ML20116L255

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Application for Amends to Licenses NPF-39 & NPF-85,revising TS Sections 3/4.3.1,3/4.3.2,3/4.3.3 & Associated TS Bases Sections 3/4.3.1 & 3/4.3.2 to Eliminate Selected Response Time Testing Requirements
ML20116L255
Person / Time
Site: Limerick  Constellation icon.png
Issue date: 08/08/1996
From: Hunger G
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20116L256 List:
References
NUDOCS 9608160181
Download: ML20116L255 (17)
v • d • e


Text

Station tupport Department k'- A 10 CFR 50.90 t

PECO NUCLEAR nm%c-Nuclear Group Headquarters I

A UMr ol PECO ENacy 965 Chesterbrook Boulevard i Wayne. PA 19087-5691 l

l August 8,1996 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 l i

s U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 i l l

Subject:

Limerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 96-09-0 Gentlemen PECO Energy Company is submitting Technical Specifications (TS) Change Request No.

9609-0, in accordance with 10 CFR 50.90, requesting an amendment to the TS (Appendix A) of Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively. This proposed TS change involves revising TS Sections 3/4.3.1, " Reactor Protection System Instrumentation," 3/4.3.2, " isolation Actuation Instrumentation," 3/4.3.3,

  • Emergency Core Cooling System Actuation Instrumentation," and the associated TS Bases Sections 3/4.3.1 and 3/4.3.2 to eliminate selected response time testing requirements. These i proposed TS changes are supported by analyses performed by the Bouing Water Reactor '

Owners' Group (BWROG) and documented in NEDO-32291, " System Analyses for the Elimination of Selected Response Time Testing Requirements," dated January 1994.

NEDO-32291 demonstrates that other periodic tests required by TS, such as channel calibrations, channel checks, channel functional tests, and logic system functional tests, in conjunction with actions taken in response to NRC Bulletin (NRCB) 90-01, " Loss of Fill-Oil in l Transmitters Manufactured by Rosemount," and NRCB 90-01, Supplement 1, are adequate to l ensure that instrument response times are within acceptable limits. 1 j By letter dated December 28,1994, the NRC indicated that response time testing requirements i can be eliminated from the TS for the selected instrumentation identified in NEDO-32291, and that this report is acceptable for reference in license amendment requests, provided licensees ,

comply with the provisions specified in the December 28,1994 letter, and supporting Safety l Evaluation Report (SER).

l l We request that, if approved, the amendments to the LGS, Units 1 and 2, TS be issued prior to January 24,1997, and become effective within 30 days following issuance.

9608160181 960808 PDR P

ADOCK 05000352.

PDR .

h)./ I 160059

- August 8,1996

, , Page 2 l

if you have any questions or require additional information, please do not hesitate to contact us.

Very truly yours, w Ag G. A. Hunger, Jr.  !

Director - Uconsing j r

Attachments l

Enclosure i l

cc: T. T. Martin, Administrator, Region I, USNRC (w/ attachments, enclosure) l N. S. Perry, USNRC Senior Resident inspector, LGS (w/ attachments, enclosure)  !

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

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COMMONWEALTH OF PENNSYLVANIA  : i

as.

COUNTY OF CHESTER  : I i

D. B. Fetters, being first duly sworn, deposes and says:

That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the foregoing Technical Specifications Change Request No.96-094 for Limerick Generating Station, Units 1 and 2, Facility Operating License Nos. NPF-39 and NPF-85, to eliminate selected response time testing requirements for Reactor Protection System Instrumentation, Isolation Actuation instrumentation, and Emergency Core Cooling System Actuation Instrumentation, 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.

( _, N .

. C Vice President Subscribed and sworn to before me this day of ( 1996.

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l ATTACHMENT 1 i

UMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50 352 50-353 ]

Ucense Nos. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 96-09-0

, Emergency Core Cooling System Actuation Instrumentation" Supporting information for Changes - 13 pages 4

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, ,, Attachment 1

, , Page 1 of 13 PECO Energy Company, Licensee under Facity Operating Ucense Nos. NPF-39 and NPF 85 for Umorick Generating Station (LGS), Units 1 and 2, respectively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating Ucenses be amended as proposed herein to revise TS to eliminate selected response time testing requirements for the Reactor Protection System (RPS),

Emergency Core Cooling System (ECCS), and Isolation Actuation instrumentation. The proposed TS changes are supported by analyses performed by the Bolling Water Reactor Owners' Group (BWROG) i and documented in NEDO-32291, " System Analyses for the Elimination of Selected Response Time Testing Requirements." The NRC indicated in a letter dated December 28,1994, that response time l requirements can be eliminated from the TS for the selected instrumentation identified in NEDO-32291,

! and that this report is acceptable for reference in license amendment requests, provided licensees satisfy the provisions specified by the NRC's Safety Evaluation Report (SER) issued for NEDO-32291. The proposed changes to the TS are shown on the attached mark-up of TS pages for Units 1 and 2, and are contained in Attachment 2.

l PECO Energy is requesting that, if approved, the amendments to the TS be issued by January 24,1997, and become effective within 30 days of issuance.

