ML20196A232

Proposed Tech Specs Consisting of Tech Spec Change 88-33, Revising Upper Head Injection Level Switch Setpoint & Tolerances of Surveillance Requirement & Heat Flux Hot Channel Factor of Limiting Condition for Operation

ML20196A232
Person / Time
Site:	Sequoyah
Issue date:	12/02/1988
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19295G773	List: ML20196A222 ML20196A232
References
NUDOCS 8812050291
Download: ML20196A232 (15)

Text

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l ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT t! NIT 2 [

DOCKET NO. 50-328 (TVA-SQN-TS-88-33)

LIST OF AFTECTED PAGES i Unit 2 3/4 2-4 3/4 2-5 3/4 2-6 3/4 5-4 l

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S312050291 881202 FDR ADOCL 05000326 p CDC

l POWER O!$TRIBUT!0N LIMITS 3/4,2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be limited by the following relationships:

W i F9 (Z) 1 (2.237) (K(Z)) for P > 0.5 i P i Fq (Z) $ (2.237] (K(Z)] for P $ 0.5 i 0.5 l

TFERMAL POWER where P =

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

APPLICABILITY: MODE 1 i ACTION:

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With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1,% q F (Z) exceeds the limit l within 15 minutes and similarly reduce the Power Range Neutron [

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 R 21 i may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION i may proceed provided the Overpower Delta T Trip Setpoints (value of f.4) have been reduced at least 1% (in AT span) for each 1% Fq(Z) h exceeds the limit. I

b. Identify and correct the cause of the out of limit condition prior to increezing THERMAL POWER; THERMAL POWER may then be incr;ased providedFh(Z)isdemonstratedthroughincorerappingtobewithin l its limit.

f "URVE!LLANCE REQUIREMENTS  !

y.g , 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.  ;

, SC 2 0139b ,.

SEQUOYAH - UNIT 2 3/424 -m : a b ,' 3

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POWER DISTRIBUTION LIMITS SURVEILLANCli REQUIREMENTS (Continued) 4.2.2.2 Fq (z) shall be evaluated to determine if qF (Z) is within its 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 seasured Fqc,) component of the power distribution ,

map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c. Satisfying the following relationship:

N 2 Fq (2) $ g) x K(z) for P > 0.5 H 2.237 Fq (z) 1 Mz) x 0.5 x K(z) for P 1 0.5 where F (2) is the measured Fq (z) increased by the allowances for manufacturing tolerances and measurement uncertainty, q F limit is the Fq limit, K(z) is given in Figure 3.2-2 P is the relative 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.

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d. Measuring Fq (z) ccording to the following schedule:

1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which Fq (z) was last determined.' or ,

2. At least once per 31 effective full power days, whichever occurs' first.

"During power escalation at the beginning of each 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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-SET' 2 01983-SEQUOYAH - UNIT 2 3/4 2 5 4. ;na:n: nfa 9

t POWER O!STRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) '

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e. With measurements indicating  !

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l over z k K(z) I has increased sint.e the previous determinatin of F N '

of the following actions shall be taken: S (z) either t

. 1. F g N(z) shall be increased by 2 percent over that specified in  !

4.2.2.2.c, or

2. F g N(z) shall be measured at least once per 7 effective full j

power days until 2 successivw maps indicate that T maximum Fn (z) is not increasing.

over z K(2) R 21

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

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1. Calculate the percent 9F (2) exceeds it's limit by the following expression: l t e . l aximum 0 I*) * (*) -1 x 100 for P 3,0.5 over z L

2 237*, g(g) (

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maximum Fn (2) x W(z) 1 x 100 for P < 0.5 (over2 2.2375,ggg) l k, =

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2. Either of the following actions shall be taken: I f

c. Place the core in an equilibrium condition where the I 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 teduced 1% AFD for each perewntg F (z) exceeded its limit, or -

Infed b. Comply with the requirements of Specification 3.t.2 for '

& 4 4 *,41 Fq (z) exceeding its limit by the percent eticulated above. , i T

SEQUQYAH - UNIT 2 3/4 g.6 L.

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Footnote "A" to be Inserted en Psges 3/4 2-4 3/4 2-5 3/4 2-6

'A The limit shall be 2.15 instead of 2.237 until an analysis in conferrance with 10 CTR 50.46. using plant operating conditions and '

showing that 4 timit of 2.237 satistles the requirerents of 10 CTR 50.46(b). has been completed and submitted to NRC.

