Text

Amends 53 & 39 to Licenses DPR-58 & DPR 74,respectively, Incorporating Std Radiological Safety Tech Specs & Recognizing Installation of Equipment for Generation & Surveillance of Low Temp Pressurizer Protection Sys
	ML20054D346
Person / Time
Site:	Cook
Issue date:	03/11/1982
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17319B298	List: ML17319B297; ML17319B299; ML20054D346;
References
	NUDOCS 8204220608
	Download: ML20054D346 (77)
	v • d • e

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ff UNITED STATES y

3e(, i NUCLEAR REGULATORY COMMISSION r, %} {2]a j-wAsmNGTON, D. C. 20655 s.

INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.53 License No. DPR-58 1.

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

A.

The application for amendment by Indiana 'and Michigan Electric Company (the licensee) dated January 23, 1979,

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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 Comission; 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 Comission's regulations; D.

The issuance of this amendment will nct be inimical to the comon defense and security or to the health and safety of the public; and i

E.

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

8204220608 820311 PDR ADOCK 05000315 P

PDR

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-58 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 53, 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 its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION R\\

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' Operating Reactors nch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: March 11,1982

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NUCLEAR REGULATORY COMMISSION 3 }\\ ] /. j wasmNGTON. D. C. 20555

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INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Arendment No. 39 License No. DPR-74 1.

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

A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated January 23, 1979, complies with the standards dr.d 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 arendrent 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 ragulations; D.

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

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

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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-74 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

In addition, license condition 2.C.(3)(d) has been satisfied with the installation of equipment and this license condition is deleted.

4 This license amendment is effective as of the date of its i,ssuance, kORTHENUCLEARREGULATORYC0." MISSION j, )

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hteven gI, Chief \\

Operating Reactors Branc} #1 Division of Licensing %

Attachment-Changes to the Technical Specifications Date of Issuance: March 11,1982 l

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o ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 53 TO FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Revise Appendix A as follows:

Remove Pages Insert Pages y

v vi vi*

xi xi xii xii*

3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12*

3/4 4-3 3/4 4-3 3/4 4-4 3/4 4-4*

3/4 4-31 thru 3/4 4-42 3/4 4-31 thru 3/4 4-44 3/4 5-7 3/4 5-7 3/4 5-8 3/4 5-8 3/4 5-9 3/4 5-9 3/4 5-10 3/4 5-10 3/4 5-11 B3/4 1-3 B3/4 1-3 B3/4 1-4 B3/4 1-4 B3/4 4-1 B3/4 4-1

)

B3/4 4-2 B3/4 4-2 B3/4 4-2a B3/4 4-3 B3/4 4-3 B3/4 4-4 B3/4 4-4 83/4 4-11 B3/4 4-11 B3/4 4-12 B3/4 4-12 B3/4 4-13 83/4 5-1 B3/4 5-1*

B3/4 5-2 B3/4 5-2 B3/4 5-3 B3/4 5-3

no change - new pages included for convenience.

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INDEX LIMITING CONDITIONS FOR OPERATION & SURVEILLANCE REQUIREMENTS SECTION P_ age 3/4.4.6 REACTOR COOLANT SYSTEM LFAKAGE Leakage Detection Systems..............................

3/4 4-14 Operational Leakage....................................

3/4 4-16 3/4.4.7 CHEMISTRY..............................................

3/4 4-18 3/4.4.8 SPECIFIC ACTIVITY......................................

3/4 4-21 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................

3/4 4-25 Pressurizer............................................

3/4 4-30 Overpressure Protection Systems........................

3/4 4-31 3/4.4.10 STRUCTURAL INTEGRITY...................................

3/4 4-33 3/4.4.11 RELIEF VALVES - 0PERATING..............................

3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS...........................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg 1 350*F.........................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350 F........................

3/4 5-7 3/4.5.4 BORON INJECTION SYSTEM Baron Injection Tank...................................

3/4 5-9 Heat Tracing...........................................

3/4 5-10 3/4.5.5 REFUELING WATER STORAGE TANK...........................

3/4 5-11 D. C. COOK - UNIT 1 V

Amendment No. 53

J INDEX LIMITING CONDITIONS FOR OPERATION & SURVEILLANCE REQUIREMENTS SECTION P3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT

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Containment Integrity..................................

3/4 6-1 Containment Leakage....................................

3/4 6-2 Containmeat Air Locks..................................

3/4 6-4 Internal Pressure......................................

3/4 6-6 Air Temperature........................................

3/4 6-7 Containment Structural Integrity.......................

3/4 6-9 3/4.6.2 DEPRESSuRIZATION AND COOLING SYSTEMS Containment Spray System...............................

3/4 6-10 Spray Additive System..................................

3/4 6-12 3/4.6.3 CONTAINMENT ISO LATION VALVES...........................

3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.....................................

3/4 6-23 Electric Hydrogen Recombiners..........................

3/4 6-24 3/4.6.5 ICE CONDENSER Ice Bed................................................

3/4 6-26 Ice Bed Temperature Monitoring System..................

3/4 6-28 Ice Condenser Doors....................................

3/4 6-30 Inlet Door Position Monitoring System..................

3/4 6-33 Divider Barrier Personnel Access Doors and Equipment Hatches....................................

3/4 6-34 Containment Air Recirculation Systems..................

3/4 6-35 Floor Drains...........................................

3/4 6-36 Refueling Canal Drains.................................

3/4 6-37 Divider Barrier Sea1...................................

3/4 6-38 D. C. COOK - UNIT 1 VI Amendment No. 53 i

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INDEX i

BASES SECTION P,,ag.e.

3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............................

B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION...............

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION..............................

B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS..................................

B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES...............................

B 3/4 4-1 3/4.4.4 PRESSURIZER.............................................

B 3/4 4-2 B 3/4 4-2 3/4.4.5 STEAM GENERATORS..............

3/4.4.6 REACTOR COOLANT SYSTEM LL'U s B 3/4 4-3 3/4.4.7 CHEMISTRY...............................................

B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.................................-.....

B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................

B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4'4-12 3/4.4.11 RELIEF VALVES..........................................

B 3/4 4-13 D. C. COOK - UNIT 1 XI Amendment No. 53 4

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INDEX BASES SECTION M

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS.............................................

B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS..............................

B 3/4 5-1 3/4.5 4 BORON INJECTION SYSTEM...................................

B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK (RWST)......................

8 3/4 5-3 i

j D. C. COOK - UNIT 1 XII Amendment No. 53

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REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an l

OPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE a.

ALTERATIONS or positive reactivity.

b.

With more than one charging pump OPERABLE or with a safety injection pump (s) OPERABLE when the temperature of any RCS cold leg is less than or equal to 188*F, unless the reactor vessel head is removed, remove the additional charging pump (s) and the safety injection pump (s~) motor circuit breakers from the electrical power circuit within one hour.

c.

The provisions of Specification 3.0.3 ar not applicable.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE at least once per 31 days by:

a.

Starting (unless already operating) the pump from the control room, b.

Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2390 psig, c.

Verifying pump operation for at least 15 minutes, and d.

Verifying that the pump is aligned to receive electrical power from an OPERABLE emergency bus.

