Text

Amends 218 & 199 to Licenses NPF-4 & NPF-7,respectively, Changing NAPS TS to Provide Allowed Outage Time of 144 Days for PORV Nitrogen Accumulators
	ML20207H438
Person / Time
Site:	North Anna
Issue date:	03/02/1999
From:	Berkow H; NRC (Affiliation Not Assigned)
To:	;
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ML20207H442	List: ML20207H438; ML20207H445;
References
	NUDOCS 9903120327
	Download: ML20207H438 (22)
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UNITED STATES p~.

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

'+9.....,o VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 218 License No. NPF-4 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated October 25,1995, as supplemented February 5,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission 3

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amended'by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:

l; (2), Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 218. are hereby incorporated in the license. The licensee ~shall operate the facility in accordance with the Technical Specifications.

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~ This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION 1

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e rt N. Be ow, Director

roject Directorate 11-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation Attachments:

8 Changes to the Technical Specifications i

Date of issuance: March 2,1999 l

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l ATTACHMENT TO LICENSE AMENDMENT NO. 218 I

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TO FACILITY OPERATING LICENSE NO. NPF-4 ~

DOCKET NO. 50-338 i

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed i

pages as indicated. The revised pages are identified by amendment number and contain l

' vertical lines indicating the areas of change.

l Remove Paaes

- Insert Paaes 3/4 4-7a 3/4 4-7a 3/4 4-7b 3/4 4-7b

- 3/4 4-32 3/4 4-32 B 3/4 4-2 9 3/4 4-2 B 3/4 4-2a B 3/4 4-?a B 3/4 4-2b B 3/4 4-8 B 3/4 4-8 6-14 6-14 l

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i REACTOR COOLANT SYSTEJ_1 t

I SAFETY AND RELIEF VALVES - OPERATING RELIEF VdLVES l

LIMITING CONDITION FOR OPERATION 3.4.3.2

Both power-operated relief valves (PORVs) and their associated block valves sha!' be OPERABLE

' APPLICABILITY:

MODES 1; 2, and 3.

ACTION:

. A. PORV(s):

1. With one or both PORV(s) inoperable solely because of excessive seat leakage, within

-J l hour either.estore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least.:wT

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STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6

. hours.

2. (Risk Informed) With one or both PORV(s) inoperable because of (an) inoperable backup nitrogen supply (ies), within 14 days either restore the PORV(s) backup

- nitrogen supply (ies) to OPERABLE status or be in HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the follow < 6 hourt

3. With one or both PORV(s) inoperable due to a malfunction in the PORV automatic

' control system, within I hour restore the affected automatic control system (s)'to OPERABLE status or place and maintain the affected PORV(s) m manual control.

-4.

With one PORV inoperable On

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. sther than those addressed in ACTIONS A.I.

A.2 or A.3 above, within I hota either restore the PORV to OPERABLE status or close i

its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the fellowing 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within i

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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5. With both PORVs inoperable such that ACTIONS A.1, A.2 or A.3 above do not apply,

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within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one PORV to OPERABLE status or close the

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associated block valves and remove power from the block valves and be in HOT STANDRY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

6. The provisions of Specification 3.0.4 are not applicable.

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NORTH ANNA - UNIT 1

. 3/4 4-7a Amendment No. 32,189, 218

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REACTOR COOLANT SYSTEM SAFETY AND RELIEF VALVES - OPERATING RELIEF VALVES LIMITING CONDITION FOR OPERATION ACTION: (Continued)

B. Block Valves:

1.

With one block valve inoperable, within I hour either restore the block valve to OPERABLE status or place its associated PORV in manual control; restore the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.

With both block valves inoperable, within I hour either restore the block valves to OPERABLE status or place the PORVs in manual control; restore at least one block valve to OPERABLE status within the next hour; restore the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1

4.4.3.2.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE:

a.

At least once per 31 days by perfomiing a CHANNEL FUNCTIONAL TEST, excluding valve operation, and b.

