ML20100K995

Amends 198 & 179 to Licenses NPF-4 & NPF-7,respectively, Revising TS 3.9.4 to Allow Both Containment Pal Doors to Remain Open During Fuel Movements or Core Alterations Provided That One Door Operable

ML20100K995
Person / Time
Site:	North Anna
Issue date:	02/27/1996
From:	Imbro E NRC (Affiliation Not Assigned)
To:
Shared Package
ML20100K999	List: ML20100K995 ML20100L003
References
NUDOCS 9603040040
Download: ML20100K995 (20)

Text

.

[pcuc uq I t UNITED STATES j

j NUCLEAR REGULATORY COMMISSION o

e WASHINGTON, D.C. 200d&0001

%,.....,6 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.198 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 17, 1995, as supplemented February 26, 1996, 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 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.

9603040040 960227 PDR ADOCK 05000338 P

PDR

o 1l' 2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and the following paragraphs under Section 2 of Facility Operating License No. NPF-4 are hereby amended to read as follows:

l 2.D.2 Technical Soecifications The Technical Specifications contained in Appendices A and B, as j

revised through Amendment No. 198, are hereby incorporated in the i

license.

The licensee shall operate the facility in accordance l

with the Technical Specifications.

l 2.G Deleted i

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION OO l

Eugene V. Imbro, Director Project Directorate 11-1 l

Division of Reactor Projects - I/II l

Office of Nuclear Reactor Regulation j

i Attachments:

(1)

Page 7 of License No. NPF-4*

i (2) Changes to the Technical l

Specifications Date of Issuance:

February 27, 1996

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Page 7 is attached, for convenience, for the composite license to reflect these changes.

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l ATTACHMENT TO LICENSE AMENDMENT NO.198 FACILITY OPERATING LICENSE NO. NPF-4 l

DOCKET NO. 50-338 l

LICENSE Remove Paae Insert Pace 7

7 TECHNICAL SPECIFICATIONS Replace the following pages of the Appendix "B" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

.The corresponding overleaf pages are also provided to maintain document completeness.

Remove Paces Insert Paaes 3/4 9-4 3/4 9-4 B 3/4 8-1 B 3/4 8-1 8 3/4 9-1 8 3/4 9-1 B 3/4 9-3 8 3/4 9-3 l

. 2.E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:

" North Anna Power Station Physical Security Plan," with revisions submitted through February 24, 1988; " North Anna Power Station Guard Training and Qualification Plan," with revisions submitted through May 14, 1987; and " North Anna Power Station Safeguards Contingency Plan," with revisions submitted through January 9, 1987.

Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

2.F.

The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittal dated November 6, 1986 (Serial No. 86-477A).

2.G.

Deleted.

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2.H This license is effective as of the date of issuance and shall expire at

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midnight on April 1, 2018.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

R. C. DeYoung, for Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation Attachments:

1.

Construction Related Items to be completed prior to Initial Criticality

. Appendices A and B Technical Specification page changes 3.

Figure 1 4.

Table 1 Date of Issuance: APR 1 1978 Amendment No. 49, 400, 4&, 4M,198

l 1

I

'I REFUELING OPERATIONS i

DECAY TIME

.l LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.

l-APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the. reactor subcritical for less than 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, suspend all l

operations involving movement of irradiated fuel in the reactor pressure-vessel. The provisions of Specification 3.0.3 are not applicable.

)

l SURVEILLANCE REQUIREMENTS l

4.9.3 The reactor shall be determined to have been subcritical for at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> by verification of the date and time of subcriticality l

prior to movement of irradiated fuel in the reactor pressure vessel.

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i NORTH ANNA - UNIT 1 3/4 9-3 Amendment No. 66

II REFT Fi NG OPERATIOb3 CONTAIWFNT MITILDNG PENETRATIONS LIMITTNG CONDITION FOR OPERATION 3.0.4 The containmem building penetrations shall be in the followi'ig stams:

The equipment door closed and held in place by a minimum of four bolts, a.

b.

A minimum of one doorin each airlock is closed

• and l

Each penetration providing direct access from the containment a.mosphere to the c.

outside atmosphere shall be eidsr;

1. Closed by an isolation valve, blind flange, or manuni valve, or 2, Be capable of being closed by an OPERABLE automati: Containment Purge and Exhaust isolation valve.

