Text

Fifth Ten-Year Inservice Testing Program and Request for Approval of IST Relief Requests, PR-03 and PR-05
	ML12342A386
Person / Time
Site:	Pilgrim
Issue date:	12/06/2012
From:	Jeffery Lynch; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	2.12.088
	Download: ML12342A386 (177)
	v • d • e

%"%E~w~n o2W600 Entergy Nuclear Operations, Inc.

Rocky Hill Road Plymouth, MA 02360 Pilgrim Nuclear Power Station December 6, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Pilgrim Nuclear Power Station Fifth Ten-Year Inservice Testing (IST)

Program and Request for Approval of IST Relief Requests. PR-03 and PR-05 LETTER NUMBER: 2.12.088

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. (Entergy) has revised the Pilgrim Nuclear Power Station (PNPS) Inservice Testing (IST) Program as required by 10CFR50.55a(f)(4)(ii) for the Fifth 10-year interval starting December 7, 2012.

This submittal dockets Pilgrim IST Program, Procedure SEP-PNPS-IST-001 and requests NRC approval of IST Relief Requests, PR-03 and PR-05, as described in this letter.

The revised IST Program complies with the 2004 Edition through 2006 Addenda of the OM Code for Operation and Maintenance of Nuclear Power Plants, Section IST requirements with the exception of two pump Relief Requests, PR-03 and PR-05. These exceptions invoke 10 CFR 50.55a(f)(5)(iii) for notification that conformance with certain code requirements are impractical for the Pilgrim facility.

The details of the IST Relief Requests, PR-03 and PR-05 are specified in the IST Program Section 7.1, Pump Testing Program Relief Requests.

The IST Relief Request PR-03 is necessary for biennial comprehensive pump testing of the High Pressure Coolant Injection (HPCI) Pump, which is due in November 2015. IST Relief Request PR-05 is necessary for biennial comprehensive pump testing of the Standby Liquid Control (SLC) System pump, which is due in July 2014.

The IST PR-03 and PR-05 were previously approved by the NRC for the Fourth IST interval and follow prior NRC approved approaches, as discussed in the description of the relief requests.

Entergy Nuclear Operations, Inc. Letter Number: 2.12.088 Pilgrim Nuclear Power Station Page 2 Accordingly, Entergy requests NRC review and approval of the IST PR-03 and PR-05 by December 31, 2013.

This submittal contains no new commitments.

If you have any questions or require additional information, please contact Mr. Joseph R. Lynch, Licensing Manager, at (508) 830-8403.

Sincerely, Joseph R. LZvnn-Pilgrim Licensing Manager

Attachment:

PNPS Inservice Pump and Valve Testing Program, SEP-PNPS-IST-001, Rev. 2, dated 12/06/12, (174 pages) cc:

Mr. Richard Guzman, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North O-8C2 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 NRC Resident Inspector Pilgrim Nuclear Power Station

Attachment to Enterqy Letter No 2.12.088 PNPS Inservice Pump and Valve Testing Program, SEP-PNPS-IST-001, Rev. 2, dated 12/06/12, (174 pages)

Program Section No: SEP-PNPS-IST-001 Rev.: 2 Page 1 of 174 PNPS INSERVICE PUMP AND VALVE TESTING PROGRAM SEP-PNPS-IST-001 ENTERGY NUCLEAR ENGINEERING PROGRAMS APPLICABLE SITES All Sites: -1 Specific Sites: ANO F-1 GGNS[] IPECE] JAF L PLP I] PNPS E RBS [] VY1 W3ME] HQN[]

Continuous Use [: Reference Use Z Informational Use El-Safety Related: Yes F- No

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 2 of 174 REVIEW AND CONCURRENCE SHEET Program Section PNPS Inservice Pump and Valve Testing Program Title:

Prepared by: Date: 12/01/12

~~-e .- Joseph M. Bohn iýt~sp sible Engine ignature Checked by: Date: \-. ,\ z.-

N a " saSvi Are, Concurred by: Date: /z 6. &-i,

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 3 of 174 REVISION STATUS SHEET PROGRAM SECTION REVISION

SUMMARY

Revision Description 2 Revision due to PNPS Inservice Pump and Valve Testing Program (ISTP) being upgraded in accordance with 10CFR50.55a(f)(4)(ii) which requires compliance with the latest approved code incorporated within 10CFR50.55a(b). The Code of Record and that which is adopted for this upgrade is ASME OM, Code for Operation and Maintenance of Nuclear Power Plants, 2004 Edition through 2006 Addenda.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 4 of 174 TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE 6 2.0 SCOPE 6 3.0 COMPLIANCE 7 4.0 GENERAL 8 4.1 GENERAL DISCUSSION 8 4.2 DEFINITIONS 8 4.3 POSTMAINTENANCE TESTING 9

4.4 REFERENCES

10 5.0 INSERVICE PUMP TESTING 12 5.1 GENERAL INFORMATION 12 5.2 HYDRAULIC CIRCUITS 17 5.3 PUMP PROGRAM TABLE 21 6.0 INSERVICE VALVE TESTING 23 6.1 GENERAL INFORMATION 23 6.2 VALVE PROGRAM TABLE 40 7.0 PROGRAM JUSTIFICATIONS AND RELIEF REQUESTS 94 7.1 PUMP TESTING PROGRAM RELIEF REQUESTS 94 7.2 VALVE TESTING PROGRAM COLD SHUTDOWN JUSTIFICATIONS 109 7.3 VALVE TESTING PROGRAM REFUEL OUTAGE JUSTIFICATIONS 116 7.4 VALVE DISASSEMBLY EXAMINATION JUSTIFICATIONS 153 7.5 VALVE TESTING PROGRAM RELIEF REQUESTS 154 7.6 VALVE TESTING PROGRAM TECHNICAL POSITIONS 155 7.7 SKID-MOUNT COMPONENT TECHNICAL POSITIONS 162

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 5 of 174 INSERVICE PUMP AND VALVE TESTING PROGRAM PLAN 5 th Ten Year Interval Effective December 7 th 2012 PILGRIM NUCLEAR POWER STATION Commercial Start Date: December 7 th, 1972 Owner:

Entergy Nuclear Operations, Inc.

600 Rockyhill Road Plymouth, MA 02360

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 6 of 174 1.0 PURPOSE This Procedure encompasses and controls the PNPS Inservice Testing (IST) Program. It identifies the scope of components (pumps and valves) and testing requirements for compliance with 10CFR50.55a(f), Inservice Testing Requirements. This Procedure will be utilized for the IST Program submittal to satisfy ISTA-3120 Inservice Test Interval and to identify impractical Code requirements in accordance with 10CFR50.55a(f)(5).

2.0 SCOPE The scope of the IST Program includes those safety-related pumps and valves which are part of the Reactor coolant pressure boundary and must meet the requirements applicable to components classified as ASME Code Class 1. Additionally, other safety-related pumps and valves that perform a function to shut down the Reactor or maintain the Reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for safety-related systems meet the test requirements applicable to components which are classified as ASME Code Class 2 or Class 3. This scope is limited to those pumps and valves identified as meeting ASME Code Class 1, 2, or 3 in accordance with Regulatory Guide 1.26 classifications. The pumps and valves not performing a function as stated above or those meeting the exclusion requirements of the OM Code need not be tested, but the bases for a component's exclusion must be justified.

Non-ASME Code Class safety-related pumps and valves that perform a function to shut down the Reactor or maintain the Reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for safety-related systems are to be tested under the requirements of 10CFR50 Appendix B. The scope of the PNPS Appendix B Test Program (SEP-PNPS-IST-002) includes those safety-related pumps and valves identified as non-ASME Code Class in accordance with Regulatory Guide 1.26, but otherwise meet the OM Code scoping criteria.

This Procedure details the following items: compliance requirements, general information, pump hydraulic circuits, and tables of the components (pumps and valves) tested. The last section (7.0) contains Valve Justifications (i.e., Cold Shutdown, Refuel Outage, Disassembly Examination, and Series Valve Pairs), Relief Requests, and Technical Positions (Valve Program and Skid-Mount Component). The Procedure also references the check valves that have been placed into the PNPS Check Valve Condition Monitoring Program.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 7 of 174 3.0 COMPLIANCE This Inservice Pump and Valve Testing Program will be in effect through the fifth 120-month test interval (ending December 7, 2022) and will be updated in accordance with the requirements of ISTA-3120(c)(2) and 10CFR50.55a(f)(4)(ii).

This Procedure outlines the PNPS IST Program based on the requirements of the ASME OM Code for Operation and Maintenance of Nuclear Power Plants, 2004 Edition through the 2006 Addenda.

10CFR50.55a(f)(1) provides guidance for inclusion of ASME Code 1, 2, and 3 components into the IST Program. The guideline states that "pumps and valves which are part of the Reactor coolant pressure boundary must meet the requirements applicable to components which are classified as ASME Code Class 1. Other safety related pumps and valves [that perform a function to shut down the Reactor or maintain the Reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for safety-related systems - in or when meeting the requirements of the OM Code, or later] must meet the test requirements applicable to components which are classified as ASME Code Class 2 or Class 3". ASME safety related components that apply to IST are those pumps and valves included within the scope of Regulatory Guide 1.26.

Impractical Code requirements are reviewed and dispositioned by the Nuclear Regulatory Commission (NRC) and documented in a Safety Evaluation Report authored by the Office of Nuclear Reactor Regulation as related to the Inservice Testing Program and Requests for Relief.

The NRC will grant program relief requests pursuant to 10CFR50.55a(a)(3)(i),

10CFR50.55a(a)(3)(ii), or 10CFR50.55a(f)(6)(i). Granting of relief ensures that the IST Program has satisfactorily demonstrated that either: 1) the proposed alternative provides an acceptable level of quality and safety, 2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, or 3) conformance with certain requirements of the applicable Code edition and addenda is impractical for its facility.

If the IST Program conflicts with PNPS Technical Specifications, a Technical Specifications amendment shall be submitted to conform the Technical Specifications to this Procedure in accordance with 10CFR50.55a(f)(5)(ii). Until approval of the Technical Specifications amendment, the most limiting requirement shall be met.

The IST Program has incorporated a Condition Monitoring Program which establishes as an alternative to the test or examination requirements of ISTC-3510, ISTC-3520, ISTC-3530, ISTC-3550, and ISTC-5221 for a check valve or check valve group.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 8 of 174 4.0 GENERAL 4.1 GENERAL DISCUSSION The Procedure's pump and valve tables provide a cross-reference between a component test requirement and a Station Procedure implementing the test. Additional information is provided within this component listing: safety class, category, test frequency, test parameters, Relief Requests, justifications, and remarks. All newly identified component/test requirements shall be initially tested during the next scheduled frequency (i.e., quarterly, cold shutdown, refueling interval, and 2 years) following the IST interval start date. Newly incorporated component/test requirements may be identified by an asterisk (***) next to the implementing Procedure. When using (***)

Procedures for postmaintenance testing, the current approved Procedure should be reviewed for applicability (i.e., is the new test requirement or component incorporated).

SEP-PNPS-IST-009, "PNPS Administrative Guidelines for the Inservice Code Testing (IST) and Appendix B Testing (ABT) Programs" covers the administrative requirements for the development, performance, and maintenance of the PNPS Inservice Test Program in accordance with the ASME OMb Code for Operation and Maintenance of Nuclear Power Plants, and includes the 2004 Edition through 2006 Addenda (OMb-2006).

Station ALARA practices have been considered when addressing ASME Code test requirements within this Procedure. When test requirements are added or revised, good ALARA practices should be incorporated to minimize personnel dose.

4.2 DEFINITIONS

[1] As-Found Condition The condition of a component between inservice tests without activities that could affect the ability to determine component degradation.

[2] Active Valve Valves that are required to change obturator position to accomplish the require function.

[3] Passive Valve Valves that are required to maintain obturator position and are not required to change obturator position to accomplish their required function.

[4] Excluded Pumps and valves that are excluded from inservice testing are those that provide one of the safety-related functions that define pumps and valves subject to IST, but are specifically excluded from testing requirements by the Exclusions in ISTB-1200 or ISTC-1200. From a programmatic point of view, components that are "excluded" typically have an exclusion basis in the IST Program or basis documentation, while components that are "exempt" typically do not have a basis in the IST Program.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 9 of 174

[5] Exempt Pumps and valves that are exempt from inservice testing are those that do not meet the scope statement in the OM Code (ISTA -1100). That is, they do not perform any of the safety-related functions that define pumps and valves subject to IST. Most exempt components are not included in the IST Program. Although ISTC-1200 is entitled "Exemptions", the corresponding section was entitled, "Exclusions" in previous Editions, and the discussion describes exclusions. ISTB-1200 is still entitled "Exclusions".

[6] Preconditioninq The modification, maintenance, manipulation, or adjustment of a component performed between inservice tests that have significant potential, either with or without the intent of enhancing the results of the inservice tests. This includes activities such as cycling, cleaning, lubricating, agitating or other specific maintenance or operational activities that may be performed prior to or during inservice testing, that has the potential to enhance component test performance, and impair the ability to monitor (determine) component degradation.

4.3 POSTMAINTENACE The IST Program requires each pump and valve to be verified as operationally ready following routine servicing, maintenance, repair, or replacement.

The detailed postmaintenance pump and valve testing guideline is provided within PNPS 8.1.1, "Administrative Guidance for Inservice Pump and Valve Testing".

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 10 of 174

4.4 REFERENCES

[1] 10CFR50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

[2] 10CFR50 Appendix J, Primary Reactor Containment Leakage Testing

[3] 10CFR50.55a(b), Code and Standards, Reference Applicability

[4] 10CFR50.55a(f), Inservice Testing Requirements

[5] ASME Code, Mandatory Appendix I, Inservice Testing of Pressure Relief Device in Light-Water Reactor Power Plants

[6] ASME Code, Subsection ISTA, General Requirements

[7] ASME Code, Subsection ISTB, Inservice Testing of Pumps in Light-Water Reactor Power Plants

[8] ASME Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Power Plants

[9] ASME Code for Operation and Maintenance of Nuclear Power Plants, Section IST (Rules for Inservice Testing of Light-Water Reactor Power Plants), 2006 Addenda, Subsections ISTA, ISTB, and ISTC

[10] Calculation IN1-299, IST Instrument Calculation

[11] EN-DC-167, "Classification of Structures, Systems, and Components" (replaces obsolete PNPS "Q"-List)

[12] ESR Response Memo ERM-90-578

[13] NRC Generic Letter 87-06, "Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves"

[14] NRC Generic Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Programs"

[15] NRC Information Notice IN-88-70, "Check Valve Inservice Testing Program Deficiencies",

August 29, 1988

[16] NRC Information Notice IN-97-90, "Use of Non-Conservative Acceptance Criteria in Safety Related Pump Surveillance Tests", December 30, 1997

[17] NRC Publications, "Minutes of the Public Meetings on Generic Letter 89-04", dated October 25, 1989

[18] NRC Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants"

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 11 of 174

[19] NUREG-0800, Standard Review Plan 3.2.2, "System Quality Group Classification"

[20] NUREG-0800, Standard Review Plan 3.9.6, "Inservice Testing of Pumps and Valves"

[21] NUREG-0800, Standard Review Plan 6.2.4, "Containment Isolation System"

[22] NUREG 1352, "Action Plans for Motor-Operated Valves and Check Valves"

[23] NUREG 1482, "Guidelines for Inservice Testing at Nuclear Power Plants" Rev 0, dated April 1995 and Rev 1, dated January 2005

[24] Procedures (a) EN-DC-332, "Inservice Testing" (b) PNPS 8.1.1, "Administrative Guidance for Inservice Pump and Valve Testing" (c) PNPS 8.1.26.3, "Administrative Guidance of Pressure Relief Valve Testing for Inservice Test and Appendix B Test Programs" (d) SEP-PNPS-IST-002, "PNPS Appendix B Pump and Valve Testing Program" (e) SEP-PNPS-IST-004, "PNPS Administration of Check Valve Condition Monitoring Program Plan" (f) SEP-PNPS-IST-005, "PNPS Administration of Inservice Pressure Relief Device Test Program" (g) SEP-PNPS-IST-008, "PNPS Administrative Controls for Limiting Stroke Time Criteria of Inservice and Appendix B Test Program Power Operated Valves" (h) SEP-PNPS-IST-009, "PNPS Administrative Guidelines for the Inservice Code Testing (IST) and Appendix B Testing (ABT) Programs"

[25] PR98.9525.01, Engineering Action Item Response

[26] Summary of NRC Workshops held in NRC Regions on Inspection Procedure 73756, dated July 18, 1997

[27] TDBD-1 21, "Topical Design Basis Document for In-Service Testing (IST)"

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 12 of 174 5.0 INSERVICE PUMP TESTING 5.1 GENERAL INFORMATION

[1] Applicable Code This Inservice Testing Program addresses those ASME Code Class 1, 2, and 3 centrifugal and positive displacement pumps that meet the requirements of Subsection ISTB of ASME OM Code 2004 Edition through Addenda 2006. The applicability of the subsection establishes the requirements for preservice and inservice testing to assess the pump's operational readiness. The pumps covered are those provided with an emergency power source and that are required in shutting down a Reactor to the safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident.

The following pump exclusions exist: drivers (except where the pump and driver form an integral unit and the pump bearings are in the driver), pumps that are supplied with emergency power solely for operating convenience and skid-mounted pumps that are tested as part of the major component. The skid-mounted pumps are excluded when they have been justified by the owner to be adequately tested.