At least once per 18 months by:

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Operating tne PORV through one complete cycle of full travel during MODES 3 or 4, and 2.

Operating the solenoid air control valves and check valves on the associated accumulators in the PORV control systems through one complete cycle of full travel, and 3.

Performing a CHANNEL CALIBRATION of the actuation instrumentation.

c.

At least once per 7 days by verifying that the pressure in the PORV nitrogen accumulators is greater than the surveillance limit.

4.4.3.2.2 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 in order to meet the requirements of ACTION A.4 or A.5 in Specification 3.4.3.2.

l NORTH ANNA - UNIT 1 3/4 4-7b Amendment No. 89, 218

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REACTOR COOL' ANT SYSTEM LOW-TEMPERATURE OVERPRESSURE PROTECTION SURVEILLANCE REQUIREMENTS.

'4,4.9.3 ^

Each PORV shall be demonstrated OPERABLE by:

a.

Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation

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channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.

b.

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

~ Verifying the PORV keyswitch is in the Auto position and the PORV isolation c.

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.

d.

At least once per 7 days by verifying that the pressure in the PORV nitrogen accumulators is greater than the surveillance limit.'

l Testing pursuant to Specification 4.0.5.

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i' Amendment No. 16,31,189, 218

.. NORTH ANNA - UNIT 1 3/4 4 t

y; 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 AND 3/4.4.3 SAFETY AND RELIEF VALVES l

The pressurizer wde safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 lbs per hour l

of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during hot shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power operated relief valves (PORVs) will provide overpressure relief capability and will prevent RCS j

overpressurization.

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During operation, all pressurizer code safety valves must be OPERABLE to prev:nt the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

The safety valve tolerance requirement for Modes 1-3 is expressed as an average value.

That is, the as-found error (expressed as a positive or negative percentage) of each tested safety valve is summed and divided by the number of valves tested. This average as-found value is compared to the acceptable range of + 2% to - 3%. In addition, no single valve is allowed to be outside ofi 3%.

An average tolerance of + 2% /- 3% was confirmed to be adequate for Modes 1-3 accident analyses. For the overpressure events, the analyses considered several combinations of valve tolerance with the arithmetic average of the three valves' tolerance equal to + 2% (with no valve outside ofi 3%). The case of a + 2% tolerance on each of the three valves provided the most limiting results. The - 3% tolerance is limiting for the DNB acceptance criterion.

The power operated relief valves (PORVs) and pressurizer steam bubble function to relieve reactor coolant system (RCS) pressure during all design transients up to and including the design step load decrease with concurrent operation of the condenser with steam dumps. Operation of the l

PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

Each PORV has a remotely operated block valve supplied from emergency power to provide isolation capability should a relief valve become inoperable due to:

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NORTH ANNA - UNIT I B 3/4 4-2 Amendment No. 32, "!,189, 20^, 2 ! 5, 218

3/4.4 REACTOR COOLANT SYSTEM BASES a) seat leakage, or b) a mechanical or control system problem which results in either the valve sticking open or the potential for a spurious opening of the valve.

The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

a) Manual control of PORVs to control reactor coolant system pressure. This function is assumed for mitigation of the steam generator tube rupture accident and is therefore considered a safety function.

b) Maintaining the integrity of the reactor coolant pressure boundary. This function is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

M Manual control of the bl.k valve to (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a, above), and (2) iso-late a PORV with excessive seat leakage (Item b, above).

d) Automatic control of PORVs to control reactor coolant system pressure. This is a function that reduces challcoges to the code safety valves for overpressurization events.

e) Manual control of a block valve to isolate a stuck PORV or a PORV which has the potential for spurious opening due to a control system malfunction.

If a PORV is inoperable because of excessive seat leakage, closure of the block valve is required to restore RCS pressure boundary integrity. Power is maintained to the block valves so that the PORVs will continue to be available for the safety related function of manual RCS depressurization in the event of a steam generator tube rupture. In addition, the automatic pressure control capability could be restored by opening the block valves. Therefore, continued operation is allowed.

