ML20117G589

Proposed Tech Specs,Supporting NRC Approval of Deletion of Requirements to Maintain Upper Head Injection Sys

ML20117G589
Person / Time
Site:	McGuire, Mcguire
Issue date:	05/09/1985
From:	DUKE POWER CO.
To:
Shared Package
ML20117G583	List: ML20117G581 ML20117G589
References
NUDOCS 8505140033
Download: ML20117G589 (22)

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j Attachment 1 DUKE POWER COMPANY MCGuire Nuclear Station Proposed Technical Specification Revision Deletion of Upper Head Injection System 8505140033 850509 PDR ADOCK 05000369 P PDR

d Attachment 1 Technical Specification Revisions

.The required revisions to the McGuire Nuclear Station Technical Specifications associated with the deactivation of the Upper Head Injection (UHI) System are provided via the attached pages and the brief discussions of each revision given below:

3/4.4.6 - Reactor Coolant System Leakage The deletion of the UHI System will involve capping of the reactor vessel upper head penetrations as close as practicable to the upper head.

Associated piping and valves are thus removed from the Reactor Coolant System and leakage verification is no longer applicable for the UHI related equipment in Table 3.4-1.

3/4.5.1.1 - ECCS, Cold Leg Injection The nitrogen cover pressure will be increased to a minimum value of 585 psig in order to enhance cold leg injection water delivery during LOCA scenerios. Other specifications related to the accumulators remain unchanged.

3/4.5.1.2 - ECCS, Upper Head Injection The specifications associated with the maintenance of the UHI System within specified tolerances will be deleted.

3/4.6.1 - Primary Containment Table 3.6-1 will be revised to reflect the sealing of UHI related containment penetrations.

3/4.6.3 - Containment Isolation Valves Table 3.6-2 will be revised to reflect the removal of containment isolation valves associated with UHI containment penetrations.

3/4.7.8 - Snubbers The tables provide in Section 3/4.7.8 which describe the types and quantity of snubbers utilized in various systems will be revised via a future submittal to reflect deletion of the UHI System.

3/4.8.4 - Electrical Equipment Protective Devices Table 3.8-la and 3.8-lb will be revised to reflect the deletion of the UHI System and related containment penetration conductor overcurrent protective devices.

~ TABLE 3.4-1

( .

REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE VALVE NUMBER FUNCTION MC-1562-2.0 *

-NI60 NI71 Accumulator Discharge NI59 Accumulator Discharge NI70 Accumulator Discharge Accumulator Discharge MC-1562-2.1 '

NI82 NI94 Accumulator Discilarge NI81 Accumulator Discharge NI93 Accumulator Discharge Accumulator Discharge MC-1562-3.0 NI134 NI159 Safety Injection (Hot Leg)

NI156 Safety Infection (Hot Leg)

NI128 Safety Injection (Hot Leg)

NI124 Safety Injection (Hot Leg)

NI160 Safety Injection (Hot Leg)

NI157 Safety Injection-(Hot Leg)

NI126 Safety Injection (Hot Leg)

Safety Injection (Hot Leg)

C NI129 NI125 Safety Injection (Hot Leg) ,

Safety Injection (Hot Leg)

MC-1562-3.1 HI165 NI167 Safety Injection / Residual Heat Removal (Cold Leg)

NI169

• Safety Injection / Residual Heat Removal (Cold Leg)

NI171 Safety Injection / Residual Heat Removal (Cold Leg)

NI175 Safety Injection / Residual Heat Removal (Cold Leg)

NI176 Safety Injection / Residual Heat Removal (Cold Leg)

NI180 Safety Injection / Residual Heat Removal (Cold Leg)

NI181 Safety Injection / Residual Heat Removal (Cold Leg)

Safety Injection / Residual Heat Removal (Cold Leg)

MC-1562-4.0 NI250 Upper Head Injection NI251 Upper Head Injection NI252 NI253 ddai. Upper Head Injection NI249 Upper Head Injection NI248 UpperHeadInjection Upper Head Injection i

MC-1561-1Y0

' N01B^

N02A* Residual Heat Removal --

Residual Heat Removal

.L

" Testing per Specification 4.4.6.2.2d not applicable due to positive indication of valve position in Control Room.

