ML20116H689

Application for Amend to License NPF-39,consisting of TS Change Request 92-15-1 to Extend Allowed Outage Time for RHR Svc Water Sys & Suppression Pool Cooling Mode of RHR Sys from 72 H to 168 H

ML20116H689
Person / Time
Site:	Limerick
Issue date:	11/06/1992
From:	Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20116H695	List: ML20116H689 ML20116H707
References
NUDOCS 9211130236
Download: ML20116H689 (14)

Text

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10 CFR 50.90 PIIILADELPIUA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 95!,-65 CHESTERBROOK BLVD.

WAYNE, PA 19087-5691 (215) 640-6000 November 6, 1992 Docket No. 50-352 NUCIIAR SERVICI,5 DEPARTMINT License No. NPF-39 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Limerick Generating Station, Unit 1 Technical Specifications Change Request Gentlemen:

Philadelphia Electric Co.npa ny (PECo) is submitting Technical Specifications (TS) Change Request No. 92-15-1, in accordance with 10 CFR 50.90, requesting an amendment to the TS (Appendix A) of Operating License No. NFF-39 for Limerick Generating Station (LGS), Unit 1. This submittal requests a one-time (i.e., temporary) TS change to extend the

, allowed outage time for the Unit 1 Residual Heat Removal Service Water l (RHRSW) system and the Suppiession Pool Cooling Mode of the Residual Heat Removal (RHR) system from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., seven days) during the second Unit 2 refueling outage in order to allow continued l Unit 1 operation while upgrades are made to the 'B' RHR heat exchanger outlet valve on both units. During this maintenance activity, Unit 2 will comply with the applicable Unit 2 TS allowed outage time.

Information supporting this Change Request is contained in Attachment 1 to this letter, and the proposed replacement pages for the LGS Unit 1 TS are contained in Attachment 2.

We request that, if approvei, the amendment to the LGS Unit ; TS be effective upon issuance.

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

I Very truly ycurs,

/J G. T. Beck, Manager Licensing Section Attachments l cc: T. T. Martin, Administrator, Region I, USNRC w/attac:hments l T. J. Kenny, USNRC Senior Resident Inspector, LGS w/ attachments

%. P. Dornsife, Director, PA Bureau of Radiological Protection '

w/ attachments l 9211130236 921106 PDR ADOCK 05000352 d

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. COMl40!iWEALTl! OF PEliNSYLVANIA  :

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COUNTY OF CHESTER  :

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

that he is Vice President, LimerSck Generating Station, Philadelphia Electric Company, the Applicant herein; that he has read the enclosed Technical Specifications Change Request No.

92-15-1 for Limerick Generating Station Unit 1, Facility ,

Operar ing License No. NPF-39 and knows the contents thereof; and tiat the statements and matters set forth therein are true and correct to the best of his knowledge, infornation and belief.

Vice Pro 1 nt Subscribed and sworn to before me this I nay of ), pp&nd /t, 1992.

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10 CFR 50.90-PIIILADELPIIIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBROOK BLVD.

WAYNE, PA 19087 5691-(215) 640-60'JO November 6, 1992 Docket No. 50-352 NUCLEAR SERVICES DEP'.RTMENT License No. NPF-39 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Limerick Generating Station, Unit 1 Technical Specifications Change Request Gentlemen:

Philadelphia Electric Company (PECo) _ is submitting Technical Specifications (TS) Change Request No. 92-15-1, in accordance with 30 CFR 50.90, requesting an amendment to the TS (Appendix A) of Operating License No. NPF-39 for Limerick Generating Station (LGS), Unit 1. This submittal requests a one-time (i.e., temporary) TS change to extend the allowed outage time for the Unit 1-Residual Heat Removal Service Water (RHRSW) system and the Suppression Pool Cooling Mede of the Residual Heat Removal (RHR) system f rom 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e. , seven days) during the second Unit 2 refueling outage in order to allow continued Unit 1 operation while upgrades are made to the 'B'_RHR heat exchanger.

outlet valve on both units. During-this maintenance activity, Unit 2 will comply _ with the applicable- Unit 2 TS allowed outage time.