4.1.2.3.2 All charging pumps and safety injection pumps, excluding the above required OPERABLE charging pump, shall be demonstrated inoperable by verifying that the motor circuit breakers have been removed from their electrical power supply circuits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except when:

a.

The reactor vessel head is removed, or b.

The temperature of all RCS cold legs is greater than 188 F.

D. C. COOK - UNIT 1 3/4 1-11 Amendment No. 53

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two 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 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS by:

Starting (unless already operating) each pump from the control room, a.

b.

Verifying, that on recirculation flow, each pump develops a discharge pressure of > 2405 psig, and Verifying that each pump operates for at least 15 minutes.

c.

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0. C. COOK - UNIT 1 3/4 1 Amendment No. 53

REACTOR COOLANT SYSTEM ACTION (Continued)

Below P-7:#

two re$$(o>r co.0, operation below P-7 may proceed provided at least With K 1

a.

olant loops and associated pumps are in operation.

coolan&7fo<op.0,operationmayproceedprovidedatleastonereactor With K l

b.

is in operation with an associated reactor coolant or residual heat removal pump; however, operation for up to 15 minutes with no pump in operation is permissible to accommodate transition between residual heat removal pump and reactor coolant pump operation.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMC.NTS 4.4.1.1.1 With one reactor coolant loop and associated pump not in operation, l

at least once per 7 days determine that:

1 a.

The applicable reactor trip system and/or ESF actuatiun system instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions, and b.

If P-8 interlock setpoint has been reset for 3 loop operation, its setpoint is 5 76% of RATED THERMAL POWER.

4.4.1.1.2 Within 30 minutes prior to the' start of a reactor coolant pump when any RCS cold leg temperature is 1 188 F, verify that:

The temperature of the secondary water of each steam generator is a.

5 50 F above the temperature of each of the RCS cold legs, or b.

The pressurizer water volume is less than 1116 cubic feet, equivalent to less than 62% indicated on the wide range level indicator.

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A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 188 F unless 1) the pressurizer water volume is less than 1116 cubic feet or 2) the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

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0.C. COOK - UNIT,1 3/4 4-3 Amendment No. 53

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REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a t

lift setting of 2485 PSIG + 1%.

APPLICABILITY:

MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4.4.3.

D.C. COOK - UNIT 1 3/4 4-4 Amendment No. 53

a REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

Two power operated relief valves (PORVs) with a lift setting of less a.

than or equal to 435 psig, or b.

One power operated relief valve (PORV) with a lift setting of less than or equal to 435 psig and the RHR safety valve with a lift setting of less than or equal to 450 psig, or A reactor coolant system vent of greater than or equal to 2 square c.

inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 188 F, except when the reactor vessel head is removed.

ACTION:

With two PORV's inoperable or with one PORV inoperable and the RHR a.

safety valve inoperable, either restore the inoperable PORV(s) or RHR safety valve to OPERABLE status within 7 days or depressurize and vent the RCS through an at least 2 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until the

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inoperable PORV or RHR safety valve has been restored to OPERABLE

status, b.

With both PORVs inoperable, depressurize and vent the RCS through an at least 2 square inch vent (s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both PORVs cr one PORV and the RHR safety valve have been restored to OPERABLE status.

c.

In the event either the PORVs, the RHR safety valve or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not applicable.

D.C. COOK - UNIT 1 3/4 4-31 Amendment No.53

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Performance o.f a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV required OPERABLE.

b.

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

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months c.

when the PORV is being used for overpressure protection.

d.

Determining the emergency air tank OPERABLE by verifying:

1.

At least once per 31 days, air tank pressure greater than or equal to 900 psig.

2.

Air tank pressure instrumentation OPERABLE by performance of a:

(a) CHANNEL FUNCTIONAL TEST at least once per 31 days, and (b) CHANNEL CALIBRATION at least once per 18 months, with the low pressure alarm setpoint > 900 psig.

4.4.9.3.2 The RHR safety valve shall be demonstrated OPERABLE by verifying that the RHR system suction is aligned to the RCS loop with the valves in the flow path open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the RHR safety valve is being used for overpressure protection.

4.4.9.3.3 Each PORV and the RHR safety valve shall be demonstrated OPERABLE by testing in accordance with ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition through Summer 1975 Addenda, for Category B and C valves, respectively. Test frequency, procedures and corrective action shall be pursuant to Subsection IWV-3410 and IWV-3510, respectively and shall be performed during COLD SHUTDOWN and REFUELING, respectively.

4.4.9.3.4 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

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

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

0.C. COOK - UNIT 1 3/4 4-32 Amendment No. 53 w.+

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REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of the Reactor Coolant System components (except steam generator tubes) shall be maintained at a level consistent with the acceptance criteria in Specification 4.4.10.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of any of the above components not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10 The following inspection program shall be performed during shutdown:

a.

Inservice Inspections The structural integrity of the Reactor Coolant System components shall be demonstrated by verifying their acceptability when inspected per the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, as outlined by the inspection program shown in Table 4.4-6.

An initial report of any abnormal degradation of the structural integrity of the Reactor Coolant System components detected 1

during the above required inspections shall be made within I

D.C. COOK - UNIT 1 3/4 4-33 Amendment No. 53

a REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 10 days after detection and the detailed report shall be submitted pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.

The Inservice Inspection Program shall be reviewed every 5 years to assure that the equipment, techniques and procedures being utilized are current and applicable. The results of these reviews shall be reported in Special Reports to the Commission pursuant to Specification 6.9.2 within 90 days of completion.

b.

Inspections Following Repairs or Replacements The structurai integrity of the reactor coolant system shall be demonstrated after completion of all repairs and/or replacements to the system by verifying the repairs and/or replacements meet the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972. When repairs and/or replacements are made which involve new strength welds on components greater than 2 inch diameter, the new welds shall receive a surface and 100 percent volumetric examination and meet applicable code requirements. When repairs and/or replacements are made which involve new strength welds on components 2 inch diameter or smaller, the new welds shall receive a surface examination and meet applicable code requirements.

c.

Inspections Following System Opening The structural integrity of the reactor coolant system shall be demonstrated after each closing by performing a leak test, with the system pressurized to at least 2235 psig, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972, and the Pressure / Temperature limits of Specification 3.4.9.1.

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,1 Amendment No. 53 D. C. COOK - UNIT 1 3/4 4-34

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P TABLE 4.4-6 n

INSERVICE INSPECTION PROGRAM 8

5!

ITEM NO.

EXAMINATION COMPONENTS AND PARTS EXTENT OF EXAMINATION DURING (1)

CATEGORY (1) 10 BE EXAMINED METHOD (2) 10-YEAR INTERVAL (3), (4)

REACTOR VESSEL AND CLOSURE HEAD 1.1 A

Longitudinal and circum-Volumetric 5% of length of the circumferential ferential shell welds in welds; 10% of the length of the longi-core region tudinal welds at or near end of the interval.

1.2 B

Lo.;gitudinal and circum-Volumetric 10% of length of longitudinal welds; ferential welds in shell 5% of length of circumferential welds (other than those of excluding those welds in the closure Category A and C).

head that lie within the CRD shroud w

3 assembly and those welds in the lower head that lie within the area contain-a a

ing the in-core instrumentation pene-trations.

