ML20086G043

Proposed Tech Spec 3/4.4.4,3/4.4.9 & 3/4.4 Re PORV & Block Valve Reliability

ML20086G043
Person / Time
Site:	Mcguire, McGuire
Issue date:	11/21/1991
From:	DUKE POWER CO.
To:
Shared Package
ML20086G040	List: ML20086G036 ML20086G043
References
NUDOCS 9112040148
Download: ML20086G043 (26)

Text

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~~ ATTACHMENT 1 Proposed McGuire Unit 1 and 2 Technical Specificatione Changes e..

l 9112040148 911121

l. PDR ADOCK 05000369

.P 999

Proposed McGuire PORV/LTOP Technical Specifications

'4 REACTOR COOLANT SYSTD1 3/4.4.* RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valvue chall be OPERADLE.

A,P_PLICABI LITY : MODES 1, 2, and 3.

ACTION:

a. -With one or more PORV(s) inoperable because of excesolve leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and maintain power to the block valve (s); otherwise, bo in at least HOT STANDB'f within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. With one PORV inoperable due to causeu other than excessive leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block vals3; otherwise, be in at least HOT STANDBY within ths next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. Witn two PORis inoperable due to causes other than excessive leakage, with;.n 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORVa to OPERABLE status or-close the associated block valves and remove power from the block valves; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the block valves have been closed and power has been removed, restore at least one PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hcurs.

d. With three PORVs inoperable due to causes other than excessive leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least une PORV to OPERABLE status or close the associated block valves and remove power frcm the block valves and bo in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

e. With one block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve to OPERABLE status or place its associated PORV McGUIRE - UNITF 1 and 2 3/4 4-10 Amendment No. (Unit 1)

Amendment No. (Unit 2)

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• REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES (continued)

LIMITING CONDITION FOR OPERATION switch in the "close" position and remove power f rom its associated solenoid valve (do not enter action statement b for the resulting inoperable PORV); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

f. With two block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valves to OPERABLE status or place their associated PORV switches in the "close" position (do not enter action statement c for the resulting inoperable PORVs); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the PORV switches have been placed in the "close" position, restore at least one block

- valve to OPERABLE statun within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

g. With three block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valves to OPERABLE status or place their associated PORV switches in the "close" position (do not enter action statement d for the resulting inoperable PORVs). Restore at least one block valve to OPERABLE status within the next hour; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

h. The provisions of Specification 3.0.4 are not applicable.

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GURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

a. Performance of a CHANNEL CALIBRATION, and

b. Operating the valve through one complete cycle of full travel during MODE 3 or MODE 4 when the temperature of all RCS cold McGUIRE - UNITS 1 and 2 3/4 4-10a Amendment No. (Unit 1)

Amendment No. (Unit 2)

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REACTOR COOLANT SYSTEM L 3/4.4.4 ~ RELIEF VALVES (continued)

SURVEILLANCE . REQUIREMENTS legs is greater.than 300F with the block valvo closed.

4.4.4.2- Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through or.a complete cycle of full-travel unless the. block valve in closed in order to meet the requiremente of ACTION a., b., c., or d..in Specification 3.4.4.

4.4.4.3 The emergency power supply for the PORVs shall be demonstrated

.OPERABtE at least once per 18 months bys

a. -Manually transferring motive power from the normal (air) oupply to the emergency (nitrogen) supply,

b. Isolating'and venting the normal (air) supply, and

-. c. Operating the valves through-a complete cycle of full travel.

r McGUIRE - UNITS 1 and 2 3/4 4-10b Amendment No. (Unit 1)

Amendment No. (Unit 2)

REAC"OR COOLANT SYSTEM Oy:RPRESSURE PRO'ECTION SYSTEMS LIMITING CONDITION FOR OPERATION

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h 3.4.9.3 At least one of the following Overpressure Protection Systema chall be OPERABLE:

a. Two power-operarq4 relief valven (PORVo) with a lift setting of 1ess than or equal to 400 peig, or 3

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 4.5 square inchen.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 300F, MODE 5, and MODE 6 when the head is on the reactor vencel.

