ML20092J394

TS Change Request 95-01-0 to Licenses NPF-39 & NPF-85, Revising TS Table 4.3.1.1-1, RPS Instrumentation SRs to Reflect Change in Calibr Frequency for LPRM Signal from Every 1,000 EFPH to Every 2,000 Megawatt Days Per Std Ton

ML20092J394
Person / Time
Site:	Limerick
Issue date:	09/18/1995
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20092J395	List: ML20092J394 ML20092J396
References
NUDOCS 9509220108
Download: ML20092J394 (9)

Text

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10CFR50.90

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. v PECO ENERGY- faf==:"L.,.

965 Chesterbrook Boulevard Wayne.PA 19087-5691 September 18,1995 -

Docket Nos. 50452 50-353 License Nos. NPF-39 NPF-85 U.S. Nuclear Regulatory Commission Attn: Document Coritrol Desk Washington, DC- 20555

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Subject:

Limerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 95-01-0 Gentlemen:

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

L- 9541-0, in accordance with 10CFR 50.90, requesting an amendmera to the TS (Appendix A) of Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and j ' 2, respectively. This proposed change will revise TS Table 4.3.1.1-1, " Reactor Protection System instrumentation Survelliance Requirements" to reflect the change in the calibration frequency for

.. the Local Power Range Monitor (LPRM) signal from every 1000 Effective Full Power Hours (EFPH) to every 2000 Megawatt Days per Standard Ton (MWD /ST). Information supporting this Change Request is contained in Attachment I to this letter, and the marked up pages showing l

[ the proposed change to the LGS Units 1 and 2 TS are contained in Attachment 2. This

. Information is being submitted under affirmation, and the required affidavit is enclosed.

I We request that, if approved, the amendment to the LGS Unit 1 TS be issued by April 1,1996,

. and become effective within 30 days of issuance. For LGS Unit 2 TS, the amendment should be issued after 3D MONICORE is implemented which is currently scheduled for May 1996.

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

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Very truly yours,

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G.- A. Hunger, Jr.

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, Director - Licensing Enclosure, Attachments

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

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

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QOMMONWEALTH OF PENNSYLVANIA  :

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COUNTY OF'CHESTER,  ; )

.r W. H. Smith, Ill, being first duly swom, deposes and says: {

t That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the foregoing Application for Amendment of Faculty Operating License Nos. NPF-39 and NPF-85 (Technical l E  : Specifications Change Request No. 9541-0), to chan0e the calibration frequency for the Local Power Range Monitor (LPRM) signal from every 1000 Effective Full Power Hours (EFPH) to every 2000

. 3 Megawatt Days per Standard Ton (MWD /ST), at Limerick Generating Station, Units 1 and 2, and knows

.,. 1 the contents thereof; and that the statements and matters set forth therein are true and correct to the '

i i best of his knowled0s, information and belief.

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jll~ f, W5 Vice President i Subscribed and swom to

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Notary Public j i

Notanal Seal i

May Lou Skrocki. Notary Puhhc Tredyttrin Twp., Chester County -l My Commission Expires May 17. f 999 i

. bcw(Penreytvarna Assocanon of Notake

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ATTACHMENT 1 UMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 50-353 LICENSE NOS. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 95-01-0

' CHANGE IN THE CAUBRATION FREQUENCY FOR THE LOCAL POWER RANGE MONITOR (LPRM) SIGNAL FROM EVERY 1000 EFPH TO EVERY 2000 MWD /ST*

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Supporting Information for Changes - 6 Pages i

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/ Docket Nos. 50-352  !

'.' 50-353 License Nos. NPF-39  ;

NPF-85  ;

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PECO Energy Company, licensee under Facuity Operating License Nos. NPF-3g and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating License be amended as proposed herein, to reflect the change in the calibration frequency for the Local Power Range Monitor (LPRM) signal from every 1000 Effective FuH Power Hours (EFPH) to every 2000 Megawatt Days per Standard Ton (MWD /ST), thereby revising TS Table 4.3.1.1 1 and TS Bases 3/4.3.1. The proposed change to the TS is indicated by mark-ups on TS page 3/4 3-8 and TS Bases page B 3/4 3-1. The TS page and TS Bases page showing the proposed change are contained in Attachment 2.