Required amount of welds will be examined at or near the end of the interval.

1.3 C

Vessel-to-Flange and Head-Volumetric Cumulative, 100% coverage, of the sub-i to-Flange circumferential ject welds which are available for welds examination during refueling outage operations.

i

@f 1.4 D

Primary nozzle-to-vessel Volumetric Inspection of all coolant outlet nozzle-welds and nozzle-to to-shell welds and inner radius sections

((

vessel inside radiused by the sixth refueling outage and inspec-

's section tion of all the coolant inlet nozzle-to-shell welds and inner radius sections by end of tenth year.

1 4

P TABLE 4.4-6 (Continued)

ITEM NO.

EXAMINATION COMPONENTS AND paris EXTENT OF EXAMINATION DURING 8

(1)

CATEGORY (1)

TO BE EXAMINED METHOD (2) 10-YEAR INTERVAL (3),(4)

E 1.5 E-1 Vessel penetrations, in-None, Penetrations will be included ciuding control rod drive in Category E-2.

c:

E penetrations and control

-d rod housing pressure boundary welds.

H 3

1.6 E-2 Vessel penetrations, in-Visual Cumulative 25% of the control rod drive cluding control rod drive and of the in-core instrumentation penetrations ano control penetrations will be visually inspected rod housing pressure for leakage, j

boundary welds.

1.7 F

Primary nozzles to safe

Visual, All of the dissimilar metal welds on the m;;

end welds surface and vessel nozzles will be examined.

j volumetric 1

1.8 G-1 Closure Studs and nuts Volumetric Cumulative 100% of the vessel flange i

and visual being examined at each refueling outage.

or surface j

1.9 G-1 Ligaments between Volumetric Cumulative, 100% of the vessel flange threaded stud holes bolt ligaments will be examined at the same time as the flange weld of Item No. 1.3.

1 g

1.10 G-1 Closure washers, Visual Cumulative, 100% with approximately 10%

g bushings being examined at each refueling outage.

Eg 1.11 G-2 Pressure retaining None, there is no bolting 2-inch and bolting less in diameter.

r*

5 1.12 11 Integrally welded None, support welds are included in El vessel supports inspection of category D.

6

4

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TABLE 4.4-6 (Continued)

ITEM NO.

EXAMINATION LOMPONENTS AND PARIS EXTENT OF EXAMINATION DURING 8

(1)

CATEGORY (1)

TO DE EXAMINED METHOD (2) 10-YEAR INTERVAL (3),(4)

E 1.13 1-1 Closure llead cladding visual and 100% of at least 6 patches (each 36 i

i surface or square inches) evenly distributed in c=

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volumetric vessel head.

H 1.14 I-l Vessel cladding Visual or 100% of at least 6 patches (each 36 volumetric square inches) evenly distributed in the vessel shell.

1.15 N

Internal surfaces and Visual An examination will be made of the internals and integrally interior surfaces made available by

i welded internal supports refueling outage operations at the 1st refueling cycle.

This will be repeated at approximately 3 year intervals with u,

3:

the amount of inspection being depen-i dent upon results of the first inspec-J, tion.

The space below the reactor

'd core not available during normal refueling outages will be inspected d

once at or near the end of the interval.

4 PRESSURIZER 2.1 B

Longitudinal and circum-Visual and 10% of length of each longitudinal and ferential welds volumetric 5% of length of each circumferential weld excluding the heads of the Unit 1

((

vessel.

All of the inner radii for the upper 2.2 D

Nozzle-to-vessel welds Volumetric and nozzle-to-vessel head shall be inspected.

inside radiused section j

N

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TABLE 4.4-6 (Continued)

ITEM NO.

EXAMIhATION COMPONENTS AND PARIS EXTENT OF EXAMINATION DURING 8

__f l )

CATEGORY (1)

TO BE EXAMINED METHOD (2) 10-YEAR INTERVAL (3),(4)

S 2.3 E-1 lleater connections None, each of the penetrations in the pressurizer for heater connections c-33 meet the exclusion criteria of para-graph 15-121 of Section XI.

~4 w

2.4 E-2 lleater connections Visual A cumulative total of 25% of the heater connections and of the instru-ment and sample penetrations shall be visually inspected for leakage, f

Pressure containing Volumetric 100% of the dissimilar metal weld.

dissimilar metal welds.

2.5 G-1 Pressure retaining bolting None, there is no bolting 2 inch and larger in diameter.

O*

2.6 G-2 Pressure retaining bolting Visual Cumulative 100% by end of interval.

2.7 ll Integrally welded vessel Visual and None, on Unit No. 1.

supports Volumetric 2.8 I-2 Vessel Cladding Visual or A selected patch of the pressurizer Volumetric ~

cladding shall be inspected by the end of the interval.

y STEAM GENERATORS a.

it 3.1 B

Longitudinal and Circum-Visual and 5% of the length of all of the welds ferential welds, includ-Volumetric joining the primary head to the tube ing tube sheet-to-head or

sheet, y

shell welds on the primary side.

t 6

9 E'

TABLE 4.4-6 (Continued)

ITEM NO.

EXAMINATION COMPONENTS AND PARTS EXTENT OF EXAMINATION DURING Q

(1)

CATEGORY (1) 10 BE EXAMINED METil0D (2) 10-YEAR INTERVAL (3),(4) x i

3.2 D

Primary nozzle-to-vessel Volumetric None, on nozzle-to-vessel head welds, c:

head welds and nozzle-to-100% of the inner radius section of 35 head inside radiused the nozzle-to-vessel juncture if section radiation levels permit and a genera-tor is opened for other reasons.

3.3 F

Primary Nozzle-to-safe

Visual, All dissimilar metal welds on the end welds surface and steam generators.

~

volumetric 3.4 G-1 Pressure Retaining None, there is no bolting 2 inches or Bolting larger in diameter.

3.5 G-2 Pressure retaining Visual Cumulative 100% by end of interval.

9 bniting w

I 3.6 11 Integrally welded vessel None, there are no supports integrally supports welded to the steam generator pressure boundary.,

i 3.7 1-2 Vessel Cladding Visual A 36-inch square patch in the heads of all generators if they are opened and decontaminated for other reasons.

1 PIPING PRESSURE BOUNDARY E

4.1 F

Vessel; Pump; and Valve Nane.

For vessel, pump and valve safe-0F Safe-ends-to primary ends see applicable sections.

There 8

pipe welds and safe-ends are no branch piping safe ends.

[

in branch piping welds.

?

10

P TABLE 4.4-6 (Continued) n ITEM NO.

EXAMINATION COMPONENTS AND PARTS EXTENT OF EXAMINATION DURING (1)

CATEGORY (1)

TO BE EXAMINED METHOD (2) 10-YEAR INTERVAL (3),(4)

S 4.2 J-l Circumferential and lon-Visual and Cumulative 25% of the butt welds in gitudinal pipe welds and Volumetric the piping system, including one c

'i branch pipe connection foot of any longitudinal weld on

-d welds larger than 4 inches either side of the butt weld by H

in diameter the end of the interval.

Cumulative 25% of the pipe branch connection welded joints including the weld metal and the base metal for one pipe wall thickness beyond the edge of the weld.