ACTION:

a. With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE atatus wi- lin 7 days or complete depronourization and venting of the RCS through at least a 4.5 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. .

b. With one PORV inoperable in MODE 5, cuppend all operations that could lead to water-solid RCS conditions. Restore the inoperable PORV to OPERABLE statun within 72 houro or complete depressurization and venting of the RCS through at least a 4.5 square inch vent (s) within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. With one PORV inoperable in MODE 6, rectore the inoperable PORV to OPERADLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or complete depreneurization and venting of the RCS through at least a 4.5 square inch vent (u) within the next B hours.

d. With both PORVs inoperable, complete depressurization and venting of the RCS through at least a 4.5 square inch vent (n) within B hours.

e. In the event either the PORVs or the RCS vent (s) are used to mitigato an RCS pressure trancient, a Special Report shall be prepared and submitted to the Commianlon purcuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the tranalent, the effect of the McGUIRE - UNITS 1 and 2 3/4 4-37 Amendment No. (Unit 1)

Amendment No. (Unit 2)

REACTOH COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS (continued)

LIMITING CONDITION FOR OPERATION PORVs or vent (s) on the transient, and any corrective action nbcepsary to prevent recurrence.

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f. The provisions of specification 3.0.4 are not applicable.

C SURVEILLANCE REQUIREMENTS

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4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by s

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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpreneure protection.

4.4.9.3.2 The RCS vent (s) shall be verified to be vi>en at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

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

• Except whnn the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify theoe valves open at least once per 31 days.

McGUIRE - UNITS 1 and 2 3/4 4-38 Amendment No. (Unit 1)

Amendment No. (Unit 2)

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n REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure-during-all design transients up to and including

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the design step load decrease with steam duttp. Each PORV has a-remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.- The OPERABILITY of the PORVs and. block valves is determined on the banis of their being capable of performing the following. functions: 1) Manual control of PORVs to control PCS pressure. This is a function tht4t is used for the steam generator tube rupture accident coincident with a loss of all-offsite power and for plant shutdown. 2) Maintaining the integrity of the reactor coolant pressure boundary. This is a functinn that is related to controlling identified leakage and ensuring ths ability to detect unidentified reactor coolant pressure boundary leakage. 3) Manual control of the block valve to unblock an isolat'ed' PORV to allow it to be used for manual-control of RCS pressure end isolate a PORV with excessive leakage. 4) Automatic control of PORVs to control RCS pressure. Tals is a function that reduces challenges to the code safety valves.for overpressurization events. 5) Manual control of a block valve to isolate a stuck-open PORV.

- 3/4.4'5 . STEAM GENERATORS (THIS SECTION REMAINS UNCHANGED.)

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

Amendment No. (Unit 2) l

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s REACT 0R COOLANT SYSTEM 3/4.4.4 RELIEF VALVES IMITING CONDITION FOR OPERATION 4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: H0 DES 1, 2, and 3.

ACTION:

a. With c.ie or more (PORV(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

b. With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s) or close the PORV and remove power from its associated solenoid valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS __ ,

4 4.4.1 4

In addition to the requirements of Specification 4,0.5, each PORV shall be Jimonstrated OPERABLE at least once per 18 months by:

a. Forformance of a CHANNEL CALIBRATION, and

b. Operating the valve through one complete cycle of full travel.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed wi'.h power removed in order to meet the requirements of ACTION a. in Specification 3.4.4.

i McGUIRE - UNITS 1 and 2 3/4 4-10 Amendment No. 92 (Unit 1)

Amendment No. 73 (Unit 2)

e REACTOR COOLANT SYSTEM OVERPRE$$URE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

a. Two power-operated relief valves (PORVs) with a lif t setting of less than or equal to 400 psig, or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 4.5 square inches.