This TS Change Request 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 l Consideration, and information supporting an Environmental Assessment Discussion and Description of the Proposed Changes l

l Background The proposed Limerick Generating Station (LGS), Units 1 and 2. Technical Specifications (TS) changes involve revising the TS, and TS Bases sections, to eliminate response time testing requirements for selected plant components The TS sections affected by these proposed TS changes are: 1) TS 3/4.3.1,

" Reactor Protection System (RPS) instrumentation"; 2) TS 3/4.3.2, " Isolation Actuation Instrumentation *;

3) TS 3/4.3.3, " Emergency Core Cooling System Actuation Instrumentation"; and associated TS Bases . l Sections 3/4.3.1 and 3/4.3.2. The proposed TS changes are supported by analyses performed by the '

Bolling Water Reactor Owner's Group (BWROG) and documented in NEDO-32291, " System Analyses for Elimination of Selected Response Time Testing Requirements," dated January,1994. NEDO-32291 demonstrates that most d the fature modes identified by response time testing can be detected by personnel performing other periodic tests required by TS, such as channel calibrations, channel functional tests, and logic system functional tests. The analysis concludes that these tests, in conjunction with the actions taken in support of NRC Bulletin (NRCB) 90-01, " Loss of Fill-OH in Transmitters Manufactured by Rosemount," and NRCB 9041, Supplement 1, are adequate to ensure that instrumentation response times are within acceptable limits.

l The analyses documented in NEDO-32291 assert that the response time tests pioposed for elimination l are of little safety significance and result in unnecessary personnel radiation exposure, reduced avalab8ky of systems during plant shutdown, increased potential for inadvertent actuations of safety l systems, and a significant burden to utuity resources. The basis for these analyses is consistent with

! Regulatory Guide 1.118, Revision 2, " Periodic Testing of Electric Power and Protection Systems,' which ,

endorses IEEE 338-1977 which states:

l ' Response time testing of aN safety related equipment, per se, is not required if, in lieu of l response time testing, the response time of safety related equipment is verified by functional testing, calibration checks, or other test, or both. This is acceptable if it can be demonstrated

! that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine tests."

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Attachment 1

,. , Page 2 of 13 The evaluations documented in NEDO-32291 demonstrate that response time testing can be eliminated l for the following system cuiripci E l 1) All ECCS actuation instrumentation,

2) Sensors for selected RPS actuation instrumentation, and
3) Sensors for selected Main Steam Une Isolation Valve (MSIV) closure actuation instrumentation.

NEDO-32291 identifies the potential faRure modes of components in the affected instrumentation loops which could potentially impact the instrument loop response time. The primary basis is based on plant personnel detecting response time degradation in functional tests and/or calibrations. In addition, industry operating experience was reviewed to identify fatures that affect response times and how they were detected. The faBure modes identified were then evaluated to determine if the affect on response time would be detected by other testing requirements contained in the TS. The results of this analysis demonstrate that other TS testing requirements (i.e., channel calibration, channel checks, channel functional tests, and logic system functional tests), and actions taken in response to NRC Bulletin (NRCB) 9041, " Loss of FNI-Oli in Transmitters Manufactured by Rosemount," and NRCS 90-01, Supplement 1, are sufficient to klentify falure modes or degradations in instrument response times and assure operation of the analyzed instrument loops within acceptable limits. Furthermore, there were no falure modes identified that can be detected by response time testing than cannot also be detected by other TS required tests.

NEDO-32291 contains an evaluation of a delayed instrumentation response on the order of five (5) seconds for the trip functions selected for response time testing elimination. The five (5) second delay was chosen based on a survey of instrument and control (l&C) technicians from participating bouing water reactor (BWR) and selected pressurized water reactor (PWR) plants. Technicians were questioned about the maximum time it would take to identify sluggish component performance, no matter what individual cuiiipcs6.1 is sticking, sluggish, or broken. Fifty percent (50%) of the technicians surveyed estimated three (3) seconds, with eighty-five percent (85%) of the total technicians' estimates within five (5) seconds. A five (5) second delay provides reasonable assurance that changes in response time are detectable during functional testing and calibration checks. This five (5) second delay provides additional assurance that the consequences of this delay are not significant. I By letter dated December 28,1994, the NRC provided ks acceptance of NEDO-32291, subject to the following condklons, and indicated that this report is acceptable for reference in license amendment requests.

When submitting plant-specific license amendment requests (i.e., TS changes), licensees must confirm the applicabilty of the generic analysis of NEDO-32291 to their plant, and in addition to the request as shown in Appendix I of the topical report, the TS markup tables as shown in Appendix H, and a list of affected instrument loop components as shown in Appendix C.1, licensees must state that they are following the recommendations from EPRI NP-7243,

" investigation of Response Time Testing Requirements,' and therefore, are requiring the following actions:

a. Prior to installation of a new tiansmitter/ switch or following refurbishment of a transmitter / switch (e.g., sensor cou or variable damping components), a hydraulic response time test shaN be performed to determine an initial sensor-specific response i time value, and i

, b. For transmitters and switches that use capulary tubes, capulary tube testing shall be )

l performed after initial installation and after any maintenance or modification activity that l could damage the lines. -

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Attachment 1 Page 3 of 13 Licensees must also state the following in their requests:

a. That calibration is being done with equipment designed to provide a step function or fast ramp in the process variable,
b. That provisions have been rnade to ensure that operators and technicians are aware of the consequences of instrument response time degradation, and that applicable procedures have been reviewed and revised as racessary to assure that technicians monitor for response time degradation during the performance of calibrations and functional tests,
c. That surveulance testing procedures have been reviewed and revised if necessary to ensure calibrations and functional tests are being performed in a manner that aHows simultaneous monitoring of both the input and output response of units under test, I
d. That for any request involviry the elimination of response time testing for Rosemount pressure transmitters, the licensee is in full compliance with the guidelines of Supplement 1 to NRCB 9041, " Loss of FNIOR in Transmitters rnanufactured by Rosemount," and
e. That for those instruments where the manufacturer recommends periodic response time testing as well as calibration to ensure correct function, the licensee has ensured that elimination of response time testing is nevertheless acceptable for the particular application involved.

Description of Chances The foNowing (.hanges to the TS are proposed

1) TS Section 3/4.3.1,
2) TS Bases Section 3/4.3.1, " Reactor Protection System Instrumentation " wul be revised to make reference to NEDO-32291, as applicable.
3) TS Section 3/4.3.2, " Isolation Actuation Instrumentation," Table 3.3.2-3, wiX be levised to eliminate response time testing for applicable sensors for Reactor Vessel Water Level - Low, Level 1, and Level 2; Main Steam Line Pressure - Low; and Main Steam Line Flow - High.

