ML20092D693

Proposed TS Re Reduction in Min Measured RCS Flow

ML20092D693
Person / Time
Site:	North Anna
Issue date:	02/10/1992
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20092D689	List: ML20092D688 ML20092D693
References
NUDOCS 9202140190
Download: ML20092D693 (50)

Text

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l ATTACHMEtJT 1 PROPOSED TECHNICAL SPECIFICATIOtJS CHANGES NORTH ANNA UNIT 1 VIRGINIA ELECTRIC AND POWER COMPANY 9202140190 920210 gDR ADOCK 0500 8 ,

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• 2.0 SAFETY UMITS AND UMITING SAFETY SYSTEM SETTINGS jlt SAFETY LIMITS BEACTOR CORE 2.1.1 The combination of THERMAL POWER, prescurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the lirrits shown in Figures 2.11* for 3 loop operation and 2.12 and 2.13 for 2 loop operation.

APPLICABILIfY MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average ,

temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR. COOLANT SYSTEM PRES 3URE 2.1.2 The Reactor Coolant System pressure st,all not exceed 2735 psig.

APPLICABILRY: MODES 1,2,3,4 and 5.

I ACTION: '

MODES 1 and 2 i Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3,4 and 5 Whenever the Reactor Coolant. System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

i For the period -of operation until steam generator replacement, the combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.11a.

NORTH ANNA- UNIT 1 21

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, PERIOD OF OPERATION UNTil STEAM GENERATOR REPLACEMENT '

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ATTACHMENT 2 DISCUSSION OF PROPOSED CUn"GES l

l' l-VIRGINIA ELECTRIC AND POWER COMPANY l

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Discustion of Proposed Changes Bacharound North Anna Power Station Unit 1 is currently conducting a mid cycle steam generator inspection outage. An extensive eddy current inspection of the North Anna Unit 1 steam generator tubes is being performed using very conservative analysin quidelines and plugging criteria. As such, a substantial number of tubes aro expected to be plugged.

Based on projections ci steam generator tube plugging levels, it was projected that the RCS total flow rate would not meet the current Technical Specifications minimum requirement of 284,000 gpm. Therefore, by letter dated January 8,1992 (Serial No.

92 018), we proposed Technical Srecifications changec to reduce the minimvn total RCS flow rate to 275,300 gpm for the operating period until the steam genentors ,

could be replaced. This approi. mate 3% reduction in minimum RC'S flow rate wn intended to bound the effect of the increased flow resistance for the projected steam generator tube plugging levels.

According to Westinghouse estimates of RCS flow rates as a function of tube plugging 93rcentage. the proposed 275,300 gpm minimum RCS flow inte corresponds to approximately 32% average steam generator tube plugging. Because measuremem uncertainty may cause measured RCS flow rates to vary by as much as 2% from their true value, there exists an expected range of steam generator tube plugging over which the proposed flow rate may be met. This range is astimated to be between 28%

and 3G% average steam generator tube plugging. However, it shoud be emphasized that this is only an estimated range which assumes that the predictions of flow versus plugging are correct. If an inaccuracy in prediction is considered, the range of plugging over which the 275,300 gpm minimum measured flow might not be met is

. further widened.

As the steam generator tube inspections progressed, the adjustments to the tube plugging projections indicated that the "C" steam generator may exceed 30% To accommodate the effect of additional tube plugging in the "C" steam generator, we requested a change to the North Anna Unit 1 operating license to limit the maximum reactor power level to 95% of rated thermal power for the interim period of operation '

until the steam generators could be replaced. This change was requested in our letter dated January 28,1991 (.'erial No. 92 042). The imposed 95% power restriction will provide sufficient margin in the large break Loss of Coolant Accident (LOCA) analysis to accommodate the interim effects of the increased steam generator tube plugging for the most restrictive steam generator.

In as much as the steam generator tube inspections are continuing and the actual tube plugging levels are not known, prudence requires us to cover the uncertainty in the range of possible tube plugging and flow rate measurement results. Therefore, we supplement our Techr; cal Specification change request, submitted on January 8, Page 1 of 3

i 1932 (Sorial No. 92 018), with e request to further reduco the minimum total RCS flow rato to 268,500 gpm. This change takes credit for the 5% power reduction discussed la our roquest for licenso amendment, submitted January 28,1992 (Serial No.92-042).

This Technical Specification change request supplomonts our January 8,1992 i submittal. The attached safety evaluation builds on tno ovaluation provided in our original submittal. Approval of this supplomontal change is conditional on the prior approval of both the previously discussed change requests, l.o., the January 8,1992 i and the January 28,1992 Technica! Snocification change submittals, i

ininducilon As required by Technical Specifications 3.2.5 and 4.2.5.2, North Anna Unit 1 performs reactor coolant system (RCS) flow rato measuroments subsequent to restart after each refueling. The North Anna Unit I safety analyses are based in part on verifying, via the Technical Specifications survoillanco, that the Reactor Coolant System (RCS) total .

flow rate is greater than or equal to 284,000 gallons per minute (gpm). The additional  !

steam generator tubo plugging anticipated during the current mid cycle inspection l cutage increases the likelihood of violating this Technical Sp scifications requiremont. ,

Therefore, safety analyses and evaluations have been perfrarmed which support this I additional reduction in the RCS total flow rato limit to 208,500 gpm at 95% Rated Thermal Power. The attached safety evaluation has boon prepared to support each of l the Technical Specifications changes associated with this reduction in the RCS total flow rate limit. ,

Discussion of Procosed Ctangn These proposed Technical Scocifichuons changes supplement the changes discussed !n our submittal datad January 8,1992 in support of a reduced RCS total. .

flow rate requirement to 2268,500 gpm, which is an approxirnato 51/2% reduction from the current Technical Specifications requiremont. Those changes would only be effectivo until the North Anna Unit 1 steam generator replacement, which is currently '

scheduled to begin in January,1993.

The proposed Technical Specifications changes affect Figure 2.11, Reactor Core Safety Limits (Specification 2.1.1), Table 2.21, Roactor Trip System Instrurrentation

. Trip Sotpoints (Technical Specification 2.2.1), and Table 3.21, DNB Parameters (Technical Specification 3.2.5).

The proposed Technical Specifications changes will revise Technical Specification >

2.1.1 by placing a icotnoto on the bottom of the page referencing Figure 2.1 1a in lieu of Figure 2.1-1 for the n9 actor Coro Safety Limits. A revised reactor coacafety '. nits graph (Figure 2.1 1a)is added with the inclusion of a new page 2 2a. The graph is

. changed to reflect the 268,500 qpm flow limit at 95% of Rated Thermal Power. Both the footnote and the graph title am worded to be effective for the period of operation i

until steam generator replacement.