Where the above requirements for ASME Class 1, 2, or 3 pumps are determined to be impractical, requests for relief have been written and are included in Section 7.0 as follows:

(a) RELIEF REQUEST (PR-XX) - Relief Requests are included in Section 7.1 and identify those impractical test requirements. A Relief Request must be reviewed and approved by the NRC prior to use.

[2] Pump Progqram Table Description The table in Section 5.3 lists all pumps included in the Pilgrim Nuclear Power Station (PNPS)

IST Program. This program defines pumps as mechanical devices used to move liquid. The program addresses centrifugal and positive displacement pumps according to ISTB-1 100, Applicability. Skid-mounted pumps have been identified for information, but are excluded from Code inservice testing. The column headings of the table are listed and explained below:

(a) System: System Title (b) Pump No.: Pump Identification Number (c) IST Class: IST Classification (Class 1, 2, or 3)

(d) P&ID/Coord.: PNPS drawing number and coordinate location (e) Pump Type A or B: Pump type as specified in OMb-2006 Section ISTB (f) Test A or B (Freq.) and Comprehensive Proc. (Freq.): The PNPS test Procedure number and frequency of inservice tests as prescribed in ISTB Table 3400-1, Inservice Test Frequency (g) Test Quantities (Speed, Press, Flow, and Vib): Inservice test quantities to be measured following the guidelines of ISTB Table 3000-1, Inservice Test Parameters

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 13 of 174

[3] Inservice Test Quantities Pump Speed (N), Discharge Pressure (Pd) Or Differential Pressure (Dp), Flow Rate (Q),

Vibration (V): When the symbol "Y" appears in a particular measured parameter column, that quantity will be measured during inservice testing in accordance with Subsection ISTB.

[4] Measurement of Inservice Test Quantities (a) Speed (N): In accordance with Table ISTB-3000-1, Inservice Test Parameters, shaft measurements are not applicable (NA) for pumps coupled to synchronous or induction type drivers, but only if variable speed. For variable speed pumps, the pump speed shall be set at the reference speed in accordance with ISTB-3510 and ISTB-3530.

(b) Discharge Pressure (Pd): For positive displacement pumps, discharge pressure will be measured in lieu of differential pressure in accordance with Table ISTB-3000-1 (all other pumps will have discharge pressure measured if it is needed to determine differential pressure).

(c) Differential Pressure (DP): Differential pressure measurements will be calculated from inlet and discharge pressure measurements or by direct differential pressure measurement. The differential pressure will be measured for centrifugal pumps, including vertical line shaft pumps, in accordance with Table ISTB-3000-1. The Service Water Pumps are an exception because the individual SSW pump suction is not configured to obtain pump suction pressure. This parameter is determined by analytical method using suction bay level instruments and is then converted to the equivalent suction pressure. The analytical method has been determined to meet the required instrument accuracy (+/-1/2%) of Table ISTB-3510-1 in accordance with ISTB-3510 (a).

(d) Vibration (V): Pump vibration will be measured on centrifugal, vertical line shaft, and reciprocating pumps. On centrifugal pumps, measurements shall be taken in a plane approximately perpendicular to the rotating shaft in two orthogonal directions on each accessible pump-bearing housing and on the axial direction on each accessible pump thrust bearing housing. On vertical line shaft pumps, measurements shall be taken on the upper motor-bearing housing in three approximately orthogonal directions, one being the axial direction. On reciprocating pumps, measurements shall be taken on the bearing housing of the crankshaft, approximately perpendicular to both the crankshaft and the line of plunger travel.

[5] Allowable Ran-qes of Test Quantities The allowable ranges specified in Table ISTB-5121-1, ISTB-5221-1, ISTB-5321-2 and Figure ISTB-5223-1 will be used, as applicable, for discharge pressure, differential pressure, flow measurements and vibration measurements for Group A, Group B and Comprehensive Pump Tests.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 14 of 174

[6] Corrective Action The operational readiness of a pump shall be determined by comparing Surveillance Procedure test results to the established Acceptable, Alert, and Required Action Ranges.

(a) If deviations fall within the Alert Range of Table ISTB-5121-1, ISTB-5221-1 or ISTB-5321-2 as applicable, the frequency of testing shall be doubled until the cause of the deviation is determined and the condition corrected.

(b) If deviations fall within the Required Action Range of Table ISTB-5121-1, ISTB-5221-1 or ISTB-5321-2 as applicable, the pump shall immediately be declared inoperative and not returned to service until the cause of the deviation has been determined and the condition corrected, or an evaluation is performed and new reference values are established in accordance with ISTB-3300 (c) In cases where the pump's test parameters are within either the Alert or Required Action Ranges of Table ISTB-5121-1, ISTB-5221-1 or ISTB-5321-2 as applicable, and the pump's continued use at the changed values is supported by an analysis, a new set of reference values may be established. This analysis shall include verification of the pump's operational readiness. The analysis shall include both a pump level and a system level evaluation of operational readiness, the cause of the change in pump performance, and an evaluation of all trends indicated by available data. The results of this analysis shall be documented in the record of tests.

(d) When a test shows a systematic error (improper system lineup or inaccurate instrumentation) such that a measured parameter value(s) falls outside the Acceptable Range, the test shall be rerun after correcting the error.

[7] Instrument Accuracy Instrument accuracy is defined as the allowable inaccuracy of an instrument loop (i.e., two or more instruments or components working together to provide a single output) based on the square root of the sum of the square of the inaccuracies of each instrument or component in the loop when considered separately. Alternatively, the allowable inaccuracy of the instrument loop may be based on the output for a known input into the instrument loop.

The plant-installed instrument accuracy bases for IST pump testing is contained within PNPS calculation number; IN1-299, "IST Instrument Calculation". An Engineering Design evaluation document may be used for an interim period to supplement IN1-299, when necessary due to an upgrade of plant instrumentation.

Allowable instrument accuracies are provided in Table ISTB-3510-1. If the accuracies of plant-installed instrumentation are not acceptable, temporary M&TE instruments meeting the acceptable accuracies will be used. For individual analog instruments, the required accuracy is a percent of full scale. For digital instruments, the required accuracy is over the calibrated range. For a combination of instruments, the required accuracy is loop accuracy.

Historical guidance provided within NUREG-1 482 and the summary of public workshops held in NRC regions on Inspection Procedure 73756 will be considered when meeting the OM Code instrument accuracy requirements.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 15 of 174 In cases where a parameter is determined by analytical methods instead of being measured, the requirements of ISTB-3510 must be adhered to.

[8] Exempt and Excluded Safety Related Pumps The Reactor recirculation pumps and the recirculation jet pumps are examples of pumps exempt (OM Code refers to exempt pumps as being excluded) from inservice testing.

Excluded pumps do not meet the applicability requirements of Subsection ISTB in that they are not required to perform a specific function in shutting down the Reactor to the safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident. Additionally, pumps that meet the exclusion requirements of ISTB-1200 (except for excluded skid-mounted pumps) have not been included into the program pump test table.

[9] Skid-Mount Pumps Skid-mounted pumps meet the exclusion requirements of ISTB-1200 and are defined as pumps that are integral to or that support operation of the major component, even though these pumps may not be located directly on the skid. Generally, the manufacturer of the major component supplies these pumps. Examples of systems having skids as a subassembly include High Pressure Coolant Injection System and Reactor Core Isolation Cooling System.

ASME Code Class 1, 2, and 3 skid-mounted pumps that are tested as part of the major component and are listed in the IST Program. In some cases supplemental skid testing may be performed. The IST program skid technical positions (STP-XX) and/or supporting basis information contain the bases used to determine that each pump is justified to be adequately tested.

[10] Test Requirements Preservice and inservice test periods are used to establish the proper test method in the establishment of an initial set of reference values for a new pump. The preservice test period is the time interval prior to or before implementing inservice testing (i.e., the start of the pump's service life). The inservice testing shall commence when the pump is required to be operable and continue until the pump is retired (i.e., removed from service).

[11] Test Group The IST Program centrifugal and positive displacement pumps are to be grouped by their inservice run time. Grouped as Type A or Type B, this allows for the development of unique test requirements due to the service the pump experiences as follows:

(a) Group A pumps are those operated continuously or routinely during normal operation, cold shutdown, or refueling operations.

(b) Group B pumps are those in standby systems that are not operated routinely except for testing.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 16 of 174

[12] Test Methods (a) Group A Test Method - shall be conducted with the pump operating at a specified reference point in accordance with ISTB-5121, ISTB-5221 or ISTB-5321 (as applicable). The test parameters shown in Table ISTB-3000-1 shall be determined and recorded.

(b) Group B Test Method - shall be conducted with the pump operating at a specified reference point. Tests shall be conducted in accordance with ISTB-5122, ISTB-5222 or ISTB-5322 (as applicable). The test parameters shown in Table ISTB-3000-1 shall be determined and recorded.

(c) Comprehensive Test Method - shall be conducted with the pump operating at a specified reference point. Tests shall be conducted in accordance with ISTB-5123, ISTB-5223 or ISTB-5323 (as applicable). The test parameters shown in Table ISTB-3000-1 shall be determined and recorded.

(d) Preservice Test Method - shall be conducted under conditions as near as practicable to those expected during subsequent inservice testing in accordance with ISTB-3100.

This test establishes the initial set of reference values (typically following the completion of construction activities and before first electrical generation by nuclear heat) and only one preservice test is required for each pump. The test parameters to be measured are specified in Table ISTB-3000-1.

[13] Test Frequency (a) Quarterly - both Group A and B pump test procedures are to be tested within a 92-day interval with an allowable extension of no more than 25%.

(b) Biennially - pumps must be tested using the Comprehensive pump test procedure within a 2-year interval with an allowable extension of no more than 25%.

(c) Pumps that are placed into Alert Status, test frequency will be tested within a 6 week test interval with No grace allowed

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 17 of 174 5.2 HYDRAULIC CIRCUITS The following IST hydraulic circuits are used to identify pump test paths and instrumentation.

Individual hydraulic and mechanical reference value test quantities are identified within each pump testing procedure.

M&TE instrumentation may be used in place of installed plant instruments provided that ISTB-3510(c) requirements are satisfied. The effects of flow losses and elevation differences due to instrument location were considered.

5.2.1 Salt Service Water (SSW) Pumps

[1] Test Group A - pumps operated continuously.

[2] Test/Frequency Method SSW pumps are tested quarterly using the Group A Test Method and biennially using the Comprehensive Test Method.

[3] Hydraulic Test Path Each pump will be tested by splitting the SSW loops and establishing single pump operation in the loop utilized for testing. Using the RBCCW and/or TBCCW Heat Exchanger Outlet Valves for throttling, the reference flow rate shall be established in accordance with current FSAR/Design Bases requirements. Pump discharge pressure shall be recorded and, using suction bay level measurement, the pump differential pressure (i.e., total dynamic head) will be calculated and compared to the established reference values. The suction pressure (bay level measurement) is determined by an analytical method instead of being measured by pressure instrument. This method is in compliance to Table ISTB-3510-1.

[4] Instrumentation Inlet Pressure, Pi (ft): EPIC Computer Points CWS154 and CWS156 (LI-38010 and LI-38011, EPIC Points CWS010 and CWS012, or LI-3831A and B may be used as backup).

DischarQe Pressure (psig): M&TE gauges at the locations of PI-3802, PI-3807, PI-3812, PI-3817, and PI-3822.

Flow rate, Q (GPM): FE-38002A/FIY-38003A, FE-38002B/FIY-38003B.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 18 of 174 5.2.2 Reactor Building Closed Cooling Water (RBCCW) Pumps

[1] Test Group A - pumps operated continuously.

[2] Test Frequency/Method RBCCW pumps are tested quarterly using the Group A Test Method and biennially using the Comprehensive Test Method.

[3] Hydraulic Test Path Each pump shall be tested by establishing single pump operation in the loop utilized for testing. Using the associated RBCCW Loop RHR Heat Exchanger Inlet Valve for throttling, the reference flow rate is established in accordance with current Technical Specifications/Design Basis requirements. Pump discharge and suction pressures shall be recorded. The differential pressure (total dynamic head) will be calculated and compared to the established value.

[4] Instrumentation (a) Inlet Pressure (psig): M&TE gauges at the locations of PI-4056A, PI-4054A, PI-4057A, PI-4006A, PI-4004A, and PI-4007A.

(b) Discharge Pressure (psig): M&TE gauges at the locations of PI-4056, PI-4054, PI-4057, PI-4006, PI-4004, and PI-4007.

(c) Flow rate, Q (GPM): FT-6265 read from EPIC point RBC002 or CPRBC002 (Loop A) or FT-6263 read from EPIC point RBC004 or CPRBC004 (Loop B).

5.2.3 Residual Heat Removal (RHR) Pumps

[1] Test Group A - pumps operated routinely during cold shutdown and refueling outages.

[2] Test Frequency/Method RHR pumps are tested quarterly using the Group A Test Method and biennially using the Comprehensive Test Method.

[3] Hydraulic Test Path Each pump shall be tested by establishing a flow path with suction from and discharge returning to the Torus (Heat Exchanger Bypass Valve open). Using the loop to the Suppression Chamber Spray Cooling Valve for throttling, the reference flow rate is established in accordance with current Technical Specifications requirements. Pump discharge and suction pressure shall be recorded. The differential pressure will be calculated and compared to the established value.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 19 of 174

[4] Instrumentation Inlet Pressure (psig): M&TE test gauges at PI-1001-70A, PI-1001-70B, PI-1001-70C, PI-1001-70D.

Discharge Pressure (psig): M&TE test gauges at PI-1001-71A, PI-1001-71iB, P1-1001-71C, PI-1001-71D.

Flow rate, Q (GPM): FI-1040-12A (Loop A) and FI-1 040-12B (Loop B). EPIC computer points RHR022 (Loop A) and RHR024 (Loop B) may also be used for flow rate measurement.

5.2.4 Core Spray (CS) Pumps

[1] Test Group B - pumps that are not operated routinely except for testing.

[2] Test Frequency/Method CS pumps are tested quarterly using the Group B Test Method and biennially using the Comprehensive Test Method.

[3] Hydraulic Test Path Each pump shall be tested by establishing a flow path with suction from and discharge returning to the Torus. Using the Core Spray Full Flow Test Valve for throttling, the reference flow rate is established in accordance with current Technical Specifications requirements. Pump discharge and suction pressures shall be recorded, and the differential pressure will be calculated and compared to the established value.

[4] Instrumentation (a) Inlet Pressure (psig): M&TE test gauges at PI-40A, PI-40B.

(b) Discharge Pressure (psig): M&TE test gauges at PT-1460A, PT-1460B.

(c) Flow rate, Q (GPM): Flow indicators FI-1450-4A (Loop A) and FI-1i450-4B (Loop B) or EPIC computer points CSP002 or CPCSP002 (Loop A) and CSP004 or CPCSP004 (Loop B).

5.2.5 High Pressure Coolant Injection (HPCI) Pump

[1] Test Group B - a pump that is not operated routinely except for testing.

[2] Test Frequency/Method The HPCI pump is tested quarterly using the Group B Test Method and biennially using the Comprehensive Test Method when adequate steam pressure is available.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 20 of 174

[31 Hydraulic Test Path The HPCI pump (main/booster integral unit) shall be tested by establishing a flow path with suction from and discharge returning to the CST. Using the HPCI Full Flow Test Valve for throttling, the reference pump speed and flow rate is established in accordance with current Technical Specifications requirements. Pump discharge and suction pressure shall be recorded. The differential pressure will be calculated and compared to the established value.

[4] Instrumentation (a) Inlet Pressure (psig): M&TE test gauge at PI-2381 (Quarterly and Biennial Comprehensive Test).

(b) Discharqe Pressure (psig): M&TE test gauge at PI-2357.

(c) Flow rate, Q (GPM): FI-2340-1.

(d) Speed, N (RPM): M&TE tachometer.

5.2.6 Reactor Core Isolation Cooling (RCIC) Pump

[1] Test Group B - a pump that is not operated routinely except for testing.

[2] Test Frequency/Method The RCIC pump is tested quarterly using the Group B Test Method and biennially using the Comprehensive Test Method when adequate steam pressure is available.

[3] Hydraulic Test Path The RCIC pump shall be tested by establishing a flow path from and returning to the CST.

Using the full flow test valve for throttling, the reference pump speed and flow rate is established in accordance with current Technical Specifications requirements. Pump discharge and suction pressures shall be recorded. The differential pressure will be calculated and compared to the established value.

[4] Instrumentation (a) Inlet Pressure (psig): M&TE test gauge at PT-1 360-19 (Quarterly and Biennial Comprehensive Test).

(b) Discharqe Pressure (psig): M&TE test gauge at P1-1360-5.

(c) Flow rate, Q (GPM): Fl-1 340-1.

(d) Speed, N (RPM): M&TE tachometer.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 21 of 174 5.2.7 Standby Liquid Control (SLC) Pumps

[1] Test Group B - pumps that are not operated routinely except for testing.

[2] Test Frequency/Method SLC pumps are tested quarterly using the Group B Test Method and biennially using the Comprehensive Test Method.

