ML20065T108

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Proposed Tech Specs,Accounting for Use of New Fuel Type for Cycle 9 Operation
ML20065T108
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 12/17/1990
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20065T104 List:
References
NUDOCS 9012270158
Download: ML20065T108 (15)
v • d • e


Text

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

., ATTACHMENT 2

. c.

PEACil BOTTOM ATOMIC POWER STATION UNITS 2 and 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 REVISED TECHNICAL SPECIPICATION PAGES List of Attached Pages Unit 2 Unit 3 9 9 17- 17

_24 24 140a 140a

.140b 140b 140c 140c

-256a 256a l

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~ _ . , _ . . . _ _, _ _ _ _ . . _ _ . _ . . . __ _ -._--~ -. . _- _ . . _ . . _ _ .

7

_ Unit 3

. m .

gj PBAPSI

_-SAFETY ' LIMIT : LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY '

{ . Applicability: Applicability:

cThe SafetyLLimits. established -

The Limiting _ Safety System Settings ito preserve ths~ fuel-cladding apply to trip settings of the integrity? apply _to those instruments and devices which-are

variables which monitor the provided to prevent the fuel fuel thermal-behavior, cladding; integrity Safety-Limits from being exceeded.

. Objectives:- Objectives:

The objective-of-the Safety The objective of the Limiting Safety LLimits is to establish' limits- System Settings is to define the '

t 1which' assure the !ntegrity of level of the process variables ~at lthe-fuel cladding. which automatic protective-action is initiated to prevent the fuel' cladding-integrity Safety Limits from being exceeded.

-Specifications. Specification:

1A. Reactor Pressure 2 800 psia -The limiting safety system settings-Land Core Flow 2 10% of Hated shall beLas specified below:-

A. Neutron Flux Scram

  • TheLexistence ofalminimum 1. APRM: Flux' Scram Trip Setting-critical
  • powe ri ratio (MCPR): (Run Mode) less than 1.06Tfor'two recirculation losp; operation,_ _W hen the: Mode Switch is-in the f jo'r 1.070forcsingle loop . RUN. position, the APRM' flux
operationh shall constitute scram-trip setting shall be:

violationtof the.fuelJcladding integrityJ safetyL11mit. : S.< 0.58W + 62%.- 0.58 AW L Tofensureethat.this safety- where:

llimitilsinot" exceeded,-neutron ofluxishallinotrbe above'the S = Setting in percent.of rated Tsc' ram? setting established-in ' thermal _ power ~(3293'MWt)-.

-i

specification ~2.1.A forLlonger -

[r thani1.15Tseconds as' indicated- W ' Loop _recirculati'ng by the proceas. computer. When flow rate 11n percent.

the proces~s' computer--is.out of- of-design.'W is'100-for-

'servict this cafety limit shal.1 core 1 flow of 102.5 1belassumedLto1be exceeded if million'1b/hr or greater.

the1 neutron flux-exceeds its iscram. setting;andEa_ control-

. rod scram =doesinot' occur.  ;

_g_.

I w .-- ~~ ~ - ,-

)

, 1 Unit 2

{

PBAPS

2. l' BASES: FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Peach Bottnm Atomic Power Station Units-have been analyzed throughout the spectrum of planned operating conditions up to or above the thermal power condition required by Regulatory Guide 1.49. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7.1 of the FSAR. In addition,.3293 MWt is the licensed maximum power level-of each Peach Bottom Atomic Power Station Unit, and this represents the maximum steady state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the controlling factors,-such as void reactivity coefficient, control-rod scram worth, scram delay time, peaking factors, and axial pow 7r shapes. These factors are selected conservatively _with respect to their effect on the applicable

. t. r a r , ent results as determined by the current analysis model.

Con crvatism incorperated into che transient analyses is documented in References 2 and 3. l

-17

l Unit 2-1

, 2.1 BASES (Cont'd) 1 1

L. References l

1. Linford, R. B., " Analytical Methods of Plant  !

Transient Evaluations for the General Electric I Boiling Water Reactor", NEDO 10802, February 1973.

2. " General Electric Standard Applicacion for Reactor Fuel", NEDE-240ll-P-A (as amended).
3. " Methods for Performing BWR Reload Safety Evaluations," PECo-FMS-0006-A (as amended).

.. _ _ _ _ . . . . ._ _ _ _ __ . _ _ . _ _ . _ . - _ _ _.m .__ ..-

1 Unit 2 PSAPS 3.5. BASES 1 Cont'd)

J. Local LHGR This specification-assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation. The maxi:aum LEGR shall be checked daily during reactor operation at > 25%

- power to determine if fuel burnup, or control red movement has caused changes in power distributicr.. For LEGR to be at the design LEGR-below 25% rated thermal power, the peak local LEGR must be a factor cf approximately ten (10) greater than the average LEGR which is precluded by a considerable margin when empicying .any permissible control rod pattern.

