ML20091P456

Proposed Tech Specs Re Max Reactor Thermal Power Level

ML20091P456
Person / Time
Site:	North Anna
Issue date:	01/28/1992
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
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ML20091P454	List: ML20091P452 ML20091P456
References
NUDOCS 9202030110
Download: ML20091P456 (52)

Text

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..L ATTACHMENT 1 l

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PROPOSED CHANGES TO THE FACILITY OPERATING LICENSE AND THE

, TECHNICAL SPECIFICATIONS FOR NORTH ANNA UNIT 1  :

L VIRGINIA ELECTRIC AND POWER COMPANY 9202030110 920120 PPR ADOCK 05000338 p PDR

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ed.

(1) Maximum Power Levoj VEPCO is authorized to operate the North Anna Power Station, Unit No.1, at reactor core power lovels not in excess of 2893 megawatts (thermal).

• l (2) Technical Spedlications ,

The Technical 3pecifications contained in Appendicos A and B, as revisoJ through Amendment No. [_] are hereby incorportited in the license. The licensee shall operato the facility in accordance with the Technical Specifications.

(3) Additional Conditlung The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated timo perloo's following the issuance of this amendment or within the operational restrictions indicated. .

The removal of these conditions shall be made by an I amendment to the license supported by a favorable evaluation by the Commission: t

c. Virginia Electric and Power Company sha'.! not operate the reactor in operational modes 1 and 2 with less that three reactor coolant pumps in operation.

d. VEPCO may use two (2) fuel assemblies containing fuel rods clad with an advanced zirconium base alloy cladding material as described in the licensee's suomittats dated February 20,1987 and September F),1988.

e. !f Virginia Electric and Power Company plans to 4 remove or to make significant changos in normal operation of equipment that controls the amount of radioactivity in effluents from the North Anna Station, the Commission shall be notified in writing regardless of whether the change affects the amount of radioactivity in the effluents.

The maximum reactor power level shall be limited to 2748 megawatts (thermal) which is 95% of RATED THERMAL POWER in accordance with the licensee's submittal dated January 28,1992 (Serial No.92-042) for the period of operation until the steam generator replacement.

Amendment No. [_)

e PERGENCYCORECOOtfJG SYSTDAS ,

ECCS SUBSYSTEMS Tavo 2 350'F l I

UMITING CONDITION FOR OPERATION _

i 3.5.2 Two indopondent ECCS subsystems shall to OPERABLE with each subsystem .1

. comprised of:  ;

a. Ono OPERABLE contrifugal charging pump,

b. One OPERABLE low head safety injection pump,

c. ' An OPERABLE flow path capablo of transferring fluid to the Roactor Coolant i System when taking sucilon from the refueling water storago tank on a safoty ,

injection signal or from the containment sump when suction is transferiod during the recirculation phase of operation or from the dischargo of the outsido recirculation spray pump.

APJ'LICABILITY: MODES 1,- 2 and 3. ,

s ACTION:

a. With one ECOS subsystem inoperablo, rostore the inoperablo subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or bo in HOT SHUTDOWN within the next 12  ;

hours.

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b. In the ovent the ECCS ise.ctuated and injects water into the Roactor Coolant System, a Special Roport shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstancos of the actuation and the tolat accumulated actuation cyclos to dato. l  !

c. Tho provisions of Specifications 3.0.4 are not applicable to 3.5.2.a and 3.5.2.b l

for one hour following heatup above 324"F or prior to cooldown below 324'F.

Adherence to ACTION "a" shall require tho following equipment OPERABILITY for the period '

of operation until steam gonorator replacement:

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With ono low head safety injection pump looperablo, two contrifugal charging .

pumps (ono in cach subsystom) and their associated flow paths shall bo OPEHABLE or bo in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and be in HOT SHUTDOWN within the next 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

NORTH ANNA UNIT 1 3/4 53 l

I ATTACitMEt4T 2 DISCUSSIOil OF PROPOSED CllAl4GE VIRGINIA ELECTRIC AND POWER COMPAt4Y

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Discussion of Proposed Change North Anna Power Station Unit 1 is currently involved in a mid cycle steam generator 4

inspection outage. An extensivo oddy current inspection of the North Anna Unit 1 steam generator tubes is being performed using very conservativo analysis guidelines and plugging critoria. As such, a substantially increased number of tubes are i expected to be plu0ged.

The prodletions of steam generator tubo plugging required during this mid cycle i outago are such that the offects of increased RCS loop resistance on the largo break Loss of Coolant Accident (LOCA) analysis would not permit full rated power operation for the romalnder of Cyclo 9 operation for North Anna Unit 1. Thoroforo, safety analysos and ovaluations have boon performed which support continued operation with an imposed reactor power restriction. The attached r,afety evaluation has been prepared to discues the changes to the largo break LOCA analysis and support this licenso amendment for the associated rostriction in reac or power.