Information supporting this Change-Request is contained in Attachment 1 to this letter, and the proposed replacement pages for the LGS Unit 1 TS are contained in Attachment 2.

We request that, if approved, tne amendment to the LGS Unit 1_TS be effective upon issuance.

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

Very truly yours, G. .-beck, Hanager

, Licenuing Section Attachments cc: T. T. Martin, Administrator, Region I USNBC w/ attachments T. J. Kenny, USNRC Senior Resident Inspector, LGS w/attacknents W. P.~Dornsife, Director, PA' Bureau of Radiological Protection w/ attachments-

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

COMNOliWEALTil OF PENNSYLVANIA  :

ss. .

• c COUNTY OF CliESTER  :

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

'I h a t he is Vice President, Limerick Generating Station, Philadelphia Electric Company, the Applicant herein; that he has read the enclosed Technical Specifications Change Request No.

92 15-1 for Limerick Genert. ting Station Unit 1, Facility ----

Operating License No. NPP-39 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.

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Vice Pre i .nt Subscribed and sworn to before rae this [F day

-of ggnc /L 1992.

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ATTACHMENT 1 LIMERICK GENERATING STATION UNIT 1 r

a Docket No. 50-352 i

License No. NPF-39 TECHNICAL SPECIF1 CATIONS CHANGE REQUEST No. 95-15-1 "One-Time Technical Specifications Change to Extend the-Allowed Outage Time for the Ros1&n:1 Heat Removal Service -Water System and the-Suppression Pool Cooling Mode of the Residual Heat Removal Sy cem -. Unit 1" Supporting Information for Changes - 9 pages.

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Attachment 1 i -

.Page.-1 PhiladeJ phia Electric Company (PECo), Licensee under Facility _ Operating License NPF-39 for_the Limerick Generating Station (LGS) Unit 1, requests that the Technical Specifications (TS) contained in - Appendix A- to the Operating License be amended as proposed herein to allow for a one-time (i.e, temporary) extension in the allowed catage tJme (A0T) for the Residual Heat Removal Service Water (RHRSW) system and the Supnression Pool Cooling (SPC) mode of the Residual Heat Removal (RHR) system from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> .

(i.e., seven days) during the second Unit 2 rnfueling outage - to - - allow 1

continued Unit 1 operation while upgrades are mede to the ' B ' RHR heat exchanger outlet valve on both units. The proposed changes would_ involve adding a one-time provision to TS Sections 3.6.2.3, " Suppression Pool Cooling," and 3.7.1.1, " Residual Heat Removal Service Water System - Commor.

System." We propose that TS Section 3.6.2.3, Action a, and TS Section 3.7.1.1, Action a.3, be changed such that a 168-hour period be authorized for continued operation of Unit I although-the 'B' RHRSW loop and the Unit 1 'B' RHR heat exchanger will be inoperable during this 168-hour period. These one-time TS changes are requested to avoid a Unit I shutdown while the SPC mode of the 'B' loop of the RHR syste.m and the RHRSW system are inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, in accordance with Unit 1 TS Sections 3.6.2.3, Action a, and 3.7.1.1, Action a.3. The proposed changes to the LGS Unit 1 TS are indicated by vertical bars in the margin of the TS pages 3/4 6-16 and 3/4 7-

1. The proposed TS changes are contained in Attac; ment 2.

During the maintenance activity described above, Unit 2 will be in a refueling outage end wil.1 comply with Unit 2 TS allowed outage time.

Therefore, no changes to Unit 2 TS are required.

I This change request for LGS Unit 1 provides a discussion and description of the proposed TS changes, a safety asressment of the proposed TS changes, ir. formation supporting a finding or No Gignificant Hazards Consideration, and information supporting an Environmental Assessment.

We request that, if approved, the amendment to the LGS Unit 1 TS be effective upon issuance.