Page 2 of 3

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The proposed change, as submitted January 8,1992, requestod a revision to the 4- footnote on the bottom of Tablo 2.21, Reactor Trip System Instrumentation Trip Setpoints, to specify that the " design flow por loop" is "one third of the minimum allowable Reactor Coolant System Total Flow Rate as specified in Table 3.2-1."

Several additionel items in Tablo 2.21 are proposed to be changed by this  :

supplement. They are:  !

Item 2, Power Rar.:Jo, Neutron Flux, the high trip setpoint is loworod from 109% to

. 103% Rated Thermal Power and the allowable value for the bl0 h trip sotpoint is ,

lowered from 110% to 104% rated thermal power. Both changes are for the period of  :

operation until steam generator replacement. These changes revice the trip setpoints so that credit can be taken for the 5% reduction in Rated Thermal Power lovel requested in our January 28,1992 submittal. The revised setpoints will allow the i necessary thetmal margin to lower the total RCS flow rate to 268,500 gpm.

Item 7, Overtemperature AY, Note 1 (page 2 0), the value for K1 is reduced from 1.264 to 1.132 for the period of operation until steam generator replacoment. The change to _l

.t his ca cu l at lion actor f rev ses i te h Overtemperature AT trip setpoints so that credit can .

b taken for the 5% reduction in Rated Thormal Power level and will ensure reactor protection with the lower the total RCS flow rate, llam 8, Overpower AT, Note 2 (page 210), the value for K4 si reduced from 1.073 to 1.016 for the period of operation until steam generator replacement. The change to this calculation factor revises the Overpower AT trip setpoints so that credit can bo taken for the 5% reduction in Rated Thermal Power level and will ensure reactor >

protection with the lower the total RCS flow rate.

Tha remaining reactor _ trip sotpoints specified in Table 2.2-1 will continue to ensure that tho safety analysis Tsumptions will be met at the reduced RCS flow rate and '

Rated Thermal Poi;o*

L The proposed cha v. a submitted January 8,1992, requested a revision to Table 3.21, DNB Paramo. 's y adding a footnote which reduces minimum limit for Reactor ,

,. Coolant System Total how Rate from 284,000 gpm to 275,300 gpm until the North '

Anna Unit 1 steam generator replacement. With this supplemental change, we request to further lower the minimum limit for Reactor Coolant System Total Flow Rate to 268,500 gpm. This change takes credit for the 5% reduction in Rated Thermal Power level requested in our January 28,1992 submittal. The proposed interim -

p reduction in the minimum measured reactor coolant system flow is necessary to accommodate the expected increase in RCS loop resistance caused by increased steam generator tube plugging levels. Upon resumption of Cycle 9 power operation, l the RCS total flow rate will be confirmed by measurement in accordance with Technical Specification 4.2.5.2.

l The attached safety evaluation supports the above changes to the Technical l

_ Specifications. The supplemental changes to Specification 2.1.1, Figure 2.1-1, Table >

2.241, and Table 3.21 are required on an interim basis until the steam generator replacement in 1993, at which time they.will no longer apply.

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. ATTACHMENT 3 SAFETY EVALUATION ,

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IABLE OF C9RIENTS 1.0 IN1RCDUCTION i 1.1 Deckground......................................... 2 <

1.2 Summary of Analyses and Evaluations to Date...... . 3

. 1.3 Purpose of this Evaluation......................... 4 2.c EVALUAT10N.......................................,... . 7 2.1 Impact of Flow Reduction on Core Thermal Limits...... 7 2.1.1 Generation of New OTAT and 0 PAT Setpoints. . . . . . 10 2.1.2 Confirimatory Analyses of Uncontrolled Rod Withdrawal at Power........................... 12 2.1.2.1 Accident Description............ ...... 12 2.1.2.2 Method o Analysis..................... 13 2.1.2.3 Results and Conclusions............ ... 14 .

2.2 Summary of Accident Evaluations. . . . . . . . . . . . . . . . . . . . . . 16 2.2.1 Overview of Assessment Process................., 16 2.2.2 Assessment of Het Flow / Thermal Power impact on DNBR...................................... 18 2.2.3 Olscussion of Accidents By Grouping.......... ., 20

- Group 1: No flow /No Plugging impact............ . . 20

- Group 2: Accidents impacted by Tube Plugging But Not RCS Flow Rate...................... 22 ,

- Group 3: Accidents !mpacted by RCS Flos Alone...... 24

- Group 1: Accidents Potentially-Impacted by RCS Flow and Plugging......................

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3.0 NSSS AND BALANCE OF PLANT SYSTEMS AND COMPONENTS. . . . . . . 34 3.1 NSSS: Systems and Components........................ 34 3.2 Balance of Plant Systems- and Components. . . . . . . . . . . . 34

4.0 CONCLUSION

S............ ................................ 36 .

REFERENCES........................ .................... 38 4

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1.0 INTRODUCTION

1.1 Background

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North Anna Power Station Unit 1 is currently involved in a mid-cycle steam generatoi inspection outage. An extensive eddy current inspection ,

of the North Anna Unit 1 Meam generator tobes is being performed using ,

very conservative analysis guidelines and plugging criteria. As $uch, a p

substantially increased number of tubes are expected to be plugged. l The physical consequences of extended SG1P are primarily (a) increased RCS loop resistance, resulting in a lower RCS flow rate, (b) decreand -

steam generator tube heat transfer area, resulting In lower steam

,c generator' cutlet steam pressure, and (c) a decreased total RCS volume.

