ML20115A550

Proposed Tech Specs,Table 2.2-1, Reactor Trip Sys Instrumentation Trip Setpoints

ML20115A550
Person / Time
Site:	Summer
Issue date:	10/06/1992
From:	SOUTH CAROLINA ELECTRIC & GAS CO.
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ML20115A538	List: ML20115A535 ML20115A550
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NUDOCS 9210140369
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Attachment I to Document Control Desk Letter TSP 920002 Page 1 of 1 PROPOSED 1ECilNICAL SPECIFICATION CllANGE - TSP 920002 VIRGIL C. SUMMER NUCEEAR $1AT10N LIST Of AffECTED PAGES AND MARKED UP 1ECilNICAL SPECIFICA110HS bgg $1ecificatign Description of Chance Brief Justification 2-5 Table 2.2-1 Change to the OTAT trip This change is a result of the values for the Total increase in the positive slope of Allowance and 7. the f(AI) penalty function and the increase in nominal AT.

2-5 Table 2.2-1 Change in the footnote This change is a result of the for lo^p design flow. increase in average SGTP ievel from 15% to 18% which impacts Minimum Measured Flow.

2-0 Table 2.2-1 Reduction of the K1 This change is a result of the term from 1.203 to reduction in flow and the 1.195. increased positive slope of the f(AI) penalty function resulting from the increase in average SGTP fiom 15% to 18%.

2-9 Table 2.2-1 Change in the slope of This change is a result of the the positive wing of reduction in flow and increase in the f(AI) penalty nominal AT resulting from the function in Note 1 from increase in average SGTP from 15%

2.13% to 2.34%. to 18%.

l 9210140369 921006

-P

_ PDR ADOCK 05000395 PDR

t m TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

' Total E Functional Unit Allowance (TA) Z S Trip Setpoint Allowable Value

~

-e

- 1. Manual Reactor Trip Not Applicable NA NA NA NA 2; Power Range, Nautron Flux 7. 5 4.56 0 $109% of RTP $111.2% of RTP High Setpoint Low Setpoint 8.3 4.56 0 125% of RTP $27.2% of RTP

3. Power Range, Neutron Flux 1. 6 0.5 0 <5% of RTP with <6.3% of RTP with High Positive Rate i time constant i time constant 32 seconds 12 seconds

4. Power Range, Neutron Flux 1. 6 0.5 0 <5% of RTP with <6.3% of RTP with

~ High Negative Rate i time constant i time constant J, 12 seconds 12 seconds

5. Intermediate Range, 17.0 8.4 0 $25% of RTP $31% of RTP Neutron Flux

6. Source Range, Neutron Flux 17.0 10.0 0 $105 cps $1.4 x 105 cps 7 21' 1. See note 1 See note 2 l

7. Overtemperature AT Overpower AT 5.2 1.96 1.6 See note 3 See note 4 F 8.

<a j g 9. Pressurizer Pressure-Low 3.1 0.71 1.5 31870 psig 31859 psig R

3 10. Pressurizer Pressure-High 3.1 0.71 1.5 $2380 psig $2391 psig k 11. Pressurizer Water Level-High 5.0 2.18 1. 5 $92% of instrument span

$93.8% of instrument span w

12. Loss of Flow 2.5 1.48 .6 >90% of loop >88.9% of loop design flow

• design flow *

~s 9 27q)

$

• Loop design flow = , gpm l RTF - RATED THERMAL. POWER

• • 1.6% span for Delta-T (RTDs) and 1.2% for Pressurizer Pressure.

w TABLE 2.2-1 (Continued)

E

_M REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 1

] tr0TATION l E Z -NOTE.1: OVERTEMPERATURE AT AT 1 AT, [K -

K2 fl{

y [T - T'] + Ka (P - P') - f (al)]

t Where: AT = Measured aT by RID Instrumentation AT, 1 Indicated AT at RATED THERMAL POWER K, <

PCT 1.19 S {

[' K2 1 0.03006

- =

m 4

f The function dynamic generated by the lead-lag controller for T,,g compensation e

=

i,12 t Time constants utilized in lead-lag controller for T, , 1: 3 28 secs.,

12 1 4 secs.

T -= Average temperature, 'F T' $ 587.4'F Reference T,yg at RATED THERMAL POWER K3 1 0.00147 i g P = Pressurizer pressure, psig E P' 1 2235 psig, Nominal RCS operating pressure L 5 5 = -Laplace transform operator, sec 8 O

k

' E . TABLE 2.2-1 ' Continued) -

{

'E REAC10R TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOINTS c NOTAY10N (Centinued) 25 w NOTE 1: . (Continued).

i

• 4 and f 3 power range(AI) is a function nuclear of the indicated ion chamber 2; dif ference between top and bottom detectors of the response during plant startup tests such that:with gains to be selected based on measured instrument 4

t.

(i) for q(

L gb between - 24 percent and + 4 percent fi (AI) = 0 where q and qb ' "'P"'C'" t' RATED THERMAL POWER in the top and bottom halves.of the core respectively, + oand' ISq

{ t. b total T;fERMAL POWER in percent of RATED THERMAL POWER.

(ii) for each percent that the magnitude of q m t gbexceeds -24 percent, the AT trip setpoint E shall be automatically reduced by 2.27 percent of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of q gb **C"

• P' " * "" O ' # P ** P shall be automatically reduced by Jet 3' percent of its value at RATED THERMAL POWER.

I NOTE 2: r\ i Thepercent 2.2 channel's maximum trip setpoint shall not exched its computed trip point by more than AT Span.

\ i NOTE 3: OVERPOWER AT.

'(z.39

g. Al 1 alo [K. - K 3

-f{['f5 T-rw [T - T"))

3

& Where: AT' =

$ as defined in Note 1 r*

AT =

.as defined in Note 1 n K4' i 1.0875

,a -

Ks >,

}% 0.02/*F for increasing average' temperature and 0 for decreasing average temperature

{ gl- =

['a3 Ihe lunction generated by the rate-lag controller for T,ygcompensation dynamic

, ,e , w ,- ,- w -n - - , , - - - --- -w.

TABLE 2.2-1 m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

5 Functional Unit Allo.fance (TA) Z S Trip Setpoint Allowable Value

1. Manual Reactor Trip Not Applicable NA NA NA NA i

- 2. Power Range, Neutron Flux High Sepoint 7.5 4.56 0 $109% of RTP $111.2% of RTP Low Setpoint 8.3 4.56 0 $25% of RTP $27.2% of RTP

3. Power Range, Neutron Flux 1.6 0.5 0 55% of RTP with a time 563% of RTP with a time High Positive Rate constant 22 seconds constant 22 seconds

4. Power Range, Neutron Flux 1.6 0.5 -

0 55% of RTP with a time 563% of RTP with a time High Negative Rate constant 2 2 seconds constant 22 seconds

5. Intermediate Range, 17.0 8.4 0 $25% of RTP $31% of RTP Neutron Flux

6. Source Range, Neutron' Flux 17.0 10.0 0 $105 cps $1.4 x 105 cps

7. Overtemperature aT 10.3 7.8 1.6 See note 1 See note 2 l

& 1.2**

p 8. Overpower aT 5.2 1.96 1.6 See note 3 See note 4 ro a 9. Pressurizer Pressure-tow 3.1 0.71 1.5 21870 psig 21859 psig

10. Pressui _r Pressu e-High 3.1 0.71 1.5 52380 psig $2391 psig j g 11. Pressurizer Water Level-High 5.0 2.18 1.5 592% of instrument $93.b. of instrument span span

$ 12. Loss of Flow 2.5 1.48 .6 290% of loop design 188.9% of loop design

y flow

• flow +

?

w

• Loop design flow = 94,870 gpm

! RTP - RATED THERMAL POWER I l **1.6% span for Delta-T (RTDs) and 1.2% for Pressurizer Pressure.

TABLE 2.2-1 (Continued) m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS c-NOTATION E

x NOTE 1: OVERTEMPERATURE AT s

H (1 + t,S) , ,

i

~ ST s $Te K 1 -E T-T *K3 ( P - P > - f,tSD 2 gi 4 1,,S 3 Where: AT = Meesured AT by RTD Instrumentation AT, 5 Indicated AT at PATED THEPPAL POWER K, 5 1.195 K, 3 0.03006 1 + 1'S = The funtion generated by the lead-lag controller 7 dynamic compensation C) 1 + if for T avg t,,12

= Time constants utilized in lead-lag controller for Tavg* 'T ? 28 SECS-+

t 2 5 4 secs.

l T = Average temperature, *F l

l T' 5 587.4'F Reference T avg at RATED THERMAL POWER K

3 3 0.00147 Pressurizer pressure, psig p P =

{2 P' 5

3

=

2235 psig, Nominai RCS operating pressure Laplace transform operator, sec.