APPLICABILITY:

During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the abose specification not satisfied,immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fud in the containment butiding.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCF. REQUIREMENTS 4.9.4 Each of the above required contamment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic Containment Purge and Exhaust isolation valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the statt of and at least once per 7 days during CORE ALTERATIONS vi niovenwat ofinadiated fuel in the containment building by:

Verifying the penetrations are in their closed / isolated condition, " or l

a.

b.

Testing the Containment Purge and Exhaust isolation valves and system per the applicable portions of specifications 4.6.3.1.2 and 4.9.9.

Both doors of the containment personnel airlock may be open provided:

One personnel airlock door is OPERABLE (i.e., the door is capable of being a.

closed and that an individual is designated to close the door), and bl. There is at least 23 feet of water above the top of the remeter preunte vcwl flange durir.g movement of fuel assemblies within the containment, or b2. There is at least 23 feet of water above the top ofirradiated fuel memblies within the reactor pressure vessel during CORE ALTERATIONS excluding movement of fuel assemblies.

If both doors of the containment personnel airlock are open pursuant to Specification 3.9.4.b above, one door shall be venfied to be capable of being closed at the above surveillance frequency.

NORTH ANNA-L3TT 1 3/4 9 4 Amendment No.

3/J.0 REFUELING OPER ATIONS

. BASES

' 3/49.1 BORON CONCENTRATION

'I lj The limitations on reactivity conditions during REFUELING ensure that: 1) the reactor will l ! remain suberitical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.

. These limitations are consistent with the initial conditions assumed for the boron dilution inc in the accident analyses. The value of 0.95 or less for K g includes a 19c Ak/k conservative e

allowance for uncertainties. Similarly, the boron concentration of 2300 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.

i l.3/49.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

31/41J DECAY TIME l

The minimum requirement for reactor suberiticality prior to movement ofirradiated fuel l

assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure

that a release of radioactive material within containment will be restricted from leakage to the

{

> environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon a lack of containment pressurization potential while in the REFUELING MODE.

I OPERABILITY of the containment airlock door requires that the door is capable of being l closed, that the door is unblocked and no cables or hoses are being run through the airlock, and that 4

' j a designated individual is continuously available to close the airlock door. This individual must I e stationed near the airlock.

b

{3/495 COMMUNICATIONS The requirement for communication capability ensures that refueling station personnel l can be promptly informed of significant changes in the facility status or core reactivity conditions j :dunng CORE ALTERATIONS.

,!I NORTH ANNA - UNIT 1 B 3/4 9-1 Amendment No. 94, 198

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3/44 ELECTRICAL POWER SYSTEMS BASES

! 3/4 81 arid 3/4 8.2 A C and D C. POWER SOURCES AND DISTRIBUTION i

The OPERABILITY of the A.C. and D.C. power sources and associated distribution 3

l 1 systems during operation ensures that sufficient power will be available to supply the safety related l i equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C.

and D.C. power sources and distribution systems satisfy the requirements of General Design I

Criteria 17 of Appendix "A" to 10 CFR 50.

,l The ACTION requirements speciGed for the levels of degradation of the power sources provide restriction upon contmued facility operation commensurate with the level of degradation.

The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least one of each of the onsite A.C. and I ! D.C. power sources and associated distribution systems OPERABLE during accident conditions

' ! coincident with an assumed loss of offsite power and single failure of the other onsite A.C. sour i

i l

i The ACTION requirements specified in Modes 5 and 6 address the condition where

! sufficient power is unavailable to recover from postulated events (i.e.. fuel handling accident).

1 I

Implementation of the ACTION requirements shall not preclude completion of actions to establish

! l a safe conservative plant condition. Completion of the requirements will prevent the occurrenc

of postulated events for which mitigating actions would be required.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods 2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status, and 3 sufficient power is available for systems necessary to recover from postulated events in these MODES e.g., the control room emergency ventilation system fans during a fuel handling l

accident.

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9 " Selection of Diesel Generator Set Capacity for Standby Power Supplies." March 10.1971, and 1.108 " Periodic Testmg of Diesel Generator Units Used as Onsite Electnc Power Systems at Nuclear Power Plants" Reusion 1. August 1977. as modified by Amendment No. 83 issued August 22.1986.

