Text

Proposed Tech Specs Re Heat Up & Cool Down Curves LTOP
	ML20082B396
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/29/1995
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20082B390	List: ML20082B387; ML20082B396;
References
	NUDOCS 9504050138
	Download: ML20082B396 (66)
	v • d • e

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ATTACHMENT 1 -

l DUKE POWER COMPANY MCGUIRE NUCLEAR STATION PROPOSED REVISIONS TO THE TECHNICAL SPECIFICATIONS l

BR"!8s!"Zi888!6, P PDR i

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

3' i

1 JNDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS l

SECTION PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................... 3/4 4-30 FIGURE 3.4-2g UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 19sEFPY......................... 3/4 4-31

/4 FIGURE 3.4-kh3 UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS-APPLICABLE UP T0'10s EFPY......................... 3/4 4-32 14 i FIGURE 3.4-1 44 UNIT 1 REACTOR COOLANT SYSTEM j C00LDOWN LIMITATIONS - APPLICABLE UP .

TO 1GL E P FY . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I&

3/4 4-33 FIGURE 3.4-31(5 UNIT 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS-APPLICABLE UP ,

TO 19s E P F Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-34 :

/W l

TABLE 4.4-5 (DELETE 0]............................. 3/4 4-35 i Pressurizer............................................... 3/4 4-36 ?

Overpressure Protection Systems........................... 3/4 4-37 !

3/4.4.10 STRUCTU RAL I NT EGRITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-39, 4 l t

3/4.4.11 REACTOR VESSEL HEAD VENT SYSTEM ......................... 3/4 4'49.42- i 3/4.5 EMERGENCY CORE COOLING SYSTEMS '

3/4.5.1 ACCUMULATORS I

Cold leg Injection........................................ 3/4 5-1 !

[0eleted]................................................. 3/4 5-3 t 3/4.5.2 ECCS SUBSYSTEMS - Tavg 2 350*F............................. 3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - Tavg s 350*F............................. 3/4 5-9 3/4.5.4 [0eleted].................................................. 3/4 5-11 -

TANK...............................

t 3/4.5.5 REFUELING WATER STORAGE 3/4 5-12 ;

f McGUIRE - UNITS I and 2 X Amendment No.139 (Unit 1)

Amendment No.121 (Unit 2)

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M A T E R I A L S A 318 CU R VE APPLIC AB L E FO R H E AT UP R AT ES UP TO SS'P/NR PO R T H E SE R VICE PE R IOD CO NTR O LLING M ATERI A L-LO NGITUDIN A L UP TO 10 E PPV CO NT AINS M A R G IN S O F C O P P,E R C O N T E N T : 0.21 w tm WELD \ l' RT yg IN ITI A L: -50*P

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- l FIGURE 3.402 M GUIRE UNIT-1 REACTOR COOLANT

SYSTEM HEATOP LIMITATIONS NRC RG 1.99 REV 2 APPLICABLE FOR THE FIRST 10 EFPY j l

McGUIRE -UNITS 1 and 2 3/4 4-31 Amendment No. 115 (Unit 1) .

Amendment No. 97 (Unit 2) l

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.ir MATERIAL PROPERTY BASIS

' LIMITING MATERIALS: LOWER SHELL LONGITUDINAL WELDS 3-442A & C AND

- LOWER SHELL PLATE B5013-2

LIMITING ART AT 16 EFPY: 1/4-t,149.5 F 6 3/4-t, 102.0 F.

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. 250 TEMPERATURE (282 F)FOR THE SERVICE PERtOo UP 1016 EFPY 0 'l'j

O 50 100 150 200 250 3C0 350 400 450 500 Rooctor Boltune Region Fluid Temperature (Deg. F) l McGuire Unit 1 Reactor Coolant System Heatup Limitations (Without Margins for Instrumentation Errors)

NRC REG GUIDE 1.99, Rev. 2 Applicable For The First 16 EFPY Figure 3.4-2 McGuire - Units 1 and 2 3/4 4-31 Amendment No. (Unit 1) ,

Amendment No. (Unit 2) i

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CU R V E S APPLIC A B L E F O R M E A TUP M ATE RIA L B ASIS R A T E S UP To se'P /M R FO R TM E SE R VIC E CONTRO LLING M ATERIAL: L O W E R S M E d.

PE RIOD UP TO to E PPY. CO NT AIN S M A R GIN S COPPER CONTENT: 0.15mt%

O F 10* P A N D 8 0 PS10 P O R PO SSIB L E ny IN IT I A L : -80* P NO INST R UM E NT E R RO R. Ry,,T ApygggggPPY: 1/4T 90*P 3 /4 T . 01

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FIGURE 3.4 3 McGUIRE UNIT 2 REACTOR COOLANT SYSTEM HEATUP LtMITATIONS NRC RG 1.90 REV 2 APPLICABLE FOR THE FIRST 10 EFPY McCUIRE - UNITS 1 and 2 3/4 4-32 Amendment No. 115 (Unit 1)

Amendment No. 97 (Unit 2)

4j. 3.4- 3 Nea r MATERIAL PROPERTY BASIS l

LIMITING MATERIALS: LOWER SHELL FORGING 04 LIMITING ART AT 16 EFPY:' 1/4-1, 104 F 3/4-t, 73 F 2500 ,

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_. HYDROSTATIC TEST TEMPERATURE G36 F) FOR 250 THE SERVICE PERIOo UP TO laEFPV lif II11 till 0 ' III IIII IIII O 50 100 150 200 250 300 350 400 450 500 Reactor Beltline Region Fluid Temperature (Deg. F)

McGuire Unit 2 Reactor Coolant System Heatup Limitations (Without Margins for Instrumentation Errors)

NRC REG GUIDE 1.99, Rev. 2 '

Applicable For The First 16 EFPY Figure 3.4-3 McGuire - Units 1 and 2 3/4 4-32 Amendment No. (Unit 1) [

Amendment No. (Unit 2) .

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IN IT I A L: -59'P INsTRumEur ERROR. AT, MBT,, APTER 10EPry: 1/47,1seg s,P S/47,113 P FIGURE 3.4-4 MCGUIRE UNIT.1, REACTOR COOLANT

!' g . SYSTEM, COOLDOWN LIMITATIONS. -

NRC RG 1.99 REV 2 .

i APPLICA8LE FOR THE'FIRST 10 EFPY McGUIRE - UNITS 1 and 2 3/4 4-33 Amendment No. 115 (Unit 1) l Amendment No. 97 (Unit 2) . .

.n T.$ 3. 4 -_4 Neu) . q MATERIAL PROPERTY BASIS LIMITING MATERlALS: LOWER SHELL LONGITUDINAL WELDS 3-442A & d AND LOWER SHELL PLATE B5013-2 .

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McGuire Unit 1 Reactor Coolant System Cooldown Limitations, Cooldown Rates up to 100 F/HR (Without Margins for Instrumentation Errors)

NRC REG GUIDE 1.99, Rev. 2 Applicable For The First 16 EFPY Figure 3.4-4 McGuire - Units 1 and 2 3/4 4-33 Amendment No. (Unit 1) :

Amendment No. (Unit 2) t

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CunvEs APrucAetE Pon CootooWN nATE: "^7EneALsAsis U, TO Ste eP/NR FOR THE SERVICE PERIOD CO NT A O L LING M AT E R1A L - LOW E R SH E L \ ,

up To to appy ANs CoNTAINs MARGIN: OF

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FIGURE 3.4-5 McGUlRE UNIT 2, REACTOR COk ) ;

SYSTEM, COOLDOWN LIMITATIONS.

NRC RG 1.90 REV 2 APPLICABLE FOR THE FIRST 10 EFPY McGUIRE - UNITS 1 and 2 3/4 4-34 Amendment No.115 (Unit 1)

Amendment No. 97- (Unit 2)

4' '

'b .4 -5 da MATERIAL PROPERTY BASIS LIMITING MATERIALS: LOWER SHELL FORGING 04 LIMITING ART AT 16 EFPY: 1/4-t,1040 F 3/4-t, 73 F.

22 ,__._

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500 - - 20 y/ //

- 40 f , )

100 250 0

0 50 100 150 200 250 300 350 400 450 500 Reactor Bemine Region Fluid Temperature (Dog. F)

McGuire Unit 2 Reactor Coolant System Cooldown Limitations, Cooldown rates up to 100 F/HR (Without Margins for instrurnentation Errors)

NRC REG GUIDE 1.99, Rev. 2 Applicable For The First 16 EFPY Figure 3.4-5 McGuire - Units 1 and 2 3/4 4-34 Arnendment No. (Unit 1)

Amendment No. (Unit 2)

Q "p o.

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION

[4.5.{f leest' erie of the fellewirig Overpressure Protectic Sy:t m; :h:lF uw vrun,m u.

. .Tw p =r Oper:ted relief- valves (ponve) with a lift settina of less th: Or eq::1 te 100 peig, er

b. The Reecter C-octerit SysteiN (RCS) depresiuri :d with : RCS = t Of'

-greeter ther, er ;;;;l to 4.5 0';=r0 i,005 :. '

APPLICABILITY: MODE 4 when the temperature of any~RCS cold leg is less1than '

or equal to 300*F, MODE 5, and MODE 6 when the head is'on the reactor vessel.

ACTION:

a. th one PORV inoperable in MODE 4, restore the inoperable PORV to 0 BLE status within 7 days or complete depressurization and vent qg of the RCS through at least a 4.5 square inch vent (s) within the nek 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one P V ino;,erable in MODE 5, suspend all operations that could ;

lead to wate solid RCS conditions. Restore the inoperable PORV to ,

OPERABLE statu ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or complete depressurization and venting of the R through at least a 4.5 square inch vent (s) within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l c. With one PORY inoperabl in MODE 6, restore the inoperable PORV to OPERABLE status within 24 ours or complete depressurization and venting of the RCS through lest a 4.5 square inch vent (s)'within -

the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

~

d. With both PORVs inoperable, comple depressurization and venting of the RCS through at lest a 4.5 square nch vent (s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

e. In the event either the PORVs oc the RCS ent(s) are used to mitigate ,

an RCS pressure transient, a Special Repor shall be prepared and submitted to the Commission pursuant to Spec ication 6.9.2 within

!- 30 days. The report shall describe the circum ances initiating the l- transient, the effect of the PORVs or vent (s) on e transient, and any corrective action necessary to prevent recurren .

f. The provisions of Specification 3.0.4 are not applicabl l

l

,3 MCGUIRE - UNITS 1 AND 2 3/4 4-37 Amendment No.150 (Unit 1)

Amendment No.132 (Unit 2)

pffacNb h/4 t' L I c1' % )

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.3 As a minimum, a Low Tsuperature Dverpressure Protection (LTOP System shall be OPERABLE as follows:

a. A maximum of one Centrifugal Charging (NV) pump nr one Safety Injection (NI) pump capable of injecting into the Reactor Coolant System (RCS) with all remaining NV and NI pump motor circuit breakers open or the discharge of the remaining NV and NI pumps isolated from the RCS by at least 2 valves with power removed #

and

b. All accumulators isolated.

and

c. One of the following conditions met:

1. Two PORVs with a lift setting of 5 385 psig QI

2. The RCS depressurized with a vent of 2 2.75 square inches.

APPLICABILITY 1 MODE 4 when the temperature of any RCS cold leg is less than er equal to 300*F; MODE 5 and MODE 6 with the reactor vessel head on.

ACTION:

a. With two or more Charging (NV) or Safety Injection (NI) pumps capable of injecting into the RCS*, immediately initiate action to restore a maximum of one NI or one NV pump capable of injecting into the RCS.

Two Charging pumps (NV or NI) maybe capable of injecting into the RCS during pump swap operation for 5 15 minutes.

One Safety Injection pump and one Charging pump, or two Charging pumps maybe operated concurrently provided:

1. RHR suction relief valve (ND-3) is OPERABLE, and the RHR suction isolation valves (ND-1 and ND_2) are open and one of the following conditions is met:

a. RCS cold leg temperature is greater than 167 F, or b RCS cold leg temperature is greater than 107 F and cooldown rate is less than 20 F per hour.

