Text

Rev 3 to DC Cook Nuclear Plant Unit 1 Inservice Testing Valve Program.
	ML17334B357
Person / Time
Site:	Cook
Issue date:	02/05/1990
From:	; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	;
Shared Package
ML17334B354	List: ML17334B353; ML17334B355; ML17334B356; ML17334B357;
References
	AEP:NRC:0969R, AEP:NRC:969R, PROC-900205-02, NUDOCS 9003090184
	Download: ML17334B357 (459)
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ASME SECTION XI VALVE TEST PRO 2ND TEN YEAR INSPECTION INTERVAL FOR THE D. C. COOK NUCLEAR POWER STATION UNIT NO. 1 Revision No: 3 Date: 2-5-90 TABLE OF CONTENTS

1. INTRODUCTION (FIGURE 1)

2. LIST OF DRAWINGS (FIGURE 2)

3. NOMENCLATURE FOR TEST METHODS - (FIGURE 3)

4. RELIEF REQUEST/COLD SHUTDOWN JUSTIFICATION NOTES

5. ATTACHMENT-A

6. VALVE,

SUMMARY

SHEETS 7 ~ FLOW DIAGRAMS PDR~0~0>84 900

~DDCK OS000S>Z

~

PNU Page 1 of 81

1i' ASME SECTION XI VALVE TEST PR 2ND TEN YEAR INSPECTION INTERVAL FOR THE D. C. COOK NUCLEAR PONER STATION UNIT NO. 1 Revision No: 3 Figure 1 Date: 2-5-90 1NTRODUCTION Valve Testin Pro ram A. The valve test program shall be conducted in accordance with Subsection IWV of Section XI of the 1983 edition of the ASME Boiler and Pressure Vessel Code through Summer 1983 Addenda, except for specific relief requests which are identified in the Valve Summary Sheet.

B. The valve test program is applicable for the second 10 year inspection interval which commences on July 1, 1986.

C. The valve test program was developed employing the classification guidelines contained in 10 CFR 50.'2(v) for Quality Group A and Regulatory Guide 1.26, Revision 3 for Quality Groups B and C.

(Quality Group A is the same as ASME Class 1, Group B is 2, and Group C is 3). NRC staff guidance was provided by memorandum dated January 16, 1978.

D. Figure 2 identifies the system flow diagrams which were used to develop this valve test program.

E' Valve Summary Sheets contain the following:

'": "Y"('

(e.g., 1-5105B-42)

Valve Number: Unique valve number (e.g., 1-DCR-310)

Page 2 of 81

D. C. Cook Nuclear Plan Valve Test Program Revision No: 3 Figure 1 Date: 2-5-90

Revision Number: Any change of valve description, function or test requirement.

pg h REL . Relief and Safety CK Check BF Butterfly GA Gate GL Globe DA Diaphragm 3W Three-Way ND Needle AG Angle BL Ball VB Vacuum Breaker (Reverse Check Valve)

Valve Size: Nominal valve size in inches

Valve Actuator T e: Type of actuator, one of the following:

SA Self Actuated (e.g., CK or REL)

MO Motor Operated A Air Operated M Manual PO Pneumatic SO Solenoid Operated

Flow Dia ram Coordinates: Alpha/Numeric grid location of valve Page 3 of 81

D. C. Cook Nuclear. Plan Valve Test Program Revision No: 3 Figure 1 Date: 2-5-90

Valve osition during normal plant operation or during performance of its safety function, one of the following:

0 Open C Closed 0/C Open/Closed or vice versa

Code Class: ASME Code class of valve, either 1, 2, or 3

Valve Status-A P: Active or passive

~ J 1 defined in IWV-2200 NOTE: Combinations are possible (e.g., AC)

Primar Test Re 'd: Test required per Section XI

Test Performed: Testing that will be performed NOTE: Test nomenclature is explained in Figure 3

Test Mode Test Fre enc : One of the following:

P Every 3 months while system is required to be operable.

C Testing will be performed at cold shutdown frequency (See Note nFn)

R - Testing will be performed at refueling outage frequency.

Page 4 of 81

D. C. Cook Nuclear Plan Valve Test Program Revision No: 3 Figure 1 Date: 2-5-90

Code Relief: Hhether or not a code relief is being requested; will be one of the following:

NO Valve is to be tested per code, no comments.

NO, NOTE X Valve is to be tested per code, but there are comments.

NO, CSZ Y Valve is to be tested per code at a cold shutdown frequency with cold shutdown justification provided in notes.

YES, NOTE Z Code relief is requested. Alternate testing is proposed in lieu of that required by code, the note explains why the code relief is requested.

E. Alternative testing performed on a check valve in accordance with GL-89-04, Attachment 1, Item 42 is indicated under relief request notes. This testing is performed in lieu of stroke testing required by Section XI, IWV-3521. This is accomplished by disassembly method in the following manner:

Disassembl Method The valve bonnet is removed, the disc is manually full stroke exercised and the valve internals are visually examined. The results of this examination are documented. This will be performed on a refueling outage frequency. The valve groupings for sample disassembly is in accordance with GL-89-04.

This alternative testing to be performed for a particular valve is indicated on the valve summary sheets and relief request notes.

Page 5 of 81

D. C. Cook Nuclear Plan Valve Test Program Revision No: 3 Figure 1 Date: 2-5-90 F. Schedulin of Valve Testin at Cold Shutdown Fre enc . Valves tested at a cold shutdown frequency shall be scheduled using the following criteria:

Valve exercising need not be done more often than once every 3 months in case of frequent cold shutdowns.

2 ~ The testing shall commence as soon as the cold shutdown condition is achieved,, but not later than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after shutdown, and continue until complete or the plant is ready to return to power.

3 ~ Completion of all valve testing is not a prerequisite to return to power. Any testing not completed during one cold shutdown should be performed during any subsequent cold shutdowns starting from the last test performed at the previous cold shutdown.

4. For planned cold shutdowns, where ample time is available and testing all the valves identified for, the cold shutdown test frequency in the IST Program will be accomplished, exceptions to the 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months commencement of testing is allowed.

G., The following criteria have been used in developing limiting values of full-stroke time for the power operated valves:

0 Review of valve's design specification and/or manufacturer's test stroke times 0 Review of system response time requirements (Technical Specification, FSAR, etc.)

0 Valve's historical stroke time values at various system conditions Page 6 of 81

D. C. Cook Nuclear Pla Valve Test Program Revision No: 3 Figure 1 Date: 2-5-90 Using the above criteria, the limiting stroke time for each valve is derived as follows:

1. Nominal Valve Stroke Time < 2 Seconds*

Historical Established Recommended Action Time Stroke Time Base Line on Curves (Limiting Stroke Time Ran e in Seconds in Seconds Values in Seconds Base Line Time x 2 + 1 Second=Recommended Action Time or Tech. Spec. Limit, whichever is less.

up to to 1.24 1.0 = 1 x 2 + 1 = 3 Seconds 1.25 to 1.74 1.5 = 1.5 x 2 + 1 = 4 Seconds 1.75 to 2.49 2.0 = 2 x 2 + 1 = 5 Seconds

2. Nominal Valve Stroke Time 3. 0 to 10. 0 Seconds 2.5 to 10.49 3 to 10 Base Line Time x 1.5 =

Action Time

3. Nominal Valve Stroke Time 11.0 Seconds and U 10.5 and Up 11.0 and Up Base Line Time x 1.25 =

Action Time (or 15 seconds, whichever is larger)

Excluding those valves designated as rapid valves per Paragraph "H" noted below.

Page 7 of 81

D. C. Cook Nuclear Plan Valve Test Program Revision No: 3 Figure 1 Date: 2-5-90 The stroke time limiting values for the power operated valves will be controlled via plant Technical Data Book.

H. Stroke Time Measurements for Ra id Fast Actin Valves. In accordance with GL-89-04, Attachment 1, Item 56, power operated valves with normal stroke times'of 2 seconds or less may be assigned 2 seconds limiting values.

second limiting value, it If a valve is assigned a 2 shall be timed only and not trended in accordance with Section XI, IWV-3417a.

second limit, it If the valve exceeds 2 will be declared inoperable and corrective actions will be taken in accordance with IHV-3417(b). The major influence in the stroke time testing of rapid acting valves is the operator's response. Therefore, the timing tolerances are influenced by the operator action and trending is not indicative of valve performance. The valve limiting values will be documented in the plant Technical Data Book. The results of these tests will be documented via plant procedures.

Page 8 of 81

DONALD C. COOK NUCLEAR P 2ND TEN YEAR INSPECTION INTERVAL OF VALVE TEST PROGRAM FOR UNIT 1 LIST OF DRAWINGS Revision No: 3 Figure 2 Date: 2-5-90 SYSTEM FLOW DIAGRAM NO. REVISION NO.

Main Steam 1-5105 29 Main Steam 1-5105B 35 Steam Generating System 1-5105D 1 Feedwater 1-5106 35 Feedwater (Auxiliary) 1-5106A 38 Essential Service Water 1-5113 41 Non-Essential Service Water 1-5114A 31 Station Drainage Containment 1-5124 22 Reactor Coolant 1-5128 19 Reactor Coolant 1-5128A 37 CVCS-Reactor Letdown & Charging 1-5129 31 1-5129A 19 Component Cooling 1-5135 35 Component Cooling 1-5 135A 30 Component Cooling 1-5135B 14 Nuclear Sampling 1-5141 29 Nuclear Sampling 1-5141A 32 Post Accident Sampling-Containment Hydrogen 1-5141D 10 Page 9 of 81

DONALD C. COOK NUCLEAR P 2ND TEN YEAR INSPECTION INTERVAL OF VALVE TEST PROGRAM FOR UNIT 1 LIST OF DRAWINGS Revision No: 3 Figure 2 Date: 2-5-90 SYSTEM FLOW DIAGRAM NO. REVISION NO.

Emergency Core .Cooling (SIS) 1-5142 25 Emergency Core Cooling (RHR) 1-5143 36 Containment Spray 1-5144 28 Containment Penetration & 1-5145 17 Weld Channel Pressurization Ice Condenser Refrigeration 1-51'46B 24 Containment Ventilation 1-5147A 34 Control Room Ventilation 1-5149 20 Emergency Diesel Generator 1-5151A 25 Emergency Diesel Generator 1-5151B 28 Emergency Diesel Generator 1-5151C 26 Emergency Diesel Generator 1-5151D 28 Make-Up Water & Primary Water System 12-5115A 41 Compressed Air System 12-5120B 22-Page 10 of 81

DONALD C. COOK NUCLEAR P 2ND TEN YEAR INSPECTION INTERVAL OF VALVE TEST PROGRAM FOR UNIT 1 LIST OF DRAWINGS Revision No: 3 Figure 2 Date: 2-5-90 SYSTEM FLOW DIAGRAM NO. REVISION NO.

CVCS-Boron Makeup 12-5131 19 Spent Fuel Pit Cooling & Clean-Up 12-5136 25 WDS Vents & Drains 12-5137A 21 Post Accident Liquid & Gas Sampling 12-5141C Post Accident Liquid Sampling Inst. Panels 12-5141F Page 11 of 81

1 I

II

DONALD C. COOK NUCLEAR P~T NOMENCLATURE-.FOR TEST METR%

USED IN COLUMNS FOR. PRIMARY TEST REQUIRED AND TEST PERFORMED UNDER ASME SECTION XI Revision No: 3 Figure 3 Date: 2-5-90 ASME CODE SECTION XI 1 CATEGORY A-B VALVES PARAGRAPH EF-1 Exercise valve (full stroke) for operability quarterly (IWV-3411)

(3 months).

EF-2 -Exercise valve (full stroke) for operability at a cold shutdown (IWV-3412) frequency or refueling outage frequency'as indicated. Code relief requests and/or cold shutdown justification are provided in the corresponding valve notes.

EF-3 Exercise valve (part stroke) for operability quarterly; (IWV-3412) exercise (full stroke) at a cold shutdown frequency or refueling outage frequency as indicated. Justification for exercising the valve at cold shutdown frequency is provided in the corresponding valve notes. Code relief request is provided is deferred to coincide with refueling frequency.

if full stroke test EF-4 Exercise valve (full stroke) for operability prior to return to ( IWV-3416) service EF-5 Valves with remote position indicator shall be observed at least (IWV-3300) once every 2 years.-to verify that valve operation is accurately indicated.

EF-6 This note was intentionally deleted.

EF-7 Exercise valve (with fail-safe actuators) to observe failure (IWV-3415) mode quarterly.

EF-8 Exercise valve (with fail-safe actuators) to observe failure (IWV-3415) mode at a cold shutdown frequency or refueling frequency as indicated.

Page 12 of 81

n DONALD C. COO~CLEAR PLANT NOMENCLATURE FOR TEST MEWED USED IN COLUMNS FOR PRIMARY TEST REQUIRED AND TEST PERFORMED UNDER ASME SECTION XI Revision No: 3 Figure 3 Date: 2-5-90 ASME CODE SECTION XI (cont'd) 1 CATEGORY A-B VALVES PARAGRAPH ET-XXX Exercise power operated valve (full stroke) to its safety position (IWV-3413 &3417) and measure time. The stroke time limiting values of these valves including rapid acting valves will be identified and controlled per plant Technical Data Book and plant procedures. Valves assigned 2 seconds limiting values are subject to relief specified in "Paragraph H, Figure l."

2 CATEGORY C VALVES CF-1 Exercise valve (full stroke) for operability quarterly. (IWV-3521)

CF-2 Exercise valve (full stroke) for operability at a cold shutdown (IWV-3521) frequency or refueling outage frequency as indicated. Code relief requests and/or cold shutdown justification are provided in the corresponding valve notes.

CF-3 Exercise valve (part stroke) for operability quarterly; exercise (IWV-3522)

(full stroke) for operability at a cold shutdown frequency or refueling frequency as indicated. Justification for exercising valves at a cold shutdown frequency is provided in the corresponding valve notes. Code relief requests are provided full stroke testing is deferred to coincide with refueling if frequency.

CF-4 Exercise valve (full stroke) for operability prior to return to (IWV-3416) service.

TF-1 Safety and relief valve tests (setpoint) to Section XI, Table (IWV-3510)

IWV-3510-1.

Page 13 of Sl

DONALD C. COOK NUCLEAR P NOMENCLATURE FOR TEST MET USED IN COLUMNS FOR PRIMARY TEST REQUIRED AND TEST PERFORMED UNDER ASME SECTION XI Revision No: 3 Figure 3 Date: 2-5-90 3 CATEGORY A or AC VALVES SLT-1 Seat leakage test valve in accordance with requirements of paragraph IWV-3420 of ASME Code,Section XI, at refueling outage frequency but not less than once every two years. Permissible leakage values for each category A or AC valve are listed in Attachment "A".

SLT-2 Seat leakage test valve in accordance with 10CFR 50, Appendix J, in lieu of ASME Code Section XI except for paragraphs IWV-3426 and IWV-3427 which are applicable. This is consistent with the NRC position described in GL-89-04, Attachment A, Item 410.

Permissible leakage values for each category A or AC valve are listed in Attachment-"A".

SLT-2A In lieu of the requirements of ASME Code Section XI, paragraphs IWV-3423 and IWV-3424, valves are seat leakage tested as part of the Appendix "J" containment isolation test by imposing a static head of water on the downstream side of the valve and verifying that the leakage within the specified value of Attachment "A" for each category valve. This testing method demonstrates that the containment spray and RHR Check Valve leakage over 30 days is limited to the water resident in the containment spray headers downstream of the check valves. The leakage specified would not deplete the water inventory so as to expose these valves to a post-LOCA environment for a minimum of 30 days in the event that a spray system must be shut down and drained. This testing method is as stated in Response to Question 22.15(5) of the original FSAR Appendix "Q", Amendment 81, dated August 1978.

Page 14 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram" No: 1-5105D-1 Revision No: 3 Date: 2-5-90 NOTE 1: FW-118-1 thru -4 Code Relief : These check valves are normally open during power operation to pass main feedwater flow to the steam generators. Their safety function (close) prevents auxiliary feedwater backflow into the main feedwater system. These valves cannot be exercised during power operation because closing of these valves would require securing feedwater flow to the steam generators. Main feedwater to the steam generators cannot be isolated on a loop basis because three loop operation is not allowed per Donald C. Cook Nuclear Plant Technical Specification 3.4.1.1. For these category "C" check valves, backflow cannot be quantified at cold shutdown due to system configuration. The only practical method to verify valve closure is by disassembly. Due to size, weight and close proximity to physical barriers (whip restraints); valve disassembly at cold shutdown would impose constraints on the manpower and scheduling that may delay essential cold shutdown related activities and the plant start-up. The valves are not equipped with position indicators. There has been no operational or maintenance adverse trend noted. Therefore, the valves will be disassembled (bonnet removed) and verified closed (disc against seat) on a sampling basis (one of four) at refueling outage frequency.

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DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5105D-1 Revision No: 3 Date: 2-5-90 NOTE 2: MRV-210 -220 -230 and -240 Cold Shutdown Justification : These steam generator stop valves cannot be full stroke exercised during power operation because this would require securing steam from a steam generator which could result in a reactor trip. Three loop operation is not allowed for D. C. Cook per Technical Specification 3.4.1.1. Valves MRV-211, -221, -231, -241, -212, -222, -232, and

-242 which activate MRV-210, -220, -230, and -240 are tested quarterly in accordance with IWV-3410; MRV-210, -220, -230 and

-240 are part stroke tested quarterly by use of hydraulics attached to valve operators. and full stroke tested during hot standby (Mode 3 with RCS temperature > 541 F) at cold shutdown frequency.

NOTE 3: MS-108-2 and 108-3 Code Relief : These check valves are located in the steam supply lines to the Auxiliary Feedwater Pump Turbine.

These valves are part stroke tested during normal IST feedwater pump testing at least on a quarterly basis .at approximately 700 gpm because flow is restricted to a maximum of approximately 700 gpm through the 3" test line used during pump test. The valves will be full stroke tested to open position at a cold shutdown frequency.

The valve is not equipped with position indicator. In addition, due to the plant design, the only method available to verify the valve closure is disassembly. The valve will be disassembled (bonnet removed) and verified closed (disc against seat) and visually examined in accordance with GL-89-04, Attachment A, Item g2 on a sampling basis (one of two) once every other refueling frequency.

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DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5106-35 Revision No: 3 Date: 2-5-90 NOTE 1: FMO-201 -202 -203 -204 & FRV-210 -220 -230 -240 Cold Shutdown Justification : The function of these valves is to provide feedwater flow from the feedwater pumps to the steam generators. These valves cannot- be exercised (part or full stroke) during power operation because closing these valves would require securing feed flow to the steam generator which may cause instability of steam generator water level which could result in reactor trip. Further, three loop operation is not allowed per Donald C. Cook Nuclear Plant Technical Specification 3.4.1.1.

These valves will be full stroke exercised and timed during unit start-up or shutdown at cold shutdown frequency.

Page 17 of 81

It DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5106A-38 Revision No: 3 Date: 2-5-90 NOTE 1: FW-132-1 3 -4 Cold Shutdown Justification : These auxiliary feedwater (AFW) check valves function to supply AFW to the steam generators whenever the AFW System is caused to operate.

These check valves cannot be full or partial stroke exercised during power operation without energizing the AFW System and delivering cold water to the steam generators. This would result in thermal shock to the steam generator nozzles. These valves are full stroke exercised during startup. The valves will be verified closed quarterly by monitoring temperature of Auxiliary Feed Line as required by the plant procedure during shift inspection tours.

NOTE 2: FW-134 & FW-135 Cold Shutdown Justification : These valves are located on the suction and discharge lines of the turbine driven auxiliary feedpump. The maximum flow rate through the turbine driven auxiliary feedpump during IST is approximately 700 gpm using the pump test line. Passing the design flow of 900 gpm through these valves would require delivering cold auxiliary feedwater to the steam generators. This would result in thermal shock to the steam generator -nozzles. Therefore, these valves will be part stroke exercised quarterly and full stroke exercised (passing design flow of 900 gpm through the valves) at cold shutdown frequency.

Page 18 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5106A-38 Revision No: 3 Date: 2-5-90 NOTE 3: FW-138-1 -2 -3 -4 Cold Shutdown Justification : These auxiliary feedwater (AFW) check valves function to supply AFW to the steam generators whenever the AFW System is caused to operate.

These check valves cannot be full or partial stroke exercised during power operation without energizing the AFW System and delivering cold water to the steam generators. This would result in thermal shock to the steam generator nozzles. The valves will be verified closed quarterly by monitoring temperature of Auxiliary Feed Line as required by the plant procedure during shift inspection tours. These valves are full stroke exercised when the plant is returned to power after cold shutdown.

NOTE 4: FW-149 and 150 Comment : The required full stroking of these check valves is satisfied when Turbine Driven Auxiliary Feedpump

,completes its required testing.

NOTE 5 FW-153 and 160 Comment : These check valves installed on the Emergency Leak Off (ELO) lines open when the Motor Driven Auxiliary Feedwater Pumps (MDAFp) start. This can be established when the MDAFP pump is operating through the test line. A pressure decrease in the pump discharge line is verified by a local pressure indicator when the parallel path ELO is opened. The pressure decrease indicates that flow is established through the ELO line and that the check valve is opened.

Page 19 of 81

DONALD C. COOK NUCLEAR P T l

VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5113-41 Revision No: 3 Date: 2-5-90 NOTE 1: ESW-ill -112 -113 -114 Comment : These valves are full stroke exercised quarterly as required by IWV-3520. In addition, they are disassembled and inspected internally in accordance with IEB 85-03 at refueling outage frequency.

NOTE 2: ESW-109 -115 -243 Cold Shutdown Justification : These valves are normally closed and are required to be open when the condensate storage tank is exhausted. Exercising the valves could cause lake water contamination of the steam generators. Lake water chemistry can potentially impact steam generator tube integrity. Therefore, the valves will be full stroke tested at a cold shutdown frequency.

Since the valves are manual, stroke timing is not required.

Page 20 of 81

DONALD C. COOK NUCLEAR P ~

VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5113-41 Revision No: 3 Date: 2-5-90 NOTE 3: HRV-721 -723 -725 -727 Code Relief : These valves are located in the essential service water supply lines to the emergency diesel generators air after coolers. These three-way valves regulate water flow to maintain the temperature at which the after cooler air discharge thermostatic controller has been set. Water flow is regulated by passing a portion of the flow through the air coolers and bypassing the excess flow around the air after coolers. Code relief is being requested from the testing requirements since (1),

these valves function only as regulating valves and not open/closed valves (2), these valves are demonstrated operable during diesel generator testing (diesel generators are tested per Technical Specification 4.8.1.1.2); and (3), these valves are demonstrated operable during diesel generator 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months runs performed each refueling outage. The valves -will be "fail-safe" tested using their control scheme that will remove air from the valve operators causing them to direct all ESN flow to the.air after coolers. This proposed test for each valve will be performed at refueling frequency. The valves cannot be stroke timed because they are thermostatic valves whose position is controlled by process fluid temperature. There is no external control available.

Page 21 of 81

DONALD C. COOK NUCLEAR. P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5114A-31 Revision No: 3 Date: 2-5-90 NOTE 1: WCR-900 throu h .-915 -920 throu h -935 -941 throu h 948 -951 throu h -958 and 960 throu h -967 Code Relief See "Attachment-A" for permissible seat leakage values.

Page 22 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5124-21 Revision No: 3 Date: 2-5-90 NOTE 1 DCR-600 & -601 and NS-357 Code Relief : See "Attachment-A" for permissible seat leakage values.

NOTE-2: NS-357 Code Relief : This check valve is located on the return line of the post accident sampling system inside the containment.

Since the line is open-ended inside the containment and the check valve is not equipped with the position indication, the valve will be full stroke exercised in the open position by performing a flow test quarterly and will-be confirmed closed in conjunction with Appendix J seat leakage testing at refueling frequency.

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'I I

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5128-19 Revision No: 3 Date: 2-5-90 NOTE 1 NSO-021 -022 -023 & -024 Cold Shutdown Justification : These four one-inch solenoid operated isolation valves are installed (two in each leg in series) in the reactor head vent. These valves cannot be tested during power operation, hot standby, or hot shutdown because the valve design is such that testing of either valve can cause "burping" (momentary opening) of the second valve, resulting in the release of radioactive fluid and create an airborne situation in containment. Therefore, the valves will be full stroke exercised and timed at cold shutdown frequency.

Exercising the solenoid operated valves for verification of valve position (valve stem movement) will be performed at refueling frequency by a flow test through each valve because the valve stem is completely enclosed and cannot be observed. The reactor coolant discharged during the flow testing of the valves is collected in a container to minimize liquid contamination spill, radiation, and potential airborne situation in deference of ALARA consideration and personnel protection. The above tests are consistent with Technical Specification requirements.

Page 24 of 81

0 P

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5128A-37 Revision No: 3 Date: 2-5-90 NOTE 1 CS-442-1 thru 4 Code Relief : These containment isolation check valves are on the seal water supply line to the RC pumps. These valves cannot be part or full stroke exercised to the closed position during power operation because cooling flow is required to the RCP seals. During cold shutdown, seal water must be maintained to prevent backflow through the seals with possible damage from dirt. The valves will be full stroke exercised in conjunction with Appendix J seat leakage testing at refueling frequency.

NOTE 2: GCR-301 NCR-252 NPX-151 RCR-100 & -101 CS-442-1 throu h 4 SI-189 PW-275 and N-159 Code Relief : See Attachment-"A" for permissible seat leakage values.

NOTE 3: NRV-151 -152 -153 Cold Shutdown Justification : These pressurizer power operated relief valves are normally closed during power operation. The valves cannot be exercised at power without inducing an RCS pressure transient which could result in reactor trip. The valves will be full stroke exercised and timed at cold shutdown frequency.

Page 25 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5128A-37 Revision No: 3 Date'. 2-5-90 NOTE 4: Code Relief : This containment .isolation check valve is located in the primary water supply line to the pressurizer relief tank. The valve is not equipped with position indication. The valve 5'W-275 cannot be full stroke tested to closed'osition during power operation or at a cold shutdown frequency due to lack'of sufficient differential pressure to back seat the valve. The valve and necessary test connections are located inside the containment. Due to the plant design, the only method available to verify the valve closure is leak testing. The valve will be verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

NOTE N-159 Code Relief : This containment isolation check valve is located in the. nitrogen supply line to the pressurizer relief tank.

The valve i:s not equipped with position indication. The valve cannot be full stroke tested to closed position during power operation or at a cold shutdown frequency due to lack of sufficient differential pressure to back seat the valve. The valve and necessary test connections are located inside the containment. Due to the plant design, the only method available to verify the valve closure is leak testing. The valve will be verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

Page 26 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5128A-37 Revision No: 3 Date: 2-5-90 NOTE 6: NSO-061 -062 -063 -064 Cold Shutdown Justification : These four one-inch solenoid operated isolation valves are installed (two in each leg in series) in the pressurizer vent. These valves cannot be tested during power operation, hot standby, or hot shutdown because the valve design is such that testing of either valve can cause "burping" (momentary opening) of the second valve resulting in the release of radioactive fluid and create an airborne situation in containment. The valves will be full stroke tested and timed at cold shutdown frequency.

Exercising the solenoid operated valves for verification of valve position (valve stem movement) will be performed at refueling frequency by performing a flow test through each valve because the valve stem is completely enclosed and cannot be observed. The reactor coolant discharged during flow testing of the valves is collected in a container to minimize contaminated liquid spill, radiation, and potential airborne situation in deference of ALARA consideration and personnel protection. The above tests are consistent with Technical Specification requirements.

Page 27 of 81

/

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5128A-37 Revision No: 3 Date: 2-5-90 NOTE 7: SI-189 Code Relief : This check valve is located in the safety valves discharge (Emergency Core Cooling SVs, RHR, SVs, centrifugal charging pump SVs, etc.) collection header leading to the pressurizer relief tank. Isolating this valve for testing would result in dead heading all safety valves in the above systems.

This would result in loss of overpressurization protection and could put the plant in an unsafe condition. Therefore, the valve will be part stroke exercised to open position using external source via test connection at a cold shutdown frequency. The valve will be disassembled, manually full stroke tested and visually examined in accordance with GL-89-04, Attachment 1, Item 52 at every third refueling outage frequency. The valve will also be verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

Page 28 of 81

S

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DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5129-31 Revision No: 3 Date: 2-5-90 NOTE 1: CS-292 Code Relief : This valve is in the emergency boration path from the boric acid system to the charging pump suction header.