The proposed license change will kmit maximum reactor power to 95% of RATED THERMAL POWER. Specifically, wo request a change to Facility Operating License No. NPF 4 to the Virginia Electric and Power Company for North Anna Power Station Unit 1 to modify license condition 2.D.(1), Maximum Power Lovel, by adding a footnote which states that:

The maximum reactor core power level shall be limited to 2748 megawatts (thermal) which is 95% of RATED THERMAL POWER in accordance with the licensee's submittal dated January 28,1992 (Serial No. 92 042) for the period of operation until the steam generator replacement. ,

in addition, an associated change to the Technical Specifications is required to accommodate the effects of the revised assumptions for the largo break LOCA  :

analysis. The proposed change to the Technical Specifications will impose more restrictive equipment operability requirements for the Emor0ency Core Cooling System (ECCS), Specifically, we request a change to Action Statomont "a" of Specification 3.5.2, ECCS Subsystems Tavg 2350'F, by addin0 a footnote which states that:

Adherence to ACTION "a" shall require the following equipment OPERABILITY for the period of operation until steam generator replacement:

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- With one low head safety injection pump inoperable, two contrifugal charging pumps (ono in each subsystem) and their associated flow paths shall be OPERABLE or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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I in offect, this proposed change will ensure that both low head safoty injection pumps '

or c 10 low head injection pump and two high hond safety injection pumps romain opotable during power oporation. This chan00 offectively maintains consistoney betwoon the Technical Specification Action Statomonts and the revised assumptiens  !

for the large break LOCA anNysis.  ;

The proposed chan00s are necessary to accommodate the expected increased steam generator tubo plugging levels. The attached safety ovaluation supports the abovo changes to the operating licenso and the Technical Specifications. The changes are required on an interim basis until the steam generator replacement in 1993, at which timo it will no longer apply, m

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• ATTACHMENT 3 SAFETY EVALUATION VIRGINIA ELECTRIC AND POWER COMPANY-l-

LARGE DREAK LOSS-OF-COOLANT ACCIDENI (UISAR Section 15.4.1) 1.0 INilK100C110N North Anna Power Station (Jnit 1 it, currently involved in 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 analysis guidelines and plugging criteria. As such, a substantially increased number of tubes are expected to be plugged.

The physical consequences of extended SGTP are primarily (a) increased RCS loop resistance, resulting in a lower RCS flow rate, (b) decreased steam generator tube heat transfer area, resulting in lower steam generator outlet steam pressure, and (c) a decreased total RCS volume. The impact of these

-changes with respect to previously analyzed design conditions must be fully assessed for both normal operating and accident conditions. This assessment is performed. following a steam generator inspection outage usually concurrent with a new reload safety evaluation, When required, revised safety anal: es are performed and a Core Operating Limits Report (COLR) is prepared as required by Technical Specification 6.9.1J.

In many cases, the incorporation of revised safety analyses into the North Anna design basis could be accomplished via Virginia Power processes employed

- to assess change per 10 CFR 50.59. However, based on current steam generator plugging projections, it is expected that the North Anna 1 Technical Specification RCS flow limit could be violated, This could potentially occur at average SG1P levels of approximately 20%. To address this concern a separate Technical $pecification Amendment request to reduce the RCS total o

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flow rate limit by approximately 3% has been submitted for review and approval (1).

The Reference (1) package in combination with the existing Chapter 15 analyses and evaluations of plant system and component design support operation (within specific key core parameter limits) with up to 30% of the tubes plugged in any steam generator. Section 2.0 provio t, additional background information regarding this existing analysis basis.

The predictions of potential steam generator tube plugging during the current mid-cycle outage are such that the ef fects of increased RCS loop resistance on the large break LOCA analysis may not permit full rated power operation for the remainder of North Anna 1. Cycle 9. The existing large

, break LOCA analysis has obtained margin by taking credit for available Cycle n

9 "re characteristics but will not support 100% power operation with more tnan 30% SGTP. The large break LOCA analysis presented in Sections 3.0 through 5.0 of this evaluation extends this SGTP limit value to 35%. but with -

a reduced power level of 95% of rated thermal power. .

2.0 ADDITIONAL EVALUATIONS Ato BACKGROUND INf0RMA110N There are a rumber of areas of plant design which are potentially impacted by operation with extended SGTP. This section presents background information relating to the key evaluations which Virginia Power has performed. Existing analyses for each design area have been evaluated for potential effects of extended SGTP. Specific limitations on such operation, where applicable, have been developed such that the results and conclusions 2

of existing design basis analyses will remain bounding for the proposed operation. The following major areas were evaluated:

• NSS$ Systems and Components a Balance of Plant Systems and Components

= NSSS Accident Analyses For each of the above areas, the key aspects of the existing analysis basis supporting extended SGTP operation is discussed here.

We:tinghouse Electric Corporation performed reviews of components and systems within their design responsibility to confirm that operation within the proposed conditions remains in compliance with the applicable codes and standards. It was concluded that all NSSS systems and components will remain within the bounds of existing design analysis results for operation with up to 40% of the tubes plugged in any steam generator.

The effect of extended SGTP operation upon balance of plant systems and components has been evaluated by Stone & Webster Engineering Corporation (SWEC). The evaluations concluded t:at effects of operation with extended SOTP will remain within the bounds of existing design analyses for operation with up.to 37% average SGTP among the steam generators.

Virginia Power staff assessed the impact of extended SGTP operation upon NSSS accident analyses. The effects of reduced RCS flow associated with 3

extended SGTP upon most UFSAR Chapter 15 events has been evaluated in the l

Reference (1) amendment request. The remaining events requiring reanalysis, I excluding small and large break LOCA, have assumed both reduced RCS flow rate and 40% average SGTP.