QB

2. Two PORVs secured in the open position with their associated block valves open and power removed.

McGuire - Units 1 and 2 3/4 4-37 Amendment No. (Unit 1)

Amendment No. (Unit 2)

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& 40 ;

EACTOR COOLANT SYSTEM

\

SURhEILLANCEREQUIREMENTS

(- ;

4.4.9. 1 Each PORV shall be demonstrated OPERABLE by: ,

a. erformance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV a tuation channel, but excluding valve operation, at least once per 31- ays;

b. Perfo ance of a CHANNEL CALIBRATION on the PORV actuation channel at '

least ce per 18 months; and .

c. Verifying he PORV isolation valve is open' at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l' when the P V is being used for overpressure protection.

4.4.9.3.2 The RCS vent ) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s is being used for overpressure protection.

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Except when the vent pathway is provided with a valve which is locked, sea ed, ,

or otherwise secured in the open position, then verify these valves open at (- '

least once per 31 days.

MCGUIRE - UNITS 1 AND 2 3/4 4-38 Amendment No.150 (Unit 1)

Amendment No.132 (Unit 2) s

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A $3 C ec A

(. u / 3.) l a t Reactor Coolant System LIMITING CONDITION FOR OPERATION ACTION: (continued)

b. With an accumulator not isolated, isolate the affected accumulator within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If required action is not met, either:

1. Depressurize the accumulator to less than the maximum RCS pressure for the existing cold leg per Specification 3/4.4.9 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , OR

2. Increase RCS cold leg temperature to greater than or equal to 300 F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days. If required action is not met, depressurize the RCS and vent through at least a 2.75 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. With one PORV inoperable in MODES 5 or 6, suspend all operations which could lead to a water-solid pressurizer.

Restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If required action is not met, either:

1. Ensure RCS temperature is greater than 167"F, and ND-3 is OPERABLE, and ND-1 and ND-2 are open within one hour.

QR

2. Depressurize the RCS and vent through at least a 2.75 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

McGuire - Units 1 and 2 3/4 4-38 Amendment No. (Unit 1)

Amendment No. (Unit 2)

J' i b 3C.N 4 c$ Dy g Y ( 3, y C Reactor Coolant System LIMITING CONDITION FOR OPERATION l ACTION (Continued)

e. With the LTOP system inoperable for any reason other than a., b., c., or d. above, depressurize the RCS and vent through at least a 2.75 square inch vent within 8 hours.

f. In the event that either the PORVs or the RCS vent are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Cossaission pursuant to the specification 6.9.2 within 30 days. The report shall describe the circumstance initiating the transient, the effect of the PORVs or vent on the transient, and any corrective action necessary,to prevent recurrence.

g. The provisions of specification 3.0.4 are not applicable.

McGuire Units 1 and 2 3/4 4-39 Amendment No. (Unit 1) Amendment No. (Unit 2)

Nad Of c 40 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days;

b. Performance of a CHANNEL CALIBRATION on PORV actuation channel at least once per 18 months; and

c. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.

4.4.9.3.2 The RCO vent (s) shall be verified te be open at least once per 12 hours' when the vent (s; is bein3 ased for everpressure protection Once every'12[ hours *, vehify that an"RCS[ vent o'f N 2.'75 square inches ' is open?

when the vent.is used for' overpressure protection. 4.4.9.3.3 ;Once every 121hourss: verify that each accumulator.is isolated and thatJonly one NV or NI. pump is capable.of-injecting into the RCS. 1 4.4.9.3.'4. Once everyJ12 hours, verify that RHR suction ~ isolation valves ND and ND-2'are open when RHR suction relief valve ND-3 is being used for overpressure protection. 4.4.9.3.5 Once'every 72. hours, verify that the PORV block valve-is'open for each required PORV.

Except when the vent pathway is provided with a valve which is locked, sealed or otherwise secured in the open position, then verify these valves open.once per.31 days.

A PORV secured in'the'open position maybe,u... sed to meet'this vent requirement provided that its associated: block valve is open and power

is removed.

~ '

McGuire Units 1 and 2 3/4 4-36 Amendment No. (Unit 1) 4o Amendment No. (Unit 2) i i 1

4. .

                                                                                                             .l
   ., -s REACTOR COOLANT SYSTEM 3/4.4.1b STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION i

3.4.10 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10. , APPLICA8ILITY: All MODES. ACTION:

a. With the structural integrity of any ASME Code class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimus' temperature required by NDT considerations. .

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected' component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant Sys. tem temperature above 200*F. ,

c. With the structural integrity of any ASME Code Class 3 component (s) -

not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or ; isolate the affected component (s) from service. i SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory , Position C.4.b of Regulatory Guide 1.14. Revision 1, August 1975. , i F 8 i { . )

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                                                                                          - ~ .                 ;

i l McGUIRE - UNITS 1 and 2 3/4 4-34 L/ [ Amendment No. , (Unit 1) {

                                                             .ggf-         Amendment No,   e (Unit 2)        l l

t

N '; .. REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR VESSEL HEAD VENT SYSTEM t LIMITING CONDITION FOR OPERATION' t 3.4.11 Two reactor vessel head vent paths,'each consisting of two valves in , series powered from emergency buses shall be OPERABLE and closed. ; APPLICA8ILITY: MODES 1, 2, 3 and 4 ACTION:

a. With one of the above reactor vessel head paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vdnt path to OPERA 8LE status within 30 days or be in HOT STAN08Y within 6 hours and in COLD SHUTDOWN within the following ,

30 hours. .

b. With both of the above. reactor vessel head vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least two of the vent paths to OPERA 8LE status within 72 hours or be-in HOT STAN08Y within 6 hours and in COLD SHUTDOWN

                   .within the following 30 hours.

SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:

1. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING. ,

2. Verifying flow through the reactor vessel head vent paths during -

venting during COLD SHUTDOWN or REFUELING. McGUIRE - UNITS 1 and 2 3/4 4-)& Amendment No. (Unit.1) qt Amendment No. (Unit 2)

                  . v  ,..--      - - , - . -    - - . . - - . . ,                    ,-,-, . -- -             - ---

r 3/4.5.3 ECCS SUBSYSTEMS - Tava F 350*F I LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump,# !

1

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, and

d. An OPERABLE flow path capable of taking suction from the refueling ;

water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of : operation. 1 APPLICABILITY: MODE 4. ACTION: l

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.

b. With no ECCS subsystem OPERABLE because of the inoperability of either the RHR heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System, pA

T,y less than 350*F by use of alternate heat removal methods. a.

c. IntheeventtheECCSisactuatedandinjectswaterintotheReactok '

Coolant System, a Special Report shall be prepared and submitted to p the Commission pursuant to Specification 6.9.2 within 90 days , describing the circumstances of the actuation and the total accumu , lated actuation cycles to date. The current value of the usage fac- 4,

. {

tor for each affected Safety Injection nozzle shall be provided in 0 this Special Report whenever its value exceeds 0.70. <^ ; s 'G ta bs l U -s p l l i ' Cap a le, o 9 i w y e c. W3 into the lCS } }te

                                                                                   .9 8       /
                                                                                   + GL.       ,
     #A maximum of one centrifugal charging pump and one Sefety Injection pump 4                t shall be ^r:rXT whenever the temperature of one or more of the RCS cold 1 f.

1 cgs is less than or equal to 300*F. Two charging pumps may be operable C and operating for s15 minutes to allow swapping charging pumps. " Q* j l l McGUIRE - UNIT 1 and 2 3/4 5-9 Amendmen,iio. 152 (Unit 1) Amendment No. 134 (Unit 2)

l l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the aooli cable , requirements of Specification 4.5.2. /jife P Cg % sis #F 4 M D QNft> The. AC1/ 4.5.3.2 All centrifugal

                     ^^'"""'"

charging pumps and Safety Injection pumpsi nn; t '. -

                                    - shall be demonstrated inoperable by verifying that the motor circuit breakers are secured in the open position or by verifying the discharge of each pump has been isolated from the RCS by at least two isolation valves with power removed from the valve operators at least once per 12 hours whenever the temperature of one or more of the RCS cold legs is less than or equal to 300*F.

l l I McGUIRE - UNITS 1 and 2 3/4 5-10

2 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) I The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with. additional reactor vessel requirements. These gqs3 el;D properties are then evaluated in accordance with Appendix G of then970 30...~. d Mg:,d % Addunde to Sectica I!! of the ASME Boiler and Pressure Vessel . Code and the cal- l culation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975." ' l Heatup and cooldown limit curves are. calculated using the most limiting value of the nil-ductility reference temperature, RT , at the end of the effective full power years (EFPY) of ervice life idNItified on the appli_ cable Q;%, technical specification figure. The EFPY service life periodIWchesca such(To e#wc that the limiting RT

  ": S m 2
                        -'s
                                 'No- I r at25the  1/4T E I -- ' .. t '.location  in the The selection       core of such   a region is 7" "  - ABowndI limiting RT   xor assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.                         qq

c RT The reactor vessel materials have been tested to determine their initial uor; the results of these tests are shown in Table B 3/4.4-1. Reactor opera-tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT xor. Therefore, an adjusted reference temperature, ba' sed upon .

the fluence, copper content, and phosphate content of the mate' r ial in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART For Unit 1,theadjustedreferencetermperaturehasbeencomputedbyRegUda. Guide 1.99, Revision 2. Fcr Unit 2, the cdjusted-refereacc tee +ar&ture has tory beca computed cs discussed in WCAP-110E9. The heatup and cooldown limit curves of Figures 3.4-2, 3.4-3, 3.4-4 and 3.4-5 include predicted adjustments for this shift in RTuor at the end of tihe identified' service life. Adjustments for possible errors in the pressure and temperature sensing instruments are included when stated on the applicable figure. Values of ART wor determined in' this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR 50, Appendix H. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the pressure vessel material by using the lead factor and the with-drawal time of the capsule. The heatup and cooldown curves must be recalcu-lated when the ART *or determined from the surveillance capsule exceeds the calculated ARTNor for the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup.and cool-down rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to l 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. Amendment No. (Unit 1) McGUIRE - UNITS I and 2 B 3/4 4-8 Amendment No. (Unit 2)

                   ~
                       . REACTOR COOLANT SYSTEM i

BASES PRESSURE / TEMPERATURE LIMITS (Continued) end of the transient, conditions may exist such that the effects of compressive. thermal stresses and.different K IR 's for steady-state and finite heatup rates . do not offset each other and the pressure-temperature curve based on steady-F state conditions no longer represents a lower bound of all similar curves for , finite heatup. rates when the 1/4T flaw is considered. - Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of 'i pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, '

the thermal gradients established at the outside surface during heatup produce - stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both I the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the themal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. : 1 Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are

produced as follows. A composite curve is constructed based on a point-by- l point comparison of the steady-state and finite heatup rate data. At any '

given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the ; course of the heatup ramp the controlling condition switches from the inside ; to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. -l i Finally, the composite curves in technical specifications for the heatup j rate data and the cooldown rate data may be adjusted for possible errors in the j pressure and temperature sensin instruments by the values indicated on the ; respective curves. Where techn cal specification curves.have not been adjusted.. I such adjustments are made by plant procedures. i. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis ) performed in accordance with the ASME Code requirements. g l The OPERA 8ILITY of two PORVs or an RCS vent opening of at least%4 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of l Amendment No-~ (Unit 2) McGUIRE ~ UNITS 1 and 2 B 3/4 4-16 Amendment No. (Unit 1)

   . . , , , , - ,                 n  _. - .     -       .       . - . . . .   . . - _ - . _ _ . ,      -

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

REACTOR COOLANT SYSTEM 1- BASES PRESSURE /TEMPERATURELIMITS(Conthnued) 1ke CS Vt DQ the RCS cold legs are less than or equal to 300 F. EItherGedWhasadequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the i RCS cold leg temperatures, or (2) the start of a HPSI pump and-its injection into a water-solid RCS. ( IggerT A Track ed pay e C 3/4.4.10 STRUCTURAL INTEGRITY j The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i). Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the.ASME ., Boiler and Pressure Vessel Code, 1971 Edition and Addenda through Winter 1972. ; 3/4.4.11 REACTOR VESSEL HEAD VENT SYSTEM ( , Reactor Vessel Head Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation i core cooling. The OPERABILITY of at least one reactor coolant system vent i path from the reactor vessel head and the pressurizer steam space ensures the i capability exists to perform this function. (Operability of the pressurizer i steam space vent path is provided by Specifications 3/4.4.4 and.3/4.4.9.3. ) l I The valve redundancy of the reactor coolant system vent paths serves to j minimize the probability of inadvertent or irreversible actuation while ensur- , ing that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path. The surveillance to verify Reactor Vessel Head Vent flowpath is qualita-tive as no specific size or flow rate is required to exhaust noncondensible gases. The function, capabilities, and testing requirements of the reactor coolant system vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements", November 1980. t McGUIRE - UNITS 1 and 2 B 3/4 4-17 Amendment No. (Unit 1) Amendment No. '(Unit.2)

2 Pa p c. The pressurizer PORV setpoints for low temperature overpressure protection are based on limiting the peak pressure during the limiting transient to 1.10 times the ASME Section XI, Appendix G limits, in accordance with ASME code case N-514. Credit is taken for the RHR suction relief valve (ND-3) during conditions where relieving capacity at rated accumulation is sufficient to prevent-

exceeding the above allowable pressure limits.