We request that, if approved, the LGS Unit 1 TS change proposed herein be issued by AprH 1,1996, and ,

become effective within 30 days of issuance of the amendment. For LGS Unit 2 TS, the amendment i should be issued after 3D-MONICORE is implemented which is currently scheduled for May,1996. ,

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

D!amaahn and Descriotion of the Proposed Chance t

The proposed change to the Limerick Generating Station (LGS) Units 1 and 2 Technical Specifications (TS) Table 4.3.1.1 1, " Reactor Protection System Instrumentation Survell!ance Requirements," and TS Bases 3/4.3.1 win reflect the change in the calibration frequency for the Local Power Range Monitor (LPRM) signal from every 1000 Effective Fun Power Hours (EFPH) to every 2000 Megawatt Days per Standard Ton (MWD /ST). This proposed TS change is currently applicable to LGS Unit 1 and win be valid for LGS Unit 2 when 3D-MONICORE is implemented as the core monitoring program.

BWR power operation relies upon readings from fixed in-core neutron detectors (i.e., Local Power Range Monitors (LPRMs)). LPRMs are small fission chambers with an approximately linear response to the local neutron flux, and thus local thermal power. The calibration of the LPRMs every 1000 EFPH employs a second set of moveable detectors (i.e., Traversing in-Core Probe (TIP) system.) The required LPRM calibration relates the power distribution, measured by the TIP system, to the LPRM readings I existing at the time. When the LPRMs are norma:lzed to one another, to the TIP readings, and to a plant  ;

heat balance calculation, the LPRMs anow determination of the local power in the core for each node (approximately 6 inches) of the fuel.

Outputs from the calibrated LPRMs are used in the Reactor Protection System (i.e., Average Power Range Monitor (APRM)), and the Rod Block Monitor (RBM), as well as daHy surveulance of the Power Distribution Umits (reactor thermal limits monitoring). Accuracy requirements on the power distribution are defined by GESTAR-il (NEDE-24011-P-A-10, Section 4.3.1.1.1) and GE Fuel Bundle Designs (NEDE-31152P), which are part of the LGS licensing basis. In particular, Table 3-3 of NEDE-31152P requires calculated nodal powers to have a root mean square (rms) uncertainty of no more than 8.7% for reload cores. This uncertainty includes uncertainties contributed by the LPRM system.

Advances in process computer monitoring include the development of new mathematical techniques and algorithms combining reactor physics theory with on-line core data, (e.g., LPRM readings). One such methodology is 3D-MONICORE which is currently in use at LGS Unit 1. The 3D-MONICORE employs an adaptive loaming algorithm using on-line as weX as historical core data inputs to improve power i

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Docket Nos. 50-352 l- 50-353 License Nos. NPF-39 NPF-85 calculations within the reactor physics model by effectively modifying the neutron leakage terms (adaptive coefficients) to force the calculated power distribution to match the measured power distribution as determined by the TIP system. Corrections made within the monitoring process account '

for decay of LPRM sensitivity due to depletion of the fissile coating within each LPRM. The 1000 EFPH calibration interval at LGS was based upon older monitoring methodology and older LPRM designs in l use at the time.

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' Part of the study discussed below investigated the sensitivity decay characteristics of the present ,

generation of LPRM detectors Calibration current and exposure data demonstrate an increase in the j predictability of LPRM sensitivity as a function of exposure. This, combined with improved monitoring methods, allows lengthening of the LPRM calibration interval as discussed below. ,

3D-MONICORE reactor physics methodology was used off-line to model four continuous months of rated operation at Plant Hatch during a recent cycle to study the effect of lengthening the LPRM

, calibration interval. This period included a control rod pattern sequence exchange. Actual plant LPRM readings were modified using TIP set (OD1) calibration currents to factor out the effects of the LPRM calibrations. The operational periods of interest were then re-depleted using the modified LPRM readings (i.e., without calibrations). Comparisons of 2D bundle and 3D nodal power distributions (with .

calibrations versus without) were made and percent rms deviations were calculated for each exposure '

point of interest. Results of this analysis show that the licensing basis nodal power uncertainty of 8.7%

was satisfied for up to approximately 3000 EFPH between LPRM calibrations. Therefore, a change in the frequency for LPRM calibrations from 1000 EFPH to 2000 EFPH is acceptable.

