ML20207J656

Proposed Tech Specs,Revising Limiting Condition for Operation 3.2.2 & Surveillance Requirement 4.2.2.2 to Reflect Reduction in Heat Flux Hot Channel Factor Limit from 2.237 to 2.15

ML20207J656
Person / Time
Site:	Sequoyah
Issue date:	09/21/1988
From:	TENNESSEE VALLEY AUTHORITY
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NUDOCS 8809280073
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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNIT 1 DOCKET NO. 50-327 (TVA-SQN-TS-88-28)

LIST OF AFFECTED PAGES Unit 1 3/4 2-5 3/4.'-6 3/4 2-7 d

8809280073 86092}27 PDR ADOCK 050003 P

PDC 1

r POWER OISTRIBUTION LIMITS 3/4.2.2 HEATFLUX-HOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION i

3.2.2 F (Z) shall be limited by the following relationships:

9 4./5" F (Z) 1 [e-9-3+2 [K(2)] for P > 0.5 4

q P

a./S' i

F (Z) 1 [P-fW) [K(Z)] for P 10.5 9

0.5 THERMAL POWER w ere P =

RATED THERMAL POWER i

and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1 n23 ACTION:

7 With F (Z) exceeding its limit:

q

• ~

Reduce.aERMAL POWER at least 1% for each 1% F (Z) exceeds the limit a.

q l

within 15 minutes and similarly reduce the Power Range Neutron 1

j Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 7? hours; subsequent POWER OPERATION 3

may proceed provided the Overpower Delta T Trip Setpoints (value of K ) have bean reduced at least 1% (in.iT span) for each 1% F (2)

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exceeds the limit.

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b.

Identify and correct.the cause of tr.e out of limit condition prior to increasing THERMAL POWER; THERMAL. POWER may ther. be it.crissed provideo F (Z) is demonstrated *?. ough incrc mapping to be within l

q l

its limit.

$URVE!LLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

w Omrt;r 23,1^Q SEQUOYAH $ UNIT 1 3/4 2-5 f = d ;r,t h w 5

e 4

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POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 1 4.2.2.2 F (z) stia11 be evaluated to determine if F (Z) is within its q

9 limit by:

a.

Using the movable incore detectors to obtain a power distribu-tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Increasing the measured Fggg) componsnt of the power distribution map by 3 percent to a.ccount for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c.

Satisfying the following relationship:

2./f F "(z) < P g W(z)* U 2) for P > 0.5 Q

i F "(z) < W(z) x 0.5) for P < 0.5 Q

where F"(z) is the measured F (z) increased by the allowances for 9

manufacturing tolerances and measurement uncertainty, F limit is q

the F limit, K(z) is given in Figure 3.2-2, P is the relative q

THERMAL POWER,and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.14.

N d.

Measuring Fq (z) accort ng to the following schedule:

1.

Upco achieving equilihefum conditions after exceeding oy 30 porcent or more cf RATED THERMAL POWER, the THERftAL POWER at which F (z) was Inst determined,* or q

2.

At Itast ence per 31 effective full power o,,s, whichever occurs first.

"During power escalation at the beginning of eact, cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

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SEQUOYAH - UNIT 1 3/4 2-6 s c m,t-A -t9-

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-.-egrem,me+-r&q y-=--

yN--=1-

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.O POWER DISTRIBUTION LIMITS

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1 SURVEILLANCE REQUIREMENTS (Continued) e.

With measurements indicatin f

F0 (2) maximum k

K(z) I over z has increased since the previous determinatin of F "(z) either 0

of the following actions shall be taken:

M(z) shall be increased by 2 percent over that specified in 1.

F 9

4.2.2.2.c, or 2

F M(z) shall be measured at least once per 7 effective full g

power days until 2 successive maps indicate that F" (z) maximum is not increasing.

R23 K(z) over I

(

j f.

With the relationships specified in 4.2.2.2.c above not being satisfied:

1.

Calculate the percent F (z) exceeds it' limit by the following expression:

9 1l r

H Fn (z) x W(z)

I

-1 x 100 for P ), 0. 5 Imaximum S' over z 2 0:7 x K(z)

M /I M

Fn (z) x W(z)

_1 x 100 for P < 0.5 maximum l

1

,y M-over z O.T'

• KC3)

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2.

Either of the following actions Shell be taken:

a.

Olact the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied.

Power level may then be increased provided the AFD limits of Figure 3.2-1 are reduced 1% AfD for each percent F (z) exceeded its limit, q

or b.

Comply"with the requirements of Specification 3.2.2 for F (2) exceeding its limit by the percent calculated above.

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SEQUOYAH - UNIT 1 3/4 2-7 5.N 5 N t Eb.' b

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ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE r

SEQUOYAH NUCLEAR PLANT UNIT 1 DOCKET NO. 50-327 (TVA-SQN-TS-88-28)

DESCRIPTION AND JUSTIFICATION FOR REDUCTION IN HEAT FLUX HOT CHANNEL FACTOR LIMIT 6

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ENCLOSURE 2 Description of Change Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant Unit i technical specifications to revise limiting condition for operation 3.2.2 and surveillance requirement (SR) 4.2.2.2 to reflect a reduction in the heat flux hot channel factor (Fq[z]) limit from 2.237 to 2.15.

Season for Change By letter dated August 15, 198R, TVA submitted proposed license amendment 88-20.

This proposed change revised the upper head injection (UHI) isolation setpoint and tolerances of SR 4.5.1.2.c.1. of the August 15 letter describes that, as part of the setpoinc change, the delivered UH1 water volume band was being expanded from the range of 1,130.5 to 900 cubic feet to the range of 1,130.5 to 850 cubic feet.

The change in the delivered UHI water volume band was supported by Wastinghouse Electric Corporation (W) evaluations, which indicated that the potential decrease in delivered water volume to the core would result in increased peak clad temperatures (PCTs); but in all cases, PCT remained below the 2,200 degree Fahrenheit (F) limit of 10 CFR 50.46.

In telephone conversations on September 1 and 2,1988, NRC informed TVA that the increased PCTs described in the August 15, 1988 submittal could not be wholly justified by the sensitivity studies provided. NRC stated that restart of unit 1 could be supported by the sensitivity studies (provided a temporary exemption to certain administrative requirements of 10 CFR 50.46(a)(1) was obtained) and that operational restriction be imposed to provide at least 100 degrees F of margin between the calculated PCT and the 10 CFR 50.46 limit.

TVA's request for a temporary exemption to certain administrative requirements of 10 CFR 50.46(a)(1) will be transmitted by separate correspondence.