The impact of these changes with ree.pect to previoualy analyied design

. conditions-must be fully assessed for both normal operating and accident conditions. This. assessment is performed following e steam generator .

inspection ' outage usually concurrent with a new reload safety evaluation. -

When required, revised safety an: lyses are performed ano a Core 0perating Limits Report (COLR) _is prepared as required by Teshnical Specification-6.9.1.7. 1 In many cases, the. incorporation of-revised safety analyses into the North Anna desQn basis could.be accomplished via Virgini_a Power processes 1 employed to. assess. change per 10 CFR 50.59.- However.-based on current st.eam generator plugging projections, it is expected that the current North. Anna 1 Technical Specification RCS flow limit (284,000 gpm) could.

be violated. This could potentially occur at average SGTP levels of I

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i approximataly 20%. To address this concern a separate lechnical Specification- Amt:ndment request to reduce the RCS total .' low rate limit.

by approximately 3% was submitted for review and approval (1,2). W

' Summary of Analyses and evaivations to Date a

in Reference 1. Virginia Power proposed a Technical Specification j minimum measured flow- of 2/5,300 gpm, which i s approximately a 3%

reduction f rom the currently licensed limit of 284,000 gpm. Because this flow rato is a critical input assumption in the UFSAR Chaptor 15 analyses,  ;

it was necessary to evaluate all. Chapter .15 analyses to support the ,

implementation of extended SGTP at North Anna Unit 1. As a rtsult, reanalyses of S OFSAR Chapter 15 non-LOCA accidents were presented (1), ,

Accidents which were rur,alyzed were: .

t Loss of External Load loss of Normal Feedwater -

Rod Bant Withdrawal at Po*er g

Complete Loss of Flow Locked Reactor Coolant Pump Rotor for the balance of. accidents, evaluations bere performed on the basis of parameter sensitivities and available thermal margins. These accident -

reanalys es and reevaluations were based on assumed operation at full rated thermal power. Supplemental information relating to this evaluation was

+

provided in Reforence-2. 3 3

s 2

• r + - - # #, -,,#. .-%,, .4- , -y,. .ri. .,..-.e-.., ,% <,,% -wwy -

y #,w w --

.- 1.

Reference 3 presented a large Break t.oss of Coolant Accident (LBLOCA) reanalysis which supports operacion of Unit I with maximum Steam Generator Tube Plugging (SGTP) level of up to 35% in any steam generator. That analysis was performed based on an assumed power level of 95% of rated thermal poser.

-1.3 Purpose of this Evaluation This evaluation is being provided to supplement and extend References 1 and 2 to- support a further reduction in RCS flow to 268,500 gpm. In this extension, operation at less than or equal to 95% of rated thermal power is essumed.

According to Westinghouse estimates of RCS flow rate as a function of tube plugging percentage, thc- Reference 1 proposed RCS flow rate of 275,300 gpm corresponds to approximately 32% average tube plugging. This estimate is based on an extrapolation'of previous measured RCS flow data, tiecause RCS flow measurement uncertainty may cause nicasured flow rates

'o vary by as much as 2% from thelt true value, there exists an expected ringe of stean generator tube plugging over which the proposed flow rate

~

may be met. This range is estimated to be betwi.an 28% and 36%. However, it should.be emphasized.tbat this is only an estimated range which does not explicitiy account for inaccuracy in the prediction of flow versus plegging. If flow prediction accuracy is considered, the' cange of plugging over which the 215,300 gpm minimum n,easured flow rate requirement might not be met is further widened.

4

- 1

, i To provide for an increased confidence level that the Technical Specification flow limit will be met, Virginia Power has performed an

- additional evaluation which suppports operation of North Anna Unit I at a thermal power level not to exceed 95% of rated thermal power, peak steam generator tube p'.ugging levels of up to 35%. and a revised minimum reactor I

coolant system total measured flow rate of 268,500 gpn. Note-that the Reference 3 submittal assumed a maximum power level of 95% of rated i 1

thermal power. I This evaluatian provides an updated assessment of all of the UFSAR  ;.

Chapter 15 accidents at the conditions stated in fable 1. To support ,

operation at the 268,500 gpm measured flow rate, revised Core Thermal Limits (Technical $pec4 fications Figure 2.1-1) have been developed.

These revised limits have been used in turn to develop new Overtemperature and Overpower AT st', points. Section 2.1.1 discusses this development. '

Consistent with es tabli shed practice, confirmatory analyses of the  ?

uncontrolled rod withdrawal at power have been performed to verify DNB protection over a wide range of core thermal / hydraulic conditions. These analyses are discussed in Section 2.1.2. i The evaluations for the balance of the non-l.0CA accidents are presented in Section 2.2, The impact of the reduced flow value .on balance of plant ,

1.

and support systems is reviewed in' Section 3,0. Conclusions and references are presented in Section 4.0 and 5.0, respectively.  ;

l .

i 5

t

.w.---,. ,. . _ . - - m. . , . _ , - . _ - , . - . _ . . . _ . . , _ . . . , ~ . . - , , , _ , _ , _ _ . , , , . . ~ . , _ . . , . . , , , ~ , . -

TABLE 1 KEY EVALUATION ASSUMP110NS Initial Conditions Statistical Deterministic DNB Method Method Power 2748.35 MWt 2803.32 MWt [

Average Temperature 586.8 'F 590.8 'F RCS Flow Rate 268,500 gpm 263,130 gpm ,

Pressure 2250 psia 2220/2280 psia FAh at Assumed Power 1 512 1.573 1,55-Cosine Axial Power Profile f

w W

V k

1 1

l:

l 5

6

.. . .. . .~ .. .-. . -. .-

1 j

2.0 EVALUATION 2.1 Impact of Flow Reduction on Core Thermal 1.imits An evaluation has been performed to assess the impact of the proposed reduction in minimum measured flow rate on North Anna Unit 1 core thermal >

limits, Overtemperature and Overpower AT trip setpoints, and the f(AI) -

function.

TN current Core Thermal Limits in Figure 2.1-1 of the Technical Sr,ecifications consist of two distinct limits. The ONBR portions of the limit lines are based on a minimum measured flow of 289,200 gpm and bound a design DNBR limit of 1.46 (as opposed to a statistical DNBR limit of 1,26). The vessel exit portio,s of the limit lines are based on a thermal design flow of 278,400 gpm.

t Reference 1, an assessment was made of operation at a total measured '

RCS flow rate of 275,300 - gpm and the current Core Thormal Limits.

Operation was _ shown. to be acceptable based on the use of available

~

retained DNBR margins.

For the proposed additional reduction in minimum measured RCS flow rate-to 268,500 gpm, revised thermal limit lines have been generated based on this reduced flow and a design DNBR limit of 1.46. In this way,-no flow-reduction- penalty need be assessed against the generic retained DNBR margin for operation at 268,500 gpm, The definition and application of retained DNBR margin in Virginia Power design analyses is described-in References 1.and 2.

7 l.

i The vessel exit boiling limited portions of the core thermal limits were evaluated with the proposed reduced non-statistical (deterministic) thermal design flow rate of 263,130 gpm (which corresponds to a minimum measured flow rate limit of 268,500gpm). The revised Thermal Limits are shown in Figure 1.

t t

e d

8

- _ ____ _ _. _ . _ . . . . . , . . _ _ _ . . , . _ . _ _ _ _ _ _ __ _ _ _ . . _ . _ . _ _ . , ~ _ .

