ML20198S133

Amends 62 & 48 to Licenses NPF-87 & NPF-89,respectively, Increasing AOT for Centrifugal Charging Pump from 72 H to 7 Days & Add Configuration Risk Mgt Program

ML20198S133
Person / Time
Site:	Comanche Peak
Issue date:	12/29/1998
From:	Polich T NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198S136	List: ML20198S133 ML20198S141
References
NUDOCS 9901110172
Download: ML20198S133 (15)

Text

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I NUCLEAR REGULATORY COMMISSION wasniwaron, o.c.=== "

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l TEXAS UTILITIES FI FCTRIC COMPANY

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- COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 2 L

DOCKET NO. 50-4_4Q AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 48 License No. NPF-89 1.

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

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A.

The application for amendment by Texas Utilities Electric Company (TU Electric, I

the licensee) dated August 2,1996 (TXX-96434), as supplemented by letters dated October 2,1998 (TXX-98215), and November 13,1998 (TXX-98241 and TXX-98244), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and i

regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the i

provisions of the Act, and the rules and regulations of the Commission; C.

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~ There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the j

public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l

D.

The issuance of this license 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 l

Commission's regulations and all applicable requirements have been satisfied.

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. NPF-89 is hereby amended to read as follows:

i-l 9901110172 981229 PDR ADOCK 05000445 c

P PDR n

, (2)

Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

48, and the Environmental Protection Plan contained in Appendix B, are hereby incorporateo into this license. TU Electric shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Timothy J. Polich, Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: December 29, 1998 4

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ATTACHMENT TO LICENSE AMENDMENT NOS. 62 AND 48 FACILITY OPERATING LICENSE NCS. NPF-87 AND NPF-89 DOCKET NOS. 50-445 AND 50-446 l

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal lines Indicating the areas of change. The corresponding overleaf pages are also provided to maintain document comp:steness.

REMOVE INSERT 3/4 1-10 3/4110 3/4 5-3 3/4 5 3 B3/41-3 B3/41-3 B3/4 5-2 B3/4 5-2 6-7 6-7 6-7a 6-7a l

REAtTIVfTY LMpfildlig

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CHARGING PUMP - SHUTDOWN I

L LIMITING CONDITION FOR OptRATION 3.1.2.3 Specification 3.1.2.1 shalAt least one charping pump in the boron injection flow path requir be O OPERABLE emergency power source.PERABLE and capable of being powered from an APPL 1tABILITY: MODES 5 and 5.

EIIB:

With no charging pump OPERABLE or capable of being powered from an'0pERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE00fREMENTS I

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4.1.2.3.1 At least once per g2 days the abovt required positive displacement charging pump shall be demonstrated OPERABLE by verifying that the flow path required by Specification 3.1.2.la. Is capable of delivering at least 30 gpa to the RCS; or 4.1.2.3.2 The abovs required centrifugal char OPERABLE by verifying, on recirculation flow, ging pump shall be demonstrated ~-

that a differential pressure across the pump of greater than or equal to 2370 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.3.3 A maximum of two charging pumps shall be OPERABLE, one charging pump shall be demonstrated inoperable

• at least onc9 per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.

• An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with power

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removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

COMANCHE PEAK - UNITS 1 AND 2 3/4 1-9

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1,2,3*, and 4* ".

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 7 days or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least the valus specified in the COLR at 200'F within the next 6. hours; and be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SOHVEILLANCE REQUIREMENTS 4.1.? 4.1 The required centrifugal charging pump (s) shall be demonstrated OPERABLE by ter':,g pursuant to Specification 4.0.5.

4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to Specification 4.1.2.2.c.

4.1.2.4.3 Whenever the temperature of one or more of the Reactor Coolant System (RCS) cold legs is less than or equal to 350*F, a maximum of two charging pumps shall be OPERABLE, except when Specification 3.4.8.3 is not applicable.

