ML20059K572

Application for Amend to Licenses NPF-39 & NPF-85,consisting of Change Request 93-18-0,revising TS to Permit Increase in Allowable Leak Rate for MSIV & to Delete MSIV Leakage Control Sys

ML20059K572
Person / Time
Site:	Limerick
Issue date:	01/14/1994
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20059K578	List: ML20059K572 ML20059K583
References
NUDOCS 9402020164
Download: ML20059K572 (14)

Text

L:!esw Group Hacdqutrtsrs YECO ENERGY  ;=;r;sm Wayne, PA 19087 5691 10 CFR 50.90 January 14,1994-Docket Nos. 50-352 50453 Ucense Nos. NPF-39 NPF-85 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Umerick Generating Station, Units 1 and 2 Technical Specifications Change Request Gentlemen:

PECO Energy Company (PECo) is submitting Technical Specifications (TS) Change Request No. 93-18-0, in accordance with 10 CFR 50.90, requesting an amendment to the TS (i.e., Appendix A) of Operating Ucense No.

NPF-39 for Limerick Generating Station (LGS), Unit 1 and Operating License No. NPF-85 for LGS, Unit 2.

This submittal requests an amendment to the TS, to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) and to delete the MSIV Leakage Control System (LCS). PECo proposes to ut!!!ze the main steam drain lines and the main condenser as an alternate MSIV leakage treatment method. While certain main steam piping and components, including the drain lines and main condenser, are not currently classified as seismic Category 1, a detailed evaluation Indicates that the main steam piping and equipment are seismically rugged and meet the Intent of Appendix A to 10 CFR 100 for seismic adequacy.

Please note that due to schedule restraints, this submittal only contains a detailed evaluation of the LGS, Unit 2 main steam drain piping and the main condenser. To ensure that the corresponding LGS, Unit 1 piping and main condenser meet the same criteria, the same walkdown and evaluation will be performed on the affected LGS, Unit 1 systems during the upcoming fiah Unit I refueling outage and, if this change request is approved, all necessary design changes will be made to both Units prior to implementation of the proposed changes.

Information supporting this Change Request is contained in Attachment 1 to this letter. The LGS, Unit 1 and Unit 2 TS pages showing the proposed changes are contained in Attachment 2. A report entitled, "MSIV Leakage Alternate Drain Pathway Evaluation," discussing the seismic adequacy of the main steam piping and main condenser, including the results of a walkdown of the Unit 2 piping and equipment, is provided as Attachment 3.

9402020164 940114 PDR ADOCK 05000352 j I

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U.S. Nuclear Reguttory Commission January 13,1994 Document Control Desk Pcge 2 To a!!ow time for completion of necessary engineering work, we request that, if approved, the amendments be issued by June 1,1994 and be effective by February 15,1995 for LGS, Unit 2, and February 15,1996 for LGS, Unit 1. This will allow sufficient time for the preparation and implementation of the necessary modifications and procedural changes during the next Unit 2 refueling outage, currently scheduled for February,1995.

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

Very truly yours, e 1-G. A. Hunger, Jr.

Director Ucensing Section Attachments cc: T. T. Martin, Administrator, Region I, USNRC w/ attachments N. S. Perry, USNRC Senior Resident inspector, LGS w/ attachments W. P. Dornsife, Director, PA Bureau of Radiological Protection w/ attachments A

COMMONWEALTH OF PENNSYLVANIA :

ss.

COUNTY OF CHESTER  :

D. R. Holwig, being first duly sworn, deposes and says: .

That he is Vice President of Philadelphia Electric Company, the Applicant herein; that he has read the enclosed Technical Specifications Change Request No. 93-18-0 for Limerick Generating Station, Unit 1 and Unit 2, Facility Operating License Nos. NPF-39 and NPF-85, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

P Vice President Subscribed and sworn to before me this c 7 [ day

>, pr 1993.