TIGURE 1 CORE THERKAL LlHITS Nominal Tay, 586.8'F Nominal RCS how - 268,500 GPM 680 660 -

N '

640 -

0 psia 620 -

N

@ pk 600 -

580 -

560 ' ' '-

0 0.2 0.4 0.6 0.8 1 1.2 1,4 POWER (fraction of nominal)

Figura 2.1 1a REACTOR CORE SAFETY UmS FOR THREE LOOP OPERATION FOR THE PERIOD OF OPERATION UtmlSTEAM GENERATOR REPLACEMENT L

NORTH ANNA UNIT 1 2 2a 1

l . . -- , . . _ _ , . . . . . . . - _ _

o'

• l l

2.1.1 Generation of New OTAT and OTAT Sotpoints j 4

New Technical Specifications OTAT and OPAT setpoints were generated t from the - Figure 1 Thermal Limits using the Methodology of WCAP-8746 l (Reference 4). The revised setpoint equations are compared to the existing equations in Tables 2 and 3.

For the OTAT function, the f(41) function currently in the Technical  ;

Specifications was demonstrated to provide bounding protection for the p ' ' osed reduced minimum measured flow rate. This evaluation was :

performed by the standard approach of demonstrating that the static power reduction associated with the f(AI) function maintains the DNBR above the design limit (1.46) for a range of positively and negatively skewed power ,

shapes at the reduced flow rate,

• Confirmatory analyses of the rod withdrawal at- power accident were performed which demonstrate that the thermal limits are not exceeded over the entire range of achievable system conditions for operation with the revised protection setpoints. The results of these analyses are presented in the next section. I l

10

TABLE 2 Revised Overtemperature AT Setpoint Equation AT(Setpoint) 1+tauls I

= AT(Nom) x (K1 - K2--------- (Tave-10) + K3(P-PO) + f(Al)l 1+ tau 2s i l

where CURRENT PROPOSED i l

K1 = 1.264

~

1.132 K2 = 0.0220 0.0220 '

K3 = 0.001152 0.001152 taul =- 25 sec 25 see tau 2 = 4 sec 4 sec ,

.- T O = 586.8 'F 586.8 'F ,

P0 = 2235 psig 2235 psig f(AI) = 0.0167x(-44%-AI) Al <-44% No change  ;

a 0.0, -44% <Al<+3% No change l

= 0.020x(Al-3%), Al > +3% No change- t Tave = Average temperature 'F P = Pressurizer pressure, psig TABLE 3 Revised-Overpower AT.-Setpoint Equation AT(Setpoint)-

tau 35 i AT(Nom)-x{K4-K5---------Tave+K6(Tave-10)l~

1+ tau 3s where CURRENT PROPOSED K4 n 1.079 1.016 r 0.02/'F, Tave increasing 0.02

6. 0.0 Tave decreasing 0.0 t Kb 0.00164 0.00164 ,

-tau 3 =-

10 sec 10 sec- .i' TO = 586.8 'F 586.8 'F f(AI) = 0.0, all Al

-Tave = Average temperature 'F-11

- . . - , . - . . - . . . . . - .. - . .. - . . _ , . . . . - . . _ _ - . -. - . . . . . . .......a. ,.......;

j

.: -1 1

1 i

1

? 3.1 Cont ~ .,ov,;es ,r hentrolled Roci Withdrawal at Power i.3 Rod With* w e v ower (RWAP) Accident has been reanalyzed to asue" 9 > . . . ' rainimum mea >ured flow of P.68,500 gpm asser-, , vi, > t' . The< sed JTl se oints described above  ;

' a, i

were e- 9 ' / sis. i ;.1etion in the high nucitar flux trio no e , consistent with c,ue proposed interim operation.

A statistica. tre a i key .-ialysis uncerteinties was utilized in c';c u dsnt with %eth Anna . implementation oi ;he c.ethodology described 4 Reference F. A ' di sc]ssion of the analysis is presented in the fotiowing sections.-

2.1.2.1 Accident 1: iription The uncontrolled rod cluster contrni assembly (RCCA) withdrawal at power is:a p stulated Condition 11 event initiated by operator action or

-control system malfunction. -The transient is choracterized by an  ::rea se in core heat flux, resulting in a mismatch between core power generation and power removal Dy the stezm generator. This power mismatch, which persists until the : team generator pressure reaches-the relief or safety.

L valve setpoint, causes an increase in the primary coolant temperature.

The transient would result in violation of the core thermal-limits if not terminated b.y either manual or automatic action. The reactor protection syster is designed to terminate the transient prior to exceeding core thermal l#mits.

. 10

L s., v.

2.1.2.2 Method of Analysi The rod withdrawal at power event was reanalyzed with the RETRAN (6) i system transient analysis code. All assumptions were consistent with or .

conservative with respect to those in the previously approved analyses.

The- RETRAN code - provided transient pressures, core inlet temperatures, heat fluxes and core flows which were used as input to a detailed thermal / hydraulic statepoint analysis using the COBRA (7) code. The WRB-1

-critical heat flux :orrelation (8) was used.

To fully evaluate the RWAP event, a wide range of initial plant conditions are analyzed - to determine those which are snost limiting.

Previous analyses showed that the transients initiated from hot full

- power were limiting (1). Therefore the revised setpoints were reconfirmed by analyzing a range of reactivity insertion rates fron an initial power level of 95fof rated thermal power, s It is assumed in the'_ analysis that a s: .am dump and rod control

. systems ao not function during _the RWAP event. However credit is taken--

' for i pressurizer LPORV's and safe'y val'ees, steam generator- atmospheric

-relief valves and safety valves, as well as pressurher spray (full flow from both valves is assumed), since studies have shown that this provides more limiting DNBR's.

l:

e, 13

..(

8

-2.1.2.3 Results and Conclusions The reanalysis of the rod withdrawal at power event demonstrated that >

, the minimum DNBR will remain above the DNBR design limit for operation wi th reduced niinimum measured flow associated with extended . steam.

generator tubh plugging. '

Figure 2, on the following page, presents the minimum RVAP DNBR result as e function of . reactivity insertion rate. The lower graph (labelled

.100% Power) shows the Ref. 1 analysis results assuming the 275,300 gpm '

minimum. measured flowrate, 100% power and the . current Technical Specification OTAT'setpoints. .The upper graph (labelled 75% power) shows

. the results from Lthe revis.ed analysis with reduced minimun measured

'lowrate, 95% power, the. preposed Technica'l specification OTAT setpoints, g and a reduction in the power r_ange . high flux trip setpoint. These-results demonstrate that the combination of Overtemperature AT and high.