Flow through this path is normally not provided at power because of the resultant large negative reactivity insertion. The valve will be full stroke exercised in the open position at a cold shutdown frequency. The check valve is not equipped with position indication. Due to the plant design, the only methods available to verify the valve closure is either radiography or disassembly which will be performed at a refueling frequency when the system is not required to be operable. The radiography method is an acceptable method to verify the valve closure (disc against the seat) under no flow condition because it provides visual observation of the valve in the closed position. The flow testing of the valve verifies that it is open. This provides assurance that the disc is free to move from the open position with flow to the closed position with no flow or reverse flow.

NOTE 2 CS-299E -299M Code Relief : These check valves located on the discharge lines of the 'E'nd 'H'harging pumps function as pressure isolation valves to protect the low pressure charging pump suction lines. These valves cannot be full-stroke exercised during: (1) power operation because the charging pumps cannot achieve maximum flow rate with the reactor at full pressure, and (2) cold shutdown because the flow required could cause a low temperature overpressure condition. The valves will be part-stroke exercised quarterly and full stroke exercised at refueling frequency. The valves will also be verified closed in conjunction with seat leakage testing per IWV-3420 at refueling frequency.

Page 29 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5129-31 Revision No: 3 Date: 2-5-90 NOTE 3 CS-321 QCR-300 and -301 Code Relief : See Attachment "A" for permissible seat leakage values.

NOTE 4: CS-321 Code Relief : This containment isolation check valve's function is to supply borated water from the volume control tank to the regenerative heat exchanger through the charging pumps for chemical shim control and reactor coolant system makeup. Isolation of this system would result in loss of control of pressurizer level which could result in reactor trip. This valve is tested in the open direction quarterly and confirmed closed in conjunction with Appendix J seat leakage testing at refueling frequency.

NOTE 5: CS-328L1 -329L1 -328L4 -329L4 Comment : These check valves function to provide the interface point between the RCS and the CVCS. Since the discharge piping of the CVCS is designed to a pressure rating higher than the RCS, these valves do-not perform a pressure isolation function. The higher pressure (RCS) to low pressure (CVCS Suction) isolation is accomplished by other valves which are tested to category "A" requirements. The valves will be full stroke exercised to open position quarterly.

NOTE 6: QCR-300 -301 Cold Shutdown Justification : These air operated containment isolation valves are located on the letdown return line. Exercising these valves during power operation would result in letdown isolation which could result in loss of pressurizer level control which could result in a plant shutdown. The valves will be full stroke exercised, timed and fail safe tested at a cold shutdown frequency and seat leakage tested per Appendix J. program at refueling frequency.

Page 30 of 81

DONALD C. COOK NUCLEAR PLA

~i VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5129-31 Revision No: 3 Date: 2-5-90 NOTE 7 QMO-200 -201 Cold Shutdown Justification : These motor operated gate valves .are installed on the CVCS charging line which provide borated water for RCS chemical shim control and reactor coolant system makeup. Isolation of this system would result in loss of control of pressurizer level which could result in reactor trip.

The valves will be full stroke tested and timed at cold shutdown frequency.

NOTE 8: QRV-200 Code Relief : This air operated valve is used to regulate charging header flow to the reactor coolant system and seal water flow to the reactor coolant pump seals. The valve cannot be full stroke exercised at power operation because it would interrupt the seal injection flow to the reactor coolant pumps which could result in reactor coolant pump seal damage. The valve will be part stroke exercised during power operation and full stroke exercised at a cold shutdown frequency. The valve cannot be stroke timed because there is no local or remote position indicator available and cycle times are directly proportional to "how fast" the operator turns the control knob. Therefore, meaningful stroke times are not achievable. This valve has no fail safe position. The alternative testing proposed is to locally observe the valve during full stroke testing for smooth operation and apparent problems which can affect the valve operation.

Page 31 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGHAM RELIEF REQUEST NOTES Flow Diagram No: 1-5129-31 Revision No: 3 Date: 2-5-90 NOTE 9: QRV-251 Code Relief : This air operated valve is used to regulate charging header flow to the reactor coolant system and seal water flow to the reactor coolant pump seals. The valve cannot be full stroke exercised at power operation because it would interrupt the RCP seal injection flow and would also, upset pressurizer level.

The valve will be part stroke exercised during power operation and full stroke exercised at a cold shutdown frequency. The valve cannot be stroke timed because there is no local or remote position indicator available and cycle times are directly proportional to "how fast" the operator can turn the control knob. Therefore, meaningful stroke times are not achievable. The control scheme of this valve functions to remove air from the valve oeprator, which duplicates the fail-safe condition, resulting in the valve going to fail-safe (open) position. Therefore, the alternative testing proposed will consist of locally observing the valve during full stroke testing for smooth operation and apparent problems which can affect the valve operation.

NOTE 10: SI-185 Code Relief : This normally closed valve functions to transfer the suction source of the charging pumps to the refueling water storage tank. .This valve cannot be full stroke- exercised during: (1) power operation without introducing a high concentration of boric acid in the RCS, and.(2) cold shutdown because the only full flow path available is into the reactor coolant system and the system does not have sufficient volume to accommodate that flow without a possible low temperature overpressure condition. The valve will be full stroke exercised at refueling frequency.

Page 32 of 81

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DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5129-31 Revision No: 3 Date: 2-5-90 NOTE ll: CS-299E&W Comment leakage values.

See Attachment "A" for permissible seat Page 33 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5129A-19 Revision No: 3 Date: 2-5-90 NOTE 1: QCM-250 -350 Cold Shutdown Justification : These motor-operated reactor coolant pump seal water return isolation valves cannot be exercised during power operation because it would interrupt reactor coolant pump seal water flow and could cause damage to the seals.

Therefore, the valves are full stroke exercised and timed at cold shutdown frequency.

NOTE 2: QCM-250 and -350 Code Relief : See "Attachment-A" for permissible seat leakage values.

NOTE 3 QMO-451 -452 Cold Shutdown Justification : These-motor-operated gate valves function as volume control tank isolation valves.

Exercising these valves during power operation could result in a loss of pressurizer level control which could cause a reactor trip.

These valves are full stroke exercised and timed at cold shutdown frequency.

Page 34 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5135-29 Revision No: 3 Date: 2-5-90 NOTE 1 CCM-451 -452 -453 -454 -458 and -459 Cold Shutdown

'"""L'peration without securing cooling water to the reactor Coolant Pumps (RCPs). Isolation of these valves could cause failure of the RCPs. The valves will be full stroke tested and timed at cold shutdown frequency.

NOTE 2 CCM-451 throu h -454 -458 -459 CCR-455 throu h -457 -460 -462 and CCH-135 Code Relief See "Attachment-A" for permissible seat leakage values.

NOTE 3: CCR-455 -456 and -457 Cold Shutdown Justification : These valves cannot be tested during power operation without securing cooling water to the reactor support coolers. These valves must remain open to prevent overheating of the concrete around the reactor supports during power operation. The valves will be full stroke tested and timed at cold shutdown frequency.

NOTE 4 CCH-135 Code Relief : This check valve cannot be tested during power operation without securing cooling water to the reactor support coolers. The valve must remain open to prevent overheating of the concrete around the reactor supports during power operation.

The valve will be verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

NOTE 5: CRV-470 Code Relief : This air operated valve is used to regulate component cooling water (CCW) to the letdown heat exchanger. The valve is normally in service during power operation.

Page 35 of 81

DONALD C. COOK NUCLEAR .P VALVE TEST PROGRAM RELIEF- REQUEST NOTES Flow Diagram No: 1-5135-29 Revision No: 3 Date: 2-5-90 CRV-470 (continued)

This valve is controlled by an auto/manual station with auto input from the letdown heat exchanger outlet temeprature sensor (QTC-302). The valve also trips closed from an SI signal via a solenoid valve. The valve will be full stroke exercised quarterly using auto/ma'nual station which will permit rapid cycling of this regulating valve resulting in minimal impact on letdown temperature. Meaningful stroke time data is not available since this valve does not have local or remote position indication.

Fail safe testing this valve closed requires a longer period of time than cycling the valve using the auto/manual station. The valve will be fail safe tested to its closed position at cold shutdown frequency with letdown flow out of service thus avoiding high letdown line temperatures that could cause flashing in the letdown heat exchanger and lifting of safety valves.

Page 36 of 81

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5135A-30 Revision No: 3 Date: 2-5-90 NOTE 1 CCW-170 Comment : This valve will be tested in accordance with IWV-3416 whenever the spare CCW pump is placed in service.

NOTE 2: CMO-411 -412 -413 -414 -415 & -416 Comment: These valves remain open during initial safety injection, but may be closed during recirculation phase or passive failure. Therefore, the valve time will be recorded from open to close position.

Page 37 of 81

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DONALD 'C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5135B-14 Revision No: 3 Date: 2-5-90 NOTE 1: CCM-430 throu h -433 CCR-440 and -441; CCW-243-25 -243-72

-244-25 and -244-72 Code Relief : See "Attachment-A" for permissible seat leakage values.

NOTE 2 CCW-243-25 CCW-243-72 CCW-244-25 and CCW-244-72 Code Relief These check valves are located in the penetration cooling supply headers of the CCW System inside the containment. -The valves are open during power operation and cold shutdown -to provide cooling water to the main steam penetrations. These valves are not equipped with position indication. The valves will be confirmed closed in conjunction with Appendix Z seat leakage testing at refueling frequency.

page 38 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5141-29 Revision No: 3 Date: 2-5-90 NOTE 1: ICR-5 -6 NCR-105 throu h -110 Code Relief : See >>Attachment-A" for permissible seat leakage values.

page 39 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5141D-10 Revision No: 3 Date: 2-5-90 NOTE 1 ECR-10 throu h -29 and NS-283 Code Relief : See "Attachment-A" for permissible seat leakage values.

NOTE 2: NS-283 Code Relief : This containment isolation check valve is located in the sample return line of the Post-Accident Containment Hydrogen Monitoring System. The. valve cannot be full stroke exercised to closed position quarterly or at a cold shutdown frequency because the line is open ended in the containment. This check valve is not equipped with position indicator. The only method available to verify the valve closure is by seat leakage testing. The valve will be full stroke exercised to the open position by a flow test quarterly and will be confirmed closed in conjunction with Appendix J seat leakage testing at refueling frequency.

Page 40 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5142-25 Revision No: 3 Date: 2-5-90 NOTE 1 ICM-250 and ICM-251 Cold Shutdown Justification : These normally closed valves cannot be operated during normal plant operation without introducing Boron into a nonheat traced line. Boron could crystallize and plug the line. The valves will be full stroke tested and timed at cold shutdown frequency.

NOTE 2: ICM-250 -251 -260 and -265 Code Relief : See "Attachment-A>> for permissible seat leakage values.

NOTE 3: IMO-261 Cold Shutdown Justification : This valve cannot be tested when SI pumps are required to be operable. Testing would result in isolation of the common suction line to both SI trains. This valve will be stroke tested and timed at cold shutdown frequency.

NOTE 4: IMO-262 and -263 Cold Shutdown Justification : These motor operated valves are located in series in the re-circulation line of the Safety Injection pumps. Exercising either of these valves will make both SI pumps inoperable. These valves will be full-stroke exercised and timed at cold shutdown frequency when SI pumps are not required to be operable.

NOTE 5: SI-llON SI-llOS and SI-101 Code Relief : Safety Injection (SI) pump discharge valves, SI-110N and -110S, cannot be exercised during power operation because the SI pumps cannot overcome reactor coolant system pressure. Therefore, no flow path exists and, because minimum flow lines branch off upstream of these valves, they cannot be part-stroke tested during pump testing. The common (SI pumps) suction check valve, SI-101 is gart-stroke exercised at Page 41 of 8l

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5142-25 Revision No: 3 Date: 2-5-90 Note 5 (continued) power operation during pump testing. These valves cannot be exercised during cold shutdown because the SI pumps are required to be inoperable by Technical Specification 3.5.3 to protect against low temperature overpressurization of the reactor. These valves will be full-stroke exercised at refueling frequency.

NOTE 6: SI-142 Ll L2 L3 and L4 Code Relief : These check valves are located in the supply lines. from the Boron Injection Tank to the reactor coolant cold legs (loop 1 through 4). These valves cannot be tested during power operation because this would require injecting highly concentrated boric acid solution from the Boron Injection Tank into the Reactor Coolant System resulting in probable plant shutdown.

These valves cannot be partially-stroke exercised using the BIT bypass line because this could result in bypassing the BIT, thereby not achieving design flow through the BIT if an accident occurred.

These valves cannot be full-stroked exercised during cold shutdown because this would require injecting the BIT into the RCS which could significantly delay startup from cold shutdown condition (the BIT would have to be brought to the proper Boron concentration and the RCS would have to be diluted sufficiently to allow startup).

These valves will be full stroke exercised at refueling frequency.

Page 42 of 81

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DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF,. REQUEST NOTES Flow Diagram No: 1-5143-36 Revision No: 3 Date: 2-5-90 NOTE 1: SI-151E&W -152N&S -158-Ll throu h L4 --161-Ll throu h L4 -166-1 throu h 4 -170Ll throu h L4 RH-133 -134 and ICM-129 Comment See "Attachment-A" for permissible seat leakage values.

NOTE 2: IMO-128 and ICM-129 Cold Shutdown Justification : These valves

. function as the normal return from the RCS to the .RHR for heatup and cooldown. These valves are normally closed and cannot be operated during normal plant operation because they are interlocked to remain closed at RCS pressure above 450 psig. The valves will be full stroke exercised and timed prior to placing them into service at cold shutdown frequency.

NOTE 3: IMO-310 -320 -314 -324 Comment : These valves remain open during injection phase of a safety injection, but will be closed during. recirculation phase. Therefore, stroke timing will be from open to close position.

NOTE 4: IMO-315 -316 -325 -326 Cold Shutdown Justification : Valves IMO-315 and -325 are normally closed valves, located in the RHR and SI Supply Header to RCS hot legs. Valves IMO-316 and -326 are normally open valves located in the RHR and SI Supply Header to RCS cold legs. These valves should not be exercised during power operation because failure in a non-conservative position would result in less than minimum number of injection flow path as required by the FSAR. The valves will be full stroke tested and timed at cold shutdown frequency.

Page 43 of 81

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5143-36 Revision No: 3 Date: 2-5-90 NOTE 5 SI-166-1 throu h 4 Code Relief : These check valves function to prevent backflow from the RCS into the accumulators during normal operation. These valves function to supply flow from the accumulators to the RCS during an accident condition. These valves cannot be exercised during power operation because the accumulators do not have sufficient head to overcome RCS pressure.

These valves cannot be exercised during cold shutdown because this would result in a possible low temperature overpressurization of the.RCS. Full stroke testing during refueling outages is not possible because of the resulting water surge into the reactor and the potential for. high airborne. radiation contamination. These valves will be part stroke exercised at refueling frequency. The

= valves will be disassembled, manually. full stroke exercised and visually examined on a sampling basis (one of four) per GL-89-04, Attachment 1, Item g2, at refueling frequency.

NOTE 6: SI-161 Ll L2 L3 L4 Code Relief : These check valves are located in the supply lines from the Residual Heat Removal and Safety Injection Pumps to the RCS cold legs (loop 1 through 4).

These valves cannot be exercised during power operation because the RHR pumps and SI pumps do not develop sufficient head to overcome RCS pressure. Full stroke of these valves (individually) cannot be verified at cold shutdown frequency because flow instrumentation is not available downstream of the flow split. These valves will be part stroke exercised at cold shutdown frequency and full stroke will be locally verified using portable instrumentation at refueling frequency.

Page 44 of 81

K DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5143-36 Revision No: 3 Date: 2-5-90 NOTE 7: RH-108E Cold Shutdown Justification : These valves cannot be W

full stroke exercised quarterly because modification to existing it would require design instrumentation to accommodate the full flow test measurements. The valves will be part stroke exercised quarterly and full stroke exercised at cold shutdown frequency (during RHR operation).

NOTE 8: SI-148 Code Relief : Check valve SI-148 is located in the Refueling Water Storage Tank (RWST) supply line to the RHR system.

.The design flow through the valve is 6000 gpm. Flow to the core is not possible when the RCS pressure is above the shut-off pressure of the RHR pumps (195 psig). In order to full stroke exercise this valve, both RHR pumps must be operated and the RHR system manually aligned to recirculate flow back to the RWST. This configuration places both RHR trains inoperable since neither train can provide design flow to the core. In order to preclude placing the unit in an unsafe condition, a partial stroke test is performed quarterly.

The valve cannot be full stroke exercised during cold shutdown since the RCS cannot accommodate the introduction of 6000 gpm from the RHR system. In addition, during cold shutdown, the RHR system is required to be operable for RCS temperature control. The valve will be full stroke exercised when the reactor cavity is being flooded at refueling frequency.

NOTE 9 SI-151 E W Cold Shutdown Justification : These check valves are located in the RHR supply lines to either the hot or cold legs.

These valves cannot be exercised during power operation because the RHR pumps do not develop sufficient head to overcome RCS pressure.

These valves will be exercised at cold shutdown frequency.

page 45 of 81

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5143-36 Revision No: 3 Date: 2-5-90 NOTE 10: SI-152 N S Code Relief : These check valves function to provide Safety Injection pump discharge to either the hot or cold legs.

These valves cannot be exercised during power operation because the SI pumps do not develop sufficient pressure to overcome RCS pressure. These valves cannot be exercised during cold shutdown because the safety injection pumps are required to be inoperable by Technical Specification Section 3.5.3, to protect against low temperature overpressurization of the reactor vessel. Also, during cold shutdown, there may not be sufficient volume in the RCS to accommodate the amount of water needed to full stroke. These valves will be full stroke exercised at refueling frecpxency.

NOTE ll: SI-158 Ll L2 L3 L4 Code Relief : Check valves SI-158 are located in the supply lines from the Residual Heat Removal and Safety Injection Pumps to the RCS hot legs (loop 1 through 4).

These valves cannot be exercised during power operation because the RHR and SI pumps do not develop sufficient head to overcome RCS pressure. Full stroke of these valves (individually) cannot be verified at cold shutdown frequency because flow instrumentation is not available downstream of the flow split. These valves will be part stroke exercised at cold shutdown frequency. Full stroke will be verified using portable instrumentation at refueling frequency.

Page 46 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5143-36 Revision No: 3 Date: 2-5-90 NOTE 12: SI-170 Ll L2 L3 and L4 Code Relief : These valves are located on the RCS cold leg (loops 1 through 4) injection lines from the accumulators, RHR, and SI systems. They cannot be exercised during-power operations because the RHR and SI pumps do not develop sufficient head to overcome RCS pressure. The valves will be part-stroke exercised at a cold shutdown frequency. Due to the plant design, the valves are sized such that full stroke testing cannot be attained without discharging the accumulators and operating SI and RHR pumps simultaneously. The only method available to verify the full stroke is by disassembly. The valves are not equipped with position indicators. The valves will be disassembled, manually full stroke exercised and visually examined on a sampling basis (one of four) per GL-89-04, Attachment 1, Item N2; at refueling frequency.

NOTE 13: N-102 Code Relief : This check valve is located in the nitrogen supply header to the accumulators for blanketing purposes. The valve cannot be full stroke tested to the closed position during power operation or cold shutdown because, due to the plant design, the only method available to verify the valve closure is leak testing. The valve and necessary test connections are located inside the containment. The valve is not equipped with a position indicator..The valve will be verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

NOTE 14: GCR-314 ICM-305 and -306 N-102 SI-171 -172 and -194 Code R~elief: See "Attachment-A" for permissible seat leakage values.

Page 47 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5143-36 Revision No: 3 Date: 2-5-90 NOTE 15: IMO-340 and -350 Code Relief : Valves IMO-340 and IMO-350 are located in east and west RHR discharge headers to the suction of charging and SI pumps, respectively. These valves are normally closed during power operation, and would be opened during the recirculation phase of a LOCA.to allow the RHR pumps to provide water from the containment recirculation sump.to charging and SI pumps. These valves cannot be full stroke exercised during power operation because they are interlocked with valves IMO-262 and

-263, located in series, in the SI pump miniflow (recirculation) line to RNST. Closing of IMO-262 and -263 would render both SI pumps inoperable and, thus, places the unit in T/S 3.0.3, which allows one hour to restore the SI pumps to operable status or begin a unit shutdown. The complicated valve and equipment lineup to perform the valve testing in one hour is highly unlikely. Therefore, the valves will be full stroke exercised and timed on a cold shutdown frequency. (For additional details, refer to Code Relief granted by the NRC dated 1-30-89, AEP:NRC:09690.)-

Page 48 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5144-28 Revision No: 3 Date: 2-5-90 NOTE 1: CTS-138E & W Code Relief : These check valves are located in the lines which supply water from'he RWST to the containment spray pumps. The valves cannot be full stroke exercised during power operation, cold shutdown or refueling without, spraying the containment. The valves are part stroke exercised during containment spray pump testing on a quarterly basis. The only practical method available to verify full stroke of these valves is by disassembly. These valves are not equipped with position indicators. The valves will be disassembled, manually full stroke exercised and visually examined on a sampling basis (one of two) per GL-89-04, Attachment 1, Item g2, once every other refueling frequency.

NOTE 2: -CTS-103 E & W Code Relief : These check valves are located in the discharge lines of containment spray pumps to the spray ring headers in the containment. These valves cannot be full stroke exercised during power operation, cold shutdown or refueling without spraying the containment. The valves are part stroke exercised during containment spray pump testing on a quarterly basis. The only practical method available to verify full stroke of these valves is by disassembly. The valves are not equipped with position indicators. The valves will be disassembled, manually full stroke exercised and visually examined on a sampling basis (one of two) per GL-89-04, Attachment 1, Item g2, once every other refueling frequency.

Page 49 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5144-28 Revision No: 3 Date: 2-5-90 NOTE 3 CTS-131E & W Code Relief : These check valves are located in the supply lines to the (upper compartment) containment spray ring headers. These valves are in closed position during normal plant operation. The valves are exposed to containment atmosphere on the downstream side and are isolated from fluid pressure in the upstream side by the closed motor operated valves. The valves cannot be part or full stroke exercised during power operation, cold shutdown or refueling because flow through these valves would result in spraying the containment. This could cause problems with wet lagging, corrosion of components inside the containment, etc.

The only practical method available to exercise these valves is by disassembly. The valves are not equipped with position indicators.

The valves will be disassembled, manually full stroke exercised and visually examined on a sampling basis (one of two) per GL-89-04, Attachment 1, Item g2, once every other refueling frequency.

NOTE 4- CTS-127E & W Code Relief : These check valves are located in the supply lines to the (lower compartment) containment spray ring headers. These valves are in closed position during normal plant operation. The valves are exposed to containment atmosphere on the downstream side and are isolated from fluid pressure in the upstream side by the closed motor operated valves. The valves cannot be part or full stroke exercised during power operation, cold shutdown or refueling because flow through these valves would result in spraying the containment. This could cause problems with wet lagging, corrosion of components inside the containment, etc.

The only practical method available to exercise these valves is by

.Page 50 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5144-28 Revision No: 3 Date: 2-5-90 Note 4 (continued) disassembly. The valves are not equipped with position indicators.

The valves will be disassembled, manually full stroke exercised and visually examined on a sampling basis (one of two) per GL-89-04, Attachment 1, Item $ 2, once every other refueling frequency.

NOTE 5: RH-141 & -142 Code Relief : These check valves are located in the supply lines to the (upper compartment) containment spray ring headers from the RHR Heat Exchangers. These valves are in closed position during normal plant operation. The valves are exposed to containment atmosphere on the downstream side and are isolated from fluid pressure in the upstream side by the closed motor operated valves. The valves cannot be part or full stroke exercised during power operation, cold shutdown or refueling because flow through these valves would result in spraying the containment. This could cause problems with wet lagging, corrosion of components inside the containment, etc. The only practical method available to exercise these valves is by disassembly. The valves are not equipped with position indicators. The valves will be disassembled, manually full stroke exercised and visually examined on a sampling basis (one of two) per GL-89-04, Attachment 1, Item $ 2, once every other refueling frequency.

NOTE 6: CTS-109 and -110 Cold Shutdown Justification : These check valves function as vacuum breakers for spray additive tank. The check valves are closed during normal plant operation to maintain the tank pressurized. The valves will be verified closed quarterly during power operation and will be verified open at cold shutdown frequency.

page 51 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5144-28 Revision No: 3 Date: 2-5-90 NOTE 7'TS-127E&W CTS-131E&W RH-141 -142 Code Relief : These valves are to be seat leakage tested in accordance with the unique testing methods established in the FSAR because of the configuration at D.C. Cook Plant. The permissible seat leakage values of these valves are listed in Attachment "A".

Page 52 of 81

II DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5145-17 Revision No: 3 Date: 2-5-90 NOTE 1: CA-181-N&S Code Relief : See "Attachment-A" for permissible seat leakage values.

Page 53 of 81

DONALD C. COOK NUCLEAR PLAN'ALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5146B-24 Revision No: 3 Date: 2-5-90 NOTE 1: R-156 and R-157 Code Relief : These check valves are installed in parallel lines to the glycol main supply and return lines to relieve glycol thermal expansion. These valves and necessary test connections are located inside the containment. Due to the plant design, the only method available to verify valve closure is leak testing. The valves are not equipped with position indicators.

The valves will be full stroke exercised in the open direction quarterly and verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

NOTE 2: R-156 and -157 VCR-10 "Attachment-A"

-ll -20 and -21 for permissible seat Code Relief See leakage values.

Page 54 of 81

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5147A-34 Revision No: 3 Date: 2-5-90 NOTE 1: SM-4 -6 -8 and -10 VCR-101 throu h -107 and VCR-201 throu h -207 Page 55 of 81

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5149-20 Revision No: 3 Date: 2-5-90 NOTE 1: VRV-325 -315 Code Relief : These thermostatically controlled valves are located at the outlet of the control room air conditioner water pump. These three-way valves function to modulate water flow through the air handler package based on cooling requirements. These valves are normally in an intermediate position based on control room cooling load. These valves can not be full stroke exercised because there is no provision to fully close the valves. The valves will be part stroke exercised from an intermediate position in conjunction with fail safe testing on quarterly basis. These valves are demonstrated operable during normal control room air conditioning operation. The valves cannot be stroked timed because they are not equipped with position indicator and stroke times are not repeatable.

Page 56 of Sl

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5151A-25 Revision No: 3 Date: 2-5-90 NOTE 1: DL-113A -115A -125A -131A and -157A Comment : The required full stroking of the check valves is satisfied when the diesel generator successfully completes its required testing per Technical Specification 4.8.1.1.2.

NOTE 2: QT-114-lAB Code Relief : This valve is located at the discharge of the engine driven lube oil pump (diesel-generator). This three-way thermostatic valve functions to maintain the correct lube oil temperature by maintaining the correct proportion of oil flowing through the lube oil cooler and bypassing the lube oil cooler to maintain a preset lube oil temperature. We are requesting exemption from testing requirements since (1) this valve

.functions only as a regulating valve and not opened/closed; (2) this valve is demonstrated operable during diesel generator testing. Diesel generators are tested basis per Technical Specification 4.8.1.1.2. The valves will be verified operable by observing proper temperatures during diesel testing.

Page 57 of 81

0 II C~.

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5151B-28 Revision No: 3 Date: 2-5-90 NOTE 1: DG-101A -103A -127A 129A -145A -151A and -153A Comment :

The required full stroking of the check valves is satisfied when the diesel generator successfully completes its required testing per Technical Specification 4.8.1.1.2.

NOTE 2: QT-132-lAB Code Relief : This valve is located at the discharge of the emergency diesel engine jacket water pump. This three-way thermostatic valve functions to maintain the correct proportion of

'ater flowing through the diesel engine water cooler and bypassing the diesel engine jacket water cooler to maintain a preset jacket water temperature. We are requesting exemption from the testing requirements since (1) this valve functions only as a regulating valve and not open/closed valve; (2) this valve is demonstrated operable during diesel generator testing. Diesel generators are tested per Technical Specification 4.8.1.1.2. The valve will be verified operable by observing proper temperatures dur'ing diesel testing.

NOTE 3: XRV-221 and 222 -Startin Air Code Relief : The starting air valves are installed on parallel air supply lines to the emergency diesel generator (EDG). The valves are not equipped with position indication devices to directly measure valve stroke times. The valves function to provide starting air which rolls the EDG. The valves are functionally redundant to each other. These valves fail "as is," and, therefore, they have no fail safe position.