APPLICABILITY: H00E 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 one PORV inoperable, restore the inoparable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 4.5 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,

b. With both PORVs inoperable, depressurize and vent the RCS through at least a 4.5 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient, and any corrective action necessary to prevent recurrence,

d. The provisions of Specification 3.0.4 are not applicable.

McGUIRE - UNITS 1 and 2 3/4 4-37

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I REACTOR COOLANT SYSTEM I

i SURVEILLANCE REQUIREMENTS s.

4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: i a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV.

actuation channel, but excluding valve operation, within 31 days '

prior to entering a condition in which_the PORV is required OPERA 8LE and at least once per 31 days thereaf ter when the PORV is required OPERA 8LE;

b. -

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

c. Verifying the PORY isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.

4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per ,

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" 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, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days..

l McGUIRE - UNITS 1 and 2 3/4 4-38

4 REACTOR COOLANT SYSTEM B A,5j $

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizec the undesirable opening of the spring-loaded pressurizer code safety valves.

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The BW process (or method) equivalent to the inspection method described in Topical Report BAW-2045(P)-A will be used.

Inservice inspection of steam generator sleeves 's also required to ensure RCS integrity. Because the sleeves introduce changes in the wall thickness and diameter, they reduce the sensitivity of eddy current testing, therefore, special inspection methods must be used. A method is described in Topical Report BAW 2045(P)-A with supporting validation data that d oonstrates the inspectability of the sleeve and underlying tube. As required by NRC for licensees authorized to use this repair process, McGuire commits to validate the adequacy of any system that is used for periodic inservice inspections of the sleeves, and will evaluate and, as deemed appropriate by Duke Power Company, implement testing methods as better methods are developed and validated for commercial use.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking, The extent of cracking during plant operation would be l hited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant Syst*m (reactor-to-secondary leakage = 500 gallons per day per steam generator).

Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postuleted accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdewn. Leakage in excess of this limit will require plant Amendment No .107 (Unit 1:

McGUIRE - UNITS 1 and 2 B 3/4 4-3 Amendment No. 89 (Unit 2)

1 REACTOR C00 TANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) end of the transient, conditions may exist such that the effects of compressive thermal stresses and different Kgg's for steady state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state o nditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/47 flaw is consicered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value heatup rates of is the allowable pressure calculated for steady-state and finite 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 dependent on both the ramp.

rate of heatup and the time (or coolant temperature) along the heatup 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 4nalyzed 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 follows.

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 De 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 4.5 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 Amendment No. 82 (Unit 2)

McGUIRE - UNITS 1 and 2 8 3/4 4-16 Amencment No.100 (Unit 1)

l REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) the RCS cold legs are less than or equal to 300'F. Either PORV has adequate relieving capability to protect the RCS f rom 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 HPS! pump and its injection into a water-solid RCS.

3/4.a.10 STRUCTURAL INTEGRITY 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 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.)

The valve redundancy of the reactor coolant system vent paths serves to minimize the- probability of inadvertent or irraversible 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.

i 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 11,B,1 of NUREG-0737, " Clarification of TMI Action Plan Requirements" November 1980.

l McGUIRE - UNITS 1 and 2 B 3/4 4-17 Amendment No. 76 (Unit 1)

Amendment No. 57 (Unit 2)

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i ATTACHMENT 2

- Justification and Safety Analyniu 1

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4 Background / Justifications-On Juno 25, 1990, t he NRC issu9d Generic Letter 90+06, Repolution of Generic immuo 70, " Power-operated Relief Valve and Block Valve En11 ability", and Gonoric Issue-94, " Additional Low-Temperature overprosauro Protection for Light-Water Reactors", Pursuant to 10 CFR

$0.64(f). This genoric letter provided staf f positions regarding actions which phould be taken to improvo the ro11 ability of PORVa and block valves, specifically:

1) Including PORVs and block valvos in the quality assurdnce program per 10 CFR t,0, Appendix D)

2) Including PORVa, valvos in PoHV control air systems, and block valves within the scope of the inservice testing program; and l

3) Modifying the technical specifications for the P0HVs and LTOP systomo. ,

Duko Power Company's responeo to the generic letter, dated Ducomber 20,  !

