ML023520437

Submittal of Oyster Creek Generating Station (Ocgs) Inservice Testing Plan

ML023520437
Person / Time
Site:	Oyster Creek
Issue date:	12/09/2002
From:	Gallagher M AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2130-02-20342
Download: ML023520437 (156)

Text

Amer Genl S

AmerGen Energy Company, LLC wwwexeloncorp corn An Exelon/BritIsh Energy Company 200 Exelon Way Suite 345 Kennett Square, PA 19348 1 0CFR50.55a 2130-02-20342 December 9, 2002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Submittal of Oyster Creek Generating Station (OCGS) Inservice Testing Plan

Dear Sir/Madam:

Attached for your information is a copy of the Oyster Creek Generating Station (OCGS)

Inservice Testing Program Plan. This plan is being submitted for your information in accordance with Section ISTA 1.4 of the ASME OM Code-1 995, Code for Operation and Maintenance of Nuclear Power Plants, including the OMa-1 996 Addenda. Based on a start date of October 14, 2002, the OCGS IST Program is required by 10CFR50.55a(f)(4) to comply with the requirements of the ASME OM Code-1 995, Code for Operation and Maintenance of Nuclear Power Plants, including the OMa-1 996 Addenda.

If you have any questions or require additional information, please do not hesitate to contact us.

Very truly yurs, Michael P. Gallagher Director, Licensing & Regulatory Affairs Mid-Atlantic Regional Operating Group Attachment - Oyster Creek Generating Station (OCGS) Inservice Testing Plan cc: H. J. Miller, Administrator, USNRC, Region I (w/attachment)

R. J. Summers, USNRC Senior Resident Inspector, LGS (w/attachment)

P. Tam, Senior Project Manager, USNRC (w/attachment)

File No. 02073 A-){7

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Fourth Ten-Year Interval

September 19, 2002 To: Rob Francis From: Joseph S. Shebby, ANTI

Subject:

A.S.M.E. Pump and Valve Up-date Article IWA-2 110, Duties of the Inspector, provides direction for the Inspector to perform a detailed review of the inspection plan prior to each inspection interval. In addition, submit a report to the Owner documenting review of items listed in IWA-2110 (c).

The Oyster Creek Nuclear Generating Station Inservice Testing Program ( !ST ) for the fourth ten year interval is required by 10 CFR 50.55a to comply with the requirements of ASME/ANSI OM Code-1995, Code of Operation and Maintenance of Nuclear Power Plants, including Oma-1966 addenda. Pumps OM Part 6 and Valves Part 10.

The fourth 10 year interval will begin on October 14,2002 and conclude October 13, 2012. A review of this program plan has found the program to contain the IST component identification, type of testing and test frequency.

Relief request submitted for this program are existing, having been approved previously by the NRC.

It has been determined that the program plan meets the applicable Codes. Review of documentation of testing will be done on regular basis, to verify compliance with testing intervals.

Date: Sept. 19,2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Effective Date 10/14/02 Revision Log Revision Description Prepared:

1ST Program Engineer Revision 10:

Program revised in its entirety for fourth 10 -year interval to comply with ASME OM Code-1995 and OMa-1996 addenda. Reformatted per the guidelines of Corporate Procedure NES-MS-08.2 Approved:

Engr. Programs Supervisor Revision 10 October 14, 2002 Date "I4o-q.01 Date t*

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan TABLE OF CONTENTS SECTION

1.0 INTRODUCTION

1.1 Purpose 1.2 Scope 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan Description 2.2 Pump Plan Table Description 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan Description 3.2 Valve Plan Table Description 4.0 ATTACHMENTS

1.

Pump Relief Request Index

2.

Pump Relief Requests

3.

Valve Relief Request Index

4.

Valve Relief Requests

5.

Cold Shutdown Justification Index

6.

Cold Shutdown Justifications

7.

Refuel Outage Justification Index

8.

Refuel Outage Justifications

9.

Station Technical Position Index

10.

Station Technical Positions

11.

Corporate Technical Position Index

12.

Corporate Technical Positions

13.

Inservice Testing Pump Table Index

14.

Inservice Testing Pump Table

15.

Inservice Testing Valve Table Index

16.

Inservice Testing Valve Table

-3 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan

1.0 INTRODUCTION

1.1 Purpose The purpose of this Inservice Testing (IST) Program Plan is to provide a summary description of the Oyster Creek Nuclear Generating Station IST Program and to document compliance with the requirements of 10 CFR 50.55a(f) either directly or by reference.

1.2 Scope This Inservice Testing Program Plan identifies the components included in the Oyster Creek Nuclear Generating Station's Inservice Testing (IST) Program for the fourth ten-year IST interval. The Third Ten-Year Interval began on October 14, 1991, and will conclude (with 1-year extension) by October 13, 2002. The Fourth Ten-Year Interval will begin on October 14, 2002, and conclude on October 13, 2012.

Testing methods and frequencies are identified; Relief Requests, Cold Shutdown Justifications, Refuel Oufage Justifications, and technical positions regarding the application of certain requirements are also provided.

The Oyster Creek NGS IST Program for the fourth 10-year interval is required by 10 CFR 50.55a to comply with the requirements of the ASME OM Code-1995, Code for Operation and Maintenance of Nuclear Power Plants, including the OMa 1996 addenda, in assessing the operational readiness of those ASME Safety Class 1, 2 and 3 pumps and valves which are required to perform a specific function or functions in:

• shutting down the Reactor to the safe shutdown condition

• maintaining the Reactor in the safe shutdown condition

• mitigating the consequences of an accident Non-ASME Safety Class components which may be required to perform such functions are not required by 10 CFR 50.55a to be included in the IST Program.

They may be optionally included herein or may be otherwise tested in a manner which satisfactorily demonstrates their ability to perform their functions commensurate with their importance to safety per the applicable portions of 10 CFR 50, Appendices A or B.

Oyster Creek NGS is licensed with the Hot Standby condition as the safe shutdown condition. Therefore, this Program must include, as a minimum, all of those ASME Class 1, 2, and 3 pumps and valves which are required to shut down the Reactor to the Hot Standby condition, maintain the Hot Standby condition, or mitigate the consequences of an accident.

-4 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan The requirements of Subsections ISTA, ISTB and ISTC and, by reference, the requirements of Appendices I and II of the ASME OM Code-1995 with OMa-1996 addenda apply to this Program. Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan Description The Oyster Creek NGS Inservice Testing Program for Pumps meets the requirements of Subsections ISTA and ISTB of the ASME OM Code-1995 with OMa-1996 addenda, with the exception of those specific applications identified in the Relief Requests contained in Attachment 2.

2.2 Pump Plan Table Description The pumps included in the Oyster Creek NGS Inservice Testing Program are listed in Attachment 14. The information contained in that table identifies those pumps required to be tested to the requirements of the ASME OM Code, the parameters measured, associated Relief Requests and comments, and other applicable information. The column headings for the Pump Table are delineated below with an explanation of the content of each column.

Pump ID Description P&ID Sht Pump Type Driver Group The unique identification number for the pump, as designated on the System P&ID or Flow Diagram The descriptive name for the pump The Piping and Instrumentation Diagram or Flow Drawing on which the pump is shown The Sheet Number of the P&ID identified in the previous column An abbreviation used to designate the type of pump:

C Centrifugal PD Positive Displacement VLS Vertical Line Shaft The type of driver with which the pump is equipped.

A or B, as defined in Paragraph ISTB 1.3 of OMa-1996.

-6 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Lists each of the test parameters which are required to be measured for the specific pump. These include:

"* Speed, N (for variable speed pumps, only)

"* Differential Pressure, AP

"* Discharge Pressure, P (positive displacement pumps)

"* Flow Rate, Q

"* Vibration Displacement, Vd Velocity, Vv NOTE: All tests are performed at the frequencies required by Code unless specifically documented by a Relief Request.

Identifies the number of the Relief Request applicable to the specified test.

Any appropriate reference or explanatory information (e.g.,

technical positions, etc.) Revision 10 October 14, 2002 Test RR#

Comments

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan Description The Oyster Creek NGS Inservice Testing Program for Valves meets the requirements of Subsections ISTA and ISTC of the ASME OM Code-1995 with OMa-1996 addenda, with the exception of those specific applications identified in the Relief Requests contained in Attachment 4.

3.2 Valve Plan Table Description The valves included in the Oyster Creek NGS Inservice Testing Program are listed in Attachment 16. The information contained in that table identifies those valves required to be tested to the requirements of the ASME OM Code, the testing methods and frequency of testing, associated Relief Requests, comments, and other applicable information. The column headings for the Valve Table are delineated below with an explanation of the content of each column.

Valve ID The unique identification number for the valve, as designated on the System P&ID or Flow Diagram.

Description The descriptive name for the valve.

Size The nominal size of the valve in inches.

Vlv Type An abbreviation which designates the body style of the valve.

Abbreviations used are:

3W 3-Way 4W 4-Way BAL Ball BTF Butterfly CK Check DAM Damper DIA Diaphragm GA Gate GL Globe PLG Plug PLT Pilot PPT Poppet RPD Rupture Disk RV Relief SCK Stop-Check SHR Shear (SQUIB)

XFC Excess Flow Check

-8 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Actu Type An abbreviation which designates the type of actuator on the valve. Abbreviations used are:

AO Air Operator DF Dual Function (Self and Power)

EXP Explosive HO Hydraulic Operator M

Manual MO Motor Operator SA Self-Actuating SAP Self-Actuated Pilot SO Solenoid Operator P&ID The Piping and Instrumentation Diagram or Flow Drawing on which the valve is shown Sh/Crd The Sheet number and coordinates on the P&ID or Flow Diagram where the valve is shown.

CIs The ASME Safety Class (i.e., 1, 2 or 3) of the valve. Non ASME Safety Class valves are designated by "NC".

Positions Abbreviations used to identify the normal, fail, and safety Norm/Fail/Safe related positions for the valve. Abbreviations used are:

Al As Is C

Closed CKL Closed/Actuator Key Locked D

De-energized D/E De-energized or Energized E

Energized LC Locked Closed LO Locked Open LT Locked Throttled O

Open O/C Open or Closed OKL Open/Actuator Key Locked SYS System Condition Dependent T

Throttled Cat The ASME Code category or categories of the valve as defined in Paragraph ISTC 1.4 of the OMa-1996 Code.

-9 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Act/Pass "A" or "P", used to designate whether the valve is active or passive in fulfillment of its safety function. The terms "active valves" and "passive valves" are defined in Paragraph ISTC 1.3 of the OMa-1996 Code.

Testing Requirements

• Test A listing of abbreviations used to designate the types of testing which are required to be performed on the valve based on its category and functional requirements. Abbreviations used are:

CC2 Exercise Closed (Check Valve)

CO2 Exercise Open (Check Valve)

CP2 Partial Exercise (Check Valve)

DT Category D Test FC Fail-Safe Closed FO Fail-Safe Open LT1 Leak Rate PI Position Indicator Verification RT Relief Valve Test SC Exercise Closed SD De-energize SE Energize SO Exercise Open SP Partial Exercise (Cat. A or B)

A third letter, following the "LT" designation for leakage rate test, may be used to differentiate between the tests. For example, Appendix J leak tests will be designated as "LTJ",

low pressure (non-Appendix J) leak tests as "LTL", and high gressure leak tests as "LTH".

Three letter designations should be used for check valve condition monitoring tests to differentiate between the various methods of exercising check valves.

The letter following "CC", "CO" or "CP" should be "A" for acoustics, "D" for disassembly and inspection, "F" for flow indication, "M" for magnetics, "R" for radiography, "U" for ultrasonics, or "X" for manual exercise.

-10 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan U

Frci An abbreviation which designates the frequency at which the associated test is performed. Abbreviations used are:

Per Appendix J Cold Shutdown Quarterly Semiannually Refuel Outage Explosive Charge Sample Sample Disassemble & Inspect Per Technical Specification Requirements Annually Biennially Once per 4 Years Once per 5 Years Once per 6 Years Once per 8 Years Once per 10 Years Identifies the number of the Relief Request applicable to the specified test.

A cross-reference to the applicable Cold Shutdown Justification or Refuel Outage Justification which describes the reasons why reduced-frequency exercise testing is necessary for the applicable valve.

Any appropriate reference or explanatory information (e.g.,

technical positions, etc.).

Revision 10 October 14, 2002 AJ CS M3 M6 RR S2 SA TS Y1 Y2 Y4 Y5 Y6 Y8 Ylo

"° RR

"* Just Comments

4.0 Revision 10 October 14, 2002 OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENTS

1.

Pump Relief Request Index

2.

Pump Relief Requests

3.

Valve Relief Request Index

4.

Valve Relief Requests

5.

Cold Shutdown Justification Index

6.

Cold Shutdown Justifications

7.

Refuel Outage Justification Index

8.

Refuel Outage Justifications

9.

Station Technical Position Index

10.

Station Technical Positions

11.

Corporate Technical Position Index

12.

Corporate Technical Positions

13.

Inservice Testing Pump Table Index

14.

Inservice Testing Pump Table

15.

Inservice Testing Valve Table Index

16.

Inservice Testing Valve Table

ATTACHMENT 1 PUMP RELIEF REQUEST INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 1 PUMP RELIEF REOUEST INDEX Description Full-scale Instrument Pressure Gauges of 4.7.1(b)(1)

Range Requirements for Suction Service Water Pumps per ISTB Approval Date 10/02/02 Designates an existing, approved Relief Request which is being resubmitted for the fourth 10-year interval. The numerical identification number (e.g., 4) has been retained; the prefix "RP-" was added to comply with the recommendations of Corporate Procedure No. NES-MS-08.2. Minor changes may have been made to reflect the new version of the Code and to uldate or enhance the description or basis for relief.

Al-1 Revision 10 October 14, 2002 Designator RP-04

• ATTACHMENT 2 PUMP RELIEF REQUESTS

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan RELIEF REQUEST RP-04 SYSTEM:

Service Water PUMPS:

Service Water P-3-1A (1-1)

P-3-1B (1-2)

CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR RELIEF:

Suction pressure gages for above pumps.

Paragraph ISTB 4.7.1(b)(1) requires that the full-scale range of each analog instrument be not greater than three times the reference value.

The Service Water Suction Pressure instruments (PI-533-1173, 1172) are permanently installed instruments with a full scale range that exceeds three times the reference value as specified by the Code. Although these instruments do not meet Code requirements, they provide the same or better indication accuracy at the reference value than that which is permitted by the Code.

For instruments to be in compliance with the Code, they must be calibrated to an accuracy of +/-2% of full scale range, and have a full scale range no greater than three times the reference value.

The combination of the two requirements (i.e. accuracy equal to

+/-2% of full-scale and full scale being up to 3 times the reference value) yields a permissible inaccuracy of +/- 6% of the reference value. Section 5.5.1 of NUREG 1482 states that the staff will grant relief when the combination of the range and accuracy yields a reading at least equivalent to the reading achieved from instruments that meet the Code requirements (i.e., up to

+/-6 percent).

The table below shows the instrument accuracy and full scale range of the suction pressure instruments used to conduct inservice testing of the Service Water pumps.

The resulting instrument tolerance and indicated accuracy are calculated and also listed in the Table. In both cases, the indicated accuracy at the reference value is shown to be within the required 6 percent.

Replacement of the existing instruments with Code compliant instruments provides no safety benefit and could actually lessen the accuracy of the test results. A similar relief request for these instruments was approved in the Safety Evaluation Report contained in the letter from J. F. Stolz (NRC) to J. J. Barton (GPU Nuclear Corporation), dated August 24, 1993 for Oyster A2-1 Revision 10 October 14, 2002 Group A

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Creek Generating Station. Additionally, a similar request was approved for Limerick Generating Station, Units 1 and 2, in the Safety Evaluation Report contained in the letter from J. W.

Clifford (NRC) to J. A. Hutton (PECO Energy Company), dated November 28, 2000.

SYSTEM INSTRUMENT REFERENCE INSTRUMENT INSTRUMENT INSTRUMENT INDICATED NUMBER VALUE RANGE ACCURACY TOLERANCE ACCURACY 533 PI-533-1172 2.3 0 - 10 psig

+/-1%

+/- 0.1 psig 4.35%

533 PI-533-1173 2.2 0- 10 psig

+/-1%

+/- 0.1 psig 4.55%

ALTERNATE TESTING:

Based on Section 5.5.1 of NUREG 1482 and the information provided herein, the existing permanently installed pump instrumentation is considered acceptable in meeting the intent of the Code. No alternate testing will be performed. Accordingly, this alternative is being requested in accordance with 10CFR50.55a(a)(3)(i) in that the proposed alternative provides an acceptable level of quality and safety.

A2 -2 Revision 10 October 14, 2002

ATTACHMENT 3 VALVE RELIEF REQUEST INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 3 VALVE RELIEF REQUEST INDEX Designator RV-51

• Description Remote position indication verification Containment Isolation Valves per ISTC 4.1 of certain Approval Date 10/02/02 Designates an existing, approved Relief Request which is being resubmitted for the fourth 10-year interval. The numerical identification number (e.g., 5, 25, etc.) has been retained; the prefix "RV-" was added to comply with the recommendations of Corporate Procedure No. NES-MS-08.2. Minor changes may have been made to reflect the new version of the Code and to update or enhance the description or basis for relief.

A3-1 Revision 10 October 14, 2002

ATTACHMENT 4 VALVE RELIEF REOUESTS

SYSTEMS:

Main Steam Reactor Building Closed Cooling Water Instrument Air Reactor Water Cleanup Drywell Floor and Equipment Drains Containment Inerting Reactor Building Ventilation Reactor Head Cooling Valves:

V-1-7 V-1-8 V-1-9 V-1-10 V-5-147 V-5-166 V-5-167 V-6-395 V-16-1 V-16-2 V-1 6-14 V-1 6-61 CATEGORY:

V-22-1 V-22-2 V-22-28 V-22-29 V-23-13 V-23-14 V-23-15 V-23-16 V-23-17 V-23-18 V-23-19 V-23-20 V-23-21 V-23-22 V-27-1 V-27-2 V-27-3 V-27-4 V-28-17 V-28-18 V-28-47 V-31-2 A

FUNCTION:

TEST REQUIREMENT:

BASIS FOR RELIEF:

Containment Isolation Valves ISTC 4.1. Valves with remote position indicators shall be observed at least once every two years to verify that valve operation is accurately indicated.

Pursuant to 1 OCFR50.55a(f)(6)(i), AmerGen Energy Company, LLC, is requesting relief in that the radiation doses make it impractical to perform the inservice testing as required in the code. The above valves are located in radiation areas. Local observation to verify the accuracy of the position indicators will result in unnecessary radiation exposure. Alternate means can be used to verify accurate position indication. A similar relief request for these isolation valves was approved in the Safety Evaluation Report contained in the letter from J. F.

Stolz (NRC) to J. J. Barton (GPU Nuclear Corporation), dated September 24, 1992 for the Oyster Creek Generating Station, third interval inservice testing program.

A4 - 1 Revision 10 October 14, 2002 OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan RELIEF REQUEST RV-51

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ALTERNATE TESTING:

The position indicators for the above valves will be verified at least once every 2 years. In lieu of local observation, the following method will be used to verify accurate position indication. Proper system operation will verify accurate open position indication and successful leak rate test results will verify accurate closed indication.

A4 - 2 Revision 10 October 14, 2002

ATTACHMENT 5 COLD SHUTDOWN JUSTIFICATION INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 5 COLD SHUTDOWN JUSTIFICATION INDEX Designator Description Approval Date CS-01 Reactor Building Closed Cooling Water Containment 10/14/02 Isolation Valves Exercise Testing CS-02 Instrument Air Containment Isolation Valve Exercise Test 10/14/02 CS-03 Control Rod Drive HCU Charging Water Accumulator 10/14/02 Check Valve Exercise Test CS-04 Scram Discharge Volume Vent and Drain Valves Exercise 10/14/02 and Fail-Safe Testing CS-05 Shutdown Cooling Containment Isolation Valves Exercise 10/14/02 Testing CS-06 Core Spray System Testable Check Valve Exercise Testing 10/14/02 CS-07 Main Steam Isolation Valve (MSIV) Full Stroke Exercise 10/14/02 Testing CS-08 Reactor Water Cleanup System Containment Isolation 10/14/02 Valves Exercise Testing A5 - I Revision 10 October 14, 2002

ATTACHMENT 6 COLD SHUTDOWN JUSTIFICATIONS

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-01 SYSTEM:

Reactor Building Closed Cooling Water (RBCCW)

VALVES:

V-5-147, V-5-166, V-5-167 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 2, Category A, Active RBCCW Containment Isolation Valves Exercise test to the closed position in accordance with ISTC 4.2 V-5-147, V-5-166, and V-5-167 are motor-operated gate valves in the RBCCW supply and return lines to various coolers located inside Primary Containment. Closure of these valves isolates cooling water flow to the Recirc Pumps and Drywell Coolers.