[3] Hydraulic Test Path Test Tank Method - each pump shall be tested by establishing a suction flow path from the main SLC Storage Tank and routing the discharge flow into the SLC Test Tank. Using the test valve for throttling, the pump reference discharge pressure is established prior to start of testing. The pump is then run for exactly 3 minutes. Initial and final test tank levels are measured before and after pump testing. The flow rate is calculated from the level change observed over the 3 minute run period, and compared to the established value. Pump flow rate is determined by an analytical method instead of being measured by a flow/rate meter.

This method for determining flow rate is in compliance to Table ISTB-351 0-1.

[4] Instrumentation (a) Inlet Pressure, Pi: Not applicable.

(b) Discharge Pressure, Pd (psig): PI-1 159 or M&TE test gauge.

(c) Flow rate, Q (inches): Test tank level change - measuring stick (graduated yardstick with minimum graduations of 1/8 inch).

5.3 PUMP PROGRAM TABLE The following pump table identifies the scope of pumps within the IST Program and allows cross-referencing specific pump test quantities to their implementing Station Procedure.

The test quantities measured include: Speed (N), Pressure - Discharge Pressure (Pd) or Differential Pressure (DP), Flow Rate (Q), and Vibration (V).

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 PUMP TABLE Rev.: 2 INSERVICE TEST QUANTITIES Page 22 of 174 PUMP IST TYPE TEST A OR B# COMPREHENSIVE# SPEED PRESS FLOW SYSTEM NO. CLASS P&ID/COORD. A or B PROC. (FREQ.) PROC. (FREQ.) (N) DP Pd RATE (Q) VIB.

SALT SERVICE P-208A 3 M212/A8 A 8.5.3.2.1 (Q) 8.5.3.2.1 (2Y) NA Y NA Y Y WATER (SSW) P-208B 3 M212/A7 A 8.5.3.2.1 (Q) 8.5.3.2.1 (2Y) NA Y NA Y Y P-208C 3 M212/A6 A 8.5.3.2.1 (Q) 8.5.3.2.1 (2Y) NA Y NA Y Y P-208D 3 M212/A4 A 8.5.3.2.1 (Q) 8.5.3.2.1 (2Y) NA Y NA Y Y P-208E 3 M212/A5 A 8.5.3.2.1 (Q) 8.5.3.2.1 (2Y) NA Y NA Y Y REACTOR P-202A 3 M215 (S1)/D6 A 8.5.3.1 (Q) 8.5.3.1 (2Y) NA Y NA Y Y BUILDING P-202B 3 M215 (S1)/E6 A 8.5.3.1 (Q) 8.5.3.1 (2Y) NA Y NA Y Y CLOSED P-202C 3 M215 (S1)/F6 A 8.5.3.1 (Q) 8.5.3.1 (2Y) NA Y NA Y Y COOLING P-202D 3 M215 ($2)/G3 A 8.5.3.1 (Q) 8.5.3.1 (2Y) NA Y NA Y Y WATER (RBCCW) P-202E 3 M215 ($2)/F3 A 8.5.3.1 (Q) 8.5.3.1 (2Y) NA Y NA Y Y P-202F 3 M215 (S2)/E3 A 8.5.3.1 (Q) 8.5.3.1 (2Y) NA Y NA Y Y RESIDUAL HEAT P-203A 2 M241 (S2)/D6 A 8.5.2.2.1 (Q) 8.5.2.2.1 (2Y) NA Y NA Y Y REMOVAL (RHR) P-203B 2 M241 ($2)/D4 A 8.5.2.2.2 (Q) 8.5.2.2.2 (2Y) NA Y NA Y Y P-203C 2 M241 ($2)/F6 A 8.5.2.2.1 (Q) 8.5.2.2.1 (2Y) NA Y NA Y Y P-203D 2 M241 (S2)/F4 A 8.5.2.2.2 (Q) 8.5.2.2.2 (2Y) NA Y NA Y Y CORE SPRAY (CS) P-215A 2 M242/C4 B 8.5.1.1 (Q) 8.5.1.1 (2Y) NA Y NA Y Y*

P-215B 2 M242/C3 B 8.5.1.1 (Q) 8.5.1.1 (2Y) NA Y NA Y Y*

HIGH PRESSURE P-205 2 M244/E5 B 8.5.4.1 (Q) 8.5.4.1 (2Y) Y Y NA Y PR-03++

COOLANT INJECTION (HPCI) P-220 2 M244/B2 NA## 8.5.4.1 (Q) NA NA NA NA NA NA REACTOR CORE P-206 2 M246/E5 B 8.5.5.1 (Q) 8.5.5.1 (2Y) Y Y NA Y Y**

ISOLATION P-221 2 M246/B3 NA ## 8.5.5.1 (Q) NA NA NA NA NA NA COOLING (RCIC)

STANDBY LIQUID P-207A 2 M249/E5 B 8.4.1 (Q) 8.4.1 (2Y) NA NA Y PR-05+ Y**

CONTROL (SLC) P-207B 2 M249/D5 B 8.4.1 (Q) 8.4.1 (2Y) NA NA Y PR-05+ Y**

The required instrument accuracy for pressure and differential pressure when performing Group A or B Tests is 2% and Comprehensive or Preservice Tests is 1/2%.

1. Skid-mounted pump excluded from testing. Refer to skid technical positions STP-02 (P-221) and STP-03 (P-220).

- Group B Test Method only requires monitoring of hydraulic parameters (i.e., vibration parameters not required).

+ PR-05 applies to SLC Biennial Comprehensive Testing only.

++ HPCI Pump Group B Test - PR-03 commits PNPS to monitor vibration parameters on a quarterly (typically) basis.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 23 of 174 6.0 INSERVICE VALVE TESTING 6.1 GENERAL INFORMATION

[1] Applicable Code This Inservice Testing Program addresses those valves and pressure relief devices (and their actuating and position indicating systems) that meet the requirements of Subsection ISTC of ASME OM Code 2004 Edition through Addenda 2006. The applicability of the subsection establishes the requirements for preservice and inservice testing to assess the valve's or pressure relief device's operational readiness. Both active and passive valves are covered by this Procedure and include those that are required in shutting down the Reactor to the safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident. The pressure relief devices covered are those for protecting systems or portions of systems that are required in shutting down the Reactor to the safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident. The following exclusions exist: valves used only for operating convenience (i.e., vent, drain, instrument, and test valves), valves used only for system control (i.e., pressure regulating valves), and valves used only for system or component maintenance. Skid-mounted valves and component subassemblies that are tested as part of the major component are excluded when justified by the owner to be adequately tested. External control and protection systems responsible for sensing plant conditions and providing signals for valve operation are excluded. Additionally, nonreclosing pressure relief devices (rupture disks) used in BWR Scram Accumulators are excluded.

Where the above requirements for ASME Class 1, 2, or 3 valves and pressure relief devices are determined to be impractical, specific justifications and requests for relief have been written and included in Section 7.0 as follows:

(a) RELIEF REQUEST (VR-XX): Relief Requests are included in Section 7.5 and identify impractical test requirements. These Relief Requests must be reviewed and approved by the NRC prior to use.

(b) COLD SHUTDOWN JUSTIFICATION (CS-XX): These justifications are located in Section 7.2. The justification provides the mechanism for documenting the bases for performing a specific test on a cold shutdown (CS) frequency.

(c) REFUELING OUTAGE JUSTIFICATION (RJ-XX): These justifications are located in Section 7.3. The justification provides the mechanism for documenting the bases for performing a specific test on a refueling interval (RI) frequency (usually during a refueling outage) as well as specifying the alternative requirements.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 24 of 174

[2] Inservice Valve Testing Program Table The tables contained in Section 6.2 list all IST Class 1, 2, and 3 valves that are tested to meet IST requirements. Skid-mounted valves have been identified, but are excluded from testing.

The tables are sorted by system and Piping and Instrumentation Diagram (P&ID) number(s) and contain the following information.

(a) Valve Number: Valve identification number.

(b) P&ID Coord: Coordinate location and sheet number on the P&ID.

(c) IST Class: IST Classification (Class 1, 2, or 3)

(d) Valve Cat: Category assigned to the valve is based on ISTC-1 300, Categories of Valves. Categories A through D are defined in the Code subsection. Category AC is specified when more than one distinguishing characteristic from both A and C categories is applicable. However, the duplication of common testing requirements (i.e., leak testing) is not necessary.

(1) Category A - valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their function.

(2) Category B - valves for which a specific amount of leakage in the closed position is inconsequential for fulfillment of their function.

(3) Category C - valves which are self-actuating in response to some system characteristic, such as pressure (safety and relief valves including vacuum relief valve) or flow direction (check valves).

(4) Category D - valves which are actuated by an energy source capable of only one operation such as rupture discs or explosive actuated valves.

(5) Category AC - valves which exhibit both Category A and Category C characteristics.

(e) Valve Size: Nominal pipe size (in inches).

(f) Valve Type: Valve Body Design; ANGLE AN PRESS. CONTROL VALVE PCV BALL BL PLUG PG BUTTERFLY BF RELIEF RL CHECK CK RUPTURE DISC RD EXCESS FLOW EF SAFETY SV GATE GA SHEAR SH GLOBE GL STOP CHECK SC GOVERNOR GV SPRING CHECK SK NEEDLE ND

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 25 of 174 (g) Actuator Type: Valve Actuator Power; MOTOR OPERATOR MO EXPLOSIVE ACTUATOR EX AIR OPERATOR AO MANUAL MA SOLENOID OPERATOR SO SELF-ACTUATED SA HYDRAULIC OPERATOR HO (h) Normal Position: Normal Position during Plant Operation; NORMAL OPEN 0 LOCKED OPEN LO NORMAL CLOSED C LOCKED CLOSED LC (i) Test Requirement: Test(s) that will be performed to fulfill the requirements of Subsection ISTC. Test definitions and abbreviations are identified in Table 1.

(j) Test Frequency: Frequency at which the tests will be performed. The frequency definitions and abbreviations used are identified in Table 2, Test Frequency.

(k) PNPS Procedure No.: The PNPS Surveillance Procedure that satisfies each specific test requirement.

(I) Safety Direction: Direction the valve is exercised to during stroke time measurement or a check valve exercise. The direction for power operated valve stroke timing and check valve exercising is that which is required to fulfill the valve's safety-related function(s) according to the Updated FSAR and/or Technical Specifications.

OPEN 0 CLOSED C (m) Relief/Justification: Refer to the following sections: Section 7.2 for cold shutdown Justifications (CS-XX), Section 7.3 for Refuel Outage Justifications (RJ-XX), and Section 7.4 for Disassembly Examination Justifications (DJ-XX), and Section 7.5 for Valve Relief Requests (VR-XX). Although they are not categorized as a relief or justification, Valve Technical Positions (VTP-XX) (refer to Section 7.6) and Skid-Mount Component Technical Positions (STP-XX) (refer to Section 7.7) will be listed in this column.

(n) Notes: Clarification to identify any of the following information:

(1) Special or unique classification (i.e., Exempt, Passive)

(2) Specific information related to that test requirement (3) Alternate test procedure(s)

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 26 of 174

[3] Exempt and Excluded Valves Valves exempt according to ISTC-1200 are not required to be included in the IST Program.

Subsection ISTC exempts valves that are not required to perform a specific function in shutting down the Reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident. Exempt and excluded valves include the following:

Valves used only for operating convenience, and/or system or component maintenance are exempt.

Valves used only for system control are excluded (and may satisfy the scoping requirements of exempt).

Skid-mounted valves and component subassemblies are excluded, provided that they are tested as part of the major component and justified by the owner to be adequately tested.

" External control and protection systems responsible for sensing plant conditions and providing signals for valve operation are excluded (and meet the scoping requirements of exempt).

Nonreclosing pressure relief devices (rupture disks) used in BWR Scram Accumulators are excluded (and meet the scoping requirements of exempt).

Control valves whose actuators are required to provide a fail-safe position have been included in the program in accordance with ISTC-5100 and are fail-safe tested only.

Exempt/excluded valves with remote position indicators have been included so their position indicators will be checked with the "exempt" status noted.

[4] Manual Valves Manual valves with an active safety function shall be full-stroke exercised at least once every two years. If not practicable to exercise a manual valve at power, it shall be exercised during each refueling outage.

If a manual valve fails to exhibit the required change of obturator position, the valve shall be immediately declared inoperable. Manual valves equipped with remote position indication shall receive a position indication verification test at least once every two years.

[5] Passive Valves Passive valves are valves that maintain obturator position and are not required to change obturator position to accomplish the required function(s) (refer to ISTA-1 100).

Passive valves will be tested in accordance with Subsection ISTC, Table ISTC 3500-1 with the "passive" status noted. Passive valves that do not; have remote position indicators, perform an identified system boundary isolation function, or have a specific maximum leakage requirement are not included in the IST Program.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 27 of 174

[6] Check Valve Condition Monitoring Program (CMP)

PNPS has developed a check valve Condition Monitoring Program in accordance with ISTC-5222. This program complies with OM Code Appendix II and is an alternative to the testing requirements of ISTC-3510, ISTC-3520, ISTC-3530, ISTC-3550, and ISTC-5221 of the OM Code. Check valves have been placed in the Condition Monitoring Program with the purpose of improving check valve performance and to optimize activities to maintain the continued acceptable performance of the selected valves.

The Valve Program Tables in Section 6.2 of this Procedure assign a test frequency of "CMP" for each valve selected as part of the Condition Monitoring Program. The Valve Program table for CMP valves may continue to specify other Code-related information, such as justifications, notes, safety direction, and test requirements for the selected valves. These items will have no impact on the CMP but will remain for historical informational purposes.

ISTC- 5222 requires that if Condition Monitoring is discontinued for a valve or valve group, then the requirements of ISTC-3510, ISTC-3520, ISTC-3530, ISTC-3550, and ISTC-5221 of the OM Code shall apply.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 28 of 174 The following check valves have been placed into the PNPS Condition Monitoring Program:

CONDITION MONITORING PROGRAM VALVES CMP No. Valve No. Valve Title CMP-01 CK-1001-2A RHR Pump A Minimum Flow Check Valve CK-1001-2B RHR Pump B Minimum Flow Check Valve CK-1001-2C RHR Pump C Minimum Flow Check Valve CK-1001-2D RHR Pump D Minimum Flow Check Valve CMP-02 CK-1001-68B RHR Loop B Injection Check Valve CMP-03 CK-1001-362B RHR Loop B Keepfill Check Valve CK-1001-363A RHR Loop A Keepfill Check Valve CK-1 400-212A CS Loop A Keepfill Check Valve CK-1 400-212B CS Loop B Keepfill Check Valve CMP-04 CK-1 400-9A CS Loop A Injection Check Valve CK-1400-9B CS Loop B Injection Check Valve CMP-05 CK-1400-13A CS Pump A Minimum Flow Check Valve CK-1400-13B CS Pump B Minimum Flow Check Valve CMP-06 CK-1400-36A CS Pump A Discharge Check Valve CK-1400-36B CS Pump B Discharge Check Valve CMP-07 CK-1 101-16 SLC Outboard Containment Isolation Valve CK-1 101 -43A SLC Pump A Discharge Check Valve CK-1101-43B SLC Pump B Discharge Check Valve CMP-08 CK-2301-40 HPCI Pump Minimum Flow Check Valve CMP-09 CK-1301-27 RCIC Torus Pump Suction Check Valve CK-2301-39 HPCI Torus Pump Suction Check Valve CMP-10 CK-1301-23 RCIC Condensate Storage Tank Pump Suction Check Valve CK-2301-20 HPCI Condensate Storage Tank Pump Suction Check Valve CMP-11 CK-1301-47 RCIC Pump Minimum Flow Check Valve CMP-12 CK-1400-35 CS Pump A Test Return Line Check Valve CK-1400-214 CS Pump B Test Return Line Check Valve CMP-1 3 CK-1001-515 Torus Makeup from Condensate Check Valve CMP-14 2-CK-125A EFCV for C2205 by Dragon 2-CK-1 25B EFCV for C2206 by Dragon CMP-15 CK-2301-218 HPCI Turbine Exhaust Drain Check Valve CMP-16 See Note 1 Chemiquip EFCV Note 1: Inservice Valve Testing Program Table contains a complete listing of Chemiquip, Excess Flow Check Valves.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 29 of 174

[7] Measurement of Test Quantities (a) Stroke Time: The preferred stroke time method is the time interval from initiation of the actuating signal to the end of the actuating cycle. Alternate stroke timing methods may be utilized and are discussed within SEP-PNPS-IST-009, "PNPS Administrative Guidelines for the Inservice Code Testing (IST) and Appendix B Testing (ABT) Programs" The stroke time acceptance value(s) for each power-operated valve is specified within the appropriate test Procedure. These times shall be measured at least to the nearest second. Additionally, abnormal or erratic action discovered during testing shall be recorded and evaluated.

(b) Position Indication: Valve disc movement is determined by exercising the valve while locally observing the appropriate indicators which signal the required change of disc position. Where local observation is not possible, other indications shall be used for verification of valve operation. A suitable method is the use of indirect evidence (such as changes in system pressure, flow rate, level, or temperature) which reflects stem or disc position to verify that remote position indicators agree with valve operation.

(c) Seat Leakagqe: Seat leakage is measured by one of the following methods:

(1) Measuring leakage through a downstream telltale connection while maintaining test pressure on one side of the valve.

