ML18039A461

Proposed Tech Specs B 3.6-2,B 3.6-7 & B 3 3.7-2 for Power Uprate Operation Omitted Pages

ML18039A461
Person / Time
Site:	Browns Ferry
Issue date:	07/31/1998
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18039A460	List: ML18039A459 ML18039A461
References
NUDOCS 9808100172
Download: ML18039A461 (11)

Text

ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 and 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-384, DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE MARKED PAGES Aff'ected Pa e List The following pages have been marked and an 'X'as been placed in the right hand margin the indicate where changes occur. The affected pages list is identical for both Unit 2 and Unit 3.

Unit 2 Bases Unit 3 Bases B 3. 6-2 B 3.6-2 B 3. 6-7 B 3.6-7 B 3.7-2 B 3.7-2 9808iOOi72 98073i PDR ADQCK 050002bO P PDR

PT'i sary Cental neent 8 a.s.l,l SAS (continued)

APPLICABLE safety design bas~s for the primary contaf~ent 1$ that

'SlfPY ANN.VS'e ft must withstand the pressures and temperatures of the ligftfng QN ~ithout <<xc<<<<ding the design leakage rate.

The DSA that postulates the aaxiauw r<<l<<ase of radioactive material ~ithin prfaary conta$ aaent 1s a LOCA. In the analysis ot th1s acc1dent. 1t is assuaged that prifsary contafnw<<nt is OPERABLK such that release of fission products to the envfronm<<nt is controlled by the rate of prfeary contafnaent leakage.

Analytical Nethods and assumptfons involvin9 the pr fisary contafnaent are pres<<nted in Refer<<nces I and ?. The safety analyses assume a nonmechanistic fission product release foll<ming a OSA, Wfch forms 4h>>Jas4s for.deterafnatfon of offsite doses. The fission product release is, fn turn, based on an assuaef leakage rate from the prfaary containment. OPERABILITY of the prfaary containamt ensures that the leakage rate assumaed in the safety analyses is not exceeded.

The maxim allovatAe leakage rate for the prfeary contafnaant (L,) is 2.0S by wight of the contafnsent a1r per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the desfgn basis LOCA eaxfea peak containment pressure (P,) oWKQpsfg (Ref. I).

t~a"go. a, satisfies Criterion Primary containment 3 of the NC Policy Statesont {Ref. 6).

primary contafnwent OPERNlllTY is aafntafned by 1 faftfng leakage to s l.0 I, except prior to the first startup after rforidng. a required Prfaary Containment Leakage Rate estino Prograa leakage test. At this tfwe, applicable leakage 1fafts est be Wet. Coeplfance Hth this LCO will ensure a prfaary contafnaent conffguratfon, including equfpeent hatches, that is structurally sound and that ~ill lfait leakage to those leakage rates assuae{l fn the safety analyses.

Individual leakage rates specffied for the priILary contifnaeit air lock are address<<d in LCO 3.6.1.2.

8 3.6-2

primary t:ontainlaent Air Lock

.. B 3.6.1.2 and leak tightness are essential for maintaining primary containment leakage rate to within limits in the event of a QBA, Not maintaining air lock integr)ty or leak tightness

~ result,in a leakage rate in excess of that assumed fn the unit safety analysis.

The ON that postulates the iaximua release of radioactive ANALYSES aaterial within primary containment is a Ugj. gn the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is contra)led by the rate of primary containment leakage. The priaary containment is designed with a maximus allowable leakage rate (L,) of.2.gL by night of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the allowable'eakage calculated naxiae peak containment pressure (PJ of 9335 psig (Ref. 3}. This rate fores the s for the acceptance criteria iaposed on <he Ns So 6 associated, with the air lock.

Prissy containeent air lock OPGNStLITY is also regired to IIiniiize the ~unt of fission product gases that Imay escape primary containaent through the air lock and contalinate and pressurize,the secondary containment.

The primary containment air lock satisfies Criterion 3 of the NIC Policy Statement (Ref. I).