When required, one charging pump shall be demonstrated inoperable # at least once per 31 days by verifying that the motor circuit breakers are secured in the open position. - ~

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1 4

. The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODES 3 and 4 for the charging pump declared inoperable pursuant to Specification 3.1.2.4 provided the charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3 or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes ~

first.

"In MODE 4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps.

1. n inoperable pump may be energized for testing provided the discharge of the pump has A

i been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

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. COMANCHE PEAK - UNITS 1 AND 2 3/4 1-10 Unit 1 - Amendment No. 5;44,62 l

Unit 2 - Amendment No. 0948 F

EMERGEWCY CORE COOLING SYSTEMS E4J.2 ECCS SUBSYSTEMS - T _ > 350'F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be '

OPERABLE with each subsystem comprised of:

One OPERABLE centrifugal charging pump, a.

b.

One OPERABLE safetyinjection pump, c.

One OPERABLE RHR heat exchanger,

d.

One OPERABLE RHR pump, and An OPERABLE flow path capable of taking suction from the refueling water storage e.

tank on a Safety injection signal and automatically opening the containment sump suction valves during the recirculation phase of operation.

APPLICABILITY: MODES 1,2, ' nd 3*.

a ACTION:

. With one ECCS subsystem inoperable because of the inoperability of a centrifugal a.

charging pump, restore the inoperable pump to OPERABLE status within 7 days 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

. _m b. -

With one ECCS subsystem inoperable for reasons other than an inoperable -.

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centrifugal charging pump, restore the inoperable subsystem to OPERABLE status -

u within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

s In the event the ECCS is actuated and injects water into the Reactor Coolant l

c.

System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstarces of the actuation and the total accumulated actuation cycles to date. The current value of

- the usage factor for each affected Safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

• The provisions of Specifications 3.0.4 and 4.0.4 ara not applicable for entry into MODE 3 for the centrifugal charging pumps and tho safety injection pumps declared inoperable pursuant to Specification 3.5.3 provided the centrifugal charging pumps and the safety injection pumps are restored to OPERABLE sta.tus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

COMANCHE PEAK - UNITS 1 AND 2 3/45-3 Unit 1 - Amendment No. 62-Unit 2 - Amendment No.48 f j..

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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUTREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves a.

are in the indicated positions with power to the valve operators removed:

Valve Ni=her Valve Function Valve PoS h 8802 A & B SI Pum Closed 8808 A, B, C, D Accum.p to Hot Legs Discharge Open*

880g A & B RHR to Cold Legs Open L135 SI Pump to Cold Legs Open 8840 RHR to Hot Legs Closed 8806 S1 Pump suction from RWST Open 8813 SI Pump Mini-Flow Valve Open b.

At least once per 31 days by verifying that each valve (manual, power-opertted, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

By a visual inspection which verifies that no loose debris (rags, c.

trash, clothing, etc.

transported to the con)tainment sump and cause restriction of thei pump suctions during LOCA conditions. This visual inspection shall s.,

be performed:

1)

For all accessible areas of the containment prior to establish-ing CONTAINMENTJNTEGRITY,_and 2)

At least once daily of the areas affected within containment by containment entr.v and during the final entry when CONTAINMENT INTEGRITY is established.

d.

At least once per 18 months by:

1)

Verifying interlock action of the RHR system from the Reactor Coolant System to ensure that with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 442 psig the interlocks prevent the valves from being opened.

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• Surveillance Requirements covered in Specification 4.5.1.1.

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COMANCHE PEAK - UNITS 1 AND 2 3/4 5-4 Unit 1 - Amendment No. 4,40 Unit 2 - Amendment No. 26 APR 211995

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REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued)

With the RCS temperature below 200*F, one Boron injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron !njection System becomes inoperable.