(2) Measuring the feed rate required to maintain test pressure in the test volume or between two seats of a gate valve, provided that the total apparent leakage rate is charged to the valve or valve combination or gate valve seat being tested and the conditions required by ISTC-3630(b) are satisfied.

(3) Determining leakage by measuring pressure decay in the test volume, provided that the total apparent leakage rate is charged to the valve or valve combination or gate valve seat being tested and the conditions required by ISTC-3630(b) are satisfied.-

Exception: Containment isolation valves are seat leak tested in accordance with the PNPS 10CFR50 Appendix J program.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 30 of 174 (d) Check Valve Exercise Test: Each OM Code exercise test shall include both exercise open and exercise close tests. Exercise open and exercise close tests need only be performed at an interval when it is practicable to perform both tests. Test order (e.g.,

whether the exercise open test precedes the exercise close test) is not important and not required to be performed at the same test frequency as long as they are both performed within the same test interval. Observations shall be made by observing a direct indicator (e.g., a position-indicating device) or by other positive means (e.g.,

changes in system pressure, flow rate, level, temperature, seat leakage, testing, or nonintrusive testing results). The following test practices will be utilized:

(1) Safety function in both open and close directions - shall be tested 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) and verify that, on cessation or reversal of flow, the obturator has traveled to the seat.

(2) Safety function in open direction only - shall be tested 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) and verify closure.

(3) Safety function in closed direction only - shall be tested by initiating flow and observing that the obturator has traveled to at least the partially open position (i.e., normal or expected system flow position) and verify that, on cessation or reversal of flow, the obturator has traveled to the seat.

(4) Mechanical Exerciser - the force(s) or torque(s) required to move the obturator and fulfill its safety functions shall meet the acceptance criteria of the Owner (PNPS). The exercise testing shall meet the requirements of ISTC-5221(b)(1), (2), and (3).

When a reference value torque or force is being established by the Owner, it shall be obtained when the valve is known to be operating properly.

Due to valve and exercise mechanism design, it is not practicable to evaluate closing torque on the two valves (CK-2301-7 and CK-1301-50) within the IST Program that undergo mechanical exercising.

(5) Vacuum Breaker - for check valves that perform as vacuum breakers, the exerciser force or torque delivered to the disc may be equivalent to the desired functional pressure differential force. Also, the disc movement shall be sufficient to prove that the disc moves freely off the seat. If no functional pressure differential force is specified, only disc movement is required.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 31 of 174 (e) Check Valve Sample Disassembly: For certain check valves in which the safety function (i.e., sufficient flow, full open position or travel to seat) cannot be satisfied by the acceptable test methods, a sample disassembly examination program may be used to verify valve obturator movement. The sample program has the following requirements:

(1) Grouping of check valves to be technically justified and consider, as a minimum, manufacturer, design, service, size, materials of construction, and orientation.

(2) The full-stroke motion of the obturator shall be verified during disassembly.

The full stroke motion of the obturator shall be reverified immediately prior to completing reassembly. Additionally, valves that have their obturator disturbed before the full-stroke motion (e.g., spring-loaded lift check valves or check valves with the obturator supported from the bonnet) are examined to determine whether a condition exists that could prevent full opening or reclosure of the obturator.

(3) One check valve of each group is disassembled and examined each refueling outage with all valves in each group being disassembled and examined at least once every 8 years.

(4) Valves that were disassembled and examined or that received maintenance that could affect their performance shall be exercised if practicable prior to being returned to service.

The disassembly of check valves for the sample disassembly program (when active) will be performed in accordance with PNPS 3.M.4-53. Currently no check valves are assigned to the sample disassembly program.

(f) Nonintrusive Testinq of Check Valves: Nonintrusive testing (NIT) is an acceptable method for verifying valve obturator movement in accordance with ISTC-5221 (a).

However, if the NIT results are inconclusive for verifying the check valve exercise test, then another acceptable method may be used. If no other methods can be used to perform the exercise test, then the affected valve may be disassembled using the guidelines within PNPS 3.M.4-53, "Check Valve Disassembly and Inspection".

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 32 of 174 (g) Excess Flow Check Valve (EFCV) Testingq Relaxation: NEDO-32977-A and the associated NRC Safety Evaluation, dated March 14, 2000, provide the basis for relaxing the EFCV testing frequency. The relaxation of the number of EFCVs tested every refueling outage is from "each" to a "representative sample" (nominally once every 24 months). The representative sample is based on (approximately) 20% of the valves tested each 2-year cycle such that each valve is tested every 10 years (nominal).

All applicable EFCVs have been placed within the PNPS Condition Monitoring Program (CMP) in accordance with ISTC-5222; therefore the aforementioned EFCV Testing Relaxation determination is currently not being applied.

(h) Relief Valves: Periodic testing of pressure relief valves is required. Detailed administrative testing guidance is provided within SEP-PNPS-IST-005 "PNPS Administration of Inservice Pressure Relief Device Test Program". No maintenance, adjustment, disassembly, or other activity which could affect "as found" set-pressure or seat tightness data is permitted prior to testing.

(1) Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices are to be tested in the following sequence: visual examination, seat tightness determination, and set-pressure determination. Then the remaining may be performed following maintenance or set-pressure adjustment: electrical characteristic and pressure integrity of solenoid valves, pressure integrity and stroke capability of air actuator, electrical characteristics of position indicators, electrical characteristics of bellows alarm switch, actuating pressure of sensing element devices and continuity, and owner's seat tightness criteria, when applicable.

(2) Class 1 Main Steam Pressure Relief Valves without Auxiliary Actuating Devices are to be tested in the following sequence: visual examination, seat tightness determination, and set-pressure determination. Then the remaining may be performed following maintenance or set-pressure adjustment:

electrical characteristics of position indicators and owner's seat tightness criteria, when applicable.

(3) Class 2 and 3 Pressure Relief Valves are to be tested in the following sequence: visual examination, seat tightness determination, and set-pressure determination. Then the remaining may be performed following maintenance or set-pressure adjustment: integrity of the balancing device on balanced valves and owner's seat tightness criteria, when applicable.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 33 of 174

[8) Allowable Ranges of Test Quantities (a) Stroke Time: Stroke time acceptance criteria are categorized as being applicable to either electric motor-operated or power-operated (e.g., air-operated, solenoid-operated, hydraulic-operated, etc.) valves. The observed stroke times will not exceed the following Acceptance Criteria:

(1) Electric motor-operated valves with reference stroke times greater than 10 seconds shall exhibit no more than +15% change in stroke time when compared to the reference value.

(2) Power-operated valves with reference stroke times greater than 10 seconds shall exhibit no more than +/-25% change in stroke time when compared to the reference value.

(3) Electric motor-operated valves with reference stroke times less than or equal to 10 seconds shall exhibit no more than a +/-25% nor +/-1 second change in stroke time, whichever is greater, when compared to the reference value.

(4) Power-operated valves with reference stroke times less than or equal to 10 seconds shall exhibit no more than +/-50% change in stroke time when compared to the reference value.

(5) Valves with stroke times less than 2 seconds may be exempted from the reference value change requirement. For these cases, the maximum limiting stroke time will be 2 seconds.

(b) Position Indication: Valves with remote position indicators will be checked to verify that remote indicators accurately reflect valve operation. The OM Code does not require the periodic surveillance position indication verification to be performed at remote panels (i.e., valve Alternate Shutdown Panels).

(c) Seat Leakage:

(1) Valve leakage rates shall not exceed the value established by PNPS Engineering. These leakage values are specified within the implementing test Procedures. Valves (or valve combinations) that fail to meet the acceptance criteria shall be declared inoperable immediately and require corrective action.

(2) The Code recommended "permissible" leakage rates for a valve or a valve combination not specified by the owner (PNPS Engineering) are as follows:

for water 0.5D gal/min or 5 gal/min, whichever is less, at function pressure differential; or for air, at function pressure differential, 7.5D standard ft3/day where D equals nominal valve size, inches. Refer to ISTC-3630(e) for further guidance.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 34 of 174

[9] Instrument Requirements Instruments used to measure stroke times shall be capable of measurement to the nearest tenth of a second.

[10] Corrective Action (a) Stroke Time:

(1) If a valve fails to exhibit the required change of obturator position or exceeds the limiting value of full-stroke time, it shall be immediately declared inoperable.

(2) Valves that do not meet the measured stroke time acceptance criteria shall be immediately retested or declared inoperable. If the valve is retested and the second set of data does not meet the acceptance criteria, the data shall be analyzed within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to verify that the new stroke time represents acceptable valve operation, or the valve shall be declared inoperable. If the second set of data meets the acceptance criteria, the cause of the initial deviation shall be analyzed and the results documented. The second stroke data analysis will be completed within an administrative time period of 30 days. When necessary a third stroke will be performed, following a predetermined wait period to address any potential preconditioning concerns.

(3) Valves declared inoperable may be repaired, replaced, or the data may be analyzed to determine the cause of the deviation and the valve shown to be operating acceptably.

(4) Before returning a repaired or replacement valve to service, a test shall be performed demonstrating satisfactory operational readiness.

(5) A completed Attachment 3 of SEP-PNPS-IST-008 combined with either a Corrective Action/Evaluation Form (Attachment 3 of SEP-PNPS-IST-009) or a Surveillance Review Sheet (reference PNPS 1.3.34 Attachment 9) with ASME Test Engineer signature, satisfies the corrective action evaluation requirements for valve operability. If necessary, a valve may be declared operable using the Surveillance Test Review process prior to revision of the affected stroke timing Procedure. A revision to the Surveillance Procedure should then be implemented on a priority basis for future operability testing.

(b) Seat Leakage: Corrective action shall be to repair or replace the valve or valve combination unless an Engineering analysis is performed which demonstrates that the leakage criteria may be increased and the new leakage rate will not impact the valve's ability to fulfill its safety function. The option to evaluate and increase leakage criteria should not be used for Reactor system pressure isolation valves (PIVs) without performance of an evaluation which includes consideration of applicable Code requirements.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 35 of 174 (c) Relief Valves:

(1) Relief valves which fail to function properly during testing shall be:

1) adjusted, 2) repaired, or 3) replaced and retested to show acceptable operation. Procedure credit may be taken for valves that are tested by the vendor/manufacturer provided that similar test methods are used and the documentation verifies that the Surveillance Procedure acceptance criteria have been met.

(2) The requirements for testing additional valves following set pressure test failures for Class 1, 2, and 3 safety and relief valves are specified within PNPS Administrative Procedures PNPS 8.1.26.3 and SEP-PNPS-IST-005.

TABLE 1 INSERVICE VALVE TESTS Test Test Name Test Description/Definition LJ Containment Containment Isolation Valves (CIVs) will be seat leak tested using air in Isolation Valves (Air) accordance 10CFR50 Appendix J as specified in ISTC-3620, Containment Isolation Valves.

LJW Containment Containment Isolation Valves (CIVs) will be seat leak tested with water at a Isolation Valves pressure not less than 1.1 Pa and are not required to be added into the Type (Water) C leak test total. These isolation valves are tested to assure the seal-water fluid inventory is sufficient to assure the sealing function for at least 30 days.

Testing will be conducted in accordance with ISTC-3620.

LP Pressure Isolation Pressure Isolation Valves (PIVs) are any two valves in series within the Valves Reactor coolant pressure boundary which separate the high pressure Reactor coolant from an attached low pressure system and are normally closed. These valves will be seat leak tested in accordance with ISTC-3610 and ISTC-3630. The basis for PIV selection is provided in PNPS's response to Generic Letter 87-06.

LSBI System Boundary An augmented seat leakage test that will be applied to owner selected Isolation valves that close (or remain closed) to perform a system boundary isolation function. This augmented leakage test will provide supplemental verification of satisfactory valve closure for valves that have been defined to provide a system boundary isolation function. This augmented (Non-Code) seat leakage testing will supplement the Code specified valve closure testing, and will be performed using the methods outlined within ISTC-3630.

Active LSBI valves - will be tested on a performance based test frequency as specified within Valve Technical Position VTP-01.

Passive LSBI valves - unless otherwise noted, will be conditionally tested only after the valves have undergone manipulation. An administrative time period of three months, following valve manipulation, has been adopted for performance of the Condition Based Testing (CBT) on passive LSBI valves.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 36 of 174 TABLE 1 INSERVICE VALVE TESTS Test Test Name Test Description/Definition LEF Excess Flow Check Excess Flow Check Valves (EFCVs) are a single instrument line isolation Valves valve within the reactor coolant pressure boundary which separates the high pressure reactor coolant from the dead-ended (Reactor Building) instrument, in the unlikely event of a line rupture. This augmented gross seat leakage test will be applied to verify each EFCV closes to limit flow when actuated.

The augmented (Non-Code) seat leakage testing will supplement the Code specified check valve closure testing, and will be performed using the methods outlined within ISTC-3630.

LX Miscellaneous Other safety related valves - Miscellaneous Isolation Valves will be seat leak Isolation Valves tested in accordance with ISTC-3610 and ISTC-3630.

FE Full Stroke Exercise Exercise testing of Category A or B valves through one complete cycle of operation.

1) Normally open: Full stroke exercise the valve closed, then return to open position.

2) Normally closed: Full stroke exercise the valve open, then return to closed position ST Stroke Time Stroke time is the measurement of the time required to exercise test a Category A or B valve through an operation. The direction for stroke time measurements is that which is required to fulfill its safety-related function in accordance with the Updated FSAR and/or Technical Specifications.

PE Partial Stroke Partial stroke exercise testing will be performed on those Category A or B valves that cannot be full stroke exercised during plant operation and have the capability to be partially exercised. Then, full stroke exercise shall be performed during cold shutdowns, unless valves can only be partially exercised during cold shutdown.

OT Exercise Open Test The open test of a check valve exercise test performed by verifying the check valve in the open position by observing a direct indicator or by other positive means.

CT Exercise Close Test The close test of a check valve exercise test performed by verifying the check valve in the closed position by observing a direct indicator or by other positive means.

RD Rupture Disc Test Rupture disks (nonclosing that are not testable) were test certified by the manufacturer or the startup testing program and no additional testing shall be required.

EX Explosive Test Testing of explosive charges by firing in accordance with ASME OM Code with at least 20% of the charges in a batch fired every 2 years with no charge exceeding 10 years.

RT Relief Setpoint Test Relief and Safety Valve setpoints will be verified in accordance with ASME OM Code, Appendix I.

FS Fail-Safe Test Valves with fail-safe actuators (e.g., air operated, spring loaded, solenoid operated, and hydraulic operated) will be tested to verify proper fail-safe operation upon loss of actuator power.

P1 Position Indication Valves with remote position indicators will be checked to verify that remote Verification valve position indicators accurately reflect valve operation. The OM Code does not require the position indication verification to be performed at remote panels (i.e., valve Alternate Shutdown Panels).

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 37 of 174 TABLE 2 TEST FREQUENCY Test Freq. Operational Condition Frequency of Testing Q Power Operation At least once per 92 days, quarterly (Q).

CS Cold Shutdown See (2) below.

RI Refueling Interval See (1) below. Testing must be performed prior to returning the plant to operation from a refueling outage.

RI# Refueling Interval See (1) below. The frequency applies to check valve exercise open and exercise close tests that are not refueling outage dependent.

The exercise open or close test must be performed within plus or minus 6 months from when the corresponding check valve refueling outage test is performed.

OBJ No operational condition Testing frequencies will be established using the performance-based limitations intervals in Option B of 10CFR50 Appendix J (OBJ). Applies to Containment Isolation Valve (CIV) seat leakage tests (LJ and LJW).

See (3) below.

PBT No operational condition Testing frequencies will be established using a Performance-Based limitations Test (PBT) interval as identified in VTP-01. See (3) below.

CBT No operational condition Conditioned-Based Test (CBT) frequency will correspond to limitations applicable component conditions that are specified (such as valve manipulation of passive manual valves). No routinely scheduled testing applies to this test frequency, PWT No operational condition Testing frequency will correspond to applicable post work tests (PWT) limitations following applicable routine servicing, preventive or corrective maintenance. No routinely scheduled testing applies to this test frequency.

CMP No operational condition Testing frequencies will be established using the Condition Monitoring limitations Program (CMP) intervals.

2Y No operational condition Every 2 years. Applies to seat leakage (excluding CIVs, LSBI tests, limitations and LEF tests) (ISTC-3630 and explosively actuated valve (sample frequency) tests.

5Y No operational condition Every 5 years. Applies to ASME Class 1 Safety/Relief Valves (see limitations ASME OMb Code, Appendix 1, 1320) and ASME Class 1, 2, and 3 Rupture Discs (see ASME OMb Code, Appendix 1, 1330 & 1360).

10Y No operational condition Every 10 years. Applies to ASME Class 2 & 3 Safety/Relief Valves, limitations Thermal Relief Valves (see ASME OMa Code, Appendix I, 1350 and 1390) and explosively actuated valve tests (ISTC-5260).

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 38 of 174 TABLE 2 TEST FREQUENCY (1) Refueling outage conditions are as contained in the definitions of the PNPS Technical Specifications. For inservice testing purposes, the refueling interval associated with refuel outage testing may be up to 2 years with an allowable extension of no more than 25%.

(2) Plant cold shutdown (reference: ISTC-3521(g), ISTC-3522(e), and PNPS Technical Specifications) testing is acceptable when the following conditions are met:

(a) Testing is to commence as soon as practical when the cold shutdown condition is achieved, but not later than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after shutdown, and continue until complete or the plant is ready to return to power.