Instrumentation response time requirements for the Residual Heat Removal (RHR) Shutdown Cooling Mode Isolation, Reactor Water Cleanup (RWCU) System isolation, High Pressure Coolant injection (HPCI) System isolation, Reactor Core isolation Cooling (RCIC) System l

isolation, and Primary Containment Isolation will be eliminated as a result of the proposed TS ~

changes. Further, table notations "a" and "**" wRl be deleted and "###" wRI be added to reflect' these changes.

4) TS Bases Section 3/4.3.2," Isolation Actuation instrumentation," wlN be revised to make reference to NEDO-32291, as applicable
5) TS Section 3/4.3.1, " Emergency Core Cooling System Actuation instrumentation," Table 3.3.3-3, l wIl be revised to include an annotation indicating that ECCS actuation instrumentation is

! eliminated from response time testing for Core Spray (CS), Low Pressure Coolant injection (LPCI) system, and HPCI system.

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, , Page 4 of 13 Safety Assessment l The elimination of response time testing requirements is based on the analysis provided in NEDO42291,

" System Analyses for Elimination of Selected Response Time Testing Requirements." The falure modes analysis documented in NEDO-32291 concludes that response time degradation of specific components can be detected by other TS required testing. The primary basis for this conclusion is based on plant personnel detecting response time degradation in functional tests and/or calibrations. The five (5) second delay evaluation discussed in NEDO-32291 provides addkional assurance that the consequences 1 of this delay are not significant. Analyses have been performed demonstrating that other periodic tests required by TS, such as channel calibrations, channel checks, channel functional tests, and logic system functional tests, in conjunction with actions taken in response to NRC Bulletin 90-01, " Loss of Ful-Oli in Transmitters manufactured by Rosemount," and Supplement 1, provide adequate assurance that instrument responses are within acceptable limits.

The proposed TS changes involve eliminating response time testing requirements associated with the following TS equipment.

1. Affected Equipment: RPS Level Low-Level 3, Sensors LT-042 *N000A-D Safety Function (s): Reactor Water Level Low-Level 3 The proposed TS changes do not affect the capabuky of the associated systems to perform their i intended functions, nor do the proposed changes affect the operation of any plant equipment l As described in LGS Updated Final Safety Analysis Report (UFSAR) Section 15.2.7, the low level trip indicates that the reactor water level has dropped, which is generally indicative of a problem with level control, or reactor feedwater system. A reactor scram is initiated on this condition by RPS to 3%tially reduce steam production. If the Residual Heat Removal (RHR) system is operating in the shutdown cooling mode, the isolation valves on the RHR suction piping are j closed to prevent further loss of vessel inventory via that path. The Automatic Depressurization '

System (ADS) receives a permissive signal for initiation to avoid inadvertent activation of the low

, pressure ECCS on a spurious high drywell pressure signal The TS requirements for this ,

function is 1.05 seconds. If this trip initiation time is increased by five (5) seconds, there would j be no significant impact on pint safety. The design basis for the Level 3 scram is the Loss d i Feedwater (LOFW) event. The Level 3 scram may occur during other events, but k would be a back-up function after other scram signals have occurred. The LOFW is a non-limking event for determination of tne core thermal limks. Therefore, a five (5) second delay in the scram )

actuation function would not affect plant thermal limits or fuel integrky. The core cooling d '

function for the LOFW event is provided by the HPCI and RCIC systems which will initiate on Level 2. A five (5) second delay in scramming would neither affect the capability of these systems to initiate nor prevent adequate core cooling.  !

2. Affected Equipment: MSIVs, Reactor Water Level Low-Level 1 Sensors LT-042 *N091 A-H Safety Function (s): Reactor Water Level Low-Level 1 The proposed changes do not affect the capabilky of the associated systems to perform their intended functions, nor do the proposed changes affect the operation of any equipment Abnormally low reactor water level is used to generate inklation signals for several functions, one of which is closure of the Main Steam Is* tion Valves (MSIVs) on Level 1. Fuel claddinr, integrity must be assured by the initiation of the ECCS systems. To limit the possibility of off-site release, the MSIVs wlN be closed at the low water level signal TS required response time for this trip is 1.0 seconds. If this trip initiation time is increased by 5 seconds, there would be no significant impact on plant safety. MSIV closure at low reactor water level would occur during

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i l i events which involve loss of reactor water inventory, such LOFW or Loss of Coolant Accident I (LOCA) events. Immediate valve closure is not required for core or plar,t safety. The reactor i would have been scrammed at Level 3 and MSIV closure does not affect core cooling.  :

j However, at these reactor water levels, there is no fuel damage and the radioactivity is limited to

the inventory in the steam lines. No fuel damage or increase in off-site releases would occur l i

j oven if there is a 5 second time delay in the MSIV closure under these conditions. Therefore,  !

the 5 second delay in the MSIV closure on low reactor water level does not affect plant safety l

3. Affected Equipment: Reactor Water Level Low-Level 1,2 Sensors LT 042 *N091 A-H I

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Safety Function (s): Reactor Water Level Low-Level 1,2 The proposed changes do not affect the r armhaity of the associated systems to perform their l Intended functions, nor do the proposed changes affect 'he operation of any equipment. Level 1 i lard =%s, i.e., Containment instrument Gas line outboard valves, Main Steam Line Drain line i valves, Reactor Enclosure Cooling Water (RECW) supply and drain line valves, Drywell Chilled l Water Supply and Retum line valves, Suppression Pool spray, Containment Instrument Gas _

Suction line valves, Core Spray test and flush line valves, RHR relief valve discharge line valves, l and Main Steam Line pressure insPmment line valves; and Level 2 isolations, i.e., Reactor Water  !