The limitation for a maximum of two charging pumps to be OPERABLE and the requirement to verify one charging pump to be inoperable below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

With one centrifugal charging pump (CCP) inoperable, the Inoperable CCP must be retumed to OPERABLE status within 7 days. The 7 day Allowed Outage Time is based on a risk-informed assessment to manage the risk ' associated with the equipment in accordance with the Configuration Risk Managerr.et Program and is a reasonable time for repair of the CCPs.

The limitation for minimum solution temperature of the borated water sources are sufficient to prevent boric acid crystallization with the highest allowable boron concentration.

The boron capability required below 200'F is sufficient to provide the required SHUT-DOWN MARGIN after xenon decay and cooldown from 200'F to 140'F. This condition requires eitMr 1,100 gallons of 7000 ppm borated water from the boric acid storage tanks or 7,113 gallons of 2400 ppm borated water from the RWST.

As listed below, the required indicated levels for the boric acid storage tanks and the RWST include allowances for required / analytical volume, unusable volume, measurement uncertainties (which include instrument error and tank tolerances, as applicable), margin, and other required volume.

Ind.

Unusable Required Measurement Tank MODES Level Volur,e Volume Uncertainty Margin Other (gal)

(gal)

(gal)

(gal)

RWST 5,6 24 %

98,000 7,113 4% of span 10,293 N/A 1,2,3,4 95%

45,494 70,702 4% of span N/A 357,535*

Boric 5,6 10%

3,221 1,100 6% of span N/A N/A Acid 5,6 20%

3,221 1,100 6% of span 3,679 N/A Storage.

(gravity _ _ -. __ __.- _. _.._ -

_ = - _

feed)

Tank 1,2,3,4 50%

3,221 15,700 6% of span N/A N/A The OPERABILITY of one Boron injection System during REFUELING ensures that inis system is available for reactivity control while in MODE 6.

• Additional volume required to meet Specification 3.5.4.

COMANCHE PEAK - UNITS 1 AND 2 B 3/41-3 Unit 1 - Amendment No. 3,10,20,44,62 Unit 2 - Amendment No. 5,12,00, 48

R[afTIVITY CONTeot 9YTTEMS

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3 /4.1.3 Novastr coNTRot AsstMattri 4

l The specifications of this section ensure that (1) acceptable power the minimum 5HUTDOWN MAR 4!N is main-j distribution limits are saintained, (t)f rod misalignment on associated accident tained, and (3) the potential effects o analyses are limited. OPERASILITY of the control rod position indicators is j

j required to determine control red positions and thereby ensure compliance with.

the control red alignment and insertion limits. Verification that the Digital Rod position Indicator agrees with the desanded position within i 12 steps at and 228 14, 48,120, and 128 steps withdrawn for the Control Banks and 18, 218, ital Rod steps withdrawn for the Shutdown Banks provides assurances that the Dig position Indicator is operating correctly over the full range of indication.

Since the Digital Rod position Indication Systes does not indicate the actual i

shutdown rod position between 18 steps and 110 steps, only points in the indi-j cated ranges are picked for verification of agreement with demanded position.

The ACTION statements which permit lialted variations from the basic j

requirements are accompanied by additional restrictions which ensure that the original design criteria are net. Misalignment of a rod requires sensurementrof i

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peaking facters and a restriction in THERMAL POWER. These restrictions provide j

assurance of fuel rod integrity during continued operation.

In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during futm e operation.

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For Specification 3.1.3.1 ACTIONS b and c it is incumbent upon the plant This say be by to verify the trippability of the inoperable control rod (s)l in nature, or that j

usually electrica verification of a control system failure,l rod stepping mechanism.

1 the failure is associated with the contro In the event j __

the plant is unable to verify the rod (s) trippability, it must be assumed to be J

untrippable and thus fall under the requirements of ACTION a. Assuming a con-

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trolled shutdown from 1005 RATED THERMAL POWER, this allows approximately four hours for this vertitcation.