The Surveillance Requirements for demonstrating the OPERABILITY of the Emergency Diesel Generator battenes and the Station batteries are based on the recommendations of Regulatory Guide 1.129. " Maintenance. Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants." February 1978. and IEEE Std. 450-1980. "IEEE Recommended

. Practice for Maintenance. Testing and Replacement of Large Lead Storage Batteries for l

Generating Stations and Substations." as modiGed by Amendment No 97 issued March 25.1988.

NORTH ANNA - UNIT I B 3/4 8-1 Amendment No. 83,9' !56, 198

FEFUELING CFEF F D S BASES 3/4.9.6 MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that:

1) mani-pulator cranes will be used for movement of control rods and fuel assemblies; 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged curing lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped, 1) the activity release will be limitea to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in opera-tion ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron l

stratification.

After the reactor has shutdown and entered into M03F ~, for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 6 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140'F.

Since the decay heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removal is provided as long as the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140*F. The reduced flow rate provides additional l

margin to vortexing at the RHR pump suction while in Mid Loop Operation. During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratifi-cation.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel lange ensures that a single failure of the operatir.g RHR loop will not result in a complete loss of residual heat removal capa-bility. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment.

The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

NORTH ANNA - UNIT 1 B 3/4 9-2 Amendment No. E,137,

i

! REFUELING OPER ATIONS T

,,! B ASES.

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.! ' 3/4 910 and 3/4 9.11 WATER LEVEL - REACTOR VESSEL AND SPENT FUEL PIT l

The restrictions on minimum water level ensure that sufficient water depth is available to l

remove 99Ce of the assumed 109c iodine gap activity released from the rupture of an irradiated fuel l

assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

. The minimum water level for movement of fuel assemblies (23 feet above the vessel flange) assures that sufficient water depth is maintained above fuel elements being moved to or i I from the vessel. With the upper internals in place, fuel assemblies and control rods cannot be l removed from the vessel. Operations involving the lifting of control rods with the vessel internals in place may proceed with less than 23 feet of water above the vessel flange provided that 23 feet of water is maintained above all irradiated fuel assemblies within the reactor vessel.

c e

1 3/4 9.12 FUEL BUILDING VENTILATION SYSTEM I

e

' I i

The limitations on the fuel building ventilation system ensure that all radioactive material l released from an irradiated fuel assembly will be filtered through the auxiliary bui

! charcoal filter assemblies prior to discharge to the atmosphere. The Fuel Handling Accident i analysis does not require filtration of the fuel building exhaust in order to meet the analysis criteria.

However, the OPERABILITY of this system and the resulting iodine removal capacity provide additional conservatism compared with the assumptions of the accident analyses.

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' NORTH ANNA UNIT !

B 3/4 9 3 Amendment No. 444,198 e

3KfCp

[

It UNITED STATES 2

E-NUCLEAR REGULATORY COMMISSION f

WASHINGTON, D.C. 20066 0001

%, *...../

VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.179 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 17, 1995, as supplemented February 26, 1996, 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, i

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.

1

. 2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and the following paragraphs under Section 2 of Facility Operating License No. NPF-7 are hereby amended to read as follows:

2.C.(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.179

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

2.I Deleted 3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

[YS Eugene V. Imbro, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachments:

(1) Page 13 of License No. NPF-7*

(2) Changes to the Technical Specifications Date of Issuance:

February 27, 1996 Page 13 is attached, for convenience, for the composite license to l

reflect these changes.

ATTACHMENT TO LICENSE AMENDMENT NO. 179 FACILITY OPERATING LICENSE NO. NPF-7 9

DOCKET NO. 50-339 l

LICENSE i

Remove Pace Insert Pace 13 13 TECHNICAL SPECIFICATIONS Replace the following pages of the Appendix "B" Technical Specifications with the enclosed pages as indicated.

The revised pages are identified by anendment number and contain vertical lines indicating the area of chsnge.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Paaes Insert Paaes 3/4 9-4 3/4 9-4 B 3/4 8-1 B 3/4 8-1 B 3/4 9-1 B 3/4 9-1

)

B 3/4 9-3 B 3/4 9-3 I

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. 2.E.

The licensee shall fully implement and maintaia in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:

" North Anna Power Station Physical Security Plan," with revisions submitted through February 24, 1988; " North Anna Power Station Guard Training and Qualification Plan," with revisions submitted through May 14, 1987; and " North Anna Power Station Safeguards Contingency Plan," with revisions submitted through January 9, 1987. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

2.F.