Additional events which are impacted by extended SUTP but are Insensitive to reduced RCS flow have also been reanalyzed. These analyses have been implemented via the Virginia Power processes for assessing change per 10 CfR 50.59. These evrznts (and the SGlP levels supported by the existing analyses) are:

EVENl _._ PLUGQltfLuf!T Small Greak LOCA 35% 50tP in any SG Boron Dilution at Power 40% average SGTP among SGs Prior to restart of North Anna Unit 1, Cycle 9, an evaluation of the key core parameters and the actual final plugging will be performed to confirm

• that all applicable limits have been met. With the exception of large break LOCA, the existing analysis basis described in this section is valid for ,

operation of North Anna Unit I at the rated thermal pcwer of 2693 MWt with

. up to -35% SGTP. in any steam generator. The current evaluation submits the m 4 ovaluation of the large break LOCA accident with 35% SGTP, -It, however, a

requires reduced power operation in order to achieve PCT results in compliance with the 10 CFR 50,46 acceptance criteria, 4

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. 4 3.0 l ARGE BREAK LOCA ACCIDLNI DESClilPI10N 1

i A reanalysis of the emergency core cooling system (LCCS) performance for the postulated large break loss of coolant accident (LOCA) has been performed in compliance with 4pe s ' K to 10 CfR 50. The results of this reanalysis are presented herv, s! are in compliance with 10CfR$0.46, " Acceptance Criteria for Drnrgracy Core Cooling Systems for Light Water' Reactors." This analysis was tedmoed with the NRC-approved version of the Westinghouse LOCA-ECCS evaluatit n wdvi denoted as the 1981 raodel with BAS 11 (2). The analytical tectMquet isre in full compliance with 10CFR50, Appendix K. j L

As required by Appendix K to 10CFR50, certain conservative assumptions were made for the LOCA-ECCS analysis. The assumptions pertain to the ,

conditions of the reactor and associated safety system equipment at the time  !

that the LOCA is assumed to occur, and include such items as the core peaking l

factors, the cientalomint pressure, and the performance of the emergency core cooling system. Selection of input parameters for Appendix K analyses is made to rep re s,ent a corservative configuration of the plant initial conditions. This was accomplished by assuming bounding input values for key parameters such as core power, FAh, FQ, steam generator tube plugging and _

RCS flaw. In general, the remaining key assumptions included in the curr<!nt analysis are consistent with previous large break' ana'yses performed by Virginia Power. ' Additional discussion of these analysis assumptions is .

provided in-Section 4.0.

4 A LOCA is the result of a rupture of the- reactor coolant system (RCS) piping or- of any line connected to the system. The system boundaries 1:

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I considered tai the lOCA analysis are defined in the UFSAR. Sensitivity studies (5) have indicated that a double-ended cold-leg guillotine (DECLG) .

pipe breal is limiting. Should a DECLG occur, rapid depressurization of the  ;

RCS occurs. The reactor trip signal subt.equently occurs when the pressurizer  ;

i low pressure trip utpoint is reached. A safety injecticn system (SIS) j signal is actuated when the appropriate setpoint is reached, activating the i

I high-head safety injection pumps. The actuation and subsequent activation of the Emergency Coro Cooling System, which occurs with the SIS signal, assumes the most limiting single-failure event. These countermeasures will .

limit the consequences of the accident in two ways:

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1. Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residua' level' corresponding to fission product decay heat. No credit  !

is taken in the analysis for the insertion of control rods +

to shut down the reactor. i L

2. Injection of borated water provides heat transfer from the a core and limits the clad temperature increrse.

i Before the break occurs, the unit is in an equilibrium condition, i.e.,

the heat generated in the core is being removed via the secondary system.

During blowdown, heat from fission product decay, hot internals and the vessel continue to_ be transferred to the reactor coolant system. At the beginning of the blowdown phase, the entire reactor coolant system contains subcooled liquid that transfers heat from the core by forced convection'with some fully developed nucleate boiling. After.the break develops, the time.

i to DNB is calculated, consistent with Apper. dix K of 10CFR50. Thereafter, l the core _ heat transfer is based on local ccnditions, with transition boiling

! and forced convection to steam as the major heat transfer mechanisms. ,

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F During the refill period, it is assumed that rod-to-rod radiation is the f only core heat transfer mechanism. The heat transfer between the reactor coolant system and the secondary system may be in either direction, depending  ;

on the relative temperatures. For the case of continued heat addition to the secondary side, secondary-side pressure increases and the :in safety valves may actuate to reduce the pressure. Makeup to the secondary side is ,

automatically provided by the auxiliary feedwater system. Coincident with the safety injection signal, normal feedwater flow is stopped by closing the main feedvater control valves and tripping thS main feedwater pumps.

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Emergency feedwater flow is initiated by starting the ,iuxiliary feedwater ,

pumps. The secondary-side flow aids in the reduction of RCS oressure, When  ;

the reactor coolant system depressurizes to (>00 psia, the accumulators begin [

-to inject borated water into the reactor coolant loops. The conservative assumption is then made that injected accumulator water bypasses the core

-and goes out through the break until the termination of bypass. This conservatisin is again consistent with Appendix K of 10CfR50, in addition,  ;

I the reactor cociant pumps are assumed to be tripped at the initiation of the accident, and effects of pump coastdown are i;;c iuded in the blowdown analysis.  ;

The- water injected by the accumulators cools the core, and subsequent operation of the low-head safety injection pumps supplies water for long-term ,

cooling. When the refueling water storage tank (RWST) is nearly empty, the long-term cooling of the core is accomplished by switching to the ,

recirculation mode of core cooling, in which the -spilled borated water is drawn from the containment sua.p by the low-bead safety injection pumps and t

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returned to the reactor vessel.

