers (ASME) Code upset limits or the TS Safety Limits for the reactor pressure vessel will be exceeded for the limiting event, i.e., Main Steam Isolation Valve (MSIV) closure with flux Scram. The reactor pressure vessel and attached piping design values will not be exceeded. The current high pressure - low pressure interface evaluation utilized nominal SRV setpoints, and therefore,is unaffected. Therefore, the probability of a malfunction of the reactor pressure vessel and attached systems and piping is not increased.

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Dockit Nos,50-352/50-353 Attachm:nt 1 January 12,1999 Page 13 of 16 The nuclear fuel has been evaluated for the impact of the proposed change. Plant specific analyses were performed which indicate that for all abnormal operational transients adequate margin to the limiting thermal limit parameter, i.e., Minimum Critical Power Ratio (MCPR), is maintained. Emergency Core Cooling System (ECCS)/LOCA performance is maintained adequate to meet the requirements of 10CFR50.46. Therefore, the probability of the malfunction of the nuclear fuel is not increased.

The SRVs have been evaluated for the impact of the proposed TS changes. No physical changes to the SRVs will be made as a result of the proposed TS changes. Adequate simmer margin will be maintained with the increased tolerance to ensure that an inadvertent lifting of a SRV does not i occur. The increase in SRV discharge flow and reactor vessel pressure due to the potential for i higher SRV lift setpoints are bounded by the SRV steam flows and reactor vessel pressure currently used in the evaluation of SRV discharge piping, quencher, quencher support, and hydrodynamic loads on the suppression pool and submerged structures; therefore, the probability of a malfunction of a SRV or associated components and structures is not increased.

The Containment response during a LOCA has been evaluated for the impact of the proposed change. The major factor in the Containment response to a LOCA is the rate of reactor vessel water inventory loss. The rate of reactor vessel water inventory loss is mainly dependent on reactor decay heat which is not affected by the proposed change. Therefore, the probability of the malfunction of the Containment is not increased.

The High Pressure Coolant Injection (HPCI) system has been evaluated for the impact of the proposed TS changes. The analysis determined that the HPCI system would not be capable of developing its design flowrate of 5600 gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint +3% tolerance) unless the HPCI turbine / pump maximum rated speed was increased. However, increasing the HPCI turbine / pump maximum rated speed is prevented due to HPCI pump discharge piping overpressurization concems. Further analysis has shown that the HPCI system is capable of meeting its required ECCS function design flowrate, and its required non-ECCS flowrate, without any change to the current system operating parameters. Therefore, the probability of a malfunction of the HPCI System is not increased.

The Reactor Core Isolation Cooling (RCIC) system has been evaluated for the impact of the proposed change. The analysis determined that in order for the RCIC system to be capable of injecting its design flowrate of 600 gpm at a reactor pressure of 1205 psig (lowest SRV setpoint of 1170 psig +3% tolerance) the maximum rated speed of the RCIC turbine / pump is required to be increased from 4575 rpm to 4625 rpm. This increase in the RCIC turbine / pump maximum rated speed will reduce the margin to the overspeed trip from 123% to 122.1%. This reduction in the margin to the overspeed trip is acceptable due to the implementation of plant Modification P00210, *RCIC System Startup Transient Improvement," which reduced the amount of turbine / pump speed overshoot during system startup. The RCIC overspeed trip setpoint will not be changed; therefore, a failure of the RCIC turbine / pump (missile hazard or system overpressurization) due to overspeed is not increased. All other RCIC System components will continue to operate within the currently specified design and operating limits. Therefore, the probability of a malfunction of the RCIC System is not increased.

Dock:t Nos. 50-352/50-353 Attachm:nt 1 January 12,1999 Page 14 of 16 1

The Standby Liquid Control (SLC) system has been evaluated for the impact of the proposed change. The SLC system capability of shutting down the reactor during a postulated event in which all or some of the control rods cannot be inserted or during a postulated Anticipated l

Transirat Without Scram (ATWS) event is not impacted by this proposed change. Therefore, the l probr.oility of a malfunction of the SLCS is not increased. l l

The Control Rod Drive (CRD) system has been evaluated for the impact of the proposed change. '

The CRD system capability of controlling reactor power durin0 normal plant operation and rapidly I inserting control rod blades (Scram) during abnormal plant conditions is not impacted by the proposed change. Therefore, the probability of a malfunction of the CRD system is not increased.

The Reactor Vessel Instrumentation System has been evaluated for the impact of the proposed change. The Reactor Vessel Instrumentation System will continue to be operated within the current design pressure / temperature requirements; therefore, the probability of a malfunction of the Reactor Vessel Instrumentation System is not increased.