The following calculations were used to convert 2000 EFPH to 2000 MWD /ST. These units are, for this application, effectively interchangeable based on the combination of two facts:

1) Straiaht Conversion: Using current core weights and rated powers, converting 2000 EFPH to units of MWD /ST results in:

81 6' rs/ Day

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24 ,145.99 ST and,

,24 rs/ Day ,145.08 ST Using projected core weights and re-rated power (as LGS Unit 2 is currently re-rated and LGS Unit 1 will be re-rated in Cycle 7), a range of future MWD /ST values can be obtained. The projected core weights range from an equilibrium full core of GE-13 at 144.4 ST; to a core of GE-12 with Shoreham bundles included, which would be 152.1 ST. Therefore, the 2000 EFPH equivalences in MWD /ST are:

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Docket Nos. 50-352  ;

50-353 License Nos. NPF-39 NPF-85 .

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GE-13 Equilibrium Full Cores 2000 FP y 3 50 pp = 1995. 6 NWD/ST and, [

GE-12 with Shoreham Bundles: 2000 F g 3 1894,6 MWD /ST '

p 5 Thus, for the 2000 EFPH equivalence, the expected range is between 1879.7 MWD /ST and 1995.6 MWD /ST. Values become larger as cores become lighter and re-rate power is Implemented. This entire range is equivalent to only approximately 116 EFPH, or less than 6%

of the original 2000 EFPH.

2) Nodal Power Uncertaintv Sensitivity to EFPH: The Hatch Study indicates that the nodal power uncertainty is relatively insensitive to the number of EFPH since the last calibration. In l fact, between approximately 2000 and 3000 EFPH, the nodal power uncertainty increases by less than 0.1% which is within the 8.7% rms uncertainty limit specified in NEDE-31152P.

Thus, while it is recognized that the actual MWD /ST value corresponding to 2000 EFPH is a function of core weight and rated core power (and is therefore cycle dependent), the selection of an even 2000 MWD /ST interval for LPRM signal calibration is reasonable and results in a nodal power uncertainty which is less than the 8.7% uncertainty limit. The proposed change in units from EFPH to MWD /ST ls consistent with NUREG 1433, Standard Technical Specifications, General Electric Plants, BWR/4,"

Revision 1, dated April 1995. ,

Therefore, we propose that TS Table 4.3.1.1 1 and TS Bases 3/4.3.1 be revised to reflect the change in >

the calibration frequency for the LPRM signal, from every 1000 EFPH to every 2000 MWD /ST.

Safety Assessment The proposed change in the calibration frequency for the Local Power Range Monitor (LPRM) signal from every 1000 EFPH to 2000 MWD /ST does not involve a physical change in the configuration,

. setpoints, or operation of any safety-related instrumentation. The proposed TS change does not make any physical change to the fuel or the manner in which the fuel responds to a transient or accident.

The LPRMs are utilized as input to the Average Power Range Monitor (APRM) and Rod Block Monitor (RBM) systems. The primary safety function of the APRM system is to initiate a scram during core-wide neutron flux transients before the actual core-wide neutron flux level exceeds the safety analysis design basis. This prevents fuel damage from single operator errors or equipment malfunctions. The APRMs are calibrated at least once per week to the plant heat balance, utilize a radially and axially diverse group 2

of LPRMs as input and are utlilzed to detect chcnges in average, not local, power changes. Therefore, the effects of changing the LPRM calibration frequency to 2000 MWD /ST on the APRM system responses will be minimal due to any individual LPRM drift being practically canceled out (due to diversity of input) and/or due to the frequent recalibration of the APRMs to an independent power calculation (the heat balance). Thus, changing the LPRM calibration frequency as proposed will not impact the capability of the APRM system to perform the scram function, and there is no impact on transient delta-CPRs.

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. Docket Nos. 50-352

- 50-353 Ucense Nos. NPF-39 NPF 85 - .

