ML19340D974
| ML19340D974 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/05/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Speis T, Zieman D Office of Nuclear Reactor Regulation |
| References | |
| IEB-80-18, NUDOCS 8101060040 | |
| Download: ML19340D974 (1) | |
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NRC PDR Local PDR DEC 5 1980 NRR Reading LB#2 File DEisenhut RPurple Docket No. 50-327 RTedesco ASchwencer CStabla Jj M ervice MEMOMNDUM FOR:
T. Speis, Chief 01&E (3)
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Reactor Systems Branch, DSI
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q BCC: NSIC D. Ziemann, Chief TERA
".i Procedures and Test Review Branch, DHFS AERS (ifd.
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FROM:
A. Schwencer, Chief Licensing Branch No. 2, DL
SUBJECT:
PROPOSED SEQUOYAH LICENSE AMENDMEtT Attached is a copy of the TVA proposed amendment to confom with IE Bulletin 80-18. TVA has detemined that in accordance with the provisions to 10 CFR 50.59 this matter is an unreviewed safety question. Therefore, a license amendment is needed prior to proceeding with the proposed modifications.
The PM has discussed the IE Bulletin 80-18 with IE Headquarters (Roy Woods) and Region II (H. Dance) to detemine the appropriate course of action. We understand that NRR reviewed the Bulletin and other licenseesse proceeding without license modifications. TVA will not proceed until we infom them of our findings.
You are requested to make a finding on this issue by January 17,198q. Use TAC No. 43227 for this effort.
Original signed by A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing Attachment.
Proposed AmenMnt 7 " 3 0 8 98YO y
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TENNESSEE VALLEY AUTHORITY c
CH AMA NCCGA. *ENN ESSEE 374o1 400 Chestnut Street Tower II
'Noventer 18, 1980 Mr. Harold R. Denten, Director Office of Nuclear Reacter fegulatica h
U.S. Nuclear Regulatcry Cc:mnission l
Washington, DC 20555 t
Dear Mr. Denton:
In the Matter of
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Docket No. 50-327 Tennessee 7 alley Authority
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- n accordance with the provisicas of 10 CFR 50.59, we have determined *2at the :nedificaticcs propcsed to maolve the concern identified in Office of Inspectica and Enforcement Bulletin (IEB) 80-18 en adequate sin %m flow through centrifugal charging pumps constitute an unreviewed safety questien. Therefore, we request an mendment to license DPR-77 to allow completien of *2e proposed :nodifica.cn on Sequoyah Nuclear Plant unit 1.
Enclosed are 41 copies of the folicwing information.
(1) The propcsed amend:nent to license DPR-77 (2) T7A's October 16, 1980, response to IEB 80-18 (3) T7A's justificatica for the :nodification In acccedance with the require:nents of 10 CFR 170.22, we have deternined the proposed amend:nent to be Class III. This classificatica is based en our belief that no significant hasards censideratica is involved. The renittance of $4,000 is being wired to the Nuclear Regulatcry Cemnission, Attention: licensing Fee Management Branch.
7ery truly yours, TENNESSEE 71LLE! AD*EORITY J.
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p L. M. Mills, Manager Nuclear Regulatien and Safety Swcru to and subscribed befers :ne this : -> day of~
1980
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2-Mr. Harcid R. Denten, Director Ncvember 18, 1980 cc:
Mr. James P. O'Reilly, Director (Enclosure)
Office of Inspectico and Enfercement U.S. Nuclear Regulatory Commission Regien II - Suite 3100 101 Marietta Street Atlanta, Gecrgia 30303 Mr. Victer Stelle, Jr., Director (Enclosure)
Office of Inspection and Enforcement U.S. Nuclear Regulatory Consnission
- dashiraten, DC 20555 i
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ENCLOSURE 1 SEQUOYAH NUCLEAR PLANT UNIT 1 PROPOSED LICENSE AMENDME!C In order to complete the modifications recommended by IE Bulletin 80-18 on adequate minimum flow through centrifugal charging pumps, we request that the following paragraph be added to license DPR-77.