Discussion and Dencription of the Proposed Changes This proposed TS char ge request involves a one-time change- to the LGS Unit 1 TS to extend the aJ 1 ond outage time (A0T' for the Residual Heat Removal Service Water F2RSW) system and Suppression Pool Cooling _(SPC) mode of the Residual Heat F.emoval (RHR) system from 721 hours0.00834 days <br />0.2 hours <br />0.00119 weeks <br />2.743405e-4 months <br />'to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.c.,

f seven days). This is to allow maintenance to be performed on the-RHR heat-l exchanger service water outlet valves HV-51-1F068B- and HV-51-2F068B on _- the Unit 1 'B' RHR heat exchanger and the Unit 2 'B'_ RHR heat exchanger, respectively. This change request proposes that " TS Section 3.7.1.1 be.

modified to allow one subsystem of RHRSW to be -inoperable for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., seven days) and TS Section 3.6.2.3 be modified to allow the SPC moca of one RHR subsystem to be inoperable for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e. , seven days) curing the second Unit 2 refueling outage.

l These one-time TS changes are required to allow adequate time for-L maintenance on the 'B' RHRSW loop (i.e., a common system) while avoiding a l Unit I sheldown. During this ma intenanca - i.ctivity , Unit 2 will be in a l refueling cutage and will comply with the applicable Unit 2 TS AOTs for RHRSW .

l and the SPC_ mode of~RHR.

l l

The maintenance to be performed is an upgrade of the HV-51-1F068B and-HV4?-2F068B valve internals with stainless steel components. _These.valvas= -

Attachment 1 Page 2 are used to -isolate and throttle RHRSW flow through the RHR heat exchangers.

During throttling operations, the valves are subje-ced to harsh flow conditions which have caused degradation of the valves. The recommended solution is to upgrade the valve internals with stainless steel components.

Maintenance on the Unit 1 and Unit 2 'B' RHR heat exchanger RHRSW inlet valves HV-51-IF014B and HV-51-2F014B may also be pertormed during this maintenance period.

These RHRSW valves (i .e. , HV-51-lF0148, HV-51-2F014B, HV-51-lF068B, and HV H 48B) are unisolable and the 'B' RHFSW loop piping configuration requires the use of multiple freeze seals (i.e., up to seven) and partial-system draining in order to perform maintenanc: on these valves. The additional time required to establish multiple freeze seals, partial draining, and subsequent system restoration of the 'B' RHRSW loop neccssitates t.'a need for these one-time TS changes.

Safety Assessment

'Ihis proposed TS change request involves a one-time (i.e., temporary) change to the LGS Unit 1 TS to extend the allowed outage time for the Unin 1 RHRSW system and the SPC mode of the RHR system from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., reven days) during the second Unit 2 refueling outage to allow continued Unit 1 operation while upgrades are made to the 'B' RHR heat exchanger outlet valve on both units. During this maintenance activity Unit 1 will remain in operation and Unit 2 will be in a refueling outage and will comply with the applicable Unit 2 TS AOTs. For the duration of this maintenance activity the 'B' RHRaw loop, the Unit 1 'B' RHR heat exchanger, and the Unit 2 'B' RHR heat exchanger will be rendered inoperable. Removal of the 'B' RHRSW loop, the Unit 1 'B' RHR heat exchangar, and the Unit 2_'B' RHR heat exchanger from service will not prevent any loop of RHR on either unit from injecting water into the respective reactor vessel in the Low Pressure Coolant Injection (LPCI) mode of operation.

The accidents potentially impacted by these proposed TS changes are the full range of Loss of Coolant Accidents (LOCAs) with and without a.

concurrent Loss of uffsite Power (LOOP). The-loss of shutdown rooling was-also considered. Any postulated accident occurring during this activity is bounded by previous analysis. The removal of the 'B' RHRSW loop from service w!l1 af fect the operability of the Unit 1 and Unit-2 'B' RHR: heat exchangers.

