Text

Proposed Tech Specs,Providing Revised Record Pages for LTOP Protection in TS Amend
	ML20098A463
Person / Time
Site:	McGuire, Mcguire
Issue date:	09/18/1995
From:	; DUKE POWER CO.
To:	;
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ML20098A461	List: ML20098A459; ML20098A463;
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	NUDOCS 9509260021
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS-Reactor Coolant System .................. 3/4 4-30 FIGURE 3.4-2 UNIT 1 REACTOR COOLANT SYSTEM l HEATUP LIMITATIONS - APPLICABLE UP T0 16 EFPY . . . . . .-. . . . . . . . . . . 3/4 4-31 l-FIGURE 3.4-3 UNIT 2 REACTOR COOLANT SYSTEM l HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY . . . . . . . . . . . . . . . . . . 3/4 4-32 l

- FIGURE 3.4-4 UNIT 1 REACTOR COOLANT SYSTEM l C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY . . . . . . . . . . . . . . . . . . 3/4 4-33 l FIGURE 3.4-5 UNIT 2 REACTOR COOLANT SYSTEM I l

C00LDOWN LIMITATIONS - APPLICABLE :

UP TO 16 EFPY . . . . . . . . . . . . . . . . . . 3/4 4-34 l j TABLE 4.4-5 [ DELETED] ...................... 3/4 4-35 P re s s u ri z e r . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-36 Overpressure Protection Systems . . . . . . . . . . . . . . 3/4 4-37 i 3/4.4.10 STRUCTURAL INTEGRITY ................... 3/4 4-40 l' 3/4.4.11 REACTOR VESSEL HEAD VENT SYSTEM . . . . . . . . . . . . . . 3/4 4-41 l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection . . . . . . . . . . . . . . . . . . . . 3/4 5-1

[ Del eted] . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - Tavg 2: 350*F . . . . . . . . . . . . . . 3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350*F . . . . . . . . . . . . . . 3/4 5-9 l 3/4.5.4 [ Deleted] . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-11 3/4.5.5 REFUELING WATER STORAGE TANK ............... 3/4 5-12 t

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

Amendment No. (Unit 2) 9509260021 950918 PDR ADOCK 05000369 A - ._ PDR

l 1AATERIAL PROPERTY BATl3 Figure 3.4-2 LIMITING MATERIALS: LOWER SHELL LONGITUDINAL WELDS 3-442A and LOWER SHELL PLATE B5013-2 LIMITING kRT AT 16 EFPY:

1/4-1,149.5 deg. F 3/4-t,102.0 deg. F 2500 ,

I n n LEAK TEST LIMIT --[ [ [

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-' BASED ON INSERVICE 500 --- HYDROSTATIC TEST TEMPERATURE (282 F)

FoR THE SERVICE PERIOD UP TO 16 EFPY 0

0 100 200 300 400 500 Reactor Beltline Region Fluid Temperature (Deg. F)

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.

1

. MATERIAL PROPERTY BA!!S Figure 3.4-3 LIMITING MATERIALS: LOWER SHELL FORGING 04 LIMITING ART AT 16 EFPY:

1/4-1,104 deg F 3/4-1, 73 deg F 2500 ; i i i, _ _

s s s sJ J J LEAK TEST LIMIT J 2250 F

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500 TEMPERATURE (236 F) FOR THE j SERVICE PERIOD UP TO 16 1 EFPY 250 l 1

0 0 100 200 300 400 500 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.

MATERIAL PROPERTY BASIS: Figure 3.4-4

~-

,, s .

i LIMITING MATERIALS: l l

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' LOW"R SHELL LONGITUDINAL WELDS 3-442A and '

LOWER SHELL PLATE B5013-2

- LIMITING ART AT 16 EFPY: l

,1/4-t,149.5 deg. F- i 3/4-t, - 102.0 deg. F . :

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~2500 , , , , , , , ,

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250 --

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0 100 200 300 400 500 !

Reactor Beltline Region Fluid Temperature (Deg. F)

McGuire Unit 1 RCS Cooldown Limitations, -

Cooldown Rates up to 100 deg. F/HR (Without Margins for instrumenation 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. l t

i Figura 3.4-5

., MATERIAL PROPERTY BASIS LIMITING MATERIALS: LOWER SHELL FORGING 04 LIMITING ART AT 16 EFPY:

1/4-t',104 deg F 3/4-t, 73 deg F 2500 ,

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l l I l I l 2250 7 J

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I Reactor Beltline Region Fluid Temperature (DEG. F) i McGuire Unit 2 RCS Cooldown Limitations, :

Cooldown Rates up to 100 deg F/HR (without Margins for Instrumentation Errors)

NRC REG GUIDE 1.99, Rev. 2 Applicable for the First 16 EFPY ,

Amendment No.

