ML20249B073

Amend 216 to License DPR-65,changing TS to Resolve Several Compliance Issues

ML20249B073
Person / Time
Site:	Millstone
Issue date:	06/16/1998
From:	Mckee P NRC (Affiliation Not Assigned)
To:
Shared Package
ML20249B074	List: ML20249B073 ML20249B075
References
NUDOCS 9806220039
Download: ML20249B073 (18)

Text

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UNITED STATES p

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NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 20066 4 001 NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 216 License No. DPR-65 1.

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

A.

The application for amendment by Northeast Nuclear Energy Company, et al.

(the licensee) dated December 8,1997., 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 requiremerr.s have been satisfied.

9906220039 990616 PDR ADOCK 05000336 p

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2-2.

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. DPR-65 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 216, 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 the date of issuance, to be implemented within 30 days ofIssuance.

FOR THE NUCLEAR REGULATORY COMMISSION V

b Phillip F. McKee Deputy Director for Licensing Special Projects Office Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

June 16, 1998 I

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I 6IIACHMENT TO LICENSE AMENDMENT NO. 216 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Inlad 1-4 1-4 3/4 1-2 3/4 1-2 3/4 1-26 3/4 1-26 3/4 1-29 3/4 1-29 3/4 5-7 3/4 5-7 3/4 6-19 3/4 6-19 3/4 8-6 3/4 8-6 3/4 8-6a 3/4 8-6a 5-1 5-1 5-4 5-4 5-6 5-6 B 3/4 1-4a B 3/41-4a B 3/4 1-5 8 3/4 1-5 B 3/4 5-2 8 3/4 5-2 B 3/4 6-3e B 3/4 6-3e

DEFINITIONS AZIMUTHAL POWER TILT - T 1.18 AZIMUTHAL POWER TILT shall be the difference between the maximum l

power generated in any core quadrant (upper or lower) and the average power of ali quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core.

Maximum nower in any core cuadrant funoer or lower)

-1 AZIMUTHAL POWER TILT

=

Average power of all quadrants (upper or lower)

DOSE E0VIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (micro-curie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109 REv.1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix I."

E-AVERAGE DISINTEGRATION ENERGY 1.20 E shall be the average sum of the beta and gamma energies per dis-integration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subinterval, and b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

BE00ENCY NOTATION 1.22 The FREQUENCY P 101 specifiea for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

NILLSTONE - UNIT 2 1-4 Amendment No. Jpp. 216 0335

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d.

When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consider-ation of the following factors:

1.

Reactor coolant system boron concentration, 2.

CEA position, 3.

Reactor coolant temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement with 1.0% Ak/k at least once per 31 Effective Full Power Days. This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.d, above. The predicted reactivity values may be adjusted (normalized) to correspor.d l

to the actual core conditions prior to exceeding a fuel burnup of 60 Effective. Full Power Days after each refueling.

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i NILLSTONE UNIT 2 3/4 1-2 Amendment No. 216 032B

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REACTIVITY CONTROL SYSTENS CEA DROP TINE LINITING CONDITION FOR OPERATION l

3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be 5 2.75 seconds from when electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a.

Tm 2 515'F, and b.

All reactor coolant pumps operating.

APPLICABILITY: MODES I and 2.

ACTION:

a.

With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operation at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full length shall be demonstrated through measurement prior to reactor criticality:

For all CEAs following each removal of the react'or vessel a.

head, 6

b.

For specifically affected individual CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and c.

At least once per 18 months.

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NILLSTONE - UNIT 2-3/41-26 Amendment No. M. 77, pp. 216 0337 L--_--_----_--_--__

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REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Continued)

SURVEILLANCE REQUIREMENTS c.

With the regulating CEA grou)s inserted between the Long Term Steady State Insertion Limits and tie Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT for intervals > 5 EFPD per 30 EFPD interval or >-14 EFPD per calendar year, except during operations pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, either:

1.

Restore the regulating groups to within the Long Term Steady State Insertion Limits provided in the CORE OPERATING LIMITS REPORT within two hours, or 2.

Be'in HOT STANDBY within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1 SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be 1

within the Transient Insertion Limits provided in the CORE OPERATING LIMITS REPORT at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the PDIL alarm is inoperable, then verify the individual CEA positions at least once l

4 per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The accumulated times during which the regulatory CEA groups are inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT shall be determined at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> specified.

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MILLSTONE - UNIT 2 3/4 I-29 Amendment JJJ, JJJ, 216 0338 -

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i ENERGENCY C0RE C0OLING SYSTEMS ECCS SUBSYSTEMS - T = < 300*F LINITING CONDITION FOR OPERATION l

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

a.

One # OPERABLE high-pressure safety injection pump, and b..

An OPERABLE flow path capable of taking suction from the refuel-ing water storage tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a sump recirculation actuation signal.***

l APPLICABILITY: MODES 3* and 4.