Note 1:

Upon the deactivation of the UHI System by removal of related components and piping and modifications to the Cold Leg Accu-mulators, this specification is no longer applicable.

McGUIRE - UNITS 1 and 2 3/4 4-21

3/4.5

~

EMERGENCY CORE COOLING SYSTEMS _

,, 3/4.5.1 ACCUMULATORS '

_COLO LEG INJECTION 4

LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumu ator shall be OPERABLE with:

a.

The isolation valve open, b.

A contained borated, water volume of between 8022 and 8256

c. ,

A baron woru vuo concentration omrm en : of between 1900 and 2100 ppm, .

d..LJA t , w innitrogen o m cover4 pressure memoven u of between 430 and 484 psig, sad-

e. .+r.4 4 c- pm w ee . w h.,een A water level and pressure channel OPERABLE.

rer =J 639 ps,ey APPLICABILITY: MODES 1, 2, and 3*. .

ACTION:

a.

With one cold leg injection accumulator inoperable, except as a of a closed isolation valve, restore the inoperable accumulator to OPERABLE the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> status'within and in HOT SHUTDOWN 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> orwithin be intheat least HOT following 6 hou b.

With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isola-SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. tion valve or SURVEILLANCE REQUIREMENTS 4.5.1.1.1

a.

Each cold leg injection accumulator shall be demonstrated OPERA At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

1)

Verifying the contained borated water volume and nitrogen cover pressure in the tanks, and 2)

Verifying valve that each cold leg injection accumulator isolation is open.

L

• Pressurizer pressure above 1000 psig.

McGUIRE - UNITS 1 and 2 3/4 5-1 Amendment No..!(Unit 1)

Amendment No.O (Unit 2)

_ EMERGENCY CORE COOLING SYSTEMS ..

' UPPER HEAD INJECTION LIMITING CONDITION FOR OPERATION ~

3.5.1.2 EachUpperHeadInjectionAccum'uiatorSystemshallbeOPER a.

The isolation valves open, b.

of borated water having a concentration of boron, and c.

The nitrogen bearing accumulator pressurized to between 120 1264.psig.

APPLICABILITY: MODES 1, 2 and 3.*

w em oai e t ~o y un s . ; f. ore ia 'i o + *rr "y *

• ACTION: p a.

With the Upper Head Injection Accumulator S as a result of a closed isolation valve (s),ystem inoperable, restore the Upperexcept Head '

r, Injection Accumulator System to OPERABLE status withinfi 1 in atthe within least HOT following STANDBY 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> '

b.

With the Upper the isolation valveHead s Injection Accumulator System inoperable

'. being closed, either immediately open the isolationwithin SHUTDOWN valve the next(s) o(r)be' in HOT STAND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  !

SURVEILLANCE REQUIREMENTS 4.5.1.2 OPERABLE: Each Upper Head Injection Accumulator System shall be demo a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

1)

Verifying the contained borated water volume and nitrogen pressure in the accumulators, and 2)

Verifying that each accumulator isolation valve is open.

• Pressurizer Pressure above 1900 psig.

(.

McGUIRE - UNITS 1 and 2 3/4 5-3 Amendment No. ' (Unit 2)

Amendment No. ' - (Unit 1)

(

IT O '

TABLE 3.6-1

?

g m SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS m

PENETRATION g NUMBER SERVICE TEST 4 RELEASE LOCATION TYPE m M216 e Pressurizer Relief Tank Makeup Auxiliary Building Type C M212 Nitrogen to' Pressurizer Relief Tank Auxiliary Building Type C N

M259 Reactor Makeup Water Tank to NV System Auxiliary Buf1 ding Type C .