As part of primary cantailnt, the air lock's safety function fs re)ated to control of conta$ naent leakage rates fol)erlng a ON. Thus, the air lock's structural integr1ty and leak tightness are essential to the successful altigation'f such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERA8LE. the air lock interlock aechanisN aust be OPERABLE, the air lock aust be in cosyliance Hth the Type B air lock leakage test, and both air lock doors must be OPHNlLE. The interlock allows only one air lock door to be opened at a tin. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be (cont$ nueo) 8$ -UNIT Z B 3.6-7 AaenCee ~ '.

RHRSM System B 3.7.1 BASES (continued)

APPLICABLE The RHRSM System removes heat from the suppression pool to SAFETY ANALYSE limit the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSM System to support long term cooling of the reactor or primary containment is discussed in the FSAR, Chapters 5 and (Refs. 2 and 3, respectively). These analyses explicitly ll assume that the RHRSM System will provide adequate cooling support t'o the equipment required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures'and considers the number of units fueled. Mith one unit fueled, the worst case single failure that would affect the performance of the RHRSM System is any failure that would disable two subsystems or pumps of the RHRSM System (e.g, the failure of an RHR Suppression Pool Cooling/Spray return line valve which effectively disables two RHRSM subsysti or pumps).

Mith two and three units fueled, a worst case single failure could also include the loss of two RHRSM pumps caused by losing a 4 kV shutdown board since there are certain alignment configurations that allow two RHRSM pumps to be powered from the same 4 kV shutdown board. As discussed in the FSAR, Section Il.6.3.3.2 (Ref. 4) for these analyses, manual initiation of the OPERABLE RHRSM subsystems and the associated RHR System is assumed to occur 10 mi af oooo DBA. The RHRSM flow assumed in the analyses is i pm per pump with two pumps operating in one loop. In th s case, the maximum suppression chamber water temper re and pressure are 177'F (as reported in Referenc an X psig, respectively, well below the design temperature f 281'F and maximum allo~able pressure of 62 psig. This is al'so below the 200'F limit imposed by Design Criteria BFN-50-706IA (Ref. 5) for all plant transients involving SRV operations.

The RHRSM System"sat sfies Criterion 3 of the NRC Policy Statement (Ref 6).

(continuedj BFN-UNIT 2 B 3.7-2 Amendment

Primary Contatnaent B 3.5.1.1

, BASg$ (contf nued)

APP VOCABLE The safety design basis for the primary contafnment it must withstand the pressures and temperatures of the fi that SaPgn ANALYSES limiting OBA ~ithout exceeding the design leakage rate.

The OBA that postulates the maximum release of radioactive material efthfn primary containment is a LOCA. In the

~

analysis of this accident; it is assumed that pr)mary contafnment is OPERABLK such that release of fission products to the environment fs controlled by the rate of primary containlent leakage.

Analytical methods and assumptions involving the primary containment are Presenged in References l and 2.- The safety analyses assume a nonmechanfst$ c fission product release follo~ing a DBA, ~hich forms the basis for deterafnatfon of offsfte doses. The fission product release is. in turn, based on an assuaed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not,-

exceeded.

The maxfaua allowable leakage rate for the primary contafnaent (lJ is 2.'0S by wight of the coniafnment afr per 2i houry at the design b sis LOCA maxfaa peak contailnt pressure (PJ of ~sfg (Ref. I).

So.6, Prfmary contafrusent satisffes Crfterfon 3 of the NK Policy Stateaent (Ref. 6).

~ ~

LCO Prfaary containment OPERABILITY fs mafntafned by lkrltfni leakage to s 1.0 Lexcept prior to the ffrst startup after rforifn9 a required Primary Contafnmat Leakage Rata estfne Program leakage test. At this tfm, applfcable lethge lfifts ant be Net. Coaplfancl fifth thfs LCD adll ensure a pessary containment configuration, including eqefpment hatches, that is structurally sound and that All if@it leakage to those leakage rates assumed in the safety analyses.

Individual leakage rates specified for the primary containment afr lock are addressect in LCO 3.6.1.2.