I ACTION:

a.

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within one hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

c.

With two or more high pressure safety injection pumps OPERABLE and the temperature of one or more of the RCS cold legs s 275'F take immediate action to have a maximum of one high pressure safety injection pump OPERABLE.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

With pressurizer pressure < 1750 psia.

A maximum of one high-pressure safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is s 275'F.

In MODE 4, the requirement for OPERABLE safety injection and sump i

recirculation actuation signals is satisfied by use of the safety injection and sump recirculation trip pushbuttons.

l NILLSTONE - UNIT 2-3/4 5-7 Amendment Not 77 17), 216 once

CONTAIMENT SYSTEMS CONTAIMENT VENTILATION SYSTEM LINITING CONDITION FOR OPERATION 3.6.3.2 The containment purge supply and exhaust isolation valves shall be sealed closed.

l APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one containment purge supply and/or one exhaust isolation valve open, close the open valve (s) within one hour or 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 />.

l SURVEILLANCE REQUIREMENTS 4.6.3.2 The containment purge supply and exhaust isolation valves shall be determined sealed closed at least once per 31 days.

MILLSTONE UNIT 2 3/4 6-19 Amendment No. #1, 216 034o

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ELECTRICAL POWER SYSTEMS 3/4 8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators with tie breakers open between redundant busses:

4160 volt Emergency Bus # 24 C 4160 volt Emergency Bus #24 0 480 volt Emergency Load Center #22 E 480 volt Emergency Load Center #22 F 120 volt A.C. Vital Bus # VA-10 120 volt A.C. Vital Bus # VA-20 120 volt A.C. Vital Bus # VA-30 120 volt A.C. Vital Bus # VA-40 APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus and/or associated load center to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from normal A.C. sources with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment and indicated power availability.

MILLSTONE - UNIT 2 3/4 8-6 Amendment No.216 out

ELECTRICAL POWER SYSTENS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION (Continued) 3.8.2.lA Inverters 5 and 6 shall be OPERABLE and available for automatic transfer via static switches VS1 and VS2 to power busses VA-10 and VA-20, l

respectively.

APPLICABILITY: NODES 1, 2 &'3 ACTION:

a.

With inverter 5 or 6 inoperable, restore the inverter to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With inverter 5 or 6 unavailable for automatic transfer via static switch VS1 or VS2 to power bus VA-10 or VA-20, l

I respectively, restore the automatic transfer capability' within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I c.

With inverters 5 and 6 inoperable or unavailable for automatic transfer via static switches VS1 and VS2 to power busses VA-10 and VA-20, respectively, restore the l

inverters to OPERABLE status or restore their automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.1A a.

Verify correct inverter voltage, frequency, and alignment -

for automatic transfer via static switches VS1 and VS2 to power busses VA-10 and VA-20, respectively, at least once l

per 7 days.

b.

Verify that busses VA-10 and VA-20 automatically transfer l to their alternate power sourdes, inverters 5 and 6, respectively, at least once per refueling during shutdown.

i NILLSTONE - UNIT 2 3/48-6a Amendment No. Jpp, 216 0341

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a.

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5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area is shown on Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone is shown on Figure 5.1-2.-

FLOOD CONTP.2L 5.1.3 The flood control provisions shall be designed and maintained in accordance with the design provisions contained in Section 2.5.4.2 l

of the FSAR.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a.

Nominal inside diameter - 130 feet.

b.

Nominal inside height - 175 feet.

c.

Minimum thickness of concrete walls - 3.75 feet.

d.

Minimum thickness of concrete dome - 3.25 feet.

e.

Minimum thickness of concrete floor pad - 8.5 feet.

f.

Nominal thickness of steel liner = 0.25 inches.

g.

Net free volume - 1.9 x 10' cubic feet,.

i NILLSTONE - UNIT 2 5-1 Amendment No. 216 OH5

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i DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained f

for a maximum internal pressure of 54 psig and an equilibrium liner temperature of 289'F.

j PENETRATIONS l

5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the design provisions contained in l Section 5.2.8 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing 176 rods. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.5 weight percent of U-235.

CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies.

The control element assemblies shall be designed and maintained in accordance with the design provisions contained inl Section 3.0 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 4.2.2 of the FSAR with allowance for normal, degradation pursuant of the applicable Surveillance Requirements, b.

For a pressure of 2500 psia, and c.

For a

temperature of 650'F except for the pressurizer which is 700*F.

gLSTONE-UNIT 2 5-4 Amendment No. 77, J77, JJJ, 7pp. 216

c.

4 nFtf PJi FFAT110FC 5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I Items in Section 5.1.1 of the FSAR shall be designed and maintained to the design provisions contained in Section 5.8 of the FSAR with allowance for l

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normal degradation pursuant to the applicable Surveillance Requirements.

5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.