The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittal dated November 6, 1986 (Serial No. 86-477A).

G.

If VEPC0 plans to remove or to make significant changes in the normal operation of equipment that controls the amount of radioactivity in effluents from the North Anna Power Station, the NRC shall be notified in writing regardless of whether the change affects the amount of radioactivity in the effluents.

H.

VEPC0 shall report any violations of the requirements contained in Section 2, Items C.(3) through C.(21), E, F and G of this license within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegram, mailgram, or facsimile transmission to the Director of the Regional Office, or his designate, no later than the first working day following the violation, with a written followup report within 14 days.

2.1.

Deleted.

i Amendment No. 87, 43, 440,179

REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.

I APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor suberitical for less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, suspend all operations I

5 involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> by verification of the date and time of suberiticality prior to l

movement of irradiated fuel in the reactor pressure vessel.

NORTH ANNA - UNIT 2 3/4 9-3 Amendmen: No.52

REFITJT NG OPERATIONS rnNTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment buildmg penetrations shall be in the following status:

a.

The equipment door closed and held in place by a minimum of four bolts, b.

A mininum of one door in each airlock is closed,

• and l

4 k.ach penetration providing direct access from the containnmut atumplace to the

{

c.

outside atmosphere shall be either-l

. Closed by an isolation valve, blind flange. or manual valve, or

2. Be capable of being closed by an OPERABLE automatic Containment Purge and Exhaust isolation valve.

APPLIC6BILITY:

During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not sansfied, immediately suspend all operations involving CORE ALTERATIONS or movement ofirradiated fuel in the containment building.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4 Cach of the aboyc rcquired cor.tainment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic Containment P.trge and Exhaust isolation valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least ence per 7 days dusing CORE ALTERATIONS or movercent of irradiated fuel in the con:ainment building br i

Verifying the penetrations are in their closed / isolated condition," or l

a.

Both doors of the containment permututel a1:194 may be open provided:

One personnel airlock door is OPERABLE (i.e., the door is capable of being a.

closed and that an individual is designated to close the door), and bl Them is nr least 2'4 feet of water above the top of the reactor pressure vessel flange during movement of fuel assemblies within the containment, or b2. Theie is at least 23 feet of water above the top ofirradiated fuct asacmblies withm the reactor pressure vessel during CORE ALTERATIONS excluding movement of fuel assunblies.

" If both doors of the containment personnel airlock are open pursuant to Specification 3.9.4.b above, one door shall be verified to be capable of being clcsed at the above surveillance frequency.

NORTH ANNA - UNIT 2 3/49-4 Amendment No.

._m

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Etj ELECTRICAL POWER SYSTEMS i -

BASES 1

3/4 8.1 and 3/4 8.2 A C. and D C. POWER SOURCES AND DISTRIBUTION The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related l

equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified indep ndent and redundant A.C.

i and D.C. power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.

l_

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation.

The OPERABILITY of the power sources are consistent with the initial condition assumptions of I

the accident analyses and are based upon maintaining at least one of each of the onsite A.C. and :

I D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source.

l The ACTION requirements specified in Modes 5 and 6 address the condition where sufficient power is unavailable to recover from postulated events (i.e., fuel handling accident).

l Implementation of the ACTION requirements shall not preclude completion of actions to establish a safe conservative plant condition. Completion of the requirements will prevent the occurrence of postulated events for which mitigating actions would be required.

i The OPERABILITY of the minimum specified A.C. and D.C, power sources and I

associated distribution systems during shutdown and refueling ensures that 1) the facility can be

. maintained in the shutdown or refueling condition for extended time periods. 2) sufficient ll mstrumentation and control capability is available for monitosing and maintaining the unit status, and 3) sufficient power is available for systems necessary to recover from postulated es ents in these MODES. e.g., the control room emergency sentilation system fans during a fuel handling l

accident.

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9 " Selection of i

Diesel Generator Set Capacity for Standby Power Supplies " March 10,1971, and 1.108 " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants" Revision I. August 1977, as modified by Amendment No.48 issued August 22,1986.