. fle'x reactor trips act .together to provide core DNB protection over the ,

l;y range of achelyable thertr.al conditions.

L L

1

't y 0 l

8

(

14

.- s-

1. * + , -

=

1 l :.

l l

l

FIGURE 2 l+

l- Effect of Renttivity .ir,sertion tels se Minimum DN82 '

1005 Power vs 96! Power, Mininum feedback 2.20-2,10-l 2.00-5I'IO~

o 95! Poser E1,80- ^-

a

~

s- 1.74- ei -rtn IDDI Pever o.s:- / 3 BT97 1,50- ---,,

L i 4s -

r --- ,---- -

IE*D6- 1[-05 1E.04 IE-03 REACT, .iYINSERT10llRATE(0X,iEC) l

)

y  ;.: -

I 2.2 Summary of Accident Evaluations 2.2.1 Overview of Assessment Process  :

i As discussed in Reference 2, the process of evaluating the accidents for the ef fects' of reduced flow is aided by what is essentially a screening process which stajects -the individual accidents to the following tests:

1) Is the acciden; tricacted-by neither RCS flow nor steam generator tube pluggirg?- -In some casts (e.g., waste gas decay tank rupture), there isino impact andithus the event need not be considered iurther.

~

2) Is the accident impacted by plugaing but.not by flow?' These events E

(e.g. chemical ^ and volume control system malfunction at power, which is-sensitive:to RCS volumn-but not flow) will be addressed under 11CFA 50.59 -

to supportLunit restart with extended plugging but have not been addressed here since they are ' not impacted by the proposed RC5 flow Technical

' Specification Change.

-3)LIs the accident impacted by RCS flow alone (i.e. and not by othat--

tube-plugging phenomena)? In some cases -thi dynamics of the event are

'not impacted by plugging effects,: and the impact is limited to the direct of fect of RCS flow on the DNBR. ' An example is accidental'depressurization

.o f. the reactor coolant system. - Accidents in this category were 7

dispositioned via ap' plication of a generic DNBR evaluation which shows that the ef fects of' a15% power reduction more than of fset the impact of-g 16 r ,' , .,

.2 .?-  ;

l l

the proposed additional flow reduction (i.e. with respect to the Reference 1proposedvalue). This evaliiation is discussed in Section 2.2.2, below.

4) Is the accident potentially impacted-by both RCS flow and steam' generator tube plugging ef fects? These are accidents which, in addition to the direct flow effect on DNBR, may be sensitive:to a) steam generator hydraulic resistance (i.e. pressure drop)

, b) steam generator heat transfer area and/o.' secondary side  :

initial condition; 1

c) reactor coolant system volume d) instrumentatio.i ef fects (i .e. OTAT trip)

Accidents in-this' category must be evaluated, not only-for the direct

.0MBR effects, as in Category 3) above, but also for the additiona'l dynamic effects.

']

7- 7 %-)

E

' 17 m> .

, t i 2.2.2 Assessment of Net Flow / Thermal Power Impact on DNBR Accidents in Category 3 above have been assessed using a generic evaluatiou of the net effect on ' calculated DNOR's of the proposed (a) reduction in core thermal power to 95% of rated thermal power and the associated increase in allowable radial power peaking, and (b) reduction o i_n thermal-design flow from 275,300 gpm (91,767 gpm/ loop) to 268,500 gpm (89,500 gpm/ loop). The analyses presented in Reference (1) support the establishment of a 275,300 gpm minimum maasured RCS Total Flow Rate.

To perform this evaluation, a series of thermal / hydraulic statepoints

'which represent normal operation and linit%g accident conditions were e

percurbed to determine-the DNBR effect of margivel changes in flow, power, and FMi, The following statepoints were considered:

1) Nominal design hbt full power conditions.

2) The statepoint corresponding:to minimum DNBn following the c%plete loss of RCS flow evert. >

3) The limiting ONOR statepoint for the-uncontrolled RCCA

~

withdrawal event,

[

t

-The worst-case sensitivities developed from_each of these statepoints

,,ere then used- to perform a conservative overall assessment of the net effect of 1 reduction .in -design flow f rom 275,300 gpm (the- Referency 1 L basis) to 268,500 gpm.(approxiaately a 2.5% reduction) coincident with a 5% reduction ,in core thermal power and the associated increase in the 18 l

.- R. 1 I ' I r . - , , ,

d t

'I.l

+

Technical' Specification allowable FMI. The result was a net DNBR benefit-

. ranging from-1.2-2.5%, depending on the statepoint examined.

2 t

-11P T

1

- \

g i

J 1 j

. ( ;

[.i :

i l '?

I-

'jL I ,

l, jji 19-i '3'e a s , ,

- 212.3 Discussion of Accidents By Grouping Group 1: No Flow /N,o Pluaoino Impact Accidents which are impacted by neither flow nor plugging are those which are insensitive to RCS thermal / hydraulic conditions or have been demonstrated not to be. credible (e.g. inactive loop start at power) for

~

North Anna Unit 1. These events may be excluded from further consideration as discussed in Reference 1, and are as follows (UFSAR section number is~provided below):

Inactive Loop Startup (15.0.6)

• Misloaded Fuel Assembly (15.3.3) +

Waste Gas Decay Tank. Rupture (15,3.5)

Volume Control lank Rupture (15.3,6).

Fuel' Handling Accident Outside Containment (15.4.5)

-* ' Fuel Handling Accident Inside Containment-(15.4.7)

Steam Generator-Tube Rupture -(15.4.0}

As discussed above, the inactive loop start at power is not considered

~

credible based on current Technical Specifications. RCS flow rate is not

- an_ analysis input parameter for the misloaded fuel assembly event, the tank rupture accidents or the fuel handling accidents. Further discussion of this category of events may be found in Reference (1).

A 20

a.

[. . .

Stea LGenerator Tube Rupturg The steam gent;rator tube rupture event has been placed in this category-beause extended- steam generator tubo plugg;ng and its accompanying ef fect on RCS flow, primary to secondary heat transfer, and RCS loop-resistance would have insignif1 cant impact on the analysis results of the J+

steam gerterator tube rupture transient, as discussed in Reference 1.