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)1 EMERCENCY CORE COOLING SYSTEMS (ECCS) i i  ;

SURVE!!, LANCE RE00!REMENTS (Continited) j J

b. At l'ast once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> afteIoach solution I volume increase of-greatas than or equal to 1% of tank volume by l

, t erifying the boron concentration of the solution in the water-filled  !

accumulator.

c. At least once per 18 months by: -

1. Verifying that each acebulator isolation valve closes automa-  :

! tically when the water level in the water-filled accumulater is '

3 M.0 +2.V-If ::. l'.t inches abo've the tank. vendor workins line/' e4. at; i

:::::;:7 : :: S?.1 ;i.5 in d:: when corrected for the mass d

, of cover gas. ,

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2. Verifying that the total dissolved nitregen and air in the i j water-filled accumulator is less than 80 SCF per 1400 cubic I feet of water (equivalent to 5 x 10 -5 pounds nitrogen per f

pounds water). , i l .  :

l d. At least once per 5 years by removing the membrane installed between i j the water-filled and nitregen bearing accumulators *and verifying l j that the removed memorane bursts at a differential pressure of  !

40 10 psi.

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LNCLOSURE 2 t

PROPOSED TEC"';,.-l SPECITICATION CHANGE SEQUOYAH N L GR P1 ANT UNIT 2 DOCKET NO. 50-328

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(TVA-SQN-TS-88-33)

DESCRIPTION AND JUST!T! CATION FOR , i REVISING UN! LEVEL SN!TCH SETPOINT AND TOLER/NCES  !

AND REDUCTION IN HEAT TLUX ROT CHANNEL TAs: TOR LIMIT i

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r ENCLOSURE 2 DESCRIPTU N OF CHANGE Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant (SQN) unit 2 technical specifications to revise the upper head injection (UHI) level switch setpoint and tolerances of surveillance requirement (SR) 4.5.1.2.c.1 and the heat flux hot channel factor (F limiting condition for operation (LCO) 3.2.2 and SR 4.2.g(z]) . 2. of This proposed revision to the SQN unit 2 UHI technical specifications is consistent with the SQN unit 1 technical specification proposed change 88-20 (submitted August 15, 1988; and supplemented by letter dated "

September 21, 1988; which NRC approved by letter dated October 14, 1938) and 88-23 (submitted September 21, 1988).

REASON FOR CHANGE Condition adverse to quality report (CAQR) SQPS71644 docunents that the level switches and setpoints that were used previously could allow more

' than the analytical limit of 1,130.5 cubic feet of UHI water to be injected during a postulated accident. Two changes in the design and configuration of the UHI system were pursued to correct this potential problem. First, the minimum delivered UHI water volume was reduced f rom 900 cubic feet to 850 cubic feet. This enange is supported by

'destinghouse Electric Corporation (W) evaluations described in a September 14, 1988 letter to IVA (included as attachment '. ) . Second, a new model of level switch is being installed in the UHI system. These new switches are essentially the same as those presently used, except for their span. 3ecause of the span differences, the switches also have difterent accuracy characteristics. Demonstrated Accuracy Calculation 1-LS-87-21 determined a new setpoint and tolerances based on the new instenment characteristics. These new values are being incorporated into SR 4.5.1.2.c.1 to ensure that the delivered UHI water volumes are bounded by the 9olumes assumed in the large-break, loss of coolant accident (LOCA) analyses. This in turn ensures that the offsite doses from a postulated LOCA are bounded by the analyses of the Final Safety Analysis Report (FSAR), section 15.5.

The change in the delivered UHI water volume band described above is supported by 'f e"aluations, which indicated that the potential decrease in delivered water voleme to the core would result in increased peak : lad temperatures (PCTs); but in all cases, PCT remained below the 2,200-degrees-Fahrenheit (F) limit of 10 CFR 50.46. NRC has indicated that operation of unit 2 could be supported by the sensitivity studies (prov.ded a temporary exemption to .ertain administrative requirements of 10 CFA 50.46(a)(1) was obtained) and that operational restrictions be imposed to provide at least 100 '- rees F of margin between the calculated PCT and the 10 CFR 50.46 limit.