4.3 G-1 Pressure Retaining None, there is no bolting 2-inch and Bolting larger in diameter.

w 2

4.4 G-2 Pressure Retaining Visual 100% of bolting by end of interval.

3 A

Bolting O

I 4.5 K-1 Integrally Welded Visual and Cumulative 25% by end of interval.

Supports Volumetric 4.6 K-2 Piping Supports and Visual Cumulative 100% of supports by end of Hangers interval.

4.7 J-2 Circuferential and long-Visual Welds in piping excluded from the itudinal pipe welds and examination by 15-121 will be examined g

branch pipe connection whenever the system boundary is sub-

'j welds jected to a hydrotest per the require-

((

ments of paragraph IS-521.

Er*

4.8 J-1 Socket welds and pipe Vicual and Cumulative 25% of the pipe branch jf branch connections welds Surface connection weld joints including the 4 in. dia and smaller.

weld metal and the base metal for one wall thickness beyond the edge of the weld.

e i

E' TABLE 4.4-6 (Continued)

ITEM NO.

EXAMINATION COMPONENTS AND PARTS EXTENT OF EXAMINATION DURING (1)

CATEGORY (1)

TO BE EXAMINED METHOD (2) 10-YEAR INTERVAL (3),(4)

PUMP PRESSURE B0UNDARY AND PUMP FLYWHEELS c3 5.1 L-1 Pump Casing Welds Visual 100% inspection of the pressure con-

-d taining welds will be made on one reactor coolant pump.

5.2 L-2 Pump Casing Visual Cumulative 100% of available inner surfaces of one rector coolant pump.

5.3 F

Nozzle-to-Safe end None, there are no dissimilar metal Welds welds in the pump.

5.4 G-1 Pressure Retaining Visual and Cumulative 100% of the bolting by the w

30 Bolting Volumetric end of interval.

1 5.5 G-2 Pressure Retaining Visual Cumulative 100% of the bolting by the Bolting end of interval.

5.6 K-1 Integrally Welded Visual and Cumulative 25% by end of interval.

~

supports Volumetric 5.7 K-2 Supports and Hangers Visual Cumulative 100% of the supports by end of interval.

5.8 5.8.1 Exposed Surface of fly-Surface Cumulative 100% of the exposed surface RI wheel and keyways with of all Reactor Coolant Pumps.

Flywheel in place

{

9 E

5.8.2 Exposed surface Flywheel Volumetric Cumulative 100% of the exposed surface and Keyways of all Reactor Coolant Pumps.

?

O O

E' TABLE 4.4-6 (Continued)

ITEM NO.

EXAMINATION COMPONENTS AND PARTS EXTENT OF EXAMINATION DURING 8

(1)

CATEGORY (1)

TO BE EXAMINED METil0D (2) 10-YEAR INTERVAL (3),(4)

E VALVE PRESSURE BOUNDARY C

25 6.1 M-1 Valve Body Welds None, there are no valves with pressure retaining welds in the valve body,

-d w

6.2 M-2 Valve Bodies Visual Cumulative 100% of one valve of each type by the end of the interval.

6.3 F

Valve-to-Safe-End None, there are no valves with bolting Welds system with dissimilar metal welds.

6.4 G-1 Pressure-Retaining Bolting None, there are no valves with bolting 2-inch and larger in diameter.

u, D

6.5 G-2 Pressure Retaining Visual Cumulative 100% of the bolting by the i

Bolting end of the interval.

m 6.6 K-1 Integrally Welded None, there are no valves with inte-Supports grally welded supports.

6.7 K-2 Supports and llangers Visual Cumulative, 100% of the supports and hangers by the end of the interval.

TABLE NOTATION y

(1) Item numbers (except 5.8) and categories refer to Tables IS-261 of Section XI ASME B&PV Code 1970 g

Edition Including Winter Addenda to 1971.

R g

(2) Examination techniques and procedures refer to Paragraph 15-210 of Section XI (See above).

p (3) Inspection interval per Paragraph 15-242 of Section XI (See above).

gn (4) All inspections subject to revision based on pre-and post-operational inspection experience.

e G

i REACTOR COOLANT SYSTEM RELIEF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.11 Three power operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s)

I and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either (1) restore the block valve (s) to OPERABLE status, or (2) close the block valve (s) and remove power from the block valve (s'), or (3) close the associated PORV(s) and remove power from the associated solenoid valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .*

4 c.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4

4.4.11.1 Each of the three PORVs shall be demonstrated OPERABLE:

a.

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

At least once per 18 months by performance of a CHANNEL CALIBRATION.

When ACTION 3.4.11.b.(3) is applied, no report pursuant to Specification 6.9.1.9 is required for the PORV.

D.C. COOK - UNIT 1 3/4 4-43 Amendment No. 53 l

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.11.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel. The block valve (s) do not have to be listed nor is a report required pursuant to Specification 6.9.1.9 when ACTION 3.4.11.a is applied.

4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel while the emergency buses are energized by the onsite diesel generators and onsite plant batteries. This testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.b and 4.8.2.3.2.c.

D.C. COOK - UNIT 1 3/4 4-44 Amendment No. 53 l

l l

L

o EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg < 350 F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump,#

,i b.

One OPERABLE residual heat removal heat exchanger, c.

One OPERABLE residual heat removal pump, and d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODE 4.

ACTION:

a.

With no ECCS subsystem OPERABLE because of the inoperability of,either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or main-tain the Reactor Coolant System T less than 350 F by use of avg alternate heat removal methods.

With more than one charging pump OPERABLE or with a safety injection c.

pump (s) 0FERABLE when the temperature of any RCS cold leg is less than or equal to 188 F, remove the additional charging pump (s) and the safety injection pump (s) motor circuit breakers from the electrical power circuit within I hour.

d.

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 describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.

l

A maximum of one centrifugal charging pump shall be OPERABLE and both safety injection pumps shall be inoperable whenever the temperature of one or more of the RCS cold legs is less than or equal to 188 F.

D. C. COOK - UNIT 1 3/4 5-7 Amendment No. 53

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

4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the RCS cold legs is less than or equal to 188 F as determined at least once per hour when any RCS cold leg temperature is between 188 F and 200 F.

i i

I D. C. COOK - UNIT 1 3/4 5-8 Amendment No. 53 lt.

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 BORON INJECTION SYSTEM BORON INJECTION TANK LIMITING CONDITION FOR OPERATION 3.5.4.1 The boron injection tank shall be OPERABLE with:

a.

A minimum contained volume of 900 gallons of borated water, b.

Between 20,000 and 22,500 ppm of boron, and c.

A minimum solution temperature of 145 F.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the boron injection tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% Ak/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.4.1 The boron injection tank shall be demonstrated OPERABLE by:

a.

Verifying the water level through a recirculation flow test at least once per 7 days, b.

Verifying the boron concentration of the water in the tank at least once per 7 days, and c.