The RHR heat exchangers -provide a method of - decay her.t removal- and ~

suppression pool /drywell temperature control. Decay heat removal is a routine shutdown cooling mode of operation when the unit is shutdown. Two loops of shutdown cooling are required to be operable in accordance with'TS Sections 3.4.9.1 (i.e. , while the reactor is in Operational Condition (OPCON) 3-hot 1: itdown), 3. 4. 9. 2 - (i .e. , OPCON 4 - cold shutdown) , _ and - 3. 9.11. 2 (i .e . ,

LOPCON t

- refueling), or an alternate method of decay . heat removal is recuirt to be demonstrated. Unit l will be in OPCON 1 _(i.e., power operaticn), therefore these TS Sections are not applicable. However,- if . Unit 1 is required to be shutdown during the _ period the 'B' RHRSW loop is inoperable, alternate- decay heat removal methods are available -'such as establishing a shutdown cooling path tP ough the automatic depressurization system (ADS) valves or using the main condenser es a heat sink if offsite power .is avaLlable. These methods - will satis fy - the__ shutdown cooling requirements while the reactor is in OPCONs 3 and 4. In OPCCN 5, alternate-decay heat removal mtrhods such as the Reactor Water Cleanup system can be utilized to satisfy, the shutdown cooling requirements af ter a suf ficient' time;

~

Attachment 1 Page-3

• after plant shutdown.

Suppression pool /drywell temperature control is an accident mitigation function of the RHR system The RHR system accomplishes this function by two modes of operation, suppression pool spray and suppression pool cooling, both of which utilize the RHR beat exchangers. TS Section ) 6.2.2 requires that two loops of the suppression pool spray mode of RHR system be operable ;n OPCOPs 1, 2, and 3. The AOT for one loop of the suppression pool spray mode is se"en days. The planned 'B' RiqSW valve loop maintenance will be comple ed within this AOT, therefore, no change is required. TS Section 3.6.2.3 requires that two loops of the SPC mode of the RHR system be operable in OPCONs 1, 2, and 3. The AOT for one loop of the SPC mode being inoperable is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This change reques' proposes that this AOT for Unit.1 be extended to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., seven days) during the second Unit 2 refueling outage.

! The LGS Updated Final Safety Analysis Report (UFSAR), Section 6.2.2, states that one operable RHR heat exchanger is adequate for accident l mitigation. Two cases of one operable RHR heat exchanger during a postulated accident are presented. In the first case, the operable RHR heat exchanger is placed in service in the RHR drywell spray mode while one RHR pump in Lo"I mode of operation and one Core Spray loop inject water into the reactor vessel. In the other case, the operable RHR heat exchanger is placed in service along with an associated RHR pump taking suction from the suppression pool and discharging to the reactor vessel. The flow from the RHR-pump is cooled by RHRSW flow through the RHR heat exchanger before being discharged into the reactor vessel while another RhR pump, in LPCI mode of operation, and one Core Spray loop inject directly into the_ reactor vessel. Both cases assume a LOOP and that the High Pressure Coolant Injection (HPCI) system-is available for the entire accident. Other assumptions include: initial suppression pool temperature and RHRSW temperature are at their maxiraum, all the decay heat fron the reactor _ vessel is rejected through the RHR heat exchanger, and the RHR heat exchanger is in a fully fouled condition. The peak contninment pressura is higher for the second case, but is-still much less than the containment design pressure. This analysir is for a rupture of a reactor recirculation lue and is the bounding event for - similar -

occurrences. TS Section 3.7.1.1 requires that two loops of RHRSW be operable ir OPCONs 1, 2, and 3. The AOT for one loop of RHRSW being inoperable, which renders the associated RHR heat exchanger inoperable is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This change request proposes that this AOT be extended to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., -seven days) during the second Unit 2 refueling outage. Thc RHRSW system was designed with sufficient capacity so that one loop of RHRSW with two RHRSW pumps-in operation and two spray pond spray networks can-mitigate a-Design-Basis Accident (DBA) on one unit and a'. low the safe shutdown of~the other unit as described in the UFSAR Section 9.2 ..