M373 Ice Condenser Glycol In Auxiliary Building Type C M372 Ice Condenser Glycol Out Auxiliary Building Type C M330 i Nitrogen to Accumulators Auxiliary Building Type C ,

7 M321 Safety Injection Test Line puxiliaryBuilding

, Type C,.

fM348 Upper Head Injection Test Line Auxiliary Building I Typ d hO M374 Containment Floor Sump Incore Instrument Sump Discharge Auxiliary Building Type C M360 Reactor Coolant Drain Tank Gas

• Space to Waste Gas System Auxiliary Building Type C M375 Reactor Coolant Drain Tank Heat -

Exchanger Discharge Auxiliary Building Type C M356 Equipment Decontamination Auxiliary Building Type C M235 Pressurizer Sample Auxiliary Building Type C

M309 Reactor Coolant Hot Leg Sample Auxiliary Building Type C l M322 Component Cooling to Component Cooling Drain Tank Auxiliary Building Type C Note 1

Upon the deactivation modifications of the to @ C*f141 LaUHI System by removal of related components and piping and

~

i .

2 TABLE 3.6-2 (Continued)

E CONTAINMENT ISOLATION VALVES c -

MAXIMUM z

VALVE NUMBER

--e FUNCTION ISOLATION

1. TIME (SEC)

[ Phase "A" Isolation (continued) '

k NI-968

, NI-120B Test NDR Outside Containment Isolation NI-1228 , Safety Injection Pump'to Accumulator Fill Line Isolation $10 NI-255B llot Leg Injection Check HIl24, nil 28 Test Isolation 510 HI-25BA UllI Check Valve Test Line Isolation $10 UllI Check Valve Test Line Isolation $10 NI-2648 y pjo{e i gill Check Valve Test Line Isolation <10 NI-266A NI-267A UllI Check Valve Test Line Isolation 310

<10

~

u UliI Check Valve Test Line Isolation g N:1-3A 310

,  ?!M-6A Pressurizer Liquid Sample Line Inside Containment Isolation 4 NM-78 Pressurizer Steam Sample Line Inside Containment Isolation $15 m (Cl-22A Pressurizer Sample lleader Outslide Containment Isolation $15 Nff-25A NC flot Leg #1 Sample Line Insidi containment Isolation 515 NM-268 NC llot Leg #4 San!ple Line Inside Containment Isolation 515 NM-728 flC llot Legs Sample lide. Outside Containment Isolation $15 flM-758 NI Accumulator A Sample Line Inside Containment Isolation $15 1111- 7 8 8 NI Accumulator B Sample Line Inside Containment Isolation $15 flM-818 NI Accumulator C Sample Line Inside Containment Isolation $15 NM-82A NI Accumulator 0 Sample Line Inside Containment Isolation $15 Hit-187A# NI Accumulator Sample Hdr. Outside Containment Isolation $15 NM-190A# SG A Upper Shell Sample Containment Isolation Inside $15 H!1-1918# SG A Blowdown Line Sample Containment Isolation Inside $15 NM-1978# SG A Sample lidr. Containment Isolation Outside $15 NM-200B# SG B Upper Shell Sample Containment Isolation Inside 515 NM-201A# SG B Blowdown Line Sample Containment Isolation Inside 515 Hit-207A# SG B Sample Hdr. Containment Isolation Inside 515 NM-210A# SG C Upper Shell Sample Containment Isolation Inside 515 NM-2118# SG C Blowdown Line Sample Containment Isolation Inside $15 NM-2178# SG C Sample lidr. Containment Isolation Outside $15 NH-220B# SG 0 Upper Shell Sample Containment Isolation Inside $15 SG 0 Blowdown Line Sample Containment Isolation Inside $15

$15 Note 1:

Upon the deactivation of the UHI System by removal of related components <

old T ee kvecemmulatorsdiunnem

t f O 5

{ TABLE 3.8-la -(Continued)

UNIT 1 CONTAINHENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES E

U TRIP SETPOINT OR RESPONSE

" CONT. RATING TlHE g DEVICE NUMBER & LOCATION (AMPERES) (SECON05) SYSTEM PCWERED

{

2. 600 VAC-HCC (Continued) 1EHXA-2 28 Primaar Bkr 20 45 e 60A Backup Fuse 20 H.A.