(b) Completion of all testing is not a prerequisite to return to power. Any testing not completed during one cold shutdown should be performed during any subsequent cold shutdown starting with those tests not previously completed.

(c) Testing need not be performed more often than once every 3 months.

(d) In the case of extended cold shutdowns, the testing need not be started within the 48-hour limitation. However, in extended cold shutdowns, all cold shutdown testing must be completed prior to returning to power.

(3) The Containment Isolation Valve seat leakage test frequencies will utilize the performance-based test intervals in Option B of 10CFR50 Appendix J. These seat leakage intervals range from test once per 30 months to 60 months and are established and controlled in accordance with PNPS 8.7.1.3.1.

Active System Boundary Isolation (SBI) valves that have been assigned an augmented seat leakage criterion to verify satisfactory valve closure (LSBI) will also utilize these performance-based test intervals. For these SBI valves, PNPS 8.7.1.3.1 has been adopted as a guideline (Reference VTP-01).

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 39 of 174 TABLE 3 TERMINOLOGY COMPARISON PNPS Technical Specifications Surveillance Interval definitions are to be applied to the following required frequencies for performing inservice testing activities. These required intervals may be extended as allowed by the PNPS Technical Specifications "Surveillance Frequency" definition.

ASME OM CODE TERMINOLOGY REQUIRED FREQUENCIES FOR FOR IST ACTIVITIES PERFORMING IST ACTIVITIES Weekly At least once per 7 days Monthly At least once per 31 days 6 Weeks* At least once per 46 days Quarterly (every 3 months) At least once per 92 days Semiannually (every 6 months) At least once per 184 days Yearly At least once per 366 days Biennial or every 2 years At least once per 732 days

Six week pump Alert Test frequency is not identified within the PNPS Technical Specifications, and therefore, does not qualify for the 25% grace period.

Pilgrim Nuclear Power Station SEP-PNPS-IST-002 Rev.: 2 Page 40 of 174 6.2 VALVE PROGRAM TABLE This table identifies the scope of valves within the IST Program and allows cross-referencing specific valve test requirements to their implementing Station Procedure. The table sequence is by ascending P&ID number. (Compressed Air System is listed on P&IDs M220, M227, and M252.)

Newly incorporated component/test requirements will be identified by an asterisk (*) next to the implementing Procedure. When using (*) Procedures for postmaintenance testing, the current, approved Procedure should be reviewed for applicability (i.e., is the new test requirement or component incorporated).

[1] Valve Test Index System P&ID Revision Page Salt Service Water (29) M212 Sh. 1 E93 42 Reactor Building Closed Cooling Water (30) M215 Sh. 1 E52 44 M215 Sh. 2 E50 M215 Sh. 3 E41 M215 Sh. 4 E46 Compressed Air (31) M220 Sh. 3 E76 46 Containment Atmosphere Control M227 Sh. 1 E59 47 (45 & 9) M227 Sh. 2 E49 Compressed Air (31) M227 Sh. 1 E59 53 Nitrogen Supply (9) M227 Sh. 2 E49 54 Radwaste Collection (20) M232 E38 55 Post-Accident Sampling & Hydrogen M239 Sh. 1 E28 56 and Oxygen Analyzer System (5065) M239 Sh. 2 E20 M239 Sh. 4 E27 M239 Sh. 5 E26 Residual Heat Removal System (1001) M241 Sh. 1 E87 62 M241 Sh. 2 E47 Core Spray System (1400) M242 E53 67 High Pressure Coolant Injection M243 E53 70 (2301 & 23) M244 Sh. 1 E31 Reactor Core Cooling System (1301) M245 E35 74 M246 Sh. 1 E32 Reactor Water Cleanup System (1201 & 12) M247 E54 77 Standby Liquid Control System (1101) M249 E29 78 Control Rod Drive Hydraulic M250 Sh. 1 E73 79 System (302) M250 Sh. 2 E17 Recirc Pump Instrumentation (262) M251 Sh. 1 E22 81 M251 Sh. 2 E23

Pilgrim Nuclear Power Station SEP-PNPS-IST-002 Rev.: 2 Page 41 of 174

[1] Valve Test Index (Continued)

System P&ID Revision Page Feedwater System (6) M252 Sh. 2 E68 82 Nitrogen Supply (9) M252 Sh. 1 E68 83 Reactor Recirculation System (202) M252 Sh. 2 E68 84 Main Steam Isol., ADS, & Safety Relief (203) M252 Sh. 1 E68 85 Nuclear Boiler System (220) M252 Sh. 1 E68 87 M252 Sh. 2 E68 Nuclear Boiler Instrumentation (261) M252 Sh. 2 E68 88 Nuclear Boiler Vessel M253 Sh. 1 E44 90 Instrumentation (263) M253 Sh. 2 E29 Traversing In-core Probe (45) M1Q-1-5 E2 93 NOTE: The drawing revision level will require changing only if the revision affects information within the valve tables or a formal review of the program is performed against a later drawing revision.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 42 of 174 P&ID: M212 (SHEET 1) SYSTEM: SALT SERVICE WATER SYSTEM (29)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 3800 G-8 3 B 18 BF MO 0 FE Q 8.5.3.11 ST Q 8.5.3.11 0 PI 2Y 8.5.3.11 3801 G-6 3 B 12 BF MO 0 PE Q 8.5.3.11 FE CS 8.1.11.14 CS-04 ST CS 8.1.11.14 C CS-04 PI 2Y 8.1.11.14 3805 G-5 3 B 12 BF MO 0 PE Q 8.5.3.11 FE CS 8.1.11.14 CS-04 ST CS 8.1.11.14 C CS-04 P1 2Y 8.1.11.14 3806 G-4 3 B 18 BF MO 0 FE Q 8.5.3.11 ST Q 8.5.3.11 0 PI 2Y 8.5.3.11 3808 C-6 3 B 12 BF MO 0 FE Q 8.5.3.11 ST Q 8.5.3.11 O/C PI 2Y 8.5.3.11 3813 C-5 3 B 12 BF MO 0 FE Q 8.5.3.11 ST Q 8.5.3.11 O/C PI 2Y 8.5.3.11 3820 E-6 2 B 12 GA MA C FE NA NA C **Passive (10) LSBI CBT 8.5.2.11 (Manual) 3823 F-7 3 B 18 BF MA C FE 2Y* 2.2.32 C 3824 G-8 3 B 18 BF MA C FE 2Y* 2.2.32 C 3827 F-6 3 B 12 BF MA 0 FE 2Y* 2.2.32 0 3828 F-6 3 B 12 BF MA C FE 2Y* 2.2.32 C 3829 F-7 3 B 12 BF MA C FE 2Y* 2.2.32 C 3832 F-5 3 B 12 BF MA 0 FE 2Y* 2.2.32 0 3833 F-4 3 B 12 BF MA C FE 2Y* 2.2.32 C 3834 F-5 3 B 12 BF MA C FE 2Y* 2.2.32 C 3837 F-4 3 B 18 BF MA 0 FE 2Y* 2.2.32 0 3838 F-3 3 B 18 BF MA C FE 2Y* 2.2.32 C

These manual valves will be administratively full stroke exercised once per year (1Y) due to system service conditions.

- 10-HO-3820 is a manual passive system code boundary valve. Manual passive LSBI valves will be tested on a conditional basis. LSBI testing will be performed only when these valves have been manipulated (initial valve status has been altered).

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 43 of 174 P&ID: M212 (SHEET 1) SYSTEM: SALT SERVICE WATER SYSTEM (29)

Valve P&ID 1ST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 3839 F-4 3 B 18 BF MA C FE 2Y* 2.2.32 C 3842 F-8 3 B 18 BF MA 0 FE 2Y* 2.2.32 0 3880A B-7 3 C 12 CK SA 0 OT Q 8.5.3.2.1 0 CT Q 8.5.3.2.1 C 3880B B-6 3 C 12 CK SA 0 OT Q 8.5.3.2.1 0 CT Q 8.5.3.2.1 C 3880C B-5 3 C 12 CK SA 0 OT Q 8.5.3.2.1 0 CT Q 8.5.3.2.1 C 3880D B-3 3 C 12 CK SA 0 OT Q 8.5.3.2.1 0 CT Q 8.5.3.2.1 C 3880E B-4 3 C 12 CK SA 0 OT Q 8.5.3.2.1 0 CT Q 8.5.3.2.1 C 3915 D-3 3 B 6 BF AO C FE Q 8.5.3.11 ST Q 8.5.3.11 FS Q 8.5.3.11 C 3925 D-3 3 B 6 BF AO C FE Q 8.5.3.11 ST Q 8.5.3.11 FS Q 8.5.3.11 C AV-38003 B-8 3 C 2 CK SA C OT Q 8.5.3.2.1 CT Q 8.5.3.2.1 C AV-38004 B-7 3 C 2 CK SA C OT Q 8.5.3.2.1 CT Q 8.5.3.2.1 C AV-38005 B-6 3 C 2 CK SA C OT Q 8.5.3.2.1 CT Q 8.5.3.2.1 C AV-38006 B-4 3 C 2 CK SA C OT Q 8.5.3.2.1 CT Q 8.5.3.2.1 C AV-38007 B-5 3 C 2 CK SA C OT Q 8.5.3.2.1 CT Q 8.5.3.2.1 C

These manual valves will be administratively full stroke exercised once per year (1Y) due to system service conditions.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 44 of 174 P&ID: M215 (SHEETS 1,2, 3, & 4) SYSTEM: REACTOR BUILDING CLOSED COOLING WATER (30)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 419 SH.1/F-5 3 C 8 CK SA 0 OT Q 8.5.3.1 0 CT Q 8.5.3.1 C 420 SH.1/E-5 3 C 8 CK SA 0 OT Q 8.5.3.1 0 CT Q 8.5.3.1 C 421 SH.1/D-5 3 C 8 CK SA 0 OT Q 8.5.3.1 0 CT Q 8.5.3.1 C 422 SH.2/G-4 3 C 8 CK SA 0 OT Q 8.5.3.1 0 CT Q 8.5.3.1 C 423 SH.2/F-4 3 C 8 CK SA 0 OT Q 8.5.3.1 0 CT Q 8.5.3.1 C 424 SH.2/E-4 3 C 8 CK SA 0 OT Q 8.5.3.1 0 CT Q 8.5.3.1 C 432 SH.3/E-8 2 AC 6 CK SA 0 OT RI# 8.1.31 CT RI 8.5.3.15 C RJ-01 8.7.1.5 LJ OBJ 8.7.1.5 4002 SH.3/E-4 2 A 6 GA MO 0 FE RI 8.1.11.13 RJ-18 ST RI 8.1.11.13 C RJ-18 PI 2Y 8.1.11.13 8.1.30 LJ OBJ 8.7.1.5 4009A SH.2/B-5 3 B 8 GA MO 0 FE RI 8.1.11.13 RJ-18 ST RI 8.1.11.13 C RJ-18 PI 2Y 8.1.11.13 4009B SH.3/E-2 3 B 8 GA MO 0 FE CS 8.1.11.13 RJ-18 ST CS 8.1.11.13 C RJ-18 PI 2Y 8.1.11.13 4010A SH.2/G-6 3 B 12 GA MO C FE Q 8.5.3.10 ST Q 8.5.3.10 0 P1 2Y 8.5.3.10 4010B SH.2/H-6 3 B 12 GA MO C FE Q 8.5.3.10 ST Q 8.5.3.10 0 PI 2Y 8.5.3.10 4020 SH.4/D-7 3 C 0.75 RL SA C RT 10Y 8.1.26.3 Thermal Relief 4031 SH.4/D-6 3 C 3 RL SA C RT 10Y 8.1.26.3 Skid-Mounted 4032 SH.2/H-7 3 C 3 RL SA C RT 10Y 8.1.26.3 Skid-Mounted

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 45 of 174 P&ID: M215 (SHEETS 1,2, 3, & 4) SYSTEM: REACTOR BUILDING CLOSED COOLING WATER (30)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 4033 SH.2/B-7 2 C 1.5 RL SA C RT 10Y 8.1.26.3 Thermal Relief 4036 SH.2/H-7 3 C 0.75 RL SA C RT 10Y 8.1.26.3 Thermal Relief 4043A SH.4/H-7 3 B 1.5 GL AO 0 FE NA NA Passive ST NA NA FS NA NA 4043B SH.4/G-7 3 B 1.5 GL AO 0 FE NA NA Passive ST NA NA FS NA NA 4044A SH.2/D-6 3 B 1.5 GL AO 0 FE NA NA Passive ST NA NA FS NA NA 4044B SH.2/C-6 3 B 1.5 GL AO 0 FE NA NA Passive ST NA NA FS NA NA 4060A SH.4/C-7 3 B 12 GA MO C FE Q 8.5.3.10 ST Q 8.5.3.10 0 PI 2Y 8.5.3.10 4060B SH.4/D-7 3 B 12 GA MO C FE Q 8.5.3.10 ST Q 8.5.3.10 0 PI 2Y 8.5.3.10 4065 SH.4/A-7 3 B 6 GA MO 0 FE Q 8.5.3.10 ST Q 8.5.3.10 C PI 2Y 8.5.3.10 4083 SH.3/G-3 3 B 10 BF MO 0 FE Q 8.5.3.10 ST Q 8.5.3.10 C PI 2Y 8.5.3.10 4084 SH.1/D-7 3 B 10 BF MO 0 FE Q 8.5.3.10 ST Q 8.5.3.10 C PI 2Y 8.5.3.10 4085A SH.1/G-3 3 B 8 GA MO 0 FE RI 8.1.11.13 RJ-20 ST RI 8.1.11.13 C RJ-20 PI - 2Y 8.1.11.13 4085B SH.1/F-4 3 B 8 GA MO 0 FE RI 8.1.11.13 RJ-20 ST RI 8.1.11.13 C RJ-20 PI 2Y 8.1.11.13

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 46 of 174 P&ID: M220 (SHEET 3) SYSTEM: COMPRESSED AIR (31)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 167 SH.3/C-8 2 AC 3 CK SA 0 OT RI# 8.7.4.4 8.7.1.10 CT RI 8.7.1.2.2 C RJ-08 8.7.1.5 LJ OBJ 8.7.1.5

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 47 of 174 P&ID: M227 (SHEETS 1 & 2) SYSTEM: CONTAINMENT ATMOSPHERE CONTROL (45 & 9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 102 SH.1/C-5 2 A 1 GL MA C FE NA NA Passive (Manual)

(45) LJ OBJ 8.7.1.5 103 SH.1/C-5 2 A 1 GL MA C FE NA NA Passive (Manual)

(45) LJ OBJ 8.7.1.5 104 SH.1/C-5 2 A 1 GL MA C FE NA NA Passive (Manual)

(45) LJ OBJ 8.7.1.5 105 SH.1/C-5 2 A 1 GL MA C FE NA NA Passive (Manual)

(45) LJ OBJ 8.7.1.5 106 SH.1/E-4 2 A 4 GA MA C FE NA NA Passive (Manual)

(45) LJ OBJ 8.7.1.5 X-201A SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A. 1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201B SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A. 1 C Pi 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201C SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A.1 C PI .2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201D SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A.1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201E SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A.1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201F SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A. 1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201G SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A.1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201H SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A. 1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 48 of 174 P&ID: M227 (SHEETS 1 & 2) SYSTEM: CONTAINMENT ATMOSPHERE CONTROL (45 & 9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes X-201J SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A.1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-201K SH.1/C-6 2 AC 18 CK SA C OT Q 8.A.1 0 (45) (Testable) CT Q 8.A. 1 C PI 2Y 8.M.2-7.1.9 8.1.30 LX RI 8.A.2 X-212A SH.1/A-7 2 AC 20 CK SA C OT Q 8.7.4.9 0 (45) CT Q 8.7.4.9 C PI 2Y 8.7.4.9 8.1.30 LJ OBJ 8.7.1.5 X-212B SH.1/A-7 2 AC 20 CK SA C OT Q 8.7.4.9 0 (45) CT Q 8.7.4.9 C PI 2Y 8.7.4.9 8.1.30 LJ OBJ 8.7.1.5 5025 SH.1/E-8 2 A 8 BF AO C NA NA NA (De-Energized in (45) PI 2Y 8.7.4.12 Closed position)