A

TABLE 2.2-1 (Continued)

.g - REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

. g. NOTATION (Continued) y x

~'

c NOTE-1: (Continued)

-2

- and f, (al) is a function of the indicated difference between top and bottom detectors of the power-range i

nuclear ion chambers; with gains to be. selected based on measured instrument response during plant startup tests such that:

4

. (1). for gt - Ub between - 24 percent and + 4 percent f, (aI) = 0 where gt and gb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + Ub is total THERMAL POWER in j

p percent of RATED THERMAL POWER.

c (ii) for each percent that the magnitude of gt - Ub exceeds -24 percent, the AT trip setpoint shall be o automatically reduced by 2.27 percent of its value at RATED THERMAL POWER.

e j (iii) for each percent that the magnitude of qt - Ub exceeds +4 percent, the AT trip setpoint shall be automatically reduced by-2.34 percent of its value at RATED THERMAL POWER. I NOTE 2: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.2 percent'aT Span.

NOTE 3: OVERPOWER AT y- (t,S) ,

A T s; a T K 4 -K T-K T-T g 5(1 + t3S) 6 p

} . S. .

Where: AT -=

as defined in Note 1 y AT, =

as defined in Note 1 m .y 5- 1.0875 m

K >

3 0.02/*F.for increasing average temperature and 0 for decreasing average +emperature 2 '

1S X 3

=

1+1S3 . The function. generated by the rate-Tag controller for T,,9 dynamic compensation i

I IL.

y m'a eye y,  ! , - - > --(- -v eied' er7 re w r4 -t-t- r y -w e st - == r e tr'

• e - - w= W vr"w- ev m es 4+m ww% a - - + * - - e- - - = -- - = -

Attachment 2 to Document Control Desk letter ISP 920002 Page 1 of 26 PROPOSED TECllNICAL SPEClflCATION CilANGC - TSP 920002 VIRGIL C. SUMMER NUCLEAR STATION DESCRIPi10N AFD SAFETY LVALVATION DESCRIP110N Of LICENSE AHLHDMENT REQUEST A) Purpose for Change Changes to the VCSNS Technical Specifications are being requested to permit an increase in the maximum permissible average level of Steam GeneratorTubePlugging(SG1P)from15%to18%. Although a value for SGTP is not specified in ine Technical Specifications,.the increase in average SGTP will result in a 1.7% decrease in the Reactor Coolant System Minimum Measured flow (MMF) value which is referenced in Technical Specification 3/4.2.3. This amendment request reflects the impact of a reduction in HMF on the VCSNS Technical Specifications.

Specifically, changes are required to Table 2.2-1 for a constant and a setpoint reduction penalty in the OTAT setpoint equation, the OTal trip total allowance and 2, and the loop design flow.

B) Current License Condition Technical Specification 3/4.2.3 defines allowable conditions of indicated Reactor Coolant System (RCS) total flow and R (a parameter related to FNAH). This region of allowable operation is specified each cycle in the Core Operating Limits Report (COLR) figure entitled "RCS Total flow Rate Versus R for Three Loop Operation."

Technical Specification Table 2.2-1 lists the reactor protection system instrumentation trip setpoints for the various trip functions, functional Unit 7 represents the trip setpoint and allowable value for the Overtemperature AT _ reactor trip. This trip function is not a-constant value but is continuously calculated by the reactor protection system. Note 1 for this trip defines the OTAT setpoint equation, functional Unit 12 (Loss of flow) provides the trip setpoint and allowable value for the loss of flow reactor trip function which-is footnoted to identify a loop design flow. The loop design flow corresponds to one-thiro of the required Minimum Measured flow (MMF) at Rated Thermal Power (RTFj - 96,500 gpm/ loop.

C) function of Identified Technical Specifications The purpose of the COLR figure for RCS flow vs R for three loop operation is to allow RCS flow rate and power to be traded off against one another while still ensuring that the actual DNBR will not be below the design DNBR value.

The OTAT trip function provides sufficient core protection to preclude i

DNB over a range of operating and transient conditions. The setpoint is automhtically varied with temperature, pressure, and the axial power l

Attachment 2 to Document Control Desk Letter TSP 920002 -

.Page 2 of 26 '

distribution, lhe f(61) penalty function adjusts tne trip setpoint for axial peaks greater than design.

• The footnote to the loss of flow reactor trip values is used to identify, in absolute terms, the flow values at which the trip function should actuate. Since the trip setpoint and allowable value are given ,

in terms of relative flow (90% and 88.9%, respectively), the footnote '

can be used to determine actual flowrates which preserve the basis of the accident analysis.

D) Description of Proposed Change Ior Cycle 8, the COLR figure for RCS flow vs R for three loop operation will be revised to reflect a reduction in Minimum Measured Flow at RTP .

from 289,500 gpm to 284,600 gpm. This COLR change will be implemented in accordance with the requirements of VCSNS Technical Specification 6.9.1.11. This amendment request reflects the impact of a reduction in MMF on the lechnical Specifications, lhe OTAT trip values for total allowance and Z are revised from 9.8 and 7.21 to 10.3 and 7.8, respectively. With respect to Note 1, the value for the K1 term and the positive wing of the f(AI) penalty are also affected by the proposed reduction in flow. The value for the K1 term in the OTAT setpoint equation is reduced from 1.203 to 1.195 and the slope of the positive wing of subsection iii of Note 1 is increased from 2.13% to 2.34%. The nominal value of loop design flow in the footnote to Table 2.2-1 functional Unit 12 is revised to reflect the change in MMF.

These changes will permit the Technical Specifications to allow acceptable plant operation with a reduced MMF of 284,600 gpm at RTP.

i

.. .. - _ _ ~ . . . , _ . , , . . . . , . - . . _ . . . . . . . _ . , , . , _ . . ._ . . . . . _ ,, .,,,,,...s., -

. . . . ,w m,-,,y,, - . . ~ .-, . , ,

Attachment 2 to Document Control Desk Letter TSP 920002

, Page 3 of 26 SAFETY EVALVATION i

1.0 BACKGROUND

The Virgil C. Summer Nuclear Station (VCSNS) has experienced tube corrosion problems in its 03 steam generators and, as a result, an  ;

increasing number of tubes have been plugged during the last several outage *.. Steam generator tube plugging decreases reactor coolant system flow due to increased flow resistances through the steam generators. As the number of plugged tubes increases, the RCS flow may be reduced to a value below that which is currently analyzed in the licensing basis.

Currently, the licensing basis analyses for the VCSNS are documented in the Final Safety Analysis Report (Reference 1) and the VANTAGE +

Licensing Submittal (Reference 2). These analyses are bounding for a maximum average steam generator tube plugging (SGTP) level of up to 15%. Table I lists the primary and secondary system parameters used in the initial conditions of the various accident analyses. Case 1 represents the 0% SGTP original design basis operating conditions.

Case 2 reflects the 15% SGTP parameters used in the current non-LOCA DNB design basis. Initial conditions for the non-DNB events (e.g.,

long term heat removal and overpressurization transients) are presented in Case 3 and are based on 15% SGTP, a Thermal Design Flow (TOF) of 92,600 gpm/ loop, and a RCS averye temperature (TAVG) of 587.4*F at RTP. The current LOCA analyses are limiting at the maximum level of permissible SGTP and were performed using Case 4 which included 20%

tube plugging and a T0F of 92,600 gpm/ loop.

Table 2 presents the new operating parameters for 18% average SGTP.

The new values of T0F and MMF are within those assumed in the current LOCA analyses but are outside the range considered for the non-LOCA DNB-related transients, in addition, by maintaining a constant de ign RCS TAVG of 587.4*F at RTP, the' reduction in primary flow results ir, n increase in the reactor vessel outlet temperature (TH0T) above t'.ot assumed in the non-LOCA and structural evaluations.

ine purpose of this safety evaluation is to assess the impact on the VCSNS licensing basis for plant operation with an increase in the maximum permissible level of average SGTP from 15% to 18%. Operation with an average SGTP level of 18% implies that the linear average of equivalent plugging (i.e., including the effects of.both plugs and sleeves) in the steam generators is equal to or less than 18%. In additiun, the evaluation will permit the maximum level of SGTF in any-one. steam generator to reach but not exceed 20.T provided the average level of plugging between the three stea.a generators A es not exceed 18%. SGTP levels of up to 18% are proposed beginning n Cycle 8. This '

is expected to provide sufficiert flexibility in-plant operation until steam generator replacement.

. . . a, , - . - _ . - - - - . . . -. I

.- -_ -.. ~. -- .

- . =. .

Attachment 2 to Document Control Desk letter  :

TSP 920002 l

. Page 4 of 26 ,

1 1A8LE 1: REFERENCE PARAMETERS VIRGIL C. SUMMER NUCLEAR STATION I

03 NSSS ACCIDENT PARAMETERS Case 1 Case 2 Case 3 Case 4 NSSS Power, MWt 2787 2787 2787 2787 Reactor Power, MWt 2775 2775 2775 2775 T0F, gpm/ loop 98000 94500 92600 92600  ;

RCS Pressure, psia 2250 2250 2250 2250 Core Dypass, % 8.9 8.9 8.9 8.9 RCS Temperaturcs, F Core Outlet 624.1 625.4 626.1 626.1 Vessel Outlet 618.7 619.8 620.5 620.5 Core Average 592.1 592.3 592.5 592.5 Vessel Aversgo 587.4 587.4 587.4 587.4' Vessel / Core inlet 556.0 555.0 554.3 554.3 S/G Outlet 555.8 554.6 554.0 554.0 >

Steam Generator 0 15 15 20 Tube Plug, %

Minimum Meas, flow, 303300 289500 283500 283500 total gpm

i Attachment 2 to Document Control Desk letter TSP 920002 Page 5 of 26 1ABLE 2: PROPOSED 18% AVERAGE TUBE PLUGGING PARAMETERS NSSS Power, MWt 2787 Reactor Power, MWt 2775 10F*,gpm/ loop 92900 RCS Pressure, psia 2250 Core Pypass, % 8.9 RCS Temperatures, f Core Outlet 626.0 Vessel Outlet 620.4 Core Average 592.4 Vessel Average 587.4 Vessel / Core Inlet 554.4 S/G Outlet 554.1 Steam Generator Tube Plugging, % 18 Minimum Measured flow 284600

• TDF incorporates 18% SGTP, with 1% extra flow margin.