The RBM system is utlized in the mitigation of a Rod Withdrawal Error (RWE) event. The RBM system is -

designed to prevent the operator from increasing the local power significantly when withdrawing a control rod. Under Average Power Range Monitor . Rod Block Monitor Technical Specifications / Maximum Extended Load Une Limit Analysis (ARTS /MELLLA) on each sewetion of a control rod, the average of the assigned, unbypassed LPRMs is adjusted to equal a 100% reference signal for each of the two RBM channels. Each RBM channel automaticaHy limits the local thermal margin changes by limiting the allowable change in local average neutron flux to the RBM setpoint. - If the local average neutron flux change is greater than that allowed by the RBM setpoint, within either  :

f RBM channel, the rod withdrawal permissive is removed preventing further rod movement Since the

' change in local neutron flux is calculated from the change in the average of the LPRM readings, and calibrated, on every rod selection to the reference signal, offsets in individual LPRM readings due to calibration differences are effectively eliminated for a given RBM setpoint. Therefore, the constraints on f

' the withdrawal of any given rod are unchanged, and there wul not be any increase in RWE delta-CPR.  !

The GE Thermal Analysis Basis (GETAB) determination of the Minimum Critical Power Ratio (MCPR)

Safety Limit allows a maximum total nodal uncertainty of the Traversing in-Core Probe (flP) readings (of which the LPRM updated uncertainty is a part) of 8.7%. The change in LPRM calibration frequeacy results in an LPRM Update uncertainty which, when combined with the other uncertainties which .

comprise the total TIP readings uncertainty, yields a total TIP readings nodal power uncertainty of less ,

than the allowed 8.7%. Thus the proposed change in LPRM calibration frequericy wHI not affect the existing MCPR Safety Limit.

The proposed TS change does not affect existing accident analyses or design assumptions, nor does it Impact any safety limits of the plant.

Information Sunoortino a Findino of No Slanificant Hazards Consideration We have concluded that the proposed change to the Limerick Generation Station (LGS), Units 1 and 2 Technical Specifications (TS), which will revise TS Table 4.3.1.1 1, " Reactor Protection System Instrumentation Surveillance Requirements" to reflect the change in the calibration frequency for the Local Power Range Monitor (LPRM) signal, from every 1000 Effective Full Power Hours (EFPH) to every l 2000 Megawatt Days per Standard Ton (MWD /ST), does not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.

1. The oroonaad Technical Snecifications frS) chance does not involve a sionificant increase in the >

orchahHitV of Consecuences of an accident DreviouslV GValuated.

The change in the calibration frequency of the Local Power Range Monitor (LPRM) signal does i not make any physical change to the fuel or the manner in which the fuel responds to a transient or accident. The proposed TS change does not affect the fundamental method by which the LPRMs are calibrated. Also, the LPRM calibration frequency is not considered an initiator of any events analyzed in the SAR. Therefore, calibrating the LPRMs on a different frequency wHl not increase the probablity of occurrence of an accident previously evaluated in ,

the SAR.

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/ Docket Nos. 50-352 I 50-353 License Nos. NPF-39 NPF-85 The resulting nodal power uncertainty does not exceed the nodal power uncertainty accounted for in the existing Minimum Critical Power Ratio (MCPR) Safety Umit; thus, the MCPR Safety  ;

Umit is not affected by this TS Change, and, therefore, the initial conditions of any accident are  ;

unchanged. Since the calibration frequency change wNI not affect the course of any evaluated j

. accident, the consequences of an accident previously evaluated in the SAR wHI not be e increased. ,

i Therefore, the proposed TS change does not involve an increase in the probabNity or  !

consequences of an accident previously evaluated. .l

- 2.' The oranaamd TS chanos dame not create the nossibHity of a new or different kind of accident  !

from any accident oreviousiv evaluated.  !

The change in the calibration frequency of the Local Power Range Monitor (LPRM) signal does  ;

not make any physical change to the plant or the manner in which the equipment responds to a transient or accident. The proposed TS change does not introduce a new mode of plant operation and does not involve the installation of any new equipment or instrumentation. The fuel wNI continue to be operated to the same safety limits since the Minimum Critical Power -l Ratio (MCPR) Safety Umit remains unchanged due to this TS change.  ;

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

3. The oronosed TS chanos does not involve a sionificant reduction in a maroin of safety. l The following TS Bases were reviewed for potential reduction in the margin of safety:

i 2.0 Safety Limits and Umiting Safety System Settings; i 3/4.1 Reactivity Control Systems; 3/4.2.1 Average Planar Linear Heat Generation Rate; .