In ccnformance with IE Bulletin 80-18, TVA shall complete interim modificatiens identified in L. M. Mills' letter to J. P. O'Reilly dated October 16, 1980, to ensure adequate minimum flew through the centrifugal charging pumps.
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ENCLOSURE 2 i
SECUOYAH NUCLEAR PLANT UNIT 1 TVA'S OCTOBER 16, 1980, RESPCNSE TO IE BULLETIN 80-18 ADECUATE MINIMUM FLOW TO CENTRIFUGAL CHARGING PUMPS Reseense to Item 1 of the Bulletin TVA has completed calculations to determine if the Sequoyah Nuclear Plant unit 1 charging system would maintain adequate pump flew during parallel safety injecticn operation and determined that adequate ficw would not be
=aintained. The detailed calculations outlined by the Westinghouse Electric Corporation letter (NS-TMA-2245) are included as Attachment 1.
Rescense to Item 2 of the Bulletin a.
Modificatiens are planned for Sequoyah unit 1 as described under Interim Ebdification I of the Westinghouse letter attached to the bulletin. These modificatiens include:
(1) Verifying that the CCP miniflow return is aligned directly to the CCP suction during normal operation with the alternate return path to the volume control tank isolated (Iceked closed). _
(2) Removing the safety injection initiation automatic closure signal from the CCP miniflow isolation valves.
(3) Modifying plant emergency operating procedures to instruct the operator to:
(a) Close the CCP miniflow isolation valves when the actual RCS pressure drops to the calculated pressure for manual reactor coolant pump trip.
(b) Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater than 2,000 i
psig.
b.
As indicated in the Westinghouse Electric Corporation safety evaluation
( Attach =ent 2), if manual operator action is taken to close the CCP minificw valves when the RCS pressure drops to the calculated pressure fer =anual reactor coolant pump trip (1,500 psig), no significant change in peak clad temperature (PCT) would be observed. Since tripping of the reactor coolant pumps is itself a =anual operator action, it is our opinien that the additicnal requirement of closing the CCP miniflow valves (two handswitches) will not burden the operator and can be accomplished in the time necessary.
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The CCP miniflow valves are supplied with shutdown power via the diesel c.
generators. The same post-accident monitoring instrumentation (powered by batteries and/or diesel generators) used to determine the reactor coolant pu=p trip pressure will be utilis:ed to determine the need for opening er closing the CCP miniflow valves.
d.
As indicated in the 'destinghouse safety evaluation, the flow available from the CCP's with the modification in place, along with the operator actlen indicated in item 2.b above, will have a negligible effect on the safety-related analysis (note that part of Attachment 2 is dedicated to UHI plants),
e.
Since the results of the safety-related analyses evaluated in item 2.d indicate the insignificant effects of the interim =cdification and pro-cedure change, all technical specifications based en these remain valid.
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c ATTACHMENT 1 SEQUOYAH NUCLEAR PLANT UNIT 1 MINI!C CENTE!?UGAL CHARGING ? UMP FLCW DURING TWO PUMP PARALLEL SAFETY IN'ECTICN CALCULATION FOR NRC IE BULLETIN No. 80-18 l
Purcose Check capability to provide minimum pump flow during parallel safety injection with two centrifugal charging pumps (CCP's).
I References 1.
NRC II Bulletin No. 80-18.
J 2.
Letter from T. M. Anderson, Westinghouse Water Reactor Division, to V. Stello, NRC, dated May 8,1980, No. NS-TMA-2245.
3.
Sequoyah Nuclear Plant Unit 1 Preoperational Test WG.lC data.
Calculations Following the for=at suggested in Reference 2, using data from Reference 3.
Step 1: Maximum developed head pump flow = 2,600 psid = 6,006 ft. @
73.1 gym (pump 1A-1A)
Minimum developed head pump flow = 2,470 psid = 5,705 7 f t.
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72.3 gym ~(pump 13-13)
Step 2: Correction for testing error.
Test gauge accuracy =.25 x 3,000 psig = 7.5 psi (17.25 f t.)