Attachment 2 to Document Control Desh Letter TSP 920002 Page 6 of 26 2.0 EVALUA110N i

2.1 NON-LOCA ACCIDENT ANAi.YSES This section will assess the impact of increased SGTP levels on all non-LOCA accidents which form the licensing basis for the VCSNS. The '

evaluation covers system average SGTP up to 18% with a maximum allowable plugging level of 20% within a steam generator. The full range of potential system asymmetries are also addressed, thus allowing the individual tube plugging levels in the VCSNS steam generators to range from their current values (8.5%, 14%, and 12%) up to 20% in any one steam generator provided the average plugging level for all three steam generators does not exceed 18%. For asymmetric events involving i a flow or coolant temperature change in one loop, operating conditions based on SGTP of 20%, 20%, and 14% are assumed. These initial conditions maximize loop asymmetries while maintaining SGTP at 18% to minimize core inlet flow. Each non-LOCA licensing basis event discussed below is either evaluated individually or is in a group which shares the same discussion.

2.1.1 DNB Events The following non-LOCA Condition 11 and !!! transients are analyzed to Condition 11 criteria to demonstrate that the DNB design basis is met. These events are impacted by 18% average ,

SGTP in that the reduction in RCS flow will potentially impact the minimum DNBR. In order to accommodate the proposed reduction in RCS flow, DNBR penalties were calculated for both the typical and thimble fuel cells. Given that the cells have different sensitivities to changes in flow, penalties of 2.2%

and 2.0% were generated for the typical and thimble cells,.

respectively, to address the 1.7% decrease in MMF reflected in Table 2. The DNBR penalties envelope the reduction in steady-state RCS flow explicitly assumed in the initial conditions of each of the DNB-related events identified below.

Based on the preliminary Cycle 8 reload design and given the allocation of the penalties shown above, there will be more #

than 6% generic DNBR margin for the limiting fuel cell type available for future use. Therefore, the DNB licensing basis criteria will continue to be met and the conclusions in the FSAR remain valid for the transients listed below.

FSAR Section Event.

15.2.2 Uncontrolled Rod Cluster ContrM Assembly (RCCA) Bank Withdrawal at Pown 15.2.3 Rod Cluster Control Assembly Misoperation 15.2.10 Excessive Heat Removal Due to Fiedwater System Malfunctions i

, ~ . - -, - --- - - - - . -~ - . - ~ , , . - .-, - . . , . . ~ . . . . - . - . . . - - - ., ,

Attachment 2 to Document Control Oesk Letter TSP 920002 Page 7 of 26 15.2.11 Excessive Load Increase Incident 15.2.12 Accidental Depressurization of the Reactor Coolant System 15.2.13 Accidental Depressurization 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 (RCCA)

Withdrawal at Full Power 15.4.2.1 Major Rupture of a Main Steam Line Note: The operability of the Overtemperature AT trip function -

under asymmetric flow and temperature conditions is assured since each channel (1 channel / loop) is calibrated for the specific loop inlet and outlet conditions. This is discussed in more detail in Section 2.1.5.

2.1.2 Long Term Heat Removal Events The non-LOCA transients listed below are analyzed to show canformance to long-term heat removal requirements and primary and secondary pressure considerations, and are potentially impacted by 18% average SGTP. For all the events listed below, the FSAR analysis was based on an RCS T0F of 92,600 gpm/ loop and an RCS temperature of 587.40F (Table 1. Case 3). However, for the 18% average SGTP effort, the T0F is reduced from 94,500 to 92,900 gpm/ loop and the RCS temperatore is maintained at-587.4of for full power operation and at'5570F at hot zero power conditions. In addition, the minimal reduction in RCS volume.

(less-than 1%) due to the increase in SGTP will.not adversely affect these transients. Therefore, the increase in the maximum average SGTP level will not invalidate the assumptions used in these analyses, and the results and conclusions presented-in the FSAR for the events listed below remain valid.

FSAR Section Event 15.2.8 Loss of Normal feedwater 15.2.9 Loss of Offsite Power to the Station Auxiliaries (StationBlackout) 15.4.2.2 Major Rupture of a Main Feedwater Line

,eN h-

~~..,-.,-..n, , ,w

Attar.hment 2 to Document Control Desk letter TSP 920002 Page 8 of 26 2.1.3 Reactivity Excursion Events The non-LOCA transients listed below are analyzed to show conformanct. to fuel geometry criteria (e.g., peak fuel and clad '

temperatures along with fuel melt limits) and are potentially impacted by 18% average SGTP. For all the events listed below, the FSAR analysis was based on an RCS TDF of 92.600 gpm/ loop and an RCS temperature of 587.4of (Table 1. Case 3). However, for the 18% average SGTP effort, the TDF is reduced from 94,500 to 92.900 gpm/ loop and the RCS temperature is maintained at 587.40F for full power operation and at 5570F at hot zero power conditions. Therefore, the increase in the maximum average SGTP level will not invalidate the assumptions used in these analyses, and the results and conclusions presented in the FSAR for the events listed below remain valid.

FSAR Section Event 15.2.1 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Subcritical Condition 15.4.6 Rupture of a Control Rod Drive Mechanism Housing 2.1.4 Remaining Non-LOCA Events The remaining non-LOCA events are uniquely impacted by the increase in SGTP and are discussed individually below.

2.1.4.1 UncontrolledBoronOllution(FSARSection15.2.4)

This Condition Il event is analyzed for all si,x modes of operation. The analysis demonstrates that sufficient negative reactivity exists, such tSat, should a dilution event occur, t.here is sufficient time following an alarm to allow operator detection l

and termination of the event prior to a complete loss of shutdown margin and return _to criticality.

Eighteen percent average SGTP can potentially impact i the boron dilution analysis for modes of plant

} operation in which one or two reactor coolant-pumps

! are required to operate by decreasing the RCS E

volumes in the analysis.

The boron dilution analysis presented in the FSAR was based upon_ minimum active RCS volumes, determined for each operating mode, using information contained in the plant Technical

_ Specifications. Since tube plugging is assumed to increase by 3% (current analyses assume 15% total SGTP), the active volume assumptions for 18% SGTP

,- for Modes 1 and 2 will decrease by less than 1%.

l l:

This will result in an estimated decrease in

Attachment 2 to Document Control Desk' Letter. 1

. TSP 920002- I zPage 9 of 26 I operator response time before a loss of shutdown-margin of one minute. However, there is still . ,

greater than 15 minutes before the loss of shutdown 'l margin for operator response.

l For Modes 3/4, the FSAR analysis is based upon the i active volume of the RCS with one reactor coolant i loop operating. In this ca$e, the FSAR results  ;

could be adversely affected since RCS flow could l occur in the loop containing the greatest. level of  ;

tube plugging. This would cause a net reduction in- i the active volume assumed for the Mode 3/4  ;

licensing-basis analysis thereby reducing _the. _  !

available operator action time. However, plant  ;

operating procedures require that two reactor-  !

coolant loops be in operation thereby increasing the  ;

assumed RCS active volume thus providing greater operator action times as opposed to one loop operation. In addition, the calculated. shutdown margins for Mode 3/4 are based on 20% SCiP with one RCP in operation. Therefore, it can be concluded that greater than 13.4 minutes are available for i operator action before the loss of shutdown margin given the increase in the maximum' average level of .

SGTPfrom15%to18%(20% peak).  ;

The active volume for the Mode 4/5 analysis,.where  ;

Mode 5 is the Cold Shutdown mode, was based on the ,

reactor coolant system (minus the pressurizer ~ f volume) with one RHR loop operating.- This scenario  :

will not be impacted by increased SGTP since the

~

steam generators are not in use. Mode 6 operation '

also is unaffected since the active volume assumed-  !

comes from the reactor versel andLone RHR loop-(i.e., the active volume of-the steu generator is '

notassumedinMode6).