'po p cf2n!otary4xPublicew e x e rs -

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ATTACHMENT 1 UMERICK GENERATING STATION UNITS 1 and 2 Docket Nos. 50-352 50452 License Nos. NPF-39 NPF-85 I

TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 93-18-0

" Increase Allowable Leakage from the Main Steam isolation Valves and the Removal of the Leakage Control System" r

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Supporting Information for Changes - 10 pages

Attachment 1 Page 1 PECO Energy Company (PECo), formerly Philadelphla Electric Company, licensee under Facility Operating Ucense Nos. NPF-39 and NPF45 for Umerick Generating Station (LGS), Units 1 and 2, requests that the Technical Specifications contained in Appendix A to the Operating Ucenses be amended as proposed herein, to perm!t an increase in the allowable leak rate for the Main Steam Isolation Valves (MSIVs) and to delete the MSIV Leakage Control System (LCS). These proposed changes are based on the General Electric (GE) report prepared for the Boiling Water Reactor Owners' (BWROG), *BWROG Report for increasing MSIV Leakage Rate Umits and Elimination of Leakage Control System," NEDC-31858P, Revision 2, submitted to the NRC by BWROG letter dated October 4, 1993.

The proposed changes to the TS pages are indicated by a vertical bar in the margin of the affected TS pages. The TS pages showing the proposed changes are contained in Attachment 2.

We request that, if approved, the amendments to the LGS, Unit 1 and Unit 2 TS be issued by June 1,1994, and be effective by February 15,1995 for Unit 2 and by February 15,1996 for Unit 1 in order to coincide with the upcoming refueling outages.

This submittal provides a discussion and description of the proposed TS changes, a Safety Assessment of the proposed TS changes, Information Supporting a Finding of No Significant Hazards Consideration, and information Supporting an Environmental Assessment.

Discussion and Descriotion of the Prooosed Chanaes The proposed changes to the Technical Specifications (TS) for Umorick Generating Station (LGS), Unit 1 and Unit 2 are as follows.

a. In the TS Table of Contents page xil, replace *MSIV Leakage Control.... 3/4 G-7" with *MSIV LEAKAGE ALTERNATE DRAIN PATHWAY.. 3/4 6-7.*

b. In TS Section 3.6.1.2, paragraph c, the following changes are proposed.

1. Umiting Condition for Operation (LCO) Statement 'c." sha!! state:

"Less than or equal to 100 scf por hour for any one main steam line through the isolation valves not to exceed 200 scf per hour for all four main steam lines, when tested at P ,22.0 psig."

2. Action Statement "with: c," shall state:

'The measured leakage rate exceeding 100 scf per hour for any one main steam line through the isolation valves, or exceeding 200 scf per hour for all four main steam lines, or"

3. Action Statement " restore: c." shall state:

"The leakage rate to less than or equal to 11.5 scf per hour for any one main steam line through the isolation valves, and the total Main Steam line leakage to less than 200 scf per hour for all Main Steam lines, and" I

t i

Attachment 1 Page 2 i

c. TS Table 3.6.3-1 is proposed to be changed as follows. .

I

1. On page 3/4 6-19 for penetration number 007A(8,C,D) eliminate the following: "HV40- j 1(2)F001B(F,K,P)(XV40-1(2)01B(F,K,P) SEE PART B, THIS TABLE)* and the associated ,

parameters and notes.  ;

2. On page 3/4 6-31 for penetration number 007A(B,C,D), eliminate the following: 'SEE PART A  ;

THIS TABLE (HV40-1(2)F001B(F,K,P) SEE PART A THIS TABLE) XV40-1(2)01B(F,K,P) and the associated parameters and notes.

d. For TS Section 3/4.6.1.4, eliminate the entire text. Replace with the following.

CONTAINMENT SYSTEM MSIV LEAKAGE ALTERNATE DRAIN PATHWAY  ;

LIMITING CONDITION FOR OPERATION 3.6.1.4 The MSIV Leakage Alternate Drain Pathway shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1,2, AND 3.

ACTION:

With the MSIV Leakage Alternate Drain Pathway inoperable, restore the pathway to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

SURVEILLANCE REOUIREMENTS i

The MSIV Leakage A!!emate Drain Pathway shall be demonstrated OPERABLE:

a. In accordance with 4.0.5, by cycling each motor operated valve, required to be repositioned, r through at least one complete cycle of full travel. ,

e. In the BASES for TS Section 3/4 6.1.4, page B 3/4 6-1, delete the reference to 3/4.6.1.4, MSIV LEAKAGE CONTROL SYSTEM. Replace with the following.