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The containment spray system and the recirculation spray system operate l to return the containment environment to subatmospheric pressure. I r

i 4,0 l#1GE. BRfAK LOCA ANALYSIS ASSLMp110NS i As required by Appendix K of 10 CTR 50, certain conservative assumptions were made for the large Dreak LOCA-LCCS analysis. The assumptions pertain f to the condition of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur, and include such items as the core peaking factors, core decay heat and the performance of the Emergency Core Cooling System. Tables 1 and 2 present the values assumed for several key parameters in this analysis, Assumptions and initial operating conditions which reflect the requirements of Appendix K to 10CFRLO have been i used in this analysis. These atsumptions include: ,

B

= The break is located in the cold leg between the pump discharge and the vessel inlet.

o = The safety injection flow spills to containment back pressure in the broken loop. Safety injection occurs only in the intact loops cold legs.

• The accumulator in the broken loop also spills to containment.

= -120 percent of 1971 ANS decay heat is assumed following reactor trip.

• Initial power is 97% of the Technical Specifications rated thermal power of 2893 MWt, which includes 2% to account for the calorimetric l uncertainty.

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r Several additional atsumptions have been incorporated into the LDLOCA [

reanalysis described below to accommodate the effects of Unit 1 operation with extended SGTP. These changes are discussed here. ,

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The analysis assumes that 35% of the tubes in each steam generator are j plugged. 1,n f $ level of tube plugging is expected to bound that which is I actually experienced at North Anna Unit 1. Since large break LOCA results are sensitive to SGip, this assumption is necessary to demonstrate continued ccmpliance with the 10 CFR 50.46 ECC$ acceptance criteria. In conjunction  ;

with extended SGTp, a reduced RCS total flowrate of 264400 gpm has been assumed. This value bounds the expected RCS flow associated with 35% SGTP, A

This analysis also assumes that reactor coolant system average temperature equals 586.8'F, the Technical Specifications nominal maximum allowed value. This bounds the actual Unit I nominal operating Tavg of 583'F

• i and has been shown in Virginia Power sensitivities to produce conservative large break LOCA results.

The analysis assumed a reference cosine axial power distribution with a ,

peak Heat Flun llot Channel Factor, FQ(z), of 2,11 at 95% power (equivalent to a 2,00 limit at 100% power). This value, which is more restrictive-than ,

the existing analysis, was assumed to obtain additional analysis margin for operation with extended $GTp. Figure 1 illustrates the power shape assumed,  ;

i In addition, a peak Nuclear Enthalpy llot Channel Factor, FN 6h, of 1.573 at 95% was assumed. .This is equivalent to the current Technical 1

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Specifications limit of 1.55 at 100% power and has also been assumed to obtain c eptable results for nperation with extended SGTP.

As required by Technical Specification 6.9.1.7, the Core Operating Limits l

Report (COLR) documents the applicable 1imit values of key core-related parameters for eact, reload core. The COLR will specify the appropriate '

limits which account for all design considerations, including large and small  :

break LOCA effects.

f As part of the safety evaluation to be performed by Virginia Power for restart and continued operation of North Anna 1, Cycle 9, a revised COLR will be issued. This safety evaluation, in conjunction with the COLR, will document acceptable limit values for Ley core parameters. Since the large break LOCA assumptions will impose the most restrictive requirements on the  !

allowable' FQ x K(Z) limit at each elevation, Z, the COLR for each reinad core will document the appropriate limit.

To obtain additional margin, this analysis assumed the fuel rod temperature and internal pressure values associated with the North Anna 1, Cycle 9 burnup at shutdown on December 23, 1991. Using core design predictions and fuel performance data based on the PAD 3.4 thermal model (4),

it was determined that the 10000 MWD /MTU accumulated cycle burnup correlated with 12000 MWD /MTV for the limiting fresh-fuel assembly. These assumptions ,

will bound the. fuel charact eristics for the remainder of North Anna 1, Cycle 9.

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This analysis also modified the means of implementing the single failure assumption as compared with that 4 the existing analysis. Appendix K of 10 CFR 50 requires that the ECCS containment pressure analysis assume the opec ation of all pressure reducing equipment sirce minimum pressure is conservative. This is without regard for any assumed single failures, since operation of all such equipreent is ac omplished only by energiting hil emergency equipment trains. In pi.it large break 10CA analyses, the $1ngle failure requirement has typically been conservatively implemented by assuming that loss of of fsite power occurs coincident with the LOCA and that '

- one emergency diesel- generator (EDG) f ails t start. This has the offect of removing a single train of safety injection pumps from senice, allowing flow from one high head and one low head safety injection pump. Howenr, past analyses also have assumed as required by Appendix K, that both trains of containment spray were operating. Vestinghouse sensitivity studies (10) _

have demonstrated that the limiting single failure (within the required assumptions of Appendix K) is the assumption that one low head safety injection pump falls. This assumption, combined with Appendix K 4

requirements, leaves flow available f rom two high head and one low head safety injection pump, and flow from both containment spray systems. Since the past single _ f ailure implementation- was unnecessarily conservative and nonphysical, the assumption was ' changed to provide additional -safety injection flow margin to help accommodate the effects of extended SGTP, The total assumed safety- injection flowrate has--been confirmed to be a conservative representation of actual system flow performance.