Cooldown limits / minimum RCS temperatures restrictions ensure the allowable pressure limits will not be exceeded. e 5 W

ATTACHMENT 2 l DUKE POWER COMPANY l MCGUIRE NUCLEAR STATION 4 TECHNICAL JUSTIFICATION l - - - - - _ - - - _ - - - - _ - - - - - - - - - - - - - - - _ - - - _ - - - - - - - - - - - - - - - - - - - - - _ - - - - - _ _ - - _ _ - _ - _ - _ _ _ _ _ a

. .. - _. . - .- - ~ . .- . - - - -- INTRODUCTION The proposed changes amend the McGuire Nuclear Station' Units 1 , and 2 Technical Specifications.3.4.9.3 " Overpressure Protection Systems", 3.5.3 "ECCS Subsystems - Tavg < 350 F" and the i Pressure / Temperature limit curves, Figures 3.4-2, 3.4-3, 3.4-4 and 3.4-5. These changes revise the Low Temperature overpressure Protection (LTOP) system maximum setpoint, minimum vent

 . requirements, and ECCS subsystem availability requirements to enhance system operation and reliability. These changes also revise the Reactor Coolant System Pressure / Temperature limit curves to include the most recent radiation surveillance capsule analysis. The format has been revised to conform to the extent possible with the Westinghouse Standard Technical Specifications (NUREG 1431).

The proposed technical specification changes provided by this amendment request can be grouped as follows: l

1) Revisions to the Limiting Condition for Operation (LCO) ,

requirements, the Action Statements (AS) and the Surveillance Requirements (SR) for the Reactor Coolant System (RCS) Overpressure Protection System during low i temperature conditions. ;

2) Reduction in the Reactor Coolant System (RCS) vent requirement from 4.5 square inches to 2.75 square inches.

3) The use of the Residual Heat Removal (RHR) suction relief valve (IND3 and 2ND3) for overpressure protection under restricted conditions. (RCS > 107 F and cooldown rate less than 20 F/hr; or RCS > 167 F) .

4) Revisions of the Pressure / Temperature curves to 16 EFPY, including the incorporation of the latest radiation !

surveillance capsule results and removal of instrumentation l margins from the technical specification figures. ;

5) Changes to format and consistency.

i i f i l 1 I 1

g...... . . . . . . . - - - - - - - - - - DISCUSSION OF PROPOSED CHANGES The following is a brief discussion of the significant changes associated with each of the five (5) identified groups: t L 1) Revisions to the Limiting Condition for Operation (LCO) l requirements, the Action Statements (AS) and the Surveillance Requirements (SR) for the Reactor Coolant System (RCS) Overpressure Protection System during low temperature conditions. . This change reduces the maximum allowable setpoint of the ' PORV during low temperature conditions (RCS cold leg temperature s 300 F) . The analysis performed for establishing the new setpoint is revised to incorporate ASME I Code Case N-514. As such, the analysis performed will l ensure that the resultant peak pressure for a LTOP event is limited to 110% of the pressure determined to satisfy Appendix G of Section XI. NRC authorization for the implementation of ASME Code Case N-514 at McGuire Nuclear I Station was provided by letter dated September 30, 1994 ! In addition, the proposed changes associated with this group 1

will define additional conditions for establishing the operability of the LTOP system, what actions to take when the additional conditions are not met and revisions to the j surveillance requirements associated with the LTOP system. For the LTOP system to be considered operable, all of the accumulators need to be isolated and only a maximum of one 1 Centrifugal Charging (NV) pump or one Safety injection (NI) pump capable of injecting into the RCS, except during pump swap operations. These additional restrictions assure that operation and configuration of the units during low temperature conditions are consistent with what was assumed in the analysis of LTOP events. I 2) A reduction in the Reactor Coolant System (RCS) vent requirement from 4.5 square inches to 2.75 square inches. The current minimum size of the vent opening specified by the Technical Specifications for McGuire is 4.5 square inches. This vent path is used for situations where the PORV's may not be available. The proposed change will reduce the specified size opening from 4.5 square inches to 2.75 square inches. l I 2

The technical specification that was approved when the Facility Operating License for McGuire Nuclear Station was initially issued, specified a vent size opening of 4.5 square inches. This size requirement of 4.5 square inches appears to be very conservative in regard to the licensing bases for establishing an RCS vent during low temperature conditions. An analysis has been performed which demonstrates that the 2.75 square inch vent size provides more than adequate overpressure protection for LTOP events. l' i

3) The use of the Residual Heat Removal (RHR) suction relief valve (1ND3 and 2ND3) for overpressure protection under restricted conditions. -(RCS > 107 F and cooldown rate less I than 20 F/hr; or RCS > 167 F) !

l J The change associated with this group is to allow for the restricted use of the RHR suction relief valve for overpressure protection. The use of the RHR suction relief I valve under the restricted conditions will provide an i equivalent means of overpressure protection when compared to j a PORV. The specific situations to be allowed by the proposed amendment in which the RHR suction relief valve will be utilized for overpressure protection are; 1) to 1 allow for a second ECCS system pump to be able to inject into the RCS; and 2) when one PORV is inoperable while in Modes 5 or 6. l The RHR system suction isolation valve autoclosure i interlocks (which would isolate the RHR suction relief valve ! upon rising pressure) have been deleted at McGuire Nuclear Station. NRC approval of the technical specification amendment which, in effect, authorized the removal of the RHR autoclosure interlock circuitry was provided by a letter i dated September 11, 1990. The RHR suction relief valve is l normally in service during LTOP conditions and the proposed amendment will ensure proper system alignment, and that the relieving capacity available for LTOP events is sufficient to protect the pressure /t.emperature limits.

4) Revisions of the Pressure / Temperature curves to 16 EFPY, including the incorporation of the latest radiation surveillance capsule results and removal of instrumentation margins from the Technical Specification figures.

l 3

I i l This change updates the heatup and cooldown curves for both ! units. The service period for the proposed heatup and cooldown curves have been extended from 10 EFPY to 16 EFPY. The new pressure / temperature limits satisfy all required material embrittlement considerations including: 10 CFR 50, Appendix G; Regulatory Guide 1.99, Revision 2; and ASME Section XI, Appendix G. The development of these curves was performed by Westinghouse Electric Corporation and was included in the appendices of the Westinghouse surveillance capsule reports WCAP-13949 and WCAP-13516. These surveillance capsule reports were submitted to the NRC for review by letters dated March 24, 1994 and January 27, 1992. This change, also, includes the relocation of the instrument error margins from the curves to other licensee controlled documents. In addition, the pressure instrument error margin will be reduced from 60 psig to 30 psig. This reduction in the pressure instrument error margin is made possible by a modification to replace the wide range reactor coolant system pressure transmitters with a narrow range pressure transmitters. The purpose of this modification is to substantially reduce the instrument uncertainty associated with the LTOP function and its related effects on the Appendix G heatup and cooldown curves. The narrow range transmitters being added by this modification will only be used during LTOP conditions. During power operation, these transmitters will be valved out. The instrumentation margins have been removed from the curves. The LTOP system will be using narrow range, high accuracy pressure transmitters for overpressure protection. The normal wide range pressure transmitters will still be used for power operation and other non-LTOP conditions. The , instrument margins will be administratively implemented by ! incorporating them into the controlling procedures for unit operations and into the LTOP system setpoint selection I calculation.

5) Changes to format and consistency:

Other changes have been made to improve consistency between specifications, incorporate Westinghouse Standard Technical Specification format, and update applicable code references. These changes are considered editorial in nature and do not affect the operation of the units or the safety functions performed by the LTOP system. 4

Specifically, the administrative and editorial changes provided by this amendment request are; 1) an updating of Page X of the technice.1 specification index; 2) a re-order or re-structuring of some of the current action statements for Specification 3.4.9.3 (LTOP); 3) a re-wording of some of the current surveillance requirements'for the LTOP system (Specification-4.4.9.3.2); 4) an updating of a footnote, providing more information on the intent of the footnote; 5) a re-numbering of some of the technical specification Pages;

6) a re-wording of a footnote and a Surveillance Requirement associated with Specification 3.5.3 (ECCS' Subsystems - T,,,s 350 F); 7) Added statement to footnote for Specification 3.5.3 to refer to Specification 3.4.9.3 for additional requirements; and 8) an updating of the technical specification basis for Specifications 3/4.4.4 (Relief Valves) and 3/4.4.9 (Pressure / Temperature Limits), providing I more information and detail.

1

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REVISION OF LTOP SETPOINT The technical specifications currently state that the PORV lift setting be less than or equal to 400 psig. The proposed technical specification amendment will lower the PORV lift setpoint to less than or equal to 385 psig. The proposed change specifies a more conservative setpoint, in that the PORV will open earlier during an LTOP event. This will lower the peak pressure resulting from an overpressure event. The licensing basis for overpressure protection during low temperature, water solid modes of operation, is to ensure that the RCS will be protected from pressure transients which could exceed the limits established per Appendix G of 10 CFR 50. This is accomplished by limiting the peak pressure as a result of an LTOP event below the pressure / temperature limits of the heatup and cooldown curves for the unit. Furthermore, in response to an NRC question during the licensing phase for McGuire Nuclear Station (Question 212.70) regarding inadvertent overpressurization when the primary system is water solid, the following response was provided; "The pressurizer power operated relief valves will be utilized to provide protection against overpressurization during water solid conditions. The adequacy of these valves for this application is verified in the Westinghouse Report, " Pressure Mitigating Systems Transient Analysis Results", dated July 1977 and the supplement to that report dated September, 1977. Determination of relief valve setpoint was accomplished using procedures outlined in these documents....." One of the proposed technical specification changes provided by this submittal is the updating of the heatup and cooldown curves for both units. As such, an analysis was performed to calculate a new setpoint for the PORV which would be acceptable in preventing a violation of the proposed heatup and cooldown curves for both units. The analysis, that was performed, evaluated three possible transients. The methodology utilized in determining the initial PORV setpoint, as outlined in the 1977 Westinghouse Reports, is followed in the present analysis. The three possible pressure transients that were evaluated in the present analysis are:

1) a mass input from an operable safety injection pump.

2) a mass input from an operable centrifugal charging pump.

3) a heat input from a 50 F temperature difference between the steam generators and the rest of the RCS.

6

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

l

-f

For-all three pressure transient cases,'the following criteria and assumptions are applicable: ,-j

1) ~ The resulting PORV setpoint must'be low enough to- !

mitigate the consequences of-the' defined mass and heat : input transients without violating the Appendix G limits

                     -(considering- ASME Code Case N-514) .: The PORV setpoint
                                                    ~

for LTOP acceptance must be set such that the peak reactor vessel beltline pressure,_ including instrument

                                                      ~
                    ' error, is no more than 1.1 times the ASME Section III Appendix G limits.                                                              '

2) The pressurizer is-water solid and the pressure drop i across the vessel is the same at hot or cold temperatures.