Isolation of cooling water during normal plant operation can cause a Drywell temperature and pressure transient, or it can cause damage to the Recirc pumps requiring plant shutdown.

The operators for these valves are not designed for partial exercise testing.

These valves will be exercise tested to the closed position during Cold Shutdowns in accordance with ISTC 4.2.2.

A6 -1 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-02 Instrument Air SYSTEM:

VALVE:

V-6-395 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class NC, Category A, Active Instrument Air Containment Isolation Valve Full-stroke and fail-safe testing to the closed position in accordance with ISTC-4.2 The Instrument Air System provides air to the actuators and accumulators of the Main Steam Isolation Valves (MSIV's).

Closure of this valve could result in one or more of the MSIV's starting to drift toward the closed position, which in turn could result in a Reactor scram. Such an event during normal full power operation would create unnecessary challenges to Plant safety systems and equipment.

This valve will be full-stroke exercise tested and fail-safe tested during Cold Shutdowns in accordance with ISTC 4.2.2.

NOTE: This valve is a non-ASME Code Class component.

Revision 10 October 14, 2002 A6 - 2

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-03 Control Rod Drive (CRD)

SYSTEM:

VALVES:

V-15 (106)

CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 2, Category AC, Active CRD Hydraulic Control Unit (HCU) charging water check valves (one valve each per 137 HCU's)

Exercise test to the closed position in accordance with ISTC 4.5 Testing these valves during power operation would require depressurizing the Control Rod Drive charging water header, resulting in a possible degradation of the primary scram system's capability. Normal cooling water flow for the CRD's would also be lost while testing was in progress which could be detrimental to the control rod drives.

All 137 of these valves will be exercise tested to the closed position during Cold Shutdowns in accordance with ISTC 4.5.2.

Testing will be conducted by depressurizing the Control Rod Drive charging water header and verifying by the depressurization rate of the associated HCU that the valves have traveled to the closed position.

A6 - 3 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-04 Control Rod Drive SYSTEM:

VALVES:

V-15-119, V-15-120, V-15-121, V-15-133, V-15-134, V-15-135, V-15-136, V-15-137 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 2, Category B, Active Isolate the SDV on initiation of a scram to contain the water discharged from above the CRD piston.

Full-stroke exercise and fail safe test to the closed position in accordance with ISTC 4.2.

These valves can be exercised every three months, but the test or exercise solenoid used to bleed off control air pressure is not the same as the solenoid that would be used in the event of a scram. A full scram signal is required to actuate the safety function solenoids.

These valves will be exercised every three months.

This is considered a "partial" exercise, since the safety-related solenoid valves for the air operators are not used.

Fail-safe and stroke-time testing using the safety-related solenoids will be performed during Cold Shutdown in accordance with ISTC 4.2.2.

A6 - 4 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-05 Shutdown Cooling SYSTEM:

VALVES:

V-17-1, V-17-2, V-17-3, V-17-19, V-17-54, V-17-55, V-17-56, V-17-57 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category B, Active Shutdown Cooling System Containment Isolation Valves Full-stroke exercise to the closed position in accordance with ISTC 4.2 Valves V-17-19 and V-17-54 are interlocked to remain closed when Reactor Coolant Temperature is above 3501F.

Therefore, these valves cannot be cycled, either fully or partially, during normal Plant operation.

In addition, the function of all these valves is to isolate the Shutdown Cooling System from the Reactor Recirc System. They are maintained in the closed position at all times during power operation.

Opening them to perform an exercise test to the closed position would result in a reduction in the safety margin provided by keeping them closed.

These valves will be exercise tested to the closed position during Cold Shutdowns when Reactor Coolant temperature is below 350OF in accordance with ISTC 4.2.2.

A6-5 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-06 Core Spray SYSTEM:

VALVES:

V-20-150, V-20-151, V-20-152, V-20-153 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category AC, Active These testable check valves open to allow Core Spray injection into the Reactor Vessel following an accident.

In the closed position, these valves function as Reactor Coolant Pressure Boundary Pressure Isolation Valves to isolate the low pressure portions of the Core Spray System from the Reactor Coolant System.

Exercise to the fully open and closed positions necessary to fulfill the functions of these valves in accordance with ISTC 4.5.

During normal operation, the differential pressure across these valves is significantly greater than the shutoff head of the Core Spray pumps and greatly exceeds the capability of the test actuators to cycle them.

Furthermore, these valves are Reactor Coolant Pressure Boundary Pressure Isolation Valves and are required to maintain a barrier between the Reactor Coolant and Core Spray Systems.

These valves will be exercised to the fully open and closed positions during Cold Shutdowns using the test operators in accordance with ISTC 4.5.2.

A6 - 6 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD. SHUTDOWN JUSTIFICATION CS-07 SYSTEM:

Main Steam VALVES:

V-1-7, V-1-8, V-1-9, V-1-10 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

I BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category A, Active Main Steam Isolation Valves (MSIV's).

Isolate the Reactor Pressure Vessel from areas outside Primary Containment in the event of several accident and abnormal event situations.

Exercise testing to the close position in accordance with ISTC 4.2 Special testing is required to perform full-stroke exercise testing of these valves, which requires major plant power reduction evolutions.

Previous quarterly surveillance testing history for these valves supports testing on a Cold Shutdown frequency.

Furthermore, Section 2.4.5 of NUREG-1482 specifically identifies conditions which could result in a plant trip or require a reduction in power as appropriate justification for deferral of testing. A note at the end of Section 4.2.4 of NUREG-1482 states that the revised technical specification bases for MSIV surveillance requirements indicates that MSIV's should not be tested at power.

These valves will be exercise tested to the closed position during Cold Shutdowns in accordance with ISTC 4.2.2 and Technical Specification 4.5F.3.

A6 - 7 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan COLD SHUTDOWN JUSTIFICATION CS-08 SYSTEM:

Reactor Water Cleanup VALVES:

V-16-1, V-16-2, V-16-14, V-16-61 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category A, Active Provide isolation of the RWCU System Containment penetrations upon receipt of isolation signals.

Exercise testing to the close position in accordance with ISTC 4.2 These valves are required to be open during normal plant operation so that the Cleanup System can maintain the water quality of the Reactor Coolant within Technical Specification limits. The valves isolate automatically on receipt of Containment and/or system isolation signals. Exercising these valves requires the RWCU filter and demineralizer to be taken out of service.

Isolating the system, performing the testing, and restoring the system to service during power operation is a complex evolution generally involving core maneuvering, power reductions, and a significant amount of time and radiation dose. Sudden changes in system temperature or flow could result in significant water chemistry changes, which challenge allowable limits. In addition, instances of resin intrusion into the RPV have occurred at other plants as a result of Cleanup System evolutions while at power.

Surveillance history of these valves supports testing on a Cold Shutdown frequency. Actuator design for these valves does not support partial stroke testing.

These valves will be exercise tested to the closed position during Cold Shutdowns in accordance with ISTC 4.2.2.

A6 - 8 Revision 10 October 14, 2002

ATTACHMENT 7 REFUEL OUTAGE JUSTIFICATION INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 7 REFUEL OUTAGE JUSTIFICATION INDEX Description Main Steam Isolation Valve Fail-Safe Testing EMRV Discharge Piping Vacuum Relief Valve Exercise Testing Feedwater Check Valve Exercise Testing Reactor Building Closed Cooling Water Containment Isolation Check Valve Closure Testing Instrument Air Containment Isolation Check Closure Testing Valve CRD Hydraulic System Discharge to RPV Containment Isolation Check Valves Closure Testing Reactor Water Cleanup Return Containment Isolation Check Valve Closure Testing Liquid Poison Injection Header Check Valve Exercise Testing Control Rod Drive Scram Valve Exercise Testing Control Rod Drive 108 Valve Exercise Testing Designator RJ-01 Revision 10 October 14, 2002 Approval Date 10/14/02 10/14/02 10/14/02 10/14/02 10/14/02 10/14/02 10/14/02 10/14/02 10/14/02 10/14/02 RJ-02 RJ-03 RJ-04 RJ-05 RJ-06 RJ-07 RJ-08 RJ-09 RJ-10 A7-1

ATTACHMENT 8 REFUEL OUTAGE JUSTIFICATIONS

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-01 SYSTEM:

Main Steam VALVES:

V-1-7, V-1-8, V-1-9, V-1-10 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category A, Active Main Steam Isolation Valves (MSIV's).

Close to isolate the Reactor from the Main Steam System outside of Primary Containment in a number of accident scenarios. Also function as Containment Isolation Valves.

Fail-safe testing in accordance with ISTC 4.2.6 These valves normally close by a combination of spring force and air pressure. Special testing is required to perform fail-safe closure of these valves, which requires plant shutdown or major power reductions. In order to satisfy Code requirements, the fail-safe test is performed without assistance from the pneumatic supply system. Access to the valves is required for pneumatic system alignment and to observe that the operators function properly during the test.

Performance of this test requires access to either the Drywell or Trunnion Room, which are both high radiation areas.

Drywell access may not be available during Cold Shutdowns.

These valves will be fail-safe tested when the proper alignments can be made and access is available. This will be on no less than a refueling outage frequency as permitted by ISTC 4.2.2(e).

Additional testing may be performed during Cold Shutdowns if access to the valves is available.

A8 - 1 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-02 SYSTEM:

Main Steam VALVES:

V-1-190, V-1-191, V-1-192, V-1-193 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 2, Category C, Active These check valves are used to equalize pressure between the EMRV discharge downcomers and the Drywell atmosphere following EMRV actuation.

The valves open to prevent siphoning of water from the Torus as the steam in the downcomers condenses, which could subject the downcomers, downcomer supports, and other Torus components to excessive forces upon a reopening of the EMRVs.

Exercise test to the open position required to fulfill their function in accordance with ISTC 4.5 Access to the Drywell is required to perform the requisite testing. These valves are not provided with any mechanisms for exercising the internals, and the test requires removal of the valve inlet screen and the use of a special tool rig to stroke the valve and measure the opening force.

Due to the inert atmosphere and high radiation conditions in the Drywell, this is not possible during normal power operation, and is impractical during most Cold Shutdown outages.

These valves will be exercise tested during each refueling outage as permitted by ISTC 4.5.2(c) using a special tool rig to open and measure the opening force of each valve.

A8 - 2 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-03 Feedwater V-2-71, V-2-72, V-2-73, V-2-74 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category AC, Active Provide Containment Isolation and prevent loss of inventory from the Reactor Vessel in the event of a loss of Feedwater.

Exercise test to the closed position in accordance with ISTC 4.5 It is essential that these valves remain open to supply inventory to the Reactor Vessel during normal power operation in order to maintain reactor level.

An interruption in Feedwater flow would result in an unnecessary shutdown of the Plant. These valves are also impractical to test during Cold Shutdowns, since they are not fitted with exercise arms or position indication. All of the valves utilize LLRT test results to demonstrate that they have traveled to the fully closed position, which requires containment deinerting and drywell access for system alignment and draining. Extensive effort is required to set up the equipment, drain the test volume, perform the test, refill the test volume, remove the test equipment and restore the system to operation. Such effort would most likely delay Plant startup and would result in unnecessary exposure to personnel.

These valves will be exercise tested to the closed position during each refueling outage as permitted by ISTC 4.5.2(c).

A8 - 3 Revision 10 October 14, 2002 SYSTEM:

VALVES:

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-04 SYSTEM:

Reactor Building Closed Cooling Water (RBCCW)

VALVE:

V-5-165 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 2, Category AC, Active RBCCW Containment Isolation Check Valve Exercise test to the closed position in accordance with ISTC 4.5 This valve is a check valve in the RBCCW supply header to the system heat loads inside Containment. Testing of this valve to the closed position requires shutdown and isolation of the RBCCW System. This valve is not fitted with an exercise arm or position indication, and special testing is required to verify the closed position of this valve. There is no practical means to perform such testing during normal Plant operation nor during Cold Shutdown.

Based on the function of this valve, partial testing in the closing direction is neither meaningful nor practical.

This valve will be exercise tested to the close position during refueling outages as permitted by ISTC 4.5.2(c).

A8 - 4 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-05 Instrument Air SYSTEM:

VALVE:

V-6-393 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class NC, Category AC, Active Instrument Air System Containment Isolation Check Valve.

Exercise test to the closed position in accordance with ISTC 4.5.

This valve is not fitted with any parts which would give an external positive indication of its position, such as an exercise arm or position indicating device. Special testing is required to test this valve to the closed position, which requires Drywell access and isolation of control air to all equipment in the Drywell, including the Main Steam Isolation Valves (MSIV's) and Drywell Coolers.

This valve will be exercise tested to the closed position during each refueling outage as permitted by ISTC 4.5.2(c).

NOTE: This valve is a non-ASME Code Class component.

Revision 10 October 14, 2002 A8 - 5

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-06 Control Rod Drive V-15-27, V-15-28 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category AC, Active Control Rod Drive Hydraulic System Discharge to Reactor Vessel Containment Isolation Valves Exercise test to the closed position in accordance with ISTC 4.5.

These valves are not fitted with any parts which would give an external positive indication of their position, such as an exercise arm or position indicating device.

In order to test these valves to the closed position, the CRD Hydraulic System must be shut down. Since the System must be operable during power operation and is normally kept operating during periods when the Reactor is shut down, it is not practical to attempt to test these valves quarterly or during Cold Shutdowns.

These valves will be exercise tested to the closed position during each refueling outage as permitted by ISTC 4.5.2(c).

A8-6 Revision 10 October 14, 2002 SYSTEM:

VALVES:

OYSTER CREEK NUCLEAR GENERATING STATION OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-07 Reactor Water Cleanup V-16-62 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category AC, Active Reactor Water Cleanup System Return Containment Isolation Check Valve.

Exercise test to the closed position in accordance with ISTC 4.5.

This valve is required to be open during normal plant operation so that the Cleanup System can maintain the water quality of the Reactor Coolant within Technical Specification limits.

Exercising this valve requires the RWCU filter and demineralizer to be taken out of service. Additionally, this valve is not fitted with any parts which would give an external positive indication of its position, such as an exercise arm or position indicating

device, hence it requires LLRT testing to demonstrate that it has traveled to the fully closed position.

This valve is located in the Drywell, requiring deinerting of the Containment for system alignment and draining to perform the LLRT. The Cleanup System is normally maintained in operation during Cold Shutdown. Extensive effort is required to set up the equipment, drain the test volume, perform the test, refill the test volume, remove the test equipment and restore the system to operation. Such an effort would most likely challenge Reactor Chemistry limits, delay Plant startup, and result in unnecessary exposure to personnel.

This valve will be exercise tested to the closed position during each refueling outage as permitted by ISTC 4.5.2(c).

A8 - 7 Revision 10 October 14, 2002 SYSTEM:

VALVE:

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-08 SYSTEM:

Standby Liquid Control (Liquid Poison)

VALVES:

V-19-16, V-19-20 CLASS/CATEGORY:

FUNCTION:

TEST REQUIREMENT:

BASIS FOR EXTENDED FREQUENCY:

ALTERNATE TESTING:

Class 1, Category AC, Active These valves open to permit the injection of liquid poison (sodium pentaborate) into the Reactor Vessel in the event that the control rods are not able to shut it down.

Exercise test to the open position in accordance with ISTC 4.5.

In order to exercise these valves to the open position, one of the two explosively-actuated Squib valves must be fired and liquid must be injected by the corresponding pump.

During power operation, this would involve the injection of cold, highly concentrated sodium pentaborate into the RCS, causing plant shutdown, and creating an undue burden on Cleanup System resources. During Cold Shutdown, flushing of the System for long periods of time would be required resulting in large quantities of hazardous waste material and very likely resulting in a delay of Plant startup.

This valve will be exercise tested to the open position during each refueling outage as permitted by ISTC 4.5.2(c) in conjunction with the test firing of one of the two Squib valves.

A8 - 8 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-09 SYSTEM:

VALVES:

CATEGORY:

FUNCTION:

TEST REQUIREMENT:

Control Rod Drive V-15 (126), V-15 (127)

B These valves (137 of each) are the scram inlet and outlet valves for each of the Hydraulic Control Units (HCU's). They function to scram the associated control rod.

Perform exercise, stroke-time and fail-safe testing in accordance with the applicable requirements of ISTC 4.2.

BASIS FOR EXTENDED FREQUENCY:

CV-126 and 127 cannot be exercised during power operation since exercising these valves will scram the associated control rod. Rapid insertion and withdrawal of the rod at power could cause fuel damage to the core.

These valves are not provided with indication for both positions and have stroke times in the order of milliseconds.

Thus, measuring of stroke time for these valves is impractical.

ALTERNATE TESTING:

Per Technical Specification requirements, a sample of 8 of these valves are tested during startup from cold shutdown if the sample has not been tested in the previous 6 months. All valves are tested at refueling.

Verifying the associated control rod meets the scram insertion time limits as defined in Technical Specifications is an acceptable method of detecting degradation of these valves.

This testing is considered to comply with the requirements of the OM-1 995 Code with OMa-1 996 addenda on the basis that (1) all valves are tested during refueling, with many being tested at a greater frequency, (2) the valves are rapid-acting and can be verified to stroke well within the 2-second limit, and (3) obturator travel is verified by indirect evidence as permitted by ISTC 4.2.3.

As documented in Section 1.2 of NUREG 1482, Generic Letter 89-04 granted approval to follow the alternate testing delineated in Position 7, provided its use is documented in the IST Program.

A8 - 9 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan REFUELING OUTAGE JUSTIFICATION RJ-10 Control Rod Drive SYSTEM:

VALVES:

CATEGORY:

FUNCTION:

TEST REQUIREMENT:

Each of these valves (1 each per 137 HCU's) opens to allow displacement of fluid from above the associated control rod drive piston in order to permit successful insertion when required for a scram.

Exercise test in accordance with ISTC 4.5 BASIS FOR EXTENDED FREQUENCY:

These valves can only be verified open during the actual scram testing. Verifying the associated control rod meets the scram insertion times specified in Technical Specifications is an acceptable alternative method of verifying the full open position of these valves according to Position 7 of NRC Generic Letter 89-04.

ALTERNATE TESTING:

Per Technical Specification requirements, a sample of 8 of the 137 valves are tested during startup from cold shutdown if the sample has not been tested in the previous 6 months. All valves are tested during each refueling outage.

As documented in Section 1.2 of NUREG 1482, Generic Letter 89-04 granted approval to follow the alternate testing delineated in Position 7, provided its use is documented in the IST Program.

A8 - 10 Revision 10 October 14, 2002 V-15 (108)

C

ATTACHMENT 9 STATION TECHNICAL POSITION INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 9 STATION TECHNICAL POSITION INDEX Description Valve Packing Adjustment Post Maintenance Test Recommendations Approval Date 10/30/01 Revision 10 October 14, 2002 Designator A9 - 1

ATTACHMENT 10 STATION TECHNICAL POSITIONS

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan VALVE PACKING ADJUSTMENT POST MAINTENANCE TEST RECOMMENDATIONS

1.0 INTRODUCTION

Engineering and Maintenance valve experts at Oyster Creek have expressed interest in waiving the valve stroke time post maintenance test (PMT) following valve packing adjustment tasks. The justification for waiving this PMT is based on the assumption that the valve is being returned to a condition for which operability has previously been demonstrated. The following discussion will address the relevant technical and regulatory requirements.

Since 1989 there have been significant changes in the maintenance and testing of valves in the nuclear industry. Generic letter 89-10, which required stations to evaluate conditions that could effect valve operability, brought into consideration the friction generated by the valve packing relative to the operability of a valve. The operability margin of the valve, type of packing, packing configuration, stem movement (rising, rising/rotating, or quarter turn), and valve stem condition are all considerations when evaluating the effects of packing friction on valve operability. In most motor operated valve (MOV) applications, a properly designed graphite packing system does not generate frictional loads large enough to effect operability.

However, in some MOV applications and most air operated valve (AOV) applications packing frictional loads can effect valve operability (Ref. 11.1).

In an effort to improve packing performance and reduce friction, several packing manufacturers, valve manufacturers, and utilities have performed extensive testing over the past decade. Numerous stations have implemented 'packing programs',

where correlations have been established between predicted packing friction values and actual running loads from diagnostically tested MOVs and AOVs. This data, coupled with independent test data from facilities such as AECL Chalk River and Wyle Laboratories, provides validation for predicting packing friction.

2.0 ASME CODE REQUIREMENTS The Inservice Test (IST) Code in effect at Oyster Creek is OM-1995 Code with OMa-1 996 Addenda.