1 3/4.6.1.4 MSIV LEAKAGE ALTERNATE DRAIN PATHWAY Calculated doses resulting from the maximum leakage allowances for the main steam line isolation .

valves in the postulated LOCA situations will not exceed the criteria of 10 CFR Part 100 guidelines, .

provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally  :

occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not .  ;

always been continuously maintained. The requirement for the MSIV Leakage Alternate Drain '

Pathway serves to reduce the off-site dose.

i 1

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Attachment 1 Page 3 Safety Assessment The Bolling Water Reactor Owners' Group (BWROG) has evaluated the availability of main steam system piping and main condenser alternate pathways for processing Main Steam isolation Valve (MSIV) leakage, and has determined that the probabl!ity of a near coincident Loss of Coolant Accident (LOCA) and a seismic event is much smaller than for other plant safety risks. Accordingly, this alternate MSIV leakage treatment pathway will be available during and after a LOCA. Nevertheless, the BWROG has also determined that main steam piping and main condenser designs are extremely rugged, and that the design requirements applied to the Limerick Generating Station (LGS), Unit 1 and Unit 2 main steam system piping and main condenser contain substantial margin, based on the original design requirements.

In order to further justify the capability of the main steam piping and main condenser altemate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers during past earthquakes. As summarized in General Electric (GE) Report, "BWROG Report for increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, *NEDC 31858P, Revision 2, submitted to the NRC by BWROG letter dated October 4,1993, this study concluded that the possibility of a failure which could cause a loss of steam or condensate in Bolling Water Reactor (BWR) main steam piping or condensers in the event of a design basis (i.e., safe shutdown) earthquake is highly unlikely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

We have evaluated the seismic adequacy of the Unit 2 main steam piping and main condenser consistent with the guidelines discussed in Section 6.7 of NEDC-31858P, Revision 2, to provide reasonable assurance of the structural Integrity of these components. The corresponding Unit 1 piping and main condenser will also be evaluated prior to e implementation of the proposed changes. This evaluation, including the walkdown report, for LGS, Unit 2 is provided in Attachment 3. The results of the evaluation clearly demonstrate that the MSIV Leakage Alternate Drain Pathway moots the intent of 10CFR100 Appendix A, with regard to seismic qualification. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any additional logic control and interlocks. The method proposed for MSIV leakage treatment is consistent with the philosophy of protection by multiply barriers used in containment design for limiting fission product release to the environment.

A plant-specific radiological analysis has been performed in accordance with NEDC-31858P, Revision 2, to assess the effects of the proposed increase to the allowable MSIV leakage rate in terms of Main Control Room (MCR) and off-site doses following a postulated design basis LOCA. This analysis utilizes the hold up volumes of the main steam piping and condenser as an a!!crnate method for treating the MSIV leakage. As discussed earlier, there is reasonable assurance that the main steam piping and condenser will remain intact following a design basis earthquake. The radiological analysis uses standard conservative assumptions for the radiological source term consistent with Regulatory Guide (RG) 1.3, ' Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-Of-Coolant Accident for Bolling Water Reactor," Revision 2, dated April 1974.

The analysis results demonstrate that dose contributions from the proposed MSIV leakage rate limit of 100 scfh per steam line, not to exceed a total of 200 scfh for all four main steam lines, along with the proposed deletion of the Leakage Control System (LCS), result in an acceptable increase to the LOCA doses previously evaluated against the regulatory limits for the off-site doses and MCR doses contained in 10CFR100 and 10CFR50, Appendix A, General Design Criterion (GDC) 19, respectively. The off-site and MCR doses resulting from a LOCA are discussed in Section 15.6.5 of the UFSAR. The off-site and MCR doses resulting from a LOCA associated with the proposed changes are the sum of the LOCA doses previously evaluated in the UFSAR and the additional doses calculated using the alternate MSIV leakage treatment method. This method of calculating the revised doses is highly conservative since the LOCA doses previously evaluated already included dose contributions from MSIV leakage at the maximum leakage rate currently permitted by the Technical Specifications (TS). Appendix 2 of Attachment 3 shows the previously calculated doses and the new calculated doses based on the proposed changes. >

Attachment 1 Page 4 In summary, the proposed changes do not result in a significant increase in the radiological consequences of a LOCA when the same assumptions and methods specified in the UFSAR are used, recognizing that radiological consequences calculated in the UFSAR and for these proposed changes are significantly higher than those using more realistic assumptions and methods. The calculated off-site and MCR doses resulting from a LOCA remain well below the regulatory limits. Although the revised LOCA doses are higher for low MSIV leakage rates, the effectiveness of the proposed a!!emate treatment method, even for leakage rates greater than the proposed increased MSIV allowable leak rate, ensures that off-site and MCR dose limits are not exceeded.