Successful starting of the EDG in accordance with Technical Specification 4.8.1.1.2 (i.e., slow start at least quarterly and fast start once every 184 days within 10 seconds) will verify the valve performance. The valve stroke timing will be verified by page 58 of 81

0 DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5151B-28 Revision No: 3 Date: 2-5-90 Note 3 (continued) measuring diesel starting times during fast start testing of EDG.

The valves on a staggered basis will be valved out one at a time to verify the operability of the opposite valve during slow start of EDG at least quarterly. Position indication will be confirmed during the above testing when only one starting air train is used to start the diesel generators.

NOTE 4: XRV-220-Jet Assist Code Relief : This valve's function is to

-facilitate the EDG fast start by providing an air boost to the turbo charger to assist in starting the EDG in its Technical Specification 4.8.1.1.2 time limitation of 10 seconds. The valve is not equipped with position indication devices; therefore, meaningful stroke times are not achievable. The valves will be

. full stroke and fail safe tested by verifying EDG starting time once per 184 days in accordance with Technical Specification 4 '.1.1.2.

Page 59 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5151C-26 Revision No: 3 Date: 2-5-90 NOTE 1: DF-108C -109C -114C -115C DL-113C .-115C -125C -131C and

-157C Comment : The required full stroking of the check valves is satisfied when the diesel generator successfully completes its required testing per Technical Specification 4.8.1.1.2.

NOTE 2: QT-114-1CD Code Relief : This valve is located at the discharge of the engine driven lube oil pump (diesel-generator). This three-way thermostatic valve functions to maintain the correct lube oil temperature by maintaining the correct proportion of oil flowing through the lube oil cooler and bypassing the lube oil cooler to maintain a preset lube oil temperature. We are requesting exemption from testing requirements since (1) this valve functions only as a regulating valve and not opened/closed valve; (2) this valve is demonstrated operable during diesel generator testing. Diesel generators are tested per Technical Specification 4 '.1.1.2. The valves will be verified operable by observing proper temperatures during diesel testing.

Page 60 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5151D-28 Revision No: 3 Date: 2-5-90 NOTE 1: DG-101C -103 -127C -129C -145C -151C and -153C Comment The required full stroking of the check valves is satisfied when the diesel generator successfully completes its required testing per Technical Specification 4.8.1.1.2.-

NOTE 2 QT-132-1CD Code Relief : This valve is located at the discharge of the emergency diesel engine jacket water pump. This three-way thermostatic valve functions to maintain the correct proportion of water flowing through the diesel engine water cooler and bypassing the diesel engine jacket water cooler to maintain'a preset jacket water temperature. He are requesting exemption from the testing requirements since (1) this valve functions only as a regulating valve and not open/closed valve; (2) this valve is demonstrated operable during diesel generator testing. Diesel generators are tested on a staggered basis, every 31 days per Technical Specification 4.8.1-.1.2. The valve will be verified operable by observing proper temperatures during diesel testing.

NOTE 3: XRV-226 and -227 -Startin Air Code Relief : The starting air valves are installed on parallel air supply lines to the emergency diesel generator (EDG). The valves are not equipped with-position indication devices to directly measure valve stroke times. The valves function to provide starting air which rolls the EDG. The valves are functionally redundant to each other. These valves fail

<<as is," and, therefore, they have no fail safe position.

Successful starting of the EDG in accordance with Technical Specification 4.8.1.1.2 (i.e., slow start at least quarterly and fast start once every 184 days within 10 seconds) will verify the valve performance. The valve stroke timing will be verified by Page 61 of 81

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 1-5151D-28 Revision No: 3 Date: 2-5-90 Note 3 (continued) measuring diesel starting times during fast start testing of EDG.

The valves on a staggered basis will be valved out one at a time to verify the operability of the opposite valve during slow start of EDG at, least quarterly. Position indication will be confirmed during the above testing when only one starting air train is used to start the diesel generators.

NOTE 4 XRV-225 -Jet Assist Code Relief : This valve's function is to.

facilitate the EDG fast start by providing an air boost to the turbo charger to assist in starting the EDG in its Technical Specification 4.8.1.1.2 time limitation of 10 seconds. The valve is not equipped with position indication devices; therefore, meaningful stroke times are not achievable. The valves will be full stroke and fail safe tested by verifying EDG starting time once per 184 days in accordance with Technical Specification 4.8.1.1.'2 ~

page 62 of 81

DONALD C. COOK NUCLEAR PLAN VALVE TEST PROGRAM RELIEF REQUEST NOTES Plow Diagram No: 12-5115A Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: QCR-919 and -920 Code Relief : See "Attachment-A" for permissible seat leakage values.

page 63 of 81

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 12-5120B-22 Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: PA-343 Code Relief : This check valve is located in the maintenance air supply line into the containment. The valve cannot be tested during power operation and cold shutdown because: 1) this line is generally isolated by removing a spool piece and inserting a blind flange, and 2) the valve and test connections are located inside the containment. The valve is not equipped with position indicator. Due to the plant design, the only method available to verify the valve closure is leak testing. The-valve will be verified closed in conjunction with Appendix J seat leakage

~

testing at refueling frequency.

NOTE 2: PA-343 PCR-40 XCR-100 throu h -103 Code Relief See "Attachment-A" for permissible seat leakage values.

NOTE 3: XCR-100 -101 -102 -103 Cold Shutdown Justification : These air operated containment isolation valves located in the control air supply lines to the containment. These valves cannot be full stroke tested during power operation without causing a loss of containment control air. Testing of these valves can potentially cause: 1) disruption of air flow to air operated valves in the containment; as a result they would go to their fail safe

.position, e.g., close position for containment isolation valves, 2) systems from performing their design function, i.e, termination of system flow and change in RCS pressure and temperature, and 3) challenge to system safeguard protection which may result in a unit trip. The valves will be full stroke exercised and timed at cold shutdown frequency.

Page 64 of 81

,I

'0

DONALD C. COOK NUCLEAR 'P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 12-5131-.19 --Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: CS-427N Cold Shutdown Justification : This valve is located in the emergency boration path. This valve cannot be tested during power operation without inserting large negative reactivity which would result in unit. shutdown. The valve will be full= stroke exercised at cold shutdown frequency.

Page 65 of Sl

DONALD C. COOK NUCLEAR PLA VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 12-5136-25 Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: SF-151 and -153 Code Relief : See "Attachment-A" for permissible seat leakage values.

Page 66 of 81

0 DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 12-5137A-21 Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: DCR-201 throu h -207 -610 -611 -620 and -621 N-160 SF-159 and

-160 Code Relief See "Attachment-A" for permissible seat leakage values.

NOTE 2: N-160 Code Relief : This containment isolation check valve is located in the Nitrogen Supply line to Reactor Coolant Drain Tank.

This valve cannot be part or full stroke exercised due to, lack of sufficient differential pressure to back seat the valve during power operation or cold shutdown. Due to the plant design, the only method available to verify the valve closure is leak testing. The valve is not equipped with position indicator. This valve will be verified closed in conjunction with Appendix Z seat leakage testing at refueling frequency.

Page 67 of 81

II' C

P

DONALD C. COOK NUCLEAR P VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: .12-5141C-8 Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: ECR-416 -417 -496 -497 -535 and -536 Code Relief :

See "Attachment-A" for permissible seat leakage values.

Page 68 of 81

DONALD C. COOK NUCLEAR PLAN VALVE TEST PROGRAM RELIEF REQUEST NOTES Flow Diagram No: 12-5141F-6 Unit-1 Revision No: 3 Date: 2-5-90 NOTE 1: ECR-36 Cold Shutdown Justification : This valve, located in the common sample .return line of the lower containment radiation monitors, cannot be part or full stroke exercised during power operation or refueling because closure of the valve would isolate both radiation monitors, which are required to be operable (Technical Specification Table 3.3-6) during power operation (Mode 1 through 4) and refueling (Mode 6). The valve will be full stroke exercised at cold shutdown frequency.

NOTE 2: -ECR-31 33 36 and SM-1 Code Relief See "Attachment-A" for permissible leakage values.

NOTE 3 SM-1 Code Relief : This containment isolation check valve for the containment radiation monitors'ample return cannot be full or part stroke exercised during power operation because these monitors are required to be operable in Modes 1, 2, 3, 4, and 6. The valve is not equipped with position indication. The valve is located in the open ended return line inside the containment. The only method available to verify the valve closure is leak testing. The .valve will be verified closed in conjunction with Appendix J seat leakage testing at refueling frequency.

Page 69 of 81

DONALD C. COOK NUCLEAR .P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT gl ATTACHMENT-A Revision No: 3 Date: 2-5-90

1. CONTAINMENT ISOLATION VALVES Cate or A or AC Testing Method: (SLT-2) Seat leakage test the valve in accordance with 10CFR50; Appendix J, in lieu of ASME Code-Section XI except for paragraphs IWV-3426 and INV-3427, which are applicable (refer to Figure 3, Item g3).

Flow Permissible Leakage Valve No ~Dia ram Size ~Te WCR-920,-922 5114A 3 DA 900 WCR-921,-923 5114A 3 DA 900 WCR-932,-934 5114A 3 DA 900 WCR-933,-935 5114A '

DA 900 WCR-941,-945 5114A DA 900 WCR-944,-948 5114A DA 900 WCR-951,-955 5114A DA 900 NCR-954,-958 5114A DA 900 WCR-924,-926 5114A DA 900 WCR-925,-927 5114A DA 900 WCR-928,-930 5114A DA 900 page 70 of 81

0 DONALD C. COOK NUCLEAR PLA ASME SECTION XI VALVE TEST PROGRAM FOR UNIT Nl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size WCR-929,-931 5114A DA 900 WCR-942,-946 5114A DA. 900 WCR-952,-956 5114A DA 900 WCR-943,-947 5114A DA 900 WCR-953,-957 5114A DA 900 WCR-960,-962 5114A DA 750 WCR-961,-963 5114A DA 750 WCR-964,-966 5114A -DA 750 WCR-965,-967 5114A DA 750 ECR-10,-20 5141B 0.50 GL 750 ECR-11,-21 5141B 0 '0 GL 750 ECR-12,-22 . 5141B 0.50 GL 750 ECR-13,-23 5141B 0.50 GL 750 ECR-14,-24 5141B 0.50 GL 750 Page 71 of 81

DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT Nl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size ~Te ECR-15,-25 5141B 0.50 GL 750 ECR-16,-26 5141B 0.50 GL 750 ECR-. 17,-27 5141B 0.50 GL 750 ECR-18,-28 5141B 0.50 GL 750 ECR-19,-29 5141B 0.50 GL 750 CS-442-1 5128A CK 750 CS-442-2 5128A CK 750 CS-442-3 5128A CK 750 CS-442-4 5128A CK 750 SI-189 5128A CK 1200 SM-1 5141F CK 750 N-102 5143 CK 750 N-159 5128A 0.75 CK 750.

PW-275 5128A CK 900 page 72 of 81

DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT gl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size ~Te CS-321 5129 3 CK 1800 VCR-10,-11 5146B DA 1200 VCR-20,-21 5146B DA 1200 DCR-203,-207 5137A DA,GL 750 N-160, DCR-201 5137A CK, DA 1125 DCR-610,-611 5137A 2.50 DA 750 DCR-620,-621 5137A DA . 750 DCR-205,-206 5137A DA 1200 DCR-600,-601 5124 DA 900 QCR-300,-301 5129 GL 750 QCM-250,-350 5129A GA 1200 QCR-919,-920 5115A DA 750 SF-152,-154 5136 2.50 DA,GL 750 SF-159,-160 5137A DA 900 Page 73 of 81

/

DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT Nl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size NCR-105,-106 5141 0.50 GL 750 NCR-107,-108 5141 0.50 750 NCR-109,-110 5141 0.50 GL 750 RCR-100,-101 5128A 0.375 GL 750 DCR-202,-204 5137A 0.75 DA 750 ICR-5,-6 5141 0.50 GL 750 ECR-33,-35 5141F 0.75,2 GL,DA 750 ICM-260 5142 GA(DD)

600 ICM-265 5142 GA(DD)

600 ECR-31,-32 5141F GL 750 XCR-100,-101 5120B GL 750 XCR-102,-103 5120B GL 750 GCR-301 5128A 0.75 DA 375 GCR-314 5143 . GL 375

.Double Discs Page 74 of 81

DONALD C. COOK NUCLEAR",P ASME SECTION XI VALVE TEST. PROGRAM FOR UNIT Nl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size ~Te SI 17lg 172'94 5143 0.75 GL 1125 NCR-252 5128A GL 450 CCR-460,-462 5135 GL 900 CCR-457,CCW-135 5135 2,2.50 GL,CK 1125 CCR-455,-456 5135 GL 750 SM-4,-6 5147A 0.50 GL 750 ICM-251 5142 GA(DD)

600 ICM-250 5142 GA(DD)

600 CA-181S 5145 0.50 CK 750 CA-181N 5145 0.50 CK 750 SM-8,-10 5147A 0.50 ND 750 CCW-243-25 5135B CK 750 CCW-244-25 5135B CK- 750

Double Discs Page 75 of 81

0 DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT gl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve 'No. ~Dia ram Size CCW-243-72 5135B CK 750 CCW-244-72 5135B CK. 750 CCM-430 5135B 1.50 GL 375 CCM-431 5135B 1.50 GL 375 CCR-440 513B 1 '0 GL 375 CCR-441 5135B 1.50 GL 375 CCM-432 5135B 1.50 GL 375 CCM-433 5135B 1.50 GL 375 R-156 5146B 0.375 CK 750 R-157 5146B 0.375 CK 750 NS-357 5124 0.50 CK 750 ECR-496,-497 5141C 0.50 GL 750 ECR-416 5141C 0 '0 GL 375 ECR-417 5141C 0.50 GL 375 ECR-535 5141C 0.50 GL 375 Page 76 of 81

DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT Nl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size ~Te ECR-536- 5141C 0.50 GL 375 ECR-36 5141F DA 375 PCR-40 5120B 2 GA 375-

,PA-342 5120B CK 750 NS-283 5141D 0.50 CK 750 NPX-151 5128A 0.50 375 WCR-900,-902 5114A DA 1800 WCR-901,-903 5114A DA 1800 WCR-912,-914 5114A DA 1800 WCR-913,-915 5114A DA 1800 WCR-904,-906 5114A DA 1800 WCR-905,-907 5114A DA 1800 WCR-908,-910 5114A DA 1800 WCR-909,-911 5114A DA 1800 VCR-101,-201 5147A 14 BF 4200 Page 77 of 81-

e l L

DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT N1 ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size ~Te VCR-102,-202 5147A 14 BF 4200 VCR-103,-203 5147A '24 BF 7200 VCR-104,-204 5147A 30 BF 9000 VCR-105,-205 5147A 30 BF 9000 VCR-106,-206 5147A 24 BF 7200 VCR-107,-207 5147A 14 BF 4200 ICM-305 5143 18 GA(DD)

2700 ICM-306 5143 18 GA(DD)

2700 CCM 452 g 454 I 458 5 135 8I.4g8 BFIGLgBF 3000 CCM 451 / 453 I 459 5135 8,4,8 BF,GL,BF 3000

Double Discs Page 78 of 81

DONALD C. COOK NUCLEAR PLA ASME SECTION XI VALVE TEST PROGRAM FOR UNIT gl ATTACHMENT-A Revision No: 3 Date: 2-5-90 2 ~ CONTAINMENT SPRAY VALVES Cate or A or AC Testing Method: As described in "SLT-2A," Figure 3, Item g3.

Flow Permissible Leakage Valve No. Diaciram Size ~Te CTS-13 1H 5144 CK 35.00 CTS-131E 5144 CK 35.00

- CTS-127H 5144 CK 22.55 CTS-127E 5144 CK 21.21 RH-141 5144 CK 20.70 RH-142 5144 CK 23.00 page 79 of 83.

DONALD C. COOK NUCLEAR PLA ASME SECTION XI VALVE TEST PROGRAM FOR UNIT Nl ATTACHMENT-A Revision No: 3 Date: 2-5-90

3. PRESSURE ISOLATION VALVES Cate o A or AC Testing Method: (SLT-1) Seat leakage test the valve per ASME Code Section XI (refer to Figure 3, Item g3).

Flow Permissible Leakage Valve No. Diaciram Size ~Te CS-299E 5129 CK 2.0 CS-299N 5129 CK 2.0 SI-152-N 5143 CK 5.0 SI-152-S 5143 CK 5.0 ICM-129 5143 14 GA(DD)

10.0 SI-161-Ll,-L4 5143 CK 10.0 SI-161-L2,-L3 5143 CK 10.0 SI-170-Ll 5143 10 CK 5.0 SI-170-L2 5143 10 CK 1.0 SI-170-L3 5143 10 CK 1.0

Double Discs Page 80 of 81

DONALD C. COOK NUCLEAR P ASME SECTION XI VALVE TEST PROGRAM FOR UNIT gl ATTACHMENT-A Revision No: 3 Date: 2-5-90 Flow Permissible Leakage Valve No. Diaciram Size ~Te SI-170-L4 5143 10 CK 5.0 SI-158-Ll,-L4 5143 CK. 10.0 SI-158-L2,-L3 5143 CK 10.0 SI-151-E 5143 CK 5.0 SI-151-W 5143 CK 5.0 SI-166-Ll 5143 10 CK 5.0 SI-166-L2 5143 10 CK 5.0 SI-166-L3 5143 10 CK 5.0 SI-166-L4 5143 10 CK 5.0 RH-133,-134 5143 CK 1.0 Page 81 of 81

-ag I

DONALD C. COOK NUCLEAR PLANT RUH DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SmARY SHEET UNIT 1 SYSTEH NAHE: HAIN STEAN FLOH DIAGRAN: 1-5105-29 VALVE I VALVE POSITION I ASHE SECTION XI HUNBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIN TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORNED NODE 1-DCR-310 3 GL 2 A 0/5 0 2 A B EF1 EF-1 P NO, EF-5 EF-5 HO EF-7 EF-7 P NO ET-XXX ET-XXX P HO 1-DC R-320 3 GL 2 A D/5 0 2 A 'B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P HO 1-DCR-330 3 GL 2 A D/5 0 2 A B EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO I-DCR-300 3 GL 2 A D/5 0 2 A B EF1 EF-1 P NO E F-5 EF-5 HO EF-7 EF"7 P HO ET-XXX ET-XXX P HO

0 0 0 0

e e e e o o e e e e o o e o e o

DONALD Co COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: HAIN STEAM FLON DIAGRAM: 1"51058-35 VALVE I VALVE POSITION I ASHE SECTION XZ NUMBER REV TYPE SIZE ACT F.D. l PONER- SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDl OPER FUNCT ICL REQUIRED PERFORMED HODE 1-QT-506 3 GL MO L/8 C A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: l5FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SU%AERY SHEET - UNIT 1 SYSTEM NAME: FEEDHATER FLOH DIAGRAM: 1-5105D-1 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F ~ D t PONER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-FH-118-1 3 CK 14 SA C/4 0 2 A C CF1 CF-2 R YESp NOTE 1 1-FH-118-2 3 CK 14 SA K/4 0 2 A C CF1 CF-2 R YES NOTE 1 1-FH-118-3 3 CK 14 SA K/8 0 2 A C CF1 CF-2 R YESp NOTE 1 1-FN-118-4 3 CK 14 SA B/9 0 2 A C CF1 CF-2 R YESp NOTE 1 I"MCM-221 3 GL 0/C 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-MCM-231 3 GL 4 0/C 2 A B EF-1 EF"1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P HO 1-MRV"210 3 GA 28 PO 8/3 0 2 A B EF1 EF-3 P HOp CSJ 2 EF-5 EF-5 NO EF-7 EF-8 C NOp CSJ 2 ET-XXX ET-XXX C NOp CSJ 2 1-MRV-211 3 AG 2 A A/1 C 2 A B EF1 EF-1 P NO Ef-5 EF-5 NO EF-7 EF"7 P NO ET-XXX ET-XXX P NO 1-MRV-212 3 AG 2 A A/1 C 2 A B EF1 EF-1 P NO EF-5 EF-5 HO EF-7 EF-7 P NO ET-XXX ET-XXX P NO

e e e e e e e e. e DONALD C. COOK NUCLEAR PLANT RUiN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SRSARY SHEET - UiNZT 1 SYSTEM NAHE: HAIN STEAM FLON DIAGRAM: 1-5105D-1 VALVE I VALVE POSITION I ASME SECTION XI NASSER REV TYPE SI2E ACT F.D ~ I PONER SAFETY I CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1"HRV-220 3 GA 28 PO L/3 0 2 A 8 EF-1 EF-3 P NO> CSJ 2 EF"5 EF-5 NO EF-7 EF-8 C NO> CSJ 2 ET-XXX ET-XXX P NO> CSJ 2 1-HRV-221 3 AG 2 A H/1 C 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-HRV-222 3 AG 2 A H/1 C 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-HRV>>230 3 GA 28 PO L/7 0 2 A B EF-1 EF-3 P NO> CSJ 2 EF-5 EF-5 NO EF-7 E F-8 C NO> CSJ 2 ET-XXX ET-XXX P NOp CSJ 2 1-HRV-231 3 AG 2 A H/5 C 2 A 8 EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-HRV-232 3 AG 2 A H/5 C 2 A B EF-1 EF-1 P NO EF-5 E F-5 NO EF-7 EF-7 P NO ET"XXX ET-XXX P NO

e e o o e o e o e o e e e e e -

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SU DARY SHEET - UNIT 1 SYSTEM NAME: HAIN STEAH FLON DIAGRAM: 1-5105D-1 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. ) PONER SAFETY )CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED NODE 1-HRV-240 3 GA 28 PO B/7 0 2 A B EF-1 EF-3 P NOp CSJ 2 EF-5 EF"5 NO EF-7 EF-8 C NOp CSJ 2 ET-XXX ET-XXX C NOp CSJ 2 1-HRV"241 3 AG 2 A A/5 C 2 A B EF1 EF-1 P NO E F-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-MRV-242 3 AG 2 A A/5 C 2 A B EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-HS-108-2 ' CK 4 SA K/4 C 0/C 3 A C CF-1 CF-2 R YESp NOTE 3" 1-MS-108-3 3 CK SA K/4 C 0/C 3 A C CF-1 CF-2 R YESp NOTE 3 1-SV-1A-1 3 REL 6 SA C/1 C 2 A C TF1 TF-1 R NO 1-SV-1A-2 3 REL 6 SA K/1 C 2 A C TF-1 TF-1 R NO 1-SV-)A-3 3 REL 6 SA K/5 C 2 A C TF-1 TF"1 1-SV-1A-4 3 REL 6 SA C/5 C 2 A C TF"1 TF-1 R NO

0 0 0

DONALD C. COOK NUCLEAR PLANT SECOND TEN YEAR INTERVAL RUN DATE AND TINE: 15FEB90:16:03 VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: HAIN STEAH FLON DIAGRAH: 1-51058-1 VALVE I VALVE POSITION I ASHE SECTION XI

'UHBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUESTERS)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED HODE 1-SV-18-1 3 REL 6 SA 8/1 C 2 A C TF1 TF-1 R NO 1-SV-18-2 3 REL 6 SA L/1 C 2 A C TF-1 TF"1 R NO 1-SV-18-3 3 REL 6 SA L/5 C 2 A C TF1 TF-1 R NO 1-SV-18-4 3 REL 6 SA 8/5 C 2 A C TF1 TF-1 R NO 1-SV-2A-1 3 RE L 6 SA 8/1 C 2 A C TF"1 TF-1 R NO 1-SV-2A-2 3 REL 6 SA L/1 C 2 A C TF1 TF-1 R NO 1-SV-2A-3 3 REL 6 SA L/5 C 2 A C TF1 TF-1 1-SV-2A-4 3 REL 6 SA 8/5 C 2 A C TF"1 TF-1 R NO 1-SV"28"1 3 REL 6 SA A/1 C 2 A C TF-1 TF-1 1-SV-28-2 3 REL 6 SA L/1 C 2 A TF 1 TF-1 R NO 1>>SV-28-3 3 REL 6 SA L/5 C 2 A C TF1 TF-1 R NO 1-SV-28-4 3 REL 6 SA 8/5 C 0 ->>

2 A C TF-1 TF-1 R NO

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TZHE: 15FEB90:16:03 SECOND TEN YEAR ZNTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: HAIN STEAH FLON DIAGRAH: 1-5105D-1 VALVE VALVE POSITION ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY (CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORHED HODE 1-SV-3-1 3 REL 6 SA A/1 C 2 A C TF-1 TF-1 R NO 1-SV-3-2 3 REL 6 SA L/1 C 2 A C TF1 TF-1 R NO 1-SV-3-3 REL 6 SA L/5 C 2 A C TF-1 TF-1 R NO I-SV-3-4 REL 6 SA A/5 C 2 A C TF1 TF"1 R NO

0 e e e e e e e e e e e e e o e o e e o e e o o o e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FE890:16:03 SECOND TEN YEAR INTERVAL VALVE SUKSLRY SHEET - UNIT 1 SYSTEH NAHE: FEEDHATER FLOH DIAGRAH: 1"5106-35 VALVE I VALVE POSITION I ASHE SECTION XI NUYBER REV TYPE SI2E ACT F.D. I POHER SAFETY lCD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1-FH0-201, 3 GA 14 HO F/5 0 C 2 A 8 EF-1 EF-2 C NOp CSJ 1 EF-5 EF"5 NO ET-XXX ET-XXX C NOp CSJ 1 1-FHO-202 3 GA 14 HO F/9 0 C 2 A 8 EF-1 EF-2 C NOp CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NO) CSJ 1 1-FHO-203 3 GA 14 HO F/9 0 C 2 A 8 EF-1 EF"2 C NOp CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NOp.CSJ 1 I"FYiO"204 3 GA 14 HO G/5 0 C 2 A 8 EF-1 C NO) CSJ 1 EF-5 E F-5 NO ET"XXX ET-XXX C NOp CSJ 1 1-FRY-210 3 AG 14 A G/5 0 C 2 A 8 EF-1 EF-2 C NOp CSJ 1 EF-5 EF-5 NO EF-7 EF-8 C NOp CSJ 1 ET-XXX ET-XXX C NO) CSJ 1 1-FRV-220 3 AG 14 A E/9 0 ' 2 A 8 EF-1 EF-2 C NOp CSJ 1 EF-5 EF-5 NO EF-7 E F-8 C NO) CSJ 1 ET-XXX ET"XXX C NOp CSJ 1 1-FRV-230 3 AG 14 A G/9 0 C 2 A 8 EF-1 EF-2 C NOp CSJ 1 EF-5 EF"5 NO EF-7 EF-8 C NOp CSJ 1 ET-XXX ET-XXX C NO) CSJ 1

0 0 0 0 0 0 0 0 0 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: FEEDHATER FLON DIAGRAH: 1-5106-35 VALVE VALVE POSITION ASHE SECTION XI NUHBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORHED NODE 1-FRV-260 3 AG 14 A H/5 0 A B EF1 EF-E C NO> CSJ 1 EF"5 EF-5 NO EF-7 EF-8 C NOy CSJ 1 ET-XXX ET-XXX C NOy CSJ 1

0 e e o o e e o e e o e o e e I I DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:D3 SECOND TEN YEAR INTERVAL VALVE SUHMARY SHEET - UNIT 1 SYSTEM NAME: FEEDHATER FLON DIAGRAM: 1-5106A-38 VALVE I VALVE POSITION I ASHE SECTION XI NUN3ER REV TYPE SIZE ACT F.D. I PONER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF. REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-FMO"211 3 GL 4 MO J/4 0 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-FMO-212 3 GL HO J/5 0 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-FMO-221 3 GL HO F/5 0 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX 1-FMO-222 3 GL 4 HO F/6 0 0 3 A 8 EF-1 EF-1 EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-FMO-231 3 .