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Evaluations by W have determined that at least 100 degrees F PCT margin can be obtained by administrative 1y limiting steam generator tube plugging (SGTP) to 5 percent and by reducing Fo(z) from 2.237 to 2.15. The proposed Fo(z) limit change is being submitted to reflect this operational restriction.

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

JUSTIFICATION FOR CHANGE Delivered UHI Water Volume The UHI system is designed to passively supply additional inventory to the reactor core during the blowdown phase of a postulated LOCA. The UHI system is described in FSAR section 6.3.2. As described in FSAR section 15.4.1.1.4, a broad spectrum of LOCA analyses has been performed to evaluate UHI performance. The various UHI performance analyses are categorized by the assumed discharge coefficient (Co) of the break and the presence or lack of UHI water mixing in the upper head region of the vessel (perfect and imperfect mixing, respectively).

The limiting case break in the UHI Evaluation Model emergency core cooling system (ECCS) analysis presented in the original SQN FSAR was the discharge coefticient Co=0.6 double-ended, cold-leg guillotine (DECLG) break with imperfect mixing of UHI water assumed in the ves* 4pper head. Compliance with regulatory limits was achieved for tr. case by reducing the allowable core peaking factor (Fq) from 2.32 to *.237.

4 Minimizing the volume of UHI water delivered maximizes PCT for imperfect mixing UHI LOCA cases. The lower bound value for UHI water volume delivery established in the original FSAR Co=0.6 DECLG imperfect mixing case is 900 cubic feet. This value also was employed in the imperfect mixing cases of the 10-percent SGTP analysis performed in the 1982-83 tiamframe.

A complete spectrum of perfect mixing cases was analyzed for the original SQN FSAR. The limiting case with perfect mixing of UHI water assumed in the vessel upper head was the Co=0.6 DECLG; the calculated PCT for this case is 2,111 degrees* F at an Fq of 2.32 with a UHI-delivered water volume of t,053 cubic feet.

Using sensitivities appropriate to UHI plant perfect mixing cases, tradeoffs have previously been made among various input assumptions to justify increasing the maximum allowable UNI-delivered water volume to 1,130.5 cubic feet. Increasing the value of UHI water delivered maximizes PCT for perfect mixing UHI f.0CA analyses. With the present technical specification Fq of 2.237 in force, 1,130.5 cubic' feet is a valid maximum delivered water volume for the SQN UHI synism because it results in a PCT of 2,163 degrees F.

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< It should be noted that separate safety evaluations performed for SQN have considered the impacts on PCT of guide tube flexure failures, increased feedwater isolation valve stroke time, reduced safety injection fl.ow from a failed residual heat removal pump miniflow, and thimble tube filling during core reflood. For tFe perfect mixing cases, these scenarios do not ,

impact PCT; and 2,163 degrees F remains the limiting PCT for perfect mixing cases.

The Cp=0.8 and Cp=0.6 DECLG imperfect mixing cases from the 1982-83 10-percent SGTP analysis have been reviewed to assess the PCI impact of reducing the delivered UHI water volume to 350 cubic feet. The calculated 3 cts for the C =0.8 D and C D=0.6 DECLG cases that compriso the current

licensing basis for SQN are 2,111 degrees F and 2,113 degrees F.

respectively. Reducing the UHI water delivery in an imperfect mixing case will reduce the cooling of the fuel as the upper head drains during blowdown. During the core reflood phase, this hotter fuel will then expel more injection water as entrained liquid, producing a degraded flooding rate. Existing SQN imperfect mixing cases performed for the FSAR identify the penalty in core fuel heatup associated with decreasing UHI water delivery to 850 cubic feet, which reduces core inlet velocity by 7 percent for the licensing basis imperfect mixing cases.

The impact of degraded flooding rates upon hot red calculated PCT has been determined by WREFLOOD/LCCTA sensitivity runs for each licensing basis imperfect mixing case. The 10-percent SGTP licensing basis imperfect mixing cases are acceptable at an 850-cubic-foot-delivered UHI water volume because the degraded reft. cod penalty only increases calculated PCT as follows:

Cp=0.8 DECLG PCT = 2,151 degrees F Cp=0.6 DECLG PCT = 2,166 degrees "

i The PCT penalties imposed upon the imperfect mixing cases are 20 degrees F for postulated guide tube flexure failures and 12 degrees F for thimble tube filling during core reflood. Because the net PCT for the limiting imperfect mixing Cp=0.6 DECLG case becomes 2,166 degrees F +

20 degrees F + 12 degrees F = 2.198 degrees F, compliance with the regulatory limit is maintained.