K. Minimum Crl.' cal Power Ratio (MCP?1

- Ocerating Limit MCPR The recuired operating limit MCPR's at eteady state operating conditions are derived from the establiched fuel cladding integrity Safety Limit MCPR a>l 'nalyses of tb~ abnc tal operational transients presented in Supplen 31 Reload - nsing Analysis and References 7

- and 10. For any abne tal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it.is r; quired that the resulting MCPR does not decrease-below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specificaticn 2.1.

l-To assure that the fuel cladding-integrity Safety Limit is net -

siolated during any anticipated abnormal cperational transient, the most ' limiting transients have been analyted to determine whien result

- in the largest reduction in critical power ratio (CPR). The transients evaluated are as described in References 7 and 10. l l

i l

-140a-1 l

1 I

5. , . < # e v s- -

e

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

x Unit-2 s

PBAPS 3.5.K. BASES (Cont'd)

The largest-reduction in critical powerfratio is then added to the

' fuel cladding integrity safety limit MCPR to establish the MCPR

Operating Limit for each fuel type.

Analysis of the abnormal operational transients is presented in References-7'and 10. Input data and operating conditions used in this analysis are shown in References 7 and 10 and in the Supplemental Reload Licensing Analysis.

-3.5.L. Averace Planar LHGR-(APLHGR), Local LHGR and F.inimum Critical Power Ratio (MCPR)

In the event that the calculated value of APLHGR, LHGR or MCPR exceeds

-its limiting value, a-determination is made to ascertain-the cause and initiate corrective actions to restore _the value to.within prescribed limits.. The status of all indicated limiting fuel bundles is reviewed as.wcll as input data associated with the limiting values such as power distribution, instrumentation data (Traversing In-Core Probe -

'TIP, Local Power Range Monitor - LPRM, and reactor heat balance

-instrumentation), control rod configuration, etc., in order to

.dete'rmine whether the calculated values are valid.

In.the event that the review indicates that the calculated value exceeding limits ^1s valid, corrective action is immediately undertaken to' restore the value to within-prescribed _ limits. Following corrective action,.which may involve alterations-to the control rod configuration and_ consequently changes.to the core power distribution, revised instrumentation data, including changes to the relative neutron flux distribution, for up to 43 in-core locations is obtained land:the powerEdistribution, APLHGR, LEGR and MCPP--calculated. *

,. Corrective action is initiated within one hour of an indicated value E < exceeding limits and verification-that the indicated value is+within p prescribed limits is obtained-within five hours of the initial indication.

h In: the event that the calculated value of APLHGR, LHGR or MCPR l exceeding 7itsflimiting value is not valid, i.e., due to an erroneous -

L instrumentation indication,. etc. , corrective action is initiated 1 m w

.~ithin one hour of an indicated value exceeding limits. Verification H that the indicated value is within prtiscribed limits is obtained

! -within.five hours of'the initial indication. Such an. invalid L . indication.would not be a violation cf the l'imiting condition for L- operation and therefore would not constitute a reportable occurrence.

l l

-140b-1

. . Unit 2 l

, PBAPS

, 3.5.L. BASES (Cont'd)

Operating experience has demonstrated that a calculated value of APLHGR, LHGR or.MCPR exceeding its limits value predominantely occurs due to this latter cause. This experience coupled with the extremely unlikely occurrence of concurrent operation exceeding APLHGR, LHGR or MCPR and a Loss-of-Coolant Accident or applicable Abnormal Operational Transients demonstrates that the times required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) are adequate.

3.5.M. References

1. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7 and 8, NEDM-10735, August 1973.
2. Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (Regulatory Staff).
3. Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27, 1974.
4. Genera] Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CPR 50, Appendix K, NEDE 20566 (Draft), August 1974.
5. General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L.

Gyorey to Victor Stello, Jr., dated December 20, 1974.