Unit 2 will be in a refueling outage wit! 'd shutdown conditions established. The 'B' RHRSW loop will tot be removed frcm service until approximate)y day 20 of the outage._ The accay heat generation of Unit 2 will have been reduced.from 146 MWt at time of shutdown to approximately 3.5 Mwt

.by day 20. The 'A' RHRSW loop may be in operation to support Unit 2 shutdown

-ccoling requirements. However, due to prior establishment of cold shutdown

-conditions, the reduction in-decay heat gaceration and tne ability to place the Spent Fuel Pool Cooling and Cleanu c .,ys tem .in service, _ along with ' a =

recirculation pump or i RHR pump for reactor core circulation as an alternate decay heat- rek. oval .aethod, the Unit 2 heat removal requirements-on-the 'A'

-RHRSW _ loop needed to maintain cold shutdown conditions will be minimal.

F.Irthermore drafning of the Unit 2 reactor cavity will not be allowed until

Attachment 1 Page 4 the 'L' hFRSW loop is returned to service or an alterne.te decay heat removal method is available. Since one . loop of RHRSW can mitigate a DBA on one unit and support the safe shutdown of the other unit, the potential heat removc1 tequirements with respect to Unit I during the period that these proposed TS changes will be in effect:is within the capacity of the 'A' RHRSW loon.

The above discussions do not account for a single failure that could render the operable 'A' RHR heat exchanger or operable 'A' RHRSW loop inoperable during the proposed extended AOT. By limiting the time the 'B' i PHRSW loop is out of service and maintaining the Unit 1 'A' RHR heat exchanger, the 'A' RHRSW loop, and assoc lated equipment- /sys tem operable during that period, the consequenccs of an accident previously evaluated will ,

remain uncharged. The components that have the potential of preventing the Unit 1 'A' HHR heat exchanger or the 'A' RHRSW loop from performing their safety function if they were to fall tre listed below.

'A' RHR heat exchanger, shell side (i.e. ,RHR system flow side) outlet valve:

liv-51-1F003A (normally open - safety function apen).

'A' RI:R heat exchanger, RHRSW inlet valve:

HV 1.' ' 14 A (normally closed - safety function open).

'A' RHP heat exchanger, shell side bypass valve:

HV-C-51-lF048A (normally open - safety function throttled / closed).

'A' RHR heat exchanger, RHRSW outlet valve:

HV-51-1F068A (normally closed - safety function throttled /open)

'A' RHRSW spray por.d spray nozzle. ir.let valve:

HV-12-032A (normally closed - safety function open).

'C' RHRSW spray pond spray nozzle, inlet valve:

IIV-12-032C (normally closed - safety function open).

I A review of the maintenance records indicates that there was one occurreace of the Rl!RSW in'et valve, HV-51-1F014A, failing to_open. This valve was later tested and it operated properly. The RHRSW outlet valve, HV-SI-lF068A, has also experienced improper operation, however, the valve internals were replaced during Unit 1 fourth refueling outage, and the valve is currently operating properly. No other occur;snces of the other valves falling-to function properly were noted. Therefore, the prooability'oi a wa1 function of any of these valves preventing the 'A' RHR heat exchanger or

'A' RHRSW loop _from performing their safety function is considered to be cdnimal.

The above mentioned valves are powered from Class 1E safeguard buses.

HV-C-051~-lF04Be, HV-51-1F014 A ar.d HV-12-032 A are powered f rom Unit i division i 1-power supplies. HV-51-lF068A is powered from . Unit 1 division 3 power supplies. HV-12-032C is powered from Unit 2 division 3 power supplies.

Therefore, to ensure emergency electrical power is available to these valve l during a LOOP, the D11, D13, and D23_ emergency _ diesel generators (EDGs)-will l- be required to remain operable during the= proposed extended AOTs. ]

l \

A complete cleaning and inspection of the tube side of th6 Unit 1 'A' 1 RHR heat exchar"er was performed during the fourth Unit I refueling outage. 1

. Although some LLoe pitting occurrer' Nhich required some tubes to be plugged,-

L the heat removal capacity was_ evaluated-to be 202 x 10 5 BTU /HR. This is 65%

l.

Attachment 1 Page 5 above the minimum design capacity of 122 x 10' BTU /HR.