GL MO F/5 0 0 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-FHO-232 3 GL 4 MO F/6 0 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-FMO-241 3 GL HO J/5 0 0 3 A 8 EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX

e e e e o e e e o e e e e e DONALD C. COOK NUCLEAR PLANT RUN DATE,AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SlÃHARY SHEET - UNIT 1 SYSTEH NAHE: FEEDWATER FLON DIAGRAH: 1-5106A-38 VALVE I VALVE POSITION l ASHE SECTION XI NJii3ER REV TYPE SI2E ACT F.D. I PONER SAFETY I CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FlliCT ICL REQUIRED PERFORHED HODE 1-FHO-242 3 GL HO J/5 0 3 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1"FRV-247 3 GL 1 A C/8 0 C/0 3 A B EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX P NO 1-FRV-257 3 GL 1 A F/8 0 C/0 3 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1"FRV"258 3 GL 1 A J/9 0 C/0 3 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-FW-124 3 CK 8 SA H/7 C 3 A C CF-1 CF-1 P NO 1" FN"128 3 CK 6 SA F/7 C 3 A C CF1 CF-1 1-FN-132-1 3 CK 4 SA H/5 C 0/C 2 A C CF-1 CF-2 NO> CSJ 1 1-FN-132-2 3 CK SA F/6 C 0/C 2 A C CF-1 CF-2 NO> CSJ 1

DONALD C. COOK NUCLEAR PLANT SECOND TEN YEAR INTERVAL

'ALVE

SUMMARY

SHEET UNIT 1 SYSTEM NAME: FEEDHATER FLOH DIAGRAM: 1-5106A-38 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. t POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORMED MODE 1" FH-132-3 3 CK SA F/6 C 0/C 2 A C CF-1 CF-2 NO> CSJ 1 1-FH-132-4 3 CK SA N/5 C 0/C 2 A C CF-1 CF-2 NO> CSJ 1 1-FH-134 CK 10 SA L/9 C 3 A C CF-1 CF-3 C NO> CSJ 2 1-FH-135 3 CK 8 SA J/8 C 3 A C CF-1 CF-3 C NO> CSJ 2 1-FH"138-1 3 CK SA K/4 C .0/C 2 A C CF-1 CF-2 NO> CSJ 3 1-FH-138-2 3 CK 4 SA F/5 C 0/C 2 A C CF-1 CF-2 NO> CSJ 3 1-FH-138-3 3 CK SA F/5 C 0/C 2 A C CF-1 CF-2 NO> CSJ 3 1-FH-138-4 3 CK SA H/4 C 0/C 2 A C CF-1 CF-2 NO) CSJ 3 1-FH-149 3 CK 0. 75 SA L/3 C 3 A C CF-1 CF-1 P NO> NOTE 4 1-FH-150 CK 0.75 SA L/4 C 3 C CF1 CF-1 P NO~ NOTE 4 1-FH-153 3 CK 1 SA F/8 C 0/C 3 A C CF-1 CF-1 P NO> NOTE 5 1" FH-159 3 CK 6 SA C/7 C 3 A C CF1 CF-1 P NO

e e e e e e e e a e 0 e e e e ll g

0 0 e o o e o o.o e e e e o o o e

III DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUaiARY SHEET UNIT 1 SYSTEH NAHE: FEEDHATER FLOH DIAGRAH: 1-5106A-38 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORNED NODE 1-FH-160 3 CK 1 SA C/8 C 0/C 3 A C CF-1 P NO> NOTE 5 1-FH-161 3 CK 8 SA E/7 C 0 3 A C CF1 P NO I-SV-160-A 3 REL 0. 75 SA H/Z C - .0 3 A C TF1 TF"1 R NO I-SV-140-B 3 REL -0. 75 SA L/Z C 0 3 A C TF1 TF-1 1-SV"169-A 3 REL 0. 75 SA 0/8 C 0 3 A C TF-1 TF-1 R NO 1-SV-169-B 3 RE L 0. 75 SA G/8 C 0 3 A C TF-1 TF-1 R NO 18-CRV-51 3 GL 8 A H/8 0/C C 3 A B EF1 EF-1 EF-5 EF-5 EF-7 EF-7 ET-XXX ET-XXX

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: ESSENTIAL SERVICE HATER F LOH DIAGRAM: 1-5113-41 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F ~ D ~ I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-ESH "101-E 3 CK 20 SA N/8 C 3 A C CF1 CF"1 P NO 1-ESH-101-H 3 CK 20 SA H/8 C 3 A C CF1 CF-1 1"ESH-109 3 BF 4 H B/5 C 3 A B EF1 EF-2 C NOp CSJ 2 1-ESH-ill 3 CK 6 SA C/6 C 3 A C CF-1 CF-1 P NOp NOTE 1 1-ESH-112 3 CK 6 SA C/6 C A C CF-l, CF-1 P NOp NOTE 1 1-ESH-113 3 CK 6 SA B/6 C 3 A C CF-1 CF-1 P NOp NOTE 1 1-ESH-114 3 CK 6 SA B/6 C 3 A C CF1 CF-1 P NOp NOTE 1 1-ESH-115 3 BF 6 N E/6 C 3 A B EF1 EF-2 C NOp CSJ 2 1-ESH-168-N 3 BF M H/1 C 3 A B EF1 EF-1 1-ESH-168-S 3 BF M H/1 C A B EF-1 EF-1 1-ESH-169-N 3 BF 3 M G/1 0 A B EF1 EF-1 1-ESH-169-S 3 BF 3 M G/1 0 3 A B EF1 EF-1

I e e e e e e e e o o e e e o o 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03

~ ~

SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEM NAME: ESSENTIAL SERVICE HATER FLON DIAGRAM: 1-5113-41 VALVE I VALVE POSITION I ASHE SECTION XZ NUMBER REV TYPE SIZE ACT F ~ D. I POHER SAFETY I CD A/P CAT PRZH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-ESN-170-N 3 BF 3 H F/1 0 3 A B EF1 EF-1 P NO 1"ESN-170-S 3 BF 3 H F/1 0 ABEf1 EF-1 P NO 1-ESN-171-N 3 Bf 3 H F/1 C 3 A B EF1 EF-1 1-ESN-171"S 3 BF 3 H F/1 C 3 A B EF-1 EF-1 I-ESN"243 3 BF H 0/6 C 3 A B EF1 EF-2 C NO) CSJ 2 1-SV-14-E 3 REL 1 SA A/1 C 3 A C TF1 TF-1 1-SV-14-N 3 RE L 1 SA C/1 C 3 A C TF1 TF-1 R NO I-SV-15-E 3 REL 0.75 SA E/4 C A C TF-1 TF-1 R NO 1-SV-15-H 3 REL 0.75 SA G/4 C A C TF-1 TF-1 R NO 1-SV-16-AB 3 REL 1 SA C/8 C 3 A C TF-1 TF-1 1-SV-16-CD 3 REL 1 SA C/8 C 3 A C TF-1 TF"1 I-NMO-701 3 BF 20 HO N/8 C 3 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

0 0 0 0 0 0 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: ESSENTIAL SERVICE HATER FLON DIAGRAM: 1-5113-41 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F ~ D. I POHER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUESTS S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-HHO-702 3 BF 20 HO H/S C 0 3 A B EF1 EF"1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-HHO-705 BF 20 HO G/6 0 0/C 3 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-NHO-707 BF 20 HO G/7 0 0/C 3 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-NMO "713 3 BF 12 HO A/5 C 3 A B ~ EF-'1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-HHO-717 3 BF 12 HO B/5 C 3 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-NO-721 3 BF 6 HO D/6 C 0 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-HMO-723 3 "BF 6 HO C/6 C 3 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

0 DONALD C. COOK NUCLEAR LANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT, 1 SYSTEM NAME: ESSENTIAL SERVICE I'%TER FLON DIAGRAM: 1-5113-41 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POMER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE I-HYiO-725 3 BF 6 HO 8/6 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-HHO-727 3 BF 6 HO 8/6 C 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-MMO-733 3 BF 16 HO C/3 0/C 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-WO-737 3 BF 16 HO E/3 0/C 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-MMO-744 3 BF MO D/6 C 3 A 8 EF1 EF-1 NO EF-5 EF-5 NO ET-XXX ET-XXX '

P NO 1-HMO-753 3 BF 6 HO D/6 C A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-HMO-754 3 BF 4 HO 8/5 C 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE %MARY SHEET - UNIT 1 SYSTEM NAME: ESSENTIAL SERVICE HATER - FLOH DIAGRAM: 1-5113-41 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-HRV-721 3 3H 4 A 0/8 0 A B EF1 NOTE 3 P YES> NOTE 3 EF-7 EF"8 R YES'OTE 3 ET-XXX ET-XXX YES>. NOTE 3 I-WRV-723 3 3H A B/8 0 3 A B EF1 NOTE 3 P YES> NOTE 3 EF-7 EF-8 R YES> NOTE 3 ET-XXX ET-XXX YES'OTE 3 1-HRV-725 3 3W 4 ~ A D/8 0 3 A B EF1 NOTE 3 P 'ES> NOTE 3 EF-7 EF-8 R YES> NOTE 3 ET-XXX ET-XXX YES> NOTE 3 1-HRV-727 3 3H A B/8 0 3 A B EF1 NOTE 3 P YES> NOTE 3 EF-7 EF-8 R YES> NOTE 3 ET"XXX ET-XXX YES> NOTE 3

I DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: NON"ESSENTIAL SERVICE HATER FLOH DIAGRAM: 1-5114A-31 VALVE I VALVE POSITION I ASME SECTION XZ NUMBER REV TYPE SIZE ACT F 0 I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-HCR-900-1 3 DA 6 A J/9 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1 "HCR-901-1 3 DA 6 A K/9 0 C 2 A A EF-1 EF-1 P NO EF-5 E F-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P -NO SLT-1 SLT-2 R YES'OTE 1 1-HCR-902-1 3 DA 6 A J/4 0 C 2 A A EF-1 EF-1 P NO EF"5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-903-1 3 DA 6 A K/4 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX - ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-904-2 3 DA 6 A J/9 0 C 2 A A EF-1 EF-1 P NO EF-5 EF"5 NO EF-7 . EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES'OTE 1

0 o o o e e e o e o o e e e e e 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AHD TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET " UNIT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE HATER FLOH DIAGRAM: I"511%A-31 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST -

TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE I-NCR-905-2 3 DA 6 A K/9 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 HO EF-7 EF-7 P HO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-NCR-906-2 3 DA 6 A J/0 0 -2 A A EF-1 EF-1 P HO EF-5 EF"5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1"HCR-907-2 3 DA 6 A K/4 0 2 A A EF-1 EF-1 s P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES'OTE 1 1-HCR-908-3 3 DA 6 A J/9 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P HO SLT") SLT-2 R YES> NOTE 1 1-HCR-909-3 3 DA 6 A K/9 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1

0 0 0 0 0 0 0 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE WATER FLOW DIAGRAM: I-511%A-31 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F D I POWER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORMED MODE 1-WCR-910-3 3 DA 6 A J/4 0 2 A A EF-1 Ef-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-WCR-911-3 3 DA 6 A K/4 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 'R YES> NOTE 1 1-WCR-912-0 3 DA 6 A J/9 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 I-WCR-913-0 3 DA 6 A K/9 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 I-WCR-910-0 3 DA 6 A J/0 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1

0 0 e o e e, e e e e o o e e e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB9D:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET UiiZT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE HATER FLOH DIAGRAM: 1-5114A-31 VALVE I VALVE POSITION l ASME SECTION XZ NUMBER REV TYPE SIZE ACT F ~ D. I POHER SAFETY ICD A/P CAT PRZM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT lCL REQUIRED PERFORMED MODE 1-HCR-915-4 3 DA 6 A K/4 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-920-1 3 DA A J/6 0 2 A A EF-1 EF-1 P NO EF-5 EF-5, NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1 "HCR-921-1 3 DA 3 A K/6 0 2 A A EF-1 EF-1 P NO E F-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-922-1 3 DA A K/2 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-923-1 3 DA 3 A J/2 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES'OTE 1

0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUGARY SHEET - UNIT 1 SYSTEN NAHE: HON-ESSENTIAL SERVICE HATER FLON DIAGRAN: 1-5114A-31 VALVE I VALVE POSITION I ASNE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIN TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUHCT ICL REQUIRED PERFORHED NODE 1-HCR-924-2 3 DA 3 A J/6 0 2 A A EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT"2 R YESp NOTE 1 1-NCR-925-2 3 DA 3 A K/6 0 2 A A EF-1 EF-1 P NO E F-5 EF-5 NO EF-7 EF"7 P HO ET".XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-NCR-926-2 3 DA 3 A K/2 0 2 A A EF-1 EF P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-HCR"927-2 3 DA 3 A J/2 0 2 A A EF-1 EF-1 P NO EF"5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES'OTE 1 1-NCR"92S-3 3 DA 3 A J/6 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 HO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1

o e e e e r e e e e o e e e e DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE HATER FLOH DIAGRAM: I-5114A-31 VALVE l VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. l POHER SAFETY lCD A/P. CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDl OPER FUNCT ICL REQUIRED PERFORMED MODE 1-HCR-929-3 3 DA 3 A K/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-930-3 3 DA 3 A K/2 0 C 2 A A EF-1 EF-1 P NO EF-5 E F-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-931-3 3 DA A J/2 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-932-4 3 DA 3 A . J/6 0 C 2 - A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT"2 R YES> NOTE 1 I-HCR-933+ 3 DA A K/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R, YES> NOTE 1

I DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE:- 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUsiQRY SHEET - UNIT 1 SYSTEH NAHE: NON-ESSENTIAL SERVICE HATER FLOH DZAGRAH: 1-5114A-31 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F D t POHER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1"HCR-934 3 DA A K/2 0 2 A A EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 I-HCR-935-4 3 DA A J/2 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P- NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-941-1 3 DA 3 A J/6 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT"2 R YES'OTE 1 1-HCR-942-2 3 DA 3 A J/6 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-943-3 3 DA 3 A J/6 0 2 A A EF-1 EF-1 ~ P NO EF"5 EF-5 NO EF-7 EF"7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

Sg 0 e e o o e e e e e e o e e e e 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: NON-ESSENTIAL SERVICE WATER FLOW DIAGRAH: 1-5114A-31 VALVE I VALVE POSITION I ASHE SECTION XI NUiiBER REV TYPE SIZE ACT F.D. I POWER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED HODE I"WCR-944M 3 DA 3 A J/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESi NOTE 1 1-WCR-945-1 3 DA A J/3 0 C 2 A A EF-1 EF-1 E F-5 EF-5 NO EF-7 EF-7 P NO ET"XXX ET-XXX P NO SLT-1 SLT-2 R YES1 NOTE 1 1-WCR-946-2 3 DA 3 A J/3 0 C 2 A A EF-1 EF-1 E F-5 EF-5 EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 I-WCR"947"3 3 DA 3 A J/3 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESi NOTE 1 1-WCR-948-4 3 DA A J/3 0 C 2 A A EF-1 EF-1 P NO EF-5, EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESi NOTE 1

0 0 0 0 , 0 0

DONALD C. COOK NUCLEAR PLANT RUN DATE- AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE HATER FLOH DIAGRAM: I-511%A-31 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER RECT ICL REQUIRED PERFORMED MODE 1-HCR-951-1 3 DA 3 A K/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-952-2 3 DA 3 A K/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF"7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-953-3 3 DA 3 A K/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT"1 SLT-2 R . YESp NOTE 1 I-HCR-954-4 3 DA 3 A K/6 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT"1 SLT-2 R YES> NOTE 1 1-HCR-955-1 3 DA 3 A J/3 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES 1 NOTE 1

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:D3

~ ~

SECOND TEN YEAR INTERVAL VALVE RRCSlRY SHEET - UNIT 1 SYSTEH NAHE: NON-ESSENTIAL SERVICE HATER FLOH DIAGRAH: I-511%A-31 VALVE VALVE POSITION I ASME SECTION XI NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1-HCR-956-2 3 DA 3 A J/3 0 2 A A EF-1 EF-1 P NO EF-5 EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-HCR-957-3 3 DA A J/3 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF"7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 I-HCR"958-4 3 DA 3 A J/3 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 I"HCR-960-4 3 DA 2 A J/7 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 I-HCR-961-4 3 DA 2 A J/7 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT"2 R YES> NOTE 1

DONALD C. COOK NUCLEAR PLANT RUH DATE AND TZME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUiiARY SHEET - UNIT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE HATER FLOW DIAGRAM: 1-5114A-31 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F AD I POWER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FRKT ICL REQUIRED PERFORMED MODE 1-WCR-962-4 3 DA 2 A J/3 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 HO EF-7

"'F-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-WCR-963-4 3 DA 2 A J/3 0 C 2 A A EF-1 EF-1 E F-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P HO SLT-1 SLT"2 R YES> NOTE 1 1-WCR-964-3 3 DA 2 A J/7 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-WCR-965-3 3 DA 2 A J/7 0 C 2 A A EF-1 EF-1 P HO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-WCR-966-3 3 DA 2 A J/3 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> HOTE 1

0 0 0 0 0 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SRSARY SHEET - UNIT 1 SYSTEM NAME: NON-ESSENTIAL SERVICE HATER FLON DIAGRAM: 1-511%A-31 VALVE VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT t CL . REQUIRED PERFORMED MODE 1-NCR-967-3 3 DA 2 A J/3 0 2 A A EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: STATION DRAINAGE - CONTAINMENT FLOH DIAGRAM: 1-51N-22 VALVE I VALVE POSITION I ASME SECTION XZ NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDl OPER FUNCT ICL REQUIRED PERFORMED MODE 1-DCR-600 3 DA 3 . A N/6 0 C 2 A A "

EF1 EF-1 P NO EF-5 EF-5 NO Ef-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R'ES'OTE 1 1-DCR-601 3 DA A N/6 0 C 2 A A- EF1 EF-1 P NO EF-5 EF-5 NO Ef-7 EF-7 P NO ET-XXX ET"XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-NS-357 3 CK 0.5 SA K/9 C 0/C 2 A AC CF-1 CF-2 R YES> NOTE 2 SLT-1 SLT-2 R YESp NOTE 1

e e e e e e e e 0

e e o e e o e e o e e o e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: REACTOR COOLANT FLOH DIAGRAM: 1-5128-19 VALVE I VALVE POSITION ASME SECTION XZ NUMBER REV TYPE SIZE ACT F.D. t POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1"NSO-021 3 GL 1 SO E/6 C 0/C 2 A B EF"1 EF-2 C HOp CSJ 1 EF-5 EF-5 HOp CSJ 1 EF-7 EF-8 C NOp CSJ 1 ET-XXX ET-XXX C NOp CSJ 1 1-HSO-022 3 GL 1 SO E/6 C 0/C 2 A B EF-1 EF-2 C NOp CSJ 1 EF-5 EF-5 HOp CSJ 1 EF-7 EF-8 C NOp CSJ 1 ET-XXX ET-XXX C HOp CSJ 1 1-HSO-023 3 GL 1 SO E/6 C 0/C 2 A B EF-1 EF-2 C NOp CSJ 1 EF-5 EF"5 HOp CSJ 1 EF-7 EF-8 C NOp CSJ 1 ET-XXX ET-XXX C HOp CSJ 1 1-NSO-ON ~ 3 GL 1 SO E/6 C 0/C 2 A B EF-1 EF-2 C NOp CSJ 1 EF-5 EF-5 HOp CSJ 1 EF-7 EF-8 C NOp CSJ 1 ET-XXX ET"XXX C HOp CSJ 1

e e e e e e e e e e 0 o e e e 0 o e o o e o e e e o e e e o e

DONALD C. COOK NUCLEAR PLANT .RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: REACTOR COOLANT FLOW DIAGRAH: I-5128A-37 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F.D. I POWER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF 'REQUEST(S) .

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE I-CS-4%2-I 3 CK 2 SA B/4 0 2 A AC CF-1 CF"2 R YESp NOTE 1 SLT-1 SLT-2 R YESp NOTE 2 I-CS-642"2 3 CK 2 SA B/0 0 2 A AC CF-1 CF-2 R YESp NOTE 1 SLT-1 SLT-2 R YESp NOTE 2 I-CS-4%2-3 3 CK 2 SA 8/0 0 2 A AC CF-1 CF-2 R YESp NOTE 1 SLT-1 SLT-2 R YESp NOTE 2 I-CS-M2-4 3 CK 2 SA B/4 0 2 A AC CF-1 CF-2 R YESp NOTE 1 SLT-1 SLT-2 R YESp NOTE 2 1-GCR-301 3 DA 0.75 A B/8 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2 1-N-159 3 CK 0.75 SA C/8 0/C C 2 A AC CF-1 CF-2 R YESp NOTE 5 SLT-1 SLT-2 R YESp NOTE 2 1-NCR-252 3 GL 3 A B/9 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF"7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2 1-NHO-151 3 GA 3 HO K/7 0 1 A B EF-1 EF-1 P NO E F-5 EF-5 NO ET-XXX ET-XXX P NO

J DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET " UiNIT 1 SYSTEM NAME: REACTOR COOLANT FLON DIAGRAM: 1-5128A-37 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-NMO-152 3 GA 3 HO K/7 0 1- A B EF-1 EF"1 P NO EF"5 E F-5 NO ET-XXX ET-XXX P NO 1-NHO-153 3 GA 3 HO K/6 0 1 A B EF-1 EF"1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-NPX-151 3 GL 0 5 M N/8 C 2 P A SLT"1 SLT"2 R YES) NOTE 2 1-NRV-151 3 GL A K/7 C C/0 1 A B EF-1 EF-2 NOp CSJ 3 EF-5 E F-5 NO EF-7 EF-8 NO) CSJ 3 ET-XXX ET-XXX NO) CSJ 3 1-NRV-152 3 GL 3 A K/7 C C/0 1 A B EF-1 EF-2 NO) CSJ 3 EF-5 E F-5 NO EF"7 EF"8 NOp CSJ 3 ET-XXX ET-XXX NOp CSJ 3 1-NRV-153 3 GL 3 A K/6 C C/0 1 A B EF-1 EF-2 NOp CSJ 3 EF"5 EF<<5 NO EF-7 EF-8 NOp CSJ 3 ET-XXX ET-XXX NOp CSJ 3 1-NSO-061 3 GL 1 SO H/6 C 0/C 2 A B EF-1 EF-2 C NO) CSJ 6 EF-5 - EF-5 NO) CSJ 6 EF"7 EF-8 C NO) CSJ 6 ET-XXX ET-XXX C NOp CSJ 6

e' e e e o o o e e e e e e e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: REACTOR COOLANT FLOH DIAGRAM: 1-5128A-37 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORHED MODE 1-NSO-062 3 GL 1 SO H/6 C 0/C 2 A 8 EF-1 EF"2 C NOp CSJ 6 EF-5 EF-5 NOp CSJ 6 EF-7 EF-8 C NOp CSJ 6 ET-XXX ET-XXX C NOp CSJ 6 1-NSO-063 . 3 GL 1 -

SO H/6 C 0/C 2 A 8 EF-1 EF-2 C NOp CSJ 6 EF-5 EF-5 NOp CSJ 6 EF-7 EF-8 C NOp CSJ 6

.ET-XXX ET-XXX C NOp CSJ 6 I-NSO-06% 3 GL 1 SO H/6 C 0/C 2 A 8 EF-1 EF-2 C NOp CSJ 6 EF-5 EF-5 NOp CSJ 6 EF-7 EF-8 C NOp CSJ 6 ET-XXX ET-XXX C NOp CSJ 6 1"PN-275 3 CK 3 SA 8/9 0/C C 2 A AC CF-1 CF-2 R YESp NOTE 4 SLT-1 SLT-2 R YESp NOTE 2 1-RCR-100 3 GL 0.375 A 8/7 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2 1-RCR-101 3 GL 0.375 A 8/7 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2

0 0 0 0 0 0 0 0 0 0 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SlRCKRY SHEET - UNIT 1 SYSTEH NAHE: REACTOR COOLANT FLON DIAGRAH: 1-5128A-37 VALVE I VALVE POSITION ASHE SECTION XZ NUiiBER REV TYPE SIZE ACT F ~ D t POHER SAFETY )CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDt OPER FUNCT ICL REQUIRED PERFORHED HiODE 1-SI-189 3 CK SA D/7 C 0/C 2 A AC CF-1 CF-2 R YES> NOTE 7 SLT-1 SLT-2 R YESp NOTE 2 1-SV-45A REL 6 SA K/6 C 1 A C TF-1 TF-1 1-SV-45B 3 REL 6 SA J/6 C 1 A C TF-1 TF-1 R NO I-SV-45C 3 REL 6 SA N/6 C 0 = 1 A C TF-1 TF-1 1"SV-50 3 REL 2 SA G/3 C 3 A C TF-1 TF-1 R'O

e e e e e e e e e e e e e o e e e e e e e e e e o e 0,

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SR$ 1ARY SHEET - UNIT 1 SYSTEM NAME: CVCS - LETDOWN & CHARGING FLON DIAGRAM: 1-5129-31 VALVE I VALVE POSITION I ASME SECTION XZ NPiSER, REV TYPE SIZE ACT F.D. I POHER SAFETY )CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-CS-292 3 CK 2 SA H/6 C 0/C 2 A C CF-1 CF-2 C YES> NOTE 1 1-CS-297-E 3 CK 2 SA H/7 0/C 0 2 A C CF1 CF-1 1-CS-297"N 3 CK 2 SA F/7 0/C 0 2 A C CF1 CF-1 P NO 1-CS"299-E 3 CK 4 SA H/7 0 0/C 2 A AC CF-1 CF-3 P 'ES> NOTE 2 SLT-1 SLT-1 R NO> NOTE 11.