Evaluations by E have determined that at least 100 degrees F PCT margin can be obtair.ed by administrative 1y limiting steam generator tube plugging to 5 percent and by reducing Fq(z) from 2.237 to 2.15.

This proposed technical specification change is being submitted to reflect the reduction l

in the Fq(z) limit.

l Justification for Change As defined in'SQN Final Safety Analysis Report (FSAR) section 4.3.2.2.1, Fg(z) is the maximum local heat flux on the surface of a fuel rod divided by the average fuel red heat flux.

Limiting this ratio minimizes the magnitude of localized "hot spots" along the fuel cladding surface.

This in turn helps ensure that PCTs will remain below the 10 CFR 50.46 limit of 2.200 degrees F during postulated loss of coolant accident (LOCA) gonditions.

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The proposed reduction in Fq(z) is a conservative change and will provide additional margin in PCT.

As described in the attached y evaluation (page 4), a reduction in Fg(z) from 2.237 to 2.15 reduces PCT by 87 degrees F for the limiting imperfect mixing case and by 96 degrees F for the limiting pdrfect mixing case. As summarized on page 5 of the evaluation, this PCT reduction, combined with the reduction obtained by administrative 1y limiting steam generator table plugging to 5 percent, i

results in PCTs of 2,089 degrees F for the limiting imperfect mixing case and 2,067 degrees F for the limiting perfect mixing case. As can be seen, these PCT values provide over 100 degrees of margin to the regulatory PCT l

1 limits.

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September 14, 1988 Westinghouse PowerSystems Electric Corporation NVC!ttf I M hn0lCgy Sr*$ Daca b 315 PmsutgiPeesyNan!a 15m 0355 Mr. P. G. Trudel

'IYA-88-761 Sequoyah Project Engineer NS-OPLS-OPL-II-88-572 Tennessee Volley Authority Ref.1) TVA RD #428873 Sequoyah Nuclear Power Plant, DSC-A

2) W G.O. CO-42680 P. O. 2000

3) TVA-88-746 Soddy Daisy, TN 37379 TDINESSEE VALLEY AUTHORITY SEQUOYAH UNITS 1 & 2 CECREASED UHI VOLUME DELIVERY LOCA SAFETY EVALU (SECL-88 1417 Revision 1)

Dear Mr. Trudel:

In accordance with our telecon of Septcrtber 7,1988, the LOCA safety evalua SGTP, and a supplemental inforr.ation doeur.cnt is be NRC request for additional infomation addressing the LOCA models refere clarification of the appropriate limiting breaks, and clarification of the effect of the postulated instrtcentation thimble and guide tube flexuro failurcs.

The revised LOCA safety evaluation, SECL-68117, Revision 1, entitled, Safe 4

Evaluation for a 50 Cubic Feet Decrease in the UHI Acomulator Volu:ne (LOCA, SOTR, Post-LOCA Long Tem Core Cooling and Hot Leg Switcho Accident), is attached.

This revision incorporates the impact of reducing F(Q) frca 2 32 to 2.15 and the Steam Generator Tube Plugging (SGTP) level tecn 10 55.

The supple:nental interv.ation doctment is also attached and is entitled Supp Infontation to SECL-88-417, Revision 1.

If you have any ecccents or questions, please contact the undersisned.

Vory tettly yours, WESTINGHOUSE ELECTRIC CORPORATION

/

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. A.

rdi, Manager ESSD Projects L. V. Tomasic/tu Hid-South Area Attach:r.cnt cc.:

D. W. Wilson W. R. Mangiante S. J. Smith R. W. Headows J. A. Vcgel M. J. Bur ynski R. C. Weir R. G. Davis M. J. Ray R. E. Ibniels i

SECL NO:_SECL-88-417 Rev, i Cucttmar R3forcnco No(o).

Westinghouse Ref. No.

WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIS"

1) NUCLEAR PLANT (S) SEOUOYAH UNITS 1 AND 2 (TVA/ TEN)

2) CHECK LIST APPLICABLE To: SAFETY EVALUATION FOR A 50 CU.FT. DE (subject of Change)

_THE UNI ACCUMULATOR DELIVERABLE WATER VOLUME 3)

The written safety evaluation of the revised procedure, design change modification required by 10CFR50.59 has been prepared to the extent or required and is attached.

If a safety evaluation is not required or is inconplete for any reason, explain on Page 2.

"s-Parts A and B of this safety Evaluation Check List are to be comploted only on the basis of the safety evaluation performed.

CHECK LIST - PART A (3.1) Yes X

No

_AchangototheplantasdescribedinthehSAR'?-~,

(3.2) Yes No X _ A change to procedures as described in tho FSAR?

(3.3) Yes No X

A test or experiment not described in the FSAR?

(3.4) Yes_ X No A change to the plant technical specifications (Appendix A to the operating License)?

4) CHECK LIST - PART B (Justification for Part B answers must be included on Page 2.)

(4.1) Yes No X _ Will the probability of an accident previously evaluated in the FSAR be increased?

(4.2) Yes No X

Will the consequences of an accident previously evaluated in the PSAR be increased?

(4.3) Yes No X

May the possibility of an accident phich is different than any already evaluated in tho,

FSAR be, created?

(4.4) Yos_.

No X'

Will the probabi.lity_of a malfunction'of equipment important to safety previously evaluated in the FSAR be increased?

i (4.5) Yen

__ Ho__X__ Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

(4.6) Yes No X

May the possibility of a malfunction of equipment j

important to safety different than any alrnady evaluated in the FSAR be created?

(4.7) Yen Ho X

Will the margin of safety as defined in the bases to any technical specification be reducod?

PAGE 1 OF 2

SECL-88-417 Ravicion 1

, a-If the answers to any of the above questions are unknown, indicate under 5) REMARKS and explain below.

If the answer to any of the above questions in 4) cannot be answered in the

negative, based on written safety evaluation, the change cannot be approved without an application for license amendment submitted to NRC pursuant to 10CFR50.90.

5) REMARKS:

The following surunarizes the justification upon the writton safety evaluation, (1) for answers given in Part B of the Safety Evaluation Check List:

seo the attachment N.