(continued)

~elIT 3 B 3.6-2 Amendment I

~ \

Pr)eat'y Containment Air Lack 5

8 3.5.1.2 g~

ROUND and leak tightness are essential <<r maintaining priiary ontinued) containment leakage rate to within limits in the event of a DBA. Hot maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assuaed in, the unit safety analysis.

AFPL,ICABLK The QBA that postulates the eaxfeua release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. ln the analysis of this accident, it is assumed that primary containmnt is OPHNSLE. such that release of fission products to the environment is controlled by the rate of priaary containaent 1eakage.'he primary containaent is designed with a maximus allowable leakage rate tLJ of 2.{S by weight of the conta$ nnant air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the calc ated maxima peak containment pressure (P,) of sag (Ref. 3). This allowable leakage rate foras the s s for the acceptance criteria ilposed on the SRs associated with the a)r lock.

Primary containment air lock OPERANLLDV is also re~irect to

~ )niaize th>> amount of fission product gases that lay escape primary containment through the a1r lock and contaminate and pressurize,the secondary containment.

The pr$ aary containment air lock satisfies Criterion 3 of the NC Policy Statement (Ref. 1).

As part of priaary containment, the air lock's safety function is related to control of containment leakage rates fo)ledni a GN. Thus, the air lock's structural 4ntegrity aef leak tightn. s are essential to the successful sit!Nation of such an event..

The priaary conta$ naent air lock is required to be OPERABLE.

For the air lock to be considered OPERASLE, the air lock .

interlock eechaniaa aust be OPERABLE, the air lock sist be in coaplfance with the Type I air lack leakage test, and both air lock doors aust be OPERABLE. The interlock allows only one air lock door to be opened, at a tim. This provision ensures that a gross breach of. priaary containsent does not exist when primary containaant is required to be (continued)

I

~ Q-NIT 3 9 '3.6-7 lagndeent

4 1

RHRSM Systee nd 8 3.7.1 BASES (continued)

APPLICABLE The RHRSW System removes heat from the suppression pool to SAFETY ANALYSE lfmft the suppression pool temperature and primary containment pressure following a LOCA. This ensures that the pr1mary conta1nment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment fs discussed in the FSAR, Chapters 5 and 14 (Refs. 2 and 3, respectively). These analyses explicitly assume that the RHRSW System will prov1de adequate cooling support to the equfpmcnt required for safe shutdown. These analyses include the evaluation of the long term primary containment response after a design basis LOCA.

The safety analyses for long term cooling were performed for various combinations of RHR System failures and considers the number of units fueled. With one un1t fueled, the worst case single failure that would affect the performance of the RHRSW System is any failure that would disable two subsystems or pumps of the RHRSW System (e.g, the failure of an RHR Suppression Pool Cooling/Spray return line valve which effectively disables two RHRSW subsystems or pumps).

With two and three units fueled, a worst case single failure could also include the loss of two RHRSW pumps caused by losing a 4 kV shutdown board since there are certain alignment configurations that allow two RHRSM pumps to be peered from the same 4 kV shutdown board. As discussed in the FSAR, Section 14.6.3.3.2 (Ref. 4) for these analyses, manual initiation of thc OPERABLE RHRSW subsystems and the associated RHR System is assumed to occur 10 mi af OBA. The RHRSW flow assumed fn the analyses is gpm per pump with two pumps operating fn one loop. In this case, the aaxfoua suppression chamber water temper rc an 3 )c pressure are 177'F (as reported in Reference an 3 4%4 psfg, respectively, well below the design temperature x of 281'F and maximum allowable pressure of 62 psig. This is

.also below the 200'F limit imposed by Oesfgn Criteria BFN-50-7064A (Ref. 5) for all lant transients 1nvolving SRV op rat1ons. y, .ii .iq ik. u The RHRSW System"satisfies Critcrfon 3 of the NRC Policy Statement (Ref 6).

(continued BFH-UNIT 3 B 3.'-2 Amend.-..".

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