{ {l 4

21

i l

- l I

I

-I Group 2: Accidents'ImpudrLbr lubtPingaina But Not RCji_Elow Rate -

Events which are insessitive to RCS flow but which can be impacted by SGTf' levels are as follows-(UFSAR section nuniber is provided below):

CVCS Malfunction (tloron Dilution) (15.2.3)

Small Breik LOCA (15.3.11

• - Large Ilreak t.0CA (15.4.1) t In general, these are event, wh ch are not assessed against the minimum DNBR criterion for moderate freq .ency events but rather by alternate cr'tecia such as available operator response time.(e.g. to avoid loss of shut,down n;argin for boron diiution 'v#ts;,

{Loron Di htion A r ,.4m. ti on' in the minimum measured flow rate has no- direct consequences on the analysis of. the boron dilution event. This hat, been discussed fully in Reference 1. The impact'on the boron dilution at power H< ;a'nalysis of the reduction in RCS volur.e associated with extended SGTP will be considered - as part of - the analysis supporting North Anna 1 Cycle -9 rsstart. Therefore,-no reanalysis of this-event is required to suppnrt the _ proposed Technical Specifications changes.

I l

22

% . , +

. . . . _ _ . . _ . , . . _ . . - _ _ __ _ . . . _ , _ - . _ . . _ _ _ _ _ .. ._ .m. . .., .

... y '

latqq_pnd Small-B Leak LOI6 1

Reference 1 presented an assessment which concluded that both small and large break loss of coclant. accident analysis results are insensitive

.o the proposed change in RCS flow- rate. The ~ prior evaluation's conclusions have been cor amed to remain applicable for the 268,500 gpm

-measured flow limit.

B l

(

s t

4 F 23 e ei-- *w *m , ,ty& y p ^ t- p y td Twy--T y k

'n Grou2 3: Airddents impacted by RCS Flow Ahqq

' Accidents which are impacted by RCS flow oni, (and not by other tube plugging phenomena) are those where primary to secondary side heat transfer and/or steam generator primary side pressure drop, which are the two major impacts of SGTP apart from flow, da not impact the dynamics of the accident. For example, ' reactivity excursion transients which 'are rapid with respect to the thermal respon30 time .of the steam generators are placed into this category. Group 3 accidents ir, this category are (UFSAR section nuuber is provided below);

• Control Rod Drop / Misalignment (15.2.3) -

• hod Withdrawal from Subcritical (15.2.1) a Controf RoJ' Ejection (15.4.6)

Control Rod Dron The Virginia Power-methodology for analysis of control rod drop was discussed in Reference 1. Reference 1 discussed the development of.new dropped rod DNBR mli it lines applicable to the 275,300 gpm total-measured flow conditi.on. lhese lines were based on a design DNBR limit which .

exceeded the statistical . DNBR limit a'nd therefor e - contained retained margin. This margin is nore than enough-to offset the reduction from the 275,300 gpm to the 268,500 gpm flow condition. Therefore the dropped rod' limit lines Referended in (1) are adequate to assess restart of Unit 'I at either_the 275,300 gpm or the 268,500 Opm r..easured flow limit.

24

s Rod Vithdrava.] From Suterifical This event is assumed to be initiated from hot zero power, It is a non-limiting transient f rom a ONBR standpoint. The current analysis yields a peak heat flux well below the hot full power steady state value, The peak heat: flux'statepoint from the UFSAR analysis was reexamined at the- proposed flow limit and the Technical Specifications radial peaking _

factor $1mit fer hot zero power. The ONBk remains well above the limit av.d the-conclusions of the UFSAR remain valid at the proposed flow ilmit for this.tJent.

-[ontrql Ro_d Election The: control rod' ejection evaluation in Reference 1 showed a small but n acceptables. increase;in peak' clad temperature for the limiting case based 1on _ application of ' flow senitivity studies,-

. Extending this assessment

,, Eto the current proposed ' flow rate _ shows that the analysis limit for peak 1 clad temperature isistill met with substantial margin for the limiting case.

01her Events Asstssed for DNB g

i: For those- events assessed against the DNBR criterion for moderate p

f requency events, the assessment in Section E.2.2 applies. - As notad itherein, ~ the impact 'of 'a S*. redpction in reactor power and the associated I

<3 25

~ -

+ -_

increase in allowable FAH has been shown to more than offset the impact of a 2.5% reduction in flow from the 275,300 gpm value provinusly assessed in Reference 1. Put another way, the assessments performed for 27E,300 gpm design flow at 100% rated thernal power are applicable to 268,500 gpm design flow at 95% rated thermal power.

\

Group 3 accidents included in this category are as follows (UFSAR section number is provided below):

• Accidental Depressurization of the RCS (15.2.1c)

• Excessive Load Increase (15.2.11)

• Excessive Heat Removal (15.2.10)

• Spurious Operation of the Safety injection System (15.2.14) 26 I

,)

l------________________

. - - 1 _,. ._ _ . _ _ . . _ _ _ _ _ . _ . . _ .. . ._

i x <

I 91.gyup 4:' Accidents' Potentia 11v I.m_pacted by RQS Flow and Pluaaing Accidents potentially affected by both tube plugging and RCS flow are as follows (UFSAR section number is provided below):

Partial less of Flow (15.2.5)

• Loss of External Load (15.2.7)

Loss of Normal feedwater (Loss of Offsite AC) (15.2.8/15.2.9) ,

Accidental Depressurizetion of the Main Steam System (15.2.13)

Minor Secondary Steam Fipe Breaks (UFSAR Section 1!.3.2) f

• Complete Loss of RCS Flow (15.3.4)

• -' Single Rod Withdrawal at Power (15.3.7) ,

Major Secondary System Pipe Ruptures (h in Steam Line Greak)

(15.4.2.1); -

Rupture of a Maia feedwater Pipe (Main.Fe90line Break)-

(15.4.P.1) ,

Locked Reactor Coolant Pump Rotor / Sheared Shaft (15.4.4)

Partial Lonj_FLqs. Small Steam Pio.9 Breaks '

t The partial loss of flow event-is bounded by the ccmplete loss of flow

q event'and.its assessment-is included within the scope of the evaluation f ,' fors.the more' limiting event.

Likewise the accidental _ depressurization-ofcthe main steam system and minor secondary steam pipe breaks are bounded V - by ' th.e main isteam line break -event and- their assessments are included g, within the scope of the evaluation of 'the more limiting event, h

L h .

I' l

r.