1 NILLSTONE - UNIT 2 5-6 Amendment No. J. JJJ, JJ7, 216 0347

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BASES 3/4.1.3 MOVEABLE CONTROL ASSEMBLIES (Continued) safety analyses and LCO and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the

.. CEA alignment and insertion limits and ensures proper operation of the rod block circuit.

The CEA " Full In" and " Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that 1

the applicable LCO's are satisfied.

l The maximum CEA drop time permitted by Specification 3.1.3.4 is the assumed CEA drop time used in the accident analyses.

Measurement with Tav 1515*Fandwithallreactorcoolantpumpsoperatingensuresthatthemeasure0l drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

- NILLSTONE - LMIT 1 B 3/4 1-4a Amendment No. 177 E216 DME-

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4.

REACTIVITY CONTROL SYSTEMS macre 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued)

The LSSS setpoints' and ' the power distribution LCOs were generated

. based upon a core burnup which would be achieved with the core operating i

in an' essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during MODES I and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA insertion permitted by the Long Tern Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption I

but will still provide sufficient reactivity control. The Transient i

Insertion Limits of Specification 3.1.3.6 are provided to ensure that

. (1) acceptable power distribution limits are maintained, (2) the minimum

. SHUTDOWN MARGIN is maintained, and-(3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configura-tion. The PDIL alarm is provided by the CEAPDS computer.

l The control rod drive mechanism requirement of specification 3.1.3.7 is provided to assure that the consequences of an uncontrolled CEA withdrawal from subcritical transient will stay within acceptable levels.

This specification assures that reactor coolant system conditions exist which are consistent with the plant safety analysis

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prior to energizing the control rod drive mechanisms.

The accident is precluded when conditions exist which are inconsistent with the safety analysis since deenergized drive mechanisms cannot withdraw a CEA.

The drive mechanisms may be energized with the boron concentration greater than or equal to the refueling concentration since, under these I

conditions, adequate SHUTDOWN MARGIN is maintained, even if all CEAs are l

fully withdrawn from the core.

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gLSTONE-UNIT 2 B 3/4 1-5 Amendment No. 77.JJp, 216

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EMERGENCY CORE COOLING SYSTEMS t.

BASES l

The purpose of the ECCS throttle valve surveillance requirements is to provide l

assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

Verification of the correct position for the mechanical and/or electrical j

valve stops can be performed by either of the following methods:

1.

Visual 11y verify the valve opens to the designated throttled position, or 2.

Manually position the valve to the designated throttled position and verify that the valve does not move when the applicable valve control switch is placed to "0 PEN."

In MODE 4 the automatic safety injection signal generated by low pressurizer pressure and high containment pressure and the automatic sump recirculation

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actuation signal generation by low refueling water storage tank level are not required to be OPERABLE.

Automatic actuation in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating engineered safety features components.

Since the manual actuation (trip pushbuttons) portion of the safety injection and sump recirculation actuation signal generation is required to be OPERABLE in MODE 4, the plant operators can use the manual trip pushbuttons to rapidly position all components to the required accident position.

Therefore, the safety injection and sump recirculation actuation trip pushbuttons satisfy the requirement for generation of safety injection and sump recirculation actuation signals in MODE 4.

Only one HPSI pump may be OPERABLE in MODE 4 with RCS temperatures less than or equal to 275'F due to the restricted relief capacity with Low-Temperature Overpressure Protection System.

To reduce shutdown risk by having additional pumping capacity readily available, a HPSI pump may be made inoperable but available at short notice by shutting its discharge valve with the key lock on the control panel.

3/4.5.4 REFUELING WATER STORAGE TANK (RWST)

The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that

1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

MILLSTONE - UNIT 2 B 3/4 5-2 Amendment No. JJ. JJ), JJJ. JJJ. 216 om

CONTli!NNENT SYSTEMS BASES l

3/4.6.3 CONTAINMENT ISOLATION VALVES (continued)

The determination of the appropriate administrative controls for these containment isolation valves included an evaluation of the expected environmental conditions. This evaluation has concluded environmental conditions will not preclude access to close the valve, and this action will prevent the release of radioactivity outside of containment through the respective penetration.

The containment purge supply and exhaust isolation valves are required to be sealed closed during plant operation since these valves have not been l

demonstrated capable of closing during a LOCA or steam line break accident.

Such a demonstration would require justification of the mechanical operability of the purge valves and consideration of the appropriateness of the electrical override circuits. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system. The containment purge supply and exhaust isolation valves are sealed closed by removing power from the valves.

This is accomplished by pulling the control power fuses for each of the valves. The associated fuse blocks are then locked. This is consistent with the guidance contained in NUREG-0737 Item.II.E.4.2 and Standard Review Plan 6.2.4, " Containment Isolation System," Item II.f.

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1 NILLSTONE - UNIT 2 8 3/4 6-3e Amendment No. JJp, JJJ, 216 0393