N2 to Prt Cont Isol Inside Viv 1NCS4A 1EHXA-2 2C Primary Bkr 20 Backup Fuse 45 e 60A RCP Mtg Brg 011 Fill Isol 20 H.A. V1v INC196A

- om IEHXA-2 3A O

" Primary Bkr 30 Backup Fuse 45 0 90A Accumulator IA Disch Isol 30 H.A. Viv 1NIS4A ,

4 IEHXA-2 38

! Primary Bkr 30 l Backup Fuse 45 0 90A Accumulator IC Disch Isol 30 N.A.  ;

i V1v 1NI76A j y IEHXA-2 3C i 5E Primary Bkr 20 i

8. $ Backup Fuse 45 0 60A Test Ildr Inside Cont Isol i 3& 20 H.A. Viv ]NI9SA i SS 1EHXA-2 4A

=" Primary Bkr ilde b

P[ Backup Fuse 20 20 45 0 60A N.A. UHI check Viv Test Line Isol

! dit V1v 1N1266A gg NOTE 1:

, 3. :.' Upon the deactivation of the UHI System by removal of

! "" related components and piping and modifications to the UU Cold Leg Accumulators, this specification is no longer applicable.

O TABLE 3.8-la (Continued) g e UNIT 1 CONTAINHENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE E

Q TRIP SETPOINT OR RESPONSE

'^ CONT. RATING TIME DEVICE NUMBER & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED

$ 2.

600 VAC-HCC (Continued) a.

1EHXA-2 48 Primary Bkr i gi 20 45 0 60A Backup Fuse J 20 H.A. Util check Viv Test Line Isol Viv INI267A 1EHXA-2 4C Primary Bkr 20 w Backup Fuse 45 0 60A A 20 N.A. Accum 1A Vent to INC34 for Blkout Viv IN1430A m 1EHXA-5 IB 4

N Primary Bkr 20 Backup Fuse 45 0j60A Pzr Steam Sample Line Inside 20 N. A. '

Cont Isol V1v 1NH3A IEHXA-5 28 .

! Primary Bkr 20

, Backup Fuse 45 0 60A Pzr Steam Sample Line Inside 20 N.A.

Cont Isol V1v It&l6A 1EHXA-5 3B Primary Bkr 20 45'O 60A Backup Fuse NC llotleg 1A Sample Line Cont 20 H.A.

Isol Viv INH 22A 1EHXA-5 2D Primary Bkr 20 Backup Fuse 45 0 60A NC llotleg 10 Sample Line Cont 20 N.A.

- Isol Vlv INH 25A IEHXA-2 7A Primary Bkr 20 Backup Fuse 45 0 60A 20 N.A. SG 1A Upper Shell Sample Cont Isol Viv INMlB7A f?

NOTE 1:

Upon the deactivation of the UHI System by. removal of related components and piping no andapplicabic.

longer modifications to.the Cold Leg Accumulators, this specification is L _ _ _-

1 y .

[

C TABLE 3.8-1b (Continued)

(( UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES E TRIP SETPOINT OR RESPONSE El CONT. RATING TIME

(( DEVICE NUMBER & LOCATION (AMPERES) .

(SECONDS) SYSTEM POWERED ll 2. 600 VAC-MCC (Continued) 2EMXA-2 48 Primary Bkr bkde d. 20 45 0 60A UHI Check Viv Test Line Isol Backup Fuse 20 N.A. Viv 2NI267A 2EMXA-2 4C Primary Bkr 20 45 0 60A Accum 2A Vent to 2NC34 for g; Backup Fuse 20 N.A. Bikout V1v 2NI430A os 2(MXA-5 IB .