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(1) Roference to document (s) containing written safety evaluationt 11S -S AT-S AI-8 8 - 3 6 2 FOR FSAR UPDATE Section:--

Pago (s) :

Tabl e (s ) : __15. 4.1-9 Reason for/ Description of Change:

_ Chance Table 15.4_.19 for UHI AccuM>1ator_ water volume delivered to reflect 850 cu.ft, mininum volume evaluated in this safety evaluation and the associated footnote. --

6) APPROVAL LADDER (6.1) Prepared by (Nuclear Safety):

4^- ( S AI i Date:

/

b Roviewed by (Nuclear Safety):.7//. 8. 23dadu ISAI)

(6.2) Coordinated with Engineer (s)D A/O 86V/GWO(SATI) Datos_f/h'/88

_Dato:

Coordinated vith Engineer (s):

A/6d N M M(TSA1_.Date:

Coordinated with Engineer (s):

PBon//OO4 AF-(coal _Date s Coordinated with Engineer (s):yPPoud<- 5 ruasAI)

Dato u-(6.3) Coordinating Group Managor(s) l A PPuc_a:;. s/W4%AII) Dates coordinating Group Managor(s) )CULY Ml4C (TSA)

Date:

Coordinating Group Managor(s)

MAL lOCA' coa /6Date [.M4A/CE~D_

(6.4) Nuclear Safety Group Managers P h b b d d /SA1) Date: 'VN/8A

{

l pAGE 2 0F 2 r

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WESTINGHOUSE PROPRIETARY CLASS 2

,y SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DE IN THE UHI ACCUMULATOR DELIVERABLE WATER VOLUME BACKGROURQ In order to accommodate relaxed UHI system tolerances at Sequoyah Units 1

and 2,

Tennessee Valley Authority (TVA) has requested a widened set limits on the allowable UHI water delivered volume.

of Specifically, a decrease in the required minimum UHI delivered water volume considering uncertainty from 900 to 850 ft3 has been requested.

The following presents the summaries of 3

safety evaluations performed to assess the effect of a

50 ft reduction in the minimum UHI delivered water volume on the LOCA-related analyses performed by Westinghouse for Sequoyah Units 1 and 2.

N._

BASES LARGE BREAK LOCA - FSAR CHAPTER 15.4.1

~~

~

In order to accommodate relaxed UHI system tolerances at Sequoyah Units 1

and 2, TVA has requested a widoned set of limits on the allowable UHI dalivered volume.

To this ond, the Sequoyah Large Break ECCS water performanco analysin has been reviewed required minimum to justify a decrease in the 850 ft3 UHI delivered water volume considering uncertainty to The limiting case break in the UHI Evaluation Model EccS analysin presented in the original sequoyah FSAR was the C =0.6 DECLG break with imperfect mixing of D

UHI water assumed in tho vasool uppor head.

Comp 11anco with regulatory limits was achieved for this case b y..

reducing the allowable core peaking factor (F ) from 2 32 to 2.237.

Minimiting the volume of UHI wator._ delivered is conservative for f

q imperfect mixing UHI LOCA cases.

The lower bound value for UHI water volume delivery established in the original FSAR C =0.6 DECLG imporfect mixing caso is 900 ft.

3 9

Thin value also was amployed in the imperfect mixing cases of the 10% steam generator tube plugging I

(SGTP) analysis performed in the 1982-83 timeframe and reported in ROferenco 1.

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WESTINGHOUSE PROPRIETARY CLASS 2 N

SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DEC IN THE UNI ACCUMULATOR DELIVERABLE WATER VOLUME A

complete spectrum of perfect mixing cases wa s.

analyzed for the original Sequoyah FSAR.

The limiting case with perfect mixing of UHI water assumed in the vessel upper head was the C =0.6 DECLG; the calculated peak clad temperature (PCT) for this case is 2111'F at an p

F of 2.32 with a

UHI delivered water volume of 1053 f t.* Using q

3 sensitivities appropriate to UHI plant perfect mixing cases, trade-offs have previously been made among various input assumptions to justify increasing the maximum allowable UMI delivered water volume to 3

1130.5 ft.

Maximizing the value of UHI water delivered is conservative for perfect mixing UHI LOCA analyses.

With a Technical Specification F

of 2.237 in

force, 1130.5 ft is a valid maximum q

3 delivered water volume for the sequoyah UNI system; the calculated -PCT of the limiting C =0.6 DECLC perfect mixing caso at 1130.5 ft3 p

water delivery is 2163*F.

UHI

~

The Cp=0.8 and C =0.6 DECLG imperfect mixing cases from the 1982-83 D

10%

SGTP analysis havo been reviewed to assess the PCT impact of reducing the delivered UHI water volume to 850 ft3 The calculated Reference 1

PCTs for the C =0.8 and C =0.6 DEcLG cases which p

D compriso the current licensing basis for Sequoyah are 2111'T and 2113'F, respectively.

Reducing the UHI water delivory in an imperfect mixing case will reduco L

the couling of the fuel as the uppor head drains during blowdown.

During the core reflood phace, thin hottor fuel will cause tho axpulsion of moro injection water as entrained

liquid, producing a

degraded flooding rate.

Existing Soquoyah imperfect mixing cases defino the penalty in coro fuel heat-up associated with decreasing UMI water delivery to 850 ft.

Expressed 3

es a

flooding rate

penalty, roducing UHI watsr delivery to 850 ft3 reduces core inlot, velocity mixing cases.

by; 7% for the licensing basis icporfect The effect of degraded flooding rates upon hot rod calculated PCT has been dotermined by WREFLc0D/LOCTA sensitivity runs.

Expressed as an imperfect mixing case PCT sensitivity relationship, a

one ft3 decreaco in UNI water delivery results in approximately a l'F incresco in calculated PCT.

The 10% SGTP licensing basia imperfect mixing cason cro acceptable at 3

an 850 ft delivered UH1 water volume becauco the degradod reflood penalty only increases calculated PCT as followst Page 2 e.-

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WESTINGHtOSE PROPRIETARY CLASS 2 SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DECRE IN THE UHI ACCUMULATOR DELIVERABLE WATER VOLUME C =0.8 DECLG PCT =2151*F D

C =0.6 DECLG PCT =2166*F D

A separate safety ' evaluation performed for Sequoyah during November, 1986 considered the impact of a possible but unlikely scenario of guide tube flexure failures.

The PCT penalties imposed upon perfect and imperfect mixing cases under this scenario are O'F and 20'F4 respectively.

The not PCT DECLC case therefore becomes for the limiting imperfect mixing C =0.6 D

2166'T + 20'F = 2186'T whhn-postulated guide tube flexure failures are considered.

t A

further phenomenon which could impact the Sequoyah Plant calculated-PCT values is filling of the instrumentation thimbles in the core ~

during the reflood phase of a

large break LOCA event.

The thimble volume which must be filled has not been explicitly treated in the Sequoyah large break LoCA analyses.