OMa - 1996 states: 'When a valve or its control system has been replaced, repaired, or has undergone maintenance that could affect the valve's performance, a new reference value shall be determined or the previous value reconfirmed by an inservice test run before the time it is returned to service or immediately if not removed from service. This is to demonstrate that performance parameters that could be affected by the replacement, repair, or maintenance are within acceptable limits.... Adjustment of stem packing...(etc.) are examples of maintenance that could affect valve performance parameters'.

Similarly, Section X1, Subsection IWV-3200, 1981 Edition states: 'When a valve or its control system has been replaced or repaired or has undergone maintenance that A10 - 1 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan could affect its performance, and prior to the time it is returned to service, it shall be tested to demonstrate that the performance parameters, which could be affected by the replacement, repair, or maintenance, are within acceptable limits. Examples of maintenance that could affect valve performance parameters are: adjustment of stem packing...(etc.)'.

On March 10, 1992, the Section XI Code Committee provided the following interpretation of Section XI, Subsection IWV-3200 (this was an interpretation of the 1980 Code with 1981 addenda):

Interpretation:

XI-1-92-16 File:

IN91-045 Question:

If it is established that adjustment of packing will not affect the stroke time of a specific valve, is a stroke time test required by Section Xl, IWV-3200?

Reply:

No.

Clearly this interpretation supports the position that PMT stroke time testing is not necessarily required for all packing adjustments. However, the wording of the question does imply that this judgement is made based on knowledge of the specific valve and the specific packing adjustment task (i.e. specified torque value, etc.).

3.0 NUREG-1482 GUIDELINE FOR INSERVICE TESTING The NRC guideline for inservice testing, NUREG-1482 (Reference 11.2, April, 1995),

addresses the issue of PMT requirements following a packing adjustment. Prior to NUREG-1482, when an MSIV (or other valve in the IST Program) required a packing adjustment, a stroke test was typically required to prove operability. This requirement sometimes necessitated a forced shutdown or downpower. In NUREG 1482, section 4.4.4, the NRC states 'If it is necessary to stop leakage, and if a required stroke test or leak rate test is not practical in the current plant operating mode, the licensee must justify that the packing adjustment is within the torque limits specified by the manufacturer for the existing configuration of packing, such that the performance parameters of the valve are not adversely effected. The licensee must evaluate any data available from previous testing with the packing torqued to the limit specified and verify the valve was previously stroked within the acceptable limits with the packing adjusted to the higher value'. The NRC is generally stating that following packing adjustments performed to stop packing leakage, by increasing packing torque above previous torque values, PMT stroke tests can be deferred until the next opportunity when the plant is in a mode where the test can be performed. The NRC also reiterates that the ASME Committee interpretation (IN-91-045) stated that a stroke time test is not required if it is established that adjustment of packing will not affect the stroke time of a specific valve. The NRC is stating in this document that if previous test data is used, certain packing adjustment PMT requirements can be waived or deferred. Section 9.0 provides recommended guidance relative to creating documentation to prove operability of a valve following a repack or packing adjustment without performing diagnostic or stroke time testing.

A10-2 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan documentation to prove operability of a valve following a repack or packing adjustment without performing diagnostic or stroke time testing.

As stated above, the NRC recognizes in NUREG-1482 that there is a reasonable technical basis for deferring PMT stroke time testing under some circumstances.

However, in NRC IN 88-51('Failures of Main Steam Isolation Valves'), the NRC also emphasized the need to carefully consider the adequacy of surveillances in establishing the operability of MSIVs or other similar valves, after excessively tight packing prevented all eight MSIVs from fully closing at Dresden.

4.0 MOV USER'S GROUP SURVEY RESULTS In July 1994, the MOV User's Group Packing Sub-Committee sent out a survey to identify what the nuclear stations in the U.S. and Canada were doing in regards to PMT following a packing adjustment or replacement. Of the 44 stations that responded to the survey, 33 retested after adjusting the packing, and 35 retested after replacing the packing. Several other stations had 'conditional' testing that generally required valves to be retested only if the original torque was exceeded.

Examples of documentation of technical justification for waiving PMT for packing adjustments and repacks was requested by the sub-committee. The stations that responded indicated that they had controlled packing programs, and they used their diagnostic-test results, as well as test data from other stations and laboratories, to support their justifications. When a valve was repacked or adjusted, a conservative packing friction value was estimated from previous DP testing data to assure adequate seat load. Written documentation was provided to eliminate or reduce post maintenance testing following a packing adjustment or replacement. In some cases, motor current was checked at the MCC to identify any major friction effects resulting from a packing adjustment or replacement. Results from the survey were used to develop a MOV User's Group Guideline on the subject. This survey shows that although the majority of stations perform PMT stroke testing following packing adjustments, there is support in the industry for the position that it is not required under some circumstances.

5.0 SUPPORTING PROCESSES AND PROGRAMS To support waiving or deferring PMT following a packing adjustment or replacement, the following should be considered:

5.1 Implement Valve Packing Program Reference 11.1 provides an overview of a valve packing program, which was generated from stations that have highly rated programs. The key areas of a valve packing program include:

5.1.1 Using quality valve packing materials - Although the program should address all types of packing, graphite packing materials are the most commonly used in nuclear stations because of their thermal stability, radiation resistance, low halogen content, and excellent sealing A10 - 3 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan properties. The only drawbacks to using flexible graphite are its tendency to extrude and potentially high frictional loads on the valve stem. Select packing materials that provide low and stable packing friction.

5.1.2 Using optimum packing configurations to maximize packing performance while keeping packing friction at the lowest level achievable. To address flexible graphite extrusion, containment ('anti extrusion') rings must be used above and below the flexible graphite seal rings.

5.1.3 Applying the proper gland stress to the packing system.

5.1.4 Using proper installation and maintenance techniques developed and implemented through training and procedures. Installation and maintenance of modern packing systems requires a high level of knowledge and expertise from the installer or maintenance technician.

The proper technique of stroking the valve and applying the correct torque to consolidate the packing system is important; also the use of calibrated preset torque wrench, or double verification of torque adjustments may be applicable.

5.1.5 Documenting the packing configuration and applied gland stress (or torque) in a controlled database. This data sheet verifies the variables that were used to calculate the packing friction. A computerized packing program can also be utilized to select the packing system and calculate packing friction and gland nut torque.

Generate evaluations as required to justify deferring or waiving PMT stroke tests per Section 9.6.

5.2 Develop Validated Formulas for Packing Friction A validated friction formula would not be needed when adjusting the packing system back to the original gland torque or stress recorded from a previous valid baseline test (see section 5.4).

Under other circumstances, it may be beneficial as part of an AOV, MOV and packing program to have the ability to predict the frictional effects of the packing system when performing a packing adjustment or repacking a valve.

Having an accurate estimate of the packing friction may be beneficial when determining the operability of valves, and quantifying the friction formula with hard data can reduce or eliminate costly PMTs. In order to predict packing friction it is important that quality packing materials have been installed and controlled as discussed in Section 5.1. Several packing manufacturers have performed extensive testing to determine the frictional limit of their packing systems. The packing supplier should be able to supply a summary of laboratory and field tested valves, and the comparison of actual versus predicted running loads. The packing supplier can provide an upper and lower expected packing friction limit for the packing systems supplied. A A10 - 4 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan sampling of in-house testing could be performed to verify the predicted packing friction is within the limits supplied by the packing supplier.

5.3 Implement MOV and AOV Programs When identifying the effects of packing related friction through diagnostics, it is important that the user understand the accuracy of the diagnostic systems in the running load region. The method and range of calibration, placement of strain gages and marking of the trace can all have a significant effect on the indicated versus the actual running load.

5.4 Valid Baseline Diagnostic or Stroke Time Test In a situation where the packing system is adjusted back to its original value, a previous baseline diagnostic test or stroke time test may be used to eliminate a diagnostic post maintenance test if the following conditions are addressed:

5.4.1 The packing system that is used has equal or lower packing friction when subjected to increased temperature and aging. Testing to prove this has been performed by most of the major packing suppliers, and stations should request test data from the packing supplier for validation, or perform in-house testing.

5.4.2 Prior to performing a baseline test, the packing system must be consolidated (see section 6.1).

5.4.3 The packing configuration and the applied gland torque at the time of the baseline test must be documented and not changed.

A valid baseline test may not be necessary if a predicted friction value is going to be used.

5.5 Operability Margin Review The available operability margin of a valve must be known if a predicted friction value is used for packing replacement or adjustment. Valves with marginal operators should be handled conservatively. The upper packing friction limit should be used to assure a conservative margin of operability.

6.0 VARIABLES THAT CONTRIBUTE TO PACKING FRICTION It is not possible through diagnostic testing to distinguish packing friction from other variables that contribute to valve running load. Factors which could effect the accuracy of diagnostics in the running load region will be discussed in Section 8.0.

Only in a test facility can the environment be controlled to the point where the indicated running load is primarily packing friction.

A10 - 5 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan Within a grouping of similar valves, there are numerous reasons why there could be a significant variations in the indicated running loads. The condition of the valve packing gland, valve stem and stud bolts can have a significant effect on how much of the gland stress actually reaches the packing system. In most cases these situations result in less packing load reaching the valve packing system and lower than expected packing friction. Additionally, the procedure and method in which the valve packing was installed and initially consolidated has a major impact on the performance and friction of the packing system.

6.1 Installation/Consolidation of the Packing System When the gland stress is applied to the packing set, much of the stress is initially absorbed by the upper packing rings. Actuating the valve stem transmits this energy through the entire packing system. Until this consolidation is complete, the packing friction is very inconsistent. If the packing system is not consolidated correctly, lower packing friction would be expected, as well as potential packing leakage. The valve should be stroked while returning the gland stud torque to its original value between strokes.

When the packing gland torque has stabilized, the packing system is initially consolidated. Running load data should never be collected until the valve packing system has been properly torqued and consolidated. If a valve is diagnostically tested, the packing gland torque should be re-established due to consolidation of the packing system.

6.2 Valve Stem Finish The finish of the valve stem can have an impact on packing friction and performance. Typically, the smoother the finish of the stem, the lower the packing friction and the better the performance of the packing system. Some packing systems do not seem to be effected as much as others when subjected to varying stem finish. Flexible graphite is a material that has the ability to flow into the microscopic pores of the valve stem. When these pores are filled with flexible graphite, it creates high 'breakaway' friction when the valve stem is actuated because the flexible graphite has to be sheared.

Also, every actuation of the valve stem results in flexible graphite being removed from the packing system which reduces packing life. The packing manufacturer can provide information on what stem finish is optimum for their packing systems.

6.3 Temperature and Cycle Life It has been noted in testing and field experience that the packing friction decreases with increasing temperature. This characteristic is important to consider when using flexible graphite since many critical valves operate at elevated temperatures. During laboratory testing of combination composite/graphite and standard yarn/graphite packing systems it was noted that packing friction decreased by 30 - 40% when the test rigs were up to temperature (approximately 700F). This aspect could be useful if a valve is on the edge of operability and the assurance of stable or reduced packing friction is important. It is important to note that the packing friction typically A10 - 6 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan did not increase with stem cycles. While this is true of most packing systems, the packing supplier should be consulted to verify the type of packing materials being used do not have increased friction when subjected to age, cycling and temperature.

6.4 Gland StresslTorque A key factor in designing a graphite packing system is applying the proper amount of gland stress to achieve sealing while keeping friction to a minimum. To accomplish this, it is important to take into consideration the system pressure, cross-sectional area of the packing ring, and bolt condition.

Torque has been the most common method of applying stress to the packing set, but more accurate methods of applying packing stress such as stud elongation and calibrated Belleville spring packs are also available. In most cases, applying a specified torque is the most economical and practical method of applying stress to the packing system. One drawback of measuring torque is that imperfect or degraded condition of the gland studs can result in poor torque-to-stress efficiency. Even in good conditions, friction developed between the gland stud nut and gland stud or flange may vary significantly between studs, and can decrease gland stud load by as much as 40%. This could result in non-uniform gland stud loading, or lower than intended gland stud loading. When determining how much stress is needed to seal, it is important to take into consideration the density of the packing rings and the system pressure. The packing supplier should provide their theories and calculations for the proper gland stress and torque that should be used.

6.5 Live Loading Live-loading is a simple, extensively used technology that uses an assembly of Belleville washers between the gland stud nut and the gland flange. The purpose of the Belleville washers is to provide a constant stress on the packing system in the event of packing material loss, thermal changes within the valve, or consolidation of the valve packing. Live-loading is a very simple and useful technique when applied properly to graphite packing systems.

Selection of the proper belleville washer should be accomplished through actual load versus deflection test reports. A properly designed live-loading system does not add any additional friction to the packing system. However, if you compare a friction profile of a live-loaded and a conventionally loaded valve, the live-loaded valve will often be higher. There are two reasons for this:

6.5.1 The live-loaded valve's graphite packing system will initially consolidate much quicker than the conventionally loaded valve. The more consistent load provided by the belleville washers drives the gland stress through the packing system more efficiently.

6.5.2 As described above, in conventionally loaded valves, imperfect or degraded condition of the gland studs can result in poor torque-to stress efficiency. In a live-loading system, hardened flat washers are A10 - 7 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan placed below the packing stud nut. This approach results in very low friction when compared to a conventionally loaded valve, and improved efficiency in transferring the gland stress to the packing system. Many stations now place flat washers between the packing stud nut and the gland follower flange to reduce friction, thereby improving torque to stress efficiency.

7.0 OTHER VARIABLES THAT CONTRIBUTE TO VALVE RUNNING LOAD Most of the factors that effect packing friction can be controlled by using good maintenance practices and quality packing materials. In the laboratory, packing friction is quite repeatable (+/- 10%), considering the variance that can be experienced in torquing on a soft joint. In practice, according to Ref. 11.1, many of the packing failures and valves that have low packing friction can be traced back to poor bolting and torque practices. Other variables that can effect running load are more complex. There are many different types of valves, and components unique to those valves that could effect running load. Close scrutiny of the diagnostic trace can often identify these effects or factors but, unfortunately, many times the packing is blamed for high running loads. If running loads are higher than the predicted upper limit, then other conditions that increase running load may be present and should be investigated.

7.1 Gate Guide Friction The friction developed in the guides in a gate valve can be significant if galling or improper tolerances exist. Gate guide friction can often be identified in a diagnostic trace by an erratic running load and/or increasing running load as more of the gate contacts the guides. Orientation of the valve stems in other than vertical position can add running load because of the friction in the gate guides.

7.2 Anti-rotation Device Friction Certain valves use a mechanical anti-rotation device that can be a significant source of running load. The friction associated with this device can be reduced by improving the sliding finish and using lubrication. A diagnostic trace can identify anti-rotation device friction. Typically, the trace will be very erratic, and will look very much like gate guide friction.

A1 0 - 8 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan 7.3 Gland FollowerlLantern RinglJunk Ring Galling Cocked gland followers are the major cause of severely damaged valve stems. Also, lantern rings and junk rings can become tilted and gall the valve stem. This problem results in high running loads, packing failure, and costly stem replacement. Fortunately, this is a problem that can be averted by using good maintenance practices and modern packing technology. The repack or packing adjustment procedure should have a step to verify that the valve stem is not cocked or contacting the gland follower. This can be accomplished by slipping a piece of.005" plastic shim stock between the follower and the stem following a repack or packing adjustment to verify clearance exists. Lantern rings in valves with an inactive leakoff should be removed. Valves with active leakoff should be targeted for removal of the leak collection system and converted to a single packing set. If this conversion cannot be accomplished, then the use of split graphite lantern rings should be considered to minimize the potential for stem scoring.

8.0 DIAGNOSTIC SYSTEMS Regardless of the type of diagnostic system used, it is important to identify the accuracy and the methods for collecting data in the running load region. Since running load may only be a fraction of the valve's seating load, accurately reading the running load values may be difficult.

8.1 Factors that Effect Diagnostic Test Accuracy 8.1.1 Load Sensor Location 8.1.2 Calibration 8.1.3 Zero Point Improper marking of the trace is very common and results in incorrect indicated running load unless the average of the open/close is taken.

In a baseline test, the only difference between the opening and closing running load should be the weight of the disk and the stem. At ambient temperature, packing friction is a constant force and will not be changed by a change in stem direction. At elevated temperatures the valve stem is expanded and may create a 'wedge' when moving in the open position. This thermal expansion creates higher friction in the up stroke.

8.2 Using Diagnostics to Improve Packing Performance As stations satisfy the regulatory requirements of valve testing, the use of valve diagnostics to improve valve performance will be emphasized. With the packing materials and technology available today it is possible to predict packing friction within a certain range. Low indicated running loads can indicate insufficient sealing force. Higher than expected running loads can A10 - 9 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan identify other anomalies such as gate guide galling, stem binding and diagnostic inaccuracy.

9.0 DEFINING PACKING POST MAINTENANCE TESTING The primary intent of this guideline is to provide an approach to effectively address PMT requirements following a packing adjustment or replacement. The packing PMT flow chart (Attachment A) provides direction on PMT requirements. There are several distinct maintenance tasks involving valve packing for which PMT requirements should be addressed. They include:

9.1 Adjusting the packing torque back to the original value on a valve that has a known packing system and a valid baseline test.

The situation occurs when a valve develops a leak or as part of a preventative maintenance program the packing nut torque is restored to its original value. If the valve was diagnostically tested at a known torque value with a known packing system, the original baseline test could be used as a prediction of the running load following the adjustment when returned to the original torque value.

9.2 Adjusting the packing torque to a higher value on a valve that has a known packing system and a valid baseline test.

This situation occurs when the packing stud torque is increased above the original value. In this situation, if it is necessary to use a predicted friction; the upper friction limit provided by the packing supplier should be used.

9.3 Adjusting the packing torque to the original value on a valve that does not have a known packing configuration or valid baseline test.

If the packing configuration and the amount of applied gland stress is not known, it is impossible to predict the packing friction. Even if the packing was torqued during static testing if the packing configuration and materials is not known, it is recommended that any packing adjustments be followed by an actual thrust measurement or a full static PMT.

9.4 Adjusting the packing torque to a higher value on a valve that does not have a known packing configuration or a valid baseline test.

If the packing configuration and the amount of applied gland stress is not known, it is impossible to predict the packing friction. Even if the packing was torqued during static testing if the packing configuration and materials is not known, it is recommended that any packing adjustments be followed by an actual thrust measurement or a full static PMT.

A10 - 10 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan 9.5 Replacing the packing system.

Repacking valves with a new packing configuration requires an evaluation of the predicted packing friction of the packing system. The predicted or calculated friction values can be used to determine if sufficient actuator thrust is available following packing replacement.

9.6 PMT Recommendations Each of these tasks is different in the approach to the PMT requirements.

Valves that have been repacked under a valve packing program with known materials, and statically baseline tested would require the least amount of documentation to justify the reduction of PMT. Valves repacked with a new packing configuration or adjusted to a higher torque value would require a more extensive review of the predicted friction against available AOV or MOV margin.

There are several different types of packing post maintenance tests that can be performed following a packing adjustment. In the flow chart (Attachment A) are references to PMT1 and PMT2. PMT1 would normally consist of a full diagnostic static trace using direct measurement. Stem or yoke mounted strain gages or 'C' clamps would be the most common method for PMT1.

PMT2 would normally consist of a functional stroke test or a current test. In conditions where a valve could not be stroked to perform PMT2, an engineering evaluation could be performed. In accordance with NUREG 1482, if the valve is in the IST Program, the PMT stroke test could then be performed when practical. If it is established that adjustment of packing will not affect the stroke time of a specific valve, a stroke time PMT is not required. An evaluation (documented in an Action Request Evaluation or other appropriate document) should be performed by the AOV, MOV or other valve expert documenting deferring or waiving the PMT, and should address the following as applicable:

9.6.1 Review of packing type & configuration 9.6.2 Review of previous maintenance history and diagnostic or stroke time test results 9.6.3 Evaluation of the calculated packing friction against available AOV or MOV margin, if applicable 9.6.4 Data review from similar valves, if applicable 9.6.5 AOV and MOV industry data, if applicable

10.0 CONCLUSION

The ASME Code and NUREG-1482 allow deferring or waiving stroke time PMT for packing adjustments, provided it is established that the adjustment will not affect the A10 - 11 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan stroke time of the specific valve. This determination should be documented per the recommendations of Section 9.6.

11.0 REFERENCES

11.1 MOV User's Group, 'Guideline for Post Maintenance Testing for Valve Packing Replacement or Valve Packing Adjustment', July 1995.

11.2 NUREG-1482, 'Guidelines for Inservice Testing at Nuclear Power Plants',

April 1995.

11.3 NUREG/CP-0137, pg. 513-522, 'Argo Packing Friction Research Update',

July 1994.