Es Primary Bkr 20 45 0 60A Pzr Steam Sample Line Inside Backup Fuse -20 N.A. Cont Isol Viv 2NM3A 2EMXA-5 2C Primary Bkr 20 45 0 60A Pzr Steam Sample Line Inside Backup Fuse 20 N.A. Cont Isol Viv 2NM6A l 2EMXA-5 3B Primary Bkr 20 45 0 60A NC Hotleg 2A Sample Line Cont Backup Fuse 20 N.A. Isol V1v 2NH22A 2EMXA-5 2D Primary Bkr 20 45 @ 60A NC Hotleg 2D Sample Line Cont Backup Fuse - 20 N.A. -

Isol Viv 2NH25A 2EMXA-2 7A Primary Bkr 20 45 0 60A SG 2A Upper Shell Sample Cont Backup Fuse 20 N.A. Isol Valve 2NM187A NOTE 1:

Upon the deactivation Q UHI System by removal of related commonen3e md] mgagm________ ___

s -

{

c TABLE 3.8-lb (Continued) k

• UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTE

_E g TRIP SETPOINT OR. -

RESPONSE

j g CONT. RATING T1HE DEVICE NUMBER & LOCATION (AMPERES)

(SECONOS) SYSTEM POWERED 1

k 2.

600 VAC-MCC (Continued) 2EMMA-2 28 Primary Bkr 20 Backup fuse 45 9 60A 20 N.A. N2 to Prt Cont Isol Inside Vlv 2NC54A 2EMXA-2 2C i Primary 8kr 20 Backup Fuse 45 9 60A RCP Mtg Brg 011 Fill Isol 20 N.A.

m Vlv 2NC196A 2EMMA-2 34 i A Primary Skr i 30 45 9 90A

, Backup Fuse 30 N.A. Accumulator 2A Disch Isol

' V1v 2NI54A  !

2EMMA-2 38 Primary Skr 30

Backup fuse 45 9 90A 30 N.A. Accumulator 2C Disch Isol Vlv 2NI76A 2EMXA-2 3C {

)i

l. { Primary 8kr 20 Backup Fuse 45 9 60A O{g 20 Test lide Inside Cont Isol s s N.A. .,

i es Vlv 2H195A L

2EMMA-2 44 4

g ," Primary Skr Noted. 20

. o Backup Fuse 45 9 60A l

b g-20 N.A. UNI check Viv Test Line Isol V1v 2NI266A '

'. 3322 NOTE 1:

i Upon the deactivation of the UHI System by removal of related components ad

.CS modifications to the Cold Leg Accumulators, n piping and 4 this specification is no longer applicable. *

- _ ~' '

I 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS c,},} [,q }.

The OPERABILITY of each Reactor Coolant System (RCS)3 8ccumulatore I that a sufficient volume of borated water will be immediately forced into the below the pressure of the accumulators. reactor core through each o This initial surge of water into the core provides the initial coc,.ing mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis ar met.

The accumulator

" operating bypasses" power operated isolation valves are considered to be in the context of IEEE Std. 279-1971, which requires that conditions are not met. Infunction bypasses of a protective be removed automatically whenever permiss addition, as these accumulator isolation valves fall to meet single failure criteria, removal of power to the valves is required

( The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that suff!

emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.

each ECCS subsystem provides long-term core cooling capability in theIn addition recirculation mode during the accident recovery period.

With the RCS temperature below 350'F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable

, reactivity condition of the reactor and the limited core cooling requirements.

\ L -

\

McGUIRE - UNITS 1 and 2 B 3/4 5-1

O

. Attachment 2 s

Justification & Safety A4.alvsis l

Introduction The Upper Head Injection System (UHI) was added to the Emergency Core System (ECCS) of the McGuire Nuclear Station during initial licensing in order to regain operating flexibility lost due to the impact of the Ice Condenser Containment design upon the 10CFR50.46 Appendix K ECCS Evaluatio Ice Condenser design results in a reduced containment back pressure during .

design basis loss of coolanc accident decreasing the amount of steam that ca be largevented break from the loss of Reactor coolant Coolant System (RCS) to the containment durin accident.

The ECCS analysis results therefore require more limiting restrictions on normal plant operations in order to for dry containments. satisfy Appendix K requirements for the Ice Condenser des to allow load follow operation, the UHI System and associated p fications were Westinghouse added to the McGuire units to supplement the conventional ECCS.

The development of improved analytical models combined with the numerous operational and design problems related to the UH1 System at McGuire have led to the decision to proceed with the deactivation of UH1 accumulators, associated components and piping.