Even though the I low head /2 high head pump configuration represe,its the most limiting single failure combination, an additional restriction on-ECCS 11 3

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equipment operability requirenents is being implenented for the remainder of North Anna 1, Cycle 9. This restriction (inplemented as 5 footnote to

1. S. 3.5.2 A: tion $tatement "a") will require that 2 WlSl pumps and their associated flow paths be OPERABLE if an LHS1 pump is out of service. This change ensures that (CCS equipmei,t operability (while 1.5. 3.5.2 Action

!tatement "a" is offective) is consistent with that assumed in the large break LOCA analysis.

Using these assumptions in the BA$ll LCCS evaluation rnodel, it has been demonstrated that operation at a rnarimum reactor power of 2748 MWt with SGTP of up to 35*. in any SG will comply with the 2200*f acceptance limit of :D CFR 50.46.

5.0 ANAL.YSIS Of ifTECIS AND CONSEQUINCES 5.1 MElll00 0F ANALYSIS The large break LOCA is divided, for analytical purposes, into three phases: blowdown, refill and reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the reactor coolant system, the pressure and temperature transient within the containment and the fuel clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis.

The description of the various aspects of the LOCA analysis methodology is givea in WCAP-8339 (6). This document describes the major phenomena modelled, the interfaces among the computer codes and the features of the codes that ensure compliance with 10CFR50, Appendix K. The 5ATAN-VI, COCO, 12

l WREFLOOD, BASH and LOCBART codes, which are used in the LOCA analysis, are dc seibed in detail in WCAP-8306 (7), WCAP-8326 (B), WCAP-8171 (9), and WCAP-10266 (3), respectlyely. BASH and LOCBART are described together in Reference (2). These codes assess whether sufficient heat transfer geometry and r ., re amenability to cooling are preserved during the time spans applicable to the blowdown, refill and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient of the reactor coolant system during blowdown, and the COCO computer code calculates the containment pressure transient during all three phases of the LOCA analysis. The thermal-hydraulic response of the reactor coolant system during refill is calculated by the WREFLOOD code; for the reflood phase, this response is calculated by the BASH code, Inte.nal to the BASH code is the previously approved BART model, which is used to provide a mechanistic estimate of the heat transfer coefficient in the core during reflood.

I SATAN-VI is used to determine the RCS pressure, enthalpy ar.d density, as ,

well as che r, ass and energy flow rates in the reactor coolant system and steam generator secondary, as a function of time durinn t M blowdown phase of the LOCA. SATAN-VI also calculates the accumulator mass and pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containmes during blowdown. At the end of the blowdown, the mass and l

energy release rates during blowdown are transi'ered to the C0C0 code for use in the determination of the containment pressure response during this first phase of the LOCA. Additional SATAN-VI output data f rom the end of blowdown, including the core inlet flow rate and enthalpy, the core pressure and the core power decay transient are input to the LOCBART code.

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With input from the 5ATAN-VI code at the end of blowdown, WREFLOOD is used to determine the vessel flooding rate, the coolare vessure and v irature and the quench of vessel metal mass during the eefill phase of the a,A (time period from end of blowdown to that time when flow enters the bottom of the core). WREFLOOD is also used to calculate the mass and energy flowrates assumed to be vented to the containment for refill and reflood phases. $1nce the mass flowrate to the containment depends on core pressure, which is a function of the containment backpressure, the WREFLOOD and COC0 codes are interactively linked.

8 The COCO code, which is used tiiroughout all three phases of the LOCA analysis, calculates the containment pressure. In;at to COCO is obtained from the mass and energy flow rates asserad to be vented to the containment, as calculated by the SATAN-VI and WREFLOOD codes. In addition, conservatively chosen initial containment condit.'ons en<i an assumed mode of operation for the containment cooling system are input to COCO. These initial containment conditions and assumed modes of operation ve provided l

t in Table 2.

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i Once -- the vessel has refilled to the bottom of the core, the ret lood portion of the transient begins. ~ Information is taken from the WREFLOOD code p

characterizing the thermal-hydraulic status of the _ vessel at this time as_

well as the containmant backpressure transient as calculated by C0C0 and L

-input into the BASH code. The BASH code is used_ to calculate the thermal-hydraulic simul;t an of the RCS for the reflood phase.

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LOCBART is used throughout the analysis of the LOCA transient to calculate the fuel and clad temperature of the hottest rod in the core. Input to LOCBART consists of appropriate thermal-hydraulic outputs f rom SATAN-VI, WREFLOOD and BASH, and conservatively selected initial RCS operating conditions. These initia5 < enditions are summarized in Table 1 and Figure

1. Using this information as boundary conditions, LO:BART computes the fluid conditions and heat transfer coef ficient for the full length of the fuel rod by emnloying mechanistic models appr4 riate to the actual flow and t. eat transfer regimes. The axial power shape of Figure 1 assumed for LOCBART is a chopped cosine curve that has been historically used as the reference axial power shape for large break LOCA analyses. Verification that the tosine shape remains limiting is performed for each reload core.