+ 10 psi (23 ft.) reading accuracy = 40.25 ft.
Maximum pu=p = 6,046.25 ft. @ 73.1 gpm Minimum pump = 5,665.45 ft. @ 72.3 gpm Step 3: From construction of pump flow curves, attached, minimum pump =
5,670 ft. @ 60 gym projection of weak pump head point on strong pump operating curve shows flow of 224 gpm.
Total flow from both CCP's guaranteeing 60 gpm to tie weak pump is 224 gpm + 60 gpa = 284 gpm Step 4: Determination of injection piping head loss.
Frem Reference 3, runout head of pu=p 1A-1A = 480 psi runout flow of pump 1A-1A = 490 gpm F
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ATTACHMENT 2 UESTINGHOUSE ET.ECTRTC CORPORATION SAFETY EVAI.UATTON-A r,
CEniRIFUGAL CHARGING PU",? OPERATI0tt FCL'.CWING SECONCARY SIDE HIGH ENERGY LINE RUPTURE 2
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i Reference 1:
NS-T".A-22:5, 5/3/50 Reference 1 notified the NRC of a concern for consequential damage of ene or mere centrifugal charging pumps (CC?) following a secondary system high energy line rupture. Reference 1 included a calculaticnal method b
and sample calculatien to permit evaluation of this concern on a plant specific basis.
Sheuld a plant specific pr:blem be identified, Westinghcuse provided several rec:=endatiens for the interim until necessary design m:difications can be implemented to resolve the problem. These rece menda-i tiens inci'uded tw: pr:p: sed interim modifications which included:
1.
Re :ve the safety injection initiation aut::atic closure signal from the CC? minif1:w isolatien valves.
- 2.
Modify plant emergency operating precedures to instruct the operator to:
Cicse the CCP miniflew isolation. valves when tne actual RCS a.
pressure drops to the calculated pressure for manual reactor
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-coolant peqp trip.
b.
Recpen the CC? miniflow isolatien valves should the wide range RCS pressure subsequently rise to greater than 2000 psig.,
c Prior to making this recc=endation, Westinghouse evaluated the impact of the rec:= ended operating procedure modifications en the results of the.
Varicus accidents which initiate safety injection and are sensit'ive to CCP ficw delivery. The accidents evaluated in detail include secondary system ruptures and the spectrum of small icss of coolant accidents.
The analytical results for steam generator tube rupture and large Icss of coolant accident are' net sensitive t's a reduction in CC? ficw of'the magnitude that results from the rec:= ended : difications. This letter functions to supplement Referencel'andidentifythesensikivityoftheac:identanalysesto the rec:= ended ::difications. This evaluati:n is generic in nature.
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Sensitivity analyses have been performed for sece'ndary high energy line evaluate the impact of reduced safety injecticn flew due to ruptures t:
normally open minificu isciatien valves.
These analyses indicate an insignificant effect en the plant transient respense.
1 A.
Feedline Rupture
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Foliewing a feedline rupture, the reacter c:olant pressure will, reach the pressuri:er safety valve setpoint within apprcximately 100 seconds assuming maximum safeguards with the pcwer-cperated relief valves With minimum safeguards, the reacter c clant pressure will inoperable.
not reach the pressurizer safety valve set;0 int until a; proximately The time that the reacter coolant system pressure remains 300 sec:nds.
at the pressuri:er safety valve setpcint is a function of the auxiliary feedwater fl:w injected into the non-faulted steam generators and the' With the mini-time at which the Operat:r is assumed to take action.
ficw isciati:n valves open,'the peak react:r c:clant system pressure and the water discharged via the pressuri:er safety valves are insignifi___
cantly changed from the FSAR results..
B.
Steamline Rupture l
The effects of maintaining the minificw isolation valves in a 'normaliy
, open position was also investigated following a main steamline rupture.
For'the conditien II " credible" steamline rupture, the results of_the transient with the minificw valves open shewed that the licensing Icriterien(noreturntocriticalityafterreactortrip)continuesto b
The c ndition III and IV main steamline ruptures were also
,e met.
reanalyzed' assuming the minifica valves were open. The results of i_h,e analysis shewed that, even with rede:ed safety injection ficw I
.,p.tothec:re,nc0.':3occurredforanyrupture.