LJ OBJ 8.7.1.5 8.1.30 5033A SH.2/D-6 2 A 1 GL AO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 C FS Q 8.7.4.2 C P1 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5033B SH.2/C-6 2 A 4 GA AO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5033C SH.2/D-6 2 A 1 GL AO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 49 of 174 P&ID: M227 (SHEETS 1 & 2) SYSTEM: CONTAINMENT ATMOSPHERE CONTROL (45 & 9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 5035A SH.1/D-3 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5035B SH.1/D-2 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5036A SH.1/C-3 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5036B SH.1/C-2 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5040A SH.1/B-7 2 A 20 BF AO C FE Q 8.7.4.9 (45) ST Q 8.7.4.9 O/C FS Q 8.7.4.9 0 PI 2Y 8.7.4.9 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 5040B SH.1/B-7 2 A 20 BF AO C FE Q 8.7.4.9 (45) ST Q 8.7.4.9 O/C FS Q 8.7.4.9 0 PI 2Y 8.7.4.9 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 5041A SH.1/C-8 2 A 2 GL AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5041B SH.1/C-7 2 A 2 GL AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 50 of 174 P&ID: M227 (SHEETS 1 & 2) SYSTEM: CONTAINMENT ATMOSPHERE CONTROL (45 & 9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 5042A SH.1/D-8 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5042B SH.1/D-7 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5043A SH.1/E-7 2 A 2 GL AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5043B SH.1/E-6 2 A 2 GL AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5044A SH.1/G-7 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5044B SH.1/G-6 2 A 8 BF AO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 C FS Q 8.7.4.2 C PI 2Y 8.7.4.2 8.1.30 LJ OBJ 8.7.1.5 5081A SH.1/G-5 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5081B SH.1/G-5 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 51 of 174 P&ID: M227 (SHEETS 1 & 2) SYSTEM: CONTAINMENT ATMOSPHERE CONTROL (45 & 9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 5082A SH.1/G-5 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5082B SH.1/G-5 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5083A SH.1/H-8 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5083B SH.1/G-8 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5084A SH.1/H-8 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5084B SH.1/G-8 2 A 1 GA SO C FE Q 8.7.4.2 (45) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5085A SH.2/E-7 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5085B SH.2/D-7 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 0/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 52 of 174 P&ID: M227 (SHEETS 1 & 2) SYSTEM: CONTAINMENT ATMOSPHERE CONTROL (45 & 9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 5086A SH.2/E-6 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5086B SH.2/D-6 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5087A SH.2/E-7 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5087B SH.2/D-7 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5088A SH.2/E-7 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5 5088B SH.2/D-7 2 A 1 GA SO C FE Q 8.7.4.2 (9) ST Q 8.7.4.2 O/C FS Q 8.7.4.2 C PI 2Y 8.7.4 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 53 of 174 P&ID: M227 (SHEET 1) SYSTEM: COMPRESSED AIR (31)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 434 SH.1/B-6 2 AC 1 CK SA C OT RI# 8.A.1 CT RI 8.7.1.2.2 C RJ-08 8.7.1.5 LJ OBJ 8.7.1.5 5046 SH.1/B-6 2 A 1 GL AO C FE/FS NA NA Exempt (Test)

ST NA NA Exempt (Test)

LJ OBJ 8.7.1.5

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 54 of 174 P&ID: M227 (SHEET 2) SYSTEM: NITROGEN SUPPLY (9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 340 SH.2/C-6 2 AC 1 CK SA C OT RI# 8.A.1 CT RI 8.7.1.5 C RJ-08 3.M.4-112 LJ OBJ 8.7.1.5 341 SH.2/C-6 2 AC 1 CK SA C OT RI# 8.A.1 CT RI 8.7.1.5 C RJ-08 3.M.4-112 LJ OBJ 8.7.1.5

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 55 of 174 P&ID: M232 SYSTEM: RADWASTE COLLCTTION (20)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 7011A G-5 2 A 2 BL AO 0 FE Q 8.7.4.3 ST Q 8.7.4.3 C FS Q 8.7.4.3 C P1 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 7011B G-5 2 A 2 BL AO 0 FE Q 8.7.4.3 ST Q 8.7.4.3 C FS Q 8.7.4.3 C PI 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 7017A D-5 2 A 2 BL AO 0 FE Q 8.7.4.3 ST Q 8.7.4.3 C FS Q 8.7.4.3 C P1 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 7017B D-5 2 A 2 BL AO 0 FE Q 8.7.4.3 ST Q 8.7.4.3 C FS Q 8.7.4.3 C PI 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 56 of 174 P&ID: M239 (SHEETS 1,2, 4, & 5) SYSTEM: POST-ACCIDENT SAMPLING & H2 & 02 ANALYZER SYSTEM (5065)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 11A SH.1/C-6 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 13B SH.1/D-3 2 A 1 GL so 0 FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 14A SH.1/D-6 2 A 1 GL so 0 FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 15B SH.1/C-3 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 18A SH.1/C-6 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 20B SH.1/D-4 2 A 1 GL so 0 FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 21A SH.1/D-6 2 A 1 GL so 0 FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 22B SH.1/C-3 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 57 of 174 P&ID: M239 (SHEETS 1,2, 4, & 5) SYSTEM: POST-ACCIDENT SAMPLING & H2 & 02 ANALYZER SYSTEM (5065)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 24A SH.1/D-6 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 25B SH.1/D-4 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 26A SH.1/D-6 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 27B SH.1/D-3 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 31B SH.1/E-3 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.5 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 33A SH.1/E-6 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 35B SH.1/E-4 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.5 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 37A SH.1/E-5 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 iST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 58 of 174 P&ID: M239 (SHEETS 1, 2, 4, & 5) SYSTEM: POST-ACCIDENT SAMPLING & H2 & 02 ANALYZER SYSTEM (5065)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 63 SH.2/H-6 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 64 SH.2/H-5 2 A 1 GL SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 65 SH.2/G-6 2 B 0.50 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 66 SH.2/G-5 2 B 0.50 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 67 SH.2/D-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 68 SH.2/D-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 69 SH.2/C-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 70 SH.2/C-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 71 SH.2/E-6 2 A 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.6 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 59 of 174 P&ID: M239 (SHEETS 1,2, 4, & 5) SYSTEM: POST-ACCIDENT SAMPLING & H2 & 02 ANALYZER SYSTEM (5065)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 72 SH.2/E-5 2 A 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.6 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 73 SH.2/A-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 74 SH.2/A-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 75 SH.2/A-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 76 SH.2/A-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 77 SH.2/E-6 2 A 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.6 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 78 SH.2/E-5 2 A 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.6 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 79 SH.2/C-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4 80 SH.2/C-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.4

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 60 of 174 P&ID: M239 (SHEETS 1, 2, 4, & 5) SYSTEM: POST-ACCIDENT SAMPLING & H2 & 02 ANALYZER SYSTEM (5065)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 81 SH.2/B-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C P1 2Y 8.7.4.8.3 82 SH.2/B-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.3 83 SH.2/F-6 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 84 SH.2/F-5 2 B 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 85 SH.2/G-6 2 A 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 LJ OBJ 8.7.1.5 86 SH.2/G-5 2 A 1 GA SO C FE Q 8.7.4.1 ST Q 8.7.4.1 O/C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.2 LJ OBJ 8.7.1.5 87 SH.1/F-4 2 B 1 GL AO 0 FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 88 SH.1/F-4 2 B 1 GL AO 0 FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 89 SH.1/F-6 2 B 1 GL AO 0 FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 90 SH.1/F-6 2 B 1 GL AO 0 FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 61 of 174 P&ID: M239 (SHEETS 1, 2, 4, & 5) SYSTEM: POST-ACCIDENT SAMPLING & H2 & 02 ANALYZER SYSTEM (5065)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 91 SH.1/B-5 2 A 1 GL AO 0 FE Q 8.7.4.1 ST a 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 LJ OBJ 8.7.1.5 92 SH.1/B-4 2 A 1 GL AO 0 FE Q 8.7.4.1 ST Q 8.7.4.1 C FS Q 8.7.4.1 C PI 2Y 8.7.4.8.1 8.1.30 LJ OBJ 8.7.1.5 122A SH.1/E-8 2 B 0.375 GA SV C FE/FS Q 8.7.4.1 O/C Skid-Mounted ST Q 8.7.4.1 PI 2Y 8.7.4.8.5 122B SH.1/E-1 2 B 0.375 GA SV C FE/FS Q 8.7.4.1 O/C Skid-Mounted ST Q 8.7.4.1 PI 2Y 8.7.4.8.5 123A SH.1/E-8 2 B 0.375 GA SV C FE/FS Q 8.7.4.1 O/C Skid-Mounted ST Q 8.7.4.1 PI 2Y 8.7.4.8.5 123B SH.1/E-1 2 B 0.375 GA SV C FE/FS Q 8.7.4.1 O/C Skid-Mounted ST Q 8.7.4.1 PI 2Y 8.7.4.8.5 124A SH.1/E-7 2 B 0.375 GA SV C FE/FS Q 8.7.4.1 O/C Skid-Mounted ST Q 8.7.4.1 PI 2Y 8.7.4.8.5 124B SH.1/E-2 2 B 0.375 GA SV C FE/FS Q 8.7.4.1 O/C Skid-Mounted ST Q 8.7.4.1 PI 2Y 8.7.4.8.5 5117A SH.4/G-4 2 B 0.375 GA SV 0 FE/FS Q 8.M.3-13/14 C Skid-Mounted ST Q 8.M.3-13/14 0 5117B SH.5/G-5 2 B 0.375 GA SV 0 FE/FS Q 8.M.3-13/14 C Skid-Mounted ST Q 8.M.3-13/14 0 5137A SH.4/F-4 2 B 0.375 GA SV 0 FE/FS Q 8.M.3-13/14 C Skid-Mounted ST Q 8.M.3-13/14 0 5137B SH.5/F-5 2 B 0.375 GA SV 0 FE/FS Q 8.M.3-13/14 C Skid-Mounted ST Q 8.M.3-13/14 0

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 62 of 174 P&ID: M241 (SHEETS 1 & 2) SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM (1001 & 10)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 2A SH.2/D-6 2 C 2 CK SA C OT CMP-01 8.5.2.2.1 0 NIT/3.M.4-112.

CT CMP-01 8.5.2.2.1 C 2B SH.2/D-3 2 C 2 CK SA C OT CMP-01 8.5.2.2.2 0 NIT/3.M.4-112*

CT CMP-01 8.5.2.2.2 C 2C SH.2/G-6 2 C 2 CK SA C OT CMP-01 8.5.2.2.1 0 NIT/3.M.4-112*

CT CMP-01 8.5.2.2.1 C 2D SH.2/G-3 2 C 2 CK SA C OT CMP-01 8.5.2.2.2 0 NIT/3.M.4-112*

CT CMP-01 8.5.2.2.2 C 7A SH.2/D-5 2 A 18 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C P1 2Y 8.5.2.3 LJW OBJ 8.7.1.2 7B SH.2/D-4 2 A 18 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 LJW OBJ 8.7.1.2 7C SH.2/F-5 2 A 18 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C P1 2Y 8.5.2.3 LJW OBJ 8.7.1.2 7D SH.2/F-4 2 A 18 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 LJW OBJ 8.7.1.2 16A SH.2/F-7 2 B 18 GL MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 16B SH.2/E-3 2 B 18 GL MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3

For CK-1001-2A, 2B, 2C, and 2D, performance of the pump biennial comprehensive test (8.5.2.2.1 / 8.5.2.2.2, as applicable) is the required postmaintenance test (PMT) following check valve disassembly. The non-intrusive testing (NIT) is performed periodically, as a supplemental condition monitoring performance test.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 63 of 174 P&ID: M241 (SHEETS 1 & 2) SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM (1001)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 18A SH.2/G-6 2 A 3 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 LJW OBJ 8.7.1.2 18B SH.2/G-4 2 A 3 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 LJW OBJ 8.7.1.2 19 SH.1/B-7 2 B 18 GA MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 C PI 2Y 8.5.2.3 21 SH.1/B-3 2 B 4 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 C PI 2Y 8.5.2.11 8.1.30 LSBI PBT 8.5.2.11 VTP-01 22A SH.1/G-8 2 AC 1 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 22B SH.1/F-2 2 AC 1 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 23A SH.1/G-6 2 A 10 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 23B SH.1/G-4 2 A 10 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 26A SH.1/G-5 2 A 10 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 26B SH.1/G-4 2 A 10 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 28A SH.1/F-6 2 A 18 GL MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 28B SH.1/F-3 2 A 18 GL MO 0 FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 64 of 174 P&ID: M241 (SHEETS 1 & 2) SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM (1001)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 29A SH.1/E-6 1 A 18 GA MO C FE CS 8.1.11.3 CS-03 ST CS 8.1.11.3 O/C CS-03 PI 2Y 8.5.2.7 8.1.30 LJ OBJ 8.7.1.5 LP 2Y 8.5.2.7 29B SH.1/F-3 1 A 18 GA MO C FE CS 8.1.11.4 CS-03 ST CS 8.1.11.4 O/C CS-03 PI 2Y 8.5.2.7 8.1.30 LJ OBJ 8.7.1.5 LP 2Y 8.5.2.7 32 SH.1/B-3 2 B 4 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 C PI 2Y 8.5.2.11 8.1.30 LSBI OBJ 8.5.2.11 VTP-01 33A SH.1/E-6 1 NA 18 GA MA LO FE NA NA 0 Passive (Manual)

P1 2Y 8.5.2.7 8.1.30 33B SH.1/E-4 1 NA 18 GA MA LO FE NA NA 0 Passive (Manual)

P1 2Y 8.5.2.7 8.1.30 34A SH.1/F-7 2 A 12 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 34B SH.1/F-3 2 A 12 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 36A SH.1/E-7 2 B 12 GL MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 36B SH.1/E-3 2 B 12 GL MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 37A SH.1/E-7 2 A 6 GL MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5 37B SH.1/E-3 2 A 6 GL MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 8.1.30 LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 65 of 174 P&ID: M241 (SHEETS 1 & 2) SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM (1001)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 43A SH.2/E-5 2 B 18 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 43B SH.2/E-4 2 B 18 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C P1 2Y 8.5.2.3 43C SH.2/G-5 2 B 18 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 43D SH.2/G-4 2 B 18 GA MO C FE Q 8.5.2.3 ST Q 8.5.2.3 O/C PI 2Y 8.5.2.3 44 SH.2/C-5 2 AC 1.5 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 47 SH. 1/C-5 1 A 20 GA MO C FE CS 8.1.11.5 CS-01 ST CS 8.1.11.5 C CS-01 PI 2Y 8.5.2.8 8.1.30 LJ OBJ 8.7.1.5 LP PBT 8.5.2.8 50 SH.1/D-5 1 A 20 GA MO C FE CS 8.1.11.5 CS-01 ST CS 8.1.11.5 C CS-01 PI 2Y 8.5.2.8 8.1.30 LJ OBJ 8.7.1.5 LP PBT 8.5.2.8 67A SH.2/D-6 2 C 12 CK SA C OT Q 8.5.2.2.1 0 CT Q 8.5.2.2.1 C 67B SH.2/D-3 2 C 12 CK SA C OT Q 8.5.2.2.2 0 CT Q 8.5.2.2.2 C 67C SH.2/F-7 2 C 12 CK SA C OT Q 8.5.2.2.1 0 CT Q 8.5.2.2.1 C 67D SH.2/F-3 2 C 12 CK SA C OT Q 8.5.2.2.2 0 CT Q 8.5.2.2.2 C

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 66 of 174 P&ID: M241 (SHEETS 1 & 2) SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM (1001)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 68A SH.1/E-6 1 AC 18 CK SA C OT RI 8.1.34 0 RJ-23 CT RI 8.5.2.7 C RJ-21 LP 2Y 8.5.2.7 68B SH.1/E-4 1 AC 18 CK SA C OT CMP-02 8.1.34 0 ##RJ-23 NIT CT CMP-02 8.5.2.7 C ## RJ-21 LP 2Y 8.5.2.7 362B SH.2/E-3 2 AC 2 CK SA C OT CMP-03 8.5.2.3 CT CMP-03 8.5.2.13 C ##RJ-21 LX 2Y 8.5.2.13 363A SH.2/G-8 2 AC 2 CK SA C OT CMP-03 8.5.2.3 CT CMP-03 8.5.2.13 C ##RJ-21 LX 2Y 8.5.2.13 515 SH.1/D-3 2 AC 6 CK SA C CT CMP-13 8.7.1.2 C Passive (Maint.)

(10) LJW 8.7.1.2 511 SH.1/D-3 2 B 8 GA HO LC FE NA NA **Passive (10) LSBI CBT 8.5.2.11 (Manual) 8004 SH.2/F-6 2 AC 1.5 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 8005 SH.2/D-6 2 AC 1.5 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 8006 SH.2/D-4 2 AC 1.5 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 8007 SH.2/G-4 2 AC 1.5 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 8008 SH.2/D-7 2 AC 1 RL SA C RT 10Y 8.1.26.3 Thermal Relief LX NA 8.1.26.3 RT Verifies 8009 SH.2/D-3 2 AC 1 RL SA C RT 10Y 8.1.26.3 Thermal Relief LX NA 8.1.26.3 RT Verifies

1. CK-1001-68B, CK-1001-362B, and CK-1001-363A are included in the PNPS Condition Monitoring Program. At the present time this refuel outage justification is not necessary but has been listed as a place keeper for reference purposes.

10-HO-511 is a manual passive system code boundary valve. Manual passive LSBI valves will be tested on a conditional basis. LSBI testing will be performed only when these valves have been manipulated (initial valve status has been altered).