6. DELETED.
7. " General Electric Standard Application for Reactor Fuel", NEDE- l 240ll-P-A (as amended),

i

8. Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, NEDO-24081, December 1977, and for Unit 3, NEDO-24082, December 1977.
9. Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, Supplement 1, NEDE-24081-P, November 1986.

l

10. " Methods for Performing BWR Reload Safety Evaluations," PECo-FMS-0006+A (as amended).

l.

t

-140c-(

I

l Unit 2 )

PBAPS 6.9.1' Routine Reports (Cont'd) l (3) PECo-FMS-0003-A, " Steady-State fuel Performance Methods Report" (4) PECo-FMS-0004-A, " Methods for Performing BWR Systems Transient Analysis" (5) PEco-FMS-0005-A, " Methods for Performing BWR Steady-State Reactor Physics Analysis" (6) PECo-FMS-0006-A, " Methods for Performing BWR Reload Safety Evaluations" (3) The core operating limits shall be determined such tnat all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

(4) The CORE'0PERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be submitted upon issuance for each Operating Cycle to the NRC Oocument Control Desk with copies to the Regional Administrator and Resident Inspector.

l l~ -256a-

l Unit 3

. PBAPS l

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

- 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Appl.icability: Applicability:

The Safety Limits established The Limiting Safety System Settings to preserve,the fuel cladding apply to trip settings of the integrity apply to those instruments and devices which are variables which monitor the provided to prevent the fuel fuel thermt> behavior. cladding integrity Safety Limits from being exceeded.

Objectives: Objectives:

The objective of the Safety The objective of the Limiting Safety Limits is to establish limita System Settings is to define the which assure the integrity of level of the process variables at the fuel cladding. which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.

Specification: Specification:

A. Reactor Pressure 1 800 psia ~ The limiting safety system settings and Core Flow 2 10% of Rated shall be as specified below:

A. Neutron Flux Scram The existence of a minimum 1. APRM Flux Scram Trip Setting critical power ratio (MCPR) LRun Mode) less than 1.06 for two recirculation loop operation, When the Mode Switch is in the 1 or 1.07 for single loop RUN position, the APRM flux operation, shall-constitute scram trip setting chall be:

violation of the fuel' cladding

-integrity safety limit. S < 0.58W + 62% - 0.58 AW To ensure that this safety where:

limit is not exceeded, neutron flux shall not be abcVe-the S = Setting in percent of rated scram setting established in thermal power (3293 MWt) specification 2.1.A for longer than 1.15 seconds as indicated W = Loop recirculating by-the process computer. When flow rate in percent

'the process computer is out of of design. W is 100 for service this safety limit shall core flow of 102.5 be assumed to be exceeded if million lb/hr or greater.

the neutron flux exceeds its scram' setting and a control l- rod scram does not occur.

_9_

Unit 3 PBAPS

' 2 .1 BASES: FUEL CLADDING INTEGRITY The abnormal operational transients applicab'.e to operation of the Peach Bottom Atomic Power Station Units have been analyzed '

throughout the spectrum of planned operating conditions up to or above the thermal power condition required by Regulatory Guide 1.49. The analyses were ba.,ed upon plant operation in accordance with the operating map given in Figure 3.7.1 of the FSAR. In addition, 3293 MWt is the licensed maximum power level of each Peach Bottom Atomic Power Station Unit, and this represents the maximum steady' state power which shall r.ot knowingly be exceeded.

Conservatism is incorporated in the transient analyses la estimating the controlling factors, such as void reactivity

~

coefficient, control rod scram worth, scram delay time, peaking factors,:and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model.

Cone ^rvatism incorporated into the transient analyses is doct';nted in References 2 and 3.

l l

I

, Unit 3 2.1 BASES (Cont'd)

L. References

1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor", NEDO 10802, February 1973.
2. " Qualification of the One-Dimensional Core '

{ Transient Model for Dolling Water Reactors", NEDO 24154 and NEDE 24154-P, Volumes I, II, and III.

3. " Safety Evaluation for the General Electric Topical Report Qualification of the'One-Dimensional Core Transient Model for Boiling Water Reactors NEDO-24154 and NEDE 24154-P, Volumes I, II, and III.

I 4. " General Electric Standard Application for Reactor Fuel", NEDE-2dO11-P-A (as amended).

5. " Methods for Performing BWR Reload Safety Evaluations," PECo-FMS-0006-A (as amended).

F F

M h

)

l

Unit 3

  • PSAPS 3.5 BASES (Cont'd)

J. Local LEGR This specification assures that the linear heat generation rate in any EX8 fuel rod is less than the design linear heat generation. Tne maximum LEGR shall be checked daily during reactor cperatien at > 25%

power to determine if fuel burnup, or control rod movement has caused changes in power distribution. Fcr LEGR to be at the design LEGR 7 below 25% rated thermal power, the peak local LEGR must be a factor of approximately ten (10) greater than the average LEGR which is precluded by a considerable margin when employing.any permissible control. rod pattern.