I"CS-299-N 3 4 F/7 0 0/C 2 CF-1 CF-3 CK SA A AC SLT-1 SLT-1 P

R NO> NOTE ll YES> NOTE 2 1-CS-321 3 CK 3 SA E/3 0 C/0 2 A AC CF-1 CF-2 R YES> NOTE 4 SLT-1 SLT"2 R YES> NOTE 3 1-CS-328"Ll "3 CK 3 SA B/2 0/C 0 1 A C CF1 CF-1 P NO> NOTE 5 1-CS-328-L4 3 CK 3 SA B/3 0/C 0 1 A C CF-1 CF-1 P NO> NOTE 5 1-CS-329-Ll 3 CK 3 SA B/2 0/C 0 1 A C CF1 CF-1 P NO> NOTE 5 1-CS-329-L4 3 CK 3 SA B/3 0/C 0 1 A C CF1 CF-1 P NO> NOTE 5 1-IMO-36D 3 GA MO H/6 C 2 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

0 0 e e e o e e e e e o e e e o e

DONALD C. COOK NUCLEAR PLANT RW DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: CVCS - LETDONN 8 CHARGING FLON DIAGRAM: 1-5129-31 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED MODE 1-IHO-910 3 GA 8 HO L/5 C 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-IHO-911 3 GA 8 HO L/6 C 2 A B EF-1 EF-1 P NO fF-5 EF-5 NO ET-XXX ET-XXX P NO 1-QCR-300 3 GL 2 A E/1 0 2 A A EF-1 EF-2 C NOp CSJ 6 EF-5 EF-5 NO-EF-7 EF C NOp CSJ 6 ET-XXX ET-XXX P NOp CSJ 6 SLT-1 SLT-2 R YESp NOTE 3 1-QCR-301 3 GL 2 A E/1 0 2 A A EF-1 EF-2 C NOp CSJ 6 EF"5 EF-5 NO EF-7 EF-8 C NOp CSJ 6 ET-XXX ET-XXX P NOp CSJ 6 SLT-1 SLT-2 R YESp NOTE 3 1-QHO-200 3 GA 3 HO J/3 0 0/C 2 A B EF-1 EF-2 C NOp CSJ 7 EF-5 EF-5 NO ET-XXX ET-XXX P NOp CSJ 7 1-QHO-201 3 GA HO J/3 0 0/C 2 A B EF-1 EF"2 C NOp CSJ 7 EF-5 EF-5 NO ET-XXX ET-XXX P NOpCSJ 7 I-QMO-225 3 GA 2 MO J/7 0 0/C 2 A B EF-1 EF-1 P NO E F-5 EF-5 NO ET-XXX ET-XXX P NO

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUGARY SHEET UNIT 1 SYSTEH HAHE: CVCS - LETDOHH 2 CHARGING FLON DIAGRAN: 1-5129-31 VALVE I VALVE POSITION I ASHE SECTION XI HmSER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIN TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORNED NODE I-QNO-226 3 GA 2 NO G/7 0 0/C 2 A 8 EF-1 EF-1 P HO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1"QRV"200 3 GL 3 A K/3 0 2 A B EF-1 EF-3 P HOp NOTE 8 EF-5 EF-5 YES> NOTE 8 EF-7 NOTE 8 YES'OTE 8 ET-XXX NOTE 8 YES'OTE 8 1-QRV-251 3 GL 3 A N/5 0 0 2 A B EF1 EF-3 P HO> NOTE 9 EF-5 EF-5 YES> NOTE 9 EF"7 EF"8 YES> NOTE 9 ET-XXX HOTE 9 YES> NOTE 9 1-QRV-61 3 GL 3 A C/2 C 2 A B EF1 EF-1 P HO EF-5 EF-5 HO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-QRV-62 3 GL 3 A C/2 0/C 0 2 A B EF1 EF-1 EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET"XXX P NO 1-SI-185 3 CK 8 SA K/5 C 2 A C CF1 CF-2 R YES> NOTE 10 1-SV-51 3 REL 2 SA E/2 C 2 A C TF-1 TF-1

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SONQRY SHEET - UNIT 1 SYSTEM NAME: CVCS LETDOHN S CHARGING FLOH DIAGRAM: 1"5129-31 VALVE VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. J PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-SV"52 3 REL 2 SA K/1 C 2 A C TF-1 TF-1 0 1-SV-55 3 REL 0.75 SA K/7 C 0 2 A C TF-1 TF-1 1-SV-56 REL 0.75 SA L/6 C 0 2 A C TF1 TF-1 R NO

0 e o e e o e e e e e e o e e o 0 DONALD C. COOK NUCLEAR PLANT RUH DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUESLRY SHEET '- UNIT 1 SYSTEM NAME: CVCS - LETDOHN A CHARGING FLOH DIAGRAM: 1-5129A-19 VALVE I VALVE POSITION I ASME SECTION XI NUiiBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FLRiCT ICL REQUIRED PERFORMED MODE 1-QCM-250 3 GA MO C/8 0 2 A A EF1 EF-2 C HOp CSJ 1 EF-5 EF-5 HO ET-XXX ET-XXX C NO> CSJ 1 SLT-1 SLT-2 R YES> NOTE 2 1-QCM"350 3 GA MO C/8 0 C,2 A A EF1 EF"5 EF-2 EF-5 C NO> CSJ HO 1

ET-XXX ET-XXX C HOp CSJ 1 SLT-1 SLT-2 R YES'OTE 2 1-QMO-451 3 GA MO J/5 0 2 A B EF1 EF-2 C NO@ CSJ 3 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 3 1-QMO-452 3 GA MO - J/5 0 2 A B EF1 EF-2 C NOp CSJ 3 EF-5 EF-5 NO ET-XXX ET-XXX C HOp CSJ 3 1-QRV-400 3 GL 2 A K/4 C 2 A B EF1 EF-1 P NO EF-5 EF-5 HO EF-7 EF"7 P NO ET-XXX ET-XXX P NO 1-SV-53 3 REL SA H/2 C 2 A C TF-1 TF-1 R NO 1-SV-54 3 REL 2 SA E/4 C 2 A C TF-1 TF-1

e e e e e e 0 e e e 0

I DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUMHARY SHEET - UNIT 1 SYSTEM NAME: COMPONENT COOLING FLON DIAGRAM: 1-5135-29 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F.D. I PONER SAFETY lCD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED .MODE 1-CCH-451 3 BF 8 HO E/4 0 2 A A EF-1 EF-2 C NO@ CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NOi CSJ 1 SLT-1 SLT-2 R YES'OTE 2 1-CCH-452 3 BF 8 HO E/5 0 2 A A EF-1 EF-2 C NOp CSJ 1 E F-5 EF-5 NO ET-XXX ET-XXX C NO) CSJ 1 SLT-1 SLT-2 R YES, NOTE 2 1-CCH-453 3 GL 4 HO E/4 0 2 A A EF-1 EF-2 C NOi CSJ 1 EF-5 EF-5 NO ET"XXX ET-XXX C NOi CSJ 1 SLT-1, SLT-2 R YES> NOTE 2 1-CCM-454 3 GL 4 HO E/5 0 2 A A EF-1 EF-2 C NO> CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NOi CSJ 1 SLT-1 SLT-2 R YES'OTE 2 1"CCH-458 3 BF 8 HO A/2 0 2 A A EF-1 EF-2 C NO> CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 1 SLT-1 SLT-2 R YES> NOTE 2 1-CCH-459 3 BF 8 HO B/2 0 2 A A EF-1 EF-2 C NO> CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 1 SLT-1 SLT-2 R YES> NOTE 2

0 0 0 0 0 '

0 ~ 0 0 0 ~ 0 0 ~ ~ ~ 0 ~ ~ 0 0 ~

III DONALD C. COOK NUCLEAR PLAtG'ECOND RUH DATE AHD TIME: 15FEB90:16:03 TEN YEAR INTERVAL VAlVE SUMtARY SHEET - UNIT 1 SYSTEM NAME: COMPONENT COOLING FLOH DIAGRAM: 1-5135-29 VALVE I VALVE POSITION ASME SECTION XZ NUMBER REV TYPE SIZE ACT F ~ D. I POHER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUHCT ICL REQUIRED PERFORMED MODE 1-CCR-455 3 GL 2 A B/3 0 2 A A EF-1 EF-2 C NOp CSJ 3 EF-5 EF-5 NO EF-7 EF-8 C NOp CSJ 3 ET-XXX ET-XXX C NOp CSJ 3 SLT-1 SLT-2 R YESp NOTE 2 1-CCR-456 3 GL 2 A D/4 0 2 A A EF-1 EF-2 C NOp CSJ 3 EF-5 EF-5 NO EF-7 EF-8 C NOp CSJ 3 ET-XXX ET"XXX C NOp CSJ 3 SLT-1 SLT-2 R YESp NOTE 2 1-CCR"457 3 GL 2 A D/4 0 ,C 2 A A EF"1 EF-2 C NOp CSJ 3 EF-5 E F-5 HO EF-7 EF-8 C NOp CSJ 3 ET-XXX ET-XXX C NOp CSJ 3 SLT"1 SLT-2 R YES> HOTE 2 1-CCR-460 3 GL 3 A C/4 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2 1-CCRM62 3 GL 3 A A/4 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET"XXX P HO SLT-1 SLT-2 R YESp NOTE 2

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SALARY SHEET UNIT 1 SYSTEH NAHE: COHPONENT COOLING FLOW DIAGRAH: 1-5135-29 VALVE I VALVE POSITION I ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I POWER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORHED HODE 1-CCW-135 3 CK 2. 5 SA B/3 0 C 2 A AC CF 1 SLT-1 -

CF-2 R 'ES> NOTE 4 SLT-2 R YES> NOTE 2 1-CRV-445 3 BL 6 A L/5 0 C/0 3 A B EF-1 EF-1 P NO EF-5 EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX -P NO 1-CRV-470 3 GL 6 A G/1 0 C 3 A B EF1 EF-1 P NO> NOTE 5 EF-5 NOTE 5 YES> NOTE 5 EF-7 NOTE 5 YES> NOTE 5 ET-XXX NOTE 5 YES NOTE 5 1-CRV"485 3 BF 10 A B/7 0 C 3 A B EF1 EF-1 P .

NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-SV"122-37 3 REL 1 SA D/3 C 3 A C TF1 TF-1 1-SV-62"1 3 REL 1 SA D/3 C 3 A C TF-1 TF-1 R NO 1-SV-62-2 3 REL 1 SA D/3 C 3 A C TF-I TF-1 R NO 0 1-SV-62-3 TF-1 TF-1 3 REL 1 SA D/3 C 3 A C I-SV-62M 3 REL 1 SA D/3 C 3 A C TF-1 TF-1

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAME! COMPONENT COOLING FLOH DIAGRAM: 1-5135-29 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SI2E ACT F.D. I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1"SV-63 3 REL 1 SA E/3 C 3 A C TF-1 TF-1 R NO I-SV-64 3 REL 1 SA C/3 ' A C TF-1 TF-1 R NO 1-SV-65 3 REL 1 SA H/1 C 0 3 A C TF-1 TF-1 1-SV-68 3 REL 1 SA J/2 C A C TF-1 TF-1 1-SV-71 REL 1 SA L/3 C 3 A C TF1 TF-1 R NO

e e e e e e e e e e 0 e o e e DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHMARY SHEET UNIT 1 SYSTEM NAME: COMPONENT COOLZNG FLON DIAGRAM: 1-5135A-30 VALVE I VALVE POSITION I ASHE SECTION XZ NUMBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRZH TEST TEST TEST RELIEF REQUEST< S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED MODE 1"CCN"176-E 3 CK 16 SA L/4 C/0 0 3 A C CF-1 CF-1

~ 1-CCN-176-N 3 CK 16 SA K/4 C/0 0 3 A C CF1 CF-1 1-CHO-410 3 BF 16 HO M/4 C/0 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO I-CHO&ll 3 BF 18 HO H/5 0 3 A 8 EF1 EF-1 EF-5 EF-5 NO ET-XXX ET-XXX P NOp NOTE 2 1-CMO-412 3 BF 16 HO L/3 0 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO> NOTE 2, 1-CHO-413 3 BF 18 HO L/5 0 3 A 8 EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO> NOTE 2 1-CHO-414 3 BF 16 HO K/3 0 3 A 8 EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO> NOTE 2 1-CHO-415 3 BF 16 - HO M/5 0 3 A 8 EF1 EF-1 EF-5 EF-5 NO ET-XXX ET-XXX P NO> NOTE 2

e e e o 0 o e e e e e o e o e e e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEM NAME: COMPONENT COOLING FLON DIAGRAM: 1-5135A-30 VALVE I VALVE POSITION I ASME SECTION XI NUHBER REV TYPE SIZE ACT F.D. t POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-CHO-416 3 BF 16 HO G/5 0 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO> NOTE 2 1-CHO "419 3 BF 14 HO E/5 C 3 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-CHO-420 3 BF 16 HO K/4 C/0 0 3 A B EF 1 EF 1 EF-5 EF-5 ET-XXX ET-XXX I-CHO&29 3 BF 14 HO E/5 C 3 A B EF1 EF-1 EF"5 E F-5 NO ET-XXX ET-XXX P NO 1-CRV-412 3 GL A K/1 0 3 A B EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-SV-60 3 REL 3 SA L/1 C 3 A C TF-1 "

TF-1 1-SV-72 3 REL 1 SA E/5 C 3 A C TF1 TF-1 R NO 12-CCN-170 3 CK 16 SA H/4 . C 3 A C CF1 CF-4 NOp NOTE 1

0 e e o o e e e e o o e e e o o 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB9D:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: COHPONENT COOLING FLON DIAGRAH: 1-5135B-14 VALVE I VALVE POSITION I ASME SECTION XI NlklBER REV TYPE SIZE ACT ~ F.D. I POMER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-CCH-430 3 GL 1.5 HO D/6 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1"CCH-431 3 GL 1.5 HO D/6 C 0/C 2 A A EF-1 EF-1 P NO E F-5 EF-5 = NO ET-XXX ET-XXX P NO SLT"1 ~ SLT-2 R YESp NOTE 1 1-CCH-432 3 GL 1.5 MO D/6 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-CCH-433 3 GL 1.5 HO D/6 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO, SLT-1 SLT-2 R YESp NOTE 1 1-CCR-440 3 GL 1.5 A D/6 0 2 A A EF-1 EF-1 P NO EF-5 EF"5 NO, EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-CCR-441 3 GL 1.5 A D/6 0 2 A A EF-1 EF-1 EF-5 EF-5 NO EF"7 EF"7 P 'NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SSARY SHEET - UNIT 1 SYSTEM NAME: COMPONENT COOLING F LOH DIAGRAM: 1-51358-14 VALVE I VALVE POSITION ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE I-CCH-243"25 3 CK 1 SA C/5 0 2 A AC CF-1 CF-2 R YES'OTE 2 SLT-1 SLT-2 R YES> NOTE 1 1-CCH-243-72 3 CK 1 SA C/5 0 2 A AC CF-1 CF-2 R YES> NOTE 2 SLT-1 SLT-2 R YES'OTE 1 1-CCH"244-25 3 CK 1 SA C/6 0 2 A AC CF-1 CF-2 R YESp NOTE 2 SLT-1 SLT-2 R YES> NOTE 1 I-CCH"244-72 3 CK 1 SA C/6 0 2 A AC CF-1 CF-2 R YES> NOTE 2 SLT-1 SLT-2 R YES> NOTE 1 1-SV-122-258 3 REL 1.5 SA 8/6 C 3 A C TF-1 TF-1 R NO 1-SV-122-728 3 REL 1.5 SA 8/6 C 3 A C TF-1 TF-1 R NO

-0 0 0 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEM NAME: NUCLEAR SAMPLING FLON DIAGRAM: I-5141-29 VALVE I VALVE POSITION I ASHE SECTION XI NUiiBER REV TYPE SIZE ACT F.D. ) PONER SAFETY !CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORMED HODE 1-ICR-5 3 GL 0.5 A C/5 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO

= EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-ICR-6 GL 0.5 A D/5 0 2 A A EF-1 EF-1 E F-5 EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-NCR-105 3 GL 0.5 A C/7 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT"2 R YES> NOTE 1 1-NCR-106 3 GL 0.5 A C/7 0 2'AA EF-1 EF-1 P NO EF-5 EF-7 ET-XXX EF-5 EF-7 ET-XXX P'O P

NO NO SLT-1 SLT-2 R YES> NOTE 1 1-NCR"107 3 GL 0.5 A D/6 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

o o e e o o e e e e e e o e e o 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SECSLRY SHEET - UNIT 1 SYSTEH NAHE: NUCLEAR SAHPLING FLOW DZAGRAH: 1-5141-29 VALVE I VALVE POSITION I ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I POWER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED 'ODE 1-NCR-108 3 GL 0.5 A D/6 0 C 2 A A EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-NCR-109 3 GL 0.5 A D/6 0 C 2 A A EF1 EF-1 P NO EF-5 EF"5 NO EF-7 EF"7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-NCR-110 3 GL 0.5 A D/6 0 C 2 A A EF1 EF-1 EF-5 EF-5 EF-7 EF-7 P NO ET"XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

l

)~O 0 0 ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ 0 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: NUCLEAR SAHPLING FLON DIAGRAH: 1-5141A"32 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F ~ DE I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1-DCR-301 3 GL 0.5 A B/2 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-DCR-302 3 GL 0.5 A B/3 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET"XXX P NO 1-DCR-303 3 GL 0.5 A B/3 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF"7 EF-7 P NO ET-XXX ET-XXX P NO 1-DCR-304 3 GL 0.5 A B/3 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-HCR"251 3 GL 0.5 A B/2 0 C 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO I-HCR-252 3 GL 0.5 A B/2 0 C 2 A B EF-1 EF-1 P NO E F-5 E F-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO

,0 e e' e e e e e e o e o e e 0 0 0 0 0 0 0 0 0

~ ~ DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: NUCLEAR SAHPLING FLON DIAGRAM: 1-5141A-32 VALVE I VALVE POSITION I ASHE SECTION XI NJiiBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE'-MCR-253 3 GL 0.5 A B/1 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET"XXX ET-XXX P NO 1-MCR-254 3 GL 0.5 A B/1 0 2 A B EF1 EF"1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO

0 0

e e e o e e e .e o e e e e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SRDlARY SHEET. - Wi IT 1 SYSTEH NAHE: PAS CONTAINMENT HYDROGEN FLON DIAGRAH: 1-51410-10 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST 'EST TEST RELIEF REQUEST(S)

/ TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-ECR-10 GL 0.5 A C/8 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-11 3 GL 0.5 A A/2 C 0/C 2 A A EF-1 EF-1 EF-5 EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-12 3 GL 0.5 A A/2 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 EF-7 ET-XXX EF"7 ET-XXX P 'O NO P NO SLT-1 SLT-2 R YES'OTE 1 1-ECR-13 3 GL 0.5 A A/1 C 0/C 2 A A EF-1 EF-1 P NO EF"5 EF-5 NO EF-7 - EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-14 3 GL 0.5 A A/3 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1

DONALD C. COOK NUCLEAR PLANT RUH DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEN NAME: PAS CONTAIKIEHT HYDROGEN FLON DIAGRAN: 1-5101D-10 VALVE I VALVE POSITION ASNE SECTIOH XZ HOSER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIN TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORNED NODE 1-ECR-15 3 GL 0.5 A A/1 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES'OTE 1 1-ECR-16 GL 0.5 A A/3 C 0/C 2 A A EF-1 EF-1 EF-5 EF-5 EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1"ECR-17 3 GL 0.5 A A/3 C 0/C 2 A A EF-1 EF-1 P HO EF-5 EF-5 HO EF-7 EF-7 "

P NO ET"XXX ET"XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-ECR"18 3 GL 0.5 A A/4 C 0/C 2 A A EF"1 EF-1 P NO EF-5 EF-5 HO EF-7 EF-7 P NO ET"XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-19 3 GL 0.5 A A/0 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT SLT-2 R YES> NOTE 1

g ~

DONALD C. COOK HUCLEAR PLANT RUN DATE AND TIME: 15FEB90:l6:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: PAS CONTAIHMENT HYDROGEN FLOW DIAGRAM: I"51410-10 VALVE I VALVE POSITION ASME SECTIOH XI NUMBER REV TYPE SIZE ACT F.D. I POWER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUHCT ICL REQUIRED PERFORMED MODE 1-ECR-20 GL 0.5 A C/8 C 0/C 2 A A EF-) EF-1 P NO EF-5 EF-5 'O EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-21 3 GL 0.5 A 8/2 C 0/C 2 A A EF-1 - - EF-1 P NO EF-5 E F-5 HO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT"2 R YES'OTE 1 1-ECR-22 3 GL 0.5 A 8/2 C 0/C 2 A A EF-1 EF"1 EF-5 EF EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT"1 SLT"2 R YESp NOTE 1 1-ECR"23 3 GL 0,5 A 8/1 C 0/C 2 A A EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-24 3 GL 0.5 A 8/3 C 0/C 2 A A EF-1 EF-1 P NO EF-5 - - EF-5 HO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

1 1

0 0 0 0 0 0 0 0 0 0 0 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: PAS CONTAINMENT HYDROGEN FLOH DIAGRAM: 1-5141D-10 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F D t POHER SAFETY lCD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUHCT ICL REQUIRED PERFORMiED MODE 1-ECR"25 GL 0.5 A B/1 C 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R - YES> NOTE 1 1-ECR-26 GL 0.5 A B/3 C 0/C 2 A A EF-1 EF>>1 P NO EF-5 E F-5 NO EF" 7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT"2 R YES> NOTE 1 1-ECR-27 3 GL 0.5 A B/3 C 0/C 2 A A EF-1 Ef-1 P NO EF-5 EF-5 NO EF-7 EF-7 P ,NO ET-XXX ET-XXX P HO SLT-1 SLT"2 R YES> HOTE 1 I-ECR-28 3 GL 0.5 A B/4 C 0/C 2 A A EF"1 EF-1 P HO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-ECR-29 3 GL 0.5 A B/4 C 0/C 2 A A EF-1 EF-1 P HO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1

, e e o o e o e e o o e e o e o DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND.TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: PAS CONTAINHENT HYDROGEN FLON DIAGRAH: I-5141D-10 VALVE VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F AD. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FIJiiCT ICL REQUIRED PERFORHED NODE 1-NS-283 3 CK 0.5 SA C/8 C 0/C 2 A AC CF-1 CF-2 R YES> NOTE 2 SLT-1 SLT"2 R YES> NOTE 1

0 0 e e e e e o e e e o e e o e o 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET UNIT 1 SYSTEM NAME: EMERGENCY CORE COOLING - SIS FLON DIAGRAM: 1-5142-25 VALVE I VALVE POSITION I ASHE SECTION XI NUYBER REV TYPE SIZE ACT F D t PONER SAFETY lCD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST[ S)

TYPE COORD I OPER FUhCT ICL REQUIRED PERFORHED MODE 1-ICH-25D 3 GA HO H/2 C 0/C 2 A A EF-1 EF"2 C NO> CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NOp CSJ 1 SLT-1 SLT-2 R YES> NOTE 2 1-ZCM-251 3 GA 4 MO H/3 C 0/C 2 A A EF-1 EF-2 C NOi CSJ 1 EF-5 EF-5 NO ET-XXX ET-XXX C NO) CSJ 1 SLT-1 SLT"2 R YES> NOTE 2 1-ICH-26D 3 GA HO C/9 0 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 2 1-ICH-265 . 3 GA HO C/8 0 0/C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 2 1-ZHO"255 3 GA HO J/7 C 2 A B EF-1 EF-1 P NO EF"5 EF"5 NO ET-XXX ET-XXX P NO 1-ZHO-256 3 GA HO J/6 C 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-ZHO-261 3 GA 8 HO H/8 0 0/C 2 A B EF-1 EF-2 C NO> CSJ 3 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 3

0 e e e e e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET UNIT 1 SYSTEM NAME: EMERGENCY CORE COOLING - SIS FLON DIAGRAM: 1-5142-25 VALVE I VALVE POSITION I ASHE SECTION XI Qk'tBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-IHO-262 3 GL 2 HO L/8 0 0/C 2 A B EF-1 EF-2 C NO) CSJ 4 EF-5 EF"5 NO ET-XXX ET-XXX C NO> CSJ 4 1-IHO-263 3 GL 2 HO L/8 0 0/C 2 A B EF-1 EF-2 C NO> CSJ 4 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 4 I"IHO-270 3 GA 4 HO E/9 0 0/C 2 A B EF"1 EF-1 EF-5 EF-5 ET-XXX ET-XXX P NO 1-IHO-275 3 GA HO E/8 0 0/C 2 A B EF-1 EF-1 EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-IHO-361 3 GA HO G/9 C C/0 2 A B EF-1 EF-1 P NO EF-5 EF"5 NO ET-XXX ET-XXX P NO 1-IHO-362 3 GA HO G/9 C C/0 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-IRV-251 3 GL 1 A H/5 0 2 A B EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO

0 0 0 0 0 0 0 0 0 0 0 0 0 0

D ONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN,YEAR INTERVAL VALVE SUHHARY SHEET UNIT 1 SYSTEH NAME: EHERGENCY CORE COOLING - SIS FLOH DIAGRAH: 1-5142-25 VALVE I VALVE POSITION I ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRZH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT tCL REQUIRED PERFORHED NODE 1-IRV"252 3 GL 1 A J/5 0 3 A B EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-ZRV-255 3 GL 1 A H/6 0 2 A B EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-SI-101 3 CK 8 SA H/8 - C 2 A C CF-1 CF-3 R YESp'OTE 5 1-SI-104-N 3 CK 0. 75 SA E/9 C 2 A C CF1 CF-1 1-SI-104 "S CK 0.75 SA J/9 C 2 A 'C CF1 CF-1 P NO 1-SI-110-N 3 CK SA E/9 C 2 A C CF1 CF-2 R YES> NOTE 5 1-SI-110-S 3 CK SA H/9 C 2 A C CF1 CF-2 R YES> NOTE 5 1-SI-126 3 CK 1 SA H/6 0 2 A C CF-1 CF-1 1-SI" 142-L1 3 CK 1.5 SA C/1 C 1 A C CF1 CF-2 R YES) NOTE 6 1-SZ-142-L2 3 CK 1.5 SA C/2 C 1 A C CF1 CF-2 R YES> NOTE 6

0 Ill DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHEl 15FEB90l16l03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: EHERGENCY CORE COOLING - SIS FLON DIAGRAH: 1-5142-25 VALVE I VALVE POSITION l ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST 'ELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1-SI-142-L3 3 CK 1.5 SA C/2 C 1 A C CF1 CF-2 R YES'OTE 6 1-SI-142-L4 3 CK 1.5 SA C/1 C 1 A C CF1 CF-2 R YES> NOTE 6 1-SV-96 REL 0.75 SA J/8 C 2 A C TF1 TF-1 R NO 1-SV-97 3 REL 0.75 SA J/4 C 2 A C TF1 TF-1 R NO 1-SV-98-N 3 REL 0.75 SA C/9 C 2 A C TF1 TF-1 R NO 1-SV-98-S 3 REL 0.75 SA E/8 C 0 2 A C TF1 TF-1

e DONALD C. COOK NUCLEAR PLANT SECOND TEN YEAR INTERVAL RUN DATE AND TIME: 15FEB90:16:03 VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: EMERGENCY CORE COOLING - RHR FLON DIAGRAM: 1-5143-36 VALVE I VALVE POSITION ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-GCR-314 3 GL 1 A G/2, 0 2 A A EF-1 EF-1 P NO Ef-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 14 1-ICH-129 3 GA 14 HO E/8 C 1 P A EF-1 EF-2 C NO> CSJ 2 EF-5 EF-5 NO ET-XXX ET"XXX C NO> CSJ 2 SLT-1 SLT-1 R NO>-NOTE 1 1-ZCH-305 3 GA 18 HO D/9 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO .

ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 14 1-ZCM-306 3 GA 18 HO D/9 C A A EF-1 EF-1 ET-XXX SLT-1

- EF-5 'F-5 ET-XXX SLT-2 P

R

'O NO YES> NOTE 14 1-IHO-128 3 GA 14 HO B/8 C 1 P B EF-1 EF-2 C NOp CSJ 2 EF-5 EF-5 NO ET-XXX ET-XXX C NOp CSJ 2 1-ZHO-310 3 GA 14 HO H/9 0 2 A B EF-1 EF-1 P NO EF-5 EF-5 . NO ET-XXX ET-XXX P NO> NOTE 3

7 0

0 o e e e e o e e e o o e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: EHERGENCY CORE COOLING - RHR FLOH DIAGRAM: 1-5143-36 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY I CD A/P CAT PRIM TEST TEST =-

TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORMED MODE 1-IHO-312 3 GL 2 HO J/5 0 0/C 2 A B EF-1 EF"1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-ZMiO"314 3 GA 8 HO K/6 0 2 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NOy NOTE 3 1-IMO-315 3 GA 8 HO C/7 C C/0 1 A B EF-1 EF-2 C NO> CSJ 4 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 4 1-IHO-316 3 GA 8 HO C/7 0 0/C 2 A B EF-1 EF-2 C NOp CSJ 4 EF-5 E F-5 NO ET-XXX ET-XXX C NOp CSJ 4 1-ZMiO-320 3 GA 14 HO L/9 0 2 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO> NOTE 3 1-IHO-322 3 GL 2 MO H/5 0 0/C 2 A B EF-1 EF-1 P NO EF-5 EF"5 NO ET-XXX ET-XXX P NO..