Westinghouse had initially assumed that the thimble plugging devices would be sufficiently tight to prevent the ingress of water into thimbles during reflood.

However, it was later identified that the plug clearances were sufficiently large i

and the flows were sufficiently low during reflood to allow the thimbles to fill with water even with plugs installed.

The impact ~

which thimble filling will exert on the calculated PCT values has been casossed for Sequoyah, and the appropriate PCT penalties to be imposed on the perfect imperfect mixing casen are established as O'F and and 12'F.

The not calculated PCT for the limiting imperfect mixing cane I

becomes 6

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{

2186'T + 12'F = 2198'T co compliance with the regulatory limit is maintained.

1 j

Ao previously

noted, the imperfect mixing cases of Reference 1 have been performed assuming 10% uniform steam generator tube plugging.

The i

cetual SGTP level at Sequoyah Units 1 and 2 is less than 3% in all oteam generators.

In performing the 1982-83 imperfect mixing analyses, WCstinghouse identified the sensitivity relationship between calculated PCT and assumed SGTP level.

For the limiting C =0.6 DECLC imperfect D

nixing ' case, increasing the SGTp level from 2% to 10% produced an increase of 36'T in calculated PCT to the 2113'T value.

Page 3 1

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WESTINGHOUSE PROPRIETARY CLASS 2

• ~

SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A D IN THE UHI ACCUMULATOR DELIVERABLE WATER VOLUME Taking credit for the excess margin in SGTP will permit the not calculated PCT for the Sequoyah Units 1 and 2 limiting imperfect mixing case to be reduced.

To retain some margin for the possible future plugging of steam generator tubes, revise the SGTP basis to reflect a 5%

uniform plugging level.

Then the calculated PCT is reduced by 5/8 (36'F),

or 22'F, for the limiting C =0.6 DECLG imperfect mixing case, giving a new not calculated PCT of D

2198'T - 22*F = 2176'T The NRC Staff has specified that at least 100*F margin to the regulatory limit of 2200*F should bo demonstrated in this safety e)aluation to justify Sequoyah Unit 1

Cycle 4 operation.

Further benefits must therefore be found to reduce both the perfect and imperfect mixing limiting case calculated PCT values.

The approach-taken is to maintain 100% power operation while reducing the technical

~

specification allowable maximum peaking factor (Fq) value.

The sensitivity relationship between calculated PCT and Fq has been ostablished for both imperfect and perfect mixing UHI evaluation model analysec.

For UHI impe:"act mixing casos, the sensitivity relationship is 10*P decrease in calculated DCT 0.01 decresse in Fq value For Sequoyah Unito 1 and 2, decrease the Fq value from 2 237 The corresponding decreano in calculated PCT is (0.087) (10'F/0,01) to 2.15.

87'F, giving a

net calculated

=

limiting C =0.6 DECL4 cane.

PCT of 2176'F - 87'F = 2089'F ter the~

p For UHI perfoot mixirg caoos, tne equivalant sensitivity has also been octablished.

The sensitivity rolationship for perfoct mixing in ll'F decreaso in calculated pct 0.01 decrease in Fq value Therefore, the reduction in Fq to 2.15 will produce a PCT benefit of (11*F/0.01)

(0.087)

=

96*F for Sequoyah perfect mixing cacou.

The limiting c =0.6 DECLG perfect mixing case exhibits a calculated PCT o

of 2163 - 96 a 2067'T at an Fq value of 2.15.

Page 4

WESTINGHOUSE PROPRIETARY CLASS 3 SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DECREA IN THE UMI ACCUMULATOR DELIVERABLE WATER VOLUME To summarize, by reducing the technical specification limit on total peaking factor (Fq) to 2.15 additional peak clad temperature margin is obtained for Sequoyah Units 1

and 2.

The net calculated PCT values become 2089'F and 2067'T for the imperfect and perfect mixing limiting cases respectively; greater than 100*F margin in PCT is available to the regulatory limit of delivered water volume band of 850 -2200'F for Segacyah Units 1 and 2 with a UHI 1130.5 cu. ft.

SMALL BREAX LOCA The Current FSAR small break LOCA analysis for Sequoyah Units 1 and 2, was performed using the NRC-approved UHI Small Break LOCA ECCS Eviluation Model (Reference 2), which resulted in the most limiting PCT of 1486'T for a

8 inch equivalent diameter break (Reference 1).

A safety evaluation which considered the ef fect of charging /SI pump flow-shortfall increased this result by 200*F, resulting in an overall licensing basis PCT of 1686'F.

The reason the 8 inch break is limiting and exhibits a low PCT value is because UHI provides enhanced safety injectica capability relativo to otandard plant systema.

Typically, 4 loop plants demonstrate the throo or four-inch equivalent diameter break to ho limiting at a higher calculated PCT than the Sequoyah valuo.

Becauso UHI injection is inherently beneficial for the evaluation model r, mall break LocA evout, the Sequoyah FSAR analysis has assumed a conservatively lov value for ~

the minimum UHI doliverable water volumo which is loss than 850 cu.

ft.

Thorofore, a decrease in the delivorablo UHI water volume to 650 cu. ft. does not adversoly affect the FSAR small break LOCA results.

ROD EJECTION HASS AND ENERGY R2 LEASE FOR DOSE CALCULATION - FSA i

CHAPTER 15. 5. 7 AND TdBIE 15. 5. 7 3 Similar to a

small break

LOCA, a rod ejection accident analysin is parformed to provida primary and secondary mass and energy releases for uce in computing the radiological consoquences of a

rod ejection accident as por Regulatory Guido 1.77.

This analysis is a long term trannient performed specifically to dete rmine primary RCS hass and energy releanos through the upper head break and secondary maus and Cnergy releases via the secondary code safoty valves.

These mass and energy releases aro then upod to computo the radiological consequences Pago 5

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WESTINZHLUSE PROPRIETARY CLASS 2

.=

SECL-88-417 REVISICH 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DECRE!3E IN THE UMI ACCUMULATOR DELIVERABLE WATER VOLUME of a rod ejection accident.

A reduction in the minimum deliverable UMI water volume will result in an increase in the mass and energy releases of the primary coolant and a decrease in the secondary mass and enorgy releases.

However, for the Sequoyah Units, a conservative assumption was used regarding the primary mass and energy releases such that the not effect of a 50 ft3 reduction in the minimum deliverable UHI water volume is a

reduction in secondary mass and energy releases.

Since a net reduction in secondary mass and energy releases would slightly reduce the computed

doses, the current doses as reported in Table 15.5.7-2 of Reference i remain bounding.