Cleanup suction valves, Drywell purge supply and exhaust line valves, Hydrogen Recombiner l inlet valves, Recirculation Loop sample line valves, Drywell H 2 /0 2 sample line valves, Tip purge )

and drive line valves, Main Steam Sample line valves, Drywell 2H 2 /0 sample retum and N2 Make-Up line valves, Drywell Radiation Monitoring supply and retum line valves, Suppression Pool Purge supply and exhaust line valves, Hydrogen recombiner exhaust line valves, HPCI test  ;

and flush line valves, H2 /02 sample retum line valves, Wetwell H2/02sample line valves, Drywell  :

Floor Drain Sump discharge line valve, Drywell Equipment Drain Tank Discharge line valves, j Suppression Pool Cleanup pump suction valves, and Suppression Pool level instrumentation line  ;

valves, have a 13 second response time requirement in accordance with TS Section 3/4.3.2,  !

the design basis evaluation for the reactor inventory release for these lines is based on the -  ;

assumption that any DC poworod valves have failed and that the plant has lost off-site power. In i this case, the AC powered isolation valve cannot close unti the on-site emergency diesel generator provides power to the valve. The TS response time is 13 seconds based on the delay .

for the emergency diesel generator which is longer than the five (5) second delay for the instrumentation. The emergency diesel generator is initiated upon loss of off-site power and is i independent of the instrumentation. The safety analysis considers an allowable inventory loss in each case which in tum determines the valve speed in conjunction with the 13 second delay. It follows that checking the valve speeds and the 13 second time for emergency power establishment wul establish the response time for the isolation functions. j

4. Affected Equipment: MSIVs, Reactor Low Pressure Sensors PT042 *N090A-H Safety Function (s): Main Steam Line Low Pressure i i

The proposed changes do not affect the capability of the associated systems to perform their intended functions, nor do the proposed changes affect the operation of any equipment As described in UFSAR Section 15.1.3, MSIV closure on low steam line pressure is provided to protect the reactor sys;am during normal power generation against transients that could cause uncontrolled depressurization. Protection is provided primarty for a pressure regulator malfunction which results in turbine control and/or bypass valve opening The Main Steam Line (MSL) low pressure trip set-point is specified to limit the duration and severity of the depressurization so that vessel thermal stresses, resulting from vessel cool down rate, remain below the appropriate safety limit and reactor water inventory loss is limited to prevent is.cs -;r.g the core. The set-point is chosen to be low enough that unnecessary isolation is avoided TS required response time for this trip is 1.0 seconds. If this trip initiation time is increased by 5 seco,Ws, there would be no significant impact on plant safety. The MSL low

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, , Page 6 of 13 pressure trip signal (i.e., setpoint 756 psig) is used primarNy to protect the reactor system in case of a pressure regulator malfunction event. The event is not a limiting event for the core thermal limits. The primary concem is the reactor water inventory loss and the thermal cyclic effect on the reactor vessel. During the event, the rapid depressurization causes an increase in reactor water level which results in the high water level trip. This in tum inklates a turbine trip and reactor scram After reactor scram, reactor water level can be maintained by the HPCI or l RCIC systems which are initiated at Level 2. The reactor vessel is designed to accommodate j more rapid depressurization than this event. Therefore, a five (5) second delay would reduce this pressure margin by approximately five (S) to ten (10) psig, but not affect vessel integrity or plant safety (IS allowable value is 736 poig).

5. Affected Equipment: Reactor Low Pressure Sensors PT-042 *N090A-H Safety Functions (s): Main Steam Line Low Pressure The proposed changes do not affect the capablity of the associated systems to perform their intended functions, nor do the proposed changes affect the operation of any equipment Reactor low pressure isolation for MSL drain valves and MSL pressure instrument line valves have a 13 second response time requirement. In accordance wkh TS Section 3/4.3.2, the <

design basis evaluation for the reactor water inventory release for these lines is based on the assumption that any DC powered v;lves have fared and that the plant has lost offske power. In this case, the AC powered valves cannot close untB the on-site emergency diesel generator provides power to the valve. The TS response time is 13 seconds based on the delay for the ,

emergency diesel generator which is longer than the five (5) seconds delay for the instrumentation. The emergency diesel generator is initiated upon loss of off-site power and is independent of the instrumentation. .The safety analysis considers an allowable inventory loss in each case which in tum determines the valve speed in conjunction with the 13 second delay. It follows that checking the valve speeds and the 13 second time for emergency power establishment wHI establish the response time for the isolation functions.

6. Affected Equipment: MSIVs, Main Steam Line Flow High Sensors PDT-041 *N006A-D,

. PDT-041 *N087A-D PDT-041 *N088A-D, PDT-041 *N089A-D Safety Function (s): Main Steam Line Flow High The proposed changes do not affect the capabRity of the associated systems to perform their intended functions, nor do the proposed changes affect the operation of any equipment As described in UFSAR Section 15.6.4, MSiv closure on high steam line flow is provided to protect the reactor system against accidents or transients that could cause unexpected increases in steam line flow. Protection is provided primarHy for a break in the steam line outside the primary containment Flow restrictors are provided to limit the maximum steam line flow to 140% of rated steam flow. The MSL high fkw trip setpoint is specified to limit the duration and severity of the high steam flow condition so that any off-site release wiu remain below the appropriate limit and inventory loss is limited to prevent uncovering the core. The setpoint is chosen to be high enough that unnecessary isolations are avoided. TS required response time for this trip is ,

0.5 seconds. If this trip initiation time is increased by five (5) seconds, there would be no {

significant impact on plant safety. The MSL high flow is designed primarBy to protect against a

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! MSL break outside containment. The high steam flow from the postulated double end break '

would result in releasing a large amount of steam and water outside the primary containment.

4 However, fuel falure would not result from this event as the break would be isolated long before i the reactor water level has any significant drop. Even with conservative MSIV closure times, the l i

offsite release for this event is only a small fraction of the aHowable 10CFR100 limits. A five (5) 1 second delay in the MSlV closure on high steam flow would stHI meet the requirments of 10CFR100. Therefore, a five (5) second delay would not affect plant safety.