Verifying the water temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. C. COOK - UNIT 1 3/4 5-9 Amendment No. 53

o EMERGENCY CORE COOLING SYSTEMS HEAT TRACING LIMITING CONDITION FOR OPERATION 3.5.4.2 At least two independent channels of heat tracing shall be OPERABLE for the boron injection tank and for the heat traced portions of the associated flow paths.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With only one channel of heat tracing on either the boron injection tank or on the heat traced portion of an associated flow path OPERABLE, operation may continue for up to 30 days provided the tank and flow path temperatures are verified to be 1145*F at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5110VEILLANCE REQUIREMENTS 4.5.4.2 Each heat tracing channel for the boron injection tank and associated flow path shall be demonstrated OPERABLE:

a.

At least once per 31 days by energizing each heat tracing channel,.

and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the tank and flow path temperatures to be 1 145*F.

The tank temperature shall be determined by measurement. The flow path temperature shall be determined by either measurement or recirculation flow until establishment of equilibrium temperatures within the tank.

l l

0. C. COOK - UNIT 1 3/4 5-10 Amendment No. 53

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=. = +- *~,

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o EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained volume of 350,000 gallons of borated water.

b.

A minimum boron concentration of 1950 ppm, and c.

A minimum water temperature of 70 F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the water level in the tank, and 2.

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is less than 70 F.

1 l

D. C. COOK - UNIT 1 3/4 5-11 Amendment No.ai l

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o

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l REACTIVITY CONTROL SYSTEMS BASES B0 RATION SYSTEMS (Continued)

With the RCS average temperature above 200 F, one 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 change in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injec-tion pumps, except the required OPERABLE charging pump, to be inoperable below 188 F, unless the reactor vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200*F to 140 F.

This condition requires either 835 gallons of 20,000 ppm borated water from the boric acid storage tanks of 9690 gallons of 1950 ppm borated water from the refueling water storage tank.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) limit the potential effects of rod ejection accident.

OPERABILITY of the control rod position indicators is required to determine cor trol rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restric-tions provide assurance of fuel rod integrity during continued operation.

The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumptions used in the accident analysis for a rod ejection accident.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with T,yg > 541 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

t D. C. COOK - UNIT 1 8 3/4 1-3 Amendment No. 53

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued)

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

These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

t 1

D. C. COOK - UNIT 1-B 3/4 1-4 Amendment No. 53

o 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant loop not in operation, THERMAL POWER is restricted to < 51 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset.

Either action ensures that the DNBR will be maintained above 1.30.

A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (51 percent of RATED THERMAL POWER).

A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a RHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs less than or equal to 188 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point.

l The relief capacity of a single safety valve is adequate to relieve any over-l pressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

D. C. COOK - UNIT 1 B 3/4 4-1 Amendment No. 53

REACTOR COOLANT SYSTEM BASES During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maxi-mum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves..or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code, 1974 Edition.

3/4.4.4 PRESSURIliR A steam bubble in the pressurizer ensures that the RCS is not a

, hydraulically solid system and is capable of accommodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.

The powe'r operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The requirement that 150 kW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation conditions.

D. C. COOK - UNIT 1 B 3/4 4-2 Amendment No, 53

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection.of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-sion pursuant to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

D. C. COOK - UNIT 1 B 3/4 4-2a Amendment No. 53 u m

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O REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitations provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 52 GPM.

This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

D. C. COOK - UNIT 1 B 3/4 4-3 Amendment No, 53 w,~,

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REACTOR COOLANT SYSTEM BASES The total steam generator tube leakage limit of 1 GPM for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isciated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion pro-tection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking correc-tive actions to restore the contaminant concentrations to within the Steady State Limits.

D. C. COOK ~ UNIT 1 B 3/4 4-4 Amendment No. 53 m.-e

REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been-tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based NDT.

upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT at the end of 5 EFPY, as well as adjustments for possible errors in NOT the pressure and temperature sensing instruments.

The actual shift in RT f the vessel material will be established NDT periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens in-stalled near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The heatup and cooldown curves must be recalculated when the ART determined NDT from the surveillance capsule is different from the calculated ART f r the NDT equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum tertperature requirements of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of ippendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 188*F.

Either PORV or RHR safety valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tempera-ture of the steam generator less than or equal to 50*F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.

D. C. COOK - UNIT 1 B 3/4 4-11 Amendment No. 53

~

a REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The required inspection programs for the Reactor Coolant System components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

To the extent applicable, the inspection program for the Reactor Coolant System components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code " Inservice Inspection of Nuclear Reactor Coolant Systems," 1971 Edition and Addenda through Winter 1972.

All areas scheduled for volumetric examination have been pre-service mapped using equipment, techniques and procedures anticipated for use during post-opera-tion examinations. To assure that consideration is given to the use of new or improved inspection equipment, techniques and procedures, the Inservice Inspection Program will be periodically reviewed on a 5 year basis.

The use of conventional nondestructive, direct visual and remote visual test techniques can be applied to the inspection of most reactor coolant loop components except the reactor vessel.

The reactor vessel requires special consideration because of the radiation levels and the requirement for remote underwater accessibility.

The techniques anticipated for inservice inspection include visual inspections, ultrasonic, radiographic, magnetic particle and dye penetrant testing of selected parts.

The nondestructive testing for repairs on ccmponents greater than 2 inches diameter gives a high degree of confidence in the integrity of the system, and will detect any significant defects in and near the new welds.

Repairs on components 2 inches in diameter or smaller receive a surface examination which assures a similar standard of integrity.

In each case, the leak test will ensure leak tightness during normal operation.

For normal opening and reclosing, the structural integrity of the Reactor Coolant System is unchanged.

Therefore, satisfactory performance of a system leak test at 2235 psig following each opening and subsequent reclosing is acceptable demonstration of the system's structural integrity.

These leak tests will be conducted within the pressure-temperature limitations for Inservice Leak and Hydrostatic Testing and Figure 3.4-1.

1 D. C. COOK - UNIT 1 B 3/4 4-12 Amendment No. 53

o REACTOR COOLANT SYSTEM I

BASES 3/4.4.11 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should the relief valve become inoperable.

The electrical power for both the j

relief valves and the block valves is supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

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I i

i D. C. COOK - UNIT 1 a 3/4 4-13 Amendment No. 53 m..

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3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each RCS accumulator ensures that a sufficient volume of borated water will be immediately 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 accumulator power operated isolation valves are considered to be " operating bypasses" 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 required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

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 assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying s 'ficient core cooling to limit the peak cladding temperatures within acceptable 1.mits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.

D. C. COOK - UNIT 1 B 3/4 5-1 Amendment No. 53

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EMERGENCY CORE COOLING SYSTEMS BASES With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injec-tion pumps, except the required OPERABLE charging pump, to be inoperable below 188*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

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 condi-tions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assump-tions 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.

3/4.S.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown.

RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.

The limits on injection tank minimum contained volume and boron concentration ensure that the assumptions used in the steam line break analysis are met.

The OPERABILITY of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron salution will be maintained above the solubility limit of 135 F at 21000 pom boron.

)

i D. C. COOK - UNIT 1 B 3/4 5-2 Arendment No. 53

m EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.5 REFUELING WATER STORAGE TANK 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 baron concentration ensure that 1) suffi-cient 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.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution'of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The ECCS analyses to determine F limits in Specifications 3.2.2 and 3.2.6 0

assumed a RWST water temperature 0f 70 F.