1-IHO-324 3 GA 8 HO H/6 0 2 A B EF1 EF1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO NOTE 3

0 e e e e e e o e e e e e e a

0 0 0 0 0 0 0

DOilALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: EMERGENCY CORE COOLING - RHR FLON DIAGRAM: 1-5143-36 VALVE I VALVE POSITION I ASME SECTZOiN XZ NUHBER REV TYPE SIZE ACT F ~ D ~ I POHER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-IMO-325 3 GA 8 MO C/7 C C/0 1 A 8 EF-1, EF-2 C NO> CSJ 4 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 4 1-IHO-326 3 GA 8 HO C/7 0 0/C 2 A B EF-1 EF-2 C NO> CSJ 4 EF-5 EF-5 NO ET-XXX ET-XXX C NO> CSJ 4 1-IMO-330 3 GA 8 HO G/4 C 2 A B EF1 EF-1 EF-5 EF-5 ET-XXX ET-XXX P NO 1-IHO-331 3 GA 8 MO L/5 C 2 A B EF-1 EF-1 EF-5 EF-5 NO ET-XXX ET"XXX NO I-IMO-340 3 GA 8 HO H/5 C 2 A B EF1 EF-2 C YES> NOTE 15 EF-5 EF-5 NO ET-XXX ET-XXX C YES> NOTE 15 1-IHO-350 GA 8 HO L/5 C 0 2 A B EF"1 EF-2 C YESi NOTE 15 EF"5 EF-5 NO ET-XXX ET-XXX C -'ES> NOTE 15 1-N-102 3 CK 1 -'A F/5 0/C C 2 A AC CF-1 CF"2 R YES> NOTE 13 SLT-1 SLT-2 R YES> NOTE 14 1-RH-108E 3 CK 8 SA K/9 C 0 2 A C CF1 CF-3 C NOi CSJ 7

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SALARY SHEET - UNIT 1 SYSTEH NAHE: EHERGENCY CORE COOLING - RHR FLON DIAGRAH: 1-5103-36 VALVE I VALVE POSITIOiN I ASHE SECTION Xl NUHBER REV TYPE SI2E ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED HODE 1-RH" 108H 3 CK 8 SA N/9 C 2 A C CF-1 CF-3 C NOp CSJ 7 1-RH-133 3 CK 8 SA C/5 C 1 P AC SLT-1 SLT-1 R NOp NOTE 1 I-RH-134 3 CK 8 SA C/5 C 1 P AC SLT-1 SLT-1 R NOp NOTE 1 I-SI-148 3 ~

CK 12 SA G/7 C 2 A C CF1 CF-3 YESp NOTE 8 I-SI-151-E 3 CK 8 SA D/7 C 2 A AC CF-1 CF-2 C NOp CSJ 9 SLT-1 SLT-1 R NOp N)TE 1 1-SI-151-H 3 CK 8 SA 0/7 C 2 -A AC CF-1 CF-2 C NOp CSJ 9 SLT-1 SLT-1 R NOp NOTE 1 1-SI-152-N 3 CK SA D/8 C 2 A AC CF1 CF-2 R YESp NOTE 10 SLT-1 SLT-1 R NOp NOTE 1 1-SI-152-S 3 CK SA D/7 C 2 A AC CF-1 CF-2 R YES NOTE 10 SLT-1 SLT-1 R NOp NOTE 1 1-SI-158-Ll 3 CK 6 SA 8/8 C 1, A AC CF-1 SLT-1 CF-2 SLT-1 R YESp NOTE NOp NOTE 1 ll 1-SI-158-L2 3 CK 6 SA 8/7 C 1 A AC CF-1 SLT-1 CF-2 SLT-1 R YESp NOTE NOp NOTE 1 ll

0 e e o o e e e e e e e e e e o 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FE890:16:03 SECOND TEN YEAR INTERVAL VALVE SUGARY SHEET - UNIT 1 SYSTEN NANE: EHERGEHCY CORE COOLING - RHR FLON DIAGRAH: I-5143-36 VALVE I VALVE POSITION I ASHE SECTION XZ QNBER REV TYPE SIZE ACT F AD I POHER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUHCT ICL REQUIRED PERFORHED NODE 1-SI-158-L3 3 CK 6 SA 8/7 C 1 A AC CF1 CF-2 YESp NOTE 11 0 SLT-1 SLT-1 R HOp NOTE 1 I-SI-158-L4 3 CK 6 SA 8/7 C 1 A AC CF 1 CF-2 YESp NOTE 11 SLT-1 SLT-1 R HOp NOTE 1 1-SI-161-L1 3 CK 6 SA 8/6 C 1 A AC CF-1 CF-2 YESp HOTE 6 i, SLT-1 SLT-1 R NOp NOTE. 1 1-SI-161-L2 3 CK 6 SA 8/5 C 0 1 A AC CF-1 CF"2 YESp NOTE 6 SLT-1 SLT-1 R NOp NOTE 1 1-SI-161-L3 3 CK 6 SA 8/5 C 1 A AC CF-1 CF-2 YESp NOTE 6 SLT-1 SLT-1 R NOp NOTE 1 1 "SI-161" LO 3 CK 6 SA 8/6 1 A CF1 CF-2 YESp NOTE 6 C AC SLT-1 SLT-1 R'Op NOTE 1 1-SI-166-1 3 CK 10 SA C/0 C 1 A AC CF 1 CF-2 R YESp NOTE 5 SLT"1 SLT-1 R NOp NOTE 1 1-SI-166"2 3 CK 10 SA C/0 C 0 1 A AC CF-1 CF-2 R YESp NOTE 5 SLT"1 SLT-1 R NOp NOTE 1 1-SI-166-3 3 CK 10 SA C/0 C 0 1 A AC CF-1 CF-2 R YESp NOTE 5 SLT"1 SLT-1 R HOp N)TE 1

o o e o e o e e o o e o e e o 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET " UNIT 1 SYSTEM NAME: EHERGENCY CORE COOLING - RHR FLON DIAGRAM: 1>>5143-36 VALVE I VALVE POSITION I ASME SECTION Xl NUMBER REV TYPE SI2E ACT F.DE I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-SI-166-4 3 CK 10 SA C/4 C 1 A AC CF1 CF-2 R YESp NOTE 5 SLT-1 SLT-1 R NOp NOTE 1 1-SI-170-Ll 3 CK 10 SA A/4 C 1 A AC CF-1 CF-3 R YESp NOTE 12 SLT-1 SLT-1 R NOp NOTE 1 1-SI-170-L2 3 CK 10 SA A/5 C 0/C 1 A AC CF-1 CF-3 R YESp NOTE 12 SLT-1 SLT-1 R NOp NOTE 1 1-SI-170-L3 3 CK 10 SA A/5 C 0/C 1 A AC CF-1 CF-3 R YESp NOTE 12 SLT-1 SLT"1 R NOp NOTE 1 1-SI-170-L4 3 CK 10 SA A/4 C 1 A AC CF1 CF-3 R YESp NOTE 12 SLT-1 SLT-1 R NOp NOTE 1 1-SI-171 3 GL 0. 75 H H/6 C 2 P A SLT-1 SLT-2 R YESp NOTE 14 1-SI-172 3 GL 0. 75 H H/6 C 2 P A SLT1 SLT-2 R YESp NOTE 14 1-SZ"194 3 GL 0.75 H G/6 C 2 P A SLT 1 SLT-2 R YESp NOTE 14 1-SV-100-1 3 RE L 1 SA D/1 C 2 A C TF-1 TF-1 R NO 1-SV-100-2 3 REL 1 SA 0/1 C 2 A C TF-1 TF"1

0

~ ~ ~ 0 ~ 0 0 0 ~ ~ 0 0 0 ~ ~ 0 QO

DONALD C. COOK NUCLEAR PLANT SECOND TEN YEAR INTERVAL RUN DATE AND TINE: 15FEB90:16:03 e

VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: EHERGENCY CORE COOLING - RHR FLON DIAGRAH: 1-5143-36 VALVE I VALVE POSITION I ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST<S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1-SV"100-3 3 REL 1 SA D/1 C 2 A C TF-1 TF-1 R NO 1-SV-100-4 3 REL 1 SA 0/1 C 2 A C TF-1 TF"1 1-SV-102 3 REL 0.75 SA E/5 C 2 A C TF1 i TF-1 1-SV-103 3 REL 3 SA F/B C 2 A C TF1 TF-1 1-SV-104E 3 REL 2 SA G/4 C 2 A C TF-1 TF-1 1-SV-104N 3 REL 2 SA K/4 C 2 A C TF1 TF-1 R NO 1-SV-105E 3 REL 2 SA 0/9 C 2 A C TF-1 TF-1 R NO 1-SV-105W 3 REL 2 SA D/9 C 2 A C TF-1 TF-1 R NO

e e e e e e e e e e e e e e ,

I e e e o o e o e r e e e e e e o 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SENARY SHEET - UNIT 1 SYSTEM NAME: CONTAINMENT SPRAY FLOH DIAGRAM: 1-5144-28 VALVE I VALVE POSITION I ASME SECTION XI NUBBER REV TYPE SIZE ACT F ~ D I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUESTERS)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-CTS-103-E 3 CK 10 SA J/9 C 0 2 A C CF-1 CF-3 YES> NOTE 2 1-CTS-103-H 3 CK 10 SA L/9 C 0 2 A C CF1 CF-3 YES'OTE 2 1-CTS-109 3 VB 1 SA H/6 C 0 2 A C CF1 CF-2 C NO> CSJ 6 1-CTS-110 3 VB 1 SA M/6 C 0 2 A C CF1 CF-2 C NOp CSJ 6 1-CTS-120-E 3 CK 2 SA H/8 C 0 2 A C CF1 CF-1 1-CTS-120-H 3 CK 2 SA K/8 C 0 2 A C CF1 CF-1 P NO 1-CTS-127-E 3 CK 6 SA E/5 C 0 2 A AC CF-1 CF-2 R YES> NOTE 4 SLT-1 SLT-2A R YES> NOTE 7 1-CTS"127-H 3 CK SA E/4 C 0 2 A AC CF-1 CF-2 R YES) NOTE 4 SLT-1 SLT-2A R YESi NOTE 7 1-CTS-131-E 3 CK 8 SA E/2 C 0 2 A AC CF-1 CF-2 R YES> NOTE SLT-1 SLT-2A R YES> NOTE 7 1-CTS-131-H 3 CK 8 SA E/2 C 0 2 A AC CF 1 CF-2 R YES> NOTE 3 SLT-1 SLT-2A R YES> NOTE 7 0 1-CTS-138-E 3 CK 12 SA G/9 C 0/C 2 A C CF-1 CF-3 YES> NOTE 1

e e e e e e o o e e e e e e DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: )5FEB90:)6:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET " UiNIT 1 SYSTEH NAME: CONTAINMENT SPRAY FLOW DIAGRAM: 1"5144-28 VALVE I VALVE POSITION I ASHE SECTION XZ NRSER REV TYPE SIZE ACT F D t POWER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REIWEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED MODE 1-CTS-138-N 3 CK 12 SA J/9 C 0/C 2 A C CF-1 CF-3 YES> NOTE 1 1-IHO-202 3 GA 2.5 HO H/7 C 0 2 A B EF-1 EF-1 EF-5 EF-5 NO ET-XXX ET-XXX P NO

) "IHO-204 3 GA 2.5 HO H/7 C 0 2 A B EF) EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-IHO-210 3 GA 10 HO J/8 C 2 A B EF1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-ZHO-211 3 GA 10 HO J/8 C 0 2 A B EF) EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

)<<IHiO-212 3 GA 2 HO H/8 0 0/C 2 A B EF-1 EF-1 P NO EF-5 E F-5 NO ET-XXX ET-XXX P NO 1-IHO-215 3 GA 12 HO G/9 0 0/C 2 A B EF-1 EF-1 P- NO EF-5 EF-5 NO ET-XXX ET-XXX P NO

)-IHO-220 3 GA 10 HO L/8 C 0 2 A B EF1 EF-1 P NO EF"5 EF-5 NO ET-XXX ET-XXX P NO

0 0 0 0 0 0 0 0 0

N DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUMi'lARY SHEET - UiNIT 1 SYSTEM NAME: CONTAItKENT SPRAY FLOH DIAGRAM: 1"5144-28 VALVE I VALVE POSITION I ASHE SECTION XZ NUMBER REV TYPE SIZE ACT F.D. t PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 0

1-IHO"221 3 GA 10 HO L/8 C 2 A 8 EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX .P NO 1-ZHO-222 3 GA 2 HO L/9 0 0/C 2, A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX P NO 1-ZHO-225 3 GA 12 MO J/9 0 0/C 2 A B EF-1 EF-1 P NO EF-5 EF-5 NO ET-XXX ET-XXX I-RH-141 3 CK 8 SA E/3 C 2 A AC CF 1 CF-2 R YES> NOTE 5 SLT-1 SLT-2A R YES> NOTE 7 1-RH-142 3 CK 8 SA E/3 C 2 A AC CF-1 CF-2 R YES> NOTE 5 SLT-1 SLT-2A R YES> NOTE 7 1-SV-107 3 REL 1 SA N/5 C 2 A C TF-1 TF-1 R NO,

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: CPN/HELD CMANNEL PRESSURIZATION FLON DIAGRAH: I-5145-17 VALVE VALVE POSITION I ASHE SECTION XZ NRSER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE 1-CA-181-N 3 CK 0. 5 SA F/2 C 2 P AC SLT-1 SLT-2 R YES'OTE 1 1-CA-181-S 3 CK 0.5 SA F/3 C 2 P AC SLT-1 SLT-2 R YES'OTE 1

0 0

0

DONALD Co COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEM NAME! ICE CONDENSER REFRIGERATION FLON DIAGRAM: 1-5146B-24 VALVE I VALVE POSITION I ASHE SECTION XZ NUMBER REV TYPE SI2E ACT F.D. I PONER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUESTERS)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-R-156 3 CK 0.375 SA L/4 C 0/C 2 A AC CF-1 CF-2 R YESp NOTE 1 SLT-1 SLT-2 R YESp NOTE 2 1-R-157 3 CK 0.375 SA L/6 C 0/C 2 A AC CF-1 CF-2 R YESp NOTE 1 SLT-1 SLT-2 R YESp,NOTE 2 1-VCR-10 DA A H/5 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2 1-VCR-11 3 DA 4 A L/5 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2 1"VCR-20 3 DA 4 A M/7 0 C 2 A A EF-1 EF-1 EF-5 EF-5 EF-7 EF-7 P NO ET-XXX ET-XXX P NO, SLT-1 SLT"2 R YESp NOTE 2 1-VCR"21 3 DA 4 A L/7 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 2

0 e e e e e e e e e 0 o e e o e'e o o e e e e e i e 0

DONALD C. COOK NUCLEAR PLANT RUH DATE AND TIME: 15FEB90:16:03 SECOHD TEN YEAR INTERVAL VALVE

SUMMARY

SHEET

'- UNIT 1 SYSTEM NAME: CONTAINMENT VENTZ LATION FLOH DIAGRAM: 1-5147A-34 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER TYPE COORDI OPER SAFETY ICD A/P FUHCT ICL CAT PRIM TEST REQUIRED

'EST PERFORMED TEST MODE RELIEF REQUEST(S) 1-SM-10 3 GA 0.5 M A/4 C 2 P A SLT-1 SLT-2 R YES p NOTE 1 1-SM-4 3 GA 0.5 M A/2 C P A SLT-1 SLT-2 R YES% NOTE 1 1-SM-6 3 GA 0.5 M A/2 C 2 P A SLT-1 SLT-2 R YES> NOTE 1 1-SM-8 GA 0.5 M A/4 C 2 P A SLT-1 SLT-2 R YES> NOTE 1 1-VCR"101 3 BF 14 A J/8 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-VCR-102 3 BF 14 A J/9 C 2 P A EF-1 EF-1 EF"5 E F-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-VCR-103 3 BF 24 A J/5 C 2 P A EF-1 EF-1 P HO EF-5 EF-5 HO EF-7 EF-7 P HO ET-XXX ET-XXX P HO SLT"1 SLT-2 R YES> NOTE 1 1-VCR"104 3 BF 30 A J/6 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

0 o o o e o o e e e e e o e a 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SmARY SHEET - UNIT 1 SYSTEM NAME: CONTAINMENT VENTILATION FLON DIAGRAM: 1-5147A-34 VALVE l VALVE POSITION ASME SECTION XI NUii3ER REV TYPE SIZE ACT F.D. I PONER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-VCR"105 3 BF 30 A J/3 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT"1 SLT-2 R YESp NOTE 1 1-VCR-106 3 BF 24 A J/3 C 2 P A EF-1 EF-1 EF-5 EF-5 HO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-VCR-107 3 BF 14 A J/4 C/0 C 2 A A EF-1 EF-1 EF-5 EF-5 EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-VCR-201 3 BF 14 A J/8 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> HOTE 1 1-VCR-202 3 BF 14 A J/9 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

0 0 0 0 0 0 0 0 DONALD C ~ COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SU%)ARY SHEET - UHIT 1 SYSTEN HAHE: CONTAINMENT VENTILATION FLON DIAGRAH: 1-5147A-34 VALVE I VALVE POSITION I ASNE SECTION Xr NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUHCT ICL REQUIRED PERFORHED NODE 1-VCR-203 3 BF 24 A J/5 C 2 P A EF-1 EF-1 P NO EF-5 EF"5 NO EF-7 EF-7 P NO ET"XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-VCR-204 3 BF 30 A J/6 C 2 P A EF-1 EF-1 P NO EF"5 EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES~ NOTE 1 1-VCR-205 3 BF 30 A J/3 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT"1 SLT-2 R YES> NOTE 1 1"VCR-206 3 BF 24 A J/3 C 2 P A EF-1 EF-1 P NO EF-5 EF-7 EF-5 EF-7 P.. NO HO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YES> NOTE 1 1-VCR-207 3 BF 14 A J/4 C/0 C- 2 A A EF-1 EF-1 P, NO EF-5 EF-5 NO EF-7 EF-7'T-XXX P NO ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUiiARY SHEET UNIT 1 SYSTEH NAHE: CONTROL ROON VENTILATION FLON DIAGRAH: I-5109-20 VALVE I VALVE POSITION I ASHE SECTION XZ RAKER REV TYPE SIZE ACT F.D. ) PQ'lER SAFETY )CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHEO NODE I-DH-163-N 3 GA 2.5 H F/2 0 0/C 3 A B EF-1 EF-1 I-DN-163-S 3 GA 2.5 H G/2 0 0/C 3 A B EF-1 EF-1 1"DN-166-N 3 GA 2.5 H E/5 0 0/C 3 . A B EF-1 EF-1 P NO 1-DN-166-S 3 GA 2.5 H J/5 0 0/C 3 A B EF-1 EF-1 1-VRV-315 3 3N 2.5 A F/5 0 0 . 3 A B EF-1 NOTE 1 P YESp NOTE 1 EF-7 EF"7 P NOp NOTE 1 ET-XXX NOTE 1 YES> NOTE 1 1-VRV"325 3W 2.5 A G/5 0 0 3 A B EF1 NOTE 1 P YES> NOTE 1 EF-7 EF-7 P NO> NOTE 1 ET-XO NOTE 1 YES> NOTE 1

0 DONALD C. COOK NUCLEAR PLANT RUH DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: EMERGENCY DIESEL GENERATOR FLOH DIAGRAM: 1-5151A-25 VALVE I VALVE POSITIOH I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. t POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1 "DL-113A 3 CK 1.5 SA B/9 0 A C CF-1 CF-1 P HOp NOTE 1 1-DL-115A 3 CK 1.5 SA 8/9 C 3 A C CF1 CF-1 P NO> NOTE 1 1-DL-125A 3 CK 2.5 SA E/9 0 A C CF 1 CF-1 P NO>NOTE 1 1-DL-131A 3 CK 1 SA F/9 0 3 A C CF1 CF-1 P NO> NOTE 1 1-DL-157A 3 CK 6 SA G/6 C 3 A C CF-1 CF-1 P NO> NOTE 1 I-QT-114-lAB 3 3H 6 " SA M/5 0 3 A B EF1 NOTE 2 P HOp NOTE 2

0 e e e e e o e e e e o e e e 0 o e e e o e e e

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: EHERGENCY DIESEL GENERATOR FLON DIAGRAH: 1-5151B-28 VALVE I VALVE POSITION I ASHE SECTION Xl NUHBER REV TYPE SIZE ACT F ~ D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE

'-DG-101A 3 CK 1. 5 SA H/4 0 0/C 3 A C CF-1 CF-1 P NOp NOTE 1 1-DG-103A 3 CK 1.5 SA F/3 0 0/C 3 A C CF-1 CF-1 P NO> NOTE 1 1-DG-127A 3 CK 1 SA C/4 C 3 A C CF1 CF-1 P NO> NOTE 1 1-DG-129A 3 CK 1 SA C/4 C 3 A C CF1 CF"1 P NO> NOTE 1 1-DG-139A 3 CK 0.5 SA F/1 C 0/C 3 A C CF-1 CF-1 ,P NO 1-DG-141A 3 CK 0. 5 SA F/1 C 0/C 3 A C CF-1 CF-1 1-DG-145A 3 CK 2 SA A/8 0 A C CF1 CF-1 P NO) NOTE 1 1-DG-151A 3 CK 4 SA D/8 C 3 A C CF1 CF-1 .

P NO> NOTE 1 1-DG-153A 3 CK 4 SA C/8 C A C CF-1 CF-1 P NOp NOTE 1 1-QT-132-lAB 3 3H 6 SA E/8 0 3 A B EF1 NOTE 2 P NO> NOTE 2 1"SV-120"lAB 3 REL 0.25 SA G/2 C 3 A C TF1 TF-1 R NO 1-SV-139-lAB 3 REL 1 SA B/2 C A C TF-1 TF-1

0 e o o e o e o o o e e e e e e DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL, VALVE

SUMMARY

SHEET - UNIT 1 SYSTEH NAME: EHERGENCY DIESEL GENERATOR FLOW DIAGRAM: 1-5151B-28 VALVE I VALVE POSITION I ASHE SECTION XI NUMBER REV TYPE SIZE ACT F.D ~ I POWER SAFETY ICD A/P CAT PRIM TEST TEST, TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1"SV-61-1AB 3 REL 1 SA A/8 C 3 A C TF-1 TF-1 1-SV-78-1AB1 3 REL 1 SA E/3 C 3 A C TF1 TF-1 1-SV-78-lAB2 3 REL 1 SA 0/3 C A C TF-1 TF-1 R NO 1-SV-79"lAB1 3 REL 0.5 SA E/1 C A C " TF-1 TF"1 R NO 1-SV-79-1AB2 3 REL 0.5 SA E/1 C A C TF1 TF-1 R NO 1-XRV-220 3 GA 1 A B/3 C 3 A B EF1 EF-1 YES> NOTE 4 EF-7 EF-7 NO> NOTE 4 ET-XXX NOTE 4 YES> NOTE 4 1-XRV"221 3 'L 3 A B/4 C A B EF1 EF-1 YES> NOTE 3 EF-7 NOTE 3 YES> NOTE 3 ET-XXX NOTE 3 YES> NOTE 3 1-XRV-222 3 GL 3 A B/4 C 3 A B EF-1 EF-1 YES> NOTE 3 EF-7 NOTE 3 YES> NOTE 3 ET-XXX NOTE 3 YES> NOTE 3

0 0 i 0

0 e e e o o e e e e e e e e e

0 DONALD C. COOK NUCLEAR PLANT SECOND TEN YEAR INTERVAL RUN DATE AND TIME: 15FEB90:16:03 VALVE SUMHARY SHEET - UNIT 1 SYSTEM NAHE: EMERGENCY DIESEL GENERATOR FLON DIAGRAM: 1-5151C-26 VALVE I VALVE POSITION I- ASHE SECTION XI NUMBER REV TYPE SIZE ACT F D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD I OPER FUNCT ICL REQUIRED PERFORMED MODE I-DF-108C 3 CK 1.5 SA J/3 C 0 3 A C CF1 CF-1 P NO> NOTE 1 1-DF-109C 3 CK 1.5 SA K/3 C 0 A C CF-1 CF-1 P NOy NOTE 1 1-DF-114C 3 CK 1.5 SA L/3 C 0 3 A C CF-1 CF-1 P NOp NOTE 1 1-DF-115C 3 CK 1.5 SA H/3 C 0 3 A C CF"1 CF-1 P N01 NOTE 1 1-DL-113C 3 CK 1.5 SA B/9 0 C A C CF-1 CF-1 P NO> NOTE 1 1-DL-115C 3 CK 1.5 SA B/9 C 0 3 A C CF-1 CF-1 P NO> NOTE 1 1-DL-125C 3 CK 2.5 SA E/9 0 C 3 A C CF1 CF-1 P NO> NOTE 1 1-DL-131C 3 CK 1 SA F/9 0 C 3 A C CF-1 CF-1 P NO> NOTE 1 1-DL-157C 3 CK 6 SA G/5 C 0 3 A C CF-1 CF-1 P NOp NOTE 1 1-QT-114-1CD 3 3H 6 SA H/5 0 0 3 A B EF-1 P NO> NOTE 2

, e e 0 e e e 0' 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TZHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUKfARY SHEET " UNIT 1 SYSTEM NAME: EMERGENCY DIESEL GENERATOR ".CD" FLOH DZAGRAH: 1-515lD-28 VALVE I VALVE POSITION I ASME SECTION XZ NUMBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P = CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-DG-101C 3 CK 1.5 SA H/4 0 0/C 3 A C CF-1 CF-1 P NO> NOTE 1 1-DG-103C 3 CK 1.5 SA F/3 0 0/C 3 A C CF-1 CF-1 P NO> NOTE 1

'-DG-127C 3 CK 1 SA C/3 C 3 A C CF1 CF-1 P NO> NOTE 1 1-DG-129C 3 CK 1 SA C/3 C A C CF-1 CF-1 P NO> NOTE 1 1-DG-139-CD 3 CK 0.5 SA F/1 C 0/C 3 A C CF1 CF-1 I-DG-101-CD 3 CK 0.5 SA F/1 C 0/C 3 A C CF-1 CF-1 I-DG-165C 3 CK 2 SA A/9 - 0 3 A C CF-1 CF-1 P NO> NOTE 1 1-DG-151C 3 CK SA D/9 C 3 A C CF-1 CF-1 P NO> NOTE 1 1-DG-153C 3 CK SA C/9 C 3 A C CF1 CF-1 P NO> NOTE 1 1-QT-132-1CD 3 3N 6 SA E/8 0 3 A B EF1 NOTE 2 P NO> NOTE 2 1-SV-120-1CD 3 REL 0.25 SA H/2 C 0 3 A C TF-1 TF-1 1-SV-139-1CD 3 REL 1 SA B/2 C A C TF1 TF-1 R NO

o e o e e o e e e e e e o e e 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET - UNIT 1 SYSTEH NAHE: EHERGENCY DIESEL GENERATOR "CD" FLON DIAGRAH: 1-5151D-28 VALVE I VALVE POSITION I ASHE SECTION XI NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUESTERS)

TYPE COORDJ OPER FUNCT ICL REQUIRED PERFORHED NODE 1-SV"61-1CD 3 REL 1 sa are c 3 A C TF-1 TF-1 1-SV-78-1CD1 3 REL 1 SA E/3 C A C TF-1 TF-1 1-SV-78-1CD2 3 REL 1 SA 0/3 C -"

0 3 A C TF"1 TF"1 R NO 1"SV-79-iCDi 3 REL 0.5 SA E/1 C 3 A C TF-1 TF-1 R -NO 1-SV-79"1CD2 3 REL 0.5 SA E/1 C A C TF-1 TF-1 R NO 1-XRV-225 3 GA 1 A 8/3 C A B EF-1 EF-1 YES> NOTE 4 EF-7 EF-7 YES> NOTE 4 ET-XXX NOTE 4 YES'OTE 4 1-XRV-226 3 GL 3 A B/4 C 3 A B EF1 EF-1 YES> NOTE 3 EF-7 NOTE 3 YES> NOTE 3 ET-XXX NOTE 3 YESp NOTE 3 1"XRV"227 3 GL 3 A B/4 C 3 A B EF1 EF-1 YES> NOTE 3 EF-7 NOTE 3 YES> NOTE 3 ET-XXX NOTE 3 YES'OTE 3

'e e o o e e .o o e e e e e e e 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET UNIT 1 SYSTEH NAHE: HAKE UP S PRIHARY HATER UNIT 1 FLON DIAGRAH: 12-5115A&1 VALVE l VALVE POSITION I ASHE SECTION XI HOSER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED HODE 1-QCR-919 3 DA . 2 A 0/7 0/C C 2 A A EF-1 EF-) P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-QCR-920 3 DA 2 A D/7 0/C C 2 A A EF-1 EF-1 EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1

0 DONALD C. COOK NUCLEAR PLANT RUH DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: COMPRESSED AIR SYSTEM UHIT 1 FLOH DIAGRAM: 12-5120B-22 VALVE VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POHER SAFETY )CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUHCT ICL REQUIRED PERFORMED MODE 1"PA-3Q 3 CK 2 SA B/7 0/C C 2 A AC CF-1 CF-2 R YES> NOTE 1 SLT-1 SLT-2 R YES'OTE 2 1-PCR-60 GA 2 A D/7 0/C C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF"7 P NO ET-XXX ET-XX P HO SLT-1 SLT-2 R YES> NOTE 2 1-XCR-100 3 GL 1 A L/3 0 C 2 A A EF-1 EF-2 C NO> CSJ 3 EF-5 E F-5 HO EF-7 EF-8 C NOp CSJ 3 ET-XXX ET-XXX C p NO> CSJ 3 SLT-1 SLT-2 R YES> NOTE 2 1"XCR-101 3 GL 1 A L/3 0 C 2 A A EF-1 EF-2 C HOp CSJ 3 E F-5 EF-5 NO EF-7 EF-8 C HOp CSJ 3 ET-XXX ET-XXX C NO> CSJ 3 SLT-1 SLT-2 R YES> NOTE 2 1-XCR-102 3 GL 1 A L/2 0 C A A EF-1 EF-2 C HOp CSJ 3 EF-5 E F-5 NO EF-7 EF-8 C NO> CSJ 3 ET-XXX ET-XXX C NO> CSJ 3 SLT-1 SLT-2 R YESp NOTE 2 1-XCR-103 3 GL 1 A L/2 0 C 2 A A EF-1 EF-2 C NO> CSJ 3 EF-5 EF-5 NO EF-7 EF-8 C NO> CSJ 3 ET-XXX ET-XXX C NO> CSJ 3 SLT-1 SLT-2 R YES> NOTE 2

e e o o o e o e o e e e e e o 0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: CVCS - BORON HAKE-UP - UNITS 1 & 2 FLON DIAGRAM: 12-5131-19 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I POMER SAFETY (CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDl OPER FUNCT ICL REQUIRED PERFORMED MODE 1-CS-415-1 3 CK 2 SA H/6 0/C '

3 A C CF1 CF-1 P NO I-CS<15-2 3 CK 2 SA H/6 0/C 0 3 A C CF1 CF-1 P NO I-CS<26-N 3 CK 1 SA G/6 0/C 0 3 A C CF-1 - CF-1 P NO 1"CS-427-N 3 CK 2 SA G/5 C 3 A C CF1 CF-2 C NO> CSJ 1 I-QMOW10 3 GL 2 HO G/5 C 3 A B EF1 EF-1 P NO EF-5 Ef-5 NO ET-XXX ET-XXX P NO 1-QRV-411 3 GL 1 A G/6 0/C 0 A B EF1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO 1-QRV-412 3 GL 2 A F/7 0 3 A B EF1 EF-1 P NO EF-5 EF-5 NO EF"7 EF-7 P NO ET-XXX ET-XXX P NO

0 0

0 o e e e e e e e e e e e e o e

DONALD C. COOK NUCLEAR PLANT SECOND TEN YEAR INTERVAL VALVE SUNDRY SHEET - UNIT 1 SYSTEM NAME: SPENT FUEL PZT COOLING 8 CLEANUP Ul FLON DIAGRAM: 12-5136-25 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORD1 OPER FUNCT ICL REQUIRED PERFORMED HODE 1-SF-151 3 DA 2.5 H K/8 C 2 P A SLT-1 SLT-2 R YES> NOTE 1 1-SF-153 3 GL 2 5 H K/8 C 2 P A SLT-1 SLT-2 R YES) NOTE 1 12-SF "118N 3 CK' SA J/5 0/C 0 3 A C CF-1 CF"1 12-SF-118S 3 CK 8, SA J/5 0/C 0 3 A C CF-1 CF-1

0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TINE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHHARY SHEET " UNIT 1 SYSTEH NAHE: NDS VENTS 8 DRAINS FLON DZAGRAH: 12-5137A-21 VALVE I VALVE POSITION I ASHE SECTION XZ NUHBER REV TYPE SIZE ACT F.D. I POHER SAFETY ICD A/P CAT PRZH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED NODE

'1-DCR"201 3 DA 1 A E/4 C C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-DCR-202 3 DA 0.75 A E/5 0 C 2 A A EF-1 EF-1 EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-DCR-203 .