Due to the complexity and rigid require-ments placed upon its performance, the UH1 System has introduced concerns regarding appropriate water volume delivery, nitrogen injection into the Reactor Coolant System, fluid mixing behavior, and concerns of the increased number of plant mode changes involved with UH1 problems and their resolutio In addition, related the plant operational performance has been impacted by the UHI problems.

Meanwhile, the ability to model and predict plant behavior during a loss of coolant accident has improved significantly since the decisi was made to install the UHI System at McGuire.

Core power peaking factor limits which provide adquate operating flexibirity may be approved by the .NRC. e In order to dr.termine the feasibility of the deletion of the Ulli System, a scoping study was performed to determine the McGuire response to a Double-Ended.

cific paraucters.Cold Leg Guillotine (DECLG) break assuming values for spe-The UHI System was assumed deactivated, but the vessel internals associated with UH1 plants were modeled and provided a benefit in plant performance during the LOCA transient simulated. The Cold Leg Accumulator cover Westinghouse. plant pressure value). was raised to 600 psia (a typical 4 loop The resistance factor applied to the Cold Leg Accumulator Discharge piping was reduced to reflect the planned removal of the flow orifice.

design to compliment the performance of the UHIThe System.)(This BART core orifice refload flow and heat thereal-hydraulics code. transfer methodology was utilized with the REFLOOD The scoping study analyzed the 0.6 DECLC break 2-1 g .O

and resulted peaking in (F factor a peak clad temperature of 1960*F with a corresponding core q) of 2.20. Figures 1, 2, and 3 provide the results of the scoping study in graphical form and also provide an indication of the benefit of the UllI related internals in comparison to the standard Westing-house design.

The proposed Technical Specification revisions included herein include those presently identified as being required once approval to delete Ulti is granted and the system is removed during a refueling outage. NRC is requested to initiate the deletion of review of the technical scope of work planned, and to approve UllI. Upon completion of the analysis described herein, it is expected that the existing value of FQ (z) will be confirmed to be valid.

PLANNED ANALYSIS EFFORT The analyses to be performed and submitted at a later date will utilize the BART and BASit computer codes. The additional benefit provided by the basil code is expected to allow a core power peaking limit of 2.32 while continuing to meet the requirements of Appendix K. (This is above the existing Technical Specification value of 2.26).

The scoping study will be confirmed by the complete analysis of the large break LOCA showing the ECCS performance of the McGuire Nucicar units with UHI deactivated is not significantly different than other Westinghouse four loop plants when current ECCS analysis methodology (BART/ basil) is utilized and the Cold Leg Accumulators are adjusted. The units can be operated with UllI deleted at a peaking factor that will allow full power and full load follow operation while continuing to satisfy the existing conservative requirements of Appendix K.

The most significant application of the Ul!I System, water delivery during the blowdown phase of a LOCA involving a large break of a cold Icg pipe, will be shown to be unnecessary using the existing Westinghouse approved Evaluation Model (including BART Technology) and the BAsti methodology which is expected to receive NRC approval during 1985. Other transients which may result in the injection of Ulli water, will also be analyzed using approved or soon to be approved methods in order to demonstrate that all safety criteria requirements remain catisfied. The spectrum of small break loss of coolant accidents will be analyzed using the NOTRUMP computer code. The Steamline Break will be analyzed using the existing methodology as described in the McGuire FSAR.

Additional details of the confirmatory evaluations to be performed follow.

Large Break LOCA Analysis The analysis of a large break LOCA transient is divided into three parts:

(1) Blowdown, (2) Refill, (3) Reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transients in the RCS, the pressure and temperature transient within the containment, and the fuel and clad temperature transient of the hottest rod in the core. Based upon these for theconsiderations, a system of interrelated computer codes has been developed analysis of a LOCA.

2-2

The description of the various aspects of the LOCA analysis methodology is provided in the references. These documents describe the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the acceptance criteria. The SATAN VI, LOTIC, BART, BASH, WREFLOOD, and LOCTA-IV codes which are used in the LOCA analyses are described in detail in the references. These codes are used to assess the core heat transfer and to determine if the core geometry remains amenable phases of theto LOCA.

cooling throughout and subsequent to blowdown, refill, and reflood The SATAN-VI transient in the RCS during computer code analyzes the thermal-hydraulic blowdown.

and mass / energy releases during reflood.WREFLOOD calculates the refill phase The BASH code is used to calculate the RCS thermal-hydraulic behavior during the reflood portion of the LOCA.