1990, addrescod itoms 1 and 2 above. The purpose of this technical i specification amendment submittal is to addrena Atom J. I t

Danen/satoty Anaj nin The NRC staff has recently determinod that over thu years, the role of t PORVo has changed such that PORVn are now rollod upon by many planta to perform one or more of the following cafety-related functions:

1)' Mitigation of a design baulo steam generator tube rupture accident,  ;

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2) Low-temperature overpronoure protection of the reactor vessel during  ;

startup and uhutdown, and

3) Plant cooldown in compliance with Branch Technical Ponition RSB 5-1  ;

to SRP 5.4.7, " Realdual Heat Removal (RHR) System". ,

At McGuiro, the PORVs atu utilized for-functions 1 and 2 above.

The proposod;tochnical npocificatione changea are being submitted in accordance with.the guidanco provided by the NRC in its resolution of

Generic Issue 70,'"PORV and Dlock Valve Reliability" and Generic Inpun 94, " Additional Low-Temperature Overpressure Protection for Light-Water Reactora". Thoso changes are conalatent.with the McGuire design and safety analynou. The epocific changes are discupeed individually in the next section.. Whore deviation from the guidance provided by the NHC oxinta, justification for the deviation in provided.

Regarding Generic Innue 94, liceneces were requented to verify whether certain administrative restrictione concerning the restart of inactive

• I reactor coolar.t pumps and concerning the operability _ of high prosaure -

safety injection pumps have. boon implemented. 'These restrictions were imposed as a result of Unrosolved Safety Iseue (USI) A-26, "Roactor Vessel Proosure Transient Protection (Overpressure Protection)". Thepo administrative restrictiono have been previously implomonted at McGuire.

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Description of Proposed Technical Specifications changes:

For this discussion, refer to the new proposed technical specifications l for the PORVs and LTOP systems provided in Attachment 1. Also included for comparison are the existing PORV/LTOP technical specifications for - t McGuire.

Limiting condition for operation 3.4.4

- Action statement a. becomes applicable only in the case of PORV inoperability due to excessive leakage. While Generic Letter 90-06 only refers to seat leakage, the new action statement a. simply refore to leakage in general, so that both seat leakage and packing leakage would constitute PORV inoperability under this action statement. The phrase dealing with the removal of power from the block valve (s) is changed to require power to be maintained to the block valve (s). This ensures that the valve (s) remain operable and capable of being opened to allow the PORV(s) to be used to control primary system pressure.

In addition, the action statement has been modified to terminate the ,

forced shutdown requirement in hot shutdown rather than cold shutdown, since the LCO only applies to Modes 1, 2, and 3.

- Action statements b., c., and d. govern the cases of PORV inoperability due to causes other than excessive leakage for the cases of one, two, or three inoperable PORVs, respectively. Action statement b. allows continued plant operation with one inoperable PORV, provided its block valve is closed and power is removed from the block valve. The basis for this position is that following a steam generator tube rupture with loss of offsite power, only one PORV is required to depressurite the Reactor Coolant System. All three of McGuire's PORVs have nitrogen br.ekup capability to cope with a loss of instrument air to the valves. Therefore, even with one PORV inoperable for an extended period of time, redundant capability exists to depressurite the Reactor Coolant System following a SGTR event.

- Action statements e., f., and g. govern block valve inoperability for the cases of one, two, or three inoperable block valves, respectively.

These action statements establish remedial measures consistent with the function of the block valves. The prime importance for the <

capability to close the block valve is to jeolate a stuck-open PORV.