11.4 NRC Information Notice No. 88-51, 'Failures of Main Steam Isolation Valves',

July 1988.

11.5 BWR Owners' Group, 'Appendix J - GL 89-10 Correlation Retest Requirement Guidelines for Appendix J Valves', April 1996.

11.6 ASME OM-1995 Code with OMa-1996 Addenda, Code for Operation and Maintenance of Nuclear Power Plants.

A10-12 Revision 10 October 14, 2002

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT A PACKING ADJUSTMENT PMT RECOMMENDATIONS YES DOES PACKING HAVE LOWER FRICTION WITH INCREASED CYCLES &

TEMPERATURE ?

Revision 10 October 14, 2002 NO PMT1 DIAGNOSTIC AND/OR STROKE TIME TEST REQUIRED. LOCAL LEAK RATE TEST MAY ALSO BE REQUIRED.

PMT2 PERFORM IST STROKE TIME TEST, OR EVALUATE AND DEFER/WAIVE TESTING.

VISUAL PACKING LEAK CHECK MAY ALSO BE REQUIRED.

A10 - 13

ATTACHMENT 11 CORPORATE TECHNICAL POSITION INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 11 CORPORATE TECHNICAL POSITION INDEX Designator TP-CWE-IST-98-03 TP-EXE-IST-00-03 TP-EXE-IST-00-04 TP-EXE-IST-00-05 TP-EXE-IST-00-06 TP-EXE-IST-01-01 TP-EXE-IST-01-02 TP-EXE-IST-0 1-03 TP-EXE-IST-01-04 Description Elimination of Position Indication and Exercise Testing for Safety and Relief Valves Check Valve Open Exercise Testing During Pump Tests Classification of Skid Mounted Components Scheduling of Tests for Class 2 and Class 3 Relief Valves Cycling Check Valves during Normal Operations Satisfies Test Requirements Non-Safety Check Valve Exercise Testing by Normal Operations Thermal Relief Valve Scoping Justification for Exception to Exercise Check Valves after Reassembly Pump Testing Set Value Tolerance Range Revision 10 October 14, 2002 Approval Date 1/20/99 10/27/00 6/8/01 10/27/00 10/27/00 4/9/01 4/27/01 4/9/01 8/10/01 All - I

ATTACHMENT 12 CORPORATE TECHNICAL POSITIONS

TP-CWE-IST-98-03 Final Status January 20, 1999 Revision 0 ComEd IST Program Technical Position Elimination of PIT and Exercise Testing for Safety and Relief Valves Purpose The purpose of this technical position is to clarify that no Inservice Testing Program requirement exists for performing Position Indication Tests and Exercise tests of safety and relief valves at CornEd stations.

Background

Inservice Testing Programs at CornEd stations currently require that safety and relief valves be tested in accordance with both OM-10 (IVW-3000 for Dresden) and OM-1 requirements.

Consequently, Exercise Tests are performed on ComEd relief valves in accordance with OM-l10 IWV-3000. For valves with auxiliary operating devices, stroke times are measured when exercising the valves. In addition, Position Indication Tests are performed for safety and relief valves with remote position indication.

Position

• OM-10 (ISTB) defers to OM-1 (OM Code Appendix 1) for relief valve testing requirements.

O OM-I and later editions of the ASME OM Code do not require Exercising Tests for safety and relief valves.

• Exercising at a frequency greater than OM-l setpoint testing shall be performed when required by Tech Specs or when valve performance history indicates that exercising is needed r"

to keep the valve setpoint from increasing. However, stroke timing during exercising is not required, unless a commitment to measure stroke time has been made to the NRC.

OM-1 and later editions of the ASME OM Code do not require Position Indication Tests for safety and relief valves.

Justification

"* NUREG 1482, Section 4.3.9 states "As licensees began applying the requirements of OM-1, it became clear that clarifications were needed. The OM working group has clarified several issues in the 1994 addenda to the 1990 OM Code. The clarifications discussed below may be used without further NRC approval. Other clarifications identified by licensees may also be used without further NRC approval if it is determined to be clarification only and is documented in the IST program or test procedures, as necessary."

0 In the ASME OMa 1996 edition of the Code, a new paragraph was added at the end of Section ISTC section 1.2. This paragraph states, "Category A and B Safety and Relief valves are excluded from the requirements of ISTC 4.1, Valve Position Verification and ISTC 4.2, Inservice Exercising Test."

"* Summary of Public Workshops, Section 2.4.18 states, "Licensees should note that the OM Code has been revised (i.e., in the 1996 Addenda) to clarify that Category A and B safety and relief valves are excluded from the requirements of ISTC 4.1, Valve Position Verification, and ISTC 4.2, Inservice Exercising Test. Therefore, these valves will only be required to be tested in accordance with Appendix I. As discussed in NUREG-1482, Section 4.3.9, clarifications may be used without further NRC approval." (emphasis added)

TP-CWE-IST-98-03 Final Status January 20, 1999 Revision 0 0 A comparison of Appendix I in ASME OM 1995 with Appendix I in ASME OMa 1996 indicates that no new testing requirements were added as replacements for ISTC 4.1, Valve Position Verification and ISTC 4.2, Inservice Exercising Test. Consequently, it is appropriate to classify the subject 1996 code change as a clarification.

0 Stroke time measurements have not benefited ComEd in preventing safety and relief valve failures. In fact, exercising tests have been attributed to subsequent seat leakage failures, especially for fast stroking relief valves. Consequently, exercising is only appropriate when valve performance history indicates that exercising on a frequency greater than OM-1 setpoint testing is needed to prevent setpoint elevation due to adhesion between the seat and disk.

0 Position indication testing serves little purpose for valves that are not susceptible to mispositioning. Position indication testing at ComEd has not identified problems with stuck open relief valves being remotely indicated as closed.

References

1. ASME OM Code-1987 with OMa-1988, Part 10; "Inservice Testing of Valves in Light Water Reactor Power Plants"

2. ASME OM Code-1987, Part 1; "Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices"

3. ASME OM Code-1995, Subsection ISTC; "Inservice Testing of Valves in Light Water Reactor Power Plants" and Appendix-I; "Inservice Testing of Pressure Relief Devices in Light-Water Reactor Plants"

4. ASME OMa Code-1996, Subsection ISTC; "Inservice Testing of Valves in Light Water Reactor Power Plants" and Appendix I; "Inservice Testing of Pressure Relief Devices in Light-Water Reactor Plants"

5. Summary of Public Workshops Held in NRC Regions on Inspection Procedure 73756, "Inservice Testing of Pumps and Valves," And Answers to Panel Questions on Inservice Testing Issues; published July 18, 1997

6. NUREG-1482; "Guidelines for Inservice Testing of Nuclear Power Plants" Assumptions None Status Final Prepared by:

Is/

1/20/99 Brian D. Bunte Corporate IST Program Manager Reviewed by:

/s/

1/20/99 Jim Krueger Corporate Relief Valve Technical Expert

TP-EXE-IST-00-03 Final Status October 27, 2000 Revision 0 Exelon IST Program Technical Position Check Valve Open Exercise Testing During Pump Tests Purpose The purpose of this technical position is to justify the use of min flow test results to demonstrate that a check valve performs its design basis function, even when flow through the min flow line is not measured.

Background

Historically, ComEd stations performed disassembly and inspection examinations of some min flow check valves because instrumentation was not available to measure flow rate through min flow lines and because non-intrusive testing techniques were not capable of determining whether min flow valves fully opened. Other utilities have successfully monitored pump pressure during min flow testing to verify that check valves open sufficiently to preclude pump damage.

Position Open exercise test requirements for Min Flow Check Valves are satisfied by testing of associated pumps provided the following conditions are satisfied:

• Baseline pump hydraulic performance curves from pre-service or inservice testing are available. The pressure at the min flow condition for the pump must be measurably different from the pressure at the zero flow (dead head) condition.

• During pump testing, pressure measurements are taken while the min flow valve is open and all other flow paths are isolated

• Test acceptance criteria (including margin for pressure measurement inaccuracy) must demonstrate that the min flow requirement for the pump is achieved with only the min flow path open.

• Check valve testing requirements are documented in a Condition Monitoring Plan. This plan should reference all information used to establish acceptance criteria for the open exercise test.

Justification Disassembly and inspection requires opening potentially contaminated systems and significant time in the field. In accordance with ALARA principals, alternative test methods requiring less radiological exposure should be used if they provide equivalent assurance of component performance.

Degradation for min flow check valves is typically not a concern since these valves are infrequently subjected to flow conditions. If degradation concerns do exist, then the Condition Monitoring Plan developed in accordance with reference 1 would identify these concerns and would require testing / inspection that provides a means to monitor the degradation rate.

TP-EXE-IST-00-03 Final Status October 27, 2000 Revision 0 The health of the pump associated with the min flow check valve is on a quarterly basis by trending of vibration data and hydraulic data. Consequently, concerns with min flow valves providing adequate flow to prevent pump damage would also be observed as degradation in pump performance.

References

1.

Nuclear Engineering Standard NES-MS-08.5, "Condition Monitoring for Inservice Testing of Check Valves" Assumptions None Status Final Prepared by:

Reviewed by:

/s/

10/27/00 Brian D. Bunte IST Program Technical Specialist 10/27/00

/s/

David Sun Exelon IST Program Manager

TP-EXE-IST-00-04 Final Status June 8, 2001 Revision I Exelon IST Program Technical Position Classification of Skid Mounted Components Purpose The purpose of this technical position is to clarify requirements for classification of various components including Diesel Oil Transfer Pumps as skid mounted components, and to clarify testing requirements of check valves designated as skid mounted.

Background

The ASME Code allows classification of some components as skid mounted when their satisfactory operation is demonstrated by the performance of major components. Testing of the major component is sufficient to satisfy IST testing requirements for skid mounted components.

In the 1996a addenda to the ASME OM Code (endorsed by 10CFR50.55(a) in October 2000), the term skid-mounted was clarified by the addition of ISTA paragraph 1.7:

ISTA 1.7 Definitions Skid mounted components and component sub assemblies - components integral to or that support operation of major components, even though these components may not be located directly on the skid. In general, these components are supplied by the manufacturer of the major component. Examples include: diesel skid mounted fuel oil pumps and valves, steam admission and trip throttle valves for high-pressure coolant injection or auxiliary feedwater turbine-driven pumps, and solenoid-operated valve provided to control the air-operated valve.

This definition was further clarified in the 1998 ASME OM Code:

ISTA-2000 DEFINITIONS Skid mounted pumps and valves - pumps and valves integral to or that support operation of major components, even though these components may not be located directly on the skid. In general, these pumps and valves are supplied by the manufacturer of the major component. Examples include:

(a) diesel fuel oil pumps and valves; (b) steam admission and trip throttle valves for high-pressure coolant injection pumps; (c) steam admission and trip throttle valves for auxiliary feedwater turbine driven pumps; (d) solenoid-operated valves provided to control an air-operated valve.

In section 3.4 of NUREG 1482, the NRC supports the designation of components as skid mounted:

The staff has determined that the testing of the major component is an acceptable means for verifying the operational readiness of the skid-mounted and component subassemblies if the licensee documents this approach in the IST Program. This is acceptable for both Code class components and non-Code class components tested and tracked by the IST Program.

Subsection ISTC of OMa-1996, "Inservice Testing of Valves in Light-Water Reactor Power Plants", Paragraph 1.2, "Exclusions" states:

"....Skid-mounted valves and component subassemblies are excluded from this Subsection provided they are tested as part of the major component and are determined by the Owner to be adequately tested."

Page 1 of 2

TP-EXE-IST-00-04 Final Status June 8, 2001 Revision 1 r

Position The 1998 ASME OM Code definition of skid mounted should be used for classification of components in the Exelon Inservice Testing Program. In addition, for a component to be considered skid mounted:

+ The major component associated with the skid mounted component must be surveillance tested at a frequency sufficient to meet ASME OM Code test frequency for the skid mounted component.

• Satisfactory operation of the skid mounted component must be demonstrated by satisfactory operation of the major component.

• The IST Bases Document should describe the bases for classifying a component as skid mounted, and the IST Program Plan should reference this technical position for the component.

For Stations committed to the 1996 addenda of the 1995 OM Code for Inservice Exercise Testing of Category C Check Valves (ISTC 4.5 and Appendix II), testing as required by ISTC 4.5 does not apply for check valves designated as skid mounted.

Justification Classification of components as skid mounted eliminates the need for testing of sub components that are redundant with testing of major components provided testing of the major components demonstrates satisfactory operation of the "skid mounted" components.

As recognized in section 3.4 of NUREG 1482:

Various pumps and valves procured as part of larger component subassemblies are often not designed to meet the requirements for components in ASME code classes 1, 2, and 3.

References All references are called out in the text of the technical position.

Assumptions None Status Final Prepared by:

John Kowalski_6

/ 8 /01 John Kowalski, LaSalle IST Program Coordinator Reviewed by: _____David Sun_6/

8/01 David Sun, Exelon IST Program Manager Page 2 of 2

TP-EXE-IST-00-05 Final Status October 27, 2000 Revision 0 Exelon IST Program Technical Position Scheduling of Tests for Class 2 and Class 3 Relief Valves Purpose The purpose of this technical position is to provide the bases for deferring testing of Class 2 and Class 3 relief valves (replaced during refueling outages) until after resumption of power.

Background

During past outages, problems with sample expansion testing of IST relief valves have occurred.

Outage schedules would be significantly impacted my making system trains containing the sample expansion relief valves available for valve removal and testing.

Position When a partial complement of class 2 and 3 relief valves from a sample group are replaced during an outage, the ASME Code only requires that the removed valves be tested within 3 months of removal. This allows testing the valves after resumption of power. If a valve fails testing after resumption of power, sample expansion tests are not required until the next opportunity (typically the next refueling outage).

Testing removed valves after startup is only recommended if potential sample expansion tests cannot be performed without impacting the outage.

Justification Part 1 of the ASME OMa-1988 Code requires testing of Class 2 and 3 relief valves within 3 months of the valves being removed. This has been clarified in later versions of the code. For example, paragraph I1.3.5(b)(1) of the ASME OMa Code-1996 (endorsed by 10CFR50.55(a) in October 2000), states that "for replacement of a partial complement of valves, the valves removed from service shall be tested within 3 months of removal from the system or before resumption of electric power generation, whichever is later."

If sample expansion for a relief valve cannot be performed without impacting outage schedule, consideration should be given to deferring testing of the removed valve until after the unit is back on line. By doing this the outage schedule is protected since sample expansion is not required until the next practical opportunity.

Shutting down to perform sample expansion testing is clearly not practical. Therefore, sample expansion testing at the next outage is supported by the "Summary of Public Workshops Held in NRC Regions on Inspection Procedure 73756, 'Inservice Testing of Pumps and Valves,' and Answers to panel Questions on Inservice Testing Issues" (July 18, 1997). The NRC answer to Question 2.4.19 states the following about when additional valves must be tested:

The Code does not specify when the additional testing should be done when a relief valve fails the acceptance criteria. Licensees would be expected to perform testing as soon as practical.

TP-EXE-IST-00-05 Final Status October 27, 2000 Revision 0 C

Prompt causal evaluation of failures is still required to verify that the failure does not represent a potential common failure mode impacting other valves. Since performing causal evaluations may be difficult without sample expansion test results, deferring testing of removed valves until after startup is only recommended if sample expansion tests cannot be performed without impacting the outage schedule.

References References for this technical position are identified in the text above.

Assumptions None Status Final Prepared by:

/s/

10/27/00 Brian D. Bunte IST Program Technical Specialist Reviewed by:

/s/

10/27/00 David Sun Exelon IST Program Manager

TP-EXE-IST-00-06 Final Status October 27, 2000 Revision 0 (U

Exelon IST Program Technical Position Cycling Check Valves during Normal Operations Satisfies Test Requirements Purpose The purpose of this technical position is to describe the process used to credit normal operation for demonstrating that a check valve is capable of moving to its safety related or non-safety related position. This technical position only applies to plants implementing the ASME OMa Code-1996 (or later) requirements for IST check valve testing.

Background

In reference 1, CornEd committed to implementing the ASME OMa Code-1996 for inservice testing of check valves. This code includes the following major changes:

+ Check valve testing is required to demonstrate that valves are capable of moving to both the open and close position regardless of whether both positions serve safety-related functions.

Previous editions of the ASME Code only required that inservice tests demonstrate a check valve could be placed in its safety-related position(s).

• Operation of check valves during the course of plant operation satisfies the exercising requirements of the Code provided that certain constraints are met.

Condition Monitoring Plans in accordance with Appendix II of the ASME OM Code can be used in lieu of check valve test requirements in the ISTC section of the code.

Position Many check valves that only have'one safety related position (open or close), are needed in their non-safety related position to support normal operation of the plant. For these check valves, it is not necessary to perform a surveillance test that demonstrates the valve can move to its non-safety related position. However, the bases for knowing that the valve can move to its non-safety related position must be documented in the IST Program Plan or in a Condition Monitoring Plan.

The matrix below provides the requirements for crediting normal operation as satisfying IST check valve exercising requirements.

TP-EXE-IST-00-06 Final Status October 27, 2000 Revision 0 Condition Monitoring Plan No frequency that would satisfy the exercising requirements of this Subsection observations required for testing are made observations required for testing are analyzed observations required for testing are recorded frequency that would satisfy the exercising requirements of this Subsection observations required for testing are made observations required for testing are analyzed observations required for testing are recorded

.4-Yes frequency that would satisfy the exercising requirements of this Subsection observations required for testing are made observations required for testing are analyzed observations required for testing are recorded frequency that would satisfy the exercising requirements of this Subsection observations required for testing are made observations required for testing are analyzed observations required for testing are recorded Justification ASME OMa Code 1996 paragraph 4.5.3 (Valves in Regular Service) states, "Check valves that operate in the course of plant operation at a frequency that would satisfy the exercising requirements of this Subsection need not be additionally exercised if the observations otherwise required for testing are made and analyzed during such operation and are recorded in the plant records at intervals not greater than specified in para. ISTC 4.5.1."

Burden of additional tests frequency that would satisfy the exercising requirements of this Subsection observations required for testing are made observations required for testing are analyzed Safety Related Direction Yes No

TP-EXE-IST-00-06 Final Status October 27, 2000 Revision 0 observations required for testing are recorded References

1.

CornEd letter dated April 18, 2000 to the NRC, "Request to implement a Portion of the 1995 Edition and the 1996 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants Regarding Appendix II, "Check Valve Condition Monitoring Program"

2.

Nuclear Regulatory Commission letter dated June 7, 2000, "Approval to Implement a Check Valve Inservice Testing Program Using ASME OM Code-1995 Edition, OMa-1996 Addenda at the Commonwealth Edison Company Nuclear Stations (TAC Nos. MA8703, MA8704, MA8715, MA8716, MA8717, MA8718, MA8803, MA8804, MA8733, and MA8734)

Assumptions None Status Final Prepared by:

/s/

10/27/00 Brian D. Bunte IST Program Technical Specialist Reviewed by:

1s/

10/27/00 David Sun Exelon IST Program Manager

TP-EXE-IST-01-01 Final Status April 9, 2001 Revision 0 Exelon IST Program Technical Position Non-Safety Check Valve Exercise Testing By Normal Operations Purpose The purpose of this Technical Position is to establish the Company position for the verification of the non-safety exercise testing of check valves by normal plant operations.

This is applicable to check valves in the Inservice Testing (IST) Program as related to the ASME OMa Code-1996 Addenda to the ASME OMa Code-1995.

Applicability This Technical Position is NOT applicable to testing the safety function (position) of IST Check Valves. Safety function here means the function of the valve that meets a scoping requirement to be in the IST Program. This Technical Position is applicable to testing the non-safety function (position) of IST check valves. This Technical Position is applicable to check valves tested under Subsection ISTC, and to Appendix II (Condition Monitoring), of the ASME OMa Code-1996 Addenda.

Background

The ASME OMa Code-1996 Addenda in section ISTC 4.5.3, "Valves in Regular Use,"

states the following:

"Check valves that operate in the course of plant operation at a frequency that would satisfy the exercising requirements of this Subsection need not be additionally exercised if the observations otherwise required for testing are made and analyzed during such operation and are recorded in the plant records at intervals not greater than specified in para. ISTC 4.5.1."

Section 4.5.1 indicates that check valves shall be exercised nominally every 3 months with exceptions (for extended exercise periods) referenced.

Section 4.5.4 (2) states that, "Check valves that have a safety function in only the open direction shall be exercised by initiating flow and observing that the obturator has traveled to either the full open position or to the position required to perform its intended function(s) (see para. ISTC 1.1), and verify closure."

Section 4.5.4 (3) states that, "Check valves that have a safety function in only the close direction shall be exercised by initiating flow and observing that the obturator has traveled to at least the partially open position2,..."