CONTAINMENT INTEGRITY (SHORT AND LONG TEKM MASS AND ENERGY RELEASE FFAR-CHAPTER 6.2 The containment analyses for the Sequoyah Units are described in FSAR- '

sections 6.2.1.3.3, 6.2.1.3.4, 6.2.1.3.6.

and 6.2.1.3.11.

Theno ~

sections

consider, respectively, containment pressure transients for long and short term analyse 4 mass and energy releases for postulated LOCAs and containment subcompertments, and containment maximum temperature renponse following a main oteamline break.

For the containment subcompartment analyses and the short term mass and Cnergy relcare analyses no modelling of tne UHI accumulator is included.

Therefore, a

50 ft3 veduction in the minimum delivorable water volume to 850 ft3 will have no effect on the current analyses.

~

The long term me r.s and enorgy rolosse analysin is performed to calculate the maximum availa'ola releasos which can enter the containnant foll owing a LOCA.

Simil.nr to the subcompartment analyses, no modelling of the UHI accumulator is included therefore, a minirum deliverable UHI waher volunk of'650 ft3.will.not effect the long term mass and energy releases to contaihmenc.

The evaluation for main steamline break concluded that there would be no change in the mass and energy releases to the containment for a reduction in minimum deliverable UHI water to less than 850 ft3 Therefore, the containment maximun temporaturo response following a tain steamline break will not be effected.

Hence, based upon the above internation, the results of the current Chaptor 6.2 containment Integrity analyses continue to be valid.

Pago 6 we,

..r

.v..a r.it WESTINGHOUSE PROPRIETARY CLASS 2

^

SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DECR IN THE UMI ACCUMULATOR DELIVERABLE WATER VOLUME STEAM GENERATOR TUBE RUPTURE - FSAR CHAPTER 15.4.3 The steam generator tube rupture event as analyzed in the Sequoyah FSAR equilibrates in pressure at a value which greatly exceeds the maximum UHI nitrogen gas pressure of 1300 psia.

Since the UMI system is not actuated during a design basis steam generator tube rupture event, any change in UHI delivered this analysis.

water volume upon actuation is irrelevant to BLOWDOWN REACTOR VESSEL AND LOOP FORCES - FSAR CHAPTER 3.9 N. m The blowdown hydraulic forcing functions resulting from a loss of coolant accident are considered in Section 3.9.1.5 (Analysis Nethods-

~

Under LOCA Loadings), and Section 3.9.3.5 (Blowdown Forces Due-to. Cold' and Hot Leg Break) of Volume 4 of the Sequoyah Units 1 and 2 FSAR.

decrease The in the UHI accumulator water volume will have no effect on the i

LOCA blowdown hydraulic loads since the maximum loads are generated within the first few tenths of a second after break initiation.

this reason the ECCS, including the UHI accumulator, is not considered For in the LOCA hydreulic forces modeling and thus the decrease in the UHI accumulator water voluno will have no of fect on the resu?.ts of the LOCA hydraulic forces calculations.

POST LOCA LONGTERM CORE C00L7NG SUBCRITICALITY REQUIREMENT; WESTINGHOUSF LICENSING POSITION - 7SAR CHAPTER 15 4.2 i

The Westinghouse licensing position for satisfying the requirements o 100FR Part 50 sect' ion 50.46 Paragraph..(b)_ Item (5) "Long Tarn cooling" is defined in WCAP-8339 (Reference 4,

pp. 4-22).

The Westinghouse commitment is that the reactor Will remain shutdown by boratnd ECCS water residing in the sump following a

LOCA (Reference 5).

Sinco i

credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the RWST and Accumulators must have a concentration that, when mixed with other sourcos of water, will result in the reactor core remaining suboritical assuming all control roda out (ARO)3 The docrease in the minimum UHI water delivered voluno or 50 ft results in a reduction of approximately 1 ppm in the mixed 1

Page 7 i

e m -

.., -.....a r 4-.oo Ao; v r.12 4

WESTINGHIUSE PROPRIETARY CLASS 2 SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR IN THE UMI ACCUMULATOR DELIVERABLE WATER VOLUME mean sump boron concentration.

boron concentration This reduction in the mixed Aean sump the current cycles of operation for Sequoyah Unita 1 and 2.can b 1

NOT LEG SWITCHOVER TO PREVENT POTENTIAL CHAPTER

' 2.2 BORON PRECIPITATION - FSAR The het reoirculation switchover time analysis has been performed

• or Sr nits 1 and 2 to determine the time following a LOCA that kot 41stion should be initiated.

This analysis addresses the com precipitation in the rea? tor vessel following a LOCA 4

oren

' nt

ir performed to support the decrease of 50 cubic feet in the j

ator volume to 850 cubic feet.

Du.

a large break LOCA the plant switches to cold leg recircula lon-afti the RWST switchover setpoint has been reached.

in the cold leg If the break is there is a concern that the cold lag injection water will fail to establish flow through the

core, safety injection entering the broke.n loop will spill out the break, wn11e SI entering the intact cold legs will, circulate around the downcomer and out the break.

With no flow path established through the core the fluid in t no coro rkeajns stagnant.

Au stcan is produced in the core from decay

heat, the boron associated with the steam will renain in the vessel'.'

Thus,

.n w water is boiled off with no circulation present in the core the bocic acid concentration incraases.

The boron concentration in tho i

vessel will increase until the solubility limit of the boric acid colution in reached, at which time boron will begin to precipitate.

the boron cdversely affect the'ir heat transfer characteristics.precipitatus As The purpose of the hot leg recirculation switchover time analysis in to provide the time at which hot leg recirculation munt be established to prevent boron precipitation in the core.

An evaluation has been performed to determine the effect of tho reduction of the deliverablo voluno of water in the UllI accunulator on the hot 3cg recirculation switchover time.

This evaluation concluded that the time for hot leg switchover contained in the FSAR (13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />) is bounding Therefore, the valuo in the FSAR need not be changed.

Page 8

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.,._v

...s WESTINEHOUSE PROPRIETARY CLASS 2 SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUOYAH UNITS 1 AND 2 FOR A DECREAS IN THE UMI ACCUMULATOR DELIVERABLE WATER VOLUME CONCLUSIONS The effect on the 00CA related analyses for Sequoyah Units 1 and 2 of a 3

50 ft reduction in the minimum deliverable UHI water volume to 850 ft3 has been evaluated by Westinghouse.