The removal of the 'B' RHRSW-loop frcm service will not affect the capability of any emergency core cooling systams (ECCS) from injecting water into the reactor vessel. The RHRSW system is manually operated and is not required during the first ten minutes of a LOCA. Therefore, the short-term (i.e., less than ten minutes) emergency core cooling capability of Unit 1 ECCS is unaffected. Long-term actions (i.e., greater than ten minutes) will '

be affected to the extent that only the 'A' RHR heat exchanger will be operable for long-tera heat removal. Long-term cooling requirements will be met by the Lait 1 'A' RHR heat e uhanger and the 'A' RHRSW loop with the RHR system in either the containment spray or the JPC mode of operation, as discussed earlier. Further removal of Unit 1 equipment / systems- will be allowed in accordance with existing TS requirements as long as the removal of equipment / systems from service does not adversely affect the operability of the 'A' RHRSW loop or the operable SPC mode of RHR operation or places the unit outside of the analysis described in UFSAR Section 6.2.

The Emergency Service Water (ESW) system will be af fected by the removal

. of the 'E' RHRSW loop from service in that all ESW return flow will be to the

'A' RHHSW loop. This alignment is within the design capabilities of the ESW system and will be controlled by approved procedures. A computer analysis of the flow distribution to components cooled by ESW was performed. This analysis indicates that sufficient ESW flow is available to support rperability of essential components. Measurements of ESW system flow will be performed on the most limiting components as determined by the comcuter

analysis to validate component / system operability. Physical work on tne 'B'

! RHRSW loop will not begin until the ESW system (i.e. , both loops) is ver:fied operable.

The Unit 2 Turbine Enclosure Cooling Water (TECW) system will also be af f ected such that the ESW system will not be available to backup the Service Water system as the cooling medium for the Unit 2 TECW system. The ESW system is ut111:ad as the cooling medium for the TECW system in the event of a LOOP. The ESW return path from the Unit 2 TECW heat exchangers to the 'B' RHRSW loop will be isolated during the 'B' RHRSW loop maintenance period preventing the ESW system from being capable of removing heat from the Unit 2 TECW system. TECW is non-safety related and is norm ally aligned to the

, Service Water system. Since Unit 2 will be in a refueling outage, loss of

the ESW system's ability to remove heat from the Unit 2 TECW system will'have no adverse impact.

The concern associated with the proposed TS change is the reduced margin L of safety incurred by extending the applicable AOTs. The RHRSW system is designed such that the AOT for operation with less than three RHRSW pumps e operable along with their associated operable EDGs is limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to ensure adequate decay heat removal capability is available for the. design-i accident scenario of a LOCA/ LOOP on one unit and simultaneous safe shutdown

.of the other unit. The reduction in the margin of safety due to increasing the applicable AOTs from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> in a degraded condition of the-plant la considered minimal as discussed below, since this reduction reflects the small increase in the probability that a LOCA/ LOOP event would occur on _

. Unit 1 within the proposed seven-day e r period as compared to ti.e

probability of a LOCA/ LOOP on Unit 1 during the three-day period allowed by TS. ,

g A Probaisilistic nisk Assessment (PRA) was performed for the conditions discussed above. _The cumulative-risk of a core damage event increased from l . - ~_ - ,

Attachment 1 Page 6 3.801-x 10 4 per reactor year to 1.012 x 10 4 per reactor year. This equates 's

--to a 1.4% increase from the baseline core damage _ovent risk for a three-day out-of-service porlod and a 3.2% increase f rom the baseline core damage event risk for a seven-day out-of-service period. Therefore, the increased risk

-(1.e., 1.8%) of extending the AOTs for TS Section 3.6.2.3, Action a, and-TS Section 3.7.1.1, Action a.3 to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., seven days) is judged to be minimal. ,

Information Supporting a rinding of No Significant Hazards consideration We have concluded that the propsed changes to the LGS Unit 1 TS to extend the AOTs for the 'B' RHRSW loop and the SPC mode of operation of the -

i

'B' RHR loop f rom 72 hot.rs to 168 hoars during the second Unit 2 refueling outage, 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 proposed changes do not involve a s;:*.ficant increase in the probability or consequences of an accider % Oviously evaluated. ,