Footnote 2 to this section indicates that the partially open position should correspond to the normal or expected system flow. NOTE: "Normal or expected," system flow rate may vary with plant conditions and configurations. The open safety function of a check valve usually requires meeting a specified, required limiting accident flow rate. As Operators are trained in recognizing normal plant conditions, Operator judgement is acceptable in ascertaining whether the non-safety open check valve position is providing normal or expected flow rates or plant conditions.

As stated in these two sections the non-safety ifunction is satisfactorily demonstrated by verifying closure, or passing normal or expected flow to verify opening, as applicable.

Position Verification of the non-safety position of IST check valves may be performed through the execution of a dedicated surveillance. Alternately this verification may be satisfied as follows:

• An appropriate means shall be determined which establishes how the open/closed non-safety function of the specified check valve is demonstrated during narmal operations. The position determination may be by direct indicator, or by other positive means such as changes in system pressure, flow rate, level, temperature, seat leakage, etc. This determination shall be documented in the respective Condition Monitoring Plan in the "Bases for Testing and Inspection Strategy," for valves in the Condition Monitoring Program. For check valves governed by Subsection ISTC and not in Condition Monitoring this determination shall be documented in the respective IST Bases Document valve group in the, "Bases Statement," section.

• Automated processes may be used to provide for the "observation and analysis," that a check valve is appropriately satisfying its' non-safety position function. An example of this would be a check valve that has a safety function in only the close direction and normally has flow through it to maintain normal plant operations. If the check valve is not opening to pass flow, alarms or indications would identify the problem to the Operator who is trained to respond to such situations and take appropriate actions. Condition Reports are normally written for abnormal plant conditions attributable to material condition concerns such as check valve failures.

• The "observation and analysis," of logs and other such records is satisfied by Operator reviews. Operating personnel are trained to look for off-normal data and adverse trends and take actions as appropriate. This would effectively determine if a check valve were satisfactorily fulfilling its' non-safety function.

+ The open/closed non-safety function shall be recorded at a periodicity required by ISTC 4.5.1, with exceptions as provided, in plant records such as Operator logs, Electronic Rounds, chart recorders, automated data loggers, etc. NOTE: The safety

function testing of these valves constitutes requiring a Quality Record. Records as indicated above are appropriate for the non-safety testing. Should any concerns arise regarding the material condition/operation of these check valves a Condition Report is written which is a Quality Record. The method in which the check valve position is recorded shall be included in the Condition Monitoring Plan or Bases Document sections as indicated above.

Justification This Technical Position requires that the method of determining the non-safety position be established. The plant systems and Operator actions provide for the observations and analysis that the valve is satisfying its' non-safety function. Finally, the recording of parameters demonstrating valve position is satisfied at a frequency specified in ISTC 4.5.1. These actions collectively satisfy demonstrating the non-safety position of IST check valves in regular use as required by ISTC 4.5.3.

Assumptions None Status Final Prepared by:

Reviewed by:

/S/

William A. Bielasco Byron Station IST Engineer

/s/

David Sun Exelon IST Program Manager

TP-EXE-IST-01-02 Final Status April 27, 2001 Revision 0 Exelon IST Program Technical Position Thermal Relief Valve Scoping Purpose The purpose of this technical position is to provide the bases for determining whether thermal relief valves should be included in the Inservice Testing (IST) Program.

Background

A thermal relief valve is a relief that protects the associated system from over pressurization due to thermal expansion. Whether these valves need to be included in the IST Program depends on the function of the system or subsystem they are in.

Position

"* If a systems or subsystem does NOT perform a required function in shutting down a reactor to the cold shutdown condition, in maintaining the cold shutdown condition, or in mitigating the consequences of an accident, then the thermal relief valve(s) in those systems/subsystems need not be placed in the IST Program.

"* If a system or subsystem performs a required function in shutting down a reactor to the cold shutdown condition, in maintaining the cold shutdown condition, or in mitigating the consequences of an accident, then the thermal relief valve(s) in those systems/subsystems shall be placed in the IST Program. Allowed exceptions to this requirement are the exclusion of valves which do not provide overpressurization protection to those systems (or portions thereof) as established by design requirements.

"* Plants whose licensing basis is to achieve Hot Standby need not include systems/components used to bring the reactor from hot standby to cold shutdown in their IST programs.

Justification ANSI/ASME OMa-1988, Part 10, "Inservice Testing of Valves in Light Water Reactor Power Plants," Section 1.1, "Scope," states the following:

"The pressure-relief devices covered are those for protecting systems or portions of systems which perform a required function in shutting down a reactor to the cold shutdown condition, in maintaining the cold shutdown condition, or in mitigating the consequences of an accident".

TP-EXE-IST-01-02 Final Status April 27, 2001 Revision 0 Exelon IST Program Technical Position Thermal Relief Valve Scoping NUREG - 1482 "Guidelines for Inservice Testing at Nuclear Power Plants", Section 4.3.1 states the following:

"The IST engineer may not have the documentation for the system design or development of the Section III overpressure analyses. However, if there are safety or relief valves that do not appear to perform a necessary safety or overpressure protection function, it may be possible to coordinate with a design engineering group for reanalyses. If the results of the overpressure protection "reanalysis" for a particular system indicate that a relief valve is not necessary, it may be removed from the scope of the IST program."

References ANSI/ASME OMa-1988, Part 10, "Inservice Testing of Valves in Light Water Reactor Power Plants" ANSI/ASME OM-1987, Part 1, "Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices."

"* NUREG 1482 Assumptions None Status Final Prepared by:

/s/

4/27/01 David Sun Exelon IST Program Manager Reviewed by:

/s/

4/27/01 John Cockroft Peach Bottom Station IST Engineer

TP-EXE-IST-01-03 Final Status April 9, 2001 Revision 0 Exelon IST Program Technical Position Justification for Exception to Exercise Check Valves after Reassembly Code Requirements The governing Code for this issue is found in the ASME OMa Code -1996 Addenda to ASME OM Code-1995, "Code for Operation and Maintenance of Nuclear Power Plants, Section ISTC, "Rules for Inservice Testing of Light-Water Reactor Power Plants."

Subsection ISTC, "Inservice Testing of Valves in Light-Water Reactor Power Plants,"

para. ISTC 6.2, "Test Plans," subpara. (e) requires documenting for check valves,

"...justification for not performing an exercise test to at least a partially open position after reassembly or periodic exercising in accordance with para. ISTC 4.5.2;"

Subsection ISTC, "Inservice Testing of Valves in Light-Water Reactor Power Plants,"

para.ISTC4.5.4 subpara. (c)(4) states, "Before return to service, valves that were disassembled for examination or that received maintenance that could affect their performance, shall be exercised if practicable (see nonmandatory Appendix J, Check Valve Testing Foll6wing Valve Reassembly)".

Discussion of Code Requirement Performing a partial open exercise after reassembly provides some assurance of the functionality of the check valve and that it has been installed in the proper flow direction.

Position There are numerous measures in place to assure that check valves are maintained properly and installed in the proper orientation and flow direction. As such there is justification to not exercise the check valves after reassembly.

Justification The following justifications demonstrate that exercising check valves after reassembly is unnecessary.

Match Marking: Match marking is the maintenance activity where the component (such as a check valve) and adjoining pipe section are marked adjacently. When the component is reinstalled it is done so the match marks align. This assures the component was reinstalled in the proper orientation/flow direction as when it was removed. The Nuclear Generation Group Maintenance Standards under "Expectations," regarding the execution of work states that parts should be match marked prior to disassembly to ensure proper orientation upon reassembly. Periodically a "Scorecard," which is a checklist

used for supervisory oversight to assure proper maintenance practices, is performed. The Scorecard has an item requiring assessment that components have been match marked prior to disassembly to help ensure proper reassembly. During practical exercises in the training of maintenance workers they are assessed to assure that they follow match marking practices where appropriate.

Procedures/Work Instructions: There are detailed procedures and work instructions to address that check valve maintenance, reassembly, and reinstallation is properly conducted.

Maintenance Oversight: Maintenance First Line Supervisors are expected to provide adequate oversight to assure the work is properly conducted Quality Assurance Program: A 10 CFR 50 Appendix B, Quality Assurance Program, is utilized at the stations to assure those quality standards are maintained.

Foreign Material Exclusion: A rigorous Foreign Material Exclusion program exists to assure no adverse impact to systems and components due to such intrusions. Keeping foreign material out of check valves assures they will not have their stroking or closure adversely impacted by those materials.

Training: Maintenance personnel are properly trained to assure they have the proper skills and follow procedures and instructions in working on plant components.

Condition Reports: Problems and concerns including those in the maintenance area are captured with Condition Reports. These are part of the corrective actions program to address such concerns.

Engineering Inspections: Engineering inspections are frequently performed to procedures and checklists to assure proper check valve maintenance and function. In addition to numerous checks to assess the material condition and functionality of the check valve, part of the Engineering inspection is to assess "as-found" and "as-left" manual full stroke capability.

Oversight Activities: Oversight activities by Quality Control, Nuclear Oversight, and other oversight organizations periodically review maintenance activities. This process helps assure the maintenance program is adequately functioning.

Applicability This Technical Position applies to all MWROG nuclear power plants except Clinton Station.

==

Conclusion:==

I There are adequate measures in place to justify that partially open testing check valves after reassembly need not be performed.

Assumptions None Status Final Prepared by:

Reviewed by:

Is/

William A. Bielasco Byron Station IST Engineer

/s/

David Sun Exelon IST Program Manager

TP-EXE-IST-01-04 Final Status August 10, 2001 Revision 0 Exelon IST Program Technical Position Pump Testing Set Value Tolerance Range Purpose The purpose of this Technical Position is to provide a method of determining an acceptable tolerance for the set value during IST pump testing.

Background

The purpose of Inservice Testing (IST) of pumps is to assess the operational readiness of the pumps, which fall into the scope of IST, as described in OM Part 6. This is accomplished by establishing reference values for flow, pressure, and vibration. The pump is operated at the nominal motor speed for constant speed drives and at a speed adjusted to the reference speed for variable speed drives. The resistance of the system is then varied until either the flow rate or pressure equals the reference value. When flow rate is the set value, the differential pressure is determined and compared to its reference value and the acceptance criteria specified in OM Part 6. When pressure is the set value, the flow rate is determined and compared to its reference value and the acceptance criteria specified in OM Part 6. Operational readiness may then be determined by the change (degradation) from the reference value of the non-set variable, which must be within the limits provided by OM Part 6. Additionally, vibration is also determined at various locations and is used to assess operational readiness by comparing the measurements to its reference value at each location and the acceptance criteria specified in OM Part 6. Where system resistance cannot be varied, both flow rate and differential pressure are determined and compared to their reference values and OM Part 6.

The OM Part 6 Code states that the set value will equal the reference value. It is recognized in the industry that it is often times impractical to exactly set and maintain for the duration of the test, an exact value. This impracticality is discussed in NUREG 1482 Guidelines for Inservice Testing at Nuclear Power Plants. NUREG 1482 was published in April 1995, and provides licensees guidelines and recommendations for developing and implementing IST Programs. As discussed in the Executive Summary of the NUREG, the guidance is voluntary. NUREG 1482 Section 5.3 states, "Certain designs do not allow for the licensee to set the flow at an exact value because of limitations in the instruments and controls for maintaining steady flow. The characteristics of piping systems in other designs do not allow for flow to be adjusted to exact values." NUREG 1482 Section 5.3 under NRC Recommendation states, "The staff has determined that, if the design does not allow for establishing and maintaining flow at an exact value, achieving a steady flow rate or differential pressure at approximately the set value does not require relief for establishing pump curves. The allowed tolerance for setting the fixed parameter must be established for each case individually including the accuracy of the instrument and the precision of its display. This will necessitate verification of the effect of precision on accuracy as considered in the design of the instrument gauge. A

TP-EXE-IST-01-04 Final Status August 10, 2001 Revision 0 total tolerance of +/-2 percent of the reference value is allowed without approval from the NRC. For a tolerance greater than +/- 2 percent (greater than +/- 2 percent may be necessary depending on the precision of the instrument), a corresponding adjustment to acceptance criteria may be made to compensate for the uncertainty, or an evaluation would be performed and documented justifying a greater tolerance. In using this guidance, the variance and the method for establishing the variance must be documented in the IST program documents or implementing procedures."

Published in the July 18, 1997 Summary Of Public Workshops Held In NRC Regions On Inspection Procedure 73756, "Inservice Testing Of Pumps And Valves," And Answers To Panel Questions On Inservice Testing Issues, there were questions related to this topic. Question 3.2.2 pertained to this issue of set value tolerance. The reply stated that the variance of+ 2% from the reference value is based on paragraph IWP - 4150 of Section XI. This paragraph allows symmetrical damping devices or averaging techniques to reduce instrument fluctuations to within + 2% of the observed readings. For a variance greater than + 2%, a corresponding adjustment to acceptance criteria may be made to compensate for the uncertainty, or an evaluation would be performed and documented justifying a greater tolerance. The variance in the acceptance criteria and the method used or establishing the variance must be documented in the IST program. Approval by the staff is not required.

The NRC recognized that many of the IST questions being raised at the Workshop were related to the ASME Code, versus questions related to the regulations or NRC requirements. These questions were referred back to the OM Code Working Groups.

Question 3.2.2 was referred to the Working Group. The reply from the Working Group is contained in the Proceedings of the Fifth NRC/ASME Symposium on Valve and Pump Testing (NUREG CP-0152 Vol. 2). The Working Group (WG) agreed with the NRC recommendation, but not the basis for the recommendation. The WG felt that while +2%

may be a reasonable tolerance for setting of the independent variable, they did not agree with the reason being based on damping devices and averaging techniques. The WG agreed that the tolerance in setting the reference value should be documented and justified; however, adjustments to the test acceptance criteria are not necessarily needed.

Additionally, the WG believed that the discussion in NUREG-1482, Section 5.3, that the allowed total tolerance for setting the fixed parameter should include consideration of the instrument accuracy, creates confusion. They mention that this was often interpreted to mean that the combination of the variation of the setting and the instrument accuracy at the reference value couldn't exceed +2%. They stated that when considering all the Code requirements pertaining to instruments, a combination of the variation and the instrument accuracy being within 2% may be virtually impossible to meet. The WG has assigned an action item to review the existing Code wording and revise it, if necessary, to provide a tolerance for establishing reference values during IST. The WG will also evaluate the need to include wording in the Code, which addresses the application of a tolerance greater than +2% of the reference value with consideration for adjusting the acceptance criteria commensurate with the existing NRC position. One solution discussed was to

TP-EXE-IST-01-04 Final Status August 10, 2001 Revision 0 simply change the wording of the Code to state that the setting of flow rate or differential pressure shall be as close as practical to the reference value.

Position For IST pump testing where resistance can be varied, the set parameter is to be set to the reference value. It is recognized that for many pump configurations within the system, it is often times extremely difficult and sometimes impossible to set and maintain the exact reference value for the duration of the test. This will often times require the set value to be set as close to the reference value as practical. The 1ST pump test procedures are to direct the test to set the set value to the target or reference value. The procedure may explain that if the exact value cannot be set or maintained, then it is acceptable to set the value as close as practical to the set value. The procedure may provide the range of what is practical for each pump. The range provided for each pump should be as close to the reference value as practical, but not to exceed +2% unless there is a documented evaluation.

If it is impractical to obtain a set value within +2% of the reference value, an evaluation must be performed, which justifies and documents the reasons for the greater than +2%

tolerance range. The evaluation is to take instrument accuracy into account and must justify why a lessor range is not practical and that with the range exceeding +/-2%, pump degradation may still be detected with the measured value. This additional justification may include observation of the pump curve (if applicable) or vendor information. The evaluation may also address other actions such as additional controls to assure more accurate testing results or adjustments to the acceptance criteria of the measured variable.

In situations where the pump is in a system for which the resistance cannot be varied, there is no set value. For these situations both the flow rate and differential pressure may be measured and compared to their reference values and the acceptance criteria of OM Part 6.

Justification This position meets the intent of the OM Code in that the set value is being set to the reference value. As recognized by the industry documents discussed in the Background Section, for most system configurations it would require an excessive amount of time and multiple cycling of throttle valves to set, and maintain for the test duration, a flow rate or pressure exactly equal to the reference value. As a result, the position allows the pump test procedures to set this as close as practical. A tolerance range is provided in the procedure to provide clear directions as to what is as close as practical for those performing the test. The tolerance range is specified as applicable to each pump; no further evaluation is required if this tolerance range does not exceed +2%. Gauge accuracy does not have to be taken into account when using this method. Various methods of addressing this issue including gauge accuracy considerations were contained

TP-EXE-IST-01-04 Final Status August 10, 2001 Revision 0 in the documents discussed in the Background Section. The NRC recognized that this issue was ASME Code related, and as a result forwarded this issue to the OM Code Main Committee. Only ASME can provide binding interpretations of the ASME Code. The OM Working Group substantiated that the intent of the Code was to set the set value as close as practical, 2% was a reasonable tolerance, and they did not require that gauge accuracy be taken into account in regard to this issue. Exceeding the +2% tolerance from the set value may affect the ability to adequately detect pump degradation, as a result using a greater than 2% tolerance range requires an evaluation which takes gauge accuracy into account and justifies the acceptability of the greater than +2% tolerance.

When using the greater than 2% tolerance, other considerations such as additional controls to assure more accurate testing results or adjustments to the acceptance criteria of the measured variable are considered.

The position is consistent with the guidance from the OM Working Group and is consistent with the methodology used in the industry. As the OM Code Part 6 recognizes there are some configurations for which pump flow cannot be set, guidance for such fixed resistance testing is provided in accordance with the OM Code Part 6.

References All references are called out in the text of the technical position.

Assumptions None Status Final Prepared by:

ISN Don Zebrauskas, Braidwood IST Program Engineer Reviewed by:

/S/

David Sun, Exelon IST Program Manager

ATTACHMENT 13 INSERVICE TESTING PUMP TABLE INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 13 INSERVICE TESTING PUMP TABLE INDEX System

System Description

Number 531 (03)

Service Water 532 (03)

Emergency Service Water 541 (05)

Reactor Building Closed Cooling Water 424 (11)

Condensate Transfer 251 (18)

Fuel Pool Cooling (including Augmented Fuel Pool Cooling) 213 (19)

Standby Liquid Control (Liquid Poison) 212 (20)

Core Spray 241 (21)

Containment Spray NOTE: Numbers in parentheses in the "System Number" column are the unique system designators used in the individual component ID numbers. Pumps in the following table are sorted on these numbers.

A13 - 1 Revision 10 October 14, 2002

ATTACHMENT 14 INSERVICE TESTING PUMP TABLE

ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION P&ID SHT PUMP TYPE DRIVER GROUP TEST RR#

COMMENTS SERVICE WATER P-03-001A SERVICE WATER PUMP A (1-1)

BR 2005 02 VLS MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE P-03-001B SERVICE WATER PUMP B (1-2)

BR 2005 02 VLS MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE EMERGENCY SERVICE WATER P-03-3A EMERGENCY SERVICE (52A)

WATER PUMP A P-03-3B EMERGENCY SERVICE (52B)

WATER PUMP B BR 2005 04 VLS MOTOR BR 2005 04 VLS MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE Page A14 - I of 7 RP-04 N/A N/A N/A RP-04 N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002

ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION EMERGENCY SERVICE WATER P&ID SHT PUMP TYPE DRIVER GROUP TEST RR# COMMENTS P-03-3C EMERGENCY SERVICE (52C)

WATER PUMP C BR 2005 04 VLS MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE P-03-3D EMERGENCY SERVICE (52D)

WATER PUMP D BR 2005 04 VLS MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE REACTOR BUILDING CLOSED COOLING P-05-1 (1-RBCCW PUMP NO. I 1)

P-05-2 (1-RBCCW PUMP NO. 2 2)

BR 2006 01 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE BR 2006 01 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A14 - 2 of 7 N/A N/A N/A N/A N/A N/A

ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION P&ID SHT PUMP TYPE DRIVER GROUP TEST RR#

COMMENTS CONDENSATE TRANSFER P-1i-l (1-CONDENSATE TRANSFER

1)

PUMP NO. 1 P-11-2 (1 2)

CONDENSATE TRANSFER PUMP NO. 2 BR 2004 BR 2004 02 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE 02 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE FUEL POOL COOLING FUEL POOL COOLING PUMP A GE 237E756 01 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE FUEL POOL COOLING PUMP B GE237E756 01 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE Page A14 - 3 of 7 N/A N/A N/A N/A N/A N/A N/A N/A P-18-IA (NN01-A)

P-1 8-1B (NN01-B)

N/A N/A N/A N/A Revision 10 10/14/2002 NIA N/A N/A N/A

d,

ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION P&ID SHT PUMP TYPE DRIVER GROUP TEST RR#

COMMENTS AUGMENTED FUEL POOL COOLING AUGMENTED FUEL POOL COOLING PUMP C GE 237E756 01 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE N/A N/A N/A N/A This pump provides heat removal when full core offload is performed, and will be tested only when it is required to be used.