2.1.4.2 PartialLossofForcedReactorCoolant' Flow (FSAR h Section15.2.5)- ,,

.- -. - s l This Condition 11 event is analyzed under full power conditions assuming that 1 of 3 operating reactor .

coolant pumps coasts down. _The-reactor-.is promptly tripped on low reactor coolant loop flow. The analysis demonstrates that-.the minimum DNBR~ remains: ,

above the limit value. The PLOF event analyzed in'  ;

ch support of-the'FSAR assumed-that forced reactorl i coolant flow was maintained through two pumps / loops: -

containing steam generators which were plugged to a'-

maximum level of 15%. For 18% SGTP, a~ reduction in  :

assumed RCS flow would occur since-forced flow-is- >

maintainedthroughthetwo.loopshaving' higher (20%) 1

-SGTP: levels. .The reduction in RCS minimum measured flow may also produce an increase in the.RCS _

p

+

w ---,-r..

ww .+ t- ,-.- -a.-, ,f.,.,-; g..%w, .v._vy ,, m., , ., w ,,- -.-e,a--tid'=- --

Attachment 2 to Document Control Desk Letter TSP 920002 Page 10 of 26 moderator temperature, both of which tend to reduce the margin to the licensing-basis DNB limit for this event. However,2.2%(TypicalCell)or2.0%  :

3 (Thimble Cell) generic DNB nargin has been allocated to offset the reduction in MMF associated with operation at 18% SGTP. With respect to transient flow conditions, increased SGTP will have a negligible effect on the minimum DNBR due to the offsetting effects of the increased pressure drops <

through the steam generators which could cause a faster flow coastdown and the reductior in RCS flow which would act to reduce the RCS pressere drops and result in a slower flow coastdown. Note that the partial loss of flow event remains bounded by the complete loss of flow event previously discussed.

Therefore, the DNB design basis will continue to be satisfied and the conclusions presented in the FSAR remain valid ive the transient for this event.

2.1.4.3 StartupofanInactiveReactorCoolantLoop(fSAR Section15.2.6)

This Condition !! event is analyzed assuming a maximum initial power level consistent with 2 loop operation and the P-8 setpoint. The startup of an inactive loop results in a reactivity insertion since the inactive loop fluid is at a lower temperature than the rest of the core. The FSAR analysis demonstrates that the minimum DNBR remains above the limit value. The automatic reactor protection system terminates this transient on low coolant loop flow when the power range neutron flux (two out-of-four channels) exceeds the P-8 setpoint which is initially set for two loop. operation. The licensing-basis analysis presented in the FSAR demonstrated that the minimum DNBR was considerably larger than the safety analysis limit _value.-

However, 18% average SGTP potentially impacts these results since the mass of cold water trom the inactive loop can hypothetically come from the loop with the lowest SGTP level, in this situation, the

increased loop flow permits a larger mass of cold

i. fluid to enter the core over a given time period and l produces a larger reactivity insertion (due to the p negative moderator temperature coefficient) relative =

to the case where symmetric SGTP was assumed.

For conservatism, this evaluation assumes that the inactive loop undergoing startup contains the lowest SGTP level. This will produce the largest insurge of cold water into the core and hence the largest positive reactivity excursion. Initid1 RCS TDF from

! the two active loops would be reduced, in comparison to the FSAR analysis, since this flow would be based l

l

Attachment 2 to Document Control Desk Letter TSP 920002 Page 11 of 26 on the two: loops containing the highest levels of SGTP rather than the 15% tube plugging level assumed in the FSAR. analysis. During the-transient, the reactor coolant flow increases due to the startup of-an inactive pump, thereby producing a ONB benefit.

"9 wever, c'DNB penalty may be introduced as a result of a larger-positive reactivity insertion in the core produced by this increased flow. Also note that since the reactivity feedback will be larger, the power will also reach the high nuclear flux (P-

8) trip setpoint more quickly rt ative to the analysis documented-in-the FSAR. Since the increases in T0F and power are competing DNB effects, the changes-in minimum-0NBR due to this scenario are estimated to be small. The results of the FSAR analysis indicate that sufficient margin exists to the safety analysis DNB limit to accommodate minor perturbations in the minimum DNBR.

'Specifically, the FSAR analysis concludes that for VANTAGE 5 fuel,-the minimum DNBR is greater than 2.20 compared tu a DNBR limit of 1.48 or, there is approximately 25% margin to the limit. Therefore, it is judged that the conclusions in the FSAR will remain valid for this event.

2.1.4.4 Loss of External Electrical Load and/or Turbine Trip (FSARSection15.2.7)

The-analyris presented in the FSAR represents a _

complete iuss of steam load from full power without a direct reactor trip.- Four_ cases are analyzed, maximum and minimum feedback, with-and without-pressure control. The analysis demonst ates _that, with the power _' mismatch between the core and turbire, the primary _and secondary system pressures remain below 110% of_ design and that the minimum DNBR remains above.the limit value. SGIP.

potentially impacts:the loss of load and/or turbine trip event by reducing the RCS flow rate = assumed in-the analysis. Reductions in the RCS flow will tend to increase the peak pressures reached in the primary system.

The_LOL/TT analysis which supports the FSAR is based on an RCS TOF of 92,600 gpm/ loop. For the 18%

average SGTP effort, the T0F is 92.900 gpm/ loop.

Since the lower analyzed flow is more conservative and RCS temperature-is unchanged.from the FSAR analysis, the primary and secondary-side peak pressure licensing basis as well as the DNB licensing basis will continue to be met and the conclusions in the FSAR recain valid.

Attachment'2 to Document Control Desk Letter-ISP: 920002 Page 12 Mf 26 2.1.4.5 Inadvertent Operation of the Emergency Core Cooling System During Power Operation (FSAR Section 15.2.14)

For this ANS Condition 11 event, a spurious Safety Injection System (SIS) signal is assumed to be generated at full power. The injection of borated water into the RCS reduces core power, temperature and pressure until the reactor trips on low pressurizer pressure. The power and temperature reduction causes a similar reduction in pressure on the secondary side. The analysis demonstrates that the minimum DNBR remains above the limit value.

SGTP potentially impacts the spurious SIS actuation event by reducing the RCS-flow rate assumed in the analysis. Reductions in RCS flow potentially decrease the minimum DNBR calculated during the event.

Small changes to the steady state RCS flow and secondary operating conditions (due to SGTP) would have no significant impact on the analysis results.

The licensing basis analyses shov that the DNBR is never less than the initial value. Also, the peak RCS pressure is well below the limit value and there is a large margin to pressurizer fill. Since the change in RCS flow will not affect the analysis results, it is concluded that the conclusions in the FSAR remain valid for 18% average SGTP.

2.1.4.6 Inadvertent Loading of a Fuel Assembly Into An Improper Position (FSAR Section 15.3.3) for the event presem ed in thc FSAR, the loading of a fuel assembly into an improper position would affect the core power shape. Since the power shape and not the total power generated would be affected, and the primary RCS conditions remain _the same from those assumed in the licensing basis analysis presented in the FSAR, the results of this analysis are unchanged. Thus, the current licensing basis cnalysis continues to be valid for average SGTP of 18%, and the conclusions in the FSAR remain' valid.

2.1.4.7 Single Reactor Coolant Pump Locked Rotor (FSAR Section15.4.4)

This Condition IV event is analyzed under full power conditions to quantify the peak RCS pressure, peak clad temperature, and determine the extent of DNB assuming the instantaneous seizure of one RCP rotor.

The locked rotor results in a rapid RCS flow reduction and pressure rise which may result in some amount of DNB. The reactor is promptly tripped on a low flow signal. The analysis currently in the FSAR

Attachment 2 to Document Control Desh Letter TSP 920002 Page 13 of 26 demonstrates that no more than 15% of the rods experience DNB and that the peak RCS pressure remains below that which would cause stresses to exceed the faulted condition stress limits. The FSAR consequences for this event may increase with 18% SGTP since the locked rotor could conceivably occur in the 7'op having the lowest level of SGTP and thus the greatest fraction of initial TDF.

Following the initiation of the event, a net reduction in the RCS TDF occurs since forced flow would be maintained through the two loops containing steam generators having greatest plugging levels.

A reduction in TDF could have a negative impact on the pressure and clad temperature results of the FSAR analysis. However, for VCSNS, the system transient is not expected to be significantly impacted by the small reduction in TDF (1.7%). The licensing basis analysis reports a PCT of 19740F and a peak RCS pressure of 2605 psia; these consequences are well below the acceptance limits of 27000F and 2750 psia, respectively. Existing sensitivity studies, applicable to VCSNS, show that for a 1.7%

decrease in flow, the peak pressures will remain unchanged while the current PCT margin will decrease by approximately 10*F. Thus, the primary and secondary side peak pressure and fuel PCT limits will continue to be met.

in addition, the Rods-in-DNB calculation, which is based on MMF similar to the other DNB-related events, will be unaffected by the flow reduction since this calculation is based on flow as a fraction of initial core flow and the generic DNB margin allotted for DNB events (Section 2.1.1) is also applicable to the Locked Rotor analysis.

Therefore, 18% average / 20% peak SGTP will not elter the conclusions presented in the FSAR for the locked rotor event.

2.1.5 Setpoint Impact The impact of 18% average / 20% peak SGTP levels on the VCSNS non-LOCA accident lyses have been presented; however, consideration of t .ffects of plugging levels on the overpower and over _..,arature reactor protection functions must also be addressed. The asynmetric steam generator plugging levels discussed in the intrcduction could create flow and inlet temperature asymmetries between the RCS loops. However, each channel is being used to determine the AT in individual coolant ', oops under specific loop inlet and outlet conditions.