This temperature value of the RWST water determines that of the spray water initially delivered to the containment following LOCA.

It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance with the provisions of 10 CFR 50.46 and Appendix K to 10 CFR 50.

D. C. COOK - UNIT 1 B 3/4 5-3 Amendment No. 53 1

J h.-

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 39 TO FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Revise Appendix A as follows:

Remove Pages Insert Pages v

v vi vi*

xii xii 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12*

3/4 4-1 3/4 4-1*

3/4 4-2 3/4 4-2 3/4 4-3 3/4 4-3 3/4 4-4 3/4 4-4*

3/4 4-29 thru 3/4 4-31 3/4 4-29 thru 3/4 4-33 3/4 5-7 thru 3/4 5-10 3/4 5-7 thru 3/4 5-11 B3/4 1-3 B3/4 1-3 B3/4 1-4 B3/4 1-4*

B3/4 4-1 B3/4 4-1 B3/4 4-2 B3/4 4-2 B3/4 4-2a B3/4 4-3 B3/4 4-3 B3/4 4-4 B3/4 4-4 B3/4 4-9a B3/4 4-9a*

B3/4 4-10 B3/4 4-10 B3/4 4-11 B3/4 4-11 B3/4 5-1 83/4 5-1 B3/4 5-2 B3/4 5-2

no change - pages included for convenience.

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INDEX LIMITING CONDITIONS FOR OPERATION & SURVEILLANCE REQUIREMENTS SECTION M

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..............................

3/4 4-14 Operational Leakage....................................

3/4 4-15 3/4.4.7 CHEMISTRY..............................................

3/4 4-17 3/4.4.8 SPECIFIC ACTIVITY......................................

3/4 4-20 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................

3/4 4-24 Pressurizer............................................

3/4 4-28 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components..................

3/4 4-31 3/4.4.11 RELIEF VALVES - 0PERATING..............................

3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS...........................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 350 F.........................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350 F.........................

3/4 5-7 3/4.5.4 BORON INJECTION SYSTEM Boron Injection Tank...................................

3/4 5-9 Heat Tracing...........................................

3/4 5-10 3/4.5.5 REFUELING WATER STORAGE TANK...........................

3/4 5-11 D. C. COOK - UNIT 2 V

Amendment No. 39

INDEX LIMITING CONDITIONS FOR OPERATION & SURVEILLANCE REQUIREMENTS SECTION P_aje 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity..................................

3/4 6-1 Containment Leakage....................................

3/4 6-2 Containment Air Locks..................................

3/4 6-4 Internal Pressure......................................

3/4 6-6 Air Temperature........................................

3/4 6-7 Containment Structural Integrity.......................

3/4 6-9 Containment Ventilation System.........................

3/4 6-9a 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System...............................

3/4 6-10 Spray Additive System..................................

3/4 6-11 3/4.6.3 CONTAINMENT ISOLATION VALVES...........................

3/4 6-13 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.....................................

3/4 6-33 Electric Hydrogen Recombiners..........................

3/4 6-34 3/4.6.5 ICE CONDENSER Ice Bed................................................

3/4 6-35 Ice Bed Temperature Monitoring System..................

3/4 6-37 Ice Condenser Doors....................................

3/4 6-39 Inlet Door Posi tion Monitoring System..................

3/4 6-42

~

Divider Barrier Personnel Access Do. ors and Equipment Hatches....................................

3/4 6-43 Containment Air Recirculation Systems..................

3/4 6-44 Floor 0 rains...........................................

3/4 6-45 Refueling Canal Drains.................................

3/4 6-46 Divider Barrier Seal...................................

3/4 6-47 i

D. C. COOK - UNIT 2 VI Amendment No. 39 i

INDEX BASES l

I SECTION P_ age a

j 3/4.4.11 RELIEF VALVES B 3/4 4-11 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) i i

3/4.5.1 ACCUMULATORS.............................................

B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS..............................

B 3/4 5-1 l

3/4.5.4 BORON INJECTION SYSTEM...................................

B 3/4 5-2 l

3/4.5.5 REFUELING WATER STORAGE TANK (RWST)......................

B 3/4 5-2 i

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT......................................

B 3/4 6-1 l

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.....................

B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................

B 3/4 6-3 l

3/4.6.4 COMBUSTIBLE GAS CONTR0L........................

B 3/4 6-4 3/4.6.5 ICE CONDENSER............................................

B 3/4 6-4 l

1 l

D. C. COOK - UNIT 2 XII knendment No. 39

l REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by l

Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

a.

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b.

With more than one charging pump OPERABLE or with a safety injection pump (s) OPERABLE when the temperature of any RCS cold leg is less than or equal to 152*F, unless the reactor vessel head is removed, remove the additional charging pLmp(s) and the safety injection pump (s') motor circuit breakers from the electrical power circuit within one hour, c.

The provisions or Specification 3.0.3 ar not applicable.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by I

verifying, that on recirculation flow, the pump develops a discharge pressure of > 2405 psig when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All charging pumps and safety injection pumps, excluding the above required OPERABLE charging pump, shall be demonstrated inoperable by verifying that'the motor circuit breakers have Leen removed from their electrical power supply circuits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except when:

a.

The reactor vessel head is removed, or b.

The temperature of all RCS cold legs is greater than 152 F.

D. C. COOK - UNIT 2 3/4 1-11 Amendment No. 39

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

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one charging pump OPERABLE, restare at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and barated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging p'mps to OPERABLE status within the next 7 days u

or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge pressure of > 2405 psig when tested pursuant to Specification 4.0.5.

D. C. COOK - UNIT 2 3/4 1-12 Amendment No. 39

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS NORMAL OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.

APPLICABILITY: As noted below, but excluding MdDE 6.*

ACTION:

Above P-7, comply with either of the following ACTIONS:

a.

With one reactor coolant loop and associated pump not in operation, STARTUP and/or continued POWER OPERATION may proceed provided THERMAL POWER is restricted to less than 46%

of RATED THERMAL POWER and the following ESF instrumentation channels associated with the loop not in operation, are placed in their tripped condition within I hour:

1.

T

-- Low-Low channel used in the coincidence circuit wiYRSteamFlow-HighforSafetyInjection.

2.

Steam Line Pressure - Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.

3.

Steam Flow-High Channel used for Safety Injection.

4.

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip all bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).

b.

With one reactor coolant loop and associated pump not in operation, subsequent STARTUP and POWER OPERATION above 46% of RATED THERMAL POWER may proceed provided:

1.

The following actions have been completed with the reactor in at least HOT STANDBY:

a)

Reduce the overtemperature aT trip setpoint to the value specified in Specification 2.2.1 for 3 loop operation.

^$ee Special Test Exception 3.10.4.

D.C. COOK - UNIT 2 3/4 4-1 Amendment No. 39

REACTOR COOLANT SYSTEM ACTION (Continueg b)

Place the following reactor trip system and ESF instrumentation channels, associated with the loop not in oparation, in their tripped conditions:

1)

Overpower AT channel.

2)

Overtemperature AT channel.

3)

T

-- Low-Low channel used in the coincidence cfE8uit with Steam Flow - High for Safety Injection.

4)

Steam Line Pressure - Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.