3) The assumed maximum instrument loop uncertainties for.

the RCS Narrow Range Pressure Transmitters is 30 psi. ,

4) The RCS pressure overshoot is calculated using the. !

methodology given by the Westinghouse Report, (" Pressure- ! Mitigating Systems. Transient Analysis Results", dated July 1977 and supplemented September, 1977). ! i

5) The difference between the indicated pressure (signal i actuating the PORV's) and the actual reactor vessel beltline pressure is calculated. This includes elevation differences between the reactor vessel i beltline and the reactor coolant pressure transmitters,.

and the differential pressure across the reactor core due to hydraulic losses (the transmitters actuating the ' PORV's are on the hot legs, while the RV beltline is on the cold leg side of the core). The initial analysis performed during the licensing phase did not include l these correction factors.

6) The ASME Section III, Appendix G heatup and cooldown !

limits for both units were calculated by Westinghouse. The information in support of generating these curves is l provided within Westinghouse surveillance capsule i reports WCAP-13949 for Unit 1 and WCAP-13516 for Unit 2. These Westinghouse surveillance capsule reports were i provided to the NRC for review by letters dated March ! 24, 1994 for Unit 1 and January 27, 1992 for Unit 2. i f ASME Code Case N-514 is utilized to establish the lift setpoint ! of the PORV for overpressure protection during low temperature -j conditions. As delineated in the code case, the LTOP system ' shall limit the maximum pressure in the vessel to 110% of the . ; pressure determined to satisfy Appendix G of Section XI. The'NRC !

           .was requested to approve the use of Code Case N-514 for McGuire                          !

Nuclear Station by a letter dated June 28, 1994. Approval for ! the use of the code case was granted on September 30, 1994. ! i 7 I e

                                        -    - -        - -     - - . - . = -        .-

In conclusion, the analysis performed verified that a PORV setpoint of 385 psig is adequate to ensure that the peak reactor vessel beltline pressure, (including instrument uncertainties, pressure corrections for the difference between the indicated pressure and the actual reactor vessel beltline pressure, and pressure corrections for the differential pressure across the reactor core), is less than 1.1 times the ASME Section III Appendix G limits during anticipated pressure transients, provided appropriate limits on the heatup and cooldown rates are established. The limiting pressure transient is the mass input, resulting from the inadvertent start of a safety injection pump. The resulting worst case peak pressure from this transient, given a PORV setpoint of 385 psig, is approximately 537 psig. Based on a worst case peak pressure of 537 psig, the following limits on the heatup and cooldown rates will be established within appropriate, controlled procedures for both units to ensure that the pressure / temperature limits will not be exceeded: l NUMBER OF REACTOR REACTOR COOLANT REACTOR COOLANT COOLANT PUMPS SYSTEM TEMPERATURE SYSTEM MAXIMUM OPERATING (NO MARGIN) HEATUP RATE ( F) ( F/ HOUR) 0 TO 4 > 75 < 60 NUMBER OF REACTOR REACTOR COOLANT REACTOR COOLANT COOLANT PUMPS SYSTEM TEMPERATURE SYSTEM MAXIMUM OPERATING (NO MARGIN) COOLDOWN RATE ( F) ( F/ HOUR)

                                                                   > 140                       100 0                              140 TO 105                           60 105 TO 75                           40
                                                                   > 140                       100 2                             140 TO 110                            60 110 TO 85                           40     I 85 TO 75                            20
                                                                   > 150                       100 4                              150 TO 140                           60 140 TO 110                           40 110 TO 77                           20 l

l l 8 ( \ .

l VERIFICATION OF LTOP ENABLE TEMPERATURE The current technical specification requires that overpressure protection system be enabled at Mode 4 when the temperature of any RCS cold leg is less than or equal to 300 F. Based on a review of licensing documents and correspondences, the basis for 300 F as the enable temperature for LTOP is to ensure that the Appendix G limits for McGuire would be adequately protected by ! either the low PORV setpoint or by the pressurizer safety relief valves (2485 psig). As such, the enable temperature would be 335 F. An alternative means of determining the enable temperature for the LTOP system has been defined by the NRC within Revision 1 to Section 5.2.2, " overpressure protection", of the Standard Review Plan (NUREG-0800). In particular, within paragraph B.2 of Branch Technical Position RSB 5-2, Revision 1 - November 1988, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures". In accordance with Branch Technical Position RSB 5-2, revision 1, paragraph B.2, the LTOP System is required to be operable at a water temperature corresponding to a metal temperature of RT ,+ 90'F at the beltline location that is controlling in the Appendix G limit calculations. For McGuire, the unit with the limiting material for establishing the enable temperature, based on RT,7, is Unit 1 as verified in the latest Westinghouse surveillance capsule reports for both units (WCAP-13949 and WCAP-13516). In reviewing WCAP 13949, the limiting RTer for Unit 1 is 14 9. 45 F (Table B-4). Thus, the metal temperature would be 239.45 F ( 14 9 . 4 5 F + 90 F = 239.45 F) . l During normal operations, as well as LTOP events, the I corresponding water temperature would not be significantly different than the temperature of the reactor vessel metal. Currently, the technical specifications requires that the LTOP l system be enabled at a temperature of 300 F. This provides a 60 F l margin to account for the difference between the metal i temperature of the reactor vessel and the corresponding I temperature of the RCS during LTOP events. The LTOP enable temperature of 300 F is, thus, conservative relative to the methodology defined within paragraph B.2 of Branch Technical ; Position RSB 5-2, Revision 1. Further, the methodology for determining the LTOP enable temperature specified within paragraph B.2 of Branch Technical Position RSB 5-2, Revision 1 will be utilized for future revisions that may occur to the pressure / temperature limits for ; McGuire. Although the enable temperature for the LTOP system j could be reduced, a technical specification change to that effect will not be proposed at this time. 9 l l

REVISIONS TO SPECIFICATION FOR LTOP The proposed revision to the technical specification for the overpressure protection system will define additional conditions, actions, and surveillance requirements that are presently not specified. Revisions to LCO Reauirements for Soecification 3.4.9.3 Two additional conditions for establishing the operability of the LTOP system are specified. For the LTOP system to be considered operable, all of the accumulators need to be isolated and only a maximum of one Centrifugal Charging (NV) pump nr one Safety injection (;N I) pump capable of injecting into the RCS. These ' actions (isolation of accumulators and only one pump available) will limit mass input into the RCS. These additional restrictions assure that operation and configuration of the units during low temperature conditions are consistent with what was assumed in the analysis of LTOP events. The isolation of all of the accumulators and only having one pump capable of injecting into the RCS during low temperature conditions has always been the practice at McGuire and is controlled by appropriate operating procedures. The inclusion of these restriction into the technical specifications is consistent with NUREC-1431, (Standard Technical Specifications Westinghouse Plants). As stated within NUREG-1431:

     "An LTOP System shall be OPERAELE with a maximum of [one]

[high pressure injection (HPI)] pump [and one charging pump] capable of injecting into the RCS and the accumulators isolated ......" Thus, the requirement to isolate the accumulators and to have only one pump available/ operable, as proposed by this amendment, is analogous with what is specified within NUREG-1431. A footnote has been added to Specification 3.4.9.3. The footnote allows for two charging pumps to be capable of injecting into the RCS for a limited time period (s 15 minutes). The purpose of the footnote is to allow for pump swap operations. The inclusion of this footnote into the Technical Specifications is consistent with NUREG-1431. Further, by letter dated October 4, 1994, a propose technical specification amendment request was submitted for McGuire regarding the operation of two charging pumps while in modes 4, 5, or 6. The proposed technical specification change would allow, for a short period of time (s 15 minutes), the operation of two charging pumps in modes 4, 5, or 6 in order to swap the pumps. NRC approval of the proposed technical specification change was provided by an NRC letter dated November 17, 1994. I The proposed footnote to Specification 3.4.9.3 is consistent with l the technical specification change that was approved by the NRC. i 10 I 1

Revisions to Action Statements for Snecification 3.4.9.3 The proposed amendment revises the Action Statements for Specification 3.4.9.3. The current technical specification has six Action Statements for Specification 3.4.9.3 (a through f). The proposed technical specification amendment will result in having seven Action Statements (a through g). Two of the Action Statements are just re-numbered, no changes to the wording of the Action Statements are proposed. Action Statements e and f'of Specification 3.4.9.3 are re-numbered. The proposeo new Action Statements designations are f and g. Two new Action Statements have been added (proposed Action Statements a and b) to define the action to be taken when more than one charging pump (NV or NI) is capable of injecting into the RCS, or if an Accumulator is not isolated. The remaining four Action Statements of the current technical specification (b through f) have been modified and re-numbered. The following discusses the proposed changes to the Action Statement of Specification 3.4.9.3 provided by this submittal: ProDOROd ACt$OD St&teMODt {&) A new Action Statement (a) is added which specifies required actions in the event two or more NV or NI pumps are capable of injecting into the RCS, i.e. the Action Statement requires that immediate action be taken to ensure that only one NV or NI pump is capable of injecting into the RCS. The proposed action to ensure only one NV or NI pump is capable of injecting into the RCS is consistent with the guidance provided by NUREG-1431. In addition, a footnote establishing the conditions by which a NI pump and NV pump or two NV pumps can be operated is also provided. In the event that a NI and NV pump or two NV pumps are operated, the footnote requires that either the RHR suction relief valve (ND3), as well as two PORVs, be operable or two PORVs secured in the open position with their associated block valves open with power removed from the block valves. A discussion regarding the utilization of the RHR suction relief valve is provided later within this document. With two PORVs and their associated block valves secured in the open position, there would be no credible single active failure that would result in loss of one of these flow paths. The relief capacity of this configuration is adequate to mitigate the consequences resulting from the i inadver. tent injection of both pumps into the RCS. Periodically, while in mode 5, an NV and an NI pump would be i operated in order to perform testing of the NI pumps. The ' preferred configuration of the plant to perform this test in a safe manner is to secure two PORVs and their associated block valves in the open position. 11 l

PrODOmed Act1On Statement (b) A new Action Statement (b) is added which specifies required actions in the event an accumulator is not isolated, i.e. requires that the affected accumulator be. isolated within one hour. For the situation when an Accumulator can not be isolated within one hour, two options are provided, either of which must be performed'within the next 12 hours. The first option is to depressurize the accumulator to less than the maximum RCS pressure for the. existing cold leg temperature. The second option.is to increase the RCS cold leg temperature to greater than or equal to 300 F. This new Action Statement is consistent with the guidance provided by the NRC within NUREG-1431. PrODOmed ACtlOn Statement (C) For the situation where a PORV is inoperable while in mode 4, the current technical specifications require that the inoperable PORV be returned to an operable status within seven days or the RCS be depressurized and vented within the next eight hours. The proposed Action Statement (c) provided by this amendment request specifies the same actions be taken for an inoperable PORV,.the only difference being the size:of the vent opening. A discussion regarding

the size of the RCS vent opening is provided later within this document. PrCDOmed ACtlOn Statement (d) The current technical specification Action Statements b and c specify what to do in the event a PORV is inoperable while in mode 5 or in mode 6, respectively. This amendment proposes to' combine these two Action Statements into a single Action Statement, applicable for both modes 5 and 6. The proposed Action Statement (d) addresses the situation of an inoperable PORV while in Mode 5 or in Mode 6 with the reactor vessel head on. The proposed Action Statement requires that all activities that could lead to a water-solid pressurizer be terminated and to restore the inoperable PORV to an operable status within 24 hours. In the event that the inoperable PORV cannot be returned to service within 24 hours, the Action Statement specifies two equally effective alternatives. One alternative action would be to utilize the RHR suction relief valve, provided that the RCS temperature is greater than 167 F. The other alternative would be to depressurize and vent the RCS within eight hours. As discussed in Generic Letter 90-06, the consequences of operational events that could overpressurize the RCS are , more severe at lower temperatures. As such, a completion time of 24 hours to restore an inoperable PORV to an i operable status represents a reasonable time to repair the l PORV without exposure to a lengthy period with only one I operable PORV to protect against an overpressure event. l l 12 l