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 67 of 174 P&ID: M242 SYSTEM: CORE SPRAY SYSTEM (1400)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 2A C-3 2 B 12 GA HO C FE NA NA Passive (Manual)

LSBI PBT 8.5.1.8 C VTP-01 2B B-2 2 B 12 GA HO C FE NA NA Passive (Manual)

LSBI PBT 8.5.1.8 C VTP-01 3A C-5 2 A 18 GA MO 0 FE Q 8.5.1.3 ST Q 8.5.1.3 C P1 2Y 8.5.1.3 LJW OBJ 8.7.1.2 3B B-4 2 A 18 GA MO 0 FE Q 8.5.1.3 ST Q 8.5.1.3 C PI 2Y 8.5.1.3 LJW OBJ 8.7.1.2 4A E-5 2 B 6 GA MO C FE Q 8.5.1.1 ST Q 8.5.1.1 C PI 2Y 8.5.1.1 8.1.30 4B F-4 2 B 6 GA MO C FE Q 8.5.1.1 ST Q 8.5.1.1 C PI 2Y 8.5.1.1 8.1.30 6A F-7 1 NA 10 GA MA LO FE NA NA 0 Passive (Manual)

PI 2Y 8.5.1.6 8.1.30 6B G-7 1 NA 10 GA MA LO FE NA NA 0 Passive (Manual)

PI 2Y 8.5.1.6 8.1.30 9A E-7 1 AC 10 CK SA C OT CMP-04 2.2.20 0 CT CMP-04 8.5.1.6 C LP 2Y 8.5.1.6 9B G-6 1 AC 10 CK SA C OT CMP-04 2.2.20 0 CT CMP-04 8.5.1.6 C LP 2Y 8.5.1.6 13A D-5 2 C 3 SC SA **LO/C OT CMP-05 3.M.4-112 0 NIT CT CMP-05 3.M.4-112 C NIT 13B D-3 2 C 3 SC SA **LO/C OT CMP-05 3.M.4-112 0 NIT CT CMP-05 3.M.4-112 C NIT

- Stop Check Valve: Handwheel Position - Locked Open, Normal Disc Position - Closed

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 68 of 174 P&ID: M242 SYSTEM: CORE SPRAY SYSTEM (1400)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 24A E-5 2 A 10 GA MO 0 FE Q 8.5.1.3 ST Q 8.5.1.3 O/C P1 2Y 8.5.1.3 8.1.30 LJ OBJ 8.7.1.5 24B G-4 2 A 10 GA MO 0 FE Q 8.5.1.3 ST Q 8.5.1.3 O/C PI 2Y 8.5.1.3 8.1.30 LJ OBJ 8.7.1.5 25A E-6 1 A 10 GA MO C FE CS 8.1.11.11 CS-03 ST CS 8.1.11.11 O/C CS-03 PI 2Y 8.5.1.6 8.1.30 LJ OBJ 8.7.1.5 LP 2Y -8.5.1.6 25B G-5 1 A 10 GA MO C FE CS 8.1.11.11 CS-03 ST CS 8.1.11.11 O/C CS-03 PI 2Y 8.5.1.6 8.1.30 LJ OBJ 8.7.1.5 LP 2Y 8.5.1.6 28A D-5 2 AC 2 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 28B F-3 2 AC 2 RL SA C RT 10Y 8.1.26.3 LX NA 8.1.26.3 RT Verifies 31A F-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 31B F-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 35 C-7 2 AC 6 CK SA C OT CMP-12 8.5.1.1 CT CMP-12 8.7.1.2 C 3.M.4-112

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 69 of 174 P&ID: M242 SYSTEM: CORE SPRAY SYSTEM (1400)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 36A D-5 2 C 10 CK SA C OT CMP-06 2.2.20 0 CT CMP-06 8.5.1.1 C 36B D-3 2 C 10 CK SA C OT CMP-06 2.2.20 0 CT CMP-06 8.5.1.1 C 212A E-5 2 AC 2 CK SA C OT CMP-03 8.5.2.15 CT CMP-03 8.5.1.7 C ## RJ-21 LX 2Y 8.5.1.7 212B G-3 2 AC 2 CK SA C OT CMP-03 8.5.2.15 CT CMP-03 8.5.1.7 C ## RJ-21 LX 2Y 8.5.1.7 214 C-7 2 AC 6 CK SA C OT CMP-12 8.5.1.1 CT CMP-12 8.7.1.2 C 3.M.4-112

1. CK-1400-212A and CK-1400-212B are included in the PNPS Condition Monitoring Program. At the present time this refuel outage justification is not necessary but has been listed as a place keeper for reference purposes.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 70 of 174 P&ID: M243 (SHEET 1) & M244 (SHEET 1) SYSTEM: HIGH PRESSURE COOLANT INJECTION (2301)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 2300-23 SH.1/F-3 2 B 10 GA HO C NA Q 8.5.4.4 O/C STP-03 Skid-Mounted (HO-1) (M244) P1 2Y 8.5.4.4 Stop Valve 2301-24 SH.1/F-3 2 NA 10 GV HO C NA Q 8.5.4.1/8.5.4.4 STP-03 Skid-Mounted (HO-2) (M244) P1 2Y 8.5.4.4 GV - Exempt 3 SH. 1/E-2 2 B 10 GA MO C FE Q 8.5.4.1 8.5.4.4 (M243) ST Q 8.5.4.1 0 8.5.4.4 PI 2Y 8.5.4.4 4 SH.1/G-6 1 A 10 GA MO 0 FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 O/C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 5 SH.1/G-5 1 A 10 GA MO 0 FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 O/C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 6 SH.1/G-3 2 B 16 GA MO 0 FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 O/C PI 2Y 8.5.4.4 7 SH.1/E-6 1 AC 14 CK SA C OT RI 8.1.11.7 0 RJ-24 (M243) CT RI 8.1.11.7 C RJ-24 LP 2Y 8.5.4.8 8 SH.1/E-5 1 A 14 GA MO C FE CS 8.1.11.11 CS-03 (M243) ST CS 8.1.11.11 O/C CS-03 P1 2Y 8.5.4.8 8.1.30 LJ OBJ 8.7.1.5 LP 2Y 8.5.4.8 9 SH.1/E-5 2 B 14 GA MO 0 P1 2Y 8.1.11.11 None Passive, 8.1.30 (M243) 10 SH.1/F-5 2 B 10 GL MO C FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 C PI 2Y 8.5.4.4 14 SH.1/E-4 2 B 4 GL MO C FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 O/C PI 2Y 8.5.4.4

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 71 of 174 P&ID: M243 (SHEET 1) & M244 (SHEET 1) SYSTEM: HIGH PRESSURE COOLANT INJECTION (2301)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 15 SH.1/G-4 NQ B 10 GA MO C FE** Q 8.5.4.4 None ST** Q 8.5.4.4 PI 2Y 8.5.4.4 20 SH.1/G-3 2 C 16 CK SA C OT CMP-10 8.5.4.1 0 (M243) CT CMP-10 8.5.4.11 C 3.M.4-112 23 SH.1/F-5 2 AC 1 RL SA C RT 10Y 8.1.26.3 (M244) LX NA 8.1.26.3 RT Verifies 26 SH.1/F-6 2 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 (M243) CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 29 SH.1/C-1 2 B 1 GL AO 0 FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 C FS Q 8.5.4.4 C PI 2Y 8.5.4.4 31 SH.1/D-1 2 B 1 GL AO C NA NA NA Skid-Mounted (M243) LCV - Exempt***

32 SH.1/C-3 2 B 1 GL SO C NA NA NA Skid-Mounted (M244) LCV - Exempt***

33 SH.1/A-8 2 A 4 GA MO 0 FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 34 SH.1/B-7 2 A 4 GA MO 0 FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 34 SH.1/C-6 2 C 2 CK SA C STP-03 Skid-Mounted (M243) OT RI 8.5.4.16 0 RJ-09 CT RI 8.7.1.2 C RJ-08 35 SH.1/F-3 2 B 16 GA MO C FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 O/C PI 2Y 8.5.4.4

LV-2301-31 and CV-2301-32 are Exempt level control valves (LCV) that function when the HPCI Pump/Turbine is in the "standby" condition, and have no safety stroke function for open or close operation.

MO-2301-15 is classified as non-quality class (NQ) and is not safety-related. This valve has no safety stroke direction, but is augmented because it receives an auto-isolation signal during HPCI initiation-

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 72 of 174 P&ID: M243 (SHEET 1) & M244 (SHEET 1) SYSTEM: HIGH PRESSURE COOLANT INJECTION (2301)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 36 SH.1/A-6 2 A 16 GA MO C FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 O/C PI 2Y 8.5.4.4 LJW OBJ 8.7.1.2 39 SH.1/A-5 2 C 16 CK SA C OT CMP-09 3.M.4-53 0 Disassemble (M243) CT CMP-09 3.M.4-53 C Disassemble 40 SH.1/C-5 2 AC 4 CK SA C OT CMP-08 3.M.4-112 0 NIT (M243) CT CMP-08 8.7.1.2 C 3.M.4-112 LJW OBJ 8.7.1.2 45 SH.1/C-6 2 AC 20 CK SA C OT RI# 8.5.4.1 0 (M243) CT RI 8.7.1.5 C RJ-08 LJ OBJ 8.7.1.5 46 SH.1/B-5 2 C 2 PCV SA 0 NA NA NA STP-03 Skid-Mounted (M244) PCV-Excluded 53 SH.1/B-4 2 AC 1 RL SA C RT 10Y 8.1.26.3 (M244) LX NA 8.1.26.3 RT Verifies 64 SH.1/A-5 2 B 1 GA AO C STP-03 Skid-Mounted (M244) FE/ST RI 8.5.4.1 & 4 C RJ-07 8.1.11.23##

FS RI 8.5.4.1 & 4 C RJ-07 8.1.11.23##

PI 2Y 8.5.4.4 8.1.30 68 SH.1/C-3 2 D 16 RD SA C RD 5Y 8.1.26.5 Non-Testable (M243) Replace 74 SH.1/C-6 2 AC 20 SC SA **LO/C OT RI# 8.5.4.1 0 (M243) CT RI 8.7.1.5 C RJ-08 LJ OBJ 8.7.1.5 75 SH.1/B-5 2 C 4 CK SA C OT Q 8.5.4.1 0 STP-03 Skid-Mounted (M244) 76 SH.1/B-5 2 C 2 CK SA C OT Q 8.5.4.1 0 STP-03 Skid-Mounted (M244) CT RI 8.1.31% C 217 SH.1/C-6 2 AC 2 CK SA C STP-03 Skid-Mounted (M243) OT RI 8.5.4.16 0 RJ-09 CT RI 8.7.1.2 C RJ-08 LJW OBJ 8.7.1.2 218 SH.1/D-6 2 AC 1 CK SA C OT CMP-15 8.5.4.13 0 (M243) CT CMP-15 8.7.1.5 C LJ OBJ 8.7.1.5

- Stop Check Valve: Handwheel Position - Locked Open, Normal Disc Position - Closed

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

1. PNPS 8.1.11.23 installs electrical jumpers to allow valve stroking during any system conditions (normal or low-low water level in HPCI GSC).

% PNPS 8.1.31 performs a supplemental skid check valve functional close test (CT) on 2301-76. This Procedure performs the postmaintenance close test for 2301-76.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 73 of 174 P&ID: M243 (SHEET 1) SYSTEM: HIGH PRESSURE COOLANT INJECTION (2301 & 23)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 220 SH.1/F-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 (M243) CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 232 SH.1/B-8 2 C 3 CK SA C OT RI 8.5.4.15 0 RJ-05 (23) (M243) CT RI 8.5.4.15 C RJ-05 233 SH.1/A-8 2 C 3 CK SA C OT RI 8.5.4.15 0 RJ-05 (23) (M243) CT RI 8.5.4.15 C RJ-05 9068A SH.1/C-7 2 A 1 GL SO C FE/ST Q 8.5.4.4 STP-03 Skid-Mounted (M243) FS Q 8.5.4.4 C PI 4Y 8.7.1.5 C 8.1.30 LJ OBJ 8.7.1.5 9068B SH.1/C-7 2 A 1 GL SO C FE/ST Q 8.5.4.4 STP-03 Skid-Mounted (M243) FS Q 8.5.4.4 C PI 4Y 8.7.1.5 C 8.1.30 LJ OBJ 8.7.1.5 9312 SH.1/C-2 2 B 1 GL AO C FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 C FS Q 8.5.4.4 C PI 2Y 8.5.4.4 9313 SH.1/C-3 2 B 1 GL AO C FE Q 8.5.4.4 (M243) ST Q 8.5.4.4 C FS Q 8.5.4.4 C PI 2Y 8.5.4.4

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 74 of 174 P&ID: M245 (SHEET 1) & M246 (SHEET 1) SYSTEM: REACTOR CORE COOLING SYSTEM (1301)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 1 SH.1/F-3 2 B 2 GA so 0 NA Q 8.5.5.1 C STP-02 Skid-Mounted (246) PI 2Y 8.5.5.1 8.1.30 2 SH.1/F-4 2 B 2 GV HO 0 NA Q 8.5.5.1 STP-02 Skid-Mounted (HYD-1 59) (246) PI 2Y 8.5.5.1 CV-Exempt 8.1.30 9 SH.1/F-5 2 D 8 RD SA C RD 5Y 8.1.26.5 Non-Testable (246) Replace 12 SH.1/B-5 2 B 1 GL AO C FE/ST Q 8.5.5.4 C STP-02 Skid-Mounted (246) FS Q 8.5.5.4 C PI 2Y 8.5.5.4 15A SH.1/G-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 (245) CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 15B SH.1/G-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 (245) CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 16 SH.1/F-6 1 A 3 GA MO 0 FE Q 8.5.5.4 (245) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 17 SH.1/F-5 1 A 3 GA MO 0 FE Q 8.5.5.4 (245) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 8.1.30 & 8.7.1.5 LJ OBJ 8.7.1.5 22 SH.1/G-4 2 B 6 GA MO 0 FE Q 8.5.5.4 (245) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 23 SH.1/G-4 2 C 6 CK SA C OT CMP-10 8.5.5.1 0 (245) CT CMP-10 8.5.5.10 C 3.M.4-112 24 SH.1/B-4 2 C 1 CK SA C OT Q 8.5.5.1 0 STP-02 Skid-Mounted (246) 25 SH.1/A-6 2 A 6 GA MO C FE Q 8.5.5.4 (245) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 LJW OBJ 8.7.1.2 26 SH.1/F-4 2 B 6 GA MO C FE Q 8.5.5.4 (245) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 27 SH.1/A-5 2 C 6 CK SA C OT CMP-09 3.M.4-53 0 Disassemble (245) CT CMP-09 3.M.4-53 C Disassemble

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 75 of 174 P&ID: M245 (SHEET 1) & M246 (SHEET 1) SYSTEM: REACTOR CORE COOLING SYSTEM (1301)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 31 SH.1/G-5 2 AC 1 RL SA C RT 10Y 8.1.26.3 (246) LX NA 8.1.26.3 RT Verifies 32 SH.1/D-2 2 B 1 GL AO C NA NA NA Skid-Mounted (M245) LCV-Exempt***

34 SH.1/C-3 2 B 1 GL AO 0 FE Q 8.5.5.4 (245) ST Q 8.5.5.4 C FS Q 8.5.5.4 C PI 2Y 8.5.5.4 40 SH.1/B-5 2 C 2 CK SA C OT Q 8.5.5.1 0 STP-02 Skid-Mounted (245) CT RI 8.5.5.12 41 SH.1/C-6 2 C 8 CK SA C OT RI# 8.5.5.1 0 (245) CT RI 8.7.1.2 C RJ-08 42 SH.1/C-5 2 AC 1 RL SA C RT 10Y 8.1.26.3 (246) LX NA 8.1.26.3 RT Verifies 43 SH.1/D-5 2 C 2 PCV SA 0 NA NA NA STP-02 Skid-Mounted (M246) PCV-Excluded 47 SH.1/C-5 2 AC 2 CK SA C OT CMP-11 3.M.4-112 0 NIT (245) CT CMP-11 8.7.1.2 C 3.M.4-112 LJW OBJ 8.7.1.2 48 SH.1/E-5 2 B 4 GA MO 0 PI 2Y 8.1.11.11 None Passive, 8.1.30 (245) 49 SH.1/E-6 1 A 4 GA MO C FE CS 8.1.11.11 CS-03 (245) ST CS 8.1.11.11 O/C CS-03 PI 2Y 8.5.5.7 8.1.30 LJ OBJ 8.7.1.5 LP 2Y 8.5.5.7 50 SH.1/E-6 1 AC 4 CK SA C OT RI 8.1.11.9 0 RJ-24 (245) CT RI 8.1.11.9 C RJ-24 LP 2Y 8.5.5.7 53 SH.1/F-5 2 B 4 GL MO C FE Q 8.5.5.4 (245) ST Q 8.5.5.4 C PI 2Y 8.5.5.4 59 SH.1/B-6 2 AC 2 CK SA C OT RI 8.5.5.12 0 RJ-12 (245) CT RI 8.7.1.2 C RJ-08 LJW OBJ 8.7.1.2 60 SH.1/C-5 2 B 2 GL MO C FE Q 8.5.5.4 (245) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 61 SH.1/F-3 2 B 3 GA MO C FE Q 8.5.5.1 8.5.5.4 (246) ST Q 8.5.5.1 O/C 8.5.5.4 PI 2Y 8.5.5.1 8.5.5.4

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

CV-1301-32 is an Exempt level control valve (LCV) that functions when the RCIC Pump/Turbine is in the "standby" condition, and has no safety stroke function for open or close operation.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 76 of 174 P&ID: M245 (SHEET 1) & M246 (SHEET 1) SYSTEM: REACTOR CORE COOLING SYSTEM (1301)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 62 SH.1/C-5 2 B 2 GL MO C FE Q 8.5.5.4 (246) ST Q 8.5.5.4 O/C PI 2Y 8.5.5.4 63 SH.1/B-6 2 C 2 CK SA C OT Q 8.5.5.1 0 STP-02 Skid-Mounted (246) CT RI 8.1.31% C 64 SH.1/B-6 2 AC 8 SC SA **LO/C OT RI# 8.5.5.1 0 (245) CT RI 8.7.1.2 C RJ-08 LJW OBJ 8.7.1.2 70 SH.1/C-3 2 AC 1.5 RL SA C RT 10Y 8.1.26.3 STP-02 Skid-Mounted (246) LX NA 8.1.26.3 RT Verifies 9067 SH.1/C-6 2 C 1 SK SA C OT Q 8.5.5.12 0 STP-02 Skid-Mounted (245) CT RI 8.5.5.1 C

- Stop Check Valve: Handwheel Position - Locked Open, Normal Disc Position - Closed

% PNPS 8.1.31 performs a supplemental skid check valve functional close test (CT) on 1301-63. This Procedure performs the postmaintenance close test for 1301-63.