Both the perfect and imperfect mixing cases of the SQN large-break LOCA i analysis remain in compliance with 10 CFR 50.46 if the UHt. water-delivered i

i . . volume is within the bounds of 850-1,130.5 cubic feet, f ,

Calculation-of Level Switch Setpoints t , ,

As described in FSAR section 6.3.2, four automatic hydraulic isolation f

valves are used to isolate the URI accumulators f rora the reactor coolant j system (RCS) after UHI has injected. These valves receive automatic .g, closure signals from level switches on the UHI water accumulator. The l 1evel switch setpoints are selected to ensure that the delivered UK ' ite r I

volume is within the limits described above. 1 l

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Demonstrated Accuracy Calculation 1-LS-87-21, included in the August 15, 1988 letter, generates the level switch setpoint and tolerances that ensure that the delivered UHI water volume is between 350 and 1,130.5 cubic feet. As seen on page 6 of the calculation, a tank levol of 95.2 inches (above the working line) equates to a delivered volume of S50 cubic feet; and a tank level of 35.1 inches equates to a delivered volume of 1,130.5 cubic feet. The calculation then continues to establish setpoint and tolerance between 95.3 and 85.1 incnes. Pages 7 through 23A are a compilation of the various inaccuracies associated with tne level switches, including drift characteristics. The limiting inaccuracies of

+3.29 inches and -6.S3 inches are calculated on page 22.

Becsase of the nature of the drift characteristics, a curve-fit program was utilized to determine the optimum setpoint for the level switches. As described on page 22 of the calculation, the optimum setpoint is calculated to be 92 inches. This yields limiting level switch setpv ints of 35.29 inches to 35.17 inches, which are within the analytical limits described above.

The tolerances used in the revised SR of +2.6/-5.3 inches represent the normal accuracy of the level switches excluding process variables that are unmeasurable at the time of calibration (see pages 3 and 25).

As calculated on page 22 and shown on page 23 the accuracy characteristics of the level switches necessitate calibration at . east every 480 days. This level switch calibration is independent of the level switch / isolation valve functional response test required by SR 4.5.1.2.c.l. As such, the level switch calibrations will be scheduled and tracked independently. This will also allow for the extension of the calibration intervals based on evaluation of the new level switch performance. The calibration evaluations are in accordance with our previous commitment made in response to NRC Sulletin 86-02. TVA will continue to monitor level switch performance through the normal reporting process.

FQ(2) Reduction As defined in SQN FSAR section 4.3.2.2.1, F (z)

Q is the maximum local heat flux on the surf ace of a fuel rod divided by the average fuel rod 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 bertv the 10 CFR 50.46 limit of 2,200 degrees F during postulated LOCA conditions.

The proposed reduction in Fq(z) is a conservative change and will provide additional margin in PCT. As described in the attached 'd evaluation (page 4), a reduction in F (z) 9 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 perfect mixing case. As summarized on page 5 of the *

. evaluation, this PCT reduction, combined with the reduction obtained by administrative 1y limiting SGTP to 5 percent, 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 limits.

o ATTACHMENT 1 j Technical Specification Change 88-33

'd Letter Dated September 14, 1988 (B25 880927 004) e 9

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B25 880927 004 y~j 0,

September 14, 1988 Westinghouse Power Systems

• Electric Corporation NucleN ItchocicD

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Scx355 Pinsburgfi Pennsytvama 15230 0355 Pr. P. G. Trudel TVA-88-761 Sequoyah Project Engineer NS-OPLS-OPL-II-88-572 Tennessee Valley Authority Ref.1) TVA RD #428373 Sequoyah Nuclear Power Plant, DSC-A 2) W G.0. CO-42680 P. O. 2000 - 3) ~TVA-88-746 Soddy Ibisy, TN 37379 C n},$ -4p '

85P6R.-465930

• TENESSEE VALLEY AUTHORITY SEQUOYAH UNITS 1 & 2 EECREASED UHI VOLUME DELIVERY LOCA SAFETY EVALUATION (SECL-88-417, Revision 1)

Dear Mr. Trudel:

In accordance with our telecon of September 7,1988, the LOCA safety evaluation provided in Reference 3 has been revised to reflect the impact or reducing F(Q) and SOTP, and a supplemental infomation document is being provided in response to the NRC request for additional information addressing the LOCA models referenced, clarification of the appropriate limiting breaks, and clarification of the effect 4 of the postulated instrumentatio'n thimble and guide tube flexure failures.