In addition, completion times specified in the current Action Statement b and c are 72 hours and 24 hours, respectively. The more conservative time period of the two was chosen. Accordingly, while in mode 6, there is no change in the time required to complete the specified actions, the specification still requires 24 hours. While in mode 5, the time required to complete the action is effectively reduced from 72 hours to 24 hours. As such, the length of time in which only one PORV is operable to protect against an overpressure event is shorter, thus enhancing safety of the unit. One alternative if the PORV can not be repaired within 24 hours is to depressurize and vent the RCS within eight hours. A completion time of eight hours is sufficient to allow for an orderly and controlled depressurization, while minimizing the amount of time in which only one PORV is available for protection of overpressurization events. Analysis performed indicate that one PORV is adequate in mitigating the consequences of an overpressure event. Within the time allowed by the proposed Action Statement, it is highly unlikely that an overpressure event, coincident with a failure of the operable PORV will occur. The size opening of the vent is revised from 4.5 square inches to 2.75 square inches. A discussion regarding the size of the RCS vent opening is provided later within this document. Finally, the proposed Action Statement (d) is similar to the NRC guidance provided by NUREG-1431, with the exception of the alternative action to use the RHR suction relief valve as an additional backup in providing overpressure protection. A discussion regarding the use of the RHR suction relief valve is provided later within this document. PrODOBed ACtlOn Statement (0) The current Action Statement d specifies what to do if both PORVs (for LTOP) are inoperable. The proposed amendment expands the situation in which this Action ?tatement would be applicable. The proposed Action Statemen: (e) specifies required actions in the event that the LTOP system is inoperable for any reason other than the conditions affiliated with Action Statements a,b,c, or d. The Action Statement requires that RCS be depressurized and vented within eight hours. The completion time of eight hours considers the time needed to place the units in this condition and the relatively low probability of an overpressure event during this time period. The size opening of the vent is revised from 4.5 square inches to 2.75 square inches. A discussion regarding the size of the RCS vent opening is provided later within this document. In addition, the proposed Action Statement (e) is consistent with the NRC guidance offered in NUREG-1431. 13

PrODOBed RCt$OD EtatBEOnt (f) The current Action Statement e of Specification 3.4.9.3 becomes the proposed Action Statement (f). This is an administrative / editorial revision in that the only change proposed is the designation letter utilized within the specification. There are no changes to the wording of the Action Statement. PrODOmed RCtion statement (a) The current Action Statement f of Specification 3.4.9.3 becomes the proposed Action Statement (g). This is an administrative / editorial revision in that the only change proposed is the designation letter utilized within the specification. There are no changes to the wording of the Action Statement. Revisions to Surveillance Reauirements for SDecification 3.4.9.3 The proposed amendment revises the Surveillance Requirements (SR) for the LTOP system. Currently, the technical specification for the LTOP system specifies four (4)' surveillance to be performed, three for the PORV and one surveillance requirement for the RCS vent (s) if being used for overpressure protection. The proposed changes provided by this submittal will: 1) modify the RCS vent surveillance requirement; 2)will add three new SR that presently. do not exist within the technical specifications; and 3)will modify the footnote associated with Specification 4.4.9.3.2. The following discusses the proposed changes to the SR for the LTOP system provided by this submittal: ProDOBOd BDOClflCation 4.4.9.3.2 The proposed changes to Specification 4.4.9.3.2 are editorial in nature, no changes to_the requirements are intended. The specification was re-worded to be consistent in the format of other SR specifications. The current verification that the RCS vent is open is still required once every 12 hours when it is being used for overpressure protection. The size opening of the RCS vent is specified, whereas the current technical specification does not specify the size of the RCS vent opening. The size of the RCS vent opening is discussed later within this document. 14

ProDomed BDOCif$ Cation 4.4.9.3.3 The proposed Specification 4.4.9.3.3 provides a new surveillance requirement that currently does not exist and-y specifies that all Accumulators be verified isolated once every 12 hours. In addition, the specification requires that once every 12 hours verify.that only one NI or NV pump is capable of injecting into the RCS. The performance of these surveillance requirements will minimize the potential for an overpressure event by verifying that the mass input capability is limited. The frequency of 12 hours is-sufficient, considering other indications and alarms available to the operators in the control room, to verify the required status of the equipment. Further, the proposed new surveillance requirement is similar to NRC guidance provided within NUREG-1431. ProDomed SDecification 4.4.9.3.4 The proposed Specification 4.4.9.3.4 provides a new surveillance requirement that currently does not exist and specifies that the RHR suction isolation valves be verified open once every 12 hours when t'.e RHR suction relief valve is being used for overpressure protection. The performance of this surveillance will ensure that the RHR suction relief valve would be available to mitigate the consequences of an overpressure event. The frequency of once every 12 hours is considered adequate in view of other indications available to the operators in the control room that verify that the RHR suction isolation valves are open. This surveillance is only required to be performed if the RHR suction relief valve is being utilized for overpressure protection. Finally, the proposed new surveillance requirement is consistent with the NRC guidance provided by NUREG-1431. ProDomed BDecification 4.4.9.3.5 The proposed Specification 4.4.9.3.5 provides a new surveillance requirement that currently does not exist and specifies that, for each PORV previding overpressure protection during low temperature conditions, its associated block valve be verified open once every 72 hours. The performance of this surveillance will ensure that the flow path for each required PORV for overpressure protection is not blocked due to the PORV block valve and, thus, can perform its function when actuated. The 72 hour frequency is considered adequate due to other indications and alarms available to the operators in the control room, such as valve position indication, that verify that the PORV block valve remains open. 15 i

l N The ability to. utilize.the PORV block valve to isolate this RCS flow path'will.not be affected by this surveillance ; requirement. The power to the PCHur block valve operator will not be removed. The PORV-operator will not be_ locked ; in the inactive position. As such, the' block valve can-be closed in the event the PORV develops excessive leakage or does not fully close (sticks open) after relieving an overpressure event. , 4 Procomed-footnote to Boecification 4.4.9.3.2 :

   ,    The footnote associated with Specification 4.4.9.3.2 has     i been modified. The current footnote provides an alternative  i frequency for determining that the RCS vent is open.

Normally, Specification 4.4.9.3.2 requires-that the RCS vent be verified open once every 12 hours. The current footnote allows the frequency to be once per 31 days provided that the vent path is established by a valve that has been locked, sealed, or otherwise secured in the open position. The proposed technical specification change modifies the footnote by adding a statement that a PORV secured in the open position with its associated block valve open and power removed from the block valve can be utilized as an acceptable RCS vent, which only needs to be verified locked or secured in the open position once every 31 days. A PORV and ita associated block valve locked or otherwise secured - in the open position provides adequate protection in mitigating the consequences of an overpressure event. The

       . analysis performed in determining the low setpoint for the PORV confirms that a PORV is capable of relieving the flow and pressure increase resulting from the most limiting overpressure event (inadvertent safety injection signal).

The securing or locking a PORV and its associated block valve in the open position is considered to be: a passive device, thus, the PORV is not-subject to an active single failure. The flow path can also be isolated, if needed, by - restoring power to the associated block valve and closing the valve. 1 16

 -N i

i REDUCTION OF RCS VENT SIZE , The current minimum size of the vent' opening specified by the technical specifications for McGuire is 4.5 square inches. This , vent path is established in the event a PORV(s) is. inoperable and l after the RCS has been depressurized. The minimum size of the ! RCS vent opening needs to be large enough to ensure that the RCS

will be protected from overpressure' transients which could exceed the pressure / temperature limits that have been established in ; accordance with Appendix G to 10 CFR 50. With the RCS ; depressurized, the vent path must be capable of relieving the flow resulting from the limiting overpressure mass or heat input transient, and maintain pressure below the pressure / temperature limits. The vent size requirement of 4.5 square inches appears to be very conservative in. regard to the licensing bases for establishing an RCS vent during low temperature conditions. An + analysis has been performed which demonstrates that the 2.75 i square inch vent size provides more than adequate overpressure ' protection for LTOP events. The proposed Technical Specification change will reduce the specified size opening, in accordance with the analysis performed, from 4.5 square inches to 2.75 square inches. i The analysis performed to determine the area of the vent _ path is based on fluid flow equations from Crane Technical Paper No. 410,

     " Flow of Fluids Through Valves Fittings and Pipe". The assumed              I i

temperature of the RCS, for this calculation, is 75 F since the i pressure / temperature curves range from 75 F to 335 F and thus will be the most conservative value for this calculation, in terms of , minimizing pressure relief capability of the vent path. The vent ; discharges directly into containment atmosphere, as such, back l pressure will have no impact on the results. For this { calculation, the limiting overpressure transient (mass input) is i an inadvertent safety injection signal, which would start a : centrifugal charging pump (565 gpm)' and a safety injection pump ~ i (660 gpm) for a total flow capacity of 1225 gpm. For the purpose i of this calculation, the opening will be modeled as a nozzle or pipe exit on a pressurized tank. , To find the area of the vent path, the following equation was utilized; h, = 0 . 0 02 5 9 KQ'/ d' i where h, is 868 feet, K = 1.0 and is the resistance coefficient for a pipe exit, and Q = 1225 gpm and is the total flow capacity. r 17

6 In determining the head loss across the pipe exit (1() , the , pressure applied at the reactor vessel beltline.was assumed to be . I

 -442.18 psig. This pressure value was obtained from the proposed revision to technical specification Figure 3.4-4, " Unit'1 Reactor      !

Coolant System Cooldown Limitations - Applicable up to 16 EFPY". ', The pressure value of 442.18 psig represents the pressure limit : for a temperature of 75 F at a cooldown rate of 40 F/hr. The location of the beltline region is at an elevation of approximately 719'-0". The typical location for an RCS vent ! would be near the top of the pressurizer, which is at an ! elevation of 806'-0". As such, the amount of pressure head associated with this elevation difference is 87 feet. A margin of 12 F and 30 psig is provided, since the data from the proposed : curve does not include margin for instrument uncertainty. This results in a temperature and pressure of 87 F and 412.18 psig. Converting the pressure of 412.18 psig to head, a conversion ; factor of 2.317 feet /psig was utilized. Finally, the head loss across the pipe exit (nozzle) is 955 feet minus elevation difference head loss (87 feet), which equates to 868 feet. To determine the size of the opening necessary for establishing an -t RCS vent, the above equation is re-arranged to solve for d. As a l result, the diameter of the opening would be 1.45 inches, which i results in an area for the vent opening of 1.66 square inches. . The calculated size of the vent path opening is 1.66 square inches. Although this could be an acceptable size for the vent ' path opening that could be referenced within the Tecnnical - Specifications, it would not provide any margin for error or uncertainty associated with operation or the analysis. In addition, the pressurizer would not be modified to create a new . opening or nozzle when establishing an RCS vent path. Typically, l a pressurizer safety valve is removed when establishing an RCS i vent path. The inlet piping (inner diameter of 5.187 inches) for the pressurizer safety valve would then be utilized for , establishing the vent path. To determine the minimum size of the i opening, the problem can be defined as determining the size of an [ orifice plate at the end of the pressurizer safety valve inlet piping. The equation for calculating the size of the opening for the orifice plate is (from Crane Technical Paper No, 410); 6 I q = A *C ( 2

g
h,) *"

where: q is the flow rate (1225 gpm) in feet'/second; A is the area of the orifice in' square feet; C is the flow coefficient for the orifice (0.61); g is gravity; h, is the head loss across the orifice, [868 feet minus the head loss due to the inlet piping (14 feet)] which is 854 feet. f 18

Solving for A results in an area of 0.01908 square feet or about 2.75 square inches. With the RCS de-pressurized, the analysis has determined that an orifice with a opening of 2.75 square inches at the end of the pressurizer safety valve inlet piping is capable of mitigating the consequences of an overpressure transient during low temperature conditions. As such, the flow capacity of a vent with a opening of 2.75 square inches is greater than the flow of the limiting transient (inadvertent safety injection signal). As added conservatism, the pressure / temperature limits, utilized in the analysis to determine the size of the vent opening, do NOT incorporate the provisions of ASME Code Case N-514 (110% of the pressure determined to satisfy Appendix G of Section XI). As shown by the analysis, an acceptable configuration for establishing an RCS vent is an orifice plate with an opening of 2.75 square inches at the end of the inlet piping for a pressurizer safety valve. Other alternative configurations which would be capable of relieving the flow resulting from the limiting mass input transient, and maintaining pressure below the pressure / temperature limits are:

1) Removing a pressurizer safety valve and utilizing the inlet piping without an orifice plate as the RCS vent path, this would provide an opening of 21.15 square inches or 5.187 inches in diameter (inner diameter of the inlet piping).