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7. Affected Equipment: Main Steam Line Flow High Sensors PDT-041 *N086A-D, PDT 041 *N087A-D, PDT 041 *N088A-D, PDT-041 *N089A-D i

Safety Function (s): Main Steam Line Flow High The proposed changes do not affect the capabilty of the anaelated systems to perform their j intended functions, nor do the proposed changes affect the operation of any equipment.

Reactor low pressure isolation for Main Steam IJne Drain valves and Main Steam Line pressure instrument line valves have a 13 second response time requirement, in accordance with TS ,

Section 3/4.3.2, the design basis evaluation for the reactor water inventory release for these lines is based on the assumption that any DC powered valves have faBed and that the plant has  ;

lost off-ske power. In this case, the AC powered isolation valve cannot close unti the on-site j emergency diesel generator provides power to the valves. The TS response time is 13 seconds '

based on the delay for the emergency diesel generator which is longer than the five (5) second i delay for the instrumentation. The emergency diesel generator is initiated upon loss of off-site - l power and is independent of the instrumentation. The safety analysis considers an allowable  ;

reactor water inventory loss in each case which in tum determines the valve speeds in r conjunction wth the 13 second delay. It follows that checking the valve speed in conjunction with the 13 second time for emergency power establishment wNI establish the response time for the isolation functions.

8. Affected Equipment: RHR Shutdown Cooling Supply and Retum isolation Valves, and the associated Reactor Water Level Low-Level 3 Sensors LT-042 *N080A  ;

Safety Function: Reactor Water Level Low-Level 3 l f

The proposed changes do not affect the capabuity of the associated systems from performing their intended functions, nor do the changes affect the operation of any plant equipment The '

, instrumentation is provided to protect the reactor against an accident that could cause i uriavpar4 art loss of reactor water inventory caused primarby by a break or a leak in the process l lines outside the primary containment. In accordance with TS 3/4.3.2, the design evaluation for the reactor water inventory release for those lines is based on the assumption that any DC powered valves havs failed and that the plant has lost offsite power. In this case, the AC powered isolation viives cannot close untN the onsite emergency diesel generators (EDGs) provide electrical po,ver to the raives. The TS response time is 13 seconds based on the delay for the EDGs which is knoer than the five (5) second delay for the instrumentation. The EDGs

. are initiated upon loss of offsite power and is independent of the instrumentation. The safety ,

l analysis considers an aHowable reactor water inventory loss in each case which in tum i determines the valve speed in conjunction with the 13 second delay. It follows that checking the valve speeds and the 13 second time for emergency power establishment wHl establish the

response time for the isolation functions.

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9. Affected Equipment
Reactor Water Cleanup (RWCU) Suction Valves, the associated RWCU Delta Flow-High instrumentation sensors, trip units and relays

, Safety Function (s): RWCU Delta Flow-High The prapnaari changes do not affect the capabHity of the associated systems to perform their i intended functions, nor do the proposed changes affect the operation of any equipment The

Instrumentation is provided to protect the reactor against an accident that could cause
unexpected loss of reactor coolant inventory caused primarHy by a break or a leak in the

! process lines outside the primary containment in accordance with TS Section 3/4.3.2, the 4

design basis evaluation for the reactor water inventory release for these lines are based on the assumption that any DC powered valves have faued and that the plant has lost off-site power. In this case, the AC powered lantatian valves cannot close unti the on-site emergency diesel generator (EDG) provides electrical power to the valves. The TS response time is 13 seconds

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based on the delay for the EDG which is longer than the fNo (5) second delay for the  !

4 instrumentation. The EDG is initiated upon loss of off-site power and is independent of the l

instrumentation. The safety analysis considers an allowable reactor water inventory loss in each i

] case which in tum determines the valve speed in conjunction with the 13 second delay. It i follows that checking the valve apseds and the 13 second time for emergency power

establishment wIl establish the response time for the leolation function.

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10. Affected Equipment: HPCI Steam Supply, Pump Suction and Vacuum Relief Valves; the f

associated HPCI Steam Line Delta Pressure-High instrumentation sensors, trip units and relays; and the HPCI Steam Supply Pressure-Low l instrumentation sensors, trip units and relays in Table 3 i

Safety Function (s): HPCI Steam Line Delta Pressure-High, HPCI Steam Supply Pressure-Low l

J The proposed changes do not affect the capabilty of the associated systems to perform their

intended functions, nor do the proposed changes affect the operation of any equipment The instrumentation is provided to protect the reactor against an accident that could cause unexpected loss of reactor coolant inventory caused primarily by a break or a leak in the

, process lines outside the primary containment. In accordance with TS Section 3/4.3.2, the j design basis evaluation for the reactor water inventory release for these lines is based on the i assumption that any DC powered valves have failed and that the plant has lost off-site power. In

this case, the AC powered isolation valves cannot close untu the on-site emergency diesel i generator provides power to the valves. The TS response time is 13 seconds based on the j delay for the emergency diesel generator which is longer than the 5 second delay for the
instrumentation. The emergency diesel generator is initiated upon loss of off-site power and is
independent of the instrumentation. The safety analysis considers an allowable reactor water j inventory loss in each case which in tum determines the valve speed in conjunction with the 13 i second delay. It foHows that checking the valve speeds and the 13-second time for emergency

] power establishment wtl establish the response time for the isolation functions.