The maximum rod drop time restriction is consistent with the assumed rod Measurement with T greater than or drop time used in the safety analyses.

equal to 551"F and with all reactor coolant pumps operating,, ensures that the i

i neasured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

required to be verified on a postnal basis of once per $esition indic Control r'od positions and CPERAl!LITY of the rod 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with more fro-quant verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

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I COMANCHE ptAK - UNITS 1 AND 2 3 3/4 1-4 I

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3/4.5 EMERGENCY CORE COOLING SYSTEMS I

BASES 3/4.5.1 ACCUMULATORS The OPERAGt.ITY of each Reactor Coolant System that a sufficient volume of borated water will be imm(RCS) acesulator ensu ediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

The required accumulator contained volume and nitrogen cover pressure specified in LC0 3.5.1 represent analytical limits.

has not been incorporated into the specified required values. Measurement uncertainty Current control room instrumentation used for indication of accumulator volumes and pressures include a 5 percent measurement uncertainty. Using this instrumentation, indicated values of between 39% and 61% for accumulator level (based on the analytical limits of 6119 gallons and 6597 gallons, respectively plus a 1%

tank tolerance) and between 623 psig and 644 psig (based on analytical limits of 603 psig and 693 psig, respectively) for accumulator nitrogen cover pressure, satisfy the acceptance criteria for SR 4.5.1.

Other methods employed to verify these values in satisfying SR 4.5.1 shall account for measurement uncercainties.

The accumulator

" operating bypasses" power operated isolation valves are considered to be in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is i

required by BTP ICSB 18. This is accomplished via key-lock control board __

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' cut-off switches.

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If the boron concentration of one accumulator is not within limits, it must be' returned to within the required limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In this condition, the ability to maintain suberiticality may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One i

accumulator below the minimum boron concentration limit, however, will have no j

e,ffect on available ECCS water and an insignificant effect on core i

suberiticality during reflood. Thus, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to return the boron concentration to within limits.

If one accumulator is inoperable for any reason other than boron concentration, the accumulator must be returned to OPERABLE status within one 4

l hour.

In this inoperable condition, the required contents of three i

accumulators cannot be assumed to reach the core during a LOCA, which may

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result in unacceptable peak cladding temperatures. Due to the severity of the consequences should a LOCA occur in these conditions, the one hour completion i

i time to open the valve, remove power to the valve, or restore the proper water j

volume or nitrogen cover pressure ensures that prompt action will be taken to

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COMANCHE PEAK - UNITS 1 AND 2 B 3/4 5-1 Unit 1 - Amendment No. 40 Unit 2 - Amendment No. 26 APR 2 71995 1

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EMERGENCY CORE COOLING SYSTEMS BASES ACCUMULATORS (Continued) return the inoperable accumulator to OPERABLE status. The completion time minirrizes the potential for exposure of.the plant to a LOCA under these conditions.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA essuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS

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subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350'F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the bas:s of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of two charging pumps to be OPERABLE and the requirement to verify one charging pump and all safety injection pumps to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

With one centrifugal charging pump (CCP) inoperable, the inoperable CCP must be returned to OPERABLE status within 7 days. The 7 day Allowed Outage Time is based on a risk informed assessment to manage the risk associated with the equipment in accordance with the Configuration Risk Management Program and is a reasonable time for repair of the CCPs.

l The requirement to remove power from certain valve operators is in accordance with Branch Technical Position ICSB-18 for valves that fail to meet single failure considerations._ _ - -

Power is removed via key-lock switches on the control board.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide 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 l

proper flow split between injection points in accordance with the assum?tions used in the j

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ECCS LOCA analyses, and (3) provide an acceptable level of total ECGS flow to all injection

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points equal to or above that assumed in the ECCS-LOCA analyses.