27

v j r -

I Loss of External Electrical Load The-Loss of External Load event was reanalyzed for SGfP levels up to 40'4 and a reduced RCS minimum measured flow rate of 275,300 gpm (1). The 'I 1

analysis demonstrated that the transient does ,ot challen n RCS and main steam system overpressure safety limits, nor does ' it approach DNB conditions, ihe evalustion of Section 2.2.2 shows that the DNBR results of Reference -(1) wI11 remain bounding for the-. proposed condition. With-respect to primary side oveipressurizat:on concerns, the reduction in RCS

~

flow rate - does' not- significantly affect primary or secondary side overpressurization results. Because the proposed revised design

-conditions also include a pe.1. power level of 95a;, the decreased load .

rejection would be expected to result in lower primary and secondary "

pressures than in the previous analysis.

It may be concluded that the effects of the proposed reduction in

- minimum meuured RCS flow rate are bounded by the both the DNBR and Loverpressurization analysis results presented in Reference (1).

3 28' a e -

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

1 Lg,ss o'r Norma 1 Fedillain The Loss of Normal Feedsater accident was reanalyzed as documented in Reference 1. The analysis demonstrated that ste- generator tube 1 plugging levels ~up to 40% (uniform) and a Tech Spee minimum measured RCS flow rate of 275,300 gpm do not adv.arsely impact the ability of the  ;

auxiliary feedwater system to deliver adequate feedwater to prevent the relief of reactor coolant water through the pressurizer relief or safety valves, and to prevent system overpressurization, for the cases with and

- without continued ' operation of reactor coolant pumps, the feedwater flow

rates required to provide adequate cooling were demonstrated to be well {

below actual deliverable pump flow rates, '

L Because this transient essentially. evaluates the capability of the jw

- steam generators to remove core dect.y heat, the resuic of the analysis are. primarily impacted by the '.evel of SGTP rather than the reactor

' coolant' system flow rate. The RCS flow rate impacts the nominal AT across

+

l the core, but does not significantly impact the steady state (or

[ transient) removal of heat in such a long-term heat removal = transierit.

L l The: results of the : loss nf normal = feedwater analysi>. there fore are insignificant 1y impacted by a further reduction of this minimum measured flow-rate to 268,500 gpm. Furthermore, the 5% reduction in initial pcwer ,

level reduces both the initial stored energy and the: post trip decay heat levels by an' amount which more than of fsets any small decrease in energy

- removal c:.pability due to reduced RCS flow. Therefore the Reference 1 analysis remains bounding. Si.tilar reasoning leads to the conclusion that 29 I

+~r

l. h 9 y i the Reference 1 evaluation for rupture of a main feedwater pipe remains

_ bounding for!the._propcsed condition.

Comojate;Lostof Flow y

The complete loss of flow event has been reassessed for the proposed conditions by both applying the DNBR sei.sitivity stud es discussed in .

Section 2.2.2 end examining the potential effects of higher loop resistance on the-normalized RCS flow coastdown vs. time curve for this event; The latter of fect was evaluated by assuming that the additional oecrease in loop flew from Reference. I to the proposed condition i s' entirely 'due to loop resistance effects. This effect t .s modelled in the flow coastdown analysis and the normalized-(1.c. to time zero) coastdown curve war compared to the previous result. Based on this comparison and ,

the DNdR ~ sensitivities of Section 2.2.2, an estimate of the net ef fect of . flow, pou r and radial peaking : on minimum DNBR at the proposed

• condition was made. :Not6 that the .ONBR sensitivities developed in Section 2.2.2 enveloped the flow coastdown statepoint. It:vas concluded that the -overall impact of the proposed flow limit at 95% power wi11 result in a- DNBR benefit with respect to the Reference I analysis. The Reference 1 analysis tharefore bounds the proposed condicions.

J Sinole Rod W1.thdray.al at Power

/ '

The Single Rod _ Withdrawal at Power event produces a system transient 1 response which is similar to the uncontrolld control bank assembly withdrawal; that is, it results in an increase n, core heat flux and a l 4 ,  ;

30 I

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1.

,4 . 4 ' e 1

-mismatch between core power generation and power removal by the steam

= generators. Thit, power mismatch, which persists-until-the staam generator pressure reaches the relief or safety valve setpoint, causes an increase j in the. primary coolant temperature; The transient would result in a '

violation of the analysis -limits if not terminated by either manual or automatic action.

The reanalysis pres 0nted in Section 2.1.2 for the uncontrciled control bank withdrawal at power demonstrates that the oroposed protection setpoints in ccmbination with the proposed RCS flow limit provide adequate f

DNS protection for - the full range of applicable thermal hydraulic ,

conditie9s. The reload: evaluatien process demonstrates that .less tMn 5% of the core wili experience hot cha.wel fUtors in axcess of the steady i

strite design limit for any sirgle withdrawn RCCA. In this manner it can d

be demonstrated that less than 5%:of the core will experience DNBR less than the design 7 limit during-a' single RCCA withdrawal event, consistent

- with the-conclusions-of the UFSAR. This' conclusion will be reconfirmed prior.to Unit I startup.

, Pain Steamline Brea.h The. main _ steamline break analysis is pef formed t'o <!emonstrate ' that

there would be no con damage due to the onset of DND, and that the energy release to containment does not cause failure of the containment structure.

m 31 M4

%N ,' . ,, . . . .

< y, . . .: .

Reference 1 demonstrated that the ' consequences of a main steamline break-event under conditions of extended SGIP and reduced RCS flow rate would not- exceed the analysis criteria as presented above. This was demonstrated on the basis that extended 5GTP reduces the steam generator's capacity to remove ensirgy fr( i the RCS. Because the primary effect on the RCS of a main steamline break is to decrease RCS temperature and pressure, and 'to increase core power (given an ond-of-cycle negative ,

moderator temperature coef ficierd), a reduced capacity to remove energy from the~ RCS du9 to extended SGTP is an analysis benefit. It was concluded that that the calculsted transient DNBR under conditions of cxtended steam generator tube piogging would be less limiting than tha current _ licensing analysis.

For an additicnal reductica in RCS. flow rete from 275,300 gp:n to 268,500 gpm (a reduction of 2.5%), an additional penalty to be taken out

l. of retained margin was developed. Utilizing a bounding flo,, sensitivity

'(+1.4% DNBR/% flow), the additio.tal flow nduction traialates to a 3.5%

penalty to be assessed against available main steamline break analysis ,

retair.ed mergin. This penalty fully accounts f or the impact of the reduced RCS flow rate on itSLB DNSO analysis results.