With one or both PORVs inoperable due to an inoperable backup nitrogen supply, continued operation for 14 days is allowed provided the normal motive force for the PORVs, i.e.,

the instrument air system, continues to be available. Instrument air has a high system reliability, and the likelihood of its being unavailable during a demand for PORV operation is low enough to justify a reasonable length of time (i.e.,14 days) to repair the nitrogen system. A Configuration Risk Management Program (CRMP) defined in Administrative Control Section 6.8.4.g is implemented to evaluate risk associated with an inoperable backup nitrogen supply.

NORTH ANNA - UNIT I B 3/4 4-2a Amendment No. '89,2T,218

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3/4.4 REACTOR COOLANT SYSTEM BASES If one or both PORVs are inoperable due to a malfunction in the PORV automatic control system, the PORV should be placed in manual control to limit the potential for a spurious opening of the PORV. Closure of the block valves is not required in this case so that the manual function remains readily available.

If one or both block valves are inoperable, the associated PORV(s) are placed in manual control to limit the potential for a spurious PORV opening due to a control system malfunction which would not be isolable via the clock valve. The time allowed to restore the block valve (s) to OPERABLE status is based on the remedial action time limits for inoperable PORV control systems since the PORVs are net capable of mitigating an overpressure event when placed in manual control.

Surveillance Requirements provide the assurance that the PORVs and block valves can perfonn their functions. Specification 4.4.3.2.1 addresses the PORVs, including the backup nitrogen supply system, and Specification 4.4.3.2.2 addresses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply wig the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable l

PORV to. operable status.

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l Surveillance limits are established for the pressure in the backup nitrogen accumulators to ensure there is adequate motive power for the PORVs to cope with a steam generator tube rupture coincident with loss of the containment instrument air system.

Surveillance Requirement 4.4.3.2.1.b provides for the testing of the mechanical and electrical aspects of the PORVs and their associated control systems. This testing is performed in MODE 3 or 4 to limit the potential of spurious or inadvertent opening of a PORV during power l

operation while maintaining a temperature and pressure environment which is representative of power operating conditions.

l 3/4.4.4 PRESSURIZER l

The limit on the maximum water volume in the pressurirtr assures that the parameter is maintained within the normal steady state envelope of operation assumed in the S AR. The limit is consistent with the initial SAR assumptions. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.

NORTH ANNA - UNIT I B 3/4 4-2b Amendment No. 218

m REACTOR COOLANT SYSTEM BASES Low-Temperature Overpressure Protection l

The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 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 the RCS cold legs are less than or equal to 235 F. Either PORV has adequate relieving capability to protect the RCS from

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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 charging pump and its injection into a water-solid RCS.

1 Automatic or pauive low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 235 F. This temperature is the water temperature corresponding to a metal temperature of at least the limiting RTsor + 50 F +

j instrument uncertainty. Above 235 F administrative contro is adequate protection to ensure the limits of the heatup curve (Figure 3.4-2) and the cooldown curve (Figure 3.4-3) are not violated.

The concept of requiring automatic LTOP at the lower end, and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11.

Surveillance limits are established for the pressure in the backup nitrogen accumulators to ensure there is adequate motive power for the PORVs to cope with an inadvertent start of a high head safety injection pump in a water solid condition, allowing adequate time for the operators to respond to terminate the event.

NORTH ANN A - UNIT I B 3/4 4-8 Amendment No. 71, ! !?,170, 489, 218

ADMINISTRATIVE CONTROLS Configuration Risk Management Program (continued)

3) Provisions for performing an assessment after entering the LCO Action Statement i

for unplanned entry into the LCO Action Statement.

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4) Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in tN LCO Action Statement.

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5) Provisions for considering other applicable risk significant contributors such as Level 2 issue and external events, qualitatively or quantitatively.