The potential effect of the change on the FSAR analysis results for each of the LOCA-related accidents was evaluated, and it was shown in all cases that the effect of the chance did not result in exceeding any design or Regulatory limit.

Therefore, it is be concluded that the proposed decrease in the minimum deliverable UHI water to 850 ft" for Sequoyah Units 1 and 2 is acceptable from the standpoint of the FSAR accident analyses discussed in this safety evaluation.

Table 1 lists the effect of the change on the various accidents which are discussed here.

REFERENCES 1.

Sequoyah Station (TVA/ TEN) FSAR - Updated 6/16/86 Amendment 3.

2.

WCAP-8479 Rev.

2 (Proprietary),

WCAP-8400 Rev.

2 (Hon-Proprietary),

Young, H.Y.,

et.

al.,

"Westinghouse Emergency Core Cooling System Evaluation Model Application to Plante Equipped with Upper Head Injection", January 1975.

3.

WCAP-9220-P-A (Proprietary),

WCAP-9221 (Non-Proprietary),

Eiche1dinger, C.,

"Westinghouse ECCS Evaluation Model - 1981 Version", Revision 1, 1981.

4.

KCAP-0339 (Non-Proprieta ry),

Btrdslon, F.H.,

9t.

al.,

"Westinghouse Ecc3 Evaluation Model - Summary", June 1974.

5.

Westinghouse Technical Bulletin NSID-TB-86-08, "Post-LCCA i

Long-Term Coolingt Boron Requirements", October 31, 1986.

)

Page 9 l

t e

9

\\

L n

..e 4.

WESTINGHOUSE PROPRIETARY CLASS 2 SECL-88-417 REVISION 1 SAFETY EVALUATION FOR SEQUCYAH UNITS 1 AND 2 FO IN THE UHI ACCUMULATOR DELIVERABLE WATER VOLUME TABLE 1 FSAR CHAPTER ACCIDENT DESCRIPTION IFFECT ON RESULTS 15.4.1 Large Break LOCA

$gk P

clad temperagure is h$tk re$uc$ng"t$egoaS 100 by 56.46bf$b$f Ss 10C maintained.

15.3.1 Small Break LOCA No adverge gffectonthe{$nk SAR 1 t oxkhat$gn$o$

$ax:,mbm'$$ add or max:. gum hydroge

$bCfR50N6b(1-hfbah,$ta$ned.

15.5.7 Rod Ejection Accident No adversg effect on magn and h$tk kbC 100.1$

lE$$t!

maintained.hR 6.2 Containment Integrity No pdverse effect on short (Short and Long Term or ton torn mass and one Mass and Energy Release) geas vgs g

on Sn$ts u

thus 15.4.3 Steam Generator Tube Rupture No adverno effect on primary-g$$$to pa$ntaLned. g goc o-s g g ary g s 1

3.9 B1 an$wdownMoactorVassal No advprse off at an LOCA Loop Torcee aydrausicforcIngfunctions.

15.4.1 g

CA Longterm core g doog in g gs CA

$0bfR$0$$6b($fI 6.3.2.2 Hot to Switchover to PSAR ont-10CA hot lea Preven Potential Boron Prec;p tation MVite over time remains Dound ng.

Pago 10 O

h

'~

..........~ r

.,.v.

.v...

c.is ATTACHMENT! SUPPLT. MENTAL INFORMATION TO SECL-88-417 REVISION 1 1ARGE BREAX 14CA ANALYSIS The UHI Evaluation Mod 71 dates from 1978 (Reference 2), and it received NRC approval in NUREG-0297. On May 15, 1981 Westinghouse via Reference 3 that changes be incorporated in its ECCS requested evaluation models.'

The primary change being made was to implement the cladding swelling and rupture medals contained in NUREG-0630.

The NRC approved the changes in Reference 3 on December 1, 1981, 1

noting that the model "is in compliance with 10 CFR 50 for all PWR's currently being supplied with Westinghou'ne manufactured

{

Zircaloy clad fuel."

The changes in Reference 3 pertinent to the NUREG-0297-approved UMI model were then in.plemented to create a revised UHI Evaluation Model.

Perfect and imperfect mixing l

analyses were performed in 1982-83 for Sequoyah Units 1 and 2 utilizing the UHI Evaluation Model computer codes which contain the NUREG-0630 fuel and other models as approved in Reference 3.

Note that Reference 2 still constitutes the UHI Evaluation Model, except for the minor changes instituted per Reference 3.

)

The 1982-83 Sequoyah UNI analysis effort perfect mixing cases werc pertorned to augment the perfect mixing cases analyzed initially 4

for the sequoyah FSAR.

That 1978 analysis parforried in accordance With Reference

?

showed the limiting case with perfect mixing of UHI water assumed in the vassol upper head was the Co*0.6 DECLGt its calculated peak clad temperature (PCT) is 2111 F at an Fq of 0

2.32 for a UHI delivered water volume of 1053 cu. ft.

The perfect mixing cases performad in conformance With Reference 3 in 1982-83 l

Here less limiting than the earlier result; tho 1978 limiting PCT 0

value of 2111 F bout.ds the perfect mixing case results obtained using the revised UHI Evoluntion Model.

The 1970 ncforenca 2 result has continued as the reference case for sequoyah perfoot nixing caso performance evaluations, and it remains conservative relative to the Reference 3 UHI Evaluation Model results.

With a Technical Specification Fq of 2.237 in force, 1130.5 cu. ft. in a valid maximum delivered water volume for the Sequoyah UHI systemt the c,alculated PCT of the limiting C =0.6 DECLC perfect mixing D

case at 1130.5 cu. ft. UHI water delivery is 2163 F.

0 page Al

y..

m-,.,

,,_,,,,-,,,,, __ ;, y ny_ -

" Tho C =0.s and Co 0.6 DECIA itportcet cixing cacco from tho D

a

,. ~.1982-83 104 sGTP onalycio havo b3cn rCvicw:d to cocG00 the PCT impact.

of reducing the delivered UHI water volume te $50 cu. ft.

I The calculated PCTs for the C *0.8 and C =0.6 DECI4 imperfect D

D i

mixing

cases, which utilize the Reference 3 version of the UHI l

Evaluation Mode and are presented in Reference 1, equal 2111 F I

and 2113'F, respectively.

Reducing the UNI water delivery in an 3

imperfect mixing case will reduce the cooling of the fuel as the hyper head drains during blowdown.

During the core reflood phase, j

this better fuel will cause the expulsion of more injection water i

as entrained liquid, producing a degraded flooding rate.