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Small less of C: lan: Ac:idents l
Sensitivity analyses have been performed to evaluife the impact of reduced safety injection flew on small break 1 css of c:clant accidents (LOCAs).
These analyses indicated : hat miniflew isolation can be delayed, but it I
must occur at sc=e time into the small break LOCA transient in order to limit the peak clad temperature (PCT) penal:y.
The pr:p: sed :dification delays minificw isolation and reduces SI flew delivered by appr:ximately 45 gpm at 1250 psia during the delay time period.
The i= pact of this : dificatien was evaluated based en two isolation times:
- 1) The time ecuivalent to the RCP trip time, and 2) approximately 10 minutes in the transient, or just prior to system drain to the break for the worst t
small break si:es.
The second time was evaluated to determine the impact if the operat:r d:es not isolate minificw within the pr:pesed prescribed time. The s:se: rum of small break si:es are considered to ence: pass all possible small break scenaries. Cnly cold leg break iccations are considired i
since they will c:ntinue to be limiting in terms of PCT.
A.
Very small breaks that do not dr-in the RCS or uncover the core, and maintain RCS pressure :beve see:ndary pres,sure (< s2" diaceter).
Fct these break si:es, it is quite pessible that the ccera:Or may never isciate the miniflow line, since the pressure setpoint will not be reached, and continued pumped SI degradation will persist.
Mcwever, this will have no adverse consequences in terms of core l
uncevery and PCT. No core uncovery will be expected for the degraded SI case. similarly to the base c:mparisen case with ful.i SI. The caly effect would be a slightly lcwer equilibration pressure for a given break'si:e.
- 13. f.ill breaks that drain the RCS and resul.t in the maximum cladding ismierstures(2"< diameter <6").
This range of break si:es represents the wors small break size f:r
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l most plants as determined utili:ing the currently approved October 1975 If m1Nificw is Evaluatien Medal versicn, as shewn in WCAP-8970-P-A.
I isolated at the RCP trip setpoint rather then the "S" signal, a reduc-l tien in safety injection flew of less than 45 spm results, averaged j
for the approximately 50 second period of time separating the two events.
This reducticn in RCS liquid inventory "results in core uncovery less 1I If mini- -
than ene secend earlier, and has a negligible impact on PCT.
ficw is isolated at the time of core uncovery, er approximately 10 1
minutes for break si:es in this range, a greater reduction in RCS liquid inventory resuits in a core uncovery 10 seconds earlier in the transients resulting in less than a 10*F PCT penalty for the worst si:e small break.
This would not result in any present FSAR small break analysis beccming l
more limiting than the corresponding large break LOCA F5AR analysis.
I 1
If minifiew isolation does net'eccur at any time into the t_ransient for
' this category of small LOCA, a'. PCT penalty of 200*F or more ceuld occur.
1 Small break si:es larger than the worst break tnrcugh the inte:::ediate j
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break si:es (> 5" diameter)'.
l Break si:es in this range have been determine /.o be non-limiting for small break utilizing the currently approved October 1975 Evaluation Model, WCAP-8970-P-A.
If miniflew isolation occurs at the RCP trip time for these break si:es, the negligible effect :n; PCT presented./
above also applies.
Similarly, if isolation occurs prior to core
- However, uncovery, the small (< 10*F) PCT penalty will resuiras well.
for these larger break sizes, the time of first care uncovery occurs If mintElew isolation is not ;srformed until prior to 10 minutes.
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10 minutes, reduced SI will be delivered during the cre uncovery time.
which can have a greater impact on PCT.
Studies indicate a potential i
PCT penalty of 40*F resulting for these non-limiting break sizes if l
minifiew'is not isolated until 10 minutes. This is act expected to shift the worst break si:e to larger bre'aks, since mese breaks are typica-lly hundreds of decrees *1.5 th.n snaller lhiding smil bt'cak-analy:ed with the currently approved Evaluati,cn Mc:rit.