2) Securing or locking a PORV in the open position. The associated block valve will be open, disabled, and power removed.

3) Opening of other RCS vent lines to the containment atmosphere, where the equivalent opening would be greater than or equal to 2.75 square inches. More than one vent path could be used to establish an equivalent opening of 2.75 square inches. Further, the vent path (s) shall be located such that the RCS can not be drained when the vent (s) is opened and that the total piping configuration has a resistance coefficient (k) equivalent to approximately 30 feet or less of straight pipe with an inner diameter of 5.187 inches.

Finally, the RCS vent shall discharge to the containment atmosphere except when a PORV is locked open and its associated block valve is open with power removed. This is considered to be a passive device and, as such, is not subject to an active single failure. I 1 I i l 19 l

USE OF RHR SUCTION RELIEF VALVE The current LTOP technical specification does not define how to use the RHR suction relief valve for overpressure protection during low temperature conditions. One prerequisite necessary in order to utilize the RHR suction relief valve for overpressure protection is the deletion of the RHR suction isolation valve autoclosure interlock. This will prevent the closure of the RHR suction isolation valves (ND1 and ND2). By letter dated September 11, 1990, the NRC approved a proposed Technical Specification amendment request for McGuire which, in effect, authorized the removal of the RHR autoclosure interlock circuitry. As such, the RHR autoclosure interlock circuitry at McGuire has been deleted. During low temperature conditions, the RHR system is operated for decay heat removal. For plant cooldown, after the reactor coolant temperature and pressure have been reduced to approximately 350 F and 385 psig, the RHR system is placed into i operation. While the RHR system is in service, the RHR suction relief valve (ND3) is exposed to the RCS and is able to relieve RCS overpressure transients. Enclosed is a summary flow diagram of the RHR system illustrating the location of the RHR suction isolation valves (ND1 and ND2) and the location of the RHR suction relief valve (ND3). The proposed changes to the technical specification will define the specific situations and conditions under which the RHR suction relief valve can be utilized for overpressure protection , during low temperature conditions. The specific situations in which the RHR suction relief valve can be used for overpressure protection are;

1) to allow for a second ECCS system pump to be able to inject into the RCS; or ,

2) when one PORV is inoperable while in Modes 5 or 6.

When these situations occurs, the RHR suction relief valve can be utilized for overpressure protection provided that the RCS temperature is greater than 167 F, and that the RHR suction isolation valves (ND1 and ND2) are open. In addition, for the first situation (use of a second ECCS system pump), the utilization of the RHR suction relief valve for overpressure protection below an RCS temperature of 167 F is allowed provided that the RCS temperature is greater than 107 F, that ,the unit is being shutdown and the rate of the cooldown is less than 20 F per hour. The use of the RHR suction relief valve under these restricted conditions provides an equivalent means of overpressure protection when compared to a PORV. 20 j

As stipulated.within Specification 3.4.9.3, only one NV pump or one NI pump can be capable of injecting into the RCS during low temperature conditions. There will be brief periods of times when it will be desirable to operate with more than one pump capable of injecting into the RCS. For example, during the swapping of one pump for.another, to be able to fill'the accumulators, and to be able to perform testing of certain ECCS check valves. An analysis was performed to show that the RHR suction relief valve can be utilized to relieve the extra capacity associated with the above situations, and, thus, provide an equivalent means of overpressure protection when compared to a PORV. The_ characteristics of the RHR suction relief valve (ND3) that were used.within the analysis are: Setpoint: 450 psig Capacity: 902 gpm Accumulation: 110% of setpoint (ASME III, NC-7000) The capacity of the RHR suction relief valve is 902 gpm. This is adequate to relieve the full flow of either a NV pump (565 gpm) or a NI pump (660 gpm), but not both pumps (NV and NI). The transient peak RCS pressure for the RHR suction relief valve relieving at maximum capacity is 555 psig. This assumes accumulation of 10% of the rated lift setpoint, a correction for the elevation differences between the location of the RHR suction relief valve and the reactor vessel beltline, and a correction for the pressure drop across the reactor _ vessel ~due to operation of the reactor coolant pumps. For the situation where two' pumps are capable of injecting into the RCS, both PORVs and the RHR suction relief valve are required to be operable. The RHR suction relief valve in conjunction with a pressurizer PORV, are adequate to relieve the combined flow of two NV_ pumps; or two NI pumps; or one NV and one NI pump. For the situation where two pumps are capable of injecting into the RCS, two of the three relief valves will be adequate to mitigate the consequences of an LTOP event. As discussed earlier, the resulting peak pressure, given a PORV setpoint of 385 psig, is approximately 537 psig. The resulting peak pressure for the RHR suction relief valve (555 psig) is somewhat higher than what would occur with the PORV. As such, restrictions associated with the use of the RHR suction relief valve are necessary to ensure compliance with the licensing requirements for LTOP. 21

Based on a review of the pressure / temperature limits (established per 10 CFR 50 Appendix G and in accordance with the provisions of ASME code case N-514), the allowable pressure for a 60 F heatup rate is approximately 570 psig. For steady state operation, the pressure limit at 75 F is 582 psig. As such, the RHR suction relief valve is adequate for all steady state and heatup operations. For cooldown rates of 100 F/ hour, the allowable pressure, established per 10 CFR 50 Appendix G and in accordance with the provisions of AS. Code Case N-514, is about 570 psig at 167 F (includes instrume t uncertainties). Below an RCS temperature of 167 F, the resulting peak pressure of an LTOP event mitigated by the RHR suction relief valve could exceed the allowable pressure for a cooldown rate of 100 F/ hour. As such, the use of the RHR suction relief valve below an RCS temperature of 167 F will be restricted. Below an RCS temperature of 167 F, the RHR suction relief valve can be utilized provided that the cooldown rate is limited to 20FF/ hour or less and that the RCS temperature is greater than 107 F (includes instrument uncertainties). The allowable pressure for a cooldown rate of 20FF/ hour or less is about 562 psig at 107 F. Below an RCS temperature of 107 F, the resulting peak pressure of an LTOP event mitigated by the RHR suction relief valve could exceed the allowable pressure for a cooldown rate of 20 F/ hour or less. As such, the use of the RHR suction relief valve below an RCS temperature of 107 F will be prohibited. These restrictions have been incorporated within the proposed amendments to the Technical Specifications. For the RHR suction relief valve the calculated peak pressure resulting from an overpressure event is 555 psig. Accordingly, the utilization of the RHR suction relief valve for mitigation of LTOP events is allowed as summarized below: RCS COLD LEG ALLOWABLE MAXIMUM CONDITION TEMPERATURE

PRESSURE

( F) (PSIG) STEADY STATE 2 87 582.14 OPERATION HEATUP RATE 2 87 560.71 5 6(fF/ HOUR HEATUP RATE 2 167 569.67 s 100 F/ HOUR COOLDOWN RATE 2 107 562.40 s 2CfF/ HOUR Indicated Temperature (with 12 F margin)

     **    In accordance with ASME Code Case N-514 and no margin for instrument uncertainty 22
                                                                   ]

REVISION OF PRESSURE / TEMPERATURE CURVES The proposed amendment to the technical specification will update the heatup and cooldown limit curves for both units. The service period for the proposed new curves have been extended from 10 EFPY to 16 EFPY. The new pressure / temperature limits satisfy all l required material embrittlement considerations including: ASME Section XI, Appendix G; 10 CFR 50, Appendix G; and Regulatory Guide 1.99 Revision 2. The development of these curves was performed by Westinghouse Electric Corporation, they were included in the appendices of the Westinghouse surveillance capsule reports WCAP-13949 and WCAP-13516. These surveillance l capsule reports were submitted to the NRC for review by letters ; l dated March 24, 1994 and January 27, 1992. These surveillance I capsule reports (WCAP-13949 and WCAP-13516) provide the detailed discussion of the technical data, and evaluations that were performed in support of the development of the proposed heatup and cooldown curves for both units. The layout of Figures 3.4-2, 3.4-3, 3.4-4, and 3.4-5 (the heatup and cooldown curves for both units) is revised by the proposed changes. The information that is found currently within each figure continues to be provided, i.e. it is just arranged differently. This is considered to be an administrative type of change, in that the proposed change does not involve any technical changes, just how the information looks on the page. 1 i ? I i 23

RELOCATION OF INSTRUMENT ERROR Currently, the instrument uncertainties are incorporated within the heatup and cooldown curves of the technical specifications (Figures 3.4-2,-3,-4, and -5). The heatup and cooldown curves conservatively include the calculated instrument uncertainties. The assumed allowable margins identified within the technical specifications to account for instrument uncertainties are 10 F and 60 psig. The proposed amendment requests that the instrument error margins be relocated to other documents that are controlled by the I licensee. The instrument margins will be administratively implemented by incorporating them into the controlling procedures for unit operations and into the LTOP system setpoint selection calculations. The relocation of the instrument error to licensee controlled documents is consistent with the NRC actions proposed within NUREG-1431, new standard technical specifications for Westinghouse plants. As prescribed within NUREG-1431, the pressure / temperature limit curves are to be relocated to a licensee controlled document-entitled " Pressure Temperature Limit Report (PTLR)". Changes to l the heatup and cooldown curves would then be performed in accordance with 10 CFR 50.59 criteria. For the situation proposed by this amendment, updates and revisions of the instrument error associated with the pressure / temperature limit curves will be processed in a similar fashion. Thus, the proposed change to relocate the instrument error to licensee controlled documents is analogous with NRC acceptable practices. Accordingly, the proposed heatup and cooldown curves provided by this submittal do not include'any margin to account for instrument uncertainties. The assumed margins for instrument error will be administratively implemented by incorporating them into the controlling procedures for unit operations and into the LTOP system setpoint analysis. The assumed margins to account for instrument uncertainties associated with this amendment request are 12 F and 30 psig. The margin assumed for the temperature instrument uncertainty is conservatively increased. This was done to provide for additional margin, there were no changes or modifications made to the temperature instruments or how they are maintained or used. The margin assumed for the pressure instrument uncertainty is reduced, from 60 psig to 30 psig. This reduction in the pressure instrument uncertainty is a result of a modification to replace the wide range RCS transmitters with a narrow range pressure transmitters. To verify if the assumed pressure instrument uncertainty is appropriate, a calculation was performed to determine the total instrument loop uncertainty for the RCS narrow range pressure instrumentation and its associated LTOP function. l 24

i i

i The instrument loops in. question are composed'of the following

                                                                                  -l components:

1) Rosemount 1153 series B Pressure' Transmitter j

2) Westinghouse NLP group 2 Isolator and Loop power

Supply card t

3) Westinghouse NAL' single signal comparator card The1 instrument loops have a process range of 0 to 600 psig (the

 -transmitters are isolated from the RCS above this pressure). .The-               l pressure transmitters have a maximum input range of 0 to 1000L psig,.therefore the upper range limit (URL) of the transmitter is 1000'psig. .The pressure transmitters and indications have a                     ,
 ' calibration range of 0 to 600 psig, while the Westinghouse 7300                 ;

cards have a calibration range of 0 to 30 volts corresponding to a process of 0 to 600 psig. l

 --The calculation utilized to develop the total loop uncertainty                   '

uses'the basic statistical equation: Z= [ A'+B'+C'+ (D+E) '] *il F l +L-M where: l Z total uncertainty.