P-18-1D AUGMENTED FUEL POOL (NN01-D)

COOLING PUMP D GE237E756 01 C

MOTOR A

DIFFERENTIAL PRESSURE FLOW VIBRATION COMPREHENSIVE N/A N/A N/A N/A This pump provides heat removal when full core offload is performed, and will be tested only when it is required to be used.

STANDBY LIQUID CONTROL P-19-001A LIQUID POISON PUMP A (NP-02A)

P-19-OO1B LIQUID POISON PUMP B (NP-02B)

GE 148F723 01 GE 148F723 01 PD MOTOR B

FLOW COMPREHENSIVE PD MOTOR B

FLOW COMPREHENSIVE Page A14 - 4 of 7 P-18-1C (NN01-C)

N/A N/A N/A N/A Revision 10 10/14/2002

ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION P&ID SHT PUMP TYPE DRIVER GROUP TEST RR#

COMMENTS CORE SPRAY CORE SPRAY PUMP A GE 885D781 01 C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE CORE SPRAY PUMP B CORE SPRAY PUMP C CORE SPRAY PUMP D GE 885D781 01 GE 885D781 01 GE 885D781 01 C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE Page A14 - 5 of 7 P-20-1A (NZO1-A)

N/A N/A N/A P-20-1 B (NZO1-B)

P-20-1 C (NZ01-C)

P-20-1 D (NZO1-D)

N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002

4, ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION P&ID SHT PUMP TYPE DRIVER GROUP TEST RR#

COMMENTS CORE SPRAY P-20-2A CORE SPRAY BOOSTER (NZ03-A)

PUMP A P-20-2B CORE SPRAY BOOSTER (NZ03-B)

PUMP B GE 885D781 01 GE 885D781 01 C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE P-20-2C CORE SPRAY BOOSTER (NZ03-C)

PUMP C GE 885D781 01 C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE P-20-2D CORE SPRAY BOOSTER (NZ03-D)

PUMP D GE 885D781 01 C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE Page A14 -6 of 7 N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 N/A N/A N/A

ATTACHMENT 14 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM PUMP TABLE PUMP ID DESCRIPTION P&ID SHT PUMP TYPE DRIVER GROUP TEST RR#

COMMENTS CONTAINMENT SPRAY P-21-1A CONTAINMENT SPRAY PUMP (51A)

A GE 148F740 01 P-21-1B CONTAINMENT SPRAY PUMP GE 148F740 01 (51B)

B P-21-1C CONTAINMENT SPRAY PUMP GE 148F740 01 (51C)

C P-21-1D CONTAINMENT SPRAY PUMP GE 148F740 01 (51D)

D C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE C

MOTOR B

DIFFERENTIAL PRESSURE FLOW COMPREHENSIVE Revision 10 10/14/2002 Page A14 - 7 of 7 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 15 INSERVICE TESTING VALVE TABLE INDEX

OYSTER CREEK NUCLEAR GENERATING STATION Inservice Testing Program Plan ATTACHMENT 15 INSERVICE TESTING VALVE TABLE INDEX System Number 411 (01) 422 (02) 531 (03) 532 (03) 541 (05) 852 (06) 811 (09) 424 (11) 211 (14) 225 (15) 215 (16) 214 (17) 251 (18) 213 (19) 212 (20) 241 (21) 573 (22) 242 (23) 551 (24) 243 (26) 822 (27) 822 (28) 216 (31) 223 (37) 666 (38) 555 (40) 622 (130) 623 (623)

System Description

Main Steam Feedwater Service Water Emergency Service Water Reactor Building Closed Cooling Water Instrument Air Fire Protection Condensate Transfer Isolation Condenser Control Rod Drive Cleanup Shutdown Cooling Fuel Pool Cooling (including Augmented Fuel Pool Cooling)

Standby Liquid Control (Liquid Poison)

Core Spray Containment Spray Drywell Floor and Equipment Drains Containment Inerting Reactor Sample Drywell and Suppression Reactor Building Ventilation Reactor Building Ventilation Head Cooling Recirculation Hydrogen/Oxygen Monitoring Post Accident Sampling Reactor Vessel Instrumentation Traveling Incore Probe NOTE: Numbers in parentheses in the "System Number" column are the unique system designators used in the individual component ID numbers. Valves in the following table are sorted on these numbers.

A15-1 Revision 10 October 14, 2002

ATTACHMENT 16 INSERVICE TESTING VALVE TABLE

• 1 It ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

411 VALVE ID MAIN STEAM DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAIUSAFE PAS TEST FRQ RR JUST COMMENTS V-01-007 NORTH HEADER INBOARD MSIV 24 GL AO (NS03A)

V-01-008 SOUTH HEADER INBOARD MSIV 24 GL AO (NS03B)

V-01-009 NORTH HEADER OUTBOARD (NS04A)

MSIV V-01-010 SOUTH HEADER OUTBOARD (NS04B)

MSIV V-01-160 SOUTH HEADER SAFETY (28D) 24 GL AO 24 GL AO 6x8 RV SA BR2002 01 B-5 1

10 C

C BR 2002 01 B-6 1

BR 2002 02 G-7 1 BR 2002 02 G-5 1

BR 2002 01 F-7 1

A A

LTJ AJ N/A N/A SP M3 N/A N/A SC CS N/A CS-07 FC RR N/A RJ-01 PI Y2 RV-51 N/A O

C C

A A

LTJ AJ N/A N/A SP M3 N/A N/A SC CS N/A CS-07 FC RR N/A RJ-01 PI Y2 RV-51 N/A O

C C

A A

LTJ AJ N/A N/A SP M3 N/A N/A SC CS N/A CS-07 FC RR N/A RJ-01 PI Y2 RV-51 N/A O

C C

A A

LTJ AJ N/A N/A SP M3 N/A N/A SC CS N/A CS-07 FC RR N/A RJ-01 PI Y2 RV-51 N/A C

0 C

A RT Y5 N/A N/A Revision 10 10/14/2002 Page A16 - 1 of 53

C.,

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

411 VALVE ID MAIN STEAM DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL TYPE TYPE POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS V-01-161 SOUTH HEADER SAFETY (28E)

V-01-162 SOUTH HEADER SAFETY (28F)

V-01-163 SOUTH HEADER SAFETY (28G)

V-01-164 NORTH HEADER SAFETY (28H)

V-01-165 NORTH HEADER SAFETY (28J)

V-01-166 NORTH HEADER SAFETY (28K)

V-01-167 NORTH HEADER SAFETY (28L)

V-01-168 NORTH HEADER SAFETY 6x8 RV SA 6x8 RV SA 6x8 RV SA 6x8 RV SA 6 x 8 RV SA 6x8 RV SA 6x8 RV SA 6x8 RV SA (28M)

BR 2002 01 F-7 1

BR 2002 01 E-7 1

BR 2002 01 E-7 1

BR 2002 01 D-4 1

BR 2002 01 E-4 1

BR 2002 01 E-4 1

BR 2002 01 F-4 1

BR 2002 01 F-4 1

Revision 10 10/1412002 Page A16 - 2 of 53 C

0 C

A RT Y5 N/A C

0 C

A RT Y5 N/A C

0 C

A RT Y5 N/A o

0 C

A RT Y5 N/A C

0 C

A RT Y5 N/A C

0 C

A RT Y5 N/A C

0 C

A RT Y5 N/A C

0 C

A RT Y5 N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

411 VALVE ID MAIN STEAM DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

NORM/FAIUSAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-01-173 ADS ELECTROMATIC RELIEF (108A)

VALVE V-01-174 ADS ELECTROMATIC RELIEF (108B)

VALVE V-01-175 ADS ELECTROMATIC RELIEF (108C)

VALVE V-01-176 ADS ELECTROMATIC RELIEF (108D)

VALVE V-01-177 ADS ELECTROMATIC RELIEF (108E)

VALVE 6x8 GL DF 6 x8 GL DF 6x8 GL DF 6x8 GL DF 6x8 GL DF BR2002 01 E-7 1

C C

0 B

A RT Y5 N/A BR2002 01 E-7 1

C C

0 B

A RT Y5 N/A BR2002 01 F-5 1

C C

0 B

A RT Y5 N/A BR2002 01 F-5 1

C C

0 B

A RT Y5 N/A BR2002 01 F-6 1

C C

0 B

A RT Y5 N/A V-01-190 NORTH HDR EMRV DISCHARGE 4

VACUUM BREAKER V-01-191 NORTH HDR EMRV DISCHARGE 4

VACUUM BREAKER V-01-192 SOUTH HDR EMRV DISCHARGE 4

VACUUM BREAKER CK SA BR2002 01 D-5 2

C 0

C A

CO RR N/A CK SA BR2002 01 C-5 2 C

-0 C

A CO RR N/A CK SA BR2002 01 C-6 2

C

-0 C

A CO RR N/A Revision 10 10/14/2002 RJ-02 RJ-02 RJ-02 Page A16 - 3 of 53 N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

411 VALVE ID MAIN STEAM DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS V-01-193 SOUTH HDR EMRV DISCHARGE 4

VACUUM BREAKER CK SA BR2002 01 D-6 2 C

0 C

A CO RR N/A RJ-02 Revision 10 10/1412002 Page A16 - 4 of 53 4

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

422 FEEDWATER VALVE ID DESCRIPTION V-02-071 FEEDWATER CHECK V-02-072 FEEDWATER CHECK V-02-073 FEEDWATER CHECK V-02-074 FEEDWATER CHECK SIZE VLV ACTU TYPE TYPE 18 CK SA 18 CK SA 18 CK SA 18 CK SA P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAIL/SAFE PAS TEST FRQ RR JUST COMMENTS BR 2003 01 G-2 1

BR 2003 01 G-1 1

BR 2003 01 G-2 1

BR 2003 01 G-1 1

O C

AC A

O C

AC A

O C

AC A

O C

AC A

LTJ AJ N/A N/A CC RR N/A RJ-03 LTJ AJ N/A N/A CC RR N/A RJ-03 LTJ AJ N/A CC RR N/A LTJ AJ CC RR N/A RJ-03 N/A N/A N/A RJ-03 Revision 10 10/14/2002 Page A16 - 5 of 53

I, ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

531 SERVICE WATER DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAILISAFE PAS TEST FRQ RR JUST COMMENTS V-03-062 P-3-001A DISCHARGE CHECK V-03-063 P-3-001B DISCHARGE CHECK 16 CK SA BR2005 02 F-8 3

O/C O/C C

A CCF Y1 N/A CO M3 N/A 16 CK SA BR2005 02 F-7 3

O/C O/C C

A CCF Y1 N/A CO M3 N/A Revision 10 10/14/2002 N/A N/A N/A NIA Page A16 - 6 of 53 VALVE ID

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

532 EMERGENCY SERVICE WATER SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

NORMIFAIL/SAFE PAS TESTING REQUIREMENTS TEST FRO RR JUST COMMENTS V-03-065 ESW PUMP P-3-3D DISCHARGE 10 CK SA BR 2005 04 C-7 3 CHECK V-03-066 ESW PUMP P-3-3C DISCHARGE 10 CK SA BR 2005 04 C-8 3 CHECK V-03-067 ESW PUMP P-3-3B DISCHARGE 10 CK SA BR 2005 04 G-8 3 CHECK V-03-068 ESW PUMP P-3-3A DISCHARGE 10 CK SA BR 2005 04 G-7 3 CHECK V-03-082 CONTAINMENT SPRAY HX H.75 x 1 RV SA BR 2005 04 F-4 3

1B TUBE SIDE RELIEF V-03-083 CONTAINMENT SPRAY HX H.75 x 1 RV SA BR 2005 04 H-4 3

1A TUBE SIDE RELIEF V-03-084 CONTAINMENT SPRAY HX H.75 x 1 RV SA BR 2005 04 E-4 3

IC TUBE SIDE RELIEF V-03-085 CONTAINMENT SPRAY HX H.75 x 1 RV SA BR 2005 04 C-4 3

1D TUBE SIDE RELIEF C

O/C C

A CC M3 N/A CO M3 N/A C

O/C C

A CC M3 N/A CO M3 N/A C

O/C C

A CC M3 N/A CO M3 N/A C

O/C C

A CC M3 N/A CO M3 N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A Revision 10 10/14/2002 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Page A16 - 7 of 53 VALVE ID DESCRIPTION I

'I, ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

532 EMERGENCY SERVICE WATER DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORMIFAILISAFE PAS TEST FRO RR JUST COMMENTS V-03-087 HEAT EXCHANGERS H-21 -1C &

1D OVERBOARD ISOLATION V-03-088 HEAT EXCHANGERS H-21-1A &

1B OVERBOARD ISOLATION V-03-131 SW TO ESW B LOOP CHECK V-03-133 SW TO ESW A LOOP CHECK 14 BTF M

14 BTF M

BR 2005 04 C-2 3

BR 2005 04 F-2 3

2 CK SA BR2005 04 D-7 3 2

CK SA BR2005 04 E-7 3

LT Al T

B P

PI Y2 N/A N/A LT Al T

B P

PI Y2 N/A N/A O

C C

A CC M3 N/A N/A COD Y1 N/A N/A O

C C

A CC M3 N/A N/A COD Y1 N/A N/A Revision 10 10/14/2002 Page A16 - 8 of 53 VALVE ID

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

541 CLOSED COOLING WATER VALVE ID DESCRIPTION V-05-147 RBCCW SUPPLY CONTAINMENT ISOLATION V-05-153 P-5-1 DISCHARGE CHECK V-05-154 P-5-2 DISCHARGE CHECK SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS NORMIFAIL/SAFE CAT ACTI TESTING REQUIREMENTS PAS TEST FRQ RR JUST COMMENTS 6

GA MO BR2006 03 B-5 2

0 Al C

A A

LTJ SC PI AJ CS Y2 N/A N/A RV-51 12 CK SA BR2006 01 C-5 3

O/C O/C C

A CCF Y1 N/A CO M3 N/A N/A CS-01 N/A N/A N/A 12 CK SA BR2006 01 B-6 3

O/C O/C C

A CCF Y1 N/A N/A CO M3 N/A N/A V-05-165 RBCCW SUPPLY CONTAINMENT ISOLATION CHECK V-05-166 RBCCW RETURN CONTAINMENT ISOLATION V-05-167 RBCCW RETURN CONTAINMENT ISOLATION 6

CK SA BR2006 03 B-5 2

0 C

AC A

LTJ AJ N/A CC RR N/A 6

GA MO BR2006 03 B-3 2

O AI C

A A

LTJ SC PI AJ CS Y2 6

GA MO BR2006 03 B-3 2

0 Al C

A A

LTJ AJ SC CS PI Y2 N/A N/A RV-51 N/A N/A RV-51 Revision 10 10/14/2002 N/A RJ-04 N/A CS-01 N/A N/A CS-01 N/A Page A16 - 9 of 53

eI¢ ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

541 VALVE ID CLOSED COOLING WATER DESCRIPTION V-05-879 CIV V-5-166 THERMAL RELIEF CHECK VALVE SIZE VLV ACTU TYPE TYPE

.375 CK SA P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS BR 2006 03 B-3 2

C 0

AC A

LTJ AJ N/A CCF Y4 N/A COF Y4 N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 10 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

852 INSTRUMENT AIR VALVE ID DESCRIPTION V-06-393 IA CONTAINMENT ISOLATION CHECK V-06-395 IA CONTAINMENT ISOLATION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS 2

CK SA NC 0-C AC A

LTJ AJ N/A N/A CC RR N/A RJ-05 2

GA AO NC 0

C C

A A

LTJ AJ N/A N/A SC CS N/A CS-02 FC CS N/A CS-02 PI Y2 RV-51 N/A Revision 10 10/14/2002 Page A16-11 of 53 I

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

811 VALVE ID FIRE PROTECTION DESCRIPTION V-09-2099 FIRE WATER MAKEUP TO ISOLATION CONDENSERS SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORM/FAII/SAFE PAS TEST FRO RR JUST COMMENTS 3

DIA M

BR2004 3

C Al 0

B A

SO M3 N/A N/A Revision 10 10/14/2002 Page A16 - 12 of 53 4,

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

424 CONDENSATE TRANSFER DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS NORM/FAILISAFE CAT ACT/

PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-1 1-003 P-1 1-1 MIN FLOW RECIRC CHECK 2

CK SA BR 2004 02 D-9 3 O/C 0

C A

CO M3 N/A N/A Continued pump operation confirms open position.

V-1 1-007 P-1 1-2 MIN FLOW RECIRC CHECK 2

CK SA BR2004 02 D-9 3

O/C 0

C A

CO M3 N/A N/A Continued pump operation confirms open position.

V-11-012 P-11-1 DISCHARGE CHECK V-1 1-013 P-1 1-2 DISCHARGE CHECK 3

CK SA BR2004 02 D-8 3

O/C O/C C

A CC M3 CO M3 3

CK SA BR2004 02 C-7 3

O/C O/C C

A CC M3 CO M3 V-1 1-033 ISOLATION CONDENSER NE01 B MAKEUP CHECK V-1 1-034 ISOLATION CONDENSER NE01 B MAKEUP ISOLATION V-1 1-035 ISOLATION CONDENSER NE01 A MAKEUP CHECK 2.5 CK SA 2.5 GL AO 2.5 CK SA BR2004 02 B-1 3

C 0

C A

CO M3 N/A N/A BR2004 02 B-1 3

C 0

B A

SO M3 N/A N/A PI Y2 N/A N/A BR2004 02 B-1 3

C 0

C A

CO M3 N/A N/A Revision 10 10114/2002 Page A16 - 13 of 53 VALVE ID N/A N/A N/A N/A N/A N/A N/A N/A

4' ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

424 CONDENSATE TRANSFER DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS NORM/FAIUSAFE CAT ACT!

TESTING REQUIREMENTS PAS TEST FRO RR JUST COMMENTS V-1i1-036 ISOLATION CONDENSER NE01 A MAKEUP ISOLATION V-11-042 ISOLATION CONDENSERS MAKEUP CHECK V-11-049 FIRE WATER MAKEUP TO ISOLATION CONDENSERS V-11-116 COND XFER TO CORE SPRAY KEEP-FULL V-1 1-117 COND XFER TO CORE SPRAY KEEP-FULL 2.5 GL AO BR2004 02 B-1 3

C 0

B A

SO M3 N/A PI Y2 N/A N/A N/A 3

CK SA BR2004 02 B-9 3

C

-0 C

A CO M3 N/A N/A 3

DIA M

BR2004 02 B-9 3

C Al 0

B A

SO M3 N/A N/A 1

CK SA BR2004 2

C-6 3

C C

C A

CCD Y4 N/A N/A 1

CK SA BR2004 2

A-8 3

C C

C A

CCD Y4 N/A N/A Revision 10 10/14/2002 VALVE ID Page A16 - 14 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

211 ISOLATION CONDENSER VALVE ID DESCRIPTION V-14-001 NE01-B VENT TO MAIN STEAM HEADER A V-14-005 NE01-A VENT TO MAIN STEAM HEADER B V-14-019 NE01-B VENT TO MAIN STEAM HEADER A V-14-020 NE01-A VENT TO MAIN STEAM HEADER B V-14-030 RV STEAM SUPPLY TO EMERGENCY CONDENSER NE01-A SIZE VLV ACTU P&ID SH CRD CL TYPE TYPE

.75 GL AO GE 148F262 01 E8 2

.75 GL AO GE 148F262 01 F5 2

.75 GL AO GE 148F262 01 F8 2

.75 GL AO GE 148F262 01 F5 2

10 GA MO GE 148F262 01 G5 1

POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAIL/SAFE PAS TEST FRQ RR JUST COMMENTS O

C C

A A

LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 N/A O

C C

A A

LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 N/A O

C C

A A

LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 N/A O

C C

A A

LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 N/A 0

Al O/C B

A SC M3 N/A PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 15 of 53

of ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

211 ISOLATION CONDENSER DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAILISAFE PAS TEST FRO RR JUST COMMENTS V-14-031 RV STEAM SUPPLY TO EMERGENCY CONDENSER NE01-A V-14-032 RV STEAM SUPPLY TO EMERGENCY CONDENSER NE01-B V-14-033 RV STEAM SUPPLY TO EMERGENCY CONDENSER NE01-B 10 GA MO GE 148F262 01 G5 1