The revised LOCA safety evaluation, SECL-88-417, Revision 1, entitled, Safety -

Evaluation for a 50 Cubic Feet Nerease in the UHI Accumulator Deliverable Water Volume (LOCA, SGTR, Post-LOCA Long. Tem Core Cooling and Hot Leg Switchover Accident), is attached. Bis revision incorporates the impact of reducing F(Q) from 2 32 to 2.15 and the Steam Generator Tube Plugging (SGTP) level from 10% to 5%.

he supplemental information document is also attached and is entitled Supplemental Information to SECL-88-417, Revision 1.

If you have any corrents or questions, please contact the undersigned. ,

Very truly yours, WESTI,NGHOUSE ELECTRIC CORPORATION

/ g.

. A. Lordi, Manager

. ESSD Projects Mid-South Area L. V. Tomasic/tu Attachment -

cc: D. W. Wilson W. R. Mangiante i

• S. J. Smith R. W. Headows J. A. Vogel H. J. Durzynski R. C. Weir R. G. kvis R. E. Daniels H 3 "^Y gg; 19 08

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SECL NO: SECL-88-417 Rev. 1

. Cuctcmar Raforsnca No(s) .

Westinghouse Ref. No.

WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIST

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

2) CHECK LIST APPLICABL2 TO:_ SAFETY EVALUATION FOR A 50 CU.PT. DECREASE IN (subject of Change) THE UMI ACCUMULATOR DELIVERABLE WATER VOLUME

3) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.S9 has been prepared to the extedt required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.

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

CHECK LIST - PART A

( 3.1) Yes X No A change to the plant as described in the FSAR?

(3.2) Yes No X A change to procedures as described in the 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 l evaluated in the FSAR be increased?

(4.3) Yes.;_-.8No X May the possibility of an accident which is different than any already evaluated in the FSAR be created?

(4.4) Yes No X Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

(4.5) Yes No 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 important to safety different than any already evaluated in the FSAR be created?

(4.7) Yes No X Will the margin of safety as defined in the bases ,

to any technical specification be reduced?

PAGE 1 OF 2 ,

, SECL-88-417 Ravision 1 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 summarizes the justification upon the written safety -

evaluation, (1) for answers given in Part B of the Safety Evaluation Check List:

See the attachment (1) Reference to document (s) containing written safety evaluation:

NS-SAT-SAI-88-362 FOR FSAR UPDATE Section: Page(s): Table (s): 15.4.1-9 Reason for/ Description of Change:

Chance Table 15.4.1-9 for UHI Accunulator water volume delivered to reflect 850 cu.ft. minimum volune evaluated in this safety evaluation and the associated footnote. _

6) APPROVAL LADDER (6.1) Prepared by (Nuc1' ear Safety): 4^- ( S A I ) 'Da te : [h/

Reviewed by"(Nuclear Safety): I/. 8. [M ad& u(SAI)_,Dat,e:.f M S3 _._

(6.2)' Coordinated with Engineer (s)D //0 86VIEACISAII) Date:

Coordinated with Engineer (s): A / 6 C N W I T S A.) p a t n t. ... .. _,_

Coordinated with Engineer (s): fBEL//OO% AP-(COA) Date:

Coordinated with Engineer (s):>PFcA/Ac. 5 r/u ISAI) Date:

(6.3) Coordinating Group Manager (s): PP/dr G $/LP43AII) _Date:

Coordinating Group Manager (s) : i Out# ZAM (TSA) Date: _

MAu_ LOC 4 coa /6Date [. N4A/CED (6,4)'Coordinating Nuclear SafetyGroup Manager Group Manager: (s) [3 d N N /M/4

- (SAI) Date:34V/81. . _

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