            'A,B,C          random and independent uncertainty terms               [

D and E random dependent uncertainty. terms that are independent of Terms A,B, and C ' F arbitrarily-distributed uncertainties and/or ; biases with unknown' sign , L and M biases with known direction - The calculated worst case total loop uncertainty associated with , the RCS pressure setpoint for an actuation of a PORV during a ; LTOP event is 21 psig. The instrument loop uncertainty ; analysis'shows that the calculated total loop uncertainty for=the ! RCS pressure with regard to the LTOP setpoint for the PORVs is ! less than the allowed total loop uncertainty that was assumed in : the analysis for determining the PORV LTOP setpoint. l t b 25 f 1 _ . . . - -

i lm

f*-136%No-Bo l T L

h J L J 685 583 [ 5bfION) 5u KC ND PUMP MECH. SEAL HX A 7e

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                                                                                                                                                                                  -                      a                                                   (NC LOOP IL2 39                                                                                                               COLD LEG) 6                                                 J                                    RHR                                                                                            NC HR                                                                    HX R                                                               56 [

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                                                                                                                                                  % (33 (LCD 3)                               "y g'                                                                                                                                                                                                       ,,

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a : ^# N ' k LOOP 2 & 3 HOT LEG) k( 18 63 NC m, 48 : - KC 159 , ( 17 { A, ]F T F 6 d E ' (c0Nv. REcrRUI sune)

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3_ _ _ _ ,_ a __sm_sA. - J m a A a h l ATTACHMENT 3 , 1 DUKE POWER COMPANY MCGUIRE NUCLEAR STATION l

NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS 1 l l ______ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ -- , - , , - - - - - - -v- -

Duke Power Company has evaluated the proposed changes associated t with this License Amendment Request and determined that they involve no significant hazards considerations. In accordance with 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility, in accordance with the proposed amendment, would not: f ) 1. Involve a significant increase in the probability or the 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.

l The proposed changes amend the McGuire Nuclear Station Units 1 and 2 Technical Specifications 3.4.9.3 " Overpressure Protection Systems", 3.5.3 "ECCS Subsystems - Tavg < 3 50 F" and the Pressure / Temperature limit curves, Figures 3.4-2, 3.4-3, 3.4-4 and 3.4-5. These changes revise the Low Temperature Overpressure Protection (LTOP) system maximum setpoint, minimum vent requirements, and ECCS subsystem availability requirements to enhance system operation and reliability. These changes also { revise the Reactor Coolant System Pressure / Temperature limit curves to include the most recent radiation surveillance capsule analysis. The format has been revised to conform to the extent possible with the new Standard Technical Specifications for Westinghouse Plants (NUREG 1431). The proposed technical specification changes can be grouped as follows:

1) Revisions to the Limiting Condition for Operation (LCO) requirements, the Action Statements (AS) and the Surveillance Requirements (SR) for the Reactor Coolant System (RCS) Overpressure Protection System during low temperature conditions.

2) Reduction in the Reactor Coolant System (RCS) vent requirement from 4.5 square inches to 2.75 square inches.

3) The use of the Residual Heat Removal (RHR) suction relief valve (1ND3 and 2ND3) for overpressure protection under restricted conditions. (RCS > 107 F and cooldown rate less than 20 F/hr; or RCS > 167 F) . l l

4) Revisions of the Pressure / Temperature curves to 16 EFPY, including the incorporation of the latest radiation surveillance capsule results and removal of instrumentation margins from the technical specification figures.

5) Changes to format and consistency.

1 1

 - - _ - _ - _ _ _ _ _ _ _ . _     - _                                                             h

The~following is a'brief discussion of the significant changes associated with each of the_five (5) identified groups:

1) Revisions to the Lbaiting Condition for Operation (LCO) requirements, the Action Statements (AS) and the_

Surveillance Requirements (SR) for the Reactor Coolant System (RCS) Overpressure Protection System during low temperature conditions. , This change reduces the maximum allowable setpoint of the PORV during low temperature conditions (RCS cold leg temperature s 300 F) . Since the maximum set-pressure is reduced, the peak pressure for Low Temperature Overpressure Protection (LTOP) events will be reduced by this amendment change. Analysis performed to calculate this setpoint ensure that the resultant peak pressure for an LTOP event is limited to 110% of the pressure determined to satisfy Appendix G of Section XI (as established by ASME Code Case N-514). This change, also, defines additional conditions for establishing the operability of the LTOP system, what actions to take when the additional conditions are not met and additional surveillance requirements. For the LTOP system to be considered operable, all of the accumulators need to be isolated and only a maximum of one Centrifugal Charging (NV) pump or one Safety injection (NI) pump capable of injecting into the RCS. _This assures that unit operation and configuration during low temperature conditions is consistent with what was assumed in the analysis of LTOP events. Further, the. incorporation of these two conditions within the technical specifications is consistent within the new standard technical specifications for Westinghouse plants (NUREG-1431).

2) A reduction in the Reactor Coolant System (RCS) vent requirement from 4.5 square inches to 2.75 square inches.

The bases for the LTOP vent size at McGuire is to ensure that the 10 CFR 50 Appendix G Pressure / Temperature limits (does nah take into consideration the impact of ASME Code Case N-514) are not exceeded for the combined capacity of a safety injection pump and centrifugal charging pump injecting into the RCS. The 4.5 square inch vent size was evaluated using this criteria and determined to be considerably larger than needed for this basis. An analysis has been performed which demonstrates that the 2.75 square inch vent size provides overpressure protection for the combined capacity of a centrifugal charging and safety injection pump. This vent path is used for conditions where the PORV's may not be available. l 1 2

3) The use of the Residual Heat Removal (RHR) suction relief valve (1ND3 and 2ND3) for overpressure protection under restricted conditions. (RCS > 107 F and cooldown rate less than 20 F/hr; or RCS > 167 F) .

The use of the Residual Heat Removal (RHR) suction relief valve under certain restricted conditions provides an

equivalent means of overpressure protection when compared to a PORV. The conditions established ensure that the peak 4 I

pressure reached by the relieving RHR suction relief valve will not exceed the most restrictive Appendix G limits for either Unit (does include the consideration of the impact of ASME Code Case N-514). The specific use of the RHR suction relief valve, as allowed by this amendment change is to i allow a second ECCS system pump to be made available, in ! l which case both PORV's and the RHR suction relief valve l would be available to protect the pressure / temperature limits established in accordance with 10 CFR 50 Appendix G. Any two of the three available relief valves have the necessary combined capacity. Another situation in which the , RHR suction relief valve could be utilized is when a PORV is l inoperable while in mode 5 or 6. The RHR suction relief l valve is a spring actuated liquid relief valve which does ' not require any support systems (such as instrument air or i electrical controls) which enhances reliable operation. l The RHR system suction isolation valve autoclosure i interlocks (which would isolate the.RHR suction relief valve upon rising pressure) have been deleted at McGuire Nuclear Station. NRC approval of the Technical Specification amendment which, in effect, authorized the removal of the RHR autoclosure interlock circuitry was provided by a letter dated September 11, 1990. The RHR suction relief valve is normally in service during LTOP conditions. The proposed technical specification changes ensure proper RHR system alignments, and that the relieving capacity is sufficient to protect the pressure / temperature limits.

4) Revisions of the Pressure / Temperature curves to 16 EFPY, including the incorporation of the latest radiation surveillance capsule results and removal of instrumentation margins from the Technical specification figures.

The pressure / temperature curves were developed by Westinghouse Electric Corporation and are included in the appendices of WCAP-13949 and WCAP-13516 for Units 1 and 2 respectively. These surveillance capsule reports, (WCAP-13949 and WCAP-13516), were submitted to the NRC for review by letters dated March 24, 1994 and January 27, 1992, respectively. These curves extend the service period from 10 EFPY to 16 EFPY. The new pressure / temperature curves have satisfied all required material embrittlement considerations including; 10CFR50 Appendix G, ASME Section XI Appendix G and Reg. Guide 1.99 Rev. 2. 3

This change, also, allows for'the' reduction in the margin for instrument error.of the'LTOP system which has been made possible by enhancements in the actuation circuitry. Specifically, the wide range-reactor' coolant system pressure transmitters will be replaced with narrow range transmitters capable of a more accurate setpoint. This modification-will add a QA-1-narrow range pressure transmitter (0-600 psig) to each LTOP train'of:the NC system. The purpose _of this modification'is to substantially reduce _the instrument uncertainty associated with the LTOP function and its-related effects on the Appendix G heatup and cooldown I i curves.-The transmitters function will'be to monitor.NC-system pressure and provide a signal to the pressurizer PORV' 7300 bistable for low temperature overpressure protection. The narrow range transmitters being added by this modification will only be used during LTOP mode. During_ normal operation, these transmitters will be valved out. These transmitters replace the LTOP function of the existing wide range NC system pressure transmitters. The impulse lines of the new transmitters will tap into the impulse lines of the existing wide range transmitters. Double isolation will be used since the transmitters will normally be isolated from the RCS during power operation. Furthermore, the instrumentation margins have been removed from the curves, since the LTOP system will.be using narrow range, high accuracy (s 30 psig uncertainty) pressure transmitters. The normal wide range pressure transmitters l (s 70 psig uncertainty) will still be used for power operation and other non-LTOP conditions. The instrument margins are implemented administratively by incorporation into the controlling procedures for unit operations. Additionally, the instrument uncertainties have been l incorporated into the LTOP system setpoint selection calculations.

5) Changes to format and consistency.

Other changes have been made to improve consistency between specifications, incorporate Westinghouse standard technical specification format, and update applicable code references. These changes are considered editorial in nature and do not affect the way the units are operated nor the safety functions performed by the LTOP system. The following discussion is a summary of the evaluation of the proposed amendments against the 10CFR50.92(c) requirements to demonstrate that all three standards are satisfied. Within each of the three standards, the five identified groups of the proposed technical specification amendment request are discussed. 4

I FIRST STANDARD (Amendment would not) involve a significant increase in the , probability or consequences of an accident previously evaluated. .I

1) Revised LCO and SR for LTOP:

The reduced maximum setpoint will prevent the violation of the 10 CFR 50 Appendix G pressure / temperature curves (as modified by the provisions of ASME Code Case N-514) during overpressure transients at low temperatures. Since the maximum setpoint is reduced, the peak pressure for LTOP events will be reduced as well. Accordingly, the consequences of an LTOP event would not change as result of the proposed changes. The analysis performed to determine the setpoint is, in accordance with the methods used in previous evaluations, found acceptable by the NRC. The three possible transients evaluated are; 1) a mass input from an operable safety injection pump; 2) a mass input from an operable centrifugal charging pump; and 3) a heat input from a 50 F temperature difference between the steam generators and the NC system. The LTOP setpoint of the PORV proposed by this technical specification change is not considered to be an initiator of any of these three transients. As such, the probability of an accident previously evaluated would not be increased as a j j result of the proposed changes. l ! Two additional conditions for operability of the LTOP system are defined (accumulator isolation and only one NV or NI pump operable) and new surveillance requirements are specified as well. They provide additional limitations, requirements and restrictions that currently do not exist within the technical specifications for McGuire. The incorporation of these proposed changes are consistent with what is specified within NUREG-1341. Therefore, these 5 changes do not increase the probability or conseg'ences of an accident previously evaluated. ! 2) Reduction in NC vent onenina: The bases for the size of the vent to be established per the technical specifications is to ensure that the 10 CFR 50, Appendix G pressure / temperature limits are not exceeded during an LTOP event. The determination of the size of the

l. opening continues to preserve the above design basis. The evaluation performed demonstrated that a 2.75 square inch opening would provide adequate overpressure protection for the combined capacity of a centrifugal charging pump and a safety injection pump.

5

The only time that.the vent path is to be established is. when the PORVs may not be available. Defining the size of the vent'is not considered to be an initiator of any LTOP events that have been previously' evaluated. As such, this change in.the size of the vent. opening does not increase the probability of an overpressure event during low temperature conditions. The analysis performed verifies'that the size opening specified is sufficient to mitigate the consequences of an LTOP' event. Accordingly, the change in the size of the opening for the vent will not impact the consequences of LTOP events.

3) Use of RHR suction relief valves:

By letter dated September 11, 1990, the NRC authorized the deletion of the RHR autoclosure interlock circuitry. A modification which removed the RHR system suction isolation valve autoclosure interlocks has been completed. As.such, the RHR suction relief valve can be exposed to NC system pressure and would be available to mitigate LTOP events. The proposed amendments specify the necessary requirements-and controls to ensure proper ND system alignments and l conditions will exist to protect the pressure / temperature I limits. This added relieving capacity will enhance the current LTOP system at McGuire in mitigating overpressure events at low temperatures. As such, the mitigation of f previously evaluated LTOP events would be improved by the proposed technical specification changes. Further, the proposed changes would not result in'the initiation of an LTOP event or cause an overpressure transient. Accordingly, the proposed amendment would not involve an increase in the . consequences or the. probability of an accident previously evaluated.