I 3/4.2.3 Minimum Critical Power Ratio:

3/4.2.4 Linear Heat Generation Rate; l 3/4.3.1 Reactor Protection System Instrumentation; 3/4.3.6 Control Rod Block Instrumentation; 3/4.3.7.7 Traversing in-Core Probe System;  ;

The GE Thermal Analysis Basis (GETAB) determination of the Minimum Critical Power Ratio (MCPR)  !

Safety Umit allows a maximum total nodal uncertainty of the Traversing in-Core Probe (TIP) readings of i which the Local Power Range Monitor (LPRM) Update uncertainty is a part. The change in LPRM calibration frequency results in an LPRM Update uncertainty which, when combined with the other uncertainties which comprise the total TIP readings uncertainty, yielo. a total TIP readings nodal power i uncertainty of less than the allowed GETAB uncertainty. Thus the change in LPRM calibration frequency [

will not affect the MCPR Safety Umit. -

The LPRMs are utilized as input to the Average Power Range Monitor (APRM) and Rod Block Monitor l (RBM) systems. The primary safety function of the APRM system is to initiate a scram during core-wide i neutron flux transients before the actual core-wide neutron flux level exceeds the safety analysis design .j i

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., Docket Nos. 50-352

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50-353 Ucense Nos. NPF-39 NPF-85 basis.- This prevents fuel damage from single operator errors or equipment malfunctions. The APRMs are calibrated at least once per week to the plant heat balance, utsize a radially and axiaHy diverse group of LPRMs as input and are utuized to detect changes in average, not local, power changes. Therefore, the effects of changing the LPRM calibration frequency on the APRM system responses wNi be minimal due to any indMdual LPRM drift being practically canceled out (due to diversity of input) and/or due to the frequent recalibration of the APRMs to an independent power calculation (the heat balance). Thus, _

changing the LPRM calibration frequency wBl not impact the capabulty of the APRM system to perform

the scram function, and there is no impact on transient delta-CPRs.

l The RBM system is ututzed in the nitigation of a Rod Withdrawal Error (RWE) event. The RBM system is l

' designed to prevent the operator trom increasing the local power significantly when withdrawing a control rod. Under Average Power Range Monitor Rod Block kicitor Technical 3

Specifications / Maximum Extended Load Une Umit Analysis (ARTS /l&LLLA) on each selection of a

! control rod, the average of the assigned, unbypassed LPRMs is adjusted to equal a 100% reference j signal for each of the two RBM channels. Each RBM channel automatica3y limits the local thermal 3- margin changes by limiting the allowable change in local average neutron fkx to the RBM setpoint. if l the local average neutron flux change is greater than that aHowed by the RBM setpoint, within either RBM channel, the rod withdrawal permissive is removed preventing further rod movement. Since the change in local neutron flux is calculated from the change in the average of the LPRM readings, and l calibrated on every rod selection to the reference signal, offsets in individual LPRM readings due to

. calibration differences are effectively eliminated for a given RBM setpoint. Therefore, the constraints on

! the withdrawal of any given rod are unchanged, and there will not be any increase in RWE delta-CPR.

Since the MCPR Safety Umit is unaffected and the delta-CPR values are unchanged, the cycle CPR Operating limits are unchanged due to this TS change. Therefore, the proposed change in the frequency of LPRM signal calibrction does not result in a reduction in a margin of safety.

Informelan Sunrwtina an Envlroninentsi Assessment An environmental assessment is not required for the change proposed by this TS Change Request because the requested change to the Umerick Generating Station (LGS), Units 1 and 2, TS conforms to the criteria for " actions eligible for categorical exclusion" as specified in 10 CFR51.22(c)(9). The requested change will have no impact on the environment. The proposed change does not involve a significant hazards consideration as discussed in the preceding section. The proposed change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed change does not involve a significant increase in individual or cumulative occupational radiation exposure.

Conclusion The Plant Operations Review Committee and the Nuclear Review Board have reviewed this proposed change to the Umerick Generating Station (LGS), Units 1 and 2. TS and have concluded that it does not involve an unreviewed safety question, and wRl not endanger the health and safety of the public.

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