10 GA MO GE 148F262 01 G4 1

10 GA MO GE 148F262 01 G4 1

"O Al O/C B

A "O Al O/C B

A "O Al O/C B

A SC M3 N/A PI Y2 N/A SC M3 N/A PI Y2 N/A SC M3 N/A PI Y2 N/A V-14-034 NE01-A CONDENSATE RETURN 10 GA MO GE 148F262 01 E4 1

TO RECIRC LOOP A V-14-035 NE01-B CONDENSATE RETURN 10 GA MO GE 148F262 01 E4 1

TO RECIRC LOOP E V-14-036 NE01-A CONDENSATE RETURN 10 GA MO GE 148F262 01 B4 1

TO RECIRC LOOP A V-14-037 NE01-B CONDENSATE RETURN 10 GA MO GE 148F262 01 C4 1

TO RECIRC LOOP E C

Al O/C B

A C

Al O/C B

A O

Al O/C B

A O

Al O/C B

A SC M3 N/A SO M3 N/A P1 Y2 N/A SC M3 N/A SO M3 N/A P1 Y2 N/A SC M3 N/A P1 Y2 N/A SC M3 N/A PI Y2 N/A Revision 10 10/14/2002 Page A16 - 16 of 53 VALVE ID N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

211 ISOLATION CONDENSER VALVE ID DESCRIPTION V-14-162 V-14-037 BYPASS CHECK V-1 4-165 V-14-036 BYPASS CHECK SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAIUSAFE PAS TEST FRO RR JUST COMMENTS

.375 CK SA GE 148F262 01 C4 1

.375 CK SA GE 148F262 01 B4 1

C 0

C A

COD Y4 N/A CCD Y4 N/A C

0 C

A COD Y4 N/A CCD Y4 N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 17 of 53 4

.7 ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

225 CONTROL ROD DRIVE VALVE ID DESCRIPTION V-15 (106)

CHARGING WATER ACCUMULATOR CHECK (TYP OF 137)

V-15 (108)

CRD DISCHARGE HEADER CHECK (TYP OF 137)

V-15 (126)

CRD SCRAM INSERTION VALVE (TYP OF 137)

V-15 (127)

CRD SCRAM EXHAUST VALVE (TYP OF 137)

V-15 (138)

CRD COOLING WATER SUPPLY CHECK (TYP OF 137)

V-15-027 CRD PUMPS TO RPV CONTAINMENT ISOLATION V-15-028 CRD PUMPS TO RPV CONTAINMENT ISOLATION SIZE VLV ACTU P&ID SH CRD CL TYPE TYPE

.5 SCK SA GE 197E871 01 D-2 2

.75 SCK SA GE 197E871 01 E-6 2

1 GL AO GE 197E871 01 D-3 2

.75 GL AO GE 197E871 01 E-5 2

.5 CK SA GE 197E871 01 F-3 2

3 CK SA GE 237E487 01 D-1 1

3 CK SA GE 237E487 01 D-1 1

POSITIONS CAT NORM/FAILISAFE ACT/

TESTING REQUIREMENTS PAS TEST FRO RR JUST COMMENTS C

O/C AC A

LTH Y2 N/A N/A CC CS N/A CS-03 C

0 C

A CO RR N/A RJ-10 C

0 0

B A

SO RR FO RR PI Y2 C

0 0

B A

SO RR FO RR P1 Y2 N/A RJ-09 N/A RJ-09 N/A N/A N/A RJ-09 N/A RJ-09 N/A N/A O

C C

A CC M3 N/A N/A O

C AC A

LTJ AJ N/A N/A CC RR N/A RJ-06 O

C AC A

LTJ AJ N/A N/A CC RR N/A RJ-06 Revision 10 10/14/2002 Page A16 - 18 of 53 4

. I ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

225 VALVE ID CONTROL ROD DRIVE DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORMIFAII/SAFE PAS TEST FRQ RR JUST COMMENTS V-15-119 NORTH SDV VENT V-15-120 SOUTH SDV VENT V-15-121 SOUTH SDV DRAIN V-15-133 NORTH SDV DRAIN V-15-134 SOUTH SDV DRAIN 1

GA AO GE 197E871 01 F-9 2

1 GA AO GE 197E871 01 F-8 2

2 GA AO GE 197E871 01 C-7 2 2

GA AO GE 197E871 01 C-9 2 2

GA AO GE 197E871 01 C-7 2 O

C C

B A

SP M3 N/A SC CS N/A FC CS N/A PI Y2 N/A O

C C

B A

SP M3 N/A SC CS N/A FC CS N/A PI Y2 N/A O

C C

B A

SP M3 N/A SC CS N/A FC CS N/A PI Y2 N/A O

C C

B A

SP M3 N/A SC CS N/A FC CS N/A PI Y2 N/A O

C C

B A

SP M3 N/A SC CS N/A FC CS N/A PI Y2 N/A Revision 10 10/14/2002 Page A16 - 19 of 53 N/A CS-04 CS-04 N/A N/A CS-04 CS-04 N/A N/A CS-04 CS-04 N/A N/A CS-04 CS-04 N/A N/A CS-04 CS-04 N/A j

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

225 CONTROL ROD DRIVE VALVE ID DESCRIPTION V-15-135 NORTH SDV DRAIN V-15-136 NORTH SDV VENT V-15-137 SOUTH SDV VENT SIZE VLV ACTU P&ID TYPE TYPE 2

GA AO GE 197E871 01 C-9 2

1 GA AO GE 197E871 01 F-9 2

1 GA AO GE 197E871 01 F-8 2

SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRQ RR JUST COMMENTS O

C C

B A

SP M3 SC CS FC CS PI Y2 O

C C

B A

SP M3 SC CS FC CS PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A O

C C

B A

SP M3 N/A SC CS N/A FC CS N/A PI Y2 N/A N/A CS-04 CS-04 N/A N/A CS-04 CS-04 N/A N/A CS-04 CS-04 N/A Revision 10 10114/2002 Page A16 - 20 of 53 j

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

215 VALVE ID DESCRIPTION V-16-001 CLEANUP SUPPLY INNER CONTAINMENT ISOLATION V-16-002 P-16-2 SUCTION ISOLATION V-16-014 CLEANUP SUPPLY OUTER CONTAINMENT ISOLATION V-16-061 CLEANUP RETURN OUTER CONTAINMENT ISOLATION V-16-062 CLEANUP RETURN INNER CONTAINMENT ISOLATION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TYPE TYPE NORMIFAILISAFE PAS 6

GA MO GE148F444 01 G-1 1

0 Al C

A A

6 GA MO GE 148F444 01 G-9 1

0 Al C

A A

6 GA MO GE148F444 01 G-9 1

0 Al C

A A

6 GA MO GE148F444 01 F-6 1

0 Al C

A A

6 CK SA GE 148F444 01 F-6 1

0 C

A/C A

TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS LTJ AJ N/A N/A SC CS N/A CS-08 PI Y2 RV-51 N/A LTJ AJ N/A N/A SC CS N/A CS-08 PI Y2 RV-51 N/A LTJ AJ N/A N/A SC CS N/A CS-08 PI Y2 RV-51 N/A LTJ AJ N/A N/A SC CS N/A CS-08 PI Y2 RV-51 N/A LTJ AJ N/A N/A CC RR N/A RJ-07 Revision 10 10/14/2002 Page A16 - 21 of 53 CLEANUP

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

214 SHUTDOWN COOLING SIZE VLV ACTU P&ID SH CRD CL TYPE TYPE POSITIONS CAT ACT/

NORMIFAIUSAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-17-001 P-17-1 SUCTION ISOLATION V-17-002 P-17-2 SUCTION ISOLATION V-1 7-003 P-17-3 SUCTION ISOLATION V-17-019 SHUTDOWN COOLING ISOLATION FROM RECIRC LOOP E V-17-054 RECIRC LOOP E RETURN ISOLATION V-1 7-055 SDC LOOP A RETURN ISOLATION V-1 7-056 SDC LOOP B RETURN ISOLATION 10 GA MO GE 148F711 01 G-3 10 GA MO GE 148F711 01 E-3 10 GA MO GE 148F711 01 C-3 14 GA MO GE 148F711 01 B-3 1

1 1

1 14 GA MO GE 148F711 01 D-8 1

8 GL MO GE 148F711 01 G-7 1 8

GL MO GE 148F711 01 E-7 1

O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A O/C Al C

B A

SC CS N/A CS-05 PI Y2 N/A N/A Revision 10 10/14/2002 Page A16 - 22 of 53 VALVE ID DESCRIPTION

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

214 SHUTDOWN COOLING VALVE ID DESCRIPTION V-17-057 SDC LOOP C RETURN ISOLATION V-17-227 SDC THERMAL RELIEF SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAILJSAFE PAS TEST FRO RR JUST COMMENTS 8

GL MO GE148F711 01 D-7 1

O/C Al C

.5xl RV SA GE148F711 01 D-2 1

C O/C B

A SC CS N/A CS-05 PI Y2 N/A N/A C

A RT Y2 N/A N/A Revision 10 10/14/2002 Page A16 - 23 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

251 FUEL POOL COOLING SIZE VLV ACTU -

P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAILISAFE PAS TEST FRO RR JUST COMMENTS V-18-007 P-18-1A DISCHARGE CHECK V-18-008 P-1 8-1 B DISCHARGE CHECK 4

CK SA GE 237E756 01 F-5 3

O/C O/C C

A 4

CK SA GE237E756 01 D-5 3

O/C O/C C

A CC M3 N/A CO M3 N/A CC M3 N/A CO M3 N/A V-18-076 P-18-1D DISCHARGE CHECK 8

CK SA GE 237E756 01 B-5 3

C O/C C

A CC M3 N/A CO M3 N/A V-18-077 P-18-IC DISCHARGE CHECK 8

CK SA GE 237E756 01 C-5 3

C O/C C

A CC M3 N/A CO M3 N/A Revision 10 10114/2002 Check valve to be tested only when pump is required to be used.

Check valve to be tested only when pump is required to be used.

Page A16 - 24 of 53 VALVE ID DESCRIPTION N/A N/A N/A NIA I

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

213 STANDBY LIQUID CONTROL DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS NORMIFAIL/SAFE CAT ACT/

PAS TESTING REQUIREMENTS TEST FRO RR JUST COMMENTS V-19-016 LIQUID POISON INJECTION HEADER CHECK V-19-019 LIQUID POISON INJECTION HEADER ISOLATION V-19-020 LIQUID POISON INJECTION HEADER CHECK V-19-037 P-19-001A DISCHARGE CHECK V-19-038 P-19-001B DISCHARGE CHECK V-19-042 P-19-001A DISCHARGE RELIEF V-19-043 P-19-001B DISCHARGE RELIEF V-19-044 LIQUID POISON HEADER A SQUIBB VALVE 1.5 CK SA GE 148F723 01 E-7 1

1.5 GA M

GE 148F723 01 D-8 1

1.5 CK SA GE 148F723 01 E-8 1

1.5 CK SA GE 148F723 01 E-6 2

1.5 CK SA GE 148F723 01 D-6 2

1x2 RV SA GE148F723 01 F-5 2

1 x2 RV SA GE 148F723 01 D-5 2

1.5 SHR EXP GE 148F723 01 E-7 2

C O/C A/C A

LTJ AJ CO RR N/A N/A N/A RJ-08 LO Al 0

B P

PI Y2 N/A N/A C

O/C A/C A

LTJ AJ CO RR N/A N/A N/A RJ-08 C

O/C C

A CCF M3 N/A N/A CO M3 N/A N/A C

O/C C

A CCF M3 N/A N/A CO M3 N/A N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A C

0 D

A DT S2 N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 25 of 53 VALVE ID

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

213 VALVE ID STANDBY LIQUID CONTROL DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORM/FAILJSAFE PAS TEST FRO RR JUST COMMENTS V-1 9-045 LIQUID POISON HEADER B SQUIBB VALVE 1.5 SHR EXP GE 148F723 01 D-7 2

C 0

D A

DT S2 N/A N/A Revision 10 10/14/2002 Page A16 - 26 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

212 CORE SPRAY DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

NORM/FAIL/SAFE PAS TESTING REQUIREMENTS TEST FRO RR JUST COMMENTS V-20-003 CS PUMP NZO1A SUCTION FROM TORUS V-20-004 CS PUMP NZ01B SUCTION FROM TORUS V-20-008 CS PUMP NZ01A DISCHARGE CHECK V-20-009 CS PUMP NZ01B DISCHARGE CHECK V-20-012 CS LOOP A INJECTION ISOLATION V-20-015 CS LOOP A INJECTION VALVE 12 GA MO GE885D781 01 B4 12 GA MO GE 885D781 01 D2 8

CK SA GE885D781 01 C4 2

2 2

8 CK SA GE885D781 01 F2 2

8 GA MO GE 885D781 01 E7 2

8 GA MO GE 885D781 01 E7 1

O Al 0

B P

PI Y2 N/A N/A O

Al 0

B P

PI Y2 N/A N/A C

O/C C

A CP M3 CC M3 CO SA C

O/C C

A O/C Al 0

C Al O/C CP M3 CC M3 CO SA B

A SO M3 PI Y2 A

A LTH SC SO PI Y2 M3 M3 Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 27 of 53 VALVE ID

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

212 CORE SPRAY DESCRIPTION SIZE VLV ' ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORMIFAILISAFE PAS TEST FRO RR JUST COMMENTS V-20-016 CS PUMP NZ01C DISCHARGE CHECK V-20-017 CS LOOP A INJECTION ISOLATION V-20-018 CS LOOP B INJECTION ISOLATION V-20-021 CS LOOP B INJECTION VALVE V-20-022 CS PUMP NZ01D DISCHARGE CHECK V-20-023 CS LOOP B INJECTION ISOLATION V-20-026 CS LOOP B TEST RETURN ISOLATION 8

CK SA GE 885D781 01 D4 2

8 GA M

GE 885D781 01 F7 1

8 GA MO GE 885D781 01 G7 2

8 GA MO GE 885D781 01 G7 1

8 CK SA GE885D781 01 F3 2

8 GA M

GE 885D781 01 F7 1

6 GL MO GE885D781 01 F6 2

C O/C LO Al 0

C A

CP M3 N/A CC M3 N/A CO SA N/A B

P PI Y2 N/A O

Al 0

B A

SO M3 N/A PI Y2 N/A C

Al O/C A

A LTH Y2 N/A SC M3 N/A SO M3 N/A PI Y2 N/A C

O/C C

A LO Al 0

CP M3 N/A CC M3 N/A CO SA N/A B

P PI Y2 N/A C

Al C

B A

SC M3 N/A PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 28 of 53 VALVE ID

of ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

212 CORE SPRAY DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

NORM/FAIL/SAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-20-027 CS LOOP A TEST RETURN ISOLATION V-20-030 CS LOOP A TEST RETURN CHECK V-20-031 CS LOOP B TEST RETURN CHECK V-20-032 CS PUMP NZ01C SUCTION FROM TORUS V-20-033 CS PUMP NZ01D SUCTION FROM TORUS V-20-040 CS LOOP A INJECTION VALVE V-20-041 CS LOOP B INJECTION VALVE 6

GL MO GE 885D781 01 D7 6

CK SA GE 885D781 01 C7 6

CK SA GE 885D781 01 C6 12 GA MO GE 885D781 01 B4 12 GA MO GE 885D781 01 D3 8

GA MO GE 885D781 01 D7 8

GA MO GE 885D781 01 G7 1

C Al C

B A

SC M3 PI Y2 N/A N/A N/A N/A C

0 C

A CO M3 N/A N/A C

0 C

A CO M3 N/A N/A O

Al 0

B P

PI Y2 N/A N/A O

Al 0

B P

PI Y2 N/A N/A C

Al O/C A

A LTH Y2 SC M3 SO M3 PI Y2 C

Al O/C A

A LTH Y2 SC M3 SO M3 PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 29 of 53 VALVE ID

dl ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

212 CORE SPRAY DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

NORM/FAILISAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-20-050 CS BOOSTER PUMPS NZ03NC BYPASS CHECK V-20-051 CS BOOSTER PUMPS NZ03B/D BYPASS CHECK 10 CK SA GE885D781 01 D5 2

10 CK SA GE 885D781 01 G5 2

C O/C C

A C

O/C C

A V-20-052 CS BOOSTER PUMP NZ03A DISCHARGE CHECK 10 CK SA GE 885D781 01 C5 2

C O/C C

A CP M3 N/A N/A CC M3 N/A N/A CO SA N/A N/A V-20-053 CS BOOSTER PUMP NZ03C DISCHARGE CHECK V-20-054 CS BOOSTER PUMP NZ03B DISCHARGE CHECK V-20-055 CS BOOSTER PUMP NZ03D DISCHARGE CHECK 10 CK SA GE 885D781 01 D5 2

10 CK SA GE885D781 01 G5 2

10 CK SA GE 885D781 01 F5 2

C O/C C

A C

O/C C

A C

O/C C

A Revision 10 10/14/2002 Page A16 - 30 of 53 VALVE ID CP M3 CC M3 CO SA CP M3 CC M3 CO SA N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A CP M3 CC M3 CO SA CP M3 CC M3 CO SA CP M3 CC M3 CO SA N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

212 CORE SPRAY SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

NORM/FAIL/SAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-20-060 FIRE PROTECTION SUPPLY TO CS LOOP A CHECK V-20-061 FIRE PROTECTION SUPPLY TO CS LOOP B CHECK V-20-082 FIRE PROTECTION SUPPLY TO CS LOOP B ISOLATION V-20-083 FIRE PROTECTION SUPPLY TO CS LOOP A ISOLATION V-20-088 FIRE PROTECTION SUPPLY TO CS LOOP A CHECK V-20-089 FIRE PROTECTION SUPPLY TO CS LOOP B CHECK V-20-116 CS FILL PUMP NZ04B TO NZ01B DISCHARGE CHECK V-20-119 CS FILL PUMP NZ04A TO NZ01A DISCHARGE CHECK 6

CK SA GE 885D781 01 E4 6

CK SA GE 885D781 01 F5 6

GA M

GE885D781 01 F5 6

GA M

GE8850781 01 E4 6

CK SA GE885D781 01 E4 6

CK SA GE885D781 01 F5 1

CK SA GE885D781 01 G3 I

CK SA GE885D781 01 D4 C

O/C C

A CC M3 CO SA C

O/C C

A CC M3 CO SA C

Al O/C B

A SC M3 SO M3 C

Al O/C B

A SC SO M3 M3 C

O/C C

A CC M3 CO SA C

O/C C

A CC M3 CO SA N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A M/A NIA M/A N/A N/A N/A N/A O

C C

A CC M3 N/A N/A O

C C

A CC M3 N/A NIA Revision 10 1011412002 Page A16-31 of 53 VALVE ID DESCRIPTION

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

212 VALVE ID CORE SPRAY DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAIUSAFE PAS TEST FRQ RR JUST COMMENTS V-20-150 CS LOOP A INJECTION TESTABLE CHECK VALVE V-20-151 CS LOOP B INJECTION TESTABLE CHECK VALVE V-20-152 CS LOOP A INJECTION TESTABLE CHECK VALVE V-20-153 CS LOOP B INJECTION TESTABLE CHECK VALVE 8

CK AO GE 885D781 01 E7 1

8 CK AO GE 885D781 01 G7 1

8 CK AO GE 885D781 01 E8 1

8 CK AO GE 885D781 01 G8 1

C O/C AC A

C O/C AC A

C O/C AC A

C O/C AC A

LTH CS N/A N/A Refer to T.S. 4.3G CC CS N/A CS-06 CO CS N/A CS-06 PI Y2 N/A N/A LTH CS CC CS CO CS PI Y2 LTH CS CC CS CO CS PI Y2 LTH CS CC CS CO CS PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Refer to T.S. 4.3G CS-06 CS-06 N/A N/A Refer to T.S. 4 3G CS-06 CS-06 N/A N/A Refer to T.S. 4.3G CS-06 CS-06 N/A Revision 10 10/14/2002 Page A16 - 32 of 53

'r ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

241 CONTAINMENT SPRAY DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

NORMIFAILUSAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-21-001 P-21-1C SUCTION ISOLATION V-21-002 P-21-1C DISCHARGE CHECK V-21-003 P-21-1D SUCTION ISOLATION V-21-004 P-21-1D DISCHARGE CHECK V-21-005 B LOOP DRYWELL SPRAY INJECTION ISOLATION V-21-007 P-21-1B SUCTION ISOLATION V-21-008 P-21-1B DISCHARGE CHECK 12 GA MO GE 148F740 01 D-5 2 10 CK SA GE 148F740 01 D-4 2 12 GA MO GE 148F740 01 D-5 2 10 CK SA GE 148F740 01 B-4 2 14 GA MO GE 148F740 01 G-4 2 12 GA MO GE 148F740 01 D-5 2 10 CK SA GE 148F740 01 D-6 2 0