The Surveillance Requirements for demonstrating the OPERABILITY of the Emergency Diesel Generator batteries and the Station batteries are based on the recommendations of Regulatory Guide 1.129. " Maintenance. Testing and Replacement of Large Lead Storage Batteries l

for Nuclear Power Plants." February 1978, and IEEE Std. 450-1980, "IEEE Recommended Practice for Maintenance, Testing and Replacement of Large Lead Storage Batteries for Generating Stations and Substations," as modified by Amendment No. 84 issued March 25.1988.

1 NORTH ANNA - UNIT 2 B 3/4 81 Amendment No. ^. ", !38.179 s.

n r-n

14 3

' REFL'ELING OPER ATIONS BASES 3/4 9 i BORON COMCENTRATION The limitations on reactivity conditions during REFUELING ensure that: !) the reactor will remain subcritical during CORE ALTERATIONS. and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.

These limitations are consistent with the initial conditions assumed for the boron d in the accident analyses. The value of 0.95 or less for K rrincludes a 1% Ak/k conservative e

' allowance for uncertainties. Similarly, the boron concentration of 2300 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

[

1/4_SJ DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel j

assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the i

radioactive decay of ti - sheet lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

31.43 4 CONTAINMENT BUIT nING PENETRATIONS 4

The requirements on containment building penetration closure and OPERABILITY ensure

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j that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon a lack of containment pressurization potential while in the REFUELING MODE.

OPERABILITY of the containment airlock door requires that the door is capable of being closed. that the door is unblocked and no cables or hoses are being run through the airlock, and that a designated individual is continuously available to close the airlock door. This individual must be stationed near the airlock.

3/49.5 COMMUNICATIONS The requirement for cornmunications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions F

during CORE ALTERATIONS.

NORTH ANNA - UNIT 2 B 3/4 9 1 Amendment No. M. U9 4

REFUELING OPE R ATIONS BASES l

3/4.9.6 MANIPULATOR CRANE OPERABILITY f

The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator i

cranes will be used for movement of control rods and fuel assemblies,2) each crane has sufficient load capacity to lif t a control rod or fuel assembly, and 3) the core intemals and pressure vessel are i

protected from excessive lifting force in the event they are inadvertently engaged dunng lifting operations.

i 3:4.9.7 CPANE TRAVEL - SPENT FUEL PIT The restnction on movement of loads in excess of the nominal weight that of a fuel and control rod assemblies and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped,1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuelin the storage racks will not result in a cntical array This assumption is consistent with the activity release assumed in the accident.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures inat (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F as required dunng the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 6 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140*F. Since the decay heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removalis provided as long as tne reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140*F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1;1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of I

water above the reactor pressure vessel flange ensures that a single f ailure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removea and 23 f eet of water above the reactor pressure vessel flange. a large heat sink is available for core cooling. Thus,in the event of a f ailure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

314.9.9 CONTAINMENT PURGE AND EXHAUST ISOL ATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon cetection of high radiation levels within the containment The OPERABILITY of this system is required to restnct the release of radioactive matenal from the containment atmcsphere to the environment.

NORTH ANNA - UNIT 2 B 3/4 9 2 Amendment No. 120, l

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REFUELING OPER ATIONS

BASES -

i 3/4 9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND SPENT FUEL PIT '

The restrictions on minimum water level ensure that sufficient water depth is available to remove 999c of the assumed 109c iodine gap activity released from the rupture of an irradiated fuel l

assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

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The minimum water level for movement of fuel assemblies (23 feet above the vessel l

Dange) assures that sufficient water depth is maintained above fuel elements being moved to or

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from the vessel. With the upper intemals in place, fuel assemblies and control rods cannot be removed from the vessel. Operations involving the lifting of control rods with the vessel upper intemals in place may proceed with less than 23 feet of water above the vessel flange provided that 23 feet of water is maintained above all irradiated fuel assemblies within the reactor vessel.

3/4 9.12 FUEL BUTI nING VENTILATION SYSTEM The limitations on the fuel building ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the auxiliary building HEPA and chmoal filter assemblies prior to discharge to the atmosphere. The Fuel Handling Accident analysis does not require filtration of the fuel building exhaust in order to meet the analysis criteria.

However the OPERABILITY of this system and the resulting iodine removal capacity provide additional conservatism compared with the assumptions of the accident analyses.

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h NORTH ANNA - UNIT 2 B 3/4 9-3 Amendment No 98.179

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