If the block valve (s) cannot be restored to operable status within one hour, the remedial action is to place the associated PORV switch (es) in the "close" position to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck-open PORV at a time t hat the block valve is inoperable. (The guidance contained in

• the generic letter states to place the PORV in manual control; however,.McGuire's PORV control switches are labeled "open", " close",

and " auto", so the proposed change is consistent with the McGuire design.) Since the proposed technical specifications do not require the plant to be shut down for the case of cne inoperable PORV or one inoperable block valve, action statement e. includes the additional requirement to remove power from the associated PORV solenoid valve.

This provides additional assurance that the PORV with the inoperable block valve will not be inadvertently opened at a later date, because if the block valve is not capable of being closed, an unisolable leak path would be created if the PORV were to fall open. The requirement

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- 1 to remove power from the PORV solenoid valves is not neceesary for action statements f. and g. since these action statements possess time limitations for restoring the block valves to operability. The time allowed to restore the block valve (s) to operable status is based upon the remedial action time limits for inoperable PORVs per action statements b., c., and d. since the PORVs are not capable of mitigating an overpressure event when the PORV switches are in the "close" position. The same basis proposed for allowing continued plant operation with one inoperable PORV in action statement b. also I applies to action statement e. Finally, action statements e., f., and

g. include a provision in parentheses which precludes entering action statements b., c., and d., respectively for inoperable PORVs once the l PORV switch (as) have been placed in the "close" position. This is necessary because placing the PORV switch (ou) in the "close" position renders the PORV(s) inoperable and the resulting requirements for inoperable PORV(s) in action statements b. c. , and d. (i.u., close the block valve (s) and remove power) are inappropriate since the block valve (s) are already inoperable under action statements e., f., and g.

- Old action statement c. becomes new action statement h.

Surveillance Requirement 4.4.4.1

- The generic letter recommends that PORVs be stroked during Mode 3 or Mode 4 in order to accurately simulate environmental effects on the valves. The basis for this reconnendation is that testing during Modo 5 may not be a representative test for assessing PORV performance under normal plant operating conditions. Surveillance Requirement 4.4.4.1.b has therefore been modified to require that the PORVs be cycled during Moden 3 or 4. It clarifies that the PORVs should only be cycled during that part of Mode 4 which is not in the LTOP regime and further specifies that the bicek valve must be closud prior to stroking the PORV to preclude the possibility of a loss of coolant should the PORV fail open with the block valve open. (The block valve would be closed just prior to stroking the PORV so that environmental conditions on the PORV are maintained.)

surveillance Requirement 4.4.4.2

- This surveillance requirement exempts the block valve from having to be demonstrated operable when it is closed to isolate an inoperable PORV. The exemption is maintained for the case when a block valve is closed with power maintained to isolate a PORV which is inoperable due to excessive leakage. This is to prevent having to open the block valve and discharging large quantities of effluent to the pressurizer relief tank in the case of a severely-leaking PORV. If the block valve is closed with power removed to isolate an otherwise inoperable.

PORV (the proposed technical specification allows McGuire to operate with one PORV/ block valve inoperable for an extended period of time),

the block valve would not need to be demonstrated operable because redundant accident mitigation capability would be provided by the two remaining PORVs/ block valves.

Surveillance Requirement 4.4.4.3

- Surveillance Requi.rement 4.4.4.3 has been added which requires

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4 demonstrating the pORV emergency power supply operable. It requires the valve motive power to be transferred from its normal instrument air supply to its emergency nitrogen supply, the air supply to be isolated and vented, and the pORV to be operated through a complete cycle of full travel. The generic letter also recommends that the surveillance requirements require operating solenoid air control salves and check valves on air accumulators in PORY control systems through a complete cycle of full travel. Surveillance Requirement 4.4.4.3 requires the PpHVs to be stroked while aligned to the emergency nitrogen supply, with the normal air supply vented; this ,

cycles the necessary valves. It is therefore not necessary to expand the surveillance requirements any further in this regard. Also, the guidance contained in the generic letter indicates that motive and control power for the p0RVe and block valves should be manually transferred from the normal to the emergency power bus. At McGuire, transferring control power does not apply because the poRVs and block valves are normally powered from an essential bus. Trasaferring motive power does not apply to the block valves; they are not pneumatic. Hence, the McGuire specification., as currently modified, enmplies with the guidance contained in the generic letter.