Existing j

Sequoyah imperfect mixing cases define the penalty in core fuel i

heatup associated with decreasing UNI delivery to 850 cu. ft.

Expressed as a flooding rate penalty, reducing UHI water delivery i

to 850 cu. ft. reduces core inlet velocity by 74 for the licensing basis iriparteet mixing cases.

4

'l The impact of degraded floodi. g rates upon hot rod calculated PCT han been detemined by WREFLOOD/LOCTA sensitivity runs for each

{

licensing basis imperfect mixing case.

Expressed as an imperfect Mixing case PCT sensitivity relationship, one cu. ft decrease in

[

UHI water delivery results in approximately a l'F increase in

[

calculated PCT for the limiting C =0.6 DECLG casa; the PCT p

l sensitivity is even less severe for the C =0.8 DECLG case.

The p

i 10%

SGTp licensing basis imperfect mixing cases are acceptable at

{

an 850 cu.

ft.

daliveed UNI water volume because the degraded j

reflood penalty only increases calculated PCT as follows:

i t

j Cp=0.8 DECI4 PCT =2151 F 0

l C =0.6 DECLG PCT =2166 F 0

p i

~

It may further be nota that a

C =0.4 DECIA imperfect mixing p

i case was perfornid in 1982 using the UHI Evaluation Model as

)

modified per Reference 3.

On an equivalent input basis, the j

calculated PCT for the C)=0.4 DECLG imperfect mixing case is over 200 degrees less limiting than the above cases.

Thorofore, i

consideration of the C =0.6 and 0.5 DECLG cas..es is sufficient in p

i this ase.essment.

Page A2 I

i l

I e

s an ene amm

• ~ ~

' An indep:nd:nt caf0ty cycluatien p;rform d for S qucych during

., *

• N vcmb r, 1986 (R3ferenc0 4 OttCchtd) consid r0d tho impact of a possible but unlikely scenario of guide tube flexura failures.

The PCT penalties imposed upon perfect and imperfect roixing cases under this scenario are 0*F and 20 F, respectively.

This 0

highly improbable event affects the initial portion of the UHI delivery transient, and it no longer has any effect by the time that the UHI sys' tem is isolated.

Therefore its magnitude of impact is unaffected by a change in the UHI delivered water volumo to 850 cu.

ft.

The not PCT for the limiting imperfect mixing C =0.6 DECL4 case including this scenario becomes o

0 0

0 2166 F + 20 F = 2186 F when postulated guide tube flexure failures are considered.

A furtner phenomenon which could impact the sequoyah Plant calculated PCT values is filling of the instrumentation thimbles in the core during the reflood phase of a large break LOCA event.

The thimble volume which must be filled has not been explicitly treated in the sequcyah large break LOCA analyses.

Westinghouse had initially assumed that the thimblo plugging devices would be sufficiently tight to prevent the ingress of water into thimbles during re f) cod.

However, it was later identified to the PRC Stuff that the plug clearancos waro sufficiently largo and the flows sufficiently low during reflood to allow the thimbles to fill were with water even with plugs installed.

The impact which thimble filling will exert on the calculated PCT values has been assessed for

sequoyah, and the appropriate PCT penalties to be imposed on the perfect and imperfect mixing cases are established as 0 0 0

and 12 F.

The imperfect mixing penalty equals the maximum value previously identiflod by Westinghouse in WCAP-9561-P-A, Addondur 3 i

because its PCT occurs during the traditional cora reflood phase.

The net calculated PCT for the limiting C =0 6 DECLC imperfect D

mixing case becomes l

0 0

0 2186 F + 12 F = 2198 F Page A3 l

1 l

i

c- - ~.

.... ~ ~ up s-.op so..a r.io and compliance with the regulatory limit is maintained.

The perfect mixing case calculated PCT occurs at the very inception of the core reflood phase, so it is not subject to a thimble filling penalty.

Both the perfect and imperfect mixing cases of the sequoyah large break loCA analysis remain in compliance with 10CFR50.46 if the UMI water delivered volume is set within the bounds 850 - 1130.5 cu.

ft.,

with calculated PCT values of 2163 F anft 2198 F 0

0 respectively.

CMALL BREAR loCA The Current FSAR small break ICCA analysis for Sequoyah Units 1 and 2, wcs performed using the NRC-approved UNI Small Break 14CA ECCS Evaluation Model (Reference 2), which resulted in the most limiting PCT of 1484'F for a 8 inch equivalent diameter break (Reference 1).

A cafety evaluation which considered the effect of charging /8I pump flow chortfall increased this result by 200'F, resulting in an overall licensing basis PCT of 1586'F.

Wostinghouse Letter NS-TMA-2147, dated November 2, 1979, responded to NRC concerns related to the fuel rod models used in the Westinghouse ICCS evalu*ation models, subsequently, in December,1979 letter j

NS-TNA-2174 identified a metho2 ology by which the impact of NUREG-0630 fuel rod models on existing analyses could be determined.

This opproach was employed to justify continued operation of Westinghouse plante until MUREG-0630 fuel rod models were implemented into the l

W stinghouse cosputer codes.

Note that in applying this *ethodology Wootinghouss was only required by the NRC Staff to examine the limiting l

cace break; in all plant analyses this was a large break IOCA case.

N0vertheless, it can be concluded that the Reference 2 model results remain applicable for sequoyah.

As indicated above, the not licensing bcois calculated FCT is icss*F for the eight-inch equivalent diameter cold leg break.

Fuel red burst le predicted to occur in not only the cight-inch break but also in the six-inch break case analyzed for i

sequoyah using Reference 2 methods.

Because rod burst has been predicted to occur in the analysis of record, applying the NUREG-0630 fuol red models would not produce a major change in the reported results.

since the magnitude of the change is minor and copious PCT nargin exists, the scall break LOCA remains non-limiting for sequoyah, and cocpliance with 10CFF50 is assured.

G

.ne

.c.

ne

..e

.n.,

o,REFEREiiCES

,7 1.

Sequoyah Station FSAR - Updated 6/16/86, Amendment 3.

2.

WCAP-8479 Revision 2

(Proprietary),

WCAP-8480 Revialon 2

(Non-Proprietary),

Young, M.

Y.

at, al.,

"Westinghouse Emergency Core Cooling system Evaluation Model Application to Plants Equipped with Upper Head Injection", January, 1975.

3.

WCAP-9220-P-A (Proprietary),

WCAP-9221 (Non-Propriettry),

Eicheldinger, C.,

"Westinghouse ECCS Evaluation Model - 1981 version", Revision 1, 1981.

4.