The proposed one-time Unit 1 TS chenges do not increase the

, consequences of an accident f rom any previausly evaluated. Extending the AOTs for the 'B' RHRSW loop and the SPC t. ode of operation of the 'B' RHR loop from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e. seven days) does not cause an (

increase in the probability of an a ciCent since the af fected systems are not accident initiators as defined ay the UFSAR. Maintaining the

'A' 'HRSW loop and the SPC mode of operation of the 'A' RHR Inop operable will ensure that the consequences of the accidents previouwiy .

evaluated will remain bounded by the UFSAR Safety Analysis. Therefore, there is no increase in the consequences of an accident. This conclusion is based on the following considcrations.

1. a. Removal of the 'B' RHRSW loop and its associated Unit 1 'B' RHR heat excnanger will not prevent any ECCS (i.e., LPCI, Core Spray, :_

HPCI) from injecting water into the reactor vessel. Short-term <

mitigation of an accident is unaf fected since RHRSW is manually

, operated and is not required to be placed in service during the .

firs

• 10 minutes of an accident.

b. For long-term response, accident analysis discussed in UFSAR Section 6.2 indicates that one loop of RHRSW and one RHR heat exchanger are capable of removing the decay heat f-om both units assuming a LOCP/LOCA on one unit and safe shutdown on the other-unit. ,

c. Unit 2 will be in a refueling outage. The heat load on the 'A' RHRSW loop from Unit 2.w111 be minimal based on long time after y shutdown when the 'B' loop of RHRSW will be removec from servi.ce, i

and the availability of. alternate decay heat removal methods that do not reject heat to RHRSW.

d. The RHRSW and ESW systems are designed with suf ficient : capacity B such that one loop of ESW (i .e. .gne pump in operation) using one spray pond return header and tv. s spray networks is the minimal alignment required to mitigate a LOCA with a concurrent LOOP on one unit (i.e., in this case Unit 1) and a safe shutdown on the other unit.

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Attachment l Page 7 The unit undergoing an accident would receive 100% of the required. '

RHRSW flow to its RHR heat exchanger and the unit undergoing ~a normal shutdown would receive 67% to 100% of the required RHRSW: flow to its RHR j

' heat exchanger. Sixty-seven percent RHRSW flow to the unit undergoing a normal shutdeen is suf ficient to remove the heat -load- transferred

.through the RHR heat exchanger as diacussed 11n UFSAR Section '9.2.3.-

However, since Unit 2 will already be in cold shutdown, the heat removal requirements ar.d therefore the required RHRSW flow will be substantially_

less than 67%.

The protability for a single failure to occur and render the operable

'A' RHR hear. exchanger or operable 'A' RHRSW loop inoperable du-ing the proposed extended ICTs, han been evaluated and the conclusion is th;n there is no increase in tN exist ng probaullity for a single failure as a result of these proposed TS changw.

Therefore, implementation of the proposed 168-hour AOS aill not result in an increase in the probability or consequences of an accident previously-ovaluated.

2) The proposed changes do not create the possibility of a new or dif ferent kind of accident from any accident previously evaluated.

Since the proposed changes will not result _ in . any new plant configuration, system alignment, or operational proced_ures, the ,

possibility of a new cr different kind of accident is rot. created.

The systems af fected are not accidc7t initiators. The plant has-been_ ,

analyzed for one RHRSW loop out of service. Plant operation and accident mitigation utilizing one loop of RHRSW and:one RHR heat exchanger is described in UFSAR Sections 5.4, 6.2, and 15.2. The operable systems that will be affected during the implementation of-these proposed one-time TS changes will be werated within their design capabilities under appro"ad procedures. The removal of-one RHRSW loop and its associated RHR heat exchanger from service is currently allowed by TS. - These proposed one.-time TS changes will only exte.td the subsystem I.OTs for the RHRSW -system and SPC mode :of.

operation of the RHR_ system from 72. hours to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> _(i.e., seven days) during the second UnJt 2 refueling outage.