Information Sucoortina a Findina of No Sianificant Hazards Consideration We have concluded that the proposed changes to the Limerick Generating Station (LGS), Unit 1 and Unit 2 Technical Specifications (TS) to increase the allowed leakage from the Main Steam isolation Valves (MSIVs), to eliminate the MSIV Leakage Control System (LCS) requirement from the TS, and to eliminate the primary containment isolation valves associated with the MSIV-LCS, do not constitute 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.

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

The proposed changes to TS Section 3.6.1.2 do not involve a change to structures, cornponents, or systems that would affect the probability of an accident previously evaluated in the Umerick Generating Station (LGS)

Updated Final Safety Analysis Report (UFSAR).

The proposed changes involve eliminating the Main Steam Isolation Valves (MSIVs) Leakage Control System (LCS) requirements from the TS. As described in Section 6.7 of the UFSAR, the LCS is manually initiated in about 20 minutes following a design basis Loss of Coolant Accident (LOCA). Since the LCS is operated only after an accident has occurred, these proposed changes have no effect on the probability of an accident.

Since MSIV leakage and operation of the LCS are included in the radiological analysis for the design basis LOCA as described in Section 15.6.5 of the UFSAR, the proposed changes will not affect the precursors of other analyzed accidents. Analysis of the affects of the proposed changes do, however, result in acceptable radiological consequences for the design basis LOCA previously evaluated in Section 15.6.5 of the UFSAR.

LGS, Units 1 and 2 have an inherent MSIV leakage treatment capability as discussed below. We propose to I

use the drain lines associated with the main steam lines and main turbine condenser as an altomative to the guidance in Regulatory Guide 1.96, " Design of Main Steam Isolation Valve Leakage Control System For Boiling Water Nuclear Power Plants," Revision 0, May 1975, for MSIV leakage treatment. If approved, we will ,

I incorporate this alternate method in the appropriate operatbnal procedures and Emergency Operating Procedures.

The Bolling Water Reactor Cwners' Group (BWROG) han evaluated the availabi!!!y of main steam system piping and main condenser a!!emate pathways for processing MSIV leakage, and has determined that the probability of a near coincident LOCA and a seismic event is much smaller than for other plant safety risks.

Accordingly, this altemate MSIV leakage treatment pahway will be available during and after a LOCA.

Nevertheless, the BWROG has also determined that main steam piping and main condenser designs are extremely rugged, and that the design requirements applied to the LGS, Unit 1 and Unit 2 main steam i system piping and main condenser contain substantial margin, based on the original design requirements. l

Attachment 1 Page 5 in order to further justify the capability of the main steam piping and main condenser a!!ernate treatment pathway, the BWROG has reviewed limited earthquake experience data on the performance of non-seismically designed piping and condensers during past earthquakes. As summarized in General Electric (GE) Report, *BWROG Report for increasing MSIV Leakage Rate Umits and Elimination of Leakage Control Systems

• NEDC 31858P, Revision 2 submitted to the NRC by BWROG letter dated October 4,1993, this study concluded that the possib2y of a failure that could cause a loss of steam or condensate in BoHing Water Reactor (BWR) malr, steam piping or condensers in the event of a design basis (i.e., safe shutdown) earthquake is highly milkely, and that such a failure would also be contrary to a large body of historical earthquake experience data, and thus unprecedented.

We have performed a verification of seismic adequacy of the Unit 2 main steam piping and main condenser consistent with the guidelines discussed in Section 6.7 of NEDC-31858P, Revision 2, to provide reasonable assurance of the structuralintegrity of these components. The corresponding Unit 1 piping and main condenser will be reviewed prior to implementation of the proposed changes. This evaluation, including the walkdown report, MSIV Leakage Altemate Drain Pathway Evaluation," for LGS, Unit 2, is provided in Attachment 3. The results of the evaluation clearly demonstrate that the MSIV Leakag; A!!emate Drain Pathway meets the intent of 10CFR100 Appendix A, wth regards to seismic qualification. Except for the requirement to establish a proper flow path from the MSIVs to the cordenser, the proposed methov e passive and does not require any additionallogic control and interiocks. The method proposed foi MSlV loakage treatment is consistent with the philosophy of protection by multiply barriers used in contabment design for limiting fission product release to the environment.