3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY of the refdeling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injec-tion by the ECCS in the event of a LOCA. The limits on RWST minimum volume i

and boron concentration ensure that: (1) sufficient water is available within l

COMANCHE PEAK - UNITS 1 AND 2 B 3/4 5-2 Unit 1 - Amendment No. 40, 62 Unit 2 - Amendment No. e6,48

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ADMINISTRATIVE CONTROLS rate acceptance criteria are s 0.60 L, for the Type B and Type C tests and s 0.75 L, for Type A tests; 2.

Air lock testing acceptance criteria are:

l a)

Overall air lock leakage rate is s 0.05 L, when tested at 2 P.

l b)

For each door, leakage rate is s 0.01 L, when pressurized to a P,.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified L

in the Containment Leakage Rate Testing Program, with the exception of the l

containment ventilation isolation valves, which is specified in Specifications 4.6.1.7.2 and 4.6.1.7.3.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

6.9 REPORTING REQUIREMENTS 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 Regional Administrator of the Regional Office of the NRC unless otherwise noted.

6.9.1.1 Not used.

ANNUAL REPORTS

• 6.9.1.2 Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be 1

- -- submitted prior to March 1 of the year following inliial criticality.-

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Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of station, utility, and other '

- a.

individuals (including contractors), for whom monitoring was performed, receiving an annual deep dose equivalent greater than 100 mrem and the associated collective deep dose equivalent (reported in person-rem) according to work and job functions" e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignmente to various duty functions may be estimated I-

- - based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge -

- measurements. Small exposures totalling less than 20% of the individual iotal dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent received from extemal sources should be assigned to specific major work functions; I

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'A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

• 'This tabulation supplements the requirements of 10 CFR 20.2206.

COMANCHE PEAK - UNITS 1 AND 2 6-7a Unit 1 - AmendmentNo. 5+,62 Amendment No. 07,48

j-ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 9)

Limitations on the annual and quarterly doses to a MEMBER OF 4

l THE PUBLIC from lodine-131, lodine-133, tritium, and all radionucliaes in particulate form with half-lives greater than 8 days in gaseous effluents

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released from each unit to areas beyond the SITE BOUNARY confonning l

to Appendix 1 to 10 CFR 50, and i

10)

Limitations on the annual dose or dose commitment to any MEMBER OF i

THE PUBLIC due to releases of radioactivity and to radiation from uranium j.

fuel cycle sources conforming to 40 CFR 190.

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f.

Configuration Risk Management Program (CRMP) l The Configuration Risk Management Program (CRMP) provides a proceduralized risk-informed assessment to manage the risk associated with equipment i

inoperability. The program applies to technical specification structures, systemt, or

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components for which a risk informed Allowed Outage Time has been granted.

The program shallinclude the following elements:

5 1)

Provisions for the control and implementation of a Level 1, at-power intemal i

events PRA informed methodology. The assessment shall be capable of evaluating the applicable plant configuration.

i 2)

Provisions for performing an assessment prior to entering the LCO Action for 8

j preplanned activities.

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3)

Provisions for performing an assessment after entering the LCO Action for unplanned entryinto the LCO Action.

4)

Provisions for assessing the need for additional actions after the discovery of 1Z _ additional equipment out of service conditions while in the LCOJction. ~ ~ ~

i 5)

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

g.

Containment Leakaae Rate Testina Proaram 7

A program shall be established to implement the leakage rate testing of the containment as required by 10CFR50.54(o) and 10CFR50, Appendix J, Option B,

- as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program, dated September 1995". _ _

The peak calculated containment intemal pressure for the design basis loss of coolant accident, P., is 48.3 psig.

The maximum allowable containment leakage rate, L, at P., shall be 0.10% of containment air weight per day.

Leakage mte acceptance criteria are:

1.

Containment leakage rate acceptanco criterion is s 1.0 L,. During the first unit startup following testing in accordance with this program, the leakage COMANCHE PEAK-UNITS 1 AND 2 6-7 Unit 1 - Amendment No.14,42.00,01,62 i

Unit 2 - Amendment No. 96;96p% 48

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