Limiting Condition for operation 3.4.9.3

- In LCO 3.4.9.3, the applicability is modified to change the phrase "with the reactor vessel head on" to "when the head is on the reactor vessel". It should be noted that in the guidance contained in the generic letter, it is suggested that the applicability of the LCO be clarified to exclude Mode 6 when the Reactor Coolant System is adequately vented and that the depressurizing and venting of the Reactor Coolant System should no longer be claesified as an ,

overpressure protection system. The generic letter recommends that an additional action statement be_added to specify verifying the vent pathway when the Reactor Coolant System is depressurized and vented.

This proposed structure appears inappropriate, because once the Reactor Coolant System is vented, LCO 3.4.9.3 would no longer apply and the action statement requiring verification of the vent pathway would therefore not have to be entered. For this reason, it is being proposed that the present structure of the McGuire Technical Specifications be maintained in that the depressurizing and venting of the Reactor Coolant System will continue to be classified as an overpressure protection system and the requirement to verify the vont pathway when the system is depressurized and vented will continue to be governed by Surveillance Requirement 4.4.9.3.2.

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j. - Action statement a. is modified to clarify that it is only applicable j in Mode 4. Also, the phrase "depressurize and vent" has been changed to "completo depressurization and venting of" to avoid any possible question as to when the required depressurizstion and venting action must be completed. (It also provides for consistency in wording throughout the LCo.)

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- A new action statement b. is added which prohibits entering a water-solid condition in the Reactor Coolant System when one PORV s inoperable while in Mode 5. An allowable outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is proposed for one PORV while in Mode 5. When the Reactor Coolant System is not water-solid, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> it viewed as a reasonable l

compromiso betwoon the 7-day allowable outage timo for Mode 4 and the ,

24-hour allowable outage time for Modo 6 (see below). Also, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

- providos sufficient time for completing crud burst activities. A 24-hour time limit is proposed for completing depropourization and venting activities for thin action statement in order to allow for a more controlled depresourization than could be accomplished in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

- A new action statomont c. is added to reduce the allowable uutage time for one inoperable PORV to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while in Mode 6.

- Old action statement b. becomes now action statomont d. and the phrase "dopressurizo and vent" has been changed to "completo depronourization and vencing of".

- Old action ntatement c. becomes new action statement u.

- Old action statement d. becomen new action utatement f.

Surveillance Requirement 4.4.9.3.1

- Survo111ance Requiroment 4.4.9.3.1.a. is nimpliflod by removing requiremontu that oxist because of gonoral requiromonts applicable to all ourveillanco requiremento an epocified in Section 4.0 of the Technical Specifications.

Bases Section 3/4.4.4

- The bacon for the PORV and block valve technical specifications have been expanded to identify the major functions of the PORVs and block valves. Those major functions are au follows:

1) Hanual control of Reactor Coolant System pronouro following accidents,

2) daintaining reactor coolant pressure boundary integrity by controlling leakage,

3) Hanual control of block valves to isolate and unblock PORVa (for manual prosaure control and for controlling PORV leakage),

4) Automatic control of Reactor coolant system prosaura to provent code anfety valve challenges, and

5) Hanual control of block valves to isolato a etuck-open PORV.

The expandud banen more clearly delineate those functior.a.

Conclusions:

The above changen to the McGuire Technical Specifications are being submitted in accordance with NRC guidanco contained in Generic Letter 90-06. Those changes impose certain additional restrictiona concerning the operability of the PORVs and block valves and therefore will result in an overall improvement in safety. Appropriate technical

justification has been provided for the cases where deviation from the generic letter's guidanco exists.

Duhe Power company has concluded that the proposed amendments will not be Animical to the health and safety of company personnel or to the public.

