ML20035A208

Safety Evaluation Supporting Amend 111 to License NPF-12

ML20035A208
Person / Time
Site:	Summer
Issue date:	03/18/1993
From:	Office of Nuclear Reactor Regulation
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ML20035A205	List: ML20035A204 ML20035A208
References
NUDOCS 9303240288
Download: ML20035A208 (7)

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'o UNITED STATES

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E NUCLE AR REGULATORY COMMICSION 37 g

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EAHlL EytMTP BY THE OFFICE OF NUCLE AR REACTOR REGUL ATION REL TED'TO AMf NDME:. NO,11110 FACit ITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELIF 1C & CAS COMPj M SOUTH CAROLINA PUCLIC SERV 1CE AUTHOR 111 VIRGIL C. SUMMER tbEll AR ST All0N VNil NO. 1 m

DOCKET NO. 50-395 I

1.0 ltHRODur i101 By letter dat ed October C,,1992, South Carolina Electric & Cas Company (the licenute) sub.:,itted a requast for changes to the Virgil C. Summer Nutlear i

Station, Unit No. 1 (Suw,ar St'at ion), Technical Specifications (15).

The amendment request proposed to revise Table 2.2-1, " Reactor Trip System I

Instrumuntation Trip Setroints," to allow an increase in the maximum permissible average level of steam generator tube plugging (SG1P) from 15 i

percent to 18 percent.

An increase in SGlP reduces reactor coolant system minimum neasured flow (MMF) and, therefore, requires (1) changes to a constant and a setpoint reduction penalty in the overtemperature delta T (OT delta T) setpoint equation, (2) changes to the OT delta 1 trip allowance and the value for column Z of Table 2.2-1 for the affected channel (Z), and (3) a revised loop design flow listed in Table 2.2-1.

2.0 EVALUATION Currently, the licensing basis analyses for Summer Stat ion, as documented in

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their final Safety Analysis Report (f SAR), are bounding for a minimum average SGlP of up to 15 percent.

Summer Station has experienced tube corrosion problems in the D3 steam generators and as a result an increasing number of tubes have been plugged during the last several outages. The increased l

plugging mayL affect the reactor coolant system (RCS) in several ways;

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specifically, (1).the RCS flow may be reduced to a value below the currently analyzed value:in the licensing basis, and (2) the reactor vessel outlet temperatureT(Tw) nay increase and exceed the assumed value in the non-LOCA (non-loss of coolant accident) and structural evaluations.

l The licensee has evaluated the impact on the Summer Station licensing basis f

for plant operation with an increase in the maximum permissible level of average SGTP from 15 percent to 18 percent.

Their evaluat icn also permits the

. maximum level of SGTP in one steam generator to reach, but not exceed 20 percent provided the average level of plugging between the three steam generators does not exceed 18 percent.

For asymmetric events involving a flow or coolant temperature change in one loop, operating conditions for the three 9303240288 Q M 95 pg ADOC pg P

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-. loops based on SGTP of 20, 20 and 14 percent were assumed.

These initial condit ions maximize loop asynnetries while maintaining average SGlP at 18 percent to' minimize core inlet flow.

l 2.1 Non-l0CA fvents 2.2.1 Departure from Nucleate Boiling (DNB) Events I

t The event s listed below are affected by the proposed 18 percent average SGlP in that the reduction in RCS flow could potentially ir, pact the DfG ratio (D M ).

Therefere, to address the reduct ion in the MMF, DIE : penalties of 2.2 l

per cent for the typical fuel cells and 2.0 percent for the thimble fuel. cells were calculated to corpensate for the 1.7 percent decrease in the MB.

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TL licensu ir.dicated that bar.ed on the preliminary Cycle 8 reload design and l

ginn the allocat ion of the ponalties shown above, there will be more than 6 I

percent generic DNCR margin fur the limiting fuel cell type available for I

fut e use.

Therefere, the DNB licensing basis criteria will continue to.be

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met i the conclusions in the FSAR remain valid for the transients listed bel ow.

l F5AR sect ion Event.

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15.2.2 Uncontrolled Rod Cluster Control Assembly

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Bank Withdrawal at Power 15.2.3 Rod Cluster Control Assembly fiisoperation i

15.2.10 Excessive lleat Removal Due to feedwater System Mal func t ions 15.2.11 Excessive toad Increase Incident i

15.2.12 Accident al Depressurization of the RCS 15.2.13 Accidental Dcpressurization of the Main Steam System 15.3.2 Minor Secondary System Pipe Breaks 15.3.4 Complete loss of Forced Reactor Coolant flow 15.3.6 Single Rod Cluster Control. Assembly Withdrawal at full Power J

15.4.2.1 Major Rupture of a Main Steam Line l

The staff finds that the DNB cvents remain bounded by the' Summer Station fSAR l

with 18 percent SGTP.