I The proposed changes will not cause the components important to '

4 safety that have been discussed above to be challenged by a dif ferent type of mal function, since no new type of malfunction will be created-by any operation associated with this activity.4 Therefore, these proposed changes do not create the possibility-of a new or different kind of accident from any accident previously evaluated.

3) The proposed changes do not involve a significant reduction in a margin of 62fety.

The RMRSW system and;th_e'RHR system are designed with-sufficient redundancy such that the removal % m service of-a componentLand/or 4 subsystem will not prevent the syscem from performing its required-safety function. Since-removal of the 'B' BHRSW loop from ser71ce with Unit- l 'in. operation ;and Unit 2 in: a ' refueling outagelis allowed by existing TS, the concern is the reduced margin of safety incurred

.by extending the applicable AOTs.

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Attachment 1- ,

Page 8~ i The RURSW system is designed such that the AOT for, operation with less than three RHRSW pumps operable along with- their associated operable EDGs is limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to ensure adequate decay heat.

removal capability is available for the design accident scenario of a LOCA/ LOOP on one unit and simultaneous safe shutdown of-the other-unit. The reduction in the margin of safety due to increasing the -.

applicable AOTs Irom 72 hourc to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> in a degraded condition of '

the plant is .:onsidered minimal as discussed below, since this reduction reflects the small increase in t he probability that a LOCA/ LOOP event would occur on Unit I within the proposed seven-day AOT period as compared to the probability of a LOCA/ LOOP on Unit I during the three day period allowed by TS.

A PRA was performed for the conditions discussed above. The cumulative risk of a core damage event increased from 3.001 x 10 4 per-reactor year to 1.012 x 10 4 per reactor year. This equates to a 1.4%

increase f rom the baseline core damage event risk for a three-day out-of-service period and a 3.2% increase from the baseline core darage event risk for a sev(a-day out-of-service period. This increased risk of a core damage event (i.e., 1.8%) of extending the AOTs for TS .

Section 3.6.2.3, Action a and TS Section 3.7.1.1, Action a.3.to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> (i.e., seven days) is judged to be minimal.

In addition, the following equipment and/or systems w411 be required to be operable for the duration of the proposed extended AOTs, or the Actions of TS Sections 3.6.2.3.b and 3.7.1.a.4 must be followed.

- Unit 1 'A' RHR heat exchanger and associated equipment.

'A' RHRSW loop 4Td associated equipment.

- HPCI (Unit 1).

- Dil, D13, and D23 EDGs and associated equipment.

- Any other equipment that, if removed, would place Unit l'outside .

the bounds of the UFSAR analysis described _ in Section . _6. 2 (1.e. , minimum number of ECCS, FHRSW loops, and ESW loops needed for accident mitigation).

One intended action is to maintain the Unit 1 suppression pool temperature as low as possible during the period these proposed changes are implemented. This will.-increase the hcat storage capacity of the suppression pool and further enhance the heat removal capacity of the 'A' RHRSW loop. Also, since Unit 2 will be in a refueling outage, the Unit 2 decay heat removal: demand or the RHRSW system will'-

be minimal. -Therefore, implementation of the-proposed, one-time TS:

changes will not involve a significant -reduction in the ~ margin of safety.

Information Supporting an Environmental Assessment An environmental assessment is not remif red for the changes proposed' by this Chango _ Request because the requestec changes _ to _ the LGS : Unit 1 TS-

~

.. conform to the-criteria for " actions eligible for categoricaliexclusion" as:

specified in'10 CFR 51.22(c)(9). The' requested 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

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s-Attachment 1 Page 9-  ;

~

chan Jes . do not involve a significant change in the types f or significant increase in the amounts of any effluents that may be released offsite. In 1

addition, the proposed changes do not involve an -incrmase in individual or rumulative occupational radiation exposure.

Conclusion The Plant Operations Review Committee and the Nuclear Review Board have s reviewed i hose proposed changes to the LGS Unit 1 TS 'and have concluded that -l they do involve an unreviewed safety question, but that they do not involve significant hazards consideration, and will not endanger-the health and safety of the public.

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