The LGS, Units 1 and 2, Generic Letter 89-10 Motor-Operated Valve (MOV) Program has been evaluated for the proposed change. The LGS MOV Program currently uses SRV nominal setpoints for differential pressure determinations for valves in which reactor pressure at the SRV l

setpoint is limiting. Use of nominal SRV setpoints is consistent with current industry practice. .

Therefore, the probability of a malfunction of a MOV is not increased. I 1

Reducing the number of SRVs allowed to be out-of-service does not make the consequences of a malfunction of a SRV more severe, since the number of SRVs required to maintain the reactor vessel within ASME Code and TS Safety Limits will be maintained OPERABLE. The proposeo change does not result in any changes to the interactions of any system, structure, or component. ,

All systems, structures, and components will continue to function as designed. l Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated.

2. The crocosed TS chanaes do not create the oossibility of a new or different kind of accident from any accident oreviously evaluated.

The proposed TS changes allow for an increase in the as-found SRV setpoint tolerance from 11%

to 13% The proposed TS changes also reduce the allowable number of SRVs to be out-of-service from three (3) to two (2). Generic and plant specific analyses and evaluations indicate that the plant response to any previously evaluated event will remain unchanged. All plant systems, structures, and components will continue to be capable of performing their required safety function as required by event analysis guidance.

The proposed TS changes do not alter the SRV nominal lift setpoints or SRV lift setpoint test frequency. The operation and response of the affected Equipment important to Safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters. No system, structure, or component will be subjected to a condition that has not been evaluated and determined to be acceptable using the guidance required for specific event analysis.

l Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Dockit Ncs. 50-352/50-353 Attachm:nt 1 January 12,1999 Page 15 of 16 l

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3. The orocosed TS chanaes do not involve a slanificant reduction in a marain of safety. '

The proposed TS changes allow for an increase in the as-found SRV setpoint tolerance from 11%

to 13% The proposed TS changes also reduce the allowable number of SRVs to be out-of-service from three (3) to two (2). The proposed TS changes do not alter the SRV nominal lift setpoints or SRV lift setpoint test frequency. The operation and response of the affected Equipment important to Safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters. While the calculated peak reactor vessel pressure for the ASME overpressure event and the ATWS Pressure Regulator Failure-Open (PREGO) event are higher than those calculated without the increase in setpoint tolerance, both are still within the respective licensing acceptance limits associated with these events. These licensing acceptance limits have been determined by the NRC to provide a sufficient margin of safety.

The increase in the RCIC system turbine / pump maximum rated speed is within the capability of the system design. The reduction in the margin to the overspeed trip is not a reduction in the margin of safety, since the operation of the RCIC System has demonstrated minimal speed overshoot on system initiation due to the installation of plant Modification P00210,"RCIC System Startup Transient Improvement."

The inability of the HPCI system to be capable of injecting 5600 gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint of 1170 psig +3% tolerance) is not a reduction in the margin of 4 safety, since analysis for events that would result in high reactor vessel pressure indicate that the HPCI System is capable of providing adequate coolant injection.

The increme in SRV steam flow and reactor vessel pressure does not reduce the margin of safety associated with the SRVs and associated components and structures since the increased SRV steam flow rate and reactor vessel pressure are bounded by the current design analysis.

The margin of safety for fuel thermallimits and 10CFR50.46 limits is unaffected by the prbposed change.

The margin of safety for the Containment is unaffected by the proposed change.

The capability of the SLC system to perform its safety function during all required events, using the required guidance for event analysis, is maintained. Therefore, the proposed changes do not reduce the margin of safety provided by the SLC system.

Therefore, these proposed TS changes do not involve a significant reduction in a margin of safety.

Information S'upportina an Environmental Assessment An environmental assessment is not required for the changes proposed by this TS Change Request because the requested changes to the Limerick Generating Station (LGS), Units 1 and 2, Technical Specifications (TS) conform to the criteria for " actions eligible for categorical exclusion," as specified in 10CFR5122(c)(9). The proposed changes will have no impact on the environment. The proposed changes do not involve a significant hazards consideration as discussed in the preceding section. The proposed changes do not involve a significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

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Docket Nos. 50-352/50-353 Attachment 1 January 12,1999 Page 16 of 16 l

l Conclusion The Plant Operations Review Committee and the Nuclear Review Board have reviewed the proposed changes to the Limerick Generating Station (LGS), Units 1 and 2, TS and have concluded that they do not involve an unreviewed safety question, and will not endanger the health and safety of the public.

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ATTACHMENT 2 Limerick Generating Station Units 1 and 2 Docket Nos. 50-352 50-353 l Licerise Nos. NPF-39 NPF-85 1

Technical Specifications Change Request No'. 98-08-0 List of Affected Pages

. Unit 1 Unit 2 3/44-7 3/4 4-7 B 3/4 4-2 B 3/4 4-2 B 3/4 5-1 B 3/4 5-1 i

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