5,2 RESULTS Tables 1 and 2 and Figure 1 present the initial conditions and the modes of ope ation that were assumed in the analysis, The results of this analysis are tabulated in Tables 3 and 4 for a double ended guillotine break with a CD=0.4 discharge coef ficient and 95% of rated thermal power. The mass and energy release for limiting cases are given in Tables 5 and 6. Prior Virginia Power and Westinghouse analyses employing the approved large break LOCA evaluation m<,dels have demonstrated that limiting PCT for North Anna is obtained for t'ais case. Results for other typical cases (CD=0.6, CD=0,8) are consistently >ounded by 150 - 200'F in PCT. The double ended guillotine break has been determined to be the limiting break size and location based on the sensitivity studies reported in Reference (5). The analysis resulted in a limiting peak clad temperature of 2140.8'F, a maximum local cladding 15 I _ _ _ _ _ _ _ _ _ _ _ ______________ _ ____ _______ _ _.___ _ _ _ _______ _ ___ _ _

oxidation level of 7.22% and a total core metal-water reaction of less than s

1%. The detailed results of the LOCA analysis are provided in Tables 3 through 6 and Figures 1 through 17. The attached figures show the following:

= Axial Power Shape - Figure 1 shows the cosine power shape used in this analysis.

= Core Mass Flow - Figure 2 shows the calculated core flow, both top and bottom.

= Core Pressure - Figure 3 shows the calculated pressure in the core.

= Accumulator Mass flow - Figure 4 shows the calculated accumulator flow. The accumulator delivery during blowdown is discarded until the end of bypass is calculated. Accumulator flow, however, is established in the refill-reflood calculations. The accumulator flow assumed is the sum of that injected in the intact cold legs.

= Core Pressure Drop - Figure 5 shows the calculated core pressure drop. The core pressure drop is interpreted as the pressure immediately before entering the core ir.let to the pressure just outside the core outlet.

= Break Mass Flow - Figure 6 shows the calculated flowrate out of the break. The flowrate out of the break is plotted as the sum of flow at both the pressure vessel end and the reactor coolant pump end of the guillotine break.

= Core power - Figure 7 shows the core power transient calculated by the SATAN-VI code.

= Containment V'll Heat Transfer Coefficient - Figure 8 shuws the t containment wall heat transfer coef ficient.

l-

= Containment Pressure - Figure 9-shows the calculated pressure transient, The analysis of this pressure transient is b1 sed on the containment data, reflood mass and energy release, and accumulator flow l to containment.

l l'

= Pumped ECCS Flow (Reflood) - Figure 10 shows the calculated flow of the emergency core cooling . ystem.

I 16

• Core and Downcomer Water Levels - Fiqure 11 shows the reactor ve sel downcomer snd core water levels.

• Raw Flooding Rate Integral - Figure 12 shows the raw flooding rate integrals and smooti.ed line segment integrals used in the LOCBART calculations.

= Core Flooding Rate - Figure 13 shows the resulting line segment integrals from previous figures.

- Hot Rod Clad Average Temperature - Figure 14 shows the calculated hot-spot clad temperature transient and the clad temper t're transient et the burst location. The peak clad temperature for ths initing discharge coefficient of 0.4 is 2140.8'F at 10.50 ft elevation in the core.

• Vapor Temperature - Figure 15 shows the calculated vapor temperature for the not spot and burst locations.

= Hot Rod Heat Transfer Coef ficient - Figure 16 shows the heat transfer coefficient at the hot spot location on the hattest rod.

• Hot Rod Mass Velocity - Figure 17 shows the mass velocity at the hot-spot location on the hottest fuel rod.

17

6,0 CONCLUSIONS This large break LOCA analysis was perfortred for a double ended rupture of a reactor coolant pipe with C0=0.4, at 95% of the Technical Specificution rated thermal power of 2893 MWt (2748 MWt), assuming the operating conditions specified in Tables 1 and 2. Based upon these results, the t nergency r. ore cooling system will meet the acceptance criteria as presented in 10CFR50.46 as follows: _

1. The calculated peak fuel rod clad temperature is below the requirement of 2200'F.

2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of Zircaloy in the reactra.

3. The clad temperature transient is terminated at a time when the core is still amenable to cooling, lhe localized cladding oxidation limits of 17% are not exceeded during or after quenching.

4. The core remains amenable to cooling during and after the break.

S. The core temperature is reduced and the long-term heat is recoved for an extended period of tinie. ,

18

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

1 l

l l

Table 1 INITIAL CORE CONDITIONS ASSUMED FOR THE DOUBLE ENDED COLD LEG GUILLOTINE BREAK (DECLG) ,

Calgulationa1 Input Core Power (MWt), 97% of 2893 Peak Linear Power (Kw/ft) 97% of 11,35 Peak Heat Flux Hot Channel factor, FQ(t) 2.11*

Peak Nuclear Enthalpy Hot Channel Factor, F NAh 1.573**

Accumulator Water Volume (f t*/ accumulator) 1025 Reactor Vessel Upper Head Temperature (Thot)

Limitina f_uel Region and Cycle Cycle q P,egio0_

Unit 1 9*** All Regions

• Equivalent to a 100% power limit of 2.00

• • Equivalent to a 100% power limit of 1,55

• • • Analysis is only applicable to Cycle 9 from 10000 MWD /MTU to EOC.