Current risk-informed action statements include: Action 3.8.1.1.b; 3.4.3.2.A.2 l

63 REPORTING IGOUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Oliice of Inspection and Enforcement unless otherwise noted.

i STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (a) receipt of an operating license, (2) amendment to the license involving a planned increase in power level,(3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details requested in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

NORTH ANNA - UNIT 1 6-14 Amendment No. 63, 2 M, 218

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UNITED STATES 6

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- WASHINGTON, D.C. 20555-0001 s,.... /

VIRGINIA ELECTRIC AND POWER COMPAN_Y OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT NO. 2 i

AMENDMENT TO FACILITY OPERATING LICENS_E Amendment No.199' License No. NPF-7 1.

- The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated. October 25,1995, as supplemented February 5,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

. B.

~ The facility _will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in comoliance with the Commission's

regulations;-

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-7 is hereby amended to read as follows:

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L (2)-. Technical Soecifications' l

The Technical Specifications contained in Appendices' A and B, as revised through Amendment No. 199, are hereby incorporated in the license..- The licensee shall operate the facility in accordance with the. Technical Specifications.

3.

^ This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION O

y H rbert N. Berkow, Director Project Directorate 11-2 Division of Licensing Project Management Off;ce of Nuclear Reactor Regulation Attachments:

Changes to the 7 ichnical Specifications Date of issuance: March 2,1999 1

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ATTACHMENTTO LICENSE AMENDMENT NO.199 i

TO FACILITY OPERATING LICENSE NO. NPF-7 I

DOCKET NO. 50-339

- Replace the following pages of the Appendix "A" Technical Specifications with the enclosed.

- pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Paaes insert Paaes 3/4 4-7a 3/4 4-7a 3/4 4-7b 3/4 4-7b i

3/4 4-31 3/4 4-31 B 3/4'4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a B 3/4 4-2b B 3/4 4-8 8 3/4 4-8 6-14d 6-14d l

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< REACTOR COOLANT SYSTEM

SAFETY AND RELIEF VALVES - OPERATING i

' RELIEF VALVES.,

LIMITING CONDITION FOR OPERATION 3.4.3.2 Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

' APPLICABILITY: MODES 1,2, and 3.

JACTION:

A: PORV(s):

1. With one or both PORV(s) inoperable solely because of excessive seat leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. (Risk Informed) With one or both PORV(s) inoperable because of (an) inoperable backup nitrogen supply (ies), within 14 days either restore the PO.RV(s) backup nitrogen supply (ies) to OPERABLE status or be in HOT STANDBY within the next 6

' hours and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3. With one or both PORV(s) inoperable due to a malfunction in the PORV automatic control system, within I hour restore the affected automatic control system (s) to-

' OPERABLE. status or place and maintain the affected PORV(s) in manual control.

l 4. ' With one PORV inoperable due to causes other than those addressed in ACTIONS A.1, I

A.2 or A.3 above, within I hour either restore the PORV to OPERABLE status or close

-j its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

1 Si With both PORVs inoperable such that ACTIONS A.1, A.2 or A'.3 above do not apply, within I hour e'ither restore at least one PORV to OPERABLE status or close the i

associated block valves and remove power from the block valves and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

6. The provisions of Specification 3.0.4 are not applicable.

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~ NORTH ANNA - UNIT 2 3/4 4-7a -

Amendment No. 4-70, ~199 r

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r REACTOR COOLANT SYSTEM SAFETY AND RELIEF VALVES - OPERATING RELIEF VALVES LIMITING CONDITION FOR OPERATION ACTION:(Continued)

B. Block Valves:

1.

With one block valve inoperable, within I hour either restore the block valve to OPERABLE status or place its associated PORV in manual control; restore the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SilUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.

With both block valves inoperable, within I hour either restore the block valves to OPERABLE status or place the PORVs in manual control; restore at least one block j

valve to OPERABLE status within the next hour; restore the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE:

a. At least once per 31 days by performing a CHANNEL FUNCTIONAL TEST, excluding valve operation, and

b. At least once per 18 months by:

1. Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and

2. Operating the solenoid air control valves and check valves on the associated accumulators in the PORV control systems through one complete cycle of full travel, and

3. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

c. At least once per 7 days be verifying tha'. the pressure in the PORV nitrogen accumulators is greater than the surveillance limit.