A plant. specific radiological analysis has been performed in accordance with NEDC-31858P, Revision 2, to assess the effects of the proposed increase to the allowable MSIV leakage rate in terms of Main Control Room (MCR) and off-site doses fo!!owing a postulated design basis LOCA. This analysis utilizes the hold-up volumes of the main steam piping and condenser as an attemate method for treating the MSIV leakage. As discussed earlier, there is reasonable assurance that the main steam piping and condenser will iemain intact following a design basis earthquake. The radiological analysis uses standard conservative assumptions for the radiological source term consistent with Regulatory Guide (RG) 1.3, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-Of-Coolant Accident for Bolling Water Reactor," Revision 2, dated April 1974.

The analysis results demonstrate that dose contributions from the proposed MSIV leakage rate ilmlt of 100 scfh per steam line, not to exceed a total of 200 scfh for all four main steam lines, and from the proposed deletion of the LCS, result in an acceptable increase to the LOCA doses previously evaluated against the regulatory limits for the off-sl'e doses and MCR doses contained in 10CFR100 ard 10CFR50, Appendix A, General Design Criterion (GDC) 19, respecth>ely. The off-site and MCR doses resulting from a LOCA are discussed in Section 15.6.5 of the UFSAR. The off-site and MCR doses resulting from a LOCA associated with the proposed changes are the sum of the LOCA doses previously evaluated in the UFSAR and the additional doses calculated using the attemate MSIV leakage treatment method. This method of calculating the revised doses is highly conservative since the LOCA doses previously evaluated already included dose contributions from MSIV leakage at the maximum leakage ratr. currently permitted by the TS. Appendix 2 to the Attachment 3 report, "MS!V Leakage Altemate Drain Pattrvay Evaluation," shows the previously calculated doses and the new calculated doses based on tha proposed changes.

The whole body doses at the Low Population Zone (LPZ) and the MCR are increased from 1.7 to 1.84 rem and from 0.38 to 0.43 rem, respectively. These increases are not significant r,ince the revised doses are well below the regulatory lim!ts, i.e.,1.84 rem calculated versus the !!mit of 25 rem at the LPZ, and 0.43 rem calculated versus the limit of 5 rem in the MCR. The associated whole body dose at the Exclusion Area

Attachment 1 Page 6 Boundary (EAB) increased insignificant!y from 0.67 to 0.671 rem. The revised thyroid dose at the LPZ increased from 0.04 rem to 36.41 rem for Unit 1, and to 36.48 rem for Unit 2. This increase is not significant since the maximum revised dose of 36.48 rem is well within than the regulatory ilmit of 300 rem. The thyroid dose at the EAB increased slightly from 0.15 rem to 0.17 rem, whereas the MCR thyroid dose incnased from

• 0.004 rem to 6.27 rem for Unit 1, and to 6.29 rem for Unit 2. The increased MCR thyroid dose is not significant since the revised dose remains well below the regulatory limit of 30 rem. The MCR beta dose increased from 7.6 rem to 8.2 rem, remaining insignificant relative to the regulatory limit of 30 rem.

The resulting revised thyroid doses discussed above are dominated by the organic radiolodine fractions of the accident source term which occur because of the conservative assumptions used in this analysis. For an assumed 100 scfh leakage per steam line, more than 95% of the off-site and MCR doses from lodine are due to organic iodine in accordance with the guidance in RG 1.3, and assuming that organic lodine is converted from the elemental iodine deposited in main steam piping systems. NRC paper SECY 92-127,

" Revised Accident Source Terms for Ught-Water Nuclear Power Plants," and draft NUREG-1465 " Accident Source Terms For Light-Water Nuclear Power Plants," identify that based on current research and data, the production of radiolodine from fuel damage is significantly less than that assumed in RG 1.3. Although the NRC has not yet defined the percentage of iodine to be assumed in the new source term model, the actual percentage of iodine in the accident radiological source term is recognized as being significantly less than previously assumed and will result in a reduction of the organic iodine source term by as much as a factor of 15. Accordingly, the calculated doses discussed above are considered to be highly conservative relative to the realistic radiological source term resulting from a postulated LOCA.