Since only one channel exists in each loop, AT's may vary from loop-to-loop; however, the K-terms in the overtemperature diT setpoint equetion (i.e., K1, K2, and K3) and in the overpower I

Attachment 2 to Document Control Desk Letter TSP 920002 Page 14 of 26 AT setpoint equation (i.e., K4,- KS, and K6) will remain constant for all three loops. Since the AT setpoints are based upon a fraction of the individual loop AT's and the loop channels are individually calibrated based upon the loop temperatures, the OTAT/0 PAT reactor trip functions will continue to remain effective under the tsymmetric conditions considered in this evaluation.

The calculations performed for this evaluation to demonstrate the adequacy of the actual 0 TAT and OPAT protection setpoints themselves concluded that the reduction in MMF will necessitate small changes to the OTAT setpoint equation presented in the Technical Specifications (e.g., the K1. TA, and Z terms). ~'

calculation also determined that the slope of the positive wing of the F(AI) penalty function, used in the OTAT setpoint, was affected. Thus, the slope, as presented in the Technicel Specifications, will be modified. Finally, the calculation showed that the current OPAT setpcint was sufficient. With the setpoint changes identified above, the related analyses and conclusions presented in the FSAR will remain valid.

To support operation with 18% average /20% peak SGTP, the condition listed below will be imposed:

(1) The overtemperature AT reactor trip channels will be calibrated during power op3 ration in terms of boch the AT and Tavg indicated by each channel at nominal full power.

(This calibration preserves the ability of the reactor protection system to prevent exceeding the : ore safety limits in the presence of a potential asymmetry in loop temperatures.)

2.1.6 Conclusion The impact on the non-LOCA licensing-basis analyses o) plant operation with an 18% average SGTP level (20% peak) has been examined. It has been concluded that operation under these conditions will have no adverse impact upon the non-LOCA licensing basis analyses provided the pihnt Technical Specification changes identified above are made. Given the above, all of the applicable licensing basis criteria will continue to be met and the conclusions presented in the FSAR will remain valid.

The impact of asymmetric SGTP on the overpower and overtemperature reactor protection functions has also been considered and it has been determined that small changes in the K1 value for the Overtemperature AT setpoint equation is necessary along with a change to the F(AI) function. These small changes will preserve the results of the current non-LOCF analyses. Therefore, all licensing-basis criteria continue to be met and the conclusions in the FSAR remain valid, l,

Attachment 2 to Document Control Desk Letter i

TSP 920002 Page 15 of 26 2.2 LOCA ACCIDENT ANALYSES Each LOCA and LOCA-related licensing basis event is discussed in detail below.

2.2.1 Large Break LOCA The current licensing basis large break LOCA analysis for VCSNS was performed with the NRC approved 1981 Evaluation Model plus BASH, Reference 3. The peak clad temperature analysis assumes a core power level of 2775 MWt with total peaking factor (FQ) of 2.45, a hot channel enthsipy rise f actor (FAH) of 1.62, and a thermal design flow of 277,800 gpm. An initial RCS pressure of 2280 psia was assumed with a hot leg temperature of 620.50F. Past sensitivity studies have demonstrated that LOCA consequences are limiting at RCS initial conditions corresponding to maximum SGTP. -Since the curred LBLOCA analysis assumes a SGTP level of 20% in each steam generator, operation of VCSNS with an average SGTP level of 18% will not advarsely impact the FSAR large break LOCA analysis results provided plugging is limited to equal to or less than 20% in any one steam generator.

2.2.2 Small Break LOCA The current licensing basis small break LOCA analysis for VU NS was performed with the NRC approved evaluation model described in Reference 4 which utilizes the NOTRUMP :omputer code, References 5 and 6. The peak clad temperature analysis assumed a core power level of 2775 MWt with a total core peaking factor (FQ ) of 2.45, a hot channel enthalpy rise factor (FaH) of 1.62, and a thermal design flow of 277,800 gpm. An initial RCS pressure of 2280 psia was assumed with a hot leg temperature of 620.SoF. Past sensitivity studies have demonstrated that LOCA consequences are limiting at RCS initial conditions corresponding to maximum SGTP. Since the current SBLOCA analysis assumcs a SGTP level of 20% in each steam gener:.cor, operation of VCSNS with ar, average SGTP level of 18% will not adversely impact the FSAR small breaN LOCA analysis results provided plugging is limited to equal to or less than 20% in any one steam generator.

2.2.3 Post-LOCA Long Term Core Cooling References 7 and 8 present the Westinghouse licensing position for satisfying the requirements of 10 CFR 50.46 Paragraph (b), Item (5), "Long Term Cooling." The Westinghouse position concludes that the core will rema'.i shut down by borated ECCS water residing in the RCS/ sump after a LOCA. Since credit for the control rods is not taken for a large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a baron concentration that, when mixed with other water sources, will result in the ,

reactor core remaining subcritical assuming all control rods l

Attachment 2 to Document Control Desk Letter

TSP 920002 Page:16 of 26 out. The calculation is based upon the steady state conditions at the initiaf Mn of a LOCA and considers sources of both borated and unborated fluid in the post-LOCA containment sump. The steady state conditions are obtained from the large break LOCA analysis. Since the RCS is a net  !

dilution source for the mixed sump, maximizing the RCS volume is conservative. The post-LOCA core cooling was calculated with 0% tube plugging and is conservative relative to any calculations with tube plugging. Therefore, operation of VCSNS at an average SGTP level of 18% will not adversely impact tha FSAR post-LOCA long term core cooling.

2.2.4 Hot Leg Switchover to Prevent Potential Boron Precipitation The post-LOCA hot leg recirculation time is determined for inclusion in emergency operating procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This time is dependent on power level, and the RCS, RWST, and accumulator water volumes and on their associated boron concentrations. Since the RCS is a-net '

dilution tource for the mixed sump, minimizing the RCS volume will yield the highest boron concentration and is conservative. Therefore, the proposed operation of VCSNS at 18% steam generator tube plugging is bounded by the 20% steam generator tube plugging assumed in the hot leg switchover calculation. Thus, operation of VCSNS at a-steam generator tube plugging of 18% will not adversely impact the FSAR post-

, LOCA long term core cooling.

L l

2.2.5 LOCA Hydraulic Forces The LOCA hydraulic forcing functions have been evaluated for an increase in the average SGTP leve' up to 18% with a peak of 20%. As a result of the cold leg temperature differences (Case 1 Table 1 vs Table 2), a postulated increase in the magnitude of the peak LOCA hydraulic forces was determined.

l An extensive reanalysis is not required since break area ,

l margin exists and a Leak Before Break analysis (Reference 9) has been performed and submitted to the NRC. With these items credited, it is concluded that an increase in the l average SGTP level to 18% (20% peak) will not adversely affect the LOCA hydraulic forcing function results for_ the l VCSNS.

2.2.6 LOCA Conclusion i The impact of operating VCSNS at 18% steam generator tube j plugging has been evaluated for the following analyses addressed in the FSAR:

l

• LARGE BREAK LOCA

!

• SMALL BREAK LOCA

• POST-LOCA LONG TERM CORE COOLING

• HOT LEG SWITCH 0VER TO PREVENT POTENTIAL BORON PRECIPITATION l-I l

Attachment'2 to Document Control Desk Letter

~

TSP 920002

. Page 17 of 26

• LOCA HYDRAULIC FORCES i

It was shown that the consequences of these accidents _will not be increased provided the peak plugging vel in a single steam generator does not exceed 20%. Thus, i' in be concluded that operation of VCSNS at up to 18% averags .TP is acceptable with respect to the LOCA and LOCA-related o 41yses and that the-VCSNS Large and Small Break analyses of record remain bounding.

Therefore, the conclusions presented in the VCSNS FSA9 remain valid.

2.3 STEAM GENERATOR TUBE RUPTURE ANALYSIS l Both the current Steam Generator Tube Rupture (SGTR) event presented in the FSAR and the SGTR analysis submitted for NRC review and approval (Reference 10) are unaffected by the proposed increase in SGTP. The small changes in RCS flow and loop temperatures as a result of increased SGTP are not critical parameters used in the SGTR analysis.

Primary pressure, safety injection flow, and MSSV set pressures are the more important accident assumptions and none of these are affected by the increase in SGTP. Since the increase in SGTP from 15% to 18%

average (20% peak) and the reduction in MMF do not adversely impact the current and pending analyses, the results and conclusions in the VCSNS FSAR and Reference 10 remain valid.

2.4 CONTAINMENT INTEGRITY 2.4.1 Short Term Mass and Energy (M&E) Release Evaluation Containment subcompartment analyses are performed-to demonstrate the adequacy of containment internal structures and attachments when subjected to dynamic localized pressurization effects that occur during the first 3 seconds following a design basis pipe break accident. Subsequent to the postulated rupture, the pressure builds up at a faster rate tnan the overall containment pressure, thus imposing differential pressure across the walls of the structure.