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i 11. Affected Equipment: RCIC Supply and Vacuum Relief Valves; the associated RCIC Steam Line l I

Delta Pressure-High instrumentation sensors, trip units and relays; l

). and the RCIC Steam Supply Pressure-Low instrumentation sensors, trip j units and relays f Safety Function (s): RCIC Steam Delta Pressure-High, RCIC Steam Supply Pressure-Low j The proposed changes do not affect the capability of the associated systems to perform their l Intended functions, nor do the proposed changes affect the operation of any equipment The instrumentation is provided to protect the reactor against an accident that could cause unexpected loss of reactor coolant inventory caused primarHy by a break or a leak in the

process lines outside the primary containment in accordance with TS Section 3/4.3.2, the

. design basis evaluation for the reactor water inventory release for these lines is based on the ,

j assumption that any DC powered valves have failed and that the plant has lost off-site power. In '

this case, the AC powered isolation vanes cannot close unti the on-site emergency diesel generator provides power to the valvas. The TS response time is 13 seconds based on the

delay for the emergency diesel generator which is longer than the five (5) second delay for the j instrumentation. The emergency diesel generator is initiated upon loss of off-site power and is ,

i independent of the instrumentation. The safety analysis considers an allowable reactor water i j inventory loss in each case which in turn determines the valve speed in conjunction with the 13

] second delay. It fouows that checking the valve spoods and the 13 second time for emergency i power establishment wBl establish the response time for the isolation functions. l

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Attachment 1

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Page 9 of 13

12. - Affected Equipment: Primary Containment, the namelated Drywell Pressure-High instrumentation sensors, trip units and relays

, Safety Function (s): Drywell Pressuro-High l

I The prnpanad changes do not affect the capsbuity of the associated systems to perform their  !

intended functions, nor do the proposed changes affect the operation of any equipment l Drywell high pressure isolation (i.e., Containment instrument Gas Supply header valves, RECW supply and retum valves from the Recirculation pumps, DryweN purge supply and exhaust i l valves, Hydrogen Recombiner inlet valves, DryweN 2 H /0 2 sample line valves, Tip purge and drive line volves, Containment instrument Gas suction line valves, DryweN Chuled Water supply  ;

and retum line valves, Drywou H,/02sample retum and N, Make-Up line valves, Drywell  !

Radiation Monitoring supply and retum line valves, Suppression Pool Purge supply and exhaust '

line valves, Hydrogen Recombiner exhaust line valves, HPCI test and flush line valves, H2/02  ;

sample retum line valves, instrument Gas to Vacuum Relief line valves, WetweN H,/02 sample  ;

line valves, HPCI Vacuum Relief line valves, DryweN Floor Drain Sump discharge line valves, j Drywell Equipment Drain Tank Discharge line valves, Suppression Pool Cleanup pump suction valves, Suppression Pool level instrumentation line valves, and RCIC Vacuum Relief line valves) ,

l Is provided to protect the reactor against an accident that could cause unexpected loss of i reactor coolant inventory caused primarNy by a break or a leak in the process lines outside the primary containment. In accordance with TS Section 3/4.3.2, the design basis evaluation for the reactor water inventory release for these lines is based on the assumption that any DC powered valves have faNed and that the plant has lost off-ske power. In this case, the AC powered isolation valves cannot close untu the on-ske emergency diesel generator provides power to the valves. The TS recponse time is 13 seconds based on the delay for the emergency diesel generator which is longer than the 5 second delay for the instrumentation. The emergency diesel generator is initiated upon loss of off-ske power and is independent of the instrumentation.

! The safety analysis considers an aNowable inventory loss in each case which in tum determines  ;

the valve speed in conjunction with the 13 second delay. It follows that checking the valve speeds and the 13 second time for emergency power establishment wiH establish the response l time for the isolation functions.

13. Affected Equipment: HPCI, LPCI mode of RHR, Core Spray; the associated ECCS Reactor Vessel Pressure-Low instrumentation sensors, trip units and relays, the associated ECCS Reactor Vessel Water Level-Low Level 1,2 instrumentation sent trip units and relays, and the associated ECCS Drywell Pressure-Higt trumentation sensors, trip units and relays >

l l Safety Function (s): Reactor Emergency Core Cooling Systems (ECCS)

The proposed changes do not affect the capabilky of the anaelated systems to perform their l Intended functions, nor do the proposed changes affect the operation of any equipment ECCS l Is provided to assure adaryda core cooling following loss of normal reactor cooling capabuity.

HPCI provides core cooling at high reactor pressure conditions. In case of a LOCA or when reactor pressure is sufficiently low, the Core Spray (CS) and LPCI systems provide core cooling.

In the event of a smaN leak in the primary coolant system in which HPCI cannet provide adaryda core cor. ting, the A.,$ would initiate to depressurize the reactor vessel to allow the low pressure ECCS systes to provide the necessary core cooling. The TS required response time for HPCI is 80 seconds, CS is 27 seconds, and LPCI is 40 seconds. The ECCS systems are required to mitigate LOCAs, The application of the GE SAFER /GESTR code for BWRs has demonstrated that there is significant safety margin for LOCA events. The realistic peak cladding temperature for the design basis LOCA is 1000'F which is significantly below the

., Attachmert 1

, , Page 10 of 13 2200*F Peak Cladding Temperature (PCT) limit. The delay in HPCI response time does not have any significant impact on the design basis because the system is not used as the primary cooling source due to the rapid reador depressurization. For isolation and smag breaks, a 5 second delay in the system response has minimal impact since the release of the reactor coolant inventory from the break is significantly reduced. The design basis LOCA analysis in EPRI, NSAC-131, " Basis for Relaxing ECCS Performance Requirements for BWR/4s," has demonstrated that an 11 second increase in the response time for the Core Spray system would increase the PCT by approximately 84'F. A 15 second increase in the LPCI response time would increase the PCT by 131'F. The combined effect of a ten (10) second delay for Core Spray and nine (9) second delay for LPCI is an increase in the PCT by 137*F, stil consirterably below the PCT limits.