Westinghouse SECL-86-461, "Impact of Guide Tube Flexure Cracking on ECCS Performance, Sequoyah Units 1 and 2.

5.

WCAP-9561-P-A, Addendum 3,

Young, M.

Y.,

"Addendum to:

BART-Alt a

Cocputer Code for the Best Estimate Analysis of Reflood Transients (Special Report:

Thimble Modeling in Westinghouse ECCS Evaluation Model), 1986.

6.

NUREG-0297, "Safety Evaluation Report on Wantinghouse Electric Company ECCS Evaluation Model for Plants Equipped with Upper Head Injection," April 1978.

7.

NS-TMA-2147, Letter from T. M. Anderson, Westinghouse to D.

G. Eisenhut, NRC, dated November 2, 1979.

8.

NS-TMA-2174, Letter from

.T. M. Anderson, Westinghouse to D.

G. Eisenhut, NRC, dated December 7, 1979.

Page AS e

i

ACCL-S G - % /

SUPPL.EMENTARY INFORMATION I

The reference documents the possib8.11ty of guide tube flexure failure at sequoyah.

In the extressly unlikely event that AIL four flexures of 1

one particular 15X15 guide tube containing an insert were to fail, the insert is calculated to lift off the housing plate upon which it resides i

during a large break 14CA event.

It may either fall directly downward back into piece mz twist and become suspended on the broken flexure stubs er be expelled completely from the guide tube assembly, increasing the flow area between the upper head and upper plenum regions.

and/or hollow support columns provide flow communication from the core Guide tubes outlet to the upper head for 185 of the 193 sequoyah fuel assemblies.

j housing plate in only millisecondst no espect of the large br The transient can be affected in the slightest during such a trivial length of time.

The location of the insert when it falls will determine if the calculated large break IccA performance is at all affected.

back into place, there is no impacts other identifiable configurations all If it falls involve an increase in flow communication with the upper head region.

i greatest impact will occur if the insert is missing completely from theThe gui'de tube in question.' The scenario in which a 15X15 guide tube insert j

et sequoyah disappears at the inception of the postulated large break LOCA ovents is evaluated below.

i As documented in the sequoyah FSAR, cases modeling perfect and imperfect mixing of UNI water in the reactor vessel upper head must both be analyzed.

ossumption will be limiting for Sequoyah with the UNI delivered waterL volume limits being adopted for the next cycle of plant operation.

perfect nixing cases are characterized by an extended interval during i

blevdown during which subcooled imI water in the upper head is heated to the boiling point by steam flowing in from the core outlet plenum through

.the guide tubec Dt. ring this reheat period core flow is low, and peak l

clad temperature (PCT) e increases monotonically., The postulated loss of a l

guide tube insert will enhance steam flow into the upper head and reduce 3

the duration of the reheat period, allowing the upper head to begin to i

drain sooner.

Since this is accomplished with no loss of : support colu:tn -

flow capability, the drain flow from the upper head will cool the core as l

previously analyzed.* The not effect of the loss of a 15X15 guide tube j

insert is a reduction in calculated PCT for the limiting perfect mixing i

ossumption.

Imperfect sixing cases arsi characterited by early flashing of the "

fluid in the upper head and un early upper head draint the top of the j

guide tubes is uncove, red approximately 10-15 seconds after the inception of the large break LOCA.

i The impact on the calculated ECCS performance of a missing 15X15 insert should be small in magnitude for the imperfect Diving assumption because liquid flow through the guide tubes exists for such a short time during the blowdown.

cases exhibit more than 80 F margin in calculated PCT to the 2200 7The sequoyah F 0

0 regulatory limit.

Inasauch as p e impact of a missing guide tube insert i

I L

10 estinated to be less than 20 F in imperfect mixing cases, compliance l

with the regulatory limit is assured.

i Therefore, both the perfect and inperfect. aixing cases of the sequoyah large break IcCA retain acceptable Eccs performance analysis results for the highly unlikely scenario of the i

l co:plete ejection of a 15X15 guide tube insert.

i I, !

'"--'*TJP9M-'t9=--'erye ww-rwm' eN--m~r

,w wmee-r---y-wism*m e-------m--w

>meaw e w^mwy-

-"---*------erT*m7 e*---

- " - ' ' -9"-"

,c*

ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNIT 1 9

DOCKET NO. 50-327 (TVA-SQN-TS-88-28)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS 1

c ~

ENCLOSURE 3 1

Significant Hazards Evaluation TVA has evaluated the proposed technical specification change and has determined that it does not represent a significant hazards ec4.ileration based on criteria established in 10 CFR 50.92(c). Operation of SQN in accordance with the proposed amendment will nott (1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

Fo(s) is defined as the maximum local heat flux on the surface of a fuel rod divided by the core average heat finx.

iq(s) is used to limit the nagnitude of hot spots and is used as a bounding input for accident analyses.

Fq(z) is not postulated as being the initiating event for any accident scenario.

Therefore, the proposed change does not affect the probability of any accident previously evaluated.

The proposed reduction in Fq(z ) frem 2.237 to 2.15 is conservative in nature, in that it results in reduced PCTs during a postulated accident.

The F (s) reduction serves as an operational restriction to ensure that 0

PCTs remain below the 10 CFR 50.46 limit of 2.200 degrees F.

Because of the reduction in calculated PCT, the proposed change will not increase the consequences of a previously evaluated accident.

l (2) Create the possibility of a new or different kind of accident from any previously analyzed. As stated above. Fq(s) is not assumed to be the initiating event for any accident scenario. The proposed change to Fq(s) provides additional PCT margin to ensure that the

)

2,200 degrees F limit is not exceeded. The presence of additional margin will not create the possibility of a new or different kind of i

accident.

(3) Involve a significant reduction in a margin of safety.

The proposed reduction in Fq(z) is conservative in nature as it lowers the calculated PCT for the limiting LOCA analysis case. As calculated by i

M, the proposed reduction in Fq/z) from 2.237 to 2.15 lowers the calculated PCT by 87 degrees F for the limiting imperfect mixing case and by 96 degrees F lar the limiting perfect mixing case. These reductions, ecmbined with PCT margin obtained by administrative 1y limiting steam generator tube plugging to 5 percent, result in 4

calculated PCTs of 2.089 degrees F for the limiting imperfect mixing i

case and 2.06/ degrees F for the limiting perfect mixing case.

I Beceuse the calculated PCT remains below the 2,200 degrees F limit of 10 CFR 50.46, there is no reduction in the margin of safety to cladding failure, and additional margin is being added.

1 6

l/