3 DA 1 A F/4 C C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET"XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 1 1-DCR-204 3 DA 0.75 A F/5 0 C 2 A A EF-1 EF-1 P NO EF"5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESi NOTE 1 1-DCR-205 3 GL A E/7 0/C C 2 A A EF-1 EF-1 P NO EF-5 EF-5 EF-7 ET-XXX

'EF-7 ET-XXX P,

P NO NO NO SLT-1 SLT-2 R YES> NOTE 1

0 e e e o e e e e e e o e o DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: HOS VENTS 8 DRAINS FLOH DIAGRAM: 12-5137A-21 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F ~ D. I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUHCT ICL REQUIRED PERFORMED MODE 1-DCR"206 3 GL A E/8 0/C C 2 A A EF-1 EF-1 P HO EF-5 EF"5 HO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-DCR-207 3 DA 1 A F/0 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P HO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-DC R-610 3 DA 2.5 A M/9 0 2 A A EF-1 EF-1 P NO Ef-5 E F-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YESp NOTE 1 1-DCR" 611 DA 2.5 A N/9 0 2 A A EF-1 EF-1 P NO EF-5 'F-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-DCR-620 3 DA 1 A M/9 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P HO SLT-1 SLT-2 R YESp NOTE 1

0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEM NAME: NDS VENTS & DRAINS FLON DIAGRAH: 12-5137A-21 VALVE I VALVE POSITION I ASME SECTION XI NUHBER REV TYPE SIZE ACT F.D. I PONER SAFETY ICD A/P CAT PRIH TEST TEST TEST RELIEf REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED MODE 1-DCR-621 3 DA 1 A N/9 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-N-160 CK 1 SA F/0 0 2 A AC CF-1 CF-2 R YESp NOTE 2 SLT-1 SLT-2 R YESp NOTE 1 1-SF-159 3 DA 3 H E/5 C 2 P A SLT-1 SLT-2 R YESp NOTE 1 1-SF-160 3 DA H F/5 C 2 P A SLT-1 SLT-2 R YESp NOTE 1

e e o e e e e e e e e o o e e DONALD CD COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE

SUMMARY

SHEET - UNIT 1 SYSTEH NAME: PAS LIQUID 8 GAS - UNIT-1 FLON DIAGRAM: 12-5141C-B VALVE I VALVE POSITION ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D. t POHER SAFETY I CD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-ECR-416 3 GL 0.5 A B/6 C 2 P A EF"1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-ECR-417 3 GL 0.5 A B/6 C 2 P A EF-1 EF-1 EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT<<2 R YESp NOTE 1 1-ECR-496 3 GL 0.5 A B/8 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET<<XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-ECR-497 3 GL 0.5 A , B/8 C 2 P A EF-1 EF-1 P . NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1 1-ECR-535 3 GL 0.5 A B/2 C 2 P A EF-1 EF-1 P NO EF-5 EF-5 NO EF"7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YESp NOTE 1

e e e e e e e e e. e e e e '

0 e o o e o o o e o e e o e e o 0

DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIHE: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUHMARY SHEET - UNIT 1 SYSTEH NAME: PAS LIQUID 2 GAS - UNIT-1 FLON DIAGRAM: 12-5141C-8 VALVE I VALVE POSITION I ASME SECTION XI NASSER REV TYPE SIZE ACT F.D. t PONER SAFETY I CD A/P CAT PRIH TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORMED HODE 1-ECR-536 3 GL 0.5 A 8/2 C 2 P A EF1 EF-1 P NO Ef-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES) NOTE 1

0 DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUMNLRY SHEET UNIT 1 SYSTEM NAME: PAL SAMPLING 8 INST. PANELS - U"1 FLON DIAGRAM: 12-5141F-6 VALVE I VALVE POSITION I ASHE SECTION XZ NUMBER REV TYPE SIZE ACT F AD. I PONER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST(S)

TYPE COORDI OPER FUNCT ICL REQUIRED PERFORHED MODE 1-ECR-31 GL 1 A B/5 0 C 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 2 1-ECR-32 3 GL 1 ' B/5 0 2 A A EF-1 EF-1 EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 2 1-ECR-33 GL 0.75 A B/5 0 2 A A EF-1 EF-1 P NO EF-5 EF-5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 2 1"ECR-35 GL 1 A B/5 0 2 A A EF-1 EF-1 P NO EF-5 EF"5 NO EF-7 EF-7 P NO ET-XXX ET-XXX P NO SLT-1 SLT-2 R YES> NOTE 2 1-ECR-36 3 GL 1 A B/6 0 2 A A EF-1 EF-2 C NO> CSJ 1 EF-5 EF-5 NO EF-7 EF-8 C NO> CSJ 1 ET-XXX ET-XXX C NOp CSJ 1 SLT-1 SLT-2 R YES@ NOTE 2

e e e e e o e o e e e e e e e DONALD C. COOK NUCLEAR PLANT RUN DATE AND TIME: 15FEB90:16:03 SECOND TEN YEAR INTERVAL VALVE SUMHARY SHEET - UNIT 1 SYSTEM NAME: PAL SAMPLING 8 INST. PANELS - U-l FLOH DIAGRAM: 12-5141F-6 VALVE I VALVE POSITION I ASME SECTION XI NUMBER REV TYPE SIZE ACT F.D I POHER SAFETY ICD A/P CAT PRIM TEST TEST TEST RELIEF REQUEST[ S)

TYPE COORDI OPER FUNCT lCL REQUIRED PERFORMED MODE 1-SH-1 3 CK 1 SA A/6 0 2 A AC CF1 CF-2 R YES>, NOTE 3 SLT-1 SLT-2 R YES> NOTE 2

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CRITICALITYANALYSIS OF THE DONALD C COOK NUCLEAR PLANT FUEL RACKS November 1989 W. A. Boyd M. W. Fecteau J. A. Penkrot W. P. Kovacik J. L Bradfute B. W. Schmidt R. C. Cobb W. A. Bordogna

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TABLE OF CONTENTS

.0 Introduction ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.1 Design Description ~ ~ ~ 1 1.2 Design Criteria ~ ~ ~ 2 2.0 Analytical Methods ~ ~ ~ 3 2.1 Criticality Calculation Methodology ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 2.2 Reactivity Equivalencing Methodology ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 Criticality Analysis of Region I 3.0 'I Spent Fuel Racks 0 ~ ~ ~ ~ ~ ~ ~ ~

3.1 Reactivity Calculations ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 3.2 Postulated Accidents ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3.3 Sensitivity Analysis ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e 9 j 4.0 4.1 4.2 4.3 Criticality Analysis of Region 2 Spent Fuel Racks Reactivity Equivalencing Reactivity Calculations Postulated Accidents ~

10 10 10 13 4.4 Sensitivity Analysis 13 5,0 Criticality Analysis of Fresh Fuel Racks 14 5.1 Full Density Moderation Analysis 15 5.2 Low Density. Optimum Moderation Analysis 16 6.0 Summary of Criticality Results ~ ~ ~ ~ ~ ~ ~ ~ 18 B ibliography ................................................. 38 Table of Contents

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LIST OF TABLES Table 1. Fuel Parameters Employed in Criticality Analysis 1S Table 2. Benchmark Critical Experiments [5,6] 20 Table 3. Comparison of PHOENIX Isotopics Predictions to Yankee Core 5 Measurements 21 Table 4. Benchmark Critical Experiments PHOENIX Comparison 22 Table 5. Data for U Metal and UO2 Critical Experiments 23 t

List of Tables

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LIST OF ILLUSTRATIONS Figure Figure 1.

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Nominal Dimensions ........................

Donald C Cook Nuclear Plant Spent Fuel Pool Storage Cell Donald C Cook Nuclear Plant SFP Region 1 Three of Four Fuel 25 Assembly Loading Schematic ..... ~............... 26 Figure 3. Donald C Cook Nuclear Plant Schematic for SFP Interface Boundary Between Regions 1 and 2 Figure 4. Donald C Cook Nuclear Plant Fresh Fuel Rack Radial Layout Figure 5. Donald C Cook Nuclear Plant Fresh Fuel Rack Axial Layout Figure 6. Sensitivity of Keff to Enrichment in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 30 Figure 7. Sensitivity of Keff to Center-to-Center Spacing in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 31 Figure 8. Sensitivity of Ken to B Loading in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 32 Figure 9. Donald C Cook Nuclear Plant SFP Region 2 Fuel Assembly Figure 10.

Minimum Burnup vs. Initial U Enrichment Curve ........

Sensitivity of Keff to Enrichment in the Donald C Cook Nuclear 33 Figure 11.

Plant SFP Region 2 Storage Area . ~...............,

Sensitivity of Keff to Center-to-Center Spacing in the Donald C 34 Cook Nuclear Plant SFP Region 2 Storage Area ... ~..... 35 Figure 12. Sensitivity of Koff to B Loading in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area ..... ~... ~........ 36 Figure 13. Sensitivity of Keff to Water Density in the Donald C Cook Nuclear Plant New Fuel Storage Vault ..................... 37 List of Illustrations