4.4.3.2.2 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 in order to meet the requirements of ACTION A.4 or A.5 in Specification 3.4.3.2.

l NORTH ANNA - UNIT 2 3/4 4-7b Amendment No. 4-70,199 i

REACTOR COOLANT SYSTEM LOW-TEMPERATURE OVERPRESSURE PROTECTION SURVEILLANCE REQUIREMENTS 4.4.9.3 Each PORV shall be demonstrated OPERABLE by:

a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which j

the PORV is required OPER.ABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.

b. Performance of a CHANNEL. CALIBRATION on the PORV actuation channel, at j

least once per 18 months.

c. Verifying the PORV keyswitch is in the AUTO position and 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 j

protection.

d. At least once per 7 days by verifying that the pressure in the PORV nitrogen accumulators is greater than the surveillance limit.

e. Testing pursuant to Specification 4.0.5 1

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NORTH ANNA - UNIT 2 3/4 4-31 Amendment No. 4-74,199

REACTOR COOLANT SYSTEM BASES 3/4.4.2 AND 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during hot shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power operated relief valves (PORVs) will provide overpressure relief capability and will prevent RCS j

overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss ofload assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

The safety valve tolerance requirement for Modes 1-3 is expressed as an average value.

That is, the as-found error (expressed as a positive or negative percentage) of each tested safety valve is summed and divided by the number of valves tested. This average as-found value is compared to the acceptable range of + 2% to - 3%. In addition, no single valve is allowed to be outside ofi 3%.

An average tolerance of + 2% /- 3% was confirmed to be adequate for Modes 1-3 accident analyses. For the overpressure events, the analyses considered several combinations of valve tolerance with the arithmetic average of the three valves' tolerance equal to + 2% (with no valve outside ofi 3%). The case of a + 2% tolerance on each of the three valves provided the most limiting results. The - 3% tolerance is limiting for the DNB acceptance criterion.

The power operated relief valves (PORVs) and pressurizer steam bubble function to relieve reactor coolant system (RCS) pressure during all design transients up to and including the design step load decrease with concurrent operation of the condenser with steam dumps. Operation of the l

PORVs minimizes the undesirable opening cf the spring-loaded pressurizer code safety valves.

Each PORV has a remotely operated block valve supplied from emergency power to pr > vide isolation capability should a relief valve become inopeable due to:

NORTH ANNA - UNIT 2 B 3/4 4-2 Amendment No. 121,170,181,199

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j 3/4.4 REACTOR COOLANT SYSTEM BASES a) seat leakage, or b) a mechanical or control system problem which results in either the valve sticking open or the potential for a spurious opening of the valve.

i The OPERABILITY of the PORVs and block valves is determined on the basis of their i

being capable of performing the following functions:

a) Manual control of PORVs to control reactor coolant system pressure. This function is assumed for mitigation of the steam generator tube rupture accident and is therefore considered a safety function.

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b) Maintaining the integrity of the reactor coelant pressure boundary. This function is related to controlling identified leakage and ansuring the ability to detect unidentified reactor coolant pressure boundary leakage.

c) Manual control of the block valve to (_1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a, above), and (2) isolate a PORV with excessive seat leakage (Item b, above).

d) Automatic control of PORVs to control reactor coolant system pressure. This is a function that reduces challenges to the code safety valves for overpressurization events.

e) Manual control of a block valve to isolate a stuck PORV or a PORV which has the potential for spurious opening due to a control system malfunction.

If a PORV is inoperable because of excessive seat leakage, closure of the block valve is required to restore RCS pressure boundary integrity. Power is maintained to the block valves so that the PORVs will continue to be available for the safety related function of manual RCS depressurization in the event of a steam generator tube rupture. In addition, the automatic pressure control capability could be restored by opening the block valves. Therefore, continued operation is e!! owed.