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ATTACitMr.NT 3 Analysis of Significant,llagarde consideration

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n Analysis of significant Hazards considerations Duio Power Company han made the determination that this amendment request involves a no significant hazards consideration by applying the standards eriablished by the Commission's regulation in 10CFR 50.92.

This ensures that operation of the f acility in accordance with the proposed amendment would noti (1) ~ 1nvolvo a significant-Ancreaso in the probability or consequences of an accident previously evaluated; or (2) creatu the tossibility of a new or different kind of accident from any accident previously evaluatod; or (3) involvo e significant reduction in a margin of safety.

- The Commission has provided guidelinos pertaining to the application of the throu standards by listing specific examples in 40FR14070. Example (ii) relatos to a chango that constitutes an additional limitation, restriction, or control not presently included in the technical  ;

specifications.

In this case the changes proposed by this request are similar to Example ,

(ii) in that +hoy result from NRC guidanco contained in Genoric Letter l 90-06. The changes connist mainly of additional restrictions in the limiting conditions for operation for the PORVs and block valvos (and l LTOP systems when the PORVs are used for LTOP), as well an expanded survn111ance requirements. l The following evaluation measures aspects of this proposal against the Part 50.92(c) requirements to demonotrate that all three standards are satisfied.

First standard The amendment would not involve a significant increase in the-probability or consequences of an accident previously evaluated, i

From an accident and transient mitigation standpoint, the PORVs can bo i utilized ~to gerform safety-related functions. For McGuire, theno i include mitigation of a steam generator tube rupture accident and low-temperature overpressure protection of the reactor vessel. In addition, the PORVs can be an accident initiator in the case where a.

failed-open PORV results in a small break loss of coolant accident (SDLOCA).

The proposed changes will increase the likelihood that the PORVs and block valves will be available for performing the above safety-related

' functions. They will also reduco the probability of accident sequences which may result from PORV and block valvo failures. For the case where 4 the PORVs are utilized as a means of LTOP, the proposed changes (specifically, the reduced allowable outage time while in Ms.'s 5 and 6) will ensure that both. channels of overpressure protection equipment maintain a high availability in the event that they are called upon to mitigate a low-temperature overpressurization transient.  !

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Dased on the above, the proposed -technical elecifications changes will not involve a significant increase in the probability or consequences of an accident that has been previously evaluated.

Second St andard The amendment would not create the possibility of a now or dif ferent kind of accident from any kind of accident previously evaluated.

With the except ion of a f allod-open PORV resulting in a SDLocA as

. indicated above, the PORVs and block valves cannot initiate accident sequences. The proposed changes will not result in the PORVs and block valves.telng operated or utilized in a deletorious manner; therefore,

. the possibility of a new or different kind of accident is not creatod.

L j T ird Standard The amendment would not involvo a significant reduction in a margin of

. safety.

Overall plant safety would be enhanced as a result of the additional l restrictions placed on the p0RVs and block valves (and LTOP channels when the PORVs are used for LTOP). In addition, for 01-70 and 01-94, the NRC has made the determination that there will be a substantial increase in overall protection of the public health and safety as a ,

result of the implomontation of actions recommendud by Generic Letter  !

90-06, including the recommanded technical specifications changes. .

Bksed on the above and the supporting technical. justification, Duke Power company has concluded that there is no significant hazard consideration involved in this amendment request.

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Document Control Desk Pago 4 -f Noventber 21, 1991  ;

l bxct (W/Attachmente)

A.V. Carr R.C. Futrell W.H. Sample ['

H.L. Gill, Jr.

P.F. ouill

- L.J. Rudy.

D.V. Ethington P.J. North i

- P.R. Herran D.W. Hurdock R.W. Revels J.J. Nolin

. - J.W. Dyrd, Jr.

T.L. McConnell - HNS 0.D. Gilbert - HHS W.O. Rooside - MNS R.O. Sharpe - MNS T.S. Barr - TTC -

QA Tech. Sve. NRC Coord.

Filos. Mc-815.07 (GL 90-06)

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