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1.0 INTRODUCTION
This report presents the results of the criticality analyses for the storage of Westinghouse 15x15 and 17x17 fuel assemblies in the Donald C Cook Nuclear Plant Spent Fuel Pool (SFP) storage rack and the New Fuel Storage Vault (NFSV).
The SFP rack design considered herein is an existing array of Donald C Cook Nuclear Plant SFP poisoned racks, which will be analyzed as two separate spent fuel arrays or regions. Region 1 will be analyzed for criticality using a three out of four assembly storage arrangement. The Region 1 analysis is presented in Section 3 of this report. Region 2 will be analyzed for criticality with as-sembly storage utilizing all locations. Region 2 will also be analyzed for burnup credit, which takes into consideration the changes in fuel and fission product inventory resulting from depletion in the reactor core. The Region 2 criticality and burnup credit analysis is presented in Section 4 of this report. Both the Region 1 and 2 analyses are based on maintaining Keff 5 0.95 for storage of Westinghouse 15x15 STD and OFA, and 17x17 STD, OFA and VANTAGE 5 fuel.
The NFSV rack design considered herein is an existing array of Donald C Cook Nuclear Plant NFSV unpoisoned racks which will be analyzed for criticality to show that Westinghouse 15x15 STD and OFA, and 17x17 STD, OFA and VANTAGE 5 fuel assemblies can be stored using all storage locations. The NFSV rack analysis is based on maintaining Keff 5 0.95 under full water density conditions and 5 0.98 under low water density (optimum moderation) conditions.
The NFSV analysis is presented in Section 5 of this report.
The Westinghouse 15x15 and 17x17 fuel parameters relevant to these analyses are given in Table 1 on page 19.
1.1 DESIGN DESCRIPTION The Region 1 and 2 spent fuel storage cell design is depicted schematically by Figure 1 on page 25 with nominal dimensions given on the figure. The Region 1 three out of four storage arrangement is shown in Figure 2 on page 26 and an example of the interface boundary between the Region 1 and 2 storage areas is given in Figure 3 on page 27.
The total number of SFP locations designated as Region 1 or 2 is left to the utility to determine. The boundary between the two regions can be drawn Introduction
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anywhere within the SFP racks, but the three of four assembly storage ar-rangement of the Region 1 area must be carried into the Region 2 area by at least one row. Therefore, even through Region 2 is analyzed for assembly storage using all cell locations, some cells may need to be left vacant near the Region 1 to 2 boundary to accomodate the Region 1 pattern carryover by one row (refer to.Figure 3 on page 27).
The fresh fuel rack storage rack radial layout is depicted in Figure 4 on page 28 and the axial layout is shown in Figure 5 on page 29.
'1.2 DESIGN CRITERIA Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor, Kerf, of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and in Reference 1. The 0.95 Ken limit applies to both the SFP and NFSV under all conditions, except for the NFSV under low water density toptimum moderation) conditions, where the Keff limit is 0.98 as recommended by NUREG<800.
Introduction
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2.0 ANALYTICALMETHODS 2.1 CRITICALITY CALCULATION METHODOLOGY The criticality calculation method and crosswection values are verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities.
i The design method which insures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPX 'ystem generation and KENO IV for reactivity determination.
of codes for crosssection t The 227 energy group crosswection library that is the common starting point r for all crosswections used for the benchmarks and the storage rack is generated from ENDF/B-V data. The NITAWL program includes, in this library, the selfwhielded resonance crosswections that are appropriate for each particular l geometry. The Nordheim integral Treatment is used. Energy and spatial weighting of crosswections is performed by the XSDRNPM program which is a one-dimensional Sn transport theory code. These multigroup cross-section sets are then used as input to KENO IV which is a three dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 33 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and uncertainty. The experiments range from water moderated, oxide fuel arrays separated by various materials (B4C, steel, water, etc) that simulate LWR fuel shipping and storage conditions to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials (Plexiglas and air) that demonstrate the wide range of applicability of the method. Table 2 on page 20 summarizes these experiments.
The average Keff of the benchmarks is 0.992 . The standard deviation of the bias value is 0.0008 dk. The 95/95 one sided tolerance limit factor for 33 values is 2.19. Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in- reactivity, due to the method, is not greater than 0.0018 Zdc.
Analytical Methods
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2.2 REACTIVITY EQUIVALENCING METHODOLOGY Spent fuel storage, in the Region 2 spent fuel storage racks, is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel depletion. A series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent Ke<<when the fuel is stored in the Region 2 racks.
The data points on the reactivity equivalence curve are generated with a trans-port theory computer code, PHOENIX . PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMS library is used with PHOENIX.
A study was done to examine fuel reactivity as a function of time following discharge from the reactor. Fission product decay was accounted for using CINDER . CINDER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for 30 years after discharge. The fuel reactivity was foun'd to reach a maximum at approx-imately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after discharge. At this point in time. the major fission product poison, Xe, has nearly completely decayed away. Furthermore, the fuel reactivity was found to decrease continuously from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 30 years following discharge. Therefore, the most reactive point in time for a fuel as-sembly after discharge from the reactor can be conservatively approximated by removing the Xe The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a reac-tor. In addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 3 on page 21.
The measurements were made on fuel discharged from Yankee Core 5 . The data in Table 3 on page 21 shows that the agreement between PHOENIX pred-ictions 'and measured isotopic compositions is good.
The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in Table 4 on page 22. Key parameters describing each of the 81 experiments are given in Table 5 on page
23. These reactivity comparisons again show good agreement between exper-iment and PHOENIX calculations.
An uncertainty associated with the burnupMependent reactivities computed with PHOENIX is accounted for in the development of the Region 2 burnup require-ments. A bias which increases linearly with burnup to 0.01 LLk at 30,000 MWD/MTU is applied to the PHOENIX calculational results. This bias is con-sidered to be very conservative since comparison between PHOENIX results and Analytical Methods
the Yankee Core experiments and 81 benchmark experiments indicates closer agreement (see Table 3 on page 21 and Table 4 on page 22). For the Donald C Cook Nuclear Plant SFP Region 2 analysis, the PHOENIX calculations for the maximum burnup of 5,550 MWD/MTU include a reactivity bias of 0.0019 M.
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Analytical Methods
I 3.0 CRITICALITY ANALYSIS OF REGION 1 SPENT FUEL RACKS This section develops and describes the analytical assumptions and models employed to perform the criticality analyses for storage of spent fuel in Region 1 of the Donald C Cook Nuclear Plant SFP.
3.1 REACTIVITY CALCULATIONS i The following assumptions were used to develop the nominal case KENO model for the Region 1 SFP rack storage of fresh fuel using three out of four storage locations as shown in Figure 2 on page 26.
l 1. The Westinghouse 17x17 OFA fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for t any burnable absorber in the fuel rods or any natural enrichment axial blankets (See Table 1 on page 19 for fuel parameters). Evaluation of the Westinghouse 15x15 and 17x17 fuel assemblies shows that the 17x17 OFA assembly is the most reactive fuel type when all assemblies have the same T enrichment. Therefore, only the Westinghouse 17x17 OFA fuel assembly waS analyzed,
2. All fuel rods contain uranium dioxide at an enrichment of 4.95 w/o U over the finite 144 inch length of each rod.. The fuel pellets are assumed to be at 96% of theoretical density, and no credit is taken for dishing or chamfering. If nominal theoretical density and pellet parameters were used, the resultant enrichment limit would be 5.06 w/o U . Therefore, the 4.95 w/o U enrichment limit can be considered a nominal enrichment limit since the conservative assumptions employed in the pellet modelling bound the standard 0.05 w/o enrichment tolerance.
3. No credit is taken for any U or U in the fuel, nor is any credit taken for the build up of fission product poison material.
l 4. The moderator is pure water at a temperature of 68 F. A conservative value of 1.0 gm/cm is used for the density of water.
5. No credit is taken for any spacer grids or spacer sleeves.
Criticalityc Analysis of Region 1 Spent Fuel Racks
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6. Fuel assemblies are loaded into three of every four cells in a checkerboard pattern in the storage cells as shown in Figure 2 on page 26.
7. The array is infinite in lateral extent and finite in axial extent which allows neutron leakage from only the axial direction.
8. The minimum poison material loading of 0.02 grams B per square centi-meter is used throughout the array.
The KENO'calculation for the nominal case resulted in a Keff of 0.9005 with a 95 percent probability/95 percent confidence level uncertainty of +0.0065. The nominal case result can be compared to the worst case result to determine the relative impact of applying the worst case assumptions. The nominal case is also used as the center point for the sensitivity analyses discussed in Section 3.3.
The maximum Kore under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of fuel assemblies within the L storage cells. Westinghouse internal studies of asymmetric positioning of fuel i assemblies within the storage cells have shown that symmetrically placed fuel assemblies yield equal or conservative results in rack Keff. The sheet metal i
tolerances are considered along with construction tolerances related to the cell I.D., and cell center-to~enter spacing. For the Region 1 racks this resulted in a reduction of the nominal center to center spacings to their minimum values.
Thus, the "worst case" KENO model of the Region 1 storage racks contains the i minimum center to center spacings with symmetrically placed fuel assemblies.
Based on the analysis described above, the following equation is used to de-velop the maximum Kon for the Donald C Cook Nuclear Plant Region 1 spent fuel storage racks with three out of four storage:
Keff = Kworst + Bmothod + Bplrt + /[ 2 2 (ks) worst + (kS) mothod ]
where:
Kworst worst case KENO Koff that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal Bmothod method bias determined from benchmark critical comparisons Criticality Analysis of Region 1 Spent Fuel Racks
I, I Bpa~ bias to account for poison particle self-shielding. This standard term accounts for the increased neutron transmission through the poison plate due to the inherent effects of poison particle selfwhielding, and has been analytically determined for poison plates similar to those used in this analysis.
ksworat 95/95 uncertainty in the worst case KENO Keff ksmeaaa = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
Keff = 0.9308 + 0.0083 + 0.0014 + / [(0.0046) + (0.0018) ] = 0.9454 Since Ke~r is less than 0.95 including uncertainties at a 95/95 probability/confidence level, the acceptance criteria for criticality is met with fuel enriched to a nominal 4.95 w/o.
3.2 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in Ke~r of the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the'ack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes interaction).
However, accidents can be postulated which would increase reactivity (i.e.,
dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by about 0.25 hK. Thus, for postulated accidents, should there be a reactivity increase, Kafka would be less than or equal to 0.95 due to the effect of the dissolved boron. Since the Donald C Cook Nuclear Plant.SFP will be maintained at a boron concentration of 2400 ppm, additional margin will exist Criticality Analysis of Region 1 Spent Fuel Racks
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3.3 SENSlTIVITY ANALYSlS To show the dependence of Keff on fuel and storage cells parameters as re-quested by the NRC ', the variation of the Keff with respect to the following parameters was developed using the KENO computer code:
1. Fuel enrichment, with a 0.50 w/o U delta about the nominal case enrichment.
2. Center-to-center spacing of storage cells, with a half inch delta about the nominal case center-to-center spacing.
3. Poison loading, with a 0.01 gm-B /cm2 delta about the nominal case poison loading, Results of the sensitivity analysis for the Region 1 storage cells are shown in Figure 6 on page 30 through Figure 8 on page 32 for three of four storage.
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Ig Criticality Analysis of Region 1 Spent Fuel Racks
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4.0 CRITICALITY ANALYSIS OF REGION 2 SPENT FUEL RACKS This section develops and describes the analytical techniques and models em-ployed to perform the criticality analyses for storage of spent fuel in Region 2 of the Donald C Cook Nuclear Plant spent fuel pool.
4.1 REACTIVITY EQUIVALENCING i Spent fuel storage, in the Region 2 spent fuel storage racks, is achievable by means of the concept of reactivity equivalencing.
equivalencing is predicated upon the reactivity The concept of reactivity decrease associated with fuel depletion. A series of reactivity calculations are performed to generate a set t of enrichment-'fuel assembly discharge burnup ordered pairs which all yield the t equivalent Keff when the fuel is stored in the Region 2 racks.
Figure 9 on page 33 shows the constant Ken contour generated for the Donald C Cook Nuclear Plant Region 2 racks. Note in Figure 9 on page 33 the endpoint i at 0 MWD/MTU where the enrichment is 3.95 w/o and at 5,550 MWD/MTU where the enrichment is 4.95 w/o. The interpretation of the endpoint data is as fol-lows: the reactivity of the Region 2 racks containing fuel at 5,550 MWD/MTU burnup which had an initial nominal enrichment of 4.95 w/o is equivalent to the reactivity of the Region 2 racks containing fresh fuel having an initial nominal enrichment of 3.95 w/o. It is important to recognize that the curve in Figure 9 on page 33 is based on a constant rack reactivity for that region and not on a constant fuel assembly reactivity.
4.2 REACTIVITY CALCULATIONS The maximum Ken for storage of spent fuel in Region 2 is determined using. the methods described in Section 2. Figure 9 on page 33 represents combinations of fuel enrichment and discharge burnup yielding the same rack multiplication factor (Ke~r) as the. enrichment of 3.95 w/o U at zero burnup. This curve was obtained by first calculating the equivalent reactivity points using PHOENIX and then normalizing the points to the KENO calculation for fresh fuel with a nom-inal.enrichment of 3.95 w/o U Criticality Analysis of Region 2 Spent Fuel Racks 10
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The following assumptions were used to develop the nominal case KENO model for the Region 2 storage of spent fuel:
1. The Westinghouse 17x17 OFA fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable absorber in the fuel rods or any natural enrichment axial blankets (See Table 1 on page 19 for fuel parameters). Evaluation. of the Westinghouse 15x15 and 17x17 fuel assemblies shows that the 17x17 OFA assembly is the most reactive fuel type when all assemblies have the same enrichment. Therefore, only the Westinghouse 17x17 OFA fuel assembly was analyzed.
2. All fuel rods contain uranium dioxide at an 8nrichment of 3.95 w/o U over the finite 144 inch length of each rod. The fuel pellets are assumed to be at 96'Vo of theoretical density, and no credit is taken for dishing or chamfering. If nominal theoretical density and pellet parameters were used, the resultant enrichment limit would be 4.04 w/o U . Therefore, the 3.95 w/o U enrichment limit can be considered a nominal enrichment limit since the conservative assumptions employed in the pellet modelling bound the standard 0.05 w/o enrichment tolerance.
3. No credit is taken for any U or U in the fuel, nor is any credit taken for the build up of fission product poison material.
4. The moderator is pure water at a temperature of 68'F. A conservative value of 1.0 gm/cm is used for the density of water.
5. No credit is taken for any spacer grids or spacer sleeves.
6. Fuel assemblies are loaded into three of every four cells in a checkerboard pattern in the storage cells as shown in Figure 2 on page 26.
7. The array is infinite in lateral extent and finite in axial extent which allows neutron leakage from only the axial direction.
8. The minimum poison material loading of 0.02 grams B per square centi-meter is used throughout the array.
The KENO calculation for the nominal case resulted in a Kelf of 0.9141 with a 95 percent probability/95 percent confidence level 'uncertainty of +0.0049. The nominal case result can be compared to the worst case result to determine the relative impact of applying the worst case assumptions. The nominal case is also used as the center point for the sensitivity analyses discussed in Section 4.3.
The maximum Kaff under normal conditions was determined with a "worst case" Igj KENO model, in the same manner as for the Region 1 storage racks (see Section 3). For the Region 2 racks, the cell center to center spacings are reduced from the nominal value to their minimum value. Thus, the "worst case" KENO model Criticality Analysis of Region 2 Spent Fuel Racks
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of the Region 2 storage racks contains minimum cell center to center spacings with symmetrically placed fuel assemblies. The uncertainty associated with the reactivity equivalence methodology was included in the development of the burnup requirements. This uncertainty was discussed in Section 2.2.
Based on the analysis described above, the following equation is used to de-velop the maximum Keft for the storage of spent fuel in the Donald C Cook Nuclear Plant Region 2 spent fuel storage racks:
2 2 Ketf = Kworst + Bmethod + Bpart + t/ f (kS) worst + (ks) method ]
where:
Kworat worst case KENO Keff that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal B method method bias determined from benchmark critical comparisons Bpart bias to account for poison particle self~hielding. This standard term accounts for the increased neutron transmission through the poison plate due to the inherent effects of poison particle selfmhielding, and has been analytically determined for poison plates similar to those used in this analysis.
'sworst 95/95 uncertainty in the worst case KENO Keff ksmethod = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
The maximum Keff for Region 2 for this configuration is less than 0.95, including all uncertainties at a 95/95 probability/confidence level. Therefore, the accept-ance criteria for criticality are met for storage of spent fuel at an equivalent fresh fuel nominal enrichment of 3.95 w/o U Criticality Analysis of Region 2 Spent Fuel Racks 12
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4.3 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in Ken of the rack. Ex-t amples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies which precludes interaction).
However, accidents can be postulated which would increase reactivity (i.e.,
misloading an assembly with a burnup and enrichment combination outside of the acceptable area in Figure 9 on page 33, or dropping a fuel assembly be-tween the rack and pool wall). For these accident conditions, the double con-tingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure pro-tection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The presence of approximately 2000 ppm boron in the pool water will decrease reactivity by, about 0.25 hK. Thus, for postulated accidents, should there be a reactivity increase, Ken would be less than or equal to 0.95 due to the effect of the dissolved boron. Since the Donald C Cook Nuclear Plant SFP will be maintained at a boron concentration of 2400 ppm, additional margin will exist to the 0.95 limit.
4.4 SENSITIVITY ANALYSIS To show the dependence of Ken on fuel and storage cells parameters as re-quested by the NRC, the variation of the Keff with respect to the following parameters was developed using the PHOENIX computer code:
1. Fuel enrichment, with a 0.50 w/o U delta about the nominal case enrichment.
2. Center-tormenter spacing of storage cells, with a half inch delta about the nominal .case center-to-center spacing.
3. Poison loading, with a 0.01 gm-B /cm2 delta about the nominal case poison loading.
Results of the sensitivity analysis for the Region 2 storage cells are shown in Figure 10 on page 34 through Figure 12 on page 36 for spent fuel occupying every'cell in the Region 2 fuel racks.
Criticality Analysis of Region 2 Spent Fuel Racks 13
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5.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS This section describes the analytical techniques and models employed to per-form the criticality analysis for storage- of fresh fuel in the Donald C Cook Nuclear Plant New Fuel Storage Vault (NFSV).
Since the fresh fuel racks are maintained in a dry condition, the criticality analysis will show that the rack Koff is less than 0.95 for the full water density condition and less than 0.98 for the low water density (optimum moderation) conditions. The criticality methodology employed in this analysis is discussed in Section 2 of this report.
The following assumptions were used to develop the nominal case KENO model for the storage of fresh fuel in the Donald C Cook Nuclear Plant NFSV under full density and low density optimum moderation conditions:
1.~ The fuel assembly contains the highest enrichment authorized, is at its most t 'eactive point in life, and no credit is taken for any burnable poison in the fuel rods or any natural enrichment axial blankets. ~
I 2. All fuel rods contain uranium dioxide at an enrichment of 4.55 w/o U over the infinite length of each.rod. The fuel pellets are assumed to be at 9696 of theoretical density, and no credit is taken for dishing or chamfering, If nominal theoretical density and pellet parameters were used, the resultant enrichment limit would be 4.65 w/o U Therefore, the 4.55 w/o U enrichment limit can be considered a nominal enrichment limit since the conservative assumptions employed in the pellet modelling bound the standard 0.05 w/o enrichment tolerance.
3. No credit is taken for any U or U 'n the fuel, nor is any credit taken for the build up of fission product poison material.
4. No. credit is taken for any spacer grids or spacer sleeves.
Criticality Analysis of Fresh Fuel Racks 14
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5.1 FULL DENSITY MODERATlON ANALYSIS In the nominal case KENO model for the full density moderation analysis, the moderator is pure water at a temperature of 68 F. A conservative value of 1.0 gm/cm is used for the density of water. The fuel array is infinite in lateral and axial extent which precludes any neutron leakage from the array. This 2D single cell modelling technique is conservative, however, the overall reactivity effect of neutron leakage from the array under full moderator density conditions is small. Calculations for the Donald C Cook Nuclear Plant NFSV array show total leakage effects to be worth only 0.005 hK.. Fuel rack calculations have shown that the Westinghouse 17x17 OFA fuel assemblies are more reactive than the other fuel types when all fuel assemblies have the same U enrichment.
Thus, only the Westinghouse 17x17 OFA fuel assembly was analyzed, The maximum Keff under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack Ksn. Since the Donald C Cook Nuclear Plant NFSV rack structure consists of part length angle irons, all of the structural steel was conservatively left out of the model. Thus, the most conservative, or "worst case", KENO model of the fresh fuel storage racks contains no structural steel with symmetrically placed fuel assemblies.
Based on the analysis described above, the following equation is used to de-velop the maximum Ksff for the Donald C Cook Nuclear Plant New Fuel Storage Vault:
Kofr= Kworst + Bmsthod + /'[(kS)2worst 2
+ (ks) method ]
where:
Kworst worst case KENO Keff that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal B method method bias determined from benchmark critical comparisons ksworst 95/95 uncertainty in the worst case KENO Ksn Criticality Analysis of Fresh Fuel Racks 15
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ksmethod = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
Ke<< = 0.9324 + 0.0083 + t/ [(0.0086) + (0.0018) ] = 0.9495 Since Ke<< is less than 0.95 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is met.
5.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS In the low density optimum moderation analysis, the fuel array is finite in the radial and axial extent. The nominal model described above is used in KENO except that the concrete walls and floor are explicitly modelled as shown in Figure 4 on page 28 and Figure 5 on page 29. The Westinghouse 17x17 STD fuel assembly was analyzed in the model (See Table 1 on page 19 for fuel parameters). Calculations have shown that the 17x17 STD fuel assembly is more reactive than other fuel assemblies under low moderator density conditions.
Analysis of the Donald C Cook Nuclear Plant racks has shown that the maximum rack Ke<<under low density moderation conditions occurs at 0.045 gm/cm water density. The KENO calculation of the Donald C Cook Nuclear Plant NFSV at 0.045 gm/cm water density resulted in a peak Ke<<of 0.8817 with a 95 percent probability and 95 percent confidence level uncertainty of +0.0072. Figure 13 on page 37 shows the NFSV reactivity as a function of the water density.
Based on the analysis described above, the following equation is used to de-velop the maximum Ke<< for the Donald C Cook Nuclear Plant fresh fuel storage racks under low water density optimum moderation conditions:
2 2 Ke<< = Kbsse + Bmethod + t/ [(kS) bose + (kS) method ]
where:
Kbsse maximum KENO Ke<<with low density optimum moderation Bmethod method bias determined from benchmark critical comparisons ksbese 95/95 uncertainty in the base case KENO Keff Criticality An'alysis of Fresh Fuel Racks 16
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ksmethod = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
Kafka = 0.8817 + 0.0083 + f [(0.0072) + (0.0018) ] = 0.8974 Since Ko<< is less than 0.98 including uncertainties at a 95/95 probability/confidence level, the acceptance criteria 'for criticality is met.
I Criticality Analysis of Fresh Fuel Racks 17
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6.0
SUMMARY
OF CRITICALITY RESULTS The acceptance criteria for criticality requires the effective neutron multipli-cation factor, Ken, to be less than or equal to 0,95, including uncertainties, under all conditions for the storage of fuel assemblies in the Spent Fuel Pool (SFP).
For the storage of fuel assemblies in the New Fuel Storage Vault (NFSV), the Ken must be less than or equal to 0.95, including uncertainties, under flooded conditions, and less than or equal to 0.98, including uncertainties, under optimum moderation conditions.
This report shows that the acceptance criteria for criticality is met for the Donald C Cook Nuclear Plant Spent Fuel Pool (SFP) and New Fuel Storage Vault (NFSV) for the the storage of Westinghouse 15x15 and 17x17 STD, OFA and VANTAGE 5 fuel assemblies with the following nominal enrichment limits:.
SFP Region 1 5 4.95 w/o U SFP Region 2 4.95 w/o U, with burnup restrictions given by Figure 9 on page 33 NFSV 4.55 w/o U The analytical methods employed herein conform with ANSI N18.2-1973, "Nu-clear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System; ANSI 57.2-1983, "Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, "Validation of Calculational Methods for Nuclear Criticality Safety"; NRC Standard Review Plan, Section 9.1.2, "Spent Fuel Storage"; and ANSI 57.3-1983, "Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."
Summary of Criticality Results 18
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Table 1 ~ Fuel Parameters Employed in Criticality Analysis Parameter W 'l5x'15 W 'l7x17 W 17x17.
STD & OFA OFA & V5 STD Number of Fuel Rods per Assembly 204 264 264 Rod Zirc-4 Clad O.D. (inch) 0.422 0.360 0.374 Clad Thickness (inch) 0.0243 0.0225 0.0225 Fuel Pellet O.D.(inch) 0.3659 0.3088 0.3225 Fuel Pellet Density (0 of Theoretical) 96 96 96 Fuel Pellet Dishing Factor 0.0 0.0 0.0 Rod- Pitch (inch) 0.563 0.496 0.496 Number of Zirc-4 Guide Tubes 20 24 24 Guide Tube O.D. (inch) 0 533 0.474 0.482 Guide Tube Thickness (inch) 0.017 0.016 0.016 Number of Instrument Tubes Instrument Tube O.D. (inch) 0 533 0.474 0.484 Instrument Tube Thickness (inch) 0.017 0.016 0.016 19
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Table 2. Benchmark Critical Experiments t.5,6]
General Enrichment Separating Soluble Oescr Ipt i an w/o U235 Ref lector Hater ial Boron ppm Kef f
1. U02 rod lattice 2.46 water 0 0.9857 +/- .0028 i 2.
3.
4.
5.
6.
7.
U02 U02 U02 U02 U02 U02 rod rod rad rod rod rod lattice lattice lattice lattice lattice latt ice 2.46 2.46 2.46 2.46 2.46 2.46 water water water water water water water'ater water'4C 84C B4C 84C pins pins pins pins 1037 764 0
0 0
0 0.9906 +/- .OO18 0.9896 i/- .0015
0. 9914 +/- .OO25 0.9891 +/- .0026 0.9955 +/- .0020 0.9889 +/- .0027
8. U02 rod lattice 2.46 water B4C pins 0 0.9983 +/- .0025 E 9. U02 rod lattice 2.46 water water 0 0.9931 +/- .0028
10. U02 rod lattice 2.46 water water 143 0.9928 +/- .0025
11. U02-rod lattIce 2.46 water stainless steel 514 0.9967 +/- .0020 i 12.
13.
14.
15.
16.
U02 rod U02 rad U02 rod U02 rod U02 rod lattice lattice lattice lattice lattice 2.46 2.46 2.46 2.46 2.46 water
.water water water water stainless steel borated aluminum borated aluminum borated aluminum boreted aluminum 217 15 92 395 121 0.9943 +/- . Q019 0.98S2 +/- .OO23 0.9884 +/- .OO23 0.9832 +/- .0021 O.'9848 +/- 0024 i lattice 0.9895 +/- .0020
~
17. U02 rod 2.46 water borated aluminum 487
18. U02 rod lattice 2.46 water barated aluminum 197 0.9885 +/- .OO22
19. U02 rod lattice 2.46 water bor ated aluminum 634 0.9921 +/- .0019
20. U02 rad lattice 2.46 water bor ated aluminum 320 0.9920 +/- .0020
21. U02 rod lattice 2.46 water borated aluminum 72 0.9939 +/- .0020
22. U metal cylinders 93.2 bare air 0 0.9905 +/- .0020
23. U metal cylinders 93.2 bare air 0 0.9976 +/- .0020
24. U metal cylinders 93.2 bare air 0 O.S947 +/- .0025
25. U metal cylinders 93.2 bare air 0 0.9928 +/- .0019
26. U metal cylinders 93.2 bare air 0 0.9922 +/- .0026
27. U metal cylinders 93.2 bare a Ir 0 0.9950 +/- .OO27
28. U metal cylinders 93.2 bare plexiglass 0 0.9941 +/- .0030
29. U metal cyl inder s 93.2 par af f in plexiglass 0 0.9928 +/- . 004 I
30. U metal cylinders 93.2 bar e plexiglass 0 0.9968 +/- .0018
31. U metal cylinders 93.2 par af f In plexiglass 0 1.0042 +/- .0019
32. U metal cyl inder s 93.2 par af f In plexiglass 0 0.9963 +/- .0030
33. U metal cylinders 93.2 par af f in plexiglass 0 0. 9919 4/- .OO32 L
20
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Table 3. Comparison of PHOENIX Isotopics Predictions to Yankee Core 5 Measurements Quantity (Atom Ratio) % Difference U235/U -0.67 U236/U -0.28 U238/U -0.03 PU239/U +3.27 PU240/U +3.63 PU241/U "7.01 PU242/U -'0.20 PU239/U238 +3.24 Mass(PU/U) +1.41 FI SSHU/TOT%V %.02 21
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Table 4. Benchmark Critical Experiments PHOENIX Comparison t Description of Experiments Number of Experiments PHOENIX Koff Using Experiment Buckling s UO2 Al clad 14 0.9947 SS clad 19 O.SS44 Borated H20 0.9940 Subtotal 40 0.9944 U-Metal AI clad 41 1.0012 TOTAL 81 0;9978 i
22
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Table S. Data for U Metal and U02 Critical Experiments (Part 1 of 2)
Fuel Pellet Clad Clad Lattice Case Cell A/0 H20/U Oens 1 ty 01 Material 00 Thickness Pitch Boron U-235 Ratio arnetet'cM)
Number Type (G/CC) Clad (cM) (cM) (cM) PPM Hexa 1.328 3.02 7.53 1. S265 Aluminum 1 ~ 6916 .07110 2.2050 0.0 Hexa 1. 328 3.95 7.53 1. 5265 -
A 1umi num 1. 6916 .07110 2.3590 0.0 Hexa 1. 328 4.95 7.53 1.5265 Aluminum 1,6916 .07110 2: 5120 0.0 Hexa 1. 328 3.92 7.52 .9855 Aluminum 1. 1506 .07110 1. 5580 0.0 i 5 6
7 8
9 10 Hexa Hexa Hexa Hexa
1. 328
1. 328
1. 328
1. 328 Square 2:734 Square 2.734 4.89 2.88 3.58 4.83
2. 18 2.92 7.52
10. 53
10. 53 10.53
10. 18
10. 18
.9855
.9728
.9728
.9728
.7620
.7620 Aluminum A 1 umi num A 1um1nue A 1 umi num 55-304 55-304 1 ~ 1506
1. 1506
1. 1506
1. 1506
.8594
.BS94
.07110
.07110
.07110
.07110
.04085
.04085 1.6520
1. 5580 1.6520 1.8060 1.0287
1. 1049 0.0 0.0 0.0 0.0 0.0 0.0 11 Square 2.734 3.86 10. 18 .7620 55-304 .8594 .04085 1. 1938 0.0 t~ 12 13 Squar e 2.734 Square 2.734 7.02 8.49
10. 18
10. 18
~ 7620
.7620 55 304 55-304
.8594
.8594
.04085
.04085 1.4554 1.5621 0.0 0.0 i 14 15 16 17 18 19 Square Square Square Square Square Square 2.734 2.734 2.734 3.745 3.745 3.745 10.38 2.50 4.51
2. 50 4.51 4.51
10. 18
10. 18
10. 18
10. 27
10. 37
10. 37
~ 7620
.7620
.7620
.7544
.7544
.7544 55 304 55-304 55-304 SS-304 55-304 55-304
~
.8594
.B594
.8594
.8600
.8600
.8600
.04085
.04085
.04085
.04060
.04060
.04060
1. 6891 1.0617
1. 2522 1.. 0617
1. 2522 1.2522 0.0 0.0 0.0 0.0 0.0 0.0 i 20 21 22 23 24 25 Square Square Square Squar e Square Square 3.745 3.745 3.745 3.745 3.745 4.069 4.51 4.51 4.51 4.51 4.51 2.55
10. 37
10. 37
10. 37
10. 37
10. 37 9.46
.7544
.7544
.7544
.7544
.7544
1. 1278 SS-304 55-304 55-304 SS-304 55-304 55-304 inurn
.8600
.8600
.8600
.8600
.8600 1.2090
.04060
.04060
.04060
.04060
.04060
.04060
1. 2522 1.2522
1. 2522
1. 2522 1.2522 1.5113 456.0 709.0 1260.0 1334. 0 1477.0 0.0 26 Square 4.069 2.55 9.46 1 . 1278 55 304 1. 2090 :o4o6o 1. 5113 3392.0 27 Square 4.069 2. 14 9.46 1. 1278 55-304 1. 2090 .04060 1. 4500 0.0 28 Square 2.490 2.84 10.24 1.0297 A 1 um1num 1. 2060 . 08130 1.5113 0.0 29 Square 3.037 2.64 9.28 1. 1268 55-304 1. 1701 .07163 1. 5550 0.0 30 Square 3.037 8. 16 9.28 1. 1268 55-304 1. 2701 .07163 2. 1980 0.0 31 Square 4.069 2. 59 9.45 1. 1268 SS-304 1. 2701 . 07163 1.5550 0.0 32 Square 4.069 3.53 9.45 1. 1268 55-304 1. 2701 .07163 1. 6840 0.0 33 Square 4.069 8.02 9.45 1. 1268 55-304 1. 2701 .07163 2. 1980 0.0 34 Square 4.069 9.90 9.45 1. 1268 SS-304 1. 2701 . 07 163 2. 3810 0.0 35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 . 08130, 1.5113 1677.0 36 Hexa 2.096 2.06 10.38 1. 5240 Aluminum 1. 6916 .07112, 2. 1737 0.0 37 Hexa 2.096 3.09 10.38 1. 5240 A 1uminura 1 ~ 6916 .07112 2.4052 0.0 38 Hexa 2.096 4. 12 10. 38 1.5240 A 1 ura 1. 6916 .07112 2.6162 0.0 39 Hexa 2.096 6. 14 10. 38 1.5240 Aluminue 1. 6916 .07112 2.9891 0.0 40 Hexa 2.096 8. 20 10. 38 1. 5240 A 1um1num 1.6916 .07112 3.3255 0.0 41 Hexa 1.307 1.01 18. 90 1.5240 Aluminue 1.6916 ~ 07112 2. 1742 0.0 42 Hexa 1.307 1.51 18. 90 1. 5240 A 1umi nura 1. 6916 .07112 2.4054 0.0 43 Hexa 1. 307 2. 02 18. 90 1. 5240 Aluminum 1. 6916 .07112 2. 6162 0.0 23
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Table 5. Data for U Metal and UO2 Critical Experiments (Part 2 Of 2)
Fuel Pe et 1 1 Clad Clad Lattice Case Cell A/0 H20/U Oens 1 ty Diameter Material 00 Thickness Pitch . Boron Number Type U-235 Ratio (6/CC) (CM) Clad (CM) (CM) (CM) PPM 44 Hexa 1. 307 3.01 18. 90 1.5240 A 1uminum 1.6916 .07112 2.9896 0.0 45 Hexa 1. 307 4.02 18. 90 1. 5240 Aluminuminurn 1. 6916 .07112 3.3249 0.0 46 Hexa 1. 160 1.01 18. 90 1. 5240 A lum1num '1. 6916 .07112 2. 1742 0.0 47 Hexa 1. 160 1.51 18. 90 1. 5240 A 1 um inurn 1.6916 .07112 2.4054 0.0 48 Hexa 1. 160 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 49 Hexa 1. 160 3.01 18. 90 1. 5240 Aluminum 1.6916 .07112 2.9896 0.0 50 Hexa 1. 160 4.02 18. 90 1. 5240 A 1um i num 1.6916 .07112 3.3249 0.0 51 Hexa 1.040 1.01 18. 90 1.5240 A 1um i num 1.6916 .07112 2. 1742 0.0 52 Hexa 1. 040 1.51 18. 90 1. 5240 A 1 ura 1 num 1.6916 .07112 2.4054 0.0 S3 Hexa 1.040 2.02 18. 90 1. 5240 A 1 um inurn 1. 6916 .07112 2. 6162 0.0 54 Hexa 1.040 3.01 18. 90 1.5240 A 1 urn 1.6916 .07112 2.9896 0.0 55 Hexa 1. 040 4.02 18. 90 1.5240 A 1um1num 1. 6916 .07112 3.3249 0.0 56 Hexa 1. 307 1,00 18. 90 .9830 A I um 1 num 1. 1506 .07112 1. 4412 0.0 57 Hexa 1.307 1.52 18. 90 .9830 Aluminum 1. 1508 .07112 1. 5926 0.0 58 Hexa 1. 307 2.02 18. 90 .9830 A 1uminum 1.1506 .07112 1. 7247 0.0 S9 Hexa 1. 307 3.02 18. 90 .9830 Aluminum 1. 1508 .07112 1. 9609 0.0 60 Hexa 1". 307 4.02 18. 90 .9830 A 1um1num 1. 1506 . 07112 2. 1742 0.0 61 Hexa 1. 160 1.52 18. 90 .9830 A 1um1num 1, 1508 .07112 1.5926 0.0 62 Hexa 1. 160 2.02 18. 90 .9830 Aluminum 1. 1506 .07112 1. 7247 0.0 63 Hexa 1. 160 3.02 18. 90 .9830 A 1 um1num 1. 1506 .07112 1. 9609 0.0 64 Hexa 1. 160 4.02 18. 90 .9830 A 1um 1 num 1. 1506 .07112 2. 1742 0.0 65 Hexa 1. 160 1.00 18. 90 .9830 Aluminum 1 . 1506 .07112 1 . 4412 0.0 66 Hexa 1. 160 1.52 18. 90 .9830 Aluminum 1 . 1506 .07112 1.5926 0.0 67 Hexa 1. 160 2.02 18. 90 .9830 A 1 urn1num 1. 1506 .07112 1. 7247 0.0 68 Hexa 1. 160 3.02 18. 90 .9830 A 1umi num 1. 1506 .07112 1. 9609 0.0 69 Hexa 1. 160 4.02 18. 90 .9830 A 1 um1num 1. 1506 .07112 2. 1742 0.0 70 Hexa 1.040 1. 33 18. 90 19.050 Aluminum 2.0574 .07620 2.8687 0.0 71 Hexa 1.040 1. 58 18. 90 19.050 A 1 umi num 2.0574 .07620 3.0086 0.0 72 Hexa 1.040 1.83 18.90 19. 050 A 1 um1num 2.0574 .07620 3. 1425 0.0 73 74 Hexa Hexa 1.040
1. 040 2 '3 2.83 18.90 18.90 19.050
19. 050 A 1umi num Aluminum 2.0574 2.0574 .07620
.07620 3.3942 3.6284 0.0 0.0 75 Hexa 1.040 3.83 18 ~ 90 19 050
~ A 1 um1num 2.0574 .07620 4.0566 0.0 76 Hexa 1. 310 2.02 18. 88 1. 5240 A umi num 1 1.6916 .07112 2. 6160 0.0 77 Hexa 1. 310 3.01 18.88 1. 5240 A 1 um1num 1. 6916 .07112 2. 9900 0.0 2.02 Aluminum 1.6916 .07112 2. 6160 0.0 78 79 80
'exa Hexa Hexa
1. 159
1. 159 1.312 3.01 2.03 18.88 18 88
~
18.88
1. 5240
1. 5240
.9830 Aluminum Aluminum 1.6916
1. 1506
.07112
.07112
2. 9900
1. 7250 0.0 0.0 81 Hexa 1.312 3.02 18.88 .9830 Aluminum 1. 1506 .07112 1. 9610 0.0 24
I I
! SEE DETAILA 8.884" 0.196" 1.224" CELL CENTER TO CENIER (10.5")
.03" Stainless steel
~ .Ol'il.
~.071" BoC-AL
.Ol'fiL BORAL 075" Stainless Steel INSIDE OF CELL DETAILA Figure 1. Donald C Cook Nuclear Plant Spent Fuel Pool Storage Cell Nominal Dimensions 25
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Empty Cell Figure 2. Donald C Cook Nuclear Plant SFP Region 'l Three of Four Fuel As-sembly Loading Schematic 26
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Region 1 to 2 Boundary I Regton 1 i
i0 L Re lon2 Empty Cell Fuel Cell Figure 3. Donald G Cook Nuclear Plant Schematic for SFP Interface Boundary Between Regions 1 and 2 27
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40 0 RC4.
~ OO OCR I~
~P 4.0 Rti ~ ~ IS 4,4 454, i,49 RSR>>I
~ .4% .0) IS 50.4t 5~0~ IS 40 i)4
~ CR Ot)AII'II IO'0 OCAI~ S QiI~ RORA~
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RNACIS 9.0
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4 SPACRS0 IPP 9>> +
~I 45<R>>+ RCR 5" 0-ROW, io 0 0 ~PA N
)IANR OR ~ I+~ 0 Figure 4. Donald C Cook Nuclear Plant Fresh Fuel Rack Radial Layout 28
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~ <<1 1<<o 4<<
lia )xiii Ill) lilt ill) Ill) Ih) IIII IE IIII illl WWneo WW<< ~ <<t <<s '%%%
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.91 I I I
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i .90 i
i .89 8
4.0 4.5 5.0 5.5 6.0 U ENRICHMENT (W/0)
BORAL HELD AT .02 GM B10/CM2 CENTER TO CENTER HELD AT 10.5" Figure 6. Sensitivity of Keff to Enrichment in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading.
30
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I
.96
.95
.93
.92 4
UJ hC
.91
.90 I
L
.89 I
L
.88 I
L 9.50 10.0 10.5 11.0 11.5 12.0 CENTER-TO-CENTER SPACING (INCMES)
BORAL HELD AT .02 GM B10/CM2 ENRICHMENT HELD AT 4.95 W/0 Figure 7. Sensitivity of Kerf to Center-to-Center Spacing in the Donald C Cook Nuclear Plant SFP Region 1 Storage Area with Three of Four Loading 31
I I I I I 1 I I
I
.93
.92 h.91 I
I I
.90 I I I I J L
.89 I
I L
00 .01 .02 .03 .04 POSION LOAO {GM 810(CM2)
CENTER TO CENTER HELD AT 10.5" ENRICHMENT HELD AT 4.95 W/0 Figure 8. Sensitivity of Ken to B Loading in the Donald C Cook Nuclear Plant SFP Region 'l Storage Area with Three of Four Loading 32
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i t
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I I I I I I I D
I T l I T I I CI I ACCEPTABLE I 4
Q D
t K IX Kl w 3 t 0 i CD U)
CI CQ I I
LJ I
M I
I NOT ACCEPTABLE I I I I I I J 1 L
'.5 I
4.1 4.3 4.7 4.9 5.1 U-235 ENRICHMENT (W/0)
Figure 9. Oonald C Cook Nuclear Plant SFP Region 2 Fuel Assembly Minimum Burnup vs. Initial U Enrichment Curve 33
l I
I
~l l
.94 I I I I I I I l l
.93 I
I I
.92 I
I I
t .9I
'I I
I i
i .90 I
I L J I
I I
r I
.89 I I I I L L I
I 3.6 3.8 4.0 4.2 4.4 U-235 ENRICHMENT (W/0)
BORAL HELD AT .02 GM 810/CM2 TO CENTER HELD AT 10.5" 'ENTER Figure 10. Sensitivity of Ko<<'to Enrichment in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 34
I
.96 I I I I
.95
.94 I I I I I I I I T' I T' I I
.93 I I I L I I I I QJ 92 I I I I l r g .91 I
I
.90
.89 80 10.0 10. 2 10. 4 10. 6 10. 8 11.0 11.2 CENTER-TO-CENTER SPACING (INCHES)
BORAL HELD AT .02 GM B10/CM2 ENRICHMENT HELD AT 3.95 W/0 Figure 11. Sensitivity of Ken to Center-to-Center Spacing in the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 35
I
.935
.930
.925
.920
,915
.910
.905 900 008 .010 .012 .014 .016 .018 .020 .022 .Q24 .026 .028 030 .032 POISON LOAD (GM B10/CM~)
CENTER TO CENTER HELD AT 10.5" ENRICHMENT HELD AT 3.95 W/0 Figure l2. Sensitivity of Ke(( to 8 l.oadlng ln the Donald C Cook Nuclear Plant SFP Region 2 Storage Area 1
36
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I
.89 I
1 I
.88 I
I
.87 I I
I I
.86
.85
.84
.83 I
I I
.82 t I J J I I
.81 I I I J. L J I I 8
.01 .02 .03 .04 .05 .06 .07 .08 H20 DENSITY (G/CC)
Figure 18. Sensitivity of KeH to Water Density in the Donald C Cook Nuclear Plant New Fuel Storage Vault 37
I
~I I
BIBLIOGRAPHY Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent Fuel Storage and Handling ApplicationsApril 14, 1978.
2. W. E. Ford I I I, CSRL-V: Processed ENDFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNL/CSD/TM-160, June 1982.
N. M. Greene, AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDFIB, ORNL/TM-3706, March 1976.
L. M. Petrie and N. F. Cross, KENO IV An Improved Monte Carlo Criticality Program, ORNL-4938, November 1975.
M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, July 1979.
J. T. Thomas, Critical Three-Dimensional Arrays of U(S3.2I Metal Cylinders, Nuclear Science and Engineering, Volume 52, pages 350-359, 1973.
7. D. E. Mueller, W. A. Boyd, and M. W. Fecteau (Westinghouse NFD), Qualification of KENO Calculations with ENDFIB-V Cross Sections, American Nuclear Society Transactions, Volume 56, pages 321-323, June 1986.
8. A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors, WCAP-10106, June 1982.
9. Askew, J. RFayers, F. J., and Kemshell, P. B., A General Description of the Lattice Code WIMS, Journal of British Nuclear Energy Society, 5, pp.
564-584, 1966.
10. England, T. R., CINDER - A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.
11. Melehan, J. B., Yankee Core Evaluation Program Final Report, WCAP-3017-6094, January 1971.
Bibliography
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t 0)
ENCLOSURE 1
,AEP (COOK) RESPONSE TO GENERIC LETTER 89-21
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ML17334B357

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1.0 INTRODUCTION

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