Additional information PECO Energy has confirmed the generic applicablity of NEDO-32291 to LGS, Units 1 and 2. As indicated in Appendix A of NEDO-32291, PECO Energy was a participating utRity in this evaluation.

j PECO Energy has also confirmed that the components discussed within the scope of this TS Change i Request have been evaluated in NEDO-32291. The components included in the scope of NEDO-32291 are described in Appendix G of the report, and in Table 1 of the NRC's SER. The LGS, Units 1 and 2, components specifically reviewed for this TS Change Request were Rosemount transmitter Models 1151 and 1153, Rosemount trip unit Models 510DU and 710DU, Amerace (Agastat) EGP and ETR relays, Bailey Model 745 transmitters, Bailey Model 750 square root extractors, Bauey Model 752 summers, and Eagle HPS timers all of which are bounded by the fature modes and effects analysis performed for the study as documerted in NEDO-32291. A review of the bases for excluding components from response time testing was performed, and all applied with one (1) exception. The bases for eliminating response time testing for RPS High Steam Dome Pressure could not be applied, since the Average Power Range Monitor-Rod Block Monitor (APRM-RBM) TS implementation ututzes the high steam dome pressure as a possible primary reactor scram signal at low reactor power. In all other cases the analyses provided in NEDO-32291 are applicable to LGS, Unks 1 and 2. The BWROG evaluation provided in NEDO-32291 -

confirms that the selected response time tests are of no safety significance and cause unnecessary personnel avi-o, and can reduce avaRability of safety systems and are a significant burden to utHRy

! resources.

PECO Energy confirms that LGS, Units 1 and 2, wil conform with the following recommendations from EPRI NP-7243, ' investigation of Response Time Testing Requirements

  • i
1) Prior to installation of a new transmitter / switch or following refurbishment of a transmitter / switch (e.g., sensor cell or variable demping components), a hydraulic response time test wNl be performed to determine an initial sensor-specific response time value. If this TS Change Request is approved, the applicable LGS, Units 1 and 2, procedures wHI be revised, as appropriate, to incorporate this recommendation in conjunction with implementing the proposed TS changes.
2) For transmitters and switches that use capRiary tubes, capulary tube testing shall be performed after initial installation and after any maintenance or modification activky that could damage the capillary tubes. For those transmitters and switches within the scope of this proposed TS change that utlize captiary tubes, capillary tube testing will be performed after installation and after any maintenance or modification activity, as appropriate.

Applicable station calibration procedures wil be revised, as appropriate, to include guidance to input a fast ramp or step change to system components during calibration. This new guidance wHl ensure that the response of the transmitter (s) to an input signal (i.e., fast ramp or step input change) is prompt, and

, in all cases occurs within less than five (5) secoex$s. The applicable procedures will be revised in 4

conjunction with implementing the proposed TS changes, if approved.

2 l

Attachment 1

.;* , , Page 11 of 13 l

l PECO Energy conducted training for operators and technicians in response to the Requested Actions identified in NRC8 9041, *l.oss of Ful45 in Transmitters Manufactured by Rosemount," as documented l in our luter dated July 13,1990, responding to NRC3 90-01. PECO Energy also provided additional

! Information regarding the loss of ful-ol in Rosemount transmitters in our response to NRCB 90-01, I Supplement 1, by letter dated March 5,1993. However, the applicable station calibration procedures wRl

, be revised, as appropriate, to assure that technicians monitor for response degradation during the i

performance of calibrations and functional tests. Any necessary procedure revisions wlN be completed in conjunction wkh implementing the proposed TS changes.

{

i Survetlance testing procedures wul be revised, as appropriate, to ensure that calibrations and functional i tests are being performed in a manner that allows simultaneous monitoring of both the input and output <

response of components being tested. As indicated above, the applicable calibration procedures wlN be l revised, as necessary, to ensure that the response of a transmitter (s) to input signals (i.e., step change I or fast ramp) occurs within less than five (5) seconds. The applicable surveRiance testing procedures wRl be revised in conjunction with implementing the proposed TS changes, if approved.

4 I

PECO Energy's compliance wkh the guidance stipulated in NRCB 90-01, Supplement 1, was reviewed by j the NRC as documented in a letter dated November 19,1993. The NRC's evaluation of our response to
NRCS 90-01, Supplement 1, concluded that PECO Energy's actions satisfied the Requested Actions i specified in Supplement 1.

i i As indicated above, the components affected by this proposed TS change are limited to Rosemount j transmitter Models 1151 and 1153, Rosemount trip unit Models 510DU and 710DU, Amerace (Agastat) )

EGP and ETR relays, Baley Model 745 transmitters, Baley Model 750 square root extractors, Baley  ;

i Model 752 summers, and Eagle HPS timers. PECO Energy reviewed the vendor recommendations for ,

these devices and confirmed that they do not contain recommendations for periodic response time testing.

i  !

Although not explicitly evaluated, the proposed TS changes wul provide an improvement to plant safety j
and operation by reducing the time safety systems are unavailable, reducing the potential for safety i i system actuations, reducing plant operating and shutdown risk, limiting radiation exposure to plant  !

i personnel, and eliminating the diversion of key personnel to conduct unnecessary testing. Therefore, j PECO Energy considers that the proposed TS changes wlN result in an overall increase in the margin of safety and that the changes do not constkute an unreviewed safety question.

Furthermore, a simlar TS change request requesting elimination of response time testing for selected 4

instrumentation in accordance with NEDO-32291 was submitted for Qinton Power Station by letter dated l January 27,1995. The NRC subsequently approved Clinton's TS change request as documented in a j

! letter dated March 9,1995. '

i information Supportina a Findina of No Sionificant Hazards Consideration

We have concluded that the proposed changes to Limerick Generating Station (LGS), Units 1 and 2 I Technical Specifications (TS) to eliminate selected response time testing requirements in accordance i with the supporting analysis provided in NEDO-32291 do 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

. 1. The oronosed Technical Soecifications (TS) chances do not involve a sioni6 cant incraana in the

probablity or consequences of an accident previously evaluated, i

j The prafmad TS changes do not make any physical alterations or modifications to the plant systems or equipment. The proposed changes do not affect the capabHity of the associated i systems to perform their intended functions within their required response times, nor do the proposed changes adversely impact the operation of any plant equipment. The affected plant

, systems wIl continue to function as designed Elimination of the response time testing

].