With one or both PORVs inoperable due to an inoperable backup nitrogen supply, continued operation for 14 days is allowed provided the normal motive force for the PORVs,i.e.,

the instrument air system, continues to be available. Instrument air has a high system reliability, and the likelihood of its being unavailable during a demand for PORV operation is low enough to justify a reasonable length of time (i.e.,14 days) to repair the nitrogen system. A Configuration Risk Management Program (CRMP) defined in Administrative Control Section 6.8.4.g is implemented to evaluate risk associated with an inoperable backup nitrogen supply.

If one or both PORVs are inoperable due to a malfunction in the PORV automatic control system, the PORV should be placed in manual control to limit the potential for a spurious opening NORTH ANNA - UNIT 2 B 3/4 4-2a Amendment No.170, !M,199

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3'/I.4 REACTOR COOLANT SYSTEM BASES of the PORV. Closure of the block valves is not required in this case so that the manual function remains readily available.

If one or both block valves are inoperable, the associated PORV(s) are placed in manual control to limit the potential for a spurious PORV opening due to a control system malfunction which would not be isolable via the block vane. The time allowed to restore the block valve (s) to OPERABLE status is based on the remedial action time limits for inoperable PORV control systems since the PORVs are not capable of mitigating an overpressure event when placed in manual control.

Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions. Specification 4.4.3.2.1 addresses the PORVs, including the backup nitrogen supply system, and Specification 4.4.3.2.2 addresses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to OPERABLE status.

Surveillance limits are established for the pressure in the backup nitrogen accumulators to ensure there is adequate motive power for the PORVs to cope with a steam generator tube rupture coincident with loss of the containment instrument air system.

Surveillance Requirement 4.4.3.2.1.b provides for the testing of the mechanical and electrical aspects of the PORVs and their associated control systems. This testing is performed in MODE 3 or 4 to limit the potential of spurious or inadvertent opening of a PORV during power operation while maintaining a temperature and pressure environment which is representative of power operating conditions.

3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE ensures that the plant will be able to establish natural circulation.

NORTH ANNA - UNIT 2 B 3/4 4-2b Amendment No. 199

- REACTOR COOLANT SYSTEM BASES Low-Temperature Overpressure Protection The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 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 the RCS cold legs are less than or equal to 270 F. Either PORV 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 charging pump and its injection into a water-solid RCS.

Automatic or passive low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 270 F. This temperature is the water temperature corresponding to a metal temperature of at least the limiting RTNDT + 50 F +

instrument uncertainty, Above 270"F administrative control is adequate protection to ensure the limits of the heatup curve (Figure 3.4-2) and the cooldown curve (Figure 3.4-3) are not violated.

The concept of requiring automatic LTOP at the lower end, and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11.

Surveillance limits are established for the pressure in the backup nitrogen accumulators to ensure there is adequate motive power for the PORVs to cope with an inadvertent start of a high head safety injection pump in a water solid condition, allowing adequate time for the operators to respond to terminate the event.

3/4.4.10 STRUCTURAL INTEGRITY 3/4.4.10.1 ASME CODE CLASS 1,2 and 3 COMPONENTS The inspection programs for ASME Code Class 1,2 and 3 Reactor Coolant System components ensure that the structural integrity of these components will be maintaine.1 at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

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NORTH ANNA-- UNIT 2 B 3/4 4-8 Amendment No. ! 19.170,199 l

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ADMINISTRATIVE CONTROLS Configuration Risk Management Program (continued)

3) Provisions for performing an assessment after entering the LCO Action Statement for unplanned entry into the LCO Action Statement.

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4) Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in the LCO Action Statement.

5) Provisions for considering other applicable risk significant contributors such as Level 2 issue and external events, qualitatively or quantitatively.

Current risk-informed action statements include: Action 3.8.1.1.b; 3.4.3.2.A.2 l

63 REPORTING REOUIREMENTS l

l ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (a) receipt of an operating license, (2) amendment to the license involving a planned increase in power level,(3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

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I NORTH ANNA - UNIT 2 6-14d Amendment No. 37, ! !1,195,199

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