. Figure 3.4-5 McGuire - Units 1 and 2 3/4 4-34

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS.

LIMITING CONDITION FOR OPERATION 3.4.9.3 As a minimum, a Low Temperature Overpressure Protection (LTOP) System shall be OPERABLE as follows:

a. A maximum of one Centrifugal Charging (NV) pump or 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 s 385 psig OB

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

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or 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 s 15 Minutes.

One Safety Injection pump and one Charging pump, or two Charging pumps may be 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.

OR

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 1

' REACTOR COOLANT SYSTEM:

LIMITING CONDITION FOR OPERATION (Continued)

ACTION:-(continued) b.: ~ With an accumulator not isolated, isolated the affected accumulator within I hour. 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 , l

E a l

2. Increase RCS cold leg temperature to greater than or equal to 300* F !

within-12 hours. -l

c. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within_7 days. If required action is not met, l i

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 .' l

?

With one PORV inoperable in MODES 5 or 6, suspend all operations which j d.

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. j E 4 l

Depressurize the RCS and vent through at least a 2.75 square' inch l 2.

vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . I

e. With the LTOP system inoperable for any reason other than a., b., c., or ;

j

d. above, depressurize the RCS and vent through a least a 2.75 square inch j

vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

i I f.- In the event that either the PORVs or the RCS vent are used to mitigate l and RCS pressure transient, a Special Report shall be prepared and r submitted to the Commission pursuant to Specification 6.9.2 within 30 l days. The report shall describe the circumstance initiating the i' 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.

i i 1

4 McGUIRE - UNITS _1:and_2 3/4 4-38 Amendment No. Unit 1 '

Amendment No. Unit 2 i

j

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 the PORV actuation channel at least once per 18 months; and

c. Verifying the 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 when the PORV is being used for overpressure protection.

4.4.9.3.2 Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months *, verify that an RCS vent of 2: 2.75 square inches is open when the vent is used for overpressure protection.

4.4.9.3.3 Once every .12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , verify that each accumulator is isolated and that l only one NV or NI pump is capable of injecting into the RCS.

4.4.9.3.4 Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , verify that RHR suction isolation valves ND-1 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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , verify that the PORV block valve is open for each required PORV.

I

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

A PORV secured in the open position may be used to meet this vent requirement provided that its associated block valve is open and power is removed.

McGUIRE - UNITS 1 and 2 3/4 4-39 Amendment No. Unit 1 Amendment No. Unit 2

REACTOR COOLANT SYSTEM

~

3/4.4.10 ' STRUCTURAL INTEGRITY ,

i LIMITING CONDITION FOR OPERATION 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.

APPLICABILITY: 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 minimum temperature required by NOT ,

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 System temperature above 200*F.

l

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.

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.

l 1

, I McGUIRE.-' UNITS 1 and 2 3/4 4-40 .

Amendment No. (Unit 1) ~ l Amendment No. (Unit 2) !

l

REACTOR COOLANT SYSTEM 3/4.4.11 ' REACTOR VESSEL HEAD VENT SYSTEM LIMITING CONDITION FOR OPERATION 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.

APPLICABILITY: 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 vent path to OPERABLE status within 30 days or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in l

COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 OPERABLE. status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

4

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.

I l

l McGUIRE - UNITS 1 and 2 3/4 4-41 Amendment No. Unit 1 ~l Amendment No. Unit 2

l EMERGENCY CORE COOLING SYSTEMS 3/4.5.3' ECCS SUBSYSTEMS - T , < 350*F j l

LIMITING CONDITION FOR OPERATION j l

3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be 1 OPERABLE: l

a. One OPERABLE centrifugal charging pump,# l l

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. <

APPLICABILITY: MODE 4.

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump' or the flow path from the refueling water i storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

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 T,,less than 350*F !

by use of alternate heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the -

Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles ,

to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its !

value exceeds 0.70.

l r

i t

A maximum of one centrifugal charging pump and one Safety Injection pump shall be capable of injecting into the RCS whenever the temperature of one or more of l the RCS cold legs is less than or equal to 300*F. Two charging pumps may be operable and operating for s 15 minutes to allow swapping charging pumps.

Additional requirements are provided by Specification 3.4.9.3. l McGUIRE - UNIT 1 and 2 3/4 5-9 Amendment No. (Unit 1) l Amendment No. (Unit 2) ;

l

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.