.. Attachment 1

, , Page 12 of 13 requirements es proposed by this TS change for selected components in RPS Instrumentation, Isolation Actuation System Instrumentation, and ECCS Actuation Instrumentation wIl not adversely affect the operation of these cowpor,erig.

The supporting analysis provided in NEDO42291, demorg.ates that response time testing is redundant to other TS required testing NEDO 32291 demonstrated that these other required tests (i e., channel checks, channel calibrations, channel functional tests, and logic system functional tests), in conjunction with actions taken in response to NRC Buustin 9041 and NRCB 9041, Supplement 1, are sufficient to identify fature modes or degradation in instrument

! response times, and ensure operation of the annar4= tad systems within acceptable limits. There are no known faNure modes that can be detected by response time testing that cannot also be detected by other TS required testing. The continued application of other existing TS required testing such as channel checks, channel calibrations, channel functional tests, and logic system

, functional tests, ensures that the response times for these systems wbl be maintained within the .

acceptance limits. The r apahalty of these systems to perform their intended functions within their required response times is not adversely impacted by this proposed TS change. NEDO-l 32291 evaluated the potential fature modes of the affected instrumentation loops which could i impact the instrument loop response times. Industry operating experience was also reviewed to

! identify falures that affect response times and how they are detected. The falure modes

! identitled were evaluated to determine if other TS required survetlances and actions taken in

! response to NRC Bulletin 90-01, and NRCB 90-01, Supplement 1, would detect any effects on

! response time. There are no falures modes identified that can be detected by response time j testing that cannot also be detected by other TS required testing.

PECO Energy has confirmed the applicabuity of the generic evaluation provided in NEDO-32291

to LGS, Units 1 and 2. By letter dated December 28,1994, the NRC concluded that response .

! time testing can be eliminated from the TS for the selected instrumentation identified in NEDO. l t 32291, with certain provisions, and that NEDO-32291 can be referenced in license amendment l i requests.  ;

i Therefore, the proposed TS changes do not involve an increase in the probabuity or j consequences of an accident previously evaluated.

j 2. The orooosed TS chanoes do not create the nossibuity of a new or different kind of accident

from any accident previously evaluated.

! The proposed TS changes do not involve any physical changes to plant systems or equipment.

! The proposed changes apply only to the testing requirements for the selected components l

! involved and do not result in any physical modifications to these components, or to other plant i system components. Elineation of the response time testing requirements as proposed by this

! TS change for selected coreponents in RPS Instrumentation, Isolation Actuation System j instrumentation, and ECCS Actuation Instrumentation wNi not adversely affect the operation of i these components. These components wil continue to function as designed Consequently, no

new falure modes are introduced as a result of the proposed TS changes.

e

Eliminating the response time testing requirements as proposed, does not create a new or j different type of accident than any previously evaluated.L No new or different type of accident
wul be created as a result of this proposed TS change
NEDO-32291 demonstrates that other required tests (i.e., channel checks, channel calibrations,

} channel functional tests, and logic system functional tests), in conjunction with actions taken in j response to NRC Bulletin 9041 and NRCS 90-01, Supplement 1, are sufficient to identify faRure j modes or degradation in instrument response times, and ensure operation of the associated i

4

1 i *

, , Attachment 1

, , Page 13 of 13 4

systems within acceptable limits. There are no known falure modes that can be detected by response time testing that cannot also be detected by other TS required testing, and therefore, response time testing for the selected coinponents is redundant to the other TS required testing.

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

3. The oronosed TS chanoma do not involve a sionificant reduction in a maroin of *_a_8-*v.

i The prnpanad TS changes do not involve any physical changes to plant systems or equipment The prapaamd TS changes do not affect the capsbuity of the associated systems or equipment l from performing their intended functions. The systems involved wil continue to respond within j their allowed response times. Elimination of the response time testing requirements are based i on the evaluation provided in NEDO42291 which demonstrates that response time degradation

can be detected by other TS required testing. The evaluation concluded that other TS required l tests (i.e., channel checks, channel calibrations, channel functional tests, and logic system
functional tests), in conjunction wkh actions taken in response to NRC Bulletin 90-01 and NRCB 90-01, Supplement 1, are sufficient to identify falure modes or degradation in instrument
response times, and ensure operation of the associated systems within acceptable limits.

In addition, although not speellically evaluated, the proposed TS changes wul provide an l improvement to plant safety and operation by reducing the time safety systems are unava8able,

! reducing the potential for safety system actuations, reducing plant operating and shutdown risk,

! limiting radiation exposure to plant personnel, and eliminating the diversion of key personnel to t

conduct unnecessary testing Therefore, PECO Energy considers that the proposed TS changes

j. wBl result in an overall increase in the margin of safety and that the changes do not constitute
an unreviewed safety question.

I j Therefore, the proposed TS changes do not involve a significant reduction in a margin of safety, i

r information Supporting an Environmental Assessment An Erwironmental Assessment is not required for the changes proposed by this TS Change Request because the requested changes to the LGS, Units 1 and 2, TS conform to the crkeria for ' actions eligible

for categorical exclusion," as specified in 10 CFR 51.22(c)(9). The requested changes will have no l Impact on the environment. The proposed changes do not involve a significant hazards cor sideration i as discussed in the preceding section. The proposed changes do not involve a significant change in the j j types or significant increase in the amounts of any effluent that may be released offsite. In addition, the
proposed changes do not involve a significant increase in individual or cumulative occupational radiation
exposure.

2 Conclusion I The Plant Operations Review Committee and the Nuclear Review Board have reviewed the proposed changes to the LGS, Units 1 and 2. TS and have concluded that they do not involve an unreviewed j- safety question, and wIl not endanger the health and safety of the public.

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