4) Revised oressure/temoerature curves to 16 EFPY:

The proposed pressure / temperature curves, provided by this amendment request, satisfy all regulatory required material embrittlement considerations including: ASME Section XI Appendix G, 10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. In addition, the margins for instrument error have been removed from the curves. Instrument error will be a?ministratively handled by incorporating them into the LTOP system setpoint selection calculations and into appropriate controlling procedures for unit operations. 6

a t The. proposed changes to the pressure /tamperature curves are not considered to be an initiator of LTOP events. The changes to the curves proposed by this amendment request will not cause an LTOP event. The curves define the new limits that have been defined in accordance with regulatory-requirenents by which both units are to be operated within. Accordingly, the proposed amendment will not increase the probability or the copaequences of previously evaluated accidents.

5) Format and consistencv: ,

The changes associated within this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, the proposed amendment request does not increase the probability or consequences of any previously evaluated accident. SECOND STANDARD '; (Amendment would not) create the possibility of a new or different kind of accident from any kind of accident previously evaluated.

1) Revised LCO and SR for LTOP:

The only potential impact to plant systems, structures and components, as a result of the proposed changes associated with this group, would be the setting of the PORV low , pressure setpoint. No other changes to plant systems, structures or components would occur. The proposed amendments, also, would not impact the plant operation. Although the value for the PORV pressure setting specified within the technical specification would be reduced per the proposed amendment, the actual settings of the PORV are now currently set low enough to comply with the proposed lower setpoint value. As such, the proposed lower setpoint would not require any changes to the plant nor how the plant is operated. The additional requirements for LTOP operability will not require any modifications to the plant nor how the plant is , operated. Currently, when entering LTOP conditions, the accumulators are isolated and only one NV or NI pump is capable of injecting into the reactor vessel. These actions are currently controlled and are specified within the operating procedures for heatup and cooldown of the respective units. The proposed changes will now specify l these current operating requirements within the technical specifications as well. l j 7

a i r Accordingly,- the proposed revisions will not create ainewfor' I different kind of accident.than what has already been t previously evaluated. t

!t

2) Reduction-in NC vent oneninca The proposed changes to the technical' specifications i associated with this group involves thet size of'the vent !

opening. The proposed' amendment reduces the size of the ! vent opening from 4.5 square inches to 2.75' square inches. The analysis.that was performed has determined that the proposed size for the vent opening is adequate for 3 ' overpressure events. Therefore, this proposed revision to-the technical specifications will not result in a new or different kind of accident from any kind of accident previously evaluated. '[

3) Use of RHR suction relief valves: '!

The proposed amendment associated with this group will j specify the necessary requirements and controls to ensure ; the appropriate use of the RHR suction relief valve for-overpressure protection. This added relieving capacity will , enhance the current LTOP system in mitigating overpressure l events during low temperature conditions. The analysis that ! has been-performed demonstrates the adequacy of the RHR suction relief valve, in conjunction with a PORV, in ; mitigating overpressure events at low temperatures,-assuming 1 ; a worst case single failure as well. As such, the use of' ; the RHR suction relief valve in the manner prescribed by the J ' proposed technical specification amendment will'not create a new or different kind of accident from those accidents'that , have been previously evaluated. !

4) Revised oressure/temoerature curves to 16 EFPY: !

The changes associated with this group, provide new heatup : and cooldown curves for both Units 1 and 2, which will . extend the service period from 10 EFPY to 16 EFPY and will remove the instrument error as well. The proposed heat up and cooldown curves were developed in accordance with all ; regulatory required material embrittlement criteria. Thus, i operation of the units in accordance with the proposed new p pressure / temperature curves will not create the possibility ; of a new or different kind of accident from those accident i that have been previously evaluated. ! i 8

     '5) Format and consistencv:

The changes associated within this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, the proposed amendment will not create the possibility of a new or different kind of accident from that previously evaluated. THIRD STANDARD (Junendment would not) involve a significant reduction in a margin -of safety.

1) Revised LCO and SR for LTOP-This proposed change will reduce the maximum PORV setpoint such that, for LTOP events, the maximum pressure in the ,

vessel would not exceed 110% of the pressure / temperature i limits that have been established in accordance with ASME Appendix G. This is congruous with the provisions of ASME - Code Case N-514. Currently, the maximum PORV setpoint for LTOP events ensure that the maximum pressure would not exceed 100% of the pressure / temperature curves. As such, the proposed change appears to involve a slight reduction in a margin of safety. Although the proposed change may involve a slight reduction in a margin of safety, the proposed change will provide an equivalent margins of safety to'the reactor vessel during LTOP transients and will satisfy the underlying purpose of 10 CFR 50.60 for fracture toughness requirements. By letter dated June 28, 1994, an exemption request and authorization to use ASME Code Case N-514 at McGuire was submitted to the NRC for review and approval. Approval for the use of the code case was granted on September 30, 1994. The proposed change to reduce the maximum PORV setpoint, coupled with the September 30, 1994 NRC approval for the use of Code Case N-514 satisfies current regulatory acceptance criteria. Therefore, the proposed change would not involve a significant reduction in a margin of safety. This change group, also, defines two additional conditions for the operability of the LTOP system (accumulator isolation and only one NV or NI pump operable) and proposes new surveillance requirements. This provides additional limitations, requirements and restrictions that currently do not exist within the technical specifications for McGuire. The incorporation of these proposed changes are consistent with what is specified within NUREG-1341. Therefore, these e changes do not involve a significant reduction in a margin ; of safety. 9

21 Reduction in NC vent onenina: The proposed changes to the technical specifications associated with this group involves the size of the vent opening. The proposed amendment reduces the size of the vent opening from 4.5 square inches to 2.75 square inches. i t The basis for the size of the vent to be established per the technical specifications is to ensure that the 10 CFR 50, Appendix G pressure / temperature limits are not exceeded during an LTOP event. The determination of the size of the opening continues to preserve the above design basis. The evaluation performed demonstrated that a 2.75 square inch opening would provide adequate overpressure protection for the combined capacity of a centrifugal charging pump and a safety injection pump. Accordingly, the proposed changes would not involve a significant reduction in a margin of safety.

3) Use of RHR suction relief valves 1 The proposed amendment associated with this group will specify the necessary requirements and controls to ensure the appropriate use of the RHR suction relief valves for overpressure protection. This added relieving capacity will enhance the current LTOP system in mitigating overpressure events during low temperature conditions. The analysis that has been performed demonstrates the adequacy of the RHR suction relief valve, in conjunction with a PORV, in mitigating overpressure events at low temperatures.

Further, by letter dated September 11, 1990, the NRC approved amendments to delete a portion of the surveillance requirements regarding periodic verification that the RHR ' suction isolation valves automatically close on a RCS signal less than or equal to 560 psig. This action, in effect, authorizes the removal of the RHR autoclosure interlock circuitry. As discussed within the NRC Safety evaluation for the amendment, the Commission and industry have recognized the safety benefits of removing the ACI circuitry from the RHR system to minimize, and thus reduce the risk associated with loss of decay heat removal events. Therefore, the proposed amendments associated with this change group will not involve a significant reduction in a margin of safety. 10

w 9

     ,  4) Revised pressure /temnerature curves to 16 EFPY-             .l The changes associated'with this group provide new heatup and cooldown curves'for both Units 1 and 2, which will            -

extend the service period.from.10'EFPY to 16.EFPY and will : relocate the instrument error as well. .The proposed -

                                                                        'l
 "      pressure / temperature curves provided by-this amendment
                 ~

K ' request satisfy.all regulatory required material embrittlement considerations including;.ASME Section XI Appendix G,.10 CFR 50 Appendix G, and Regulatory Guide.l.99, ' Revision 2. zThe instrument error will be administrative 1y handled by incorporating them into the-LTOP system setpoint selection calculations and into the controlling procedures for unit operations. l The relocation of the instrument error to licensee controlled documents is consistent with the NRC actions proposed within NUREG-1431, new standard technical i specifications for Westinghouse plants. As prescribed , within NUREG-1431, the pressure / temperature limit curves are ! to be relocated to a licensee controlled document entitled !

        " Pressure Temperature Limit Report (PTLR)". Changes _to the heatup and cooldown curves would then be performed in accordance with 10 CFR 50.59 criteria. For the situation          I proposed by this amendment, updates and revisions of the          i instrument error associated with the pressure / temperature    . -;

limit curves will be processed in a similar' fashion. Thus, the proposed change to relocate the instrument error to i licensee controlled documents is analogous with NRC acceptable practices. i Accordingly, the proposed changes will not reduce a margin of safety. -

5) Format and consistency:

The changes associated within this group are considered to } be administrative in nature. They do not affect station t operability or require any modifications to the facility. ' Accordingly, there is no reduction in the margin of safety of the LTOP system due to the incorporation of these editorial / administrative changes. Based on the above and the supporting technical justification, Duke has concluded that there is no significant hazard consideration involved in this request. r t f P i 11 i

4_ ,,m,j _. #.--*_ , a _ __ .@,1 .2.u. - -_ _ _ . -- 4 l l ATTACHMENT 4 J DUKE POWER COMPANY MCGUIRE NUCLEAR STATION LTOP TS SUBMITTAL HISTORY . P 4 i a

                    ..            -           -         -      -- - - - .-- -- ---      >n -

LTOP TS SUBMITTAL HISTORY In addition to this technical specification amendment request, other submittals concerning the LTOP system and the pressure / temperature curves have been made. The following provides a synopsis of the other submittals that have been made regarding the LTOP system and the pressure / temperature curves at McGuire.

1. By letters dated March 24, 1994 and January 27, 1992, Westinghouse surveillance capsule reports for Units 1 and 2 (WCAP-13949 and WCAP-13516), respectively, were submitted to the NRC for review. These surveillance capsule reports provided the technical data and analysis utilized to update the heatup and cooldown curves for both units.

2. By letter dated March 24, 1994, information which discusses the feasibility of utilizing the Diablo Canyon, Unit 2 surveillance weld metal as a credible data source was submitted for NRC review and approval. Additional information supporting this licensing action was provided by a May 18, 1994 letter. The use of the Diablo Canyon, Unit 2 data provides an enhancement to the McGuire Unit 1 reactor '

vessel surveillance program by improving the reliability of the data utilized in predicting the amount of radiation damage.

3. By letter dated June 28, 1994, an exemption from certain requirements of 10 CFR 50.60 regarding the use of American Society of Mechanical Engineers (ASME) Code Case N-514 and NRC authorization to utilize ASME Code Case N-514 was submitted. This licensing action will provide an ;

alternative criteria for calculating an acceptable setpoint for the LTOP system at McGuire. The exemption was granted j by the Commission on September 30, 1994.

4. By letter dated November 21, 1991, a proposed license ,

amendment to the McGuire technical specification was submitted. The proposed amendment request was submitted in . response to the NRC recommendations provided by Generic l Letter 90-06, and provided proposed changes to Specification 3.4.9.3 (LTOP). NRC approval of the proposed license amendment submitted by the November 21, 1991 Duke Power letter was provided by an NRC letter dated October 27, 1994. This change has been incorporated into the McGuire i technical specification. This amendment makes further changes to the same technical specifications. i

U > e

                                                                                            'l S.By'1etter dated October 4, 1994,fa proposed exigent'.

technical specification amendment. request was' submitted to the NRC for review and> approval. The proposed changes ( provided by the October 4,:1994 submittal was to add a ~

                                                                                            'l provicion to the technical specification ~to allow for the
                                    ~

swapping of:NV pumps while in modes 4,5, and 6. NRC [ approval of?the proposed license ~ amendment submitted by the-

October 4, 1994l Duke Power letter was'provided by an,NRC

         ' letter dated November 1 7 , '1 9 9 4 ~.         The October 4, 1994-submittal provided a proposed revision to;the footnote of Specification 3.5.3. The proposed amendments provided by                            -

this submittal, also, proposes changes'to the footnote of-Specification 3.5.3. ; f P I h s i t

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ML20082B396

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