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It For all FSAR s=311 LCCA analyses, one ccmplete train failure is, assumed.
is clear that two charging pumps without minificw isolation provides more l,
flew than ene pump with miniflew isolation.
The 15. pact presented in this evaluatien maintains tne one train failure and assumes no minificw iso If both pumps were operating, the PCT results tien for the remaining pump.
l would be much Icwer than present FSAR calculations even if miniflow isola-In this situation, the I
tien is not assumed to occur for the two pump case.
I plant FSAR small break calculations remain conservative.
1 hhese sensitivity studies form the basis for the rece= ended interim The accide_sts evalu -
modifications to the emergency operating procedures.
- Further, ated are relatively insensitive-to' the recc= ended modifications.
the accidents evaluated will give'results that satisfy acceptance criteria *.
as icng as the CCp minificw is isciated within 10 minutes of event initiation.
i However, small LCCA sensitivity studies with one SI train operating confirm that small 1.0CA analyses require minificw isolation within 10 minutes.
To comply with the recc= ended modifications, the operator can isciate mini-ficw at any peint in the depressuri:ation transient prior to RCS pressure
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l reaching the RCP trip setpoint.
Should a repressurization transient occ,ur, f
the operator can open CCP minificw at any point between the RCP trip set-f point and 2000 psig.
Such operater actions will ensure that plant accidents satisfy acceptance criteria and protect the CCPs frem consequential damage i
during the repressuri:stien transient that accompanies a secondary system i
f high energy line rupture at high initial pcwer levels.
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CENTRIFUGAL CHARGING FUMP CPERATION FOLLCWI"G SECCSCARY SIDE HIGH ENERGY LINE BREAK (UH r
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The small 1 css of ccolant accident (LOCA) secticn of the main report was generated primarily for plant applications which do not include upper J
' head injectica (UHI) as part of the ECCS design.
This supplement pro-I; vides additicnal small LOCA information for UHI plants and, together with the main report, assesses the impact of delayed minificw isolation for small LCCAs fer UHI plants.
T he ecdel utili:ed to determine the SI sensitivities and *o identify the worst small break si:e discussed in the main report was the Octcber This model 1975 Model (WCAP-3970 o-A) version of the Evaluation Medei.
UHI small break analyses are is not yet approved for UHI plant analyses.
Hewever, sensi-performed with the December 1974 small break version.
tivity studies performed to determine the effect of pumped SI on small break LOCA PCTs utilizing the December ecdel yielded,nearly identical S results as presented in the main report. This is expected since the mcdei changes included in the Octcber =cdei do not affect the basic vessel inventcry and core boiloff relationships that determine the impact of changes in pumped safety injection on PCT.
An important difference in UHI plant small break analysis results as cc pared to similar ncn-UHI plant analysis results is the small break si:e resulting 'in the highest PCT.
This break size is generally greater for UHI plants than for similar non-UHI plants because of the additional safety injecticn ficw provided by the UHI accumulater ~at relatively high The worst small break si:e for UHI plants may be a RCS pressures.
six inch diameter break or larger.
The main report identified breaks of this size and larger as non-limiting small break sizes. While this is-true for non-UHI plants, it is not accurate for typical UHI plant small break analyses. Therefore, the stated 40*F potential penalty for e
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g 5 six inch breaks applies to the worst break for UHI plants for,the case l
l where minifl:w isolation is delayed until 10 minutes.
It is
'Jestinghouse's epinion, however, that the stated' penalty of 40*F is conservatively hign and beunding for UHI plants, for the folicwing e
reasens: a) The 40*F penalty was based cn, sensitivity studies performed assuming an appr:ximate 20'; reduction in total HPI flow.
However, the anticipated 20" reduction actually applies only to the charging pumps.
Intermediate head SI pumps are not affected.
Therefore, total HPI for plants with intermediate head SI pumps, which includes all UHI plants, will result. in less total degradatier., and thus a smaller PCT penalty.