The BART computer code is used to calculate fluid and heat transfer conditions in the core during reflood. The LOTIC code is used to calculate the containment pressure transient during all three phases of the LOCA analysis.

the LOCTA-IV computer code is used to compute the thermal transient of theSimilarly, hottest fuel rod during the three phases. Fuel parameters input to the LOCTA-IV code are taken from the most recently approved version of the PAD code.

The large break LOCA analyses to be completed in order. to justify the deletion of the UHI System will include a range of Moody break discharge coefficients for a Double-Ended Cold Leg Guillotine (DECLG) break.

The bases used to select the numerical values that are input to the analysis have been conservatively determined from extensive sensitivity studies (Refer to Reference 10). In addition, the requirements of Appendix K regarding specific model features are met by selecting models which provide a significant overall conservatism in the analysis. The assumptions made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core peaking factors, the containment pressure, and the performance of the ECCS. Decay heat generated throughout the transient is also conservatively calculated as required by Appendix K.

Small Break LOCA Analysis The small break LOCA analysis will be performed per the most recent SBLOCA evaluation Model as described in Reference 9. The model and associated computer code, NOTRUMP (Reference 8), are expected to be approved by the NRC in May 1985.

The new Evaluation Model incorporates the requirements of NUREG-0737 Item II.K.3.30.

The analysis shall be submitted at a later date and shall demonstrate that the deletion of UHI does not impact the ability of the McGuire units to satisfy all pertinent safety and regulatory requirements.

Steamline Break A Chapter 15 Steamline Break analysis will be conducted using the LOFTRAN computer code.

Statepoints will be generated and DNB design basis limits will be verified using the THINC computer code. Although DNB and possible clad perforation follow-ing a Steamline Break are not necessarily unacceptable, the analysis will demon-strate that McGuire Safety Systems are adequate to prevent DNB from occuring for any rupture assuming the UHI System has been removed while maintaining other con-servative assumptions such as having the most reactive control rod stuck in the fully withdrawn position.

2-3

o Containment Analysis For LOCA The Ice Condenser shall be shown to limit the containment pressure to a value less than the design pressure of 15 psig for all reactor coolant pipe break sizes up to and including a double-ended severance. The containment peak, pressure calculation will be performed using a modified version of the most recent ANS Decay Heat Standard (1.2 conservative factor is maintained). The use of this decay heat curve has been previously sumbitted and approved by the NRC. The analysis will also employ the Westinghouse 1979 Mass / Energy Release Model which includes credit for steam condensation in the cold legs by safety injection water.

The 1979 Model has been previously submitted for several dry containment analyses and approval is expected in the near future.

The use of the revised decay heat standard and the 1979 Mass / Energy Release Model is not required design limit. to ensure the calculated peak pressure results are less than the The modifications are made to update the McGuire analysis with model improvements made since the initial McGuire calculations. The analysis will demonstrate that the deletion of UHI is justifiable in light of the calculated peak containment pressure remaining below the 15 psig design limit.

Containment Analysis For Steamline Break Prior to the NRC resolution of the "Superheat" issue and approval of the Ice Con-denser Model, a sensitivity study will be conducted for Mass / Energy cases where UHI is determined to impact the results. The study will be used to assess the impact of UHI removal upon the containment temperature and pressure calculations related licensingtodiscussions.

the Steamline Break Transient and will provide the basis for initial Af ter the NRC approval of the Mass / Energy and Ice Condenser Models (Third Quarter 1985), a complete reanalysis of Mass / Energy Release and containment integrity will be performed using the approved models to demonstrate all applicable safety and regulatory requirements remain satisfied.