K. Minimum Crl[ical Power Ratio (MCPR)

Operating Limit MCPR The required operating limit MCPR's at steady state operating conditiens are derived frcm the established fuel cladding integrity Safety Limit MCPR and analyses of the abr.crmal cperational transien:s presented in supplemental Relcad Licensing Analysis and References 7 and 10. For any abnormal operating transient analysis evaluation with the initial condition of the reacter being at the steady state operating limit it is recuired that the resulting MCPR does nct decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in specification 2.1.

l To assure that the fuel cladding integrity Safety Limit is .c:

violated during any anticipated abncrmal cperational transien:, the most limiting transients have been analyzed to de: ermine which resul:

in the largest reduction in critical power ra:ic (CPR). The transients evaluated are as described in References 7 and 10. j l

1:

l

)

l l

l

-140a-1

Unit 3 s PBAPS 3.5.K. BASES (Cont'd)

The largest reduction in critical power ratio is then added to the fuel cladding integrity safety limit MCPR to establish the MCPR Operating Limit for each fuel type.

Analysis of the abnormal operational transients is presented in References 7 and 10. Input data and operating conditions used in this analysis are shown in References 7 and 10 and in the Supplemental Reload Licensing Analysis.

3.5.L. Average Planar LHGR (APLHGR), Local LHGR and Minimum Critical Power Ratio (MCPR)

In the event that the calculated value of APLHGR, LHGR or MCPR exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective action to restore the value to within prescribed limits. The status ec all indicated lir.iting fuel bundles is reviewed as well as input data associated with the limiting values such as power distribution, instrumentation data (Traversing In-Core Probe -

TIP, Local Power Range Monitor - LPRM, and reactor heat balance instrumentation), control rod configuration, etc., in order to determine whether the calculated values are valid.

Innthe event-that the review indicates that the calculated value

-exceeding limits is valid, corrective action is immediately undertaken

-to. restore the value to within prescribed limits. Following corrective action, which may involve alterations to the control rod

. configuration and consequently changes to the core power distribution, '

revised instrumentation data, including changes to the relative neutron flux distribution, for up to 43 in-core locations is obtained and the power distribution, APLH1R, LEGR and MCPR calculated.

Corrective-action -is initiated within one hour of an indicated value exceeding limits and verification that the-indicated value is within prescribed limits is obtained within five hours of the initial t indication.

In the event that the calculated value of APLHGR, LHGR or MCPR exceeding its-limiting value is not valid, i.e., due to an erroneous instrumentation indication, etc., corrective action is initiated within one hour of an indication value exceeding limits. Verification

.that the indicated value is within prescribed limits is obtained

.within five hours of the initial indication. Such an invalid

-indication would not be a violatio- of the limiting condition for operation and therefore would not constitute a reportable occurrence.

-140b-

Unit 3

~

PBAPS

. 3.5.L. BASES (Cont'd)

Operating experience has demonstrated that a calculated value of APLHGR, LHGR or MCPR exceeding its limiting value predcminantely occurs due to this latter cause. This experience coupled with the 7 extremely unlikely occurrence of concurrent operation exceeding APLHGR, LHGR or MCPR and a Loss-of-Coolant Accident or applicable Abnormal Operational Trancients demonstrates that the times required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) are adequate.

3.5.M. References

1. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7 and 8, NEDM-10735, August 1973.
2. Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (Regulatory Staff).
3. Communication: V. A. Moore to I. S. Mitchell, " Modified CE Model for Fuel Densification", Docket 50-321, March 27, 1974.
4. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CPR 50, Appendix K, NEDE 20566 (Draft), August 1974.
5. General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L.

Gyorey to Victor Stello, Jr., dated December 20, 1974.

6. DELETED.
7. " General Electric Standard Application for Reactor Fuel", NEDE- l 240ll-P-A (as amended).
8. Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, NEDO-24081, December 1977, and for Unit 3, NEDO-24082, December 1977.
9. Loss-of-doolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, Supplement 1, NEDE-24081-P, November 1986, and for Unit 3, NEDE-24082-P, December 1987.
10. " Methods for Performing BWR Reload Safety Evaluations," PECo-l FMS-0006-A 'as amended).

-140c-

Unit 3

^

PBAPS 6,9.I Routine Repor!' (Cont'd)

[a-FMS-0003-A, " Steady-State fuel Performance Methods Report" (4) PECo-FMS-0004-A, " Methods for Performing BWR Systems

. Transient Analysis" (5) PECo-FMS-0005-A, " Methods for Performing BWR Steady-State Reactor Physics Analysis" l

(6) PECo-FMS-0006-A, " Methods for Performing BWR Reload Safety Evaluations" (3) The core operating limits shall be determined such that all applicable limits.(e.g., fuel thermal-mechanical limits, core

> thermal-hydraulic limits, ECC5 limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

(4) The CORE OPERATING LIMITS REPORT, including any mid-cyclc revisions or supplements, shall be submitted upon issuance for each Operating Cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector, a

E

-256a-

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