2.1.2 hnghTe'rm Heat Removal and React _ivit y [xcursion Events l

l The events 1isted below are impacted by the.18 percent average SGTP by a -

l reduction in RCS flow.

In the case of-18 percent average SGlP the total dynamic flow (TDF) is reduced from 94,500 gpm/ loop to 92,900 gpm/ loop and the RCS is maintained at 587.4*F during full power operation and at 557 'F during j

hot zero power operation.- The FSAR analysis is based on 92,600 gpm/ loop and an RCS temperature of 587.4*F.

Additionally, the minimal reduction in RCS volume (less than 1 percent), due l

to the increase in SGTP, will not adversely affect these transients.

Therefore, the staff finds acceptable the licensee's conclusion that the i

increase in maximum average SGTP level will not invalidate the' assumptions used in these analyses and finds that the results and conclusions. presented in the FSAR remain valid for the following events:

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[MR Section-LvcM 1-l 15.2.8

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Loss of Normal feedwater l

15.2.9 Loss of Offsite Power to the Station Auxiliaries 15.4.2.4 Major Rapture of a Main feedwater Line 15.2.1 Uncontrolled Rod Cluster Control Assembly Dank l

Withdrawal from a Subtritical Condition i

15.4.6 Rupture of a Control Rod Drive Mechanism flousing 1

Pgirat ning Hon-10CA Events

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i 2.1.3 l

The licensee assumed the worst case scenario for SGTP with regard im evaluat ing the remaining non-LOCA events - either the maximum SGTP of 18 f

percent or the minimum SGlP of 14 percent.

In all cases, it was determined-

-l that the inact of the 18 percent average SGlP on the original licensing basis 1

analysis was negligible and, for the reasons indicated, the conclusions in the l

FSAR remain valid.

l Or. cont rolled Doron Dilut ion ( ?

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During an uncontrolled baron dilution event, the current licensing basis'still remains valid and sufficient time exists for operator act ion (15 minutes during %fes 1 and 2 and 13.4 minutes during Mode 3/4 based on Refercrice 3).

Also Mode 3/4 was initially analyzed for 20 percent SGTP.

i Eart ial loss of Forced Reaglorloolan1_[ low (15,2.51 l

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This event remains bounded by the complete loss-of-flow event (15.3.4) l discussed previously in sect ion 2.2.1.

.Sigrtup_of an Inact ive Reactor Coolant t o.op_(1546) j A sufficient DNBR margin exists (25 percent) to accommodate the competing DNBR effects caused by the increase in total dynamic flow and power.

l L9ss of Fxt ernal Elgetrical Load artdlor Turbine Trio _(15.2.71 j

The analysis supporting the current FSAR is based on 92,600 TDF.

The TDF for 18 percent SGTP is 92,900 TDF which is bounded by the more conservative current FSAR analysis.

Inadvertent _ Operation of the_Erarggnty Core Coolina System During Power Operation (LS A 14)_

j The change in RCS steady state flow is not enough to impact the.the peak-RCS

-l pressure or the margin to pressurizer fill.

t inadvertent loading of a fuel Assembly into an Improper Positionl15.3.31

'i generated, therefore the analysis in the current FSAR remains' unchanged..

l This event would affect the core power shape and not the total power-t

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Sinale Reactor Coolant Pump Locked Rotgr_(15_.4&

j Based on exist'ing: sensitivity studies, applicable to Summer Station, with the I

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1.7 percent jncrease in TDF the peak RCS pressure remains at 2605 psia and the peak cladding' temperature (PCI) margin is decreased by approximately 10*f.

f Both remain well within the limits.

o The staff has reviewed the licensen's evaluation and finds this conclusion accept able.

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1 2.1.4 ht aint Impac_t i

f Another area that may he impacted by the 18 percent average /20 percent peak SGlP is the effect r symmetrical tube plugging on the delta T calculation.

The asymmetrical SGL un create flew and inlet temperature asymmetries between RCS lc 4

Each loop has only orie channel and it is used to determine the delta T in 'the individual coolant l'

under specific loop inlet and-1 outlet conditions.

The delta 1 may vat rom loop to loop, but the K-terms in the OT delta T and the overpressure delu 1 (OP delta T) 1."tpoint. equations.

i remain constant for all three loops.

Since the delta T setpoints are based on t

a fraction of the individual 1oop delta T and the loop channels are individually calibrated based on the loop temperatures, the OT delta T/0P delta T reactor trip functions will continue to-remain ef fective urider the asymetric conditions.

l The licensee perfortred an analysis to determine the adequacy of the actual OT 1

delta T and OP delta T protection setpoints.

It was concluded that' changes to the K[, lA, and 7 terms along with the slope of the positive portion of the.

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F(del a 1) penalty funct ion of the OT delta 1 setpoint equation were necessary-t due to the reduction in the MMF The new values - 1.195 from 1.203 for K,-

1 10.3 from 9.8 for TA, 7.8 from 7.21 for 7, and 2.34 from 3.13 for. the positive i

portion of F(al) - were calculated by approved methods.