If Leak Before Break (LBB) technology is credited (Reference 9), a reanalysis of RCS loop breaks is not required since they are eliminated from consideration. Smaller RCS nozzle breaks become limiting. The break sizes associated with the surge line, RHR, and the accumulator nozzles, are significantly less than the RCS loop breaks. The decrease in mass and energy releases from the smaller RCS nozzle breaks (as compared to the larger break sizes) more than offsets the initial RCS condition penalties associated with the tube plugging. For the current M&E releases for the RCS' loop breaks, the original design basis would remain bounding.

Surge and spray line breaks are also a part of the VCSNS L licensing basis. The mass and energy releases for the breaks

! are an issue for the pressurizer and pressurizer surge tank

! compartments and are impacted by the RCS temperature changes i

t i

~

Attachment 2 to Document Control Desk Lett'r TSP 920002'

. Page 18 of 26 due to increased SG tube plugging. Analyses which bound the impacts of increased plugging have been performed.

Subcompartment pressurization studies, assuming the current >

mass and energy releases increase by 15% for the pressurizer surge line break and 10% for the spray line break, demonstrate that the structural design differential pressures for the compartments are not compromised. The margins between the calculated pressure differential pressures and the structural design differential pressures remain high: 53.2% for the pressurizer compartment and 67.6% for the surge tank compartment. Based on the results of these bounding analyses, it is concluded that the changes in short term mass and energy releases due to increased plugging are acceptable. _

2.4.2 Long Term Mass and Energy Release Evaluation The long term mass and energy analyses conservatively neglect steam generator tube sleeving or plugging. The effect of sleeving / plugging in relation to containment integrity analyses would be to reduce the available reactor coolant system volume which could be displaced into the containment, and also to increase the resistance to flow through the str a generator.

This effect would exhibit less severe conditi' , in the containment. The vessel average temperature remains unchanged for the 18% average tube plugging case therefore the initial RCS inventory remains essentially unchanged. Based on this information, it is concluded that the steam generator tube plugging will have no adverse effect on the LOCA mass and energy releases for VCSNS. The current licensing basi 3 analysis results would remain bounding for this application. 1 2.4.3 Steamline Bredk Hass/ Energy Release - Inside/Outside Containment Various steam line break cases are analyzed for the purposes of.

generating mass and energy release rates which are then applied to containment response or compartment environmental analyses.

Cases are performed assuming various break sizes and initial power levels. Four major factors influence the release of mass

, and energy following a steam line break. These are steam generator fluid inventory, primary to secondary heat transfer, protective system operation, and the state of the secondary 4 fluid blowdown. For M/E releases inside containment, the reduction in TDF and increase in SGTP reduce the initial mass in the steam generators and the primary to secondary heat-transfer. Considering these effects, the reduction in TDF and an increase in SGTP to 18% plant average (20% peak) will not adversely affect the existing steamline break mass and energy releases inside containment.

For the M/E release outside containment used for equipment environmental qualification, the reduction in TDF and increase in SG1P reduce the initial mass in the steam generators resulting in earlier tube bundle uncovery and subsequent i

l

Attachment 2 to Document Control Desk Letter TSP 920002 Page 39 of 26 superheat. However, the reduction in T0F and increased SGTP also reduce the primary to secondary heat transfer. -Given these competing effects, the reduction in TOF and an increase in SGTP to 18% plant average (20% peak) will not adversely affect the existing steamline break mass and energy releases outside containment. Reference 11 also indicates that small variations in the RCS loop flow will have a negligible impact on the generation of the mass and energy releases outside containment. Therefore, 18% plent average SGTP (20% peak) will have little or no impact on the mass and energy releases both inside and outside containment.

2.5 COMP 0NENTS EVALUATION The scope of this safety evaluation is limited to the impact of the 18%

SGTP on the primary system mechanical components and fluid systems.

VCSNS is currently licensed for 15% average SGTP. Table 1 contains the Power Capability Parameters through 15% average steam generator tube plugging. Table 2 contains the parameters for 18% average SGTP. It is important to note that the change.in the vessel outlet temperature (THOT) with 18% SGTP is less than 20F greater than the TH0T temperature previously evaluated.

2.5.1 NSSS Design Transients The FSAR NSSS design transients, generated for component stress evaluations, were evaluated to determine the impact of the increase in SGTP. The proposed operating conditions result in a slight increase in the hot leg temperature and slightly lower secondary steam pressure and temperature than those used in the existing component stress analyses.

However, since this evaluation is only required until steam generator replacement is implemented and given that the plant will be operating in a base load mode for the majority of the duration, it is judged that the design transients as currently defined do not require modification.

With respect to auxiliary equipment, the applicable D3 steam generator design transients for auxiliary equipment are either unchanged or the current analyses remain bounding.

2.5.2 Reactor Vessel The VCSNS reactor vessel has been evaluated for the structural and fatigue effects of the revised parameters associated with 18% SGTP until steam generator replacement. The reactor vessel design analysis conservatively considered a vessel inlet temperature (TCOLD) of 5500F and a vessel outlet temperature (TH0T) of 6200F as the reactor vessel normal operating temperatures.

The inlet temperature with 18% SGTP remains within the envelope of TCOLD temperatures covered by the design analysis barring any changes to the design transients. This TCOLD envelope includes vessel inlet temperatures in the range of 5500F up to zero load temperature of 5570F.

Attachment 2 to Document Control Desk Letter F TSP 920002

, Page 20 of 26 l

The vessel outlet temperature with 18% SGTP is 620.40F, which is slightly higher than the THOT considered in the reactor vessel stress report. The outlet nozzles, which are the only reactor vessel pressure boundary regions with a normal operating temperature at TH0T were evaluated for the effects of the increase in outlet temperature. The evaluation concluded that neither the maximum range of primary plus secondary stress intensity nor the maximum cumulative fatigue usage factor for the outlet nozzles increase as a result of the increased outlet temperature. Therefore, all of the locations in the VCSWS reactor vessel analyzed in the stress report continue to satisfy the stress intensity and fatigue usage factor limits of Section III of the ASME Boiler and Pressure Vessel Code (1971 Edition). Reactor operation with 18% average '

and 20% peak steam generator tube plugging levels is fully reconciled with the reactor vessel stress report.

2.5.3 Reactor Pressure Vessel System The reactor pressure vessel system analysis review consists of a thermal / hydraulic assessment and a structural assessment.

The reactor internals components remain in compliance with the current design requirements with the slightly higher reactor coolant temperatures until steam generator replacement.

Fatigue usage for the limiting components in the reactor intern &ls are not significantly affected by the slight changt in the reactor coolant system conditions, since the design transients and LOCA blowdown forcing functions do not change.

2.5.4 Pressurizer The pressurizer components were evaluated for 18% average steam generator tube plugging. The pressurizer continues to meet the stress analysis and fat'gue analysis requiicments of Section III of the ASME Code. The evaluation showed that there is no significant effect on the component stresses and fatigue analysis as presented in the original stress reports.

2.5.5 Auxiliary Equipment 2.5.5.1 Auxiliary Heat Exchanger / Tanks The regenerative heat exchanger, residual heat removal heat exchanger, moderating heat removal heat exchanger, seal water heat exchanger, excess letdown heat exchanger, letdown heat exchanger, RC drain tank heat exchanger, letdown chiller heat exchanger, and letdown reheat heat exchanger were evaluated for the 18% SGTP level. The evaluation concluded that there is no adverse impact from the increased SGTP level.

In addition to the auxiliary heat exchangers, various auxiliary tanks were provided to VCSNS. The i

Attachment 2 to Document Control Desk-Letter TSP 920002 Page'21 of 26.

only tanks that have transients identified are the boron injection tank (BIT) and the safety injection accumulators. Due to BIT elimination at VCSNS, the boron injection tank is not being used for its original design purpose, and therefore is'not 1 impacted by the increased SGTP level. Also, the transients identified for the safety injection accumulator vessels are not significant from a fatigue standpoint, and are not adversely impacted by the increased SGTP.

2.5.5.2 Auxiliary Valves The original design and qualification requirements of the auxiliary valves at VCSNS were evaluated, and it was concluded that the increased SGTP level will.

not adversely impact the original design parameters of the auxiliary valves.

2.5.5.3 Auxiliary Pumps The charging / safety injection pumps, residual heat removal pumps, chemical drain tank pump, waste evaporator condensate pump, RCS drain tank pump, waste monitor tank pump, waste evaporation feed pump, spent resin sluice pump, recycle evaporator feed pump, boron injection recirculation _ pumps (removed by BIT elimination), boric acid transfer pump, gas decay pumo, floor drain tank pump, boron-thermal generation chiller pump, waste gas compressor pack, and laundry and hot shower were evaluated for the reduced MMf. The specifications require the pumps to be qualified for pressure and temperature transients, or, if the equipment was not expected to be significantly affected by the transients, it was designed for maximum steady _ state pressures and temperatures only. The evaluation concluded that the increased-SGTP level will not affect the qualification of the auxiliary pumps.