!l .

l D (!Lqbjd Reaglor Coolant Pump lo,t_gI n

l D

- The locked reactor coolant puay rotor event was analyzed in Reference

-1 for-a-fiow rate corresponding to a measurement limit 275,300 pm and hot full power operation. The results showed that less than 13% of ' the rods in the core experience DNBR lest than the design limit, crnsistent i

32 \

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vith th'e UFSAR. Also, the peak RCS pressure remained well within the '

I acceptance limit. '

l i

for the cise of an additforal reduction of 2,5% in RCS flow and a 5%

reduction in pnwer, use of the DNBR sensitivities in Section 2.2.2 shows that thr. fractior, af rods experiencing DNRR less than the design limit 4

- would be reduced. Also for the overpressure protection case, the effects of a. 5% reduction in thermal power will more than offsat the flow reduction effects. Since the dominant resistance to flow during.the event is the iocked rotor itself, the normalized (i.e. to time zero) flow

- coastdown curve for locked rotor is expectri to- remain essentially unchanged for the revised condition. The Reference 1 analysis remains 9

- bounding.

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8 t I-33 p, , .

) g-_ . L .

3.0 -NSSS and Balance of Plant Systems and Components 3.1 NSSS Systems and Components As documented in Reference 1, Westinghouse Electric Corporation performed reviews of the following NSS$ components and systems to ' confirm that operation within the proposed conditions remains in compliance with

• the applicable codes and standards.

- Reactor Vessel and Internals

. Control Rod < 0 rive Mechanisms

- Main 1.oop Isolation Valves

" Reactur Coolant Pump and Motor

-. Pressurizer

- Steam Generator

- Auxiliary Systems Components (tanks, valves, heat exchangers)

- Fluid Systems

- Reactor Protection

• and Control Systems.

-*See Sectio ~n 2,1 for an assessment of _new protection setpoints for the' proposed conditions.

- A_. r3yiew of the Reference 1 evaluation .shows that the Westinghouse studies envelope operation at the proposed design flow rate of-268,500 R gpm; 3.2. ~ Balance of Plant _ Systems and. Components-

. Stone and Webster Engineering Corporation has- evaluated the effects

^

of' operating North Anna Unit 1_ with reduced :RCS flow and. extended SGlP upon. -th'e balance . c f plant _ systems and components. The changes of significance ? for this assessment involve reductions in RCS flow, RCS volume, steam temperature and steam pressure. Engineering evaluations -

have been performed to - demonstrate - that these parameter - changes . and

, 34

7 T

resulting - ef fects un plant systems and components will be bounded by existing analyses end will continue *.o meet applicable design l criteria.

These major balance of plant design areas were evaluated (1):

- Accident Analyses

- Balance of Plant (B0P) Systems and Components

- Class I Piping

- Electrical Distribution System A review of the Reference 1 assessment shows that the evaluation bounds the proposed operation at 95%.of rated thermal power with an RCS total flow rate of 268,500 gpm, L

l f

(

Il l

i I

35-J'

,x. .

4,0 CONCLUSIONS A review of the accident analyses presented in UFSAR Chapter 15 has demonstrated that a reduction in mir.imum measured flowrate fo: North Anna T

Unit 1 to 268,500 gpm at 95% of rated thermal power is accommodated by current thermal margins or by the assessment of a penalty against available retained DNBR margins for all accidents. Explicit reanalyses were performed for ". rod withdrawal at power .tvent to confirm the adequacy of proposed revisions to the Overtemperature AT and high flux trip setpoints to be implemented for the balance of Unit 1 Cycle 9 operation. In addition,- the remaining Engineered Safety Features and Reactor Protection System setpoints set forth in tbc Unit 1 Technical Specifications have been demonstrated to provide adequate plant protection at the reduced flow condition, The evaluation showed that all of-the acceptance criteria previously-established in the UFSAR continue to be met for all of the events analyzed in Chapter 15. This conclusion will be reinforced by continued

-verification that core physics characteristics for operation with a reduced RCS flowrate remain within the envelope established by the reload safety evaluation to be performed prior to resumption of Cycle 9 operation.

The proposed temporary Core Tharmal Limits have been vcrified to bound Ur.it 1 Cycle 9 operation with the new RCS flow rate.

36

. o . .

A review of the NSSS design transients; NSSS fluid and control systems; reactor control and protection systems; NSSS primary components (including thermal and structural effects); and steam generator thermal / hydraulic performance has been performed. It was concluded that NSSS systems and components will continue to meet applicable acceptance criteria for operation wits the reduced design flow rates and the associated steam generator tube plugging levels.

An engineering evaluation has also been performed to assess the impact of reduced flow and tube plugging on the existing containment integrity analyses (including the impact on Net Positive Sucti:n Head-NPSH of engineered safeguards numps) and containment subcompartment integrity analyses. The existing analyses were shown to remain bounding.

A balance of plant systems review .shows continued acceptable performance under the reduced RCS flow / extended tube plugging condition.

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REFERENCES

1. cetter from W. L. Stewart, Virginia Power to USNRC, Serial No.92-018, Virginia Electric and Power Company, North Anna Power Station Unit 1, Proposed Technical Specification Change for Reduced Minimum RCS Flow Rate Lir.it, January 8,1992.

2. Letter from W. L. Stewart, Virginia Power to USNRC, Serial No. '

92-018A, Virginia Electric and Power Company, North Anna Power Station Unit 1, Supplemental Information Regarding Our Proposed Technical Sr,ecification Change for Reduced Minimum RCS Flow Rate January 31,1992.

3. Letter from W. L. Stewart, Virginia Power to USNRC, Serial No.92-042, Virginia Electric and Power Company, North Anna Power Station Unit 1, Proposed Technical Specification Change, Reduction in Maximum Reactor Power Level Due te Increased Steam Generator Tube Plugging Level, January 28, 1992.

4. WCAP-8746, Design Basis for the Thermal Overpower AT and Thermal Overtemperature AT Trip F.nctions, Westinghouse Electric Corporation, March 1977.

5.- VEP-NE-2-A,- Statistical DNBR Evaluation Methodology, Virginia Power, June 1987.

6, VEP-FRD-41-A, Vepco Reactor System Transient Analysis Using the RETRAN Computer Code, Virginia Power, May 1985.-

7. VEP-FRD-33-A, Vepco Reactor Core Thermal Hydraulic Analysis Using the COBRA IIIc/MIT Computer Code Virginia Power, October 1983,

8. VEP-NE-3-A, Qualification of the WRB-1 CHF Correlation in the Virginia Pcser COBRA Code, Virginia Power, Joiy,1990,

9. WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event. Westinghouse Electric Corporation, January,1990.

10. VEP-NFE-2-A, Vepco Evaluation of the control Rod Ejection Transient, Virginia Power, December 1984. .

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