The reduction in mF did not have an impact on the OP delta 1 setpoint, therefore, the current OP delta T setpoint remains sufficient.

I In addition, the licensee has committed to the following in their station procedures to prev rve the reactor protection system's ability to ensure that the core safety limits are not violated in the presence of a potential

.i asymmetry in the loop temperatures The'oveFiemperture delta T reactor trip channels will be calibrated duririg power. operation in terms of both the delta T and T indicated m

by each channel at nominal full power.

l The staff finds the identified changes acceptable and agrees that the FSAR analyses and conclusions will remain valid.

2.2 Lospof-Coolant AccidentJOCA) Analyses 2.2.1 Large and Small Break LOCA

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i The current analyses for both large break (LB) and small break _(SB) LOCA j

(LBLOCA and SBt0CA) assumes 20 percent plugging in each steam generator.

Summer Station operation with the proposed 18 percent average SGTP is bounded l

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. by the current LOCA analyses. The proposed change will not adversely impact the FSAR LDLOCA and SBLOCA analyses results provided the plugging is limited to equal; toloFLless than 20 percent in any one steam generator.

2.2.2 Post-LOCA Long Term Core Cooling In post-LOCA ccoling, the t. ore remains shutdown b.f the borated emergency core cooling systea (ECCS) water in the RCS sump.

The analysis does not take credit for control rods, therefore, the boron contentration of the ECCS must 3

be such that when mixed with other pater sources the reactor core will remain

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Since the RCS is a net dilution source for the mixed sump, maximizing the RCS vtlume is conservative - increased SGlP results in reduced RCS flow. The post-LOCA core cooling was calculated assuming 20 percent SGIP and, ther.

re, bounds the ca of 18 percent average SGTP.

3.0 SUM 3ARY The staff has reviewed the SCESG submittal proposing to increase the steam generator tube plugging from 15 percent to 18 percent average and 20 percent peak.

The submittal considered the effect the increased plugging would have on non-LOCA and LOCA accidents, and the asymnetrical SGTP on OT delta T and OP delta T setpoints.

It was determined by the licensee that sligbt changes in the K value for thb 01 delta T setpoint equat ion and the f(delta-1) function were n,ecessary to preserve the results of the current non LOCA analyses. The licensee also added an 01 delta T trip channel calibration to their Core Operability Limits Rcpart to preserve the core safety limits in the event of the loop temperature a syre t ry.

Both the L0r.A and non~10CA accident analyses for the proposed steaa generator tube plugging are bounded by the current analyses and conclusions in the Summer Stat ion FSAR.

The staff has reviewed the submittal and finds the 18 percent average /20 percent peak SGTP acceptable and all related analyses and conclusion 'in the FSAR remain valid.

The described changes to the EST response times will not invalidate the analyses or subsequent conclusions in the FSAR for any LOCA or non-LOCA transient. gThe proposed amendment to the Summer Station TS is, therefore, acceptabledt '

]gq 5.0 STAiE EdidSULTATIOJ 1 +:

In accordance with the Commission's regulations, the State of South Carolina official was notified of the proposed issuance of the amendment-. The State official had no comments.

6.0 EtNIRONMENTAL CONSIDERAT10N The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

The NRC staff has determined that the amendment involves no

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significantlincreaseintheamounts,andnosignificantchangeinthe-types, of any~ effluents that may be released offsite, and that there is no significant ' increase in individual or cumulative occupational. radiation exposure.. The Commission has previously-issued a proposed finding that the

_i amendment involves no significant hazards consideration, and there has been no public comment on such finding (57 FR 58251). Accordingly, the amen.'m W meets the eligibility criteria for categorical exclusion set forth in 1 ifR j

51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statent ot or -

<t environmental assessment necd be prepared in connection with the issuance of j

the aroendient.

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7.0 00F LUSION l1 1

The Commission has concluded, based on the considerations discussed above, j

that :

(1) there is reasonable assurance that the health and safety of the -

L public will not be endangered by operation in the proposed manner, (2). such.

l activities will be conducted in compliance with.the Commission's regulations, i

and (3) the issuance of the amendr.,ent will not be inimical to the common defense and security or to the health and safety of the public.

Principal Centributor:

S. Brewer l

s Date: March 18, 93 i

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-t REFERENCES ^

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Letter from J.L. Skolds, South Carolina Electric & Gas Company, to-USHRC, "ls Change Request - Increase in Steam Generator Tube Plugging i

froar15% to IBE", dated October 6, 1992.

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UFSAR, " Virgil C. Sumn er Nuclear Power Station, Unit 1", South Carolina Electric 3as Company.

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NUREG-0717, "SER Related to the Operation Of V.C. Sunuer Nuclear Station i

Unit 1*, Supplement 3, dated January 1982.

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