In summary, the proposed changes discussed above do not result in a significant increase in the radiological consequences of a LOCA when the same assumptions and methods specified in the UFSAR are used, recognizing that radiological consequences calculated in the UFSAR and for these proposed changes are significantly higher than those using more realistic assumptions and methods. Nevertheless, the calculated off-site and MCR doses resulting from a LOCA remain well below the regulatory limits. Although the revised LOCA doses are higher for low MSIV leakage rates, the effecth/eness of the proposed attemate treatment method, even for leakage rates greater than the proposed increase in the MSIV allowable leak rate, ensures that off-site and MCR dose limits are not exceeded.

The proposed change to TS Table 3.6.3-1 involves the deletion of LCS valves from the list of primary containment isolation valves. This proposed change is consistent with the proposed deletion of the LCS.

The LCS lines that are connected to the main steam piping wl!l be welded and/or capped closed to assure primary containment integrity is maintained. The welding and post weld examination procedures will be in - .

accordance with Amencan Society of Mechanical Engineers (ASME) Code, Section 111 requirements. These welds and/or caps will be periodically tested as part of the Containment Integrated Leak Rate Test (CILRT).

This proposed change does not involve an increase in the probability of equipment malfunction previously evaluated in the UFSAR. In fact, this proposed change reduces the probability of equipment rnalfunction since, upon implementation of these proposed changes, the plant will be operated with less primary containment isolation valves subjected to postulated failure. This proposed change has no effect on the +

consequences of an accident since the LCS lines will be welded and/or cap closed, thus assuring that the containment integrity, isolation, and leak test capabl!!!y are not compromised.

Therefore, as discussed above, the proposed changes do not involve a significant increase in the probab!Ilty or consequences from any accident previously evaluated.

h Attachment 1 Page 7

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

Although the proposed changes wl!I introduce and take credit for a new level of operational performance for existing plant systems and components that have not been previously evaluated in the accident analysis, the affect on this equipment has been evaluated and found to provide an acceptable level of reliability that will provide the required level of protection. This conclusion is based on the evaluation performed in NEDC 31858P, Revision 2, and the seismic evaluation provided in the Attachment 3 report, *MSIV Leakage Altemate Drain Pathway Evaluation." Therefore, re!!ance on different equipment than previously assumed to mitigate the consequences of an accident does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The BWROG evaluated MSIV performance and concluded that MSIV leakage rates up to 200 scfh will not inhibit the capability and isolation performance of the MSIVs to effectively isolate the primary containment.

Implementation of the proposed changes will not result in modifications which could adversely impact the operability of the MSIVs. The LOCA has been analyzed using the main steam piping and main condenser as a treatment method to process MSIV leakage at the proposed maximum rate of 100 scfh per main steam line, not to exceed 200 scfh total for all four main steam lines. Therefore, the proposed change to increase the allowed MSIV leakage rate does not create any new or different kind of accident from any accident previously evaluated.

The proposed change to eliminate the LCS does not create the possibility of a new or different kind of accident from any accident previously evaluated because the removal of the LCS does not affect any of the remaining LGS, Unit 1 and Unit 2 systems, and the LOCA has been re-analyzed using the proposed attemate l method to process MSIV leakage. The associated proposed change to delete the LCS isolation valves from i TS Table 3.6.3-1 does not create the possibility of a new or different kind of accident, since the affected main

'~

steam piping will be welded and/or capped closed to assure that the primary containment integrity, isolation, and leak testing capabliity are not compromised.

Therefore, as discussed above, the proposed changes do not create the possiblilty for any new or different l kind of accident from any accident previously evaluated.

l 3. The orooosed TS chances do not involve a sionificant reduction in the marcin of safety.