The high pressure accumulator en UHI plants has a similar effect of k
educing the total HPI degradation due to the delay in miniflew isolation.
b) The UHI ac:umulator is a significant source of liquid mass inventary for breaks greater than o: equal to six inches in diameter. This addi-tienal mass delays the core uncovery tima as'ccmpared to the same size break ec:uring on a similar non-UHI plant, since more liquid mass must exit fr:m the break prior to core uncovery.
The delay in core uncovery'.
results in clad heatup at a lcwer power level caused by the decay in residual c re heat.
Therefore, clad heatup rates are slower which also tends ta reduce the sensitivity to changes in HPI delivery rate.
In conclusien, the sensitivity provided for six inch diameter and larger break sizes in the main report represents the worst break size range for UHI plants.
The stated 40*F PCT penalty for breaks of this size ',
resultant from a 10 minute dehy in miniflow isolation is a conservatively high and bounding value for UHI plants, for the reasons stated above.
If miniflow is isolated at the time of RCP trip, the negligible impact on PCT discussed in the main report applies for UHI, plants as well.
The <10*F penalty resultant if minificw isolatien occurs prior to core f
une:very also applies to UHI plants, with the added benefit that this event cccurs later in a UHI plant transient than for a non-UHI plant transient of the same break size, allcwing more time for the operator to act.
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1 ENCLCSURE 3 SEIUOYAH NUCLEAR PLANT UNIT 1 JUSTIFICATICN FCR CENTRIFUGAL CHARGINP PUMP MINIFLOW VALVE MODIFICATION The centrifugal charging pu=p =iniflew valve modification has been deter-
=ined to be an unreviewed safety questien caused by uncertainties involved with the =anual operator action required.
In the Westinghouse Safety Evaluation, sensitivity studies were used to determine the effects of reduced ECCS flew (s45 sp=) en the design base accidents. Since Sequoyah l
is an upper head injection plant, the hypothesized peak clad te=perature (PCT) increase of 40cF applies to the worst case small break if no operater action occurs until 10 minutes after the break initiation. How-ever, =anual closure of the =iniflew valves at the time the RCP's are f
tripped would result in a negligible increase in peak clad te=perature.
Since the PCT penalty is based en sensitivity studies done for non-UHI plants, a nu=ber of assumptions =ade do not apply to UHI plant s and indicate the existing FSAR s=all break analysis =ay be bounding or close enough so to be insignificant. These assumptions include:
(1) A 20-percent reduction in HPI ficw - Since Sequoyah has both high head and inter ediate head SI pu=ps, the 20-percent reduction in HPI flow applies only to the high head pu=ps so t'.at the overall HPI flow c
reduction is en the ceder of 10 percent (or less for worst case s=all break).
(2) No upper head injection ficw assu=ed - This additional flew would act to delay the core uncovery time resulting in clad heatup at a lower pcwer level caused by the decay in residual decay heat and result in a lower peak clad te=perature increase.
In addition to these assumptiens, the s=all break analysis Westinghouse utilized assu=ed a core power peaking factor of 2.32.
Sequoyah is presently limited by technical specification to a peaking facter of 2.237 adding an additional =argin in PCT.
If, disregarding the above arguments, the bounding PCT increase of 40cp is assumed for the Sequoyah worst case break (8"), the resulting PCT would go frc= the FSAR valve of s1490 F to s,1530 F.
The PCT is still well belcw the worst case large break valve of 21900F.
Positien indicatien for the CCP =iniflow valves is provided by lights on the handswitches and separate status lights en the =ain control panel.
Since the proper positien of this valve it dependent en syste= pressure, the valve position alarm ryste= described in section 6 3.2.2 of the Sequoyah FSAR cannot be applied to these valves. It is our opinion that the redundant valve position indication along with explicit instructicns in the E=ergency Operating Instructicns for =anual operation of these valves are sufficient to ensure proper valve position in the event of a safety injec tion. The final =odification will address the need for a permanent valve positien alar: syste=.
We believe that the propcsed interi: =cdification does not adversely effect th consequences of the accidents analyzed in the FSAR and does not con-stiwute a significant hazard censideration.