2.5.6 Reactor Coolant Loop Piping and Primary Equipment Supports The Power Capability Parameters, thermal design transients, and LOCA loop forces are parameters that have a potential impact on the qualification of the reactor coolant loop piping and primary equipment supports. Since there is no impact on the -

l_ LOCA loop forces and the design transients do not require modification due to the increase in SGTP, there is no overall impact on the evaluation / qualification of the RCS piping and supports.

l l

Attachm nt 2 to Document Control Desk Letter TSP 9s0002 Page 22 of 26 2.5.7 Control Rod Drive Mechanism (CRDM)

The VCSNS CRDMs were evaluated for the conditions associated with 18% average SGTP. There are no changes to the NSSS design transients, and the small change in the vessel / core inlet cemperature will have a nealigible effect on the analysis of the pressure boundary components. 1herefore, it is concluded that compliance with the design criteria is not affected.

2.5.8 Reactor Coolant Pump (RCP) and RCP Motor The changes which would affect the RCP are very small. The reactor coolant temperature change is small, ar.d there is no change in pressJre or to the design transients. The temperature change has a negligible effect on the analysis of tue pressure boundary components. Compliance with the design-criteria is not affected.

The RCP motor was evaluated for operation with the revised loads, caused by the revised Power Capability-Parameters. The evaluation showed that the motor is acceptable for continuous operation at a new hot loop load of 6925 hp, since this load is less than the nameplate rating of Sou0 hp. The new cold loop load of 8890 hp represents a 1.6% increase over the motor cold loop nameplate rating of 8750 hp. However, calculations show that operation at a maximum of 8890 hp is acceptable. The motors are acceptable for operation during the worst case starting scenario; a cold loop 75% voltage start with maximum reverse flow of 30100 gpm. Based on the RCP motor analysis, the motors are suitable for operation at the revised loads a'.sociated with 18% SGTP.

2.5.9 Steam Generator

2. 5.9. 's Structural Considerations The critical components of the steam generator that were evaluated are as follows:

- Tubesheet

- Channel and Channel Head Divider Plate

- Feedwater Nozzles

- Primary Manway

- Secondary Manway

- Tubes

- Tube-to-Tubesheet Weld

- Steam Nozzle The evaluation assumed that primary side reactor coolant pressure remains unchanged at 2250 psia, while steam pressure and steam aperature values decrease to the lowest acceptable values, 894 psia l and 531.20F, respectively. These values were determined to achieve 100% power with 18% average l SGTP and 20% peak SGTP.

Attachment 2 to Document Control fesk Letter TSP 920002

. Page 23 of 26 for the tubesheet, the center of the tubesheet and tubesheet/shell are considered for the evaluation of the effect of steam generator tube plugging level.

The results of the analysis show that the maximum stress intensities expected for the VCSNS steam generator tubesheet and tubesheet/shell junction remain within the allowable stress intensities for both normal and upset condition loadings.

For the steam generator channel head and channel head divider plate, in particular for the divider plate weld which is typically highly stressed, resultant stress intensity ranges have been determined to be within acceptable limits for both normal and upset condition loadings. An assessment of the resulting fatigue usage in the critical areas of the divider plate shows acceptable usage.

Relative to the main and auxiliary feedwater nozzles, the analyses show that the conditions associated with 18% SGTP level do not significantly impact the existing stress analyses, it is shown that the stress intensities and fatigue usage remain within allowable limits for both normal and upset condition loadings.

The primary and secondary manway stress intensities and fatigue considerations have been reevaluated taking into account the 18% SGTP level. The impact of the revised parameters is found not to De significant. The cxisting analysis for the manway fasteners is judged to be conservative.

The impact of the reduced steam pressure on the tube stresses has been assessed. Existing stress analyses show that the highest calculated tube stresses are within acceptable limits. Nonetheless, ihe fatigue analysis has been modified to account for the revised steam pressure due to the 18% SGTP level. Results of this analysis show that the tubes meet the ASME Code requirements.

The tube-to-tubesheet weld has been evaluated in order to determine the stresses under the conditions associated with the 18% SGTP level. The VCSNS steam i generator tubes extend beyond the primary surface of the tubesheet cladding by 0.22 inch and are fillet welded to the tubesheet cladding. It was judged that the geometry of the tube-to-tubesheet weld required an evaluation for the Normal / Upset cundition transients. The results of this evaluation show that the Normal / Upset condition transient loadings are acceptable.

Attachment 2'to Document Control Desk Letter TSP 920002 Page 24 of- 26 Finally, the impact of the_18% SGTP~1evel on the steam outlet nozzle has been assessed. The VCSNS steam genera' rs have steam outlet nozzles with integral type flow restrictors. Normal / Upset condition transients were evaluated to show that existing analyses are not significantly impacted by the parameters associated with the 18% SGTP.

In summary, a structural evaluation of the critical components of the steam generator has been performed to address safe plant operation at conditions associated with the 18% SGTP. The evaluation-concludes that the critical cotaponents of the steam generator meet the requirements of the ASME Code,Section III, Subsection'NB.

2.5.9.2 Tube Fatigue Considerations The VCSNS steam generators have been e sluated for--

the st:sceptibility to a f atigue rupture of the type experienced at Row 9 Column 51 (R9C51) of Steam Generator C at North Anna Unit 1. The evaluhtion used-Eddy Current Test (ECT) data supplied by SCE&G and interpreted by Westinghouse.

Based upon a review of the operational, inaintenance, and modification installation factors-associated

,_ with the modification options, SCE&G decided to l install cable _ dampers into-the hot legs of seven tubes listed below. Some row dependency was observed in previous damper testing, and data of-the -

added damping provided by the cable dampers were not

available for tube rows requiring ameliorativc-L action at VCSNS. Therefore, it could not be demonstrated that the cable dampers provided sufficient added damping to assure that a tube viith l, the cable damper and solid plugs installed would not I

fail.due to high cycle fatigue and subsequently-create the potential for damage to adjacent tubes.

-Therefore, in April 1990, sentinel _ plugs were

installed on the cold leg side of each of the seven tubes in addition to hot leg solid plugs and the cable dampers. The tubes with sentinel plugs installed do not require reanalysts for the 18% tube 91ugging conditions. .The seven tubes are

SG-A

'C83, SG-B R9C56, SG-C R12C5, SG-C R13C5, SG-C M10C23, SG-C R9C29 and SG-C R9C106.

Based on an evaluation of the tubes remaining in service _in the VCSNS steam generators,'the. remainder of the tubes are not expected to be susceptible to high-cycle fatigue rupture at the top tube support plate in a manner similar to the rupture which occurred at North Anna Unit 1. This is based on the

+

Attachment 2 to Document Control Desk Letter TSP 920002

. Page 25 of 26 assumption that the Turbine Valves Wide Open limit-is not reduced below 905 psia through the end of the current operating license with power levels at or below 2787 MWt.

An additional fatigue evaluation of the VCSNS tubing has been performed to examine the effect of reduced steam pressure (894 psia) at 18% SGTP level until steam generator replacement is implemerted, with the objective of demonstrating that the tube with-highest _ stress ratio remaining in service does not exceed t' 1.0 fatigue usage criterion. A parametric analysis has been performed to determine the relative stability ratio (RSR) multipliers and stress ratios for full power steam pressures below the reference steam pressure used for prior 3D analyses. It has been determined that the limiting tube, SG-A R9C55, does not exceed the 1.0 fatigue usage (end of Cycle 8) criterion for steam pressures at or abo,e 894 psia. . For the reduced power levels associated with operation below the assumed 894 psia limit, the RSR multipliers oecrease, and the fatigue results remain bounded. Therefore, assuming that the worst case tube (SG-A R9C55) is subjected to the most stringent. tube stability condition since startup, all tubes remaining in service at VCSNS are acceptable for operation through the completion of

! steam generator replacement with power levels at or l

below 27S7 MWt.

2.5.10 Fluid Systems The Reactor Coolant System, Chemical and Volume Control System, Residual Heat Removal System, and Safety Injection System were evaluated to determine if any parameters which changed as a result of the increased SGTP level would affect the design C

adequacy of those systems. The results of the evaluation showed that the systems reviewed are adequate and acceptable for 100% power operation l with 18% average steam generator tube plugging.

2.5.11 Componer.ts Conclusion

! Operation of VCSNS with 18% aurage steam generator tube plugging will not have an adverse effect on primary pressure boundary integrity. The increased SGTP ooes not represent an unreviewed safety question in accordance with 10 CFR 50.59 criteria, and the reduced Minimum' Measured Flow does not represent a significant hazard as defined in 10CFR50.92.

Attachment 2 to Document Control Desk Letter ETSP 920002

. Page-26 of 26 REFERENCES

1. - "Virgil.C. Summer Nuclear Station Final Safety Analysis Report" tN ough Amendment 92-06.

2. USNRC Letter, G. F. Wunder to J. L. Skolds, " Issuance of Amendment No.

105 to facility Operating License No..NPF-12 Regarding Vantage Plus Fuel Reload - Virgil C. Summer Nuclear Station, Unit No. 1 TAC NO. 79121,"

October 22, 1991.

3. WCAP-10266 REV.2 with Addenda (Proprietary), "The 1981 Version of the Westinghouse ECCS Evaluation Model using the BASH Code", August, 1986.

4. WCAP-10054-P-A (Proprietary), and WCAP-10081-A (Hon-Proprietary),

"Westinchouse Small Break LOCA ECCS Evaluation Mosel Using the NOTRUMP-Code", August, 1985.