The proposed change to TS Section 3.6.1.2 to increase the MSIV a!!owable leakage does not involve a significant reduction in the margin of safety. As discussed in the current Bases for TS Section 3/4.6.1.2, the

( allowable leak rate limit specified for the MSIVs is used to quantify a maximum amount of leakage assumed l_ to bypass primary containment in the LOCA radiological analysis. Accordingly, results of the re-analysis

! supporting these proposed changes are evaluated against the dose limits contained in 10CFR100 for the off-site doses, and 10CFR50, Appendix A, GDC 19, for the MCR dosea. As discussed above, sufficient margin relative to the regulatory limits is maintained even when assumptions and methods (e.g., RG 1.3) that are considered highly conservative relative to more realistic assumptions and methods, are used in the analysis.

Attachment 1 Page 8 l

Results of the radiological analysis demonstrate that the proposed changes do not involve a significant reduction in the margin of safety. The whole body doses, in terms of margin of safety, are insignificantly reduced by 0.1% at the LPZ,1.0% in the MCR, and by 0.004% at the EAB. The margin of safety for thyroid doses is reduced by 12.1% at the LPZ,20.9% in the MCR, and 0.007% at the EAB. The margin of safety for beta dose is insignificantly reduced by 2.0% in the MCR. These reductions in the margin of safety are not significant since the revised calculated doses are highly conservative yet remain well below the regulatory limits, and therefore a substantial margin to the regulatory limits is maintained.

Furthermore, while the proposed changes will result in a calculated reduction in the margin of safety, this reduction is not significant when considering the increased reliability and capability of the proposed MSIV leakage treatment system, and also that the resulting increases in thyroid doses are dominated by the organic lodine fractions which occur because of the conservative source term assumptions used in this analysis. For a MSIV leakage rate of 100 scfh per main steam ilne, more than 95% of the off-site and MCR radiolodine doses are due to organic lodine resulting from using the RG 1.3 radiological source term and organic lodine converted from the elemental iodine deposited in main steam piping systems. NRC paper SECY 92-127, " Revised Accident Source Terms for Ught-Water Nuclear Power Plants," and draft NUREG-1465, " Accident Source Terms For Light-Water Nuclear Power Plants," identify that, based on current research and data, the production of radiolodine from fuel damage is significantly less than that assumed in RG 1.3. Although the NRC has not yet defined the percentage of iodine to be assumed in the new radiological source term model, the percentage is generally accepted as being significantly less than is currently assumed and will result in a reduction of the organic lodine term by as much as a factor of 15.

The proposed change to eliminate the LCS from TS does not reduce the margin of safety. In fact, the overall margin of safety is increased. The function of the LCS for MSIV leakage treatment wl!! be replaced i by alternate main steam drain lines and condenser equipment. This treatment method is effective in l

reducing the dose consequences of MSIV leakage over an expanded operating range compared to the capability of the LCS and will, thereby, resolve the safety concem that the LCS wl!I not function at MSIV leakage rates higher than the LCS design capacity. Except for the requirement to establish a proper flow path from the MSIVs to the condenser, the proposed method is passive and does not require any new logic control and interlocks. This proposed method is consistent with the philosophy of protectbn by multiple l barriers used in containment design for limiting fission product release to the environment. Furthermore, as previously identified, based on the evaluations discussed in NEDC-31858P, Revision 2, and the seismic evaluation provided in the Attachment 3 report, *MSIV Leakage Attemate Drain Pathway Evaluation," the l

design of the MSIV leakage attemate drain pathway, meets the intent of the 10CFR100, Appendix A t requirement for seismic qualification. Therefore, the proposed method is highly reliable and effective for MSIV leakage treatment.

1 l The revised calculated LOCA doses remain within the regulatory limits for the off-site and the MCR.

l Furthermore, the revised calculation shows that MSIV leakage rates greater than 200 scfh for all four main steam lines would not exceed the regulatory limits. Therefore, the proposed method maintains a margin of i safety for mitigating the radiological consequences of MSIV leakage beyond the proposed TS leakage rate j limit of 100 scfh per main steam line, not to exceed a total of 200 scfh for all four main steam lines.

The proposed change to delete LCS isolation valves from TS Table 3.6.3-1 does not reduce the margin of safety. Welded and/or capped closure of the LCS lines assures that the primary containment integrity and leak testing capability are not compromised. These welds and/or caps will be periodically leak tested as l

part of the CILRT. Therefore, the proposed deletion of the LGS isolation valves does rot involve a reduction in a margin of safety, l

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Attachment 1 Page 9 Accordingly, based on the above reasons, the proposed changes do not involve a significant reduction in a margin of safety.