4.5.3.2 All centrifugal charging pumps and Safety Injection pumps, not capable of injecting into the RCS shall be demonstrated inoperable by verifying that the motor circuit breakers are secured in the open position or by verifying the discharge of cach 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the RCS cold legs is less than or equal to 300*F.

McGUIRE - UNIT 1 and 2 3/4 5-10 Amendment No. Unit 1 ~

Amendment No. Unit 2

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

i l

.- 3 REACTOR COOLANT SYSTEM j i

BASES- i PRESSURE / TEMPERATURE LIMITS (Continued) l 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 properties are then evaluated in accordance with Appendix G of the 1989 Edition ;

of Section XI of the ASME Boiler and Pressure Vessel Code and the calculation l

. methods described in WCAP-7924-A, " Basis for Heatup'and Cooldown Limit Curves, !

April 1975."

Heatu) and cooldown limit curves are calculated using the most-limiting i value of t1e nil-ductility reference temperature, RTuor, at the end of the 7 effective full power years (EFPY) of service life identified on the applicable !

technical specification figure.- The 16 EFPY service life period continues to ensure that the limiting RT or at the 1/47 location in the core region is a bounding'value. The select $on of such a limiting RTuor assures that all- !

. components in the Reactor Coolant System will be operated conservatively in !

accordance with applicable Code requirements. l The reactor vessel materials have been tested to determine their initial RTuor; the results of these tests are shown in Table B 3/4.4-1. Reactor opera- l tion and resultant fast neutron'(E greater than 1 MeV) irradiation can cause an !

increase'in the RTuor. Therefore, an adjusted reference temperature, based ;

upon the fluence, copper content, and phosphate content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of l For Unit 1, the adjusted reference temperature has been computed by MTul7Regu atory Guide 1.99, Revision The 2.

heatup and cooldown limit curves of l 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 the identified service life. Adjustments for possible errors in the pressure and temperature sensing instruments are included when stated on the applicable figure.

) Values of MTuor 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 ,

l' ASTM E185-73 and 10 CFR 50, Appendix H. The lead factor represents the rela- !

I 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- 1 drawal time of the capsule. The heatup and cooldown curves must be recalcu- I lated when the MTuor determined from the surveillance capsule exceeds the .

calculated MTuor for the equivalent capsule radiation exposure. l l

t 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.

f McGUIRE - UNITS.1 AND 2 8 3/4 4-8 Amendment No. (Unit 1) ~

. Amendment No. (Unit 2) l l

-. al- , - e - ,-e.-t wm-- e+ - - - - - -

er-'- we - + + - e- - r e---:-u:'a '

REACTOR COOLANT SYSTEM BASES-PRESSURE / TEMPERATURE LIMITS (Continued) end of the transient, conditions may exist such that the effects of compressive thermal stresses and different Km's for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-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 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 de)endent on both the rate of heatup and the time (or coolant temperature) along tie heatup ramp.

Furthermore, since the thermal 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.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as fo.llows. A composite curve is constructed based on a point-by-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.

Finally, the composite curves in technical specifications for the heatup rate data and the cooldown rate data may be adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. Where technical specification curves have not been adjusted, such adjustments are made by plant procedures.

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.

The OPERABILITY of two PORVs or an RCS vent opening of at least 2.75 square l 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 McGUIRE - UNITS 1 AND 2 B 3/4 4-16 Amendment No. Unit 1 ~

Amendment No. Unit 2

e REACTOR COOLANT SYSTEM BASES. I PRESSURE / TEMPERATURE LIMITS (Continued) the RCS cold legs are less than or equal to 300*F. Either of the PORVs or the RCS vent opening has adequate 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 RCS cold leg temperatures, or (2) the start of a HPSI !

pump and its injection into a. water-solid RCS. The Pressurizer PORV setpoints j 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.

l Credit is taken for the RhR suction relief valve (ND-3) during conditions where i relieving ca)acity at rated accumulation is sufficient to prevent exceeding the i above allowa)le pressure limits. ;

Cooldown limits / minimum RCS temperature restrictions ensure the allowable' pressure limits will not be exceeded.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 l 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 I 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 core cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function. (Operability of the pressurizer steam space vent path is provided by Specifications 3/4.4.4 and 3/4.4.9.3.) {

i l

The valve redundancy of the reactor coolant system vent paths serves to j minimize the probability of inadvertent or irreversible actuation while ensuring '

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 qualitative 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.

McGUIRE - UNITS 1 AND 2 B 3/4 4-17 Amendment No. Unit 1 ~

Amendment No. Unit 2

ML20098A463

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