Summary s

The evaluations performed to date provide reasonable assurance that the full scope of reanalysis related to deletion of the UHI System will provide adequate operating flexibility ments. while continuing to satisfy all design, safety, and regulatory require-Model improvements either approved or soon to be approved by the NRC which have been developed since the existing McGuire analyses were performed will be introduced in order to modernize the analytical bases for McGuire operating re-strictions.

2-4

REFERENCES 1, . Bordelon, F. M., Massey, H. W. and Zorden, T. A., " Westinghouse ECCS July Evaluation Model - Summary", WCAP-8339, (Non-Proprietary) 1974

2. Schwarz, W. R. " Addendum to BART-A1: A computer code for the Best Estinate Analysis of Reflood Transients", WCAP-9561, Addendum 1, (Proprietary) November 1984

3. Bordelon, F.,M.,

et al., " SATAN-VI Program; Comprehensive Space Time Dependent Analysis of Loss of Coolant", WCAP-8320, (Proprietary)

June 1974

4. Kabadi, J. N.,

et al., " BASH: An Integrated Core and RCS Reflood Code for Analysis of PWR Loss of Coolant Accidents", WCAP-10337, (Proprietary)

July, 1983

5. . Hsieh, T. and Raymond, M., "Long Tern Ice Condenser Containment LOTIC code Supplement 1", WCAP-8355 Supplement 1, May 1975

6. Bordelon, F. M.,

et al., "LOCTA-IV Program: Loss of Coolant Transient Analysis", WCAP-8301, (Proprietary) February 1979, and WCAP-8305 (Non-Proprietary) February 1978

7. Rahe, E. P., Westinghouse Letter.to Thomas, C. 0., NRC, October 27, 1982,

Subject:

" Westinghouse Revised PAD Code Thermal Safety Model".

WCAP-8720 Addendum 2 (Proprietary)

8. Myer, P. E and Kornfilt, J.,

"NOTRUMP: A Nodal Transient Small Break and General Network Code", WCAP-10079, November 1982

9. Lee, N., Tauche, W. D., and Schwarz, W.

R., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10054, December 1982 10.

" Westinghouse Emergency Core Cooling System - Plant Sensitivity Studies" WCAP-8356, July 1974

~

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Attachment 3 Analysis of Significant Hazards Consideration As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed chan8es to the technical specifications involve significant hazards considerations, as defined by 10 CFR 50.92.

The probability of an accident previously evaluated is unaffected by the proposed amendments because the UHI System serves only to mitigate acci-dents after they occur and performs no function with respect to preventing accidents. The consequences of an accident previously evaluated are not increased because all applicable conservative criteria will be satisfied.

The proposed amendments would not create the possibility of a new or differ-ent type of accident, from any accident previously evaluated. NRC has pro-viously accepted partial power operation with the UHI inoperable at McGuire (Amendment No. 37 to Facility Operating License NPF-9 and Amendment 18 to Facility Operating License NPF-17 dated October 31, 1984). Furthermore, NRC has licensed plants of similar design without UHI using computer codes similar to that utilized in support of this submittal. In fact, removal of UHI eliminates any concerns regarding improper operation of UHI.

Finally, the proposed amendments do not involve a significant reduction in a margin of safety. The principal critorion on emergency core cooling system performance for a LOCA is adherence to the peak clad temperature limit of 2200*F. The 2200*F criterion was established to provide a sufficient safety margin between gross failure condition of the fuel cladding and calculated results. Inasmuch as the peak clad temperature will be limited to 12200*F with or without the UHI, there is no significant reduction in the margin of safety.

The Commission has provided guidance concerning the application of standards of no significant hazard determination by providing certain examples (48 FR

' 1487). One of the examples of actions likely to involve no significant hazards considerations relates to a chango which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduco in some way a safety margin, but where the results of the change are cicarly within all acceptable critoria with respect to the system or component specified in the Standard Review Plan. Because the analysis described in the application for the proposed amendments shows that the results of the changes are cicarly within the applicable acceptanco criteria, the examplo described above can be applied to this situation.

In summation, it has been determined that the proposed deletion of the UHI system would nott l

-o o

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3)

Involve a significant reduction in a margin of safety.

Based upon the preceding analysis, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.

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