Information Sucoortina an Environmental Assessment The proposed Technical Specifications (TS) changes have been evaluated against the criteria in 10 CFR 51.22 for the identification of licensing and regulatory actions requiring an environmental assessment. We have concluded that the proposed changes may not meet the criteria for categorical exclusion as defined in 10CFR51.22(c) (9).

Therefore, in accordance with the requirements in 10 CFR 51.30, the following information is provided to support an Environmental Assessment.

1) .Need for the Prooosed Chance The proposed Technical Specifications (TS) changes are requested for Umerick Generating Station (LGS),

Units 1 and 2, because the elimination of the Main Steam Isolation Valves (MSIVs) Leakage Control System (LGS) and the proposed increased a!!owable leakage for the MSIVs will provide a more effective and realistic means to ensure that the resulting doses from the proposed increased MSIV leakage in the un!!kely event of a Loss of Coolant Accident (LOCA) do not exceed regulatory limits.

Additionally, the proposed changes will provide an economic benefit to PECO Energy Company (PECo) by eliminating the high maintenance and operational expenses associated with the LCS. Although the calculated accident doses resulting from the proposed changes are higher than those previously calculated, the capacity of the a!!smate pathway is significantly higher than that of the LCS. Furthermore, the proposed changes wl!I result in a reduction in occupational exposure from the reduction in necessary maintenance work.

2) Altematives and Attemative Use of Resources if the proposed TS changes are not approved, the LCS and MSIVs will continue to be maintained to the current requirements. Maintenance of the LCS and the MSIVs will continue to require the expenditure of manpower and the exposure of maintenance workers to occupational radiation. For LGS, Unit 1 and Unit 2, several hundred man-hours per cycle is spent on rnalntaining the LCS and the MSIVs. Compared with maintaining the current requirements, implementation of the proposed changes will require less occupatfonal radiation exposure per operating cycle and a significant reduction in the expenditure of resources.

Inasmuch as there are no unresolved conflicts concerning the availability or use of the above described attemative associated with the proposed changes, no further evaluation of attematives is required.

3) Environmentalimoact of the Prooosed Action Approval of the proposed TS changes will not result in any significant effect on the human environment.

l This conclusion includes consideration of tf a potential impact of increased off-site and Main Control Room (MCR) doses that would result in the unlikely event of a Loss of Coolant Accident (LOCA). Although the l resulting calculated off-site and MCR doses will increase, the dose levels will remain well within the l

applicable regulatory limits. In addition, as discussed in NUREG 1169,

• Resolution of Generic issue C-8l'

, dated August,1986, the proposed attemative pathway provides a capacity significantly greater than that of i I

the currently installed LCS and therefore, will provide greater protection against potential radiological releases in the unlikely event of a LOCA.

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Attachment 1 70ge 10 The impact on the radioactive effluent discharged from the LGS site during and following a LOCA will remain within the existing regulatory limits. As shown by calculation based on conservative assumptions and methods, the resulting off-site and MCR doses will be increased in the unlikely event of a LOCA, but the calculated doses remain well within the regulatory limits. Since the systems involved in the proposed changes only serve to mitigate the consequences of an accident, the proposed changes or the attematives  ;

will not directly result in a release to the environment during normal plant operation.

The impact of the proposed changes on the generation of low level radioactive waste will be positive. By ,

reducing required maintenance on contaminated systems, there should be a net reduction in the amount of radioactive waste generated.

The impact of the proposed changes on occupational exposure will also be positive. By reducing the amount or eliminating the need for maintenance required to be performed on the affected contaminated .

systems, there will be a net reduction in the occupational exposure.  !

Finally, there are no non-radiological impacts resulting from the implementation of the proposed changes.

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4) Conclusion Therefore, we have concluded that the NRC does not need to prepare a supplemental environmental impact statement in connection with the issuance of the proposed TS Changes, and that a finding of no significant impact is supported by the above discussion.

Conclusion The Plant Operations Review Committec and the Nuclear Review Board have reviewed these proposed changes to i the Limerick Generating Station, Unit 1 and Unit 2 Technical Specifications, and have concluded that they do involve an unreviewed safety question; however, they do not involve a significant hazards consideration, and will not endanger the health and safety of the pub!!c. i t

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