5)

Steam Flow-High channel used for Safety Injection.

6)

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip.all bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).

c)

Change the P-8 interlock setpoint from the value specified in Table 3.3-1 to 1 76% of RATED THERMAL POWER.

2.

THERMAL POWER is restricted to 1 71% of RATED THERMAL POWER.

Below P-7:#

l With K 1.0, operation may proceed provided at least two reactofbo>olantloopsandassociatedpumpsareinoperation.

a.

7 With K l.0, operation may proceed provided at least one b.

reactof bo<lant loop is in operation with an associated reactor f

o coolant or residual heat removal pump.*

c.

The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

All reactor coolant pumps and residual heat removal pumps may be de-energized for up to I hour, provided no operations are permitted which could cause dilution of the reactor coolant system boron concentration.

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 152*F unless 1) the pressurizer water volume is less than 1116 cubic feet or 2) the secondary water temperature of each steam generator is less than 50%F above each of the RCS cold leg temperatures.

0.C. COOK - UNIT 2 3/4 4-2 Amendment No. 39

c l

1 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 With one reactor coolant loop and associated pump not in operation, at least once per 7 days determine that:

a.

The applicable reactor trip system and/or ESF actuation system instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions, and b.

If the P-8 interlock setpoint has been reset for 3 loop operation, its setpoint is 176% of RATED THERMAL POWER.

4.4.1.1.2 Within 30 minutes prior to the start of a reactor coolant pump when any RCS cold leg temperature is i 152*F, verify that:

a.

The temperature of the secondary water of each steam generator is i 50*F above the temperature of each of the RCS cold legs, or b.

The pressurizer water volume is less than 1116 cubic feet, equivalent to less than 62% indicated on the wide range level indicator.

l D. C. COOK - UNIT 2 3/4 4-3 Amendment No. 39

_ _ _ _ _ _ - _ _ - _ _ _ _ ~

a REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG + 1%.*

~

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

^The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

D.C. COOK - UNIT 2 3/4 4-4 Amendment No. 39

- o REACTOR COOLANT SYSTEM l

OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a.

Two power operated relief valves (PORVs) with a lift setting of less

~

than or equal to 435 psig, or b.

One power operated relief valve (PORV) with a lift setting of less than or equal to 435 psig and the RHR safety valve with a lift setting of less than or equal to 450 psi, or c.

A reactor coolant system vent of greater than or equal to 2 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 152*F, except when the reactor vessel head is removed.

ACTION:

a.

With two PORV's inoperable or with one PORV inoperable and the RHR safety valve inoperable, either restore the inoperable PORV or RHR safety valve to OPERABLE status within 7 days or depressurize and vent the RCS through an at least 2 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until the inoperable PORV or RHR safety valve has been reste ed to OPERABLE status.

b.

With both PORVs inoperable, depressurize and vent the RCS through an at least 2 square inch vent (s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both PORVs or one PORV and the RHR safety valve have been restored to OPERABLE status.

c.

In the event either the PORVs, the RHR safety valve or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not acplicable.

D. C. COOK - UNIT 2 3/4 4-29 Amendment No. 39

o -

o REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

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

b.

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

c.

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

d.

Testing in accordance with the inservice test requirements for ASME Category B valves pursuant to Specification 4.0.5.

e.

Determining the emergency air tank OPERABLE by verifying:

1.

At least once per 31 days, air tank pressure greater than or equal to 900 psig.

2.

Air tank pressure instrumentation OPERABLE by performance of a:

(a) CHANNEL FUNCTIONAL TEST at least once per 31 days, and (b) CHANNEL CALIBRATION at least once per 18 months, with the low pressure alarm setpoint > 900 psig.

4.4.9.3.2 The RHR safety valve shall be demonstrated OPERABLE by:

a.

Verifying that the RHR system suction is aligned to the RCS loop with the valves in the flow path open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the RHR safety valve is being used for overpressure protection.

b.

Testing in accordance with the inservice test requirements for ASME Category C valves pursuant to Specification 4.0.5.

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least cnce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

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

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

D. C. COOK - UNIT 2 3/4 4-30 Amendment No. 39

v.

REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.

APPLICABILITY: ALL MODES ACTION:

a.

With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.

b.

With the structural integrity of any ASME Code Class 2 component (s) not conforming to the abnve requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F.

l c.

With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

D.C. COOK - UNIT 2 3/4 4-31 Amendment No. 39 on-

m REACTOR COOLANT SYSTEM RELIEF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.11 Three power operator relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

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

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either (1) restore the block valve (s) to OPERABLE status, or (2) close the block valve (s) and remove power from the block valve (s-), or (3) close the associated PORV(s) and remove power from the associated solenoid valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .*

c.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.11.1 Each of the three PORVs shall be demonstrated OPERABLE:

a.

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

At least once per 18 months by performance of a CHANNEL CALIBRATION.

When ACTION 3.4.11.b.(3) is applied, no report pursuant to Specification 6.9.1.9 is required for the PORV.

D.C. COOK - UNIT 2 3/4 4-32 Amendment No. 39

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.11.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel. The block valve (s) do not have to be tested nor is a report required pursuant to Specification 6.9.1.9 when ACTION 3.4.11.a is applied.

4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel while the emergency buses are energized by the onsite diesel generators and onsite plant batteries. This testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.b ar.d 4.8.2.3.2.c.

D.C. COOK - UNIT 2 3/4 4-33 Amendment No. 39

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg < 350'F 1

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

[

a.

One OPERABLE centrifugal charging pump,#

1 b.

One OPERABLE residual heat removal heat exchanger, c.

One OPERABLE residual heat removal pump, and d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY:

MODE 4.

ACTION:

i a.

With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within i hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or main-tain the Reactor Coolant System T less than 350 F by use of alternate heat removal methods.

avg c.

With more than one charging pump OPERABLE or with a safety injection pump (s) OPERABLE when the temperature of any RCS cold leg is less i

than or equal to 152 F, remove the additional charging pump (s) and the safety injection pump (s) motor cirr' tit breakers from the electrical power circuit within I hour.

d.

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 describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.

A maximum of one centrifugal charging pump shall be OPERABLE and both safety injection pumps shall be inoperable whenever the temperature of one or more of the RCS cold legs is less than or equal to 152*F.

D. C. COOK - UNIT 2 3/4 5-7 Amendment No.39

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

4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the RCS cold legs is less than or equal to 152*F as determined at least once per hour when any RCS cold leg temperature is between 152*F and 200 F.

D. C. COOK - UNIT 2 3/4 5-T.-

Amendment No. 39

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 BORON INJECTIGN SYSTEM BORON INJECTION TANK LIMITING CONDITION FOR OPERATION 3.5.4.1 The boron injection tank shall be OPERABLE with:

a.

A minimum contained borated. water volume of 900 gallons, b.

Between 20,000 and 22,500 ppm of baron, and c.

A minimum solution temperature of 145*F.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the baron injection tank inoperable, restore the tank to OPERABLE status within I hour or be in HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% ak/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.4.1 The boron injection tank shall be demonstrated OPERABLE by:

a.

Verifying the contained borated water volume at least once per 7 days, b.

Verifying the boron concentration of the water in the tank at least once per 7 days, and c.