5. WCAP-10079-P-A (Proprietary), and WCAP-10080-P-A (Non-Proprietary),

"NOTRUMP, A Nodal Transient Small Break and General Network Code", ,

August 1985.

6. WCAP-11145-P-A (Proprietary), and WCAP-11373-A (Non-Proprietary),

" Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the 10 TRUMP Code", October 1986.

7. WCAP-8339 (Non-Proprietary), " Westinghouse Emergency Core Cooling System Evaluation Model - Summary", June 1974.

8. Westinghouse Technical Bulletin NS10-TB-86-08, " Post-LOCA Long-Term f 'ing: Boron Requirements", October 31, 1986.

9. ..ter from J.L. Skolds, SCE&G, to USNRC." Virgil C. Summer Nuclear Station, Docket.No. 50/395 Operating _ License No. NPF-12, Request to L Utilize Leak Before Break ~For Reactor Coolant System Piping,"~ June 22, 1992.

10. Letter from J.L. Skolds, SCE&G, to USNRC," Virgil C. Summer-Nuclear Station, Docket No. 50/395 Operating License No. NPF-12, Technical Specifications Change Request - Main Steam Safety Valve Setpoint Tolerance Change ~(TSP 880018-0)," February 4, 1991.

11. WCAP-10961. Revision 1 "Steamline Break Mass / Energy Releases for l Equipment Environmental Qualification Outside Containment," October k 1986, l

l

H Attachment 3 to Document Control Desk letter i

TSP 920002

. Page 1 of 4 PROPOSED TECliNICAL SPECIFICATION CHANGE - TSP 920002 VIRGIL C. SUMMER HUCLEAR STATION DESCRIPTION AND NO SIGNIFICANT HAZARDS DETERMINATION DESCRIPTION OF LICENSE AMENDMEH1 REQUEST A) Purpose for Change Changes to the VCSNS Technical Specifications are being requested to permit an increase in the maximum permissible average level of Steam Generator Tube Plugging (SGTP) from 15% to 18%. Although a value for SGTP is not specified in the Technical Specifications, the increase in average SGTP will result in a 1.7% decrease in the Minimum Measured Flow (tNF) value which is referenced in Technical Specification 3/4.2.3. This amendment request reflects the impact of a reduction in HMF on the VCSNS Technical Specifications. Specifically, changes are required te Table 2.2-1 for a constant and a setpoin! reduction penalty in the OTAT setpoint equation, the OTAT trip total allowance And Z, and the loop design flow.

B) Current License Condition Technical Specification 3/4.2.3 defines allowable conditions of

• indicated Reactor Coolant System (RCS) total flow and R (a parameter related to FNAH). This region of allowable operation is specified each cycle in the Core Operating Limits Report (COLR) figure entitled "RCS Total flow Rate Versus R for Three Loop Operation."

Technical Specification Table 2.2-1 lists the reactor protection system instrumentation trip setpoints for the various trip functions.

Functional Unit 7 represents the trip setpoint and allowable valve for the Overtemperature AT reactor trip. This trip function is not a constant value but is continuously calculated by the reactor protection system. Note 1 for this trip defines the OTAT setpoint equation.

Functional Unit 12 (Loss of Flow) provides the trip setpcint and allowable value for the loss of flow reactor trip function which is footnoted to identify a loop design flow. The loop design flow corresponds to one-third of the required Minimum Measured Flow (MMF) at Rated Thermal Power (RTP) - 96,500 gpm/ loop.

C) Function of Identified Technical Specifications Tha purpose of the COLR figure for RCS flow vs R for three loop operation is to allow RCS flow rate and power to be traded off-against one another while still ensuring that the actual DNBR will not be bclow the design DNBR value.

Attachment 3 to Document Control Desk-Letter TSP 920002 A Page 4 of 4 respect to the primary pressure boundary is provided, in part, by the safety factors included in the ASME Code. The components evaluated remain in compliance with the codes and standards in effect when VCSNS was originally licensed. Thus, there is no reduction in the margin of safety as defined in the bases of the VCSNS Technical Specifications.

Attachment 3 to Document Control Desk Letter TSP 920002 i Page 2 of 4 The W AT trip function provides sufficient core protection to preclude DNB over a range of operating and transient conditions. The setpoint is automatically varied with temperature, pressure, and the axial power distribution. The F(41) penalty function adjusts the trip setpoint for axial peaks greater than design.

The footnote to the loss of flow reactor trip values is used to identify, in absolute terms, the flow values at which the trip function should actuate. Since tM trip setpoint and allowable value are given in terms of relative flow (90% eno 88.9%, respectively), the footnote can be used to determine actual flowrates which preserve the basis of the accident analysis.

D) Description of Proposed Change for Cycle 8, the COLR figure for RCS flow vs R for three loop operation will be revised to reflect a reduction in Minimum Measured Flow at RTP from 289,500 gpm to 284,600 gpm. This COLR change will be implemented in accordance with the requirements of VCSNS Technical Specification 6.9.1.11. This amendment request reflects the impact of a reduction in MMF on the Technical Specifications. The OTAT trip values for total allowance and Z are revised from 9.8 and 7.21 to 10.3 and 7.8, respectively. With respect to Note 1, the value for the K1 term and the positive wing of the F(AI) penalty are also affected by the proposed reduction in flow. The value for the K1 term in the OTAT setpoint equation is reduced from 1.203 to 1.195 and the slope of the positive wing of subsection iii of Note 1 is increased from 2.13% to 2.34%. The nominal value of loop design flow in the footnote to Table 2.2-1 Functional Unit 12 is revised to reflect the change in MMF, These changes will permit the Technical Specifications to allow acceptable plant operation with a reduced MMF of 284,600 gpm at RTP.

NO SIGNIFICANT HAZARDS DETERMINAT od The technical specification changes associated with the reduced Minimum Measured Flow have been reviewed and deemed not to involve significant hazards considerations. As discussed below, all applicable LOCA, non-LOCA, and SGTR design basis acceptance criteria are satisfied and the conclusions presented in the VCSNS FSAR remain valid. Thus, the proposed Technical specification changes do not constitute an unreviewed safety question and the accident analyses support the changes. The basis for this determination follows:

1) Operction of VCSNS in accordance with the proposed license amendment-does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Since the RCS flowrates (MMF and TDF) are lot reduced to a value below that assumed in or evaluated against the licensing bases of the plant and the proposed modification involves only minor changes to primary side operating parameters and the heat transfer capabilities of the steam generators, an increase in the maximum average SGTP level from 15% to 18% (including 20% peak) will not result in any additional challenges to plant equipment. The assessment of the NS$$ primary

Attachment 3 to Document 'introl Desk letter TSP G20002 li Pace 3 of 4 components, including the re3ctor pressure vessel system, reactor coolant pump, steam generator, pressurizer, Control Rod Drive Mechanisms, and RCS piping, concluded that the integrity of the components will be unaffected by the increase in average SGTP level.

Also, evaluations of the Reactor Coolant System, Chemical and Volume Control System, Residual Heat Removal System, and Safety Injectior.

System concluded that the increase in SGTP will not adversely impact the adequacy of the systems. In addition, SGTP does not adversely impact steam generator tube integrity.

With respect to the accident analyses, all of the applicable LOCA, non-LOCA, and SCTR design basis acceptance criteria remain valid for the transients evaluated. The DNBR and PCT values remain within the specified limits of the licensing basis, and the current mass and energy release data used for contair,nent integrity and equipment qualification remain bounding. Finally, no new limiting single failure is introduced by the proposed change.

Since the design requirements and safety limits continue to be met, system functions are not adversely impacted, and the integrity of the reactor coolant system pressure boundary is not challenged, the assumptions employed in the calculation of the offsite radiological doses remain valid. Thus, the probability of occurrence and the radiological consequences of the accidents previously evaluated in the VCSNS FSAR remain unchanged.

2) The proposed license amendment does not create the possibility of a new or dif ferent kind of accident from any accident previously evaluated.

The increase it the maximum average level of SGTP from 15% to 18% (20%

peak) and subsequent reduction in MMF will not introduce any new accident initiator mechanisms. No new failure modes or limiting single failures have been identified. Since the safety and design -

requirements continue to be met and the integrity of the reactor coolant system pressure boundary is not challenged, no new accident scenarios have been created. Therefore, an accident which is different from any already evaluated in the FSAR will not be created as a result of this change.

J

3) The proposed license amendment does not involve a significant reduction in a margin of safety.

As discussed in the evaluation above although the proposed increase in the average SGTP level to 18% (20% peak) will require a change to the plant Technical Specifications, it will not invalidate the LOCA, non-LOCA, or SG Tube Rupture conclusions presented in the FSAR accident analyses. For all the FSAR non-LOCA transients, the DNB design basis, primary and secondary pressure limits, and dose limits continue to be met. The LOCA peak cladding temperatures remain below the limits specified in 10CFR50.46. The calculated doses resulting from a SG Tube Rupture event will not change and continue to remain within a small fraction of the 10 CFR 100 permissible releases. The current mass and energy release data used for containment integrity and environmental qualifica$ ion will remain applicable. The margin of safety with

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