19

, L Table 2 CONTAlisMENf DATA Net Free Volume (ft') 1.916 x 106 a

Initial Conditions Pressure (psia) 9,608 Temperature ('F) 86.0 RVST Temperature ('F) 40.0 Outside Temperature ('F) -10.0 Spray Systema-Number of Pumps Operating 2 Runout Flow Rate (per pump) 2000 gpm Time in Which Spray is Effective 59 sec Structural Heat Sinks a '

Thickness (in.) Area (ft'), with allowance for uncertainties 6- concrete 8,393 12 concrete 62,271 18- concrete- 55,365 24 _cuncrete 11,591 27 concrete 9,404 36 ' concrete 3,636

.375 steel, 54 concrete 22,039

.375 -steel, 54 concrete 28,393

.500- steel, 30 concrete 25,673 26.4 concrete, ,25 steel, 120 concrete 12,110

.407 stainless steel 10,527

.371 steel 160,328

.882 steel 9,894

.059 steel 60,875 a .ESee I.'FSAR Section 6.3.3.12 for a detailed breakdown of the corttinment bcat sinks and for justification of the_other input parametcFs use to calcualte containment pressure.

20

Table 3 11ME SEQUENCE OF EVLNTS DECLG (Cd=0.4)

Description of Parameters (seconds)

End of Bypass / End of Blowdown (sec)-

30.6685 Safety System Actions Reactor Trip (Sec) 0.549 Accumulator Injection (Sec) 13.6 SI Signal Generated (Sec)- 3.8 Pump SI Starts.(Sec) 30.8 Bottom of Core Recovery (sec) 44.65 Accumulator Empty (sec) 54.16 l

l l

21

- w- s- *r--e==

Table 4 RESULTS FOR DECLG DECLG (Cd=0.4)

Description of Parameters (seconds)

Pc ' Clad Temperature e c) 2140.8 Peat Clad Location (ft) 10.50 Hot Rod Burst Data Location (ft) 6.00 Time (Sec) 51.63 Zr/H2O Results Data Local Maximum Reaction (%) 7.22 Location of Maximum (ft) 10.25 Total Reaction (%) < 1.0 22

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

Table 5 REFLOOD MASS AND ENERGY RELEASES DECLG (C[bO.4)

Time lotal Mass Total Energy (Sec) Flow Rate Flow Rate (lbm/sec) (10' Btu /sec) 44.f ac 0.0 0.0 45 32 .00687

-45.4h .0008' 45.521 .00b60 45.721 .n .00615 45.821 .5176 .00671 56.982 61.86 0.6939 75.582 117.18 0.9059 97,032 282.84 1,3010 120.782 296.24 1.2622 146.832 319.50 1.2341 220.082 350.41- 1.2003 23

1 Table 6 BROKEN LOOP ACCUMULATOR FLOW 10 CONIAINMENT DECLG (00:0.4)

TIME (sec) MASS FLOW RATE (Ib/sec) 0.0 4096.76 1.01 3692.74 3.01 3157.39 5.01 2804.86 7.01 2546.44 10.01 2257.25 15.01 1928.86 20.01 1706.87 25.01 1550.55 29.01 1622.40 24

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

i I

HEfERENCES r

(1) Letter from W. L. Stewart (Virginia Power) to U.S. NRC, " North Anna Power Station Unit 1 Proposed Technical Specification Change - ,

Reduced Minimum RCS flow Rate Limit to Support Increased Steam .

Generator Tube Plugging level," Serial No.92-018, January 8, 1992.

(2) WCAP-10266-P-A, Rev. 2, "The 1981 Version of the Westinghouse ECfS Evaluation Hodel Using the BA5H Code," March 1987.

i 9

(3) WCAP-10444-P-A, Addendum 2, " Vantage $H fuel Assembly," April 1988.  :

i

. (4) WCAP-8720, and Addendum 2. " Improved Analytical Models Used in Westinghouse fuel Rod Design Computations " October 1976.  ;

I (5) WC/P-8356, " Westinghouse ECCS Plant Sensitivity Studies," July 1974.

(6) WCAP-8339, "Westingnouse EC' Evaluation Model-Summary," July 1974  ;

(7) WCAP-8306, " SATAN-VI Program: Comprehensive Space-lime Dependent Analysis of Eoss-of-Coolant," June 1974.

(8) WCAP-8326, " Containment Pressure Analysis Code (C0CO)," June 1974. .

(9) WCAP-8171, " Calculational Model for Core Reflooding After a Loss-of-Coolant Accident (WREfl.00D Code) " June 1914.

'(10) WCAP-8471-P-A,'"The Westinghouse ECCS Evaluation Model:

Supplementary information," April 1975.

i i

i r

42 i i

- y , - - - - - . - - - - - - - - - - - - - - - - - --

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t I

h ATTACHMENT 4 f

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I 10 CFR 50.92 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION l t

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i VIRGINIA ELECTRIC AND POWER COMPANY' P

%m-<- . . - ,re- . .+ ar-c ya rm y-.f - .,y enwe,mw.,s n ,- r,w-n,.,: .._,.e.,,m~mm.----I+_b_.m_-m

10 CFR 50.92 No Significant Hazards Consideration Evaluation In accordance with the requirements of 10 CFR 50.91(a), the proposed change to the North Anna Power Station Unit 1 Faci l;ty Operating License has been evaluated against the entena described in 10 CFR 50.92 and it has been determined that the proposed amendment to the operating license involves no significant hazards consideration. The basis for this determination is as follows:

North Anna Power Station Unit 1 is currently involved in a mid cycle steam generator inspection outage. An extensive oddy current inspection of the North Anna Unit 1 steam generator tubes is being performed using very conservative analysis guidelines and plugging criteria. As such, a substantially increased number of tubes are expected to be plugged.