Al 0

B P

P1 Y2 N/A N/A C

O/C C

A CC CO M3 M3 N/A N/A N/A N/A O

Al 0

B P

PI Y2 N/A N/A C

O/C C

A CC M3 N/A N/A CO M3 N/A N/A C

Al O/C B

A SC M3 N/A N/A SO M3 N/A N/A P1 Y2 N/A N/A 0

Al 0

B P

P1 Y2 N/A N/A C

O/C C

A CC CO M3 M3 N/A N/A N/A N/A Revision 10 10114/2002 Page A16 - 33 of 53 VALVE ID

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

241 CONTAINMENT SPRAY DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORMIFAIL/SAFE PAS TEST FRO RR JUST COMMENTS V-21-009 P-21-1A SUCTION ISOLATION V-21-010 P-21-1A DISCHARGE CHECK V-21-011 A LOOP DRYWELL SPRAY INJECTION ISOLATION V-21-013 B LOOP SUPPRESSION POOL COOLING & TEST ISOL V-21-015 B LOOP TORUS SPRAY INJECTION ISOLATION V-21-017 A LOOP SUPPRESSION POOL COOLING & TEST ISOL 12 GA MO GE 148F740 01 D-5 2

10 CK SA GE 148F740 01 B-6 2

14 GA MO GE 148F740 01 G-6 2

6 GA MO GE 148F740 01 G-4 2 4

GA MO GE 148F740 01 G-4 2 6

GA MO GE 148F740 01 G-6 2 O

Al 0

B P

PI Y2 N/A N/A C

O/C C

A CC M3 N/A N/A CO M3 N/A N/A C

Al O/C B

A SC M3 N/A N/A SO M3 N/A N/A PI Y2 N/A N/A O

Al O/C B

A C

Al O/C A

A O

Al O/C B

A SC M3 SO M3 PI Y2 LTL Y2 SC M3 SO M3 PI Y2 SC M3 SO M3 PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 -34 of 53 VALVE ID

1)

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

241 CONTAINMENT SPRAY DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT NORMIFAIL/SAFE ACT/

TESTING REQUIREMENTS PAS TEST FRQ RR JUST COMMENTS V-21-018 A LOOP TORUS SPRAY INJECTION ISOLATION V-21-021 HEAT EXCHANGE H-21-1A SHELL SIDE RELIEF V-21-022 HEAT EXCHANGE H-21-1B SHELL SIDE RELIEF V-21-023 HEAT EXCHANGE H-21-1C SHELL SIDE RELIEF V-21-024 HEAT EXCHANGE H-21-1D SHELL SIDE RELIEF 4

GA MO GE 148F740 01 F-7 2

.75 x 1 RV SA GE 148F740 01 D-7 2

.75x1 RV SA GE148F740 01 C-7 2

.75x1 RV SA GE148F740 01 D-3 2

.75xl RV SA GE148F740 01 C-3 2 C

O/C A

A LTL Y2 N/A SC M3 N/A SO M3 N/A PI Y2 N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A C

O/C C

A RT Y10 N/A Revision 10 10114/2002 Page A16 - 35 of 53 VALVE ID N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

573 DRYWELL FLOOR AND EQUIP DRAINS DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORM/FAII/SAFE PAS TEST FRO RR JUST COMMENTS V-22-001 DRYWELL EQUIP DRAIN TK PUMPS DISCH CONTAINMENT ISO V-22-002 DRYWELL EQUIP DRAIN TK PUMPS DISCH CONTAINMENT ISO V-22-028 DRYWELL FLR DRN SUMP PUMPS DISCH CONTAINMENT ISOL V-22-029 DRYWELL FLR DRN SUMP PUMPS DISCH CONTAINMENT ISOL 2

GA AO JC 147434 02 D-7 2

O/C C

C 2

GA AO JC 147434 02 D-7 2

O/C C

C 2

GA AO JC 147434 03 D-9 2

O/C C

C 2

GA AO JC 147434 03 D-9 2

O/C C

C A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 Revision 10 10/14/2002 Page A16 - 36 of 53 VALVE ID N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

242 VALVE ID CONTAINMENT INERTING DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORM/FAIL/SAFE PAS TEST FRQ RR JUST COMMENTS V-23-013 NITROGEN PURGE/HARDENED 8

BTF AO SN 13432.19- 01 VENT ISOLATION I

V-23-014 NITROGEN PURGE/HARDENED 8

BTF AO SN 13432.19- 01 VENT ISOLATION 1

V-23-015 NITROGEN PURGE/HARDENED 8

BTF AO SN 13432.19- 01 VENT ISOLATION 1

V-23-016 NITROGEN PURGE/HARDENED 8

BTF AO SN 13432.19- 01 VENT ISOLATION 1

V-23-017 NITROGEN MAKEUP ISOLATION 2

GL AO SN 13432.19- 01 1

NC O/C C

C NC O/C C

C NC O/C C

C NC O/C C

C NC O/C C

C A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A Revision 10 10/14/2002 Page A16 - 37 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

242 CONTAINMENT INERTING SIZE VLV ACTU P&ID SH CRD CL TYPE TYPE POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS V-23-018 NITROGEN MAKEUP ISOLATION 2

GL AO SN 13432.19- 01 1

V-23-019 NITROGEN MAKEUP ISOLATION 2

GL AO SN 13432.19- 01 1

V-23-020 NITROGEN MAKEUP ISOLATION 2

GL AO SN 13432.19- 01 1

V-23-021 DRYWELL VENT 2

GA AO BR 2011 02 V-23-022 DRYWELL VENT 2

GA AO BR 2011 02 NC O/C C

C NC O/C C

C NC O/C C

C NC O/C C

C NC O/C C

C A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A P1 Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A P1 Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A Revision 10 10/14/2002 Page A16 -38 of 53 VALVE ID DESCRIPTION

4 ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

242 CONTAINMENT INERTING DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REOUIREMENTS TYPE TYPE NORM/FAILISAFE PAS TEST FRQ RR JUST COMMENTS V-23-070 TIP PURGE ISOLATION V-23-357 N2 LINE ISOLATION V-23-358 HARDENED VENT LINE ISOLATION

.25 PLG SO SN 13432.19- 01 1

8 BTF M

SN 13432.19- 01 1

8 BTF M

SN 13432.19- 01 1

NC 0

C C

NC 0

Al C

NC C

Al 0

A A

LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A B

A SC M3 N/A N/A B

A SO M3 N/A N/A Revision 10 10/14/2002 Page A16 -39 of 53 VALVE ID

'1 ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

551 REACTOR SAMPLE VALVE ID DESCRIPTION V-24-029 REACTOR SAMPLE CONTAINMENT ISOLATION V-24-030 REACTOR SAMPLE CONTAINMENT ISOLATION SIZE VLV ACTU P&ID TYPE TYPE

.75 GL AO BR M0012

.75 GL AO BR M0012 SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAILJSAFE PAS TEST FRO RR JUST COMMENTS 1

0 C

C A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A 1

0 C

C A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A Revision 10 10/14/2002 Page A16 - 40 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

243 DRYWELL AND SUPPRESSION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS NORMIFAIL/SAFE CAT ACT/

TESTING REQUIREMENTS PAS TEST FRQ RR JUST COMMENTS V-26-001 TORUS TO DRYWELL VACUUM BREAKER V-26-002 TORUS TO DRYWELL VACUUM BREAKER V-26-003 TORUS TO DRYWELL VACUUM BREAKER V-26-004 TORUS TO DRYWELL VACUUM BREAKER V-26-005 TORUS TO DRYWELL VACUUM BREAKER 18 CK SA GU 3E-243 21-1000 01 E-6 2

18 CK SA GU 3E-243-01 E-6 2

21-1000 18 CK SA GU 3E-243-01 E-6 2

21-1000 18 CK SA GU 3E-243-01 E-6 2

21-1000 18 CK SA GU 3E-243-01 E-6 2

21-1000 IC O/C A/C A

LTL Y2 CCX M3 COX M3 PI Y2 C

O/C A/C A

C O/C A/C A

C O/C NC A

C O/C A/C A

LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A LTL Y2 N/A CCX M3 N/A COX M3 N/A PI Y2 N/A LTL Y2 CCX M3 COX M3 PI Y2 N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 41 of 53 VALVE ID DESCRIPTION N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

4 ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

243 DRYWELL AND SUPPRESSION DESCRIPTION SIZE VLV ACTU P&ID TYPE TYPE SH CRD CL POSITIONS CAT ACT/

NORMIFAIL/SAFE PAS TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS V-26-006 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243- 01 F-6 2

BREAKER 21-1000 V-26-007 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243- 01 F-6 2

BREAKER 21-1000 V-26-008 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243- 01 F-6 2

BREAKER 21-1000 V-26-009 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243-01 F-4 2

BREAKER 21-1000 V-26-010 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243-01 E-4 2

BREAKER 21-1000 C

O/C A/C A

C O/C A/C A

C O/C A/C A

C O/C A/C A

C O/C A/C A

LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A NIA NIA N/A Revision 10 10/14/2002 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Page A16 - 42 of 53 VALVE ID

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

243 DRYWELL AND SUPPRESSION DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORM/FAILJSAFE PAS TEST FRO RR JUST COMMENTS V-26-011 TORUSTODRYWELLVACUUM 18 CK SA GU3E-243-01 E-4 2 BREAKER 21-1000 V-26-012 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243- 01 E-4 2 BREAKER 21-1000 V-26-013 TORUS TO DRYWELL VACUUM 18 CK SA GU 3E-243- 01 E-4 2 BREAKER 21-1000 V-26-014 TORUS TO DRYWELL VACUUM BREAKER V-26-015 REACTOR BLDG TO TORUS VACUUM BREAKER 18 CK SA GU3E-243- 01 E-4 2

21-1000 20 CK SA GU 3E-243- 01 G-3 2 21-1000 C

O/C NC A

C O/C A/C A

C O/C A/C A

C O/C A/C A

C O/C NC A

Revision 10 10/14/2002 Page A16 - 43 of 53 VALVE ID LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 P1 Y2 LTL Y2 CCX M3 COX M3 PI Y2 LTL Y2 CCX M3 COX M3 PI Y2 LTJ AJ CCX M3 COX M3 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

243 DRYWELL AND SUPPRESSION VALVE ID DESCRIPTION V-26-016 REACTOR BLDG TO TORUS VACUUM BREAKER V-26-017 REACTOR BLDG TO TORUS VACUUM BREAKER V-26-018 REACTOR BLDG TO TORUS VACUUM BREAKER SIZE VLV ACTU P&ID TYPE TYPE 20 BTF AO GU 3E-243-01 G-3 2 21-1000 20 CK SA GU 3E-243-01 G-3 2 21-1000 20 BTF AO GU 3E-243- 01 G-3 2 21-1000 SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAILJSAFE PAS TEST FRO RR JUST COMMENTS C

C O/C A

A LTJ AJ N/A N/A SC M3 N/A N/A SO M3 N/A N/A FC M3 N/A N/A PI Y2 N/A N/A C

O/C A/C A

LTJ AJ N/A N/A CCX M3 N/A N/A COX M3 N/A N/A C

C O/C A

A LTJ AJ N/A N/A SC M3 N/A N/A SO M3 N/A N/A FC M3 N/A N/A PI Y2 N/A N/A Revision 10 10/14/2002 Page A16 -44 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

822 VALVE ID REACTOR BUILDING VENTILATION DESCRIPTION SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS TYPE TYPE NORM/FAIL/SAFE PAS TEST FRQ RR JUST COMMENTS V-27-001 RB VENTILATION EXHAUST CONTAINMENT ISOLATION V-27-002 RB VENTILATION EXHAUST CONTAINMENT ISOLATION V-27-003 RB VENTILATION SUPPLY CONTAINMENT ISOLATION V-27-004 RB VENTILATION SUPPLY CONTAINMENT ISOLATION 18 BTF AO 18 BTF AO 18 BTF AO 18 BTF AO BR2011 02 D-6 NC O/C C

C BR2011 02 D-6 NC O/C C

C BR2011 02 E-8 NC O/C C

C BR2011 02 E-8 NC O/C C

C A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A A

A LTJ AJ N/A N/A SC M3 N/A N/A FC M3 N/A N/A PI Y2 RV-51 N/A Revision 10 10/14/2002 Page A16 - 45 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

822 REACTOR BUILDING VENTILATION VALVE ID DESCRIPTION V-28-017 TORUS TO VENT EXHAUST CONTAINMENT ISOLATION V-28-018 TORUS TO VENT EXHAUST CONTAINMENT ISOLATION V-28-047 TORUS TO VENT EXHAUST CONTAINMENT ISOLATION SIZE VLV ACTU P&ID TYPE TYPE 12 BTF AO BR 2011 12 BTF AO BR 2011 2

GA AO BR2011 SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS NC O/C C

C NC 0/C C

C NC O/C C

C A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 A

A LTJ AJ N/A SC M3 N/A FC M3 N/A PI Y2 RV-51 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 46 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

216 VALVE ID HEAD COOLING DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORMIFAIL/SAFE PAS TEST FRQ RR JUST COMMENTS V-31-002 REACTOR HEAD COOLING CONTAINMENT ISOLATION V-31-005 REACTOR HEAD COOLING CONTAINMENT ISOLATION CHECK 2

GL AO GE 237E487 01 F-8 1

2 CK SA GE237E487 01 H-7 1

C C

C A

P LTJ AJ N/A PI Y2 RV-51 C

C AC P

LTJ AJ N/A N/A Revision 10 10/14/2002 Page A16 -47 of 53 N/A N/A

IN ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

666 HYDROGEN/OXYGEN MONITORING VALVE ID DESCRIPTION V-38-009 DRYWELL 02 SAMPLE CONTAINMENT ISOLATION V-38-010 DRYWELL 02 SAMPLE CONTAINMENT ISOLATION V-38-016 CONT PART MON RETURN CONTAINMENT ISOLATION V-38-017 CONT PART MON RETURN CONTAINMENT ISOLATION V-38-022 TORUS 02 SAMPLE CONTAINMENT ISOLATION V-38-023 TORUS 02 SAMPLE CONTAINMENT ISOLATION SIZE VLV ACTU P&ID TYPE TYPE

.75 GL SO BR M0012

.75 GL SO BR Mo012

.75 GL SO GU 3E-666 21-1000

.75 GL SO GU 3E-666 21-1000

.25 GL SO GU 3E-666 21-1000

.25 GL SO GU 3E-666 21-1000 SH CRD CL POSITIONS CAT ACTI TESTING REQUIREMENTS NORM/FAILJSAFE PAS TEST FRO RR JUST COMMENTS NC 0

C C

NC 0

C C

NC C

C C

NC C

C C

NC C

C C

NC 0

C C

A A

LTJ AJ N/A SC RR N/A FC RR N/A A

A LTJ AJ N/A SC RR N/A FC RR N/A A

A LTJ AJ N/A SC RR N/A FC RR N/A A

A LTJ AJ N/A SC RR N/A FC RR N/A A

A LTJ AJ N/A SC RR N/A FC RR N/A A

A LTJ AJ N/A SC RR N/A FC RR N/A Revision 10 10/14/2002 Page A16 - 48 of 53 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

666 HYDROGEN/OXYGEN MONITORING VALVE ID DESCRIPTION V-38-037 H2/02 SAMPLE SUPPLY CONTAINMENT ISOLATION V-38-038 H2/02 SAMPLE SUPPLY CONTAINMENT ISOLATION V-38-039 H2/02 SAMPLE RETURN CONTAINMENT ISOLATION V-38-040 H2/02 SAMPLE RETURN CONTAINMENT ISOLATION SIZE VLV ACTU P&ID TYPE TYPE 1

GL SO GU 3E-666 21-1000 1

GL SO GU 3E-666 21-1000 1

GL SO GU 3E-666 21-1000 1

GL SO GU3E-666 21-1000 SH CRD CL POSITIONS CAT ACT/

NORMIFAILJSAFE PAS NC O/C C

O/C A

A NC O/C C

O/C A

A NC O/C C O/C A

A NC O/C C O/C A

A TESTING REQUIREMENTS TEST FRQ RR JUST COMMENTS LTJ AJ SC M3 SO M3 FC M3 PI Y2 LTJ AJ SC M3 SO M3 FC M3 PI Y2 LTJ AJ SC M3 SO M3 FC M3 P1 Y2 LTJ SC SO FC PI AJ M3 M3 M3 Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Revision 10 10/14/2002 Page A16 - 49 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

666 HYDROGEN/OXYGEN MONITORING VALVE ID DESCRIPTION V-38-041 H2/02 SAMPLE SUPPLY CONTAINMENT ISOLATION V-38-043 H2/02 SAMPLE SUPPLY CONTAINMENT ISOLATION V-38-044 H2/02 SAMPLE RETURN CONTAINMENT ISOLATION V-38-046 H2/02 SAMPLE RETURN CONTAINMENT ISOLATION SIZE VLV ACTU P&ID TYPE TYPE 1

GL SO GU 3E-666 21-1000 1

GL SO GU 3E-666 21-1000 1

GL SO GU 3E-666 21-1000 1

GL SO GU 3E-666 21-1000 SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIUISAFE PAS TEST FRO RR JUST COMMENTS NC O/C C

O/C A

A NC O/C C

O/C A

A NC O/C C

O/C A

A NC O/C C

O/C A

A LTJ AJ SC M3 SO M3 FC M3 PI Y2 LTJ AJ SC M3 SO M3 FC M3 PI Y2 LTJ AJ SC M3 SO M3 FC M3 PI Y2 LTJ SC SO FC PI AJ M3 M3 M3 Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A NIA N/A N/A N/A N/A N/A N/A Revision 10 10114/2002 Page A16 - 50 of 53

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

555 POST ACCIDENT SAMPLING DESCRIPTION SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAILJSAFE PAS TEST FRO RR JUST COMMENTS V-40-006 PASS CONTAINMENT ISOLATION V-40-008 PASS CONTAINMENT ISOLATION

.75 GL SO BR M0012

.75 GL SO BR M0012 1

LC C

C 1

LC C

C A

P LTJ AJ N/A PI Y2 N/A A

P LTJ AJ N/A PI Y2 N/A V-40-137 CONTAINMENT PENETRATION

.375 CK SA BR M0012 THERMAL RELIEF CHECK C

O/C AC A

LTJ AJ CCF Y4 COF Y4 Revision 10 10/14/2002 N/A N/A N/A N/A N/A N/A Page A16 - 51 of 53 VALVE ID N/A N/A N/A N/A

'4 ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

623 TRAVELING INCORE PROBE SIZE VLV ACTU P&ID SH CRD CL POSITIONS CAT ACT/

TYPE TYPE NORM/FAIL/SAFE PAS V-623-001 TIP CONTAINMENT ISOLATION

.375 BAL SO GE 147D8699 BALL VALVE V-623-002 TIP CONTAINMENT ISOLATION 375 BAL SO GE 147D8699 BALL VALVE V-623-003 TIP CONTAINMENT ISOLATION

.375 BAL SO GE 147D8699 BALL VALVE V-623-004 TIP CONTAINMENT ISOLATION

.375 BAL SO GE 147D8699 BALL VALVE NC O/C C

C NC O/C C

C NC O/C C

C NC O/C C

C TESTING REQUIREMENTS TEST FRO RR JUST COMMENTS A

A LTJ AJ SC M3 FC M3 PI Y2 A

A LTJ AJ SC M3 FC M3 PI Y2 A

A LTJ AJ SC M3 FC M3 PI Y2 A

A LTJ AJ SC M3 FC M3 PI Y2 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A V-623-005 TIP SHEAR VALVE

.375 SHR EXP GE 147D8699 NC 0

C D

A DT S2 N/A N/A V-623-006 TIP SHEAR VALVE

.375 SHR EXP GE 147D8699 NC 0

C D

A DT S2 N/A N/A Revision 10 10/14/2002 Page A16 - 52 of 53 VALVE ID DESCRIPTION

ATTACHMENT 16 OYSTER CREEK NUCLEAR POWER STATION INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM:

623 TRAVELING INCORE PROBE SIZE VLV ACTU TYPE TYPE P&ID SH CRD CL POSITIONS CAT ACT/

TESTING REQUIREMENTS NORM/FAIL/SAFE PAS TEST FRO RR JUST COMMENTS V-623-007 TIP SHEAR VALVE V-623-008 TIP SHEAR VALVE

.375 SHR EXP GE 147D8699

.375 SHR EXP GE 147D8699 NC 0

NC 0

C D

A DT S2 N/A N/A C

D A

DT S2 N/A N/A Revision 10 10/14/2002 Page A16 - 53 of 53 VALVE ID DESCRIPTION

'k, Kl