Verifying the water temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. C. COOK - UNIT 2 3/4 5-9 Amendment No. 39

EMERGENCY CORE COOLING SYSTEMS HEAT TRACING LIMITING CONDITION FOR OPERATION 3.5.4.2 At least two independent channels of heat tracing shall be OPERABLE for the boron injection tank and for the heat traced portions of the associated flow paths.

~

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With only one channel of heat tracing on either the boron injection tank or on the heat traced portion of an associated flow path OPERABLE, operation may continue for up to 30 days provided the tank and flow path temperatures are verified to be > 145*F at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise, be in HOT SHUTDOWN within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sT SURVEILLANCE REQUIREMENTS 4.5.4.2 Each heat tracing channel for the boron injection tank and associated flow path shall be demonstrated OPERABLE:

a.

At least once per 31 days by energizing each heat tracing channel, and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the tank and flow path temperatures to be > 145*F.

The tank temperature shall be deter-mined by measurement. The flow path temperature shall be determined by either measurement or recirculation flow until establishment of equilibrium temperatures within the tank.

D. C. COOK - UNIT 2 3/4 5-10 Amendment No. 39

,e

.g, a

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A contained borated water volume of between 350,000 and 420,000 gallons, b.

Between 2000 and 2200 ppm of baron, and c.

A minimum water temperature of 80*F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the contained borated water volume in the tank, and 2.

Verifying the boron concentration of the water.

b.~

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is < 80 F.

D. C. COOK - UNIT 2 3/4 5-11 Amendment No. 39

~ _

3/4.1 REACTIVITY CONTROL SYSTEMS BASES With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoper-able. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injec-tion pumps, except the required OPERABLE charging pump, to be inoperable below 152*F, unless the reactor vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% ak/k after xenor.

decay and cooldown to 200 F.

The maximum expected boration capability require-ment occurs at EOL from full power equilibrium xenon conditions and requires 4720 gallons of 20,000 ppm borated water from the boric acid storage tanks or 64,000 gallons of 2000 ppm borated water from the refueling water storage tank.

With the RCS temperature below 200 F, one 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 ALTERA-TIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200 F to 140 F.

This condition requires either 835 gallons of 20,000 ppm borated water from the boric acid storage tanks or 9690 gallons of 1950 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

D. C. COOK - UNIT 2 B 3/4 1-3 knendment No. 39 m-

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated accident analyses.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.

Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restric-tions provide assurance of fuel rod integrity during continued operation.

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

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with T

> 541 F and avg with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

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

These verification frequencies are adequate for assuring that the applicable LC0's are satisfied.

D. C. COOK - UNIT 2 B 3/4 1-4 Amendment i.o.~ 39 >

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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain calculated DNBR above the design DNBR value during Condition I and II events. With one reactor coolant loop not in operation, THERMAL POWER is restricted to < 51 percent of RATED THERMAL POWER until the Overtemperature aT trip is reset.

Either action-ensures that the calculated DNBR will be maintained above the design DNBR value.

A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (51 percent of RATED THERMAL POWER).

A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a RHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold lags less than or equal to 152 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point.

The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

D. C. COOK - UNIT 2 B 3/4 4-1,

Amendment No. 39

REACTOR COOLANT SYSTEM BASES During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maxi-mum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

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

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.

The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The requirement that 150 kW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

D. C. COOK - UNIT 2 B 3/4 4-2 Amendment No. 39

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REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrit/ of this portion of the RCS will be main-tained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant opera-tion would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plug-ging limit of 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-sion pursuant to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

s D. C. COOK - UNIT 2 B 3/4 4-2a.

Amendment No. 39

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure ~ Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.

This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitations provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

D. C. COOK - UNIT 2 B 3/4 4-3 Amendment No. 39

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REACTOR COOLANT SYSTEM BASES The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 52 GPM.

This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The total steam generator tube leakage limit of 1 GPM for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion pro-tection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.

Corrosion stud'es show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking correc-tive actions to restore the contaminant concentrations to within the Steady State Limits.

0. C. COOK - UNIT 2 8 3/4 4-4 Amendment No. 39 e e w

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{j TABLE B 3/4.4-1 (Continued) x REACTOR VESSEL TOUGHNESS i

'c5 MINIMUM

-4 50 FT-LB/35 MIL TEMP F AVG. UPPER SHELF (FT-LB)

Parallel Normal Parallel Normal to Major to Major to Major to Major HEAT MATERIAL.

CU P

NDTT Working Working RT W rking Working COMPONENT NO.

TYPE F

Direction Direction

kDT Direction Direction I

BOT. HD.

B0018-1B A533BCL1 NA NA

-50

-11 9*

-50 177 115*

BOT. PEEL SEG. C5823-2 A533BCL1 NA NA

-10 25 45*

-10 129 84*

BOT. PEEL SEG. A4957-3 A533BCL1 NA NA

-10 0

20*

-10 149 97*

a3 WELD L5WC39 C5592-1 TO w

I 3:

C5521-2

.05.019

-40 NA 25

-35 NA 97 HAZ C5592-1 T0 C5521-2 NA NA

-10 NA 80 20 NA 109 g

j '

Estimate Based on USAEC Regulatory Standard Review Plan, Section 5.3.2 and MTEB 5-2.

i (A) 60% Shear (B) 70% Shear NA - Not available or not applicable as appropriate.

F 3

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REACTOR COOLANT SYSTEM BASES The actual shift in RT f the vessel material will be established NDT periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens in-stalled near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The hcatup and cooldown curves must be recalculated when the ART determined NDT from the surveillance capsule is different from the calculated ART f r the NDT equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria' assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 152 F.

Either PORV or RHR safety valve has adequate relieving capability to protect.the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tempera-ture of the steam generator less than or equal to 50 F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.

3/4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

D. C. COOK - UNIT 2 8 3/4 4-10 Amendment no. 39

o 0

i REACTOR COOLANT SYSTEM BASES 3/4.4.11 RELIEF VALVES The power operated relief valves.(PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should the relief valve become inoperable.

The electrical power for both the relief valves and the block valves is supplied from an emergency power source

~

to ensure the ability to seal this possible RCS leakage path.

i D. C. COOK - UNIT 2 B 3/4 4-11 krieadment No. 39 e

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I 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each RCS accumulator ensures that a sufficient volume of borated water will be immediately 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 accumulator power operated isolation valves are considered to be " operating bypasses" 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 requi. red.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mcde where this capability is not required.

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 assuming 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 subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.

D. C. COOK - UNIT 2 B 3/4 5-1 Amendment No. 39

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EMERGENCY CORE COOLING SYSTEMS BASES With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE

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and the Surveillance Requirement to verify all charging pumps and safety injec-tion pumps, except the required OPERABLE charging pump, to be inoperable below 152*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

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 condi-tions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assump-tions 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.

3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown.

RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.

The limits on injection tank minimum contained volume and boron concentration l

ensure that the assumptions used in the steam line break analysis are met.

l The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

l The OPERABILITY of the redundant heat tracing channels associated with the baron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of 135 F at 21000 ppm boron.

0. C. COOK - UNIT 2 B 3/4 5-2 Amendment No. 39

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