The predictions of potential steam genmator tube plugging during the current mid-cycle outage are such that the effects of increased RCS toop resistance on the large break LOCA analysis would not permit full rated power operation for the remainder of North Anna Unit 1, Cycin 9. The existing large break LOCA analysis has obtained margin by taking credit for available Cycle 9 core characteristics and will not support 100% power operation with more than 30% steam generator tube plugging. The large break LOCA analysis presented in Sectiors 3.0 through 5.0 of the attached safety evaluation extenas this steam generator tube plugging limit value to 35%, but with a reduced power lovel of 96% of rated thermal power. At this reduced power level, all analyses meet the requirements of 10 CFR 50.46 and Appendix K to 10 CFR Part 50.

Because the large break LOCA presents the limiting considerations for core power and total core power peaking, it was necessary to reduco the maximum core power level to 2748 megawatts (thermal) and the maximum allowable Hot Channel Peaking Factor (Fq) to 2.11 at the core mid plane. The change to the power levelis proposed as a modification to license condition 2.D.(1), Maximum Power Level, by adding a footnote limiting maximum reactor power to 2748 megawatts (thermal) until steam generator replacement is accomplished.

In addition, an associated change to the Technical Specifications is required to accommodate the effects of the revised assumptions for the large break LOCA analysis. The proposed change to the Technicel Specifications will impose more t restrictive equipment operability requirements for the Emergency Core Cooling System (ECCS). This is accomplished by modifying the Action Statemen; "a" of Specification 3.5.2 to ensure that both low head safety injection pumps or one low head injection pump and two high head safety injection pumps remain operable during power operation. This change effectively maintains consistency betwean the Technical Specification Action Statements and the revised assumptions for thG iarge break LOCA analysis.

Page 1 of 3 l

____~i--___-_-_--.---__--_-____.__---_____ _ _ . - _ - - - _ _ - - - _ - - _ _ _ . _ - _ _ - - _ _ - _ . - - - - -

e i, 6 Further, a revised K(Z) surveillance function and a reduced Enthalpy Rise Hot Channel Factor were utilized to provida additional analysis margin. With these changes, the analysis supports power operation at up to 95% of rated thermal power for North Anna Unit 1 for the remainder of Cycle 9. Changes to the peaking factor and K(Z) survel!!ance function will be accomplished via the Technical Specifications Core Operating Limits Report (COLR).

The large break LOCA analysis assumed uniform steam gene ator tube plugging of 35% which supports operation with peak steam generator tube plugging levels up to 35%. With the exceptlon of the parameters described above, which will be

-incorporated via the proposed license change and the forthcoming COLR, all analysis parameters were equivale.11 to, or conservative with respect to, those assumed in the existing analyses. All analysis parameters are expected to be conservative with respect to actual plant conditions for the remainder of North Anna Unit 1 Cycle 9.

Virginia Electric and Power Company has reviewed the proposed license condition change relative to operation of North Anna 1.' nit 1 with increased steam ganerator tube plugging and determined that the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is that this change:

1. Does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The impact of the increased level of steam generator tube plugging (up to 35%

peak) with a maximum reactor power of 95% on the large break LOCA was analyzed. The analysis demonstrated that operstion with increased steam generator tube plugging will not result in more severo consequences than those of the currently applicable analyses.

The probability of occurrence of these accidaits is not increased, because an increased level of steam generator tube plugging as an initial condition for the accident has no bearing on the probability of occurrence of these accidents.

2. Does not creato the possibility of a new or different kind of accident from any accident previously evaluated.

The implementation of the increased eMam generator tube plugging large break LOCA analysis into the North Anna unit 1 design basis will not create the possibility of an accident of a different type than was previously evaluated in the UFSaR. No changes to plant configuration or modes of operailon are implemented by the revised accident analysis. Therefore, no new mechanisms for the initiation of accidents are created by the implementation of the analysis.

L 3. Does not involve a significant reduction in a margin of safety.

L l The North Anna Unit 1 operating characteristics, and accident analyses which

support Unit 1 operation, have been fully assessed. The results of the revised i

large break LOCA analysis demonstrates that the consequences of this accident Page 2 of 3 i

l t

s___ . _ _ . _ _ _ _ . _ _ . . _ _ .

. _ . . . _ _ . . ~ . .

soo are not increased as a result of the increased steam generator tube plugging up to 35% with a maximum reactor power of 95% The results of the accident analysis remain below the limits established by the currently applicable analyses.

Therefore, there is no significant reduction in the margin of safety, Based on the above significant hazards consideration evaluation, Vir0i nia Electric and Power Company concludes that the activities associated with this proposed license condition change satisfies the no significant hazards cons!deration standards of 10 CFR 50,92(c) and, accordingly, a no significant hazards consideration finding is justified.

4 l

Page 3 of 3

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