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 77 of 174 P&ID: M247 SYSTEM: REACTOR WATER CLEANUP SYSTEM (1201 & 12)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 2 G-7 1 A 6 GA MO 0 FE Q 8.6.5.2 ST Q 8.6.5.2 C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 5 G-7 1 A 6 GA MO 0 FE Q 8.6.5.2 ST Q 8.6.5.2 C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 80 G-5 1 A 4 GL MO 0 FE Q 8.6.5.2 ST Q 8.6.5.2 C PI 2Y 8.7.1.5 8.1.30 LJ OBJ 8.7.1.5 81 H-7 1 C 6 CK SA 0 OT RI# 8.1.31 CT RI 8.1.21 C RJ-16 85 G-8 1 B 6 GA MO 0 FE Q 8.6.5.2 ST Q 8.6.5.2 C PI 2Y 8.6.5.2 360 F-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 (12) CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 361 F-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 (12) CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 78 of 174 P&ID: M249 SYSTEM: STANDBY LIQUID CONTROL SYSTEM (1101)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 1 C-7 1 NA 1.5 GA MA LO NA NA NA 0 Passive (Manual)

PI 2Y 8.7.1.2.1 8.1.30 & 8.7.1.5 15 C-7 1 C 1.5 CK SA C OT RI 8.4.6 0 RJ-15 CT RI 8.4.2.2 C RJ-21 LSBI 2Y** 8.4.2.2 VTP-01 16 D-8 1 AC 1.5 CK SA C OT CMP-07 8.4.6 0 CT CMP-07 8.7.1.5 C 8.7.1.2.1 LJ OBJ 8.7.1.5 43A E-6 2 C 1.5 CK SA C OT CMP-07 8.4.1 0 CT CMP-07 8.4.2.1 C LSBI PBT 8.4.2.1 VTP-01 43B D-6 2 C 1.5 CK SA C OT CMP-07 8.4.1 0 CT CMP-07 8.4.2.1 C LSBI PBT 8.4.2.1 VTP-01 1105A E-6 2 C 1 RL SA C RT 10Y 8.1.26.3 1105B D-6 2 C 1 RL SA C RT 10Y 8.1.26.3 1106-A E-8 2 D 1.5 SH EX C EX 10Y 8.4.6 O/C 1106-B E-8 2 D 1.5 SH EX C EX 10Y 8.4.6 O/C

- VTP-01 specifies an administrative two year (2Y) test frequency.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 79 of 174 P&ID: M250 (SHEET 2) SYSTEM: CONTROL ROD DRIVE HYDRAULIC (302)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 21A SH.2/G-3 2 B 1 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 21B SH.2/G-7 2 B 1 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 22A SH.2/D-4 2 B 2 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 22B SH.2/D-7 2 B 2 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 23A SH.2/G-3 2 B 1 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 23B SH.2/G-7 2 B 1 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 P1 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 24A SH.2/D-4 2 B 2 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01 24B SH.2/D-7 2 B 2 PG AO 0 FE Q 8.3.3 ST Q 8.3.3 0 ST RI 8.M.1-31 C RJ-17 FS Q 8.3.3 PI 2Y 8.3.3 LSBI PBT 8.3.3.2 VTP-01

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 80 of 174 P&ID: M250 (SHEET 1) SYSTEM: CONTROL ROD DRIVE HYDRAULIC UNITS (305)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 114 (EP) SH.1/H-3 2 C 0.75 CK SA C OT NA 9.9 0 STP-01 Skid-Mounted 115 (EP) SH.1/G-2 2 C 0.5 CK SA C OT NA 9.9 STP-01 Skid-Mounted CT 8.1.25 C 120 (EP) SH.1/H-2 2 B 0.5 SV SO C FE/FS NA 8.3.2 C STP-01 Skid-Mounted 121 (EP) SH.1/H-2 2 B 0.5 SV SO C FE/FS NA 8.3.2 C STP-01 Skid-Mounted 122 (EP) SH.1/H-3 2 B 0.5 SV SO C FE/FS NA 8.3.2 C STP-01 Skid-Mounted 123 (EP) SH.1/H-3 2 B 0.5 SV SO C FE/FS NA 8.3.2 C STP-01 Skid-Mounted 126 (EP) SH.1/G-3 2 B 1 PG AO C FE/FS NA 9.9 O/C STP-01 Skid-Mounted 127 (EP) SH.1/H-3 2 B 0.75 PG AO C FE/FS NA 9.9 O/C STP-01 Skid-Mounted 132 (EP) SH.1/G-3 2 D 0.5 RD NA C RD NA NA+ C STP-01 Skid-Mounted Exempt 138 (EP) SH.1/G-3 2 C 0.5 CK SA 0 CT NA 8.3.2 C STP-01 Skid-Mounted

+ The 305-RD-132 rupture discs are exempt from IST scope per ISTC-1200.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 81 of 174 P&ID: M251 (SHEET 1 & SHEET 2) SYSTEM: RECIRC PUMP INSTRUMENTATION (262)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 25A SH.1/C-7 2 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 25B SH.2/C-7 2 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 26A SH.1/C-7 2 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 26B SH.2/C-7 2 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 F013A SH.1/E-6 2 AC 0.75 CK SA 0 OT RI# 8.1.31 CT RI 8.7.1.2.1 C RJ-03 8.7.1.5 LJ OBJ 8.7.1.5 F013B SH.2/E-6 2 AC 0.75 CK SA 0 OT RI# 8.1.31 CT RI 8.7.1.2.1 C RJ-03 8.7.1.5 LJ OBJ 8.7.1.5 F017A SH.1/E-7 2 AC 0.75 CK SA 0 OT RI# 8.1.31 CT RI 8.7.1.2.1 C RJ-03 8.7.1.5 LJ OBJ 8.7.1.5 F017B SH.2/E-7 2 AC 0.75 CK SA 0 OT RI# 8.1.31 CT RI 8.7.1.2.1 C RJ-03 8.7.1.5 LJ OBJ 8.7.1.5

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 82 of 174 P&ID: M252 (SHEET 2) SYSTEM: FEEDWATER SYSTEM (6)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freq Number Direction Justification Notes 58A SH.2/F-6 1 AC 18 CK SA 0 OT RI# 9.3 0 Note 1 CT RI 8.7.1.5 C RJ-02 LJ OBJ 8.7.1.5 58B SH.2/E-6 1 AC 18 CK SA 0 OT RI# 9.3 0 Note 1 CT RI 8.7.1.5 C RJ-02 LJ OBJ 8.7.1.5 62A SH.2/F-7 1 AC 18 CK SA 0 OT RI# 9.3 Note 1 CT RI 8.7.1.5 C RJ-02 LJ OBJ 8.7.1.5 62B SH.2/E-7 1 AC 18 CK SA 0 OT RI# 9.3 Note 1 CT RI 8.7.1.5 C RJ-02 LJ OBJ 8.7.1.5 548A SH.2/F-7 2 C 0.25 CK SA C NA NA NA Note 2 548B SH.2/E-7 2 C 0.25 CK SA C NA NA NA Note 2

Denotes a check valve exercise open test or exercise close test that is not refueling outage dependent and may be performed more frequently than once a refueling interval.

Note 1: Performed when the Reactor is _> 70% rated power. PMT open exercise test (OT) for these valves may be performed using applicable steps contained within PNPS 3.M.4-49 or PNPS 3.M.4-53.

Note 2: Check valves CK-6-548A and CK-6-548B, perform a maintenance and operational function only, and have no active or passive safety related functions.

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 83 of 174 P&ID: M252 (SHEET 1) SYSTEM: NITROGEN SUPPLY (9)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 378 SH.1/A-8 2 A 1 GA MA C FE NA NA Passive (Manual)

LJ OBJ 8.7.1.5 379 SH.1/A-8 2 A 1 GA MA C FE NA NA Passive (Manual)

LJ OBJ 8.7.1.5

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 84 of 174 P&ID: M252 (SHEET 2) SYSTEM: REACTOR RECIRCULATION SYSTEM (202)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 4A SH.2/C-5 1 NA 28 GA MO 0 NA NA NA Passive PI 2Y 8.1.11.1 4B SH.2/A-5 1 NA 28 GA MO 0 NA NA NA Passive PI 2Y 8.1.11.2 5A SH.2/C-3 1 B 28 GA MO 0 FE RI 8.1.11.1 RJ-19 ST RI 8.1.11.1 C RJ-19 PI 2Y 8.1.11.1 5B SH.2/A-3 1 B 28 GA MO 0 FE RI 8.1.11.2 RJ-19 ST RI 8.1.11.2 C RJ-19 PI 2Y 8.1.11.2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 85 of 174 P&ID: M252 (SHEET 1) SYSTEM: MN STM ISOL., ADS, & SAFETY RELIEF (203)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 1A SH.1/G-6 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS RI 8.1.11.21 C RJ-11 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 1B SH.1/E-6 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS RI 8.1.11.21 C RJ-11 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 1C SH.1/D-6 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS RI 8.1.11.21 C RJ-11 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 1D SH.1/B-6 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS RI 8.1.11.21 C RJ-11 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 2A SH.1/G-4 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS CS 8.1.11.21 C CS-05 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 2B SH.1/E-4 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS CS 8.1.11.21 C CS-05 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 2C SH.1/D-4 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS CS 8.1.11.21 C CS-05 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6 2D SH.1/B-4 1 A 20 GL AO 0 FE CS 8.7.4.4 CS-05 ST CS 8.7.4.4 C CS-05 FS CS 8.1.11.21 C CS-05 PI 2Y 8.1.11.21 8.M.1-15 LJ OBJ 8.7.1.6

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 86 of 174 P&ID: M252 (SHEET 1) SYSTEM: MN STM ISOL., ADS, & SAFETY RELIEF (203)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 3A SH.1/G-7 1 AC 6 RL/ AO/SA C FE 5Y 8.5.6.2 SV PI 2Y 8.5.6.2 RT 5Y SEP-PNPS-IST-005 LX NA SEP-PNPS-IST-005 RT Verifies 3B SH.1/C-7 1 AC 6 RL/ AO/SA C FE 5Y 8.5.6.2 SV P1 2Y 8.5.6.2 RT 5Y SEP-PNPS-IST-005 LX NA SEP-PNPS-IST-005 RT Verifies 3C SH.1/A-7 1 AC 6 RL/ AO/SA C FE 5Y 8.5.6.2 SV PI 2Y 8.5.6.2 RT 5Y SEP-PNPS-IST-005 LX NA SEP-PNPS-IST-005 RT Verifies 3D SH.1/F-7 1 AC 6 RL/ AO/SA C FE 5Y 8.5.6.2 SV PI 2Y 8.5.6.2 RT 5Y SEP-PNPS-IST-005 LX NA SEP-PNPS-IST-005 RT Verifies 4A SH.1/G-7 1 AC 6 SV SA C RT 5Y SEP-PNPS-IST-005 LX NA SEP-PNPS-IST-005 RT Verifies 4B SH.1/D-7 1 AC 6 SV SA C RT 5Y SEP-PNPS-IST-005 LX NA SEP-PNPS-IST-005 RT Verifies

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 87 of 174 P&ID: M252 (SHEET 1 & SHEET 2) SYSTEM: NUCLEAR BOILER SYSTEM (220)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 1 SH.1/G-5 1 A 3 GA MO C FE Q 8.7.4.3 ST Q 8.7.4.3 C PI 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 2 SH.1/G-5 1 A 3 GA MO C FE Q 8.7.4.3 ST Q 8.7.4.3 C PI 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 44 SH.2/D-6 1 A 1 GA AO C FE Q 8.7.4.3 ST Q 8.7.4.3 C FS Q 8.7.4.3 C PI 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 45 SH.2/D-7 1 A 1 GL AO C FE Q 8.7.4.3 ST Q 8.7.4.3 C FS Q 8.7.4.3 C PI 2Y 8.7.4.3 8.1.30 LJ OBJ 8.7.1.5 46 SH.2/H-3 1 B 1 GL SV C FE CS 8.1.11.11 CS-02 ST CS 8.1.11.11 C CS-02 FS CS 8.1.11.11 C CS-02 PI 2Y 8.1.11.11 8.1.30 47 SH.2/H-3 1 B 1 GL SV C FE CS 8.1.11.11 CS&02 ST CS 8.1.11.11 C CS-02 FS CS 8.1.11.11 C CS-02 PI 2Y 8.1.11.11 8.1.30

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 88 of 174 P&ID: M252 (SHEET 1 & SHEET 2) SYSTEM: NUCLEAR BOILER (261)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 1-17A SH.1/G-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-17B SH.1/E-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-17C SH.1/C-5 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-17D SH.1/B-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-18A SH.1/F-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-18B SH.1/E-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-18C SH.1/C-5 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 1-18D SH.1/B-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 19A SH.2/D-2 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 19B SH.2/E-2 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 20A SH.2/D-2 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 20B SH.2/E-2 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 21A SH.2/C-1 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 89 of 174 P&ID: M252 (SHEET 2) SYSTEM: NUCLEAR BOILER (261)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 21B SH.2/C-1 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 22A SH.2/C-1 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 22B SH.2/C-1 1 C 1 EF SA O OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67A SH.2/F-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67B SH.2/F-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67C SH.2/F-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67D SH.2/F-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67E SH.2/G-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67F SH.2/G-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67G SH.2/G-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 67H SH.2/G-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 110A SH.2/C-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 110B SH.2/E-2 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1, 2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 90 of 174 P&ID: M253 (SHEET 1 & SHEET 2) SYSTEM: NUCLEAR BOILER VESSEL INSTRUMENTATION (263)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 2-125A SH.1/F-6 1 C 1 EF SA 0 OT CMP-14 8.M.3-2 0 CT CMP-14 8.M.3-2 C LEF CMP-14 8.M.3-2 2-125B SH.1/F-3 1 C 1 EF SA 0 OT CMP-14 8.M.3-2 0 CT CMP-14 8.M.3-2 C LEF CMP-14 8.M.3-2 38 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 44 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 45 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 51 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 53 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 55 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 57 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 59 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 61 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 69 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 71 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 91 of 174 P&ID: M253 (SHEET 1 & SHEET 2) SYSTEM: NUCLEAR BOILER VESSEL INSTRUMENTATION (263)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 73 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 75 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 77 SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 79 SH.2/E-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 81 SH.2/E-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 83 SH.1/E-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 90 SH.2/E-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 92 SH.2/D-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 215A SH.1/F-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 215B SH.1/F-3 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 217A SH.1/E-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 217B SH.1/E-3 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 219A SH.1/E-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 92 of 174 P&ID: M253 (SHEET 1 & SHEET 2) SYSTEM: NUCLEAR BOILER VESSEL INSTRUMENTATION (263)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 219B SH.1/E-3 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 220A SH.2/D-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 220B SH.2/D-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 223A SH.2/E-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 223B SH.2/E-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 225 SH.2/C-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 227 SH.2/C-7 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 231A SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 231B SH.2/B-8 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 233 SH.2/D-4 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 237 SH.1/E-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 242A SH.1/D-6 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2 242B SH.1/D-3 1 C 1 EF SA 0 OT CMP-16 8.M.3-2 0 CT CMP-16 8.M.3-2 C LEF CMP-16 8.M.3-2

TABLE INSERVICE VALVE TESTING PROGRAM SEP-PNPS-IST-001 IST CLASS 1,2 AND 3 VALVES Rev.: 2 PILGRIM NUCLEAR POWER STATION Page 93 of 174 P&ID: MIQ-1-5 (VENDOR DRAWING) SYSTEM: TRAVERSING IN-CORE PROBE (45)

Valve P&ID IST Valve Valve Valve Actuator Normal Test Test PNPS Proc. Safety Relief/

Number Sh/Coor Class Cat Size Type Type Position Rqmt. Freg Number Direction Justification Notes 300A NA 2 A 0.25 BL SO C FE/FS Q 8.7.4.3 C STP-04 Skid-Mounted (Ball A) PI NA 9.6 LJ OBJ 8.7.1.5 300B NA 2 A 0.25 BL SO C FE/FS Q 8.7.4.3 C STP-04 Skid-Mounted (Ball B) PI NA 9.6 LJ OBJ 8.7.1.5 300C NA 2 A 0.25 BL SO C FE/FS Q 8.7.4.3 C STP-04 Skid-Mounted (Ball C) PI NA 9.6 LJ OBJ 8.7.1.5 300D NA 2 A 0.25 BL SO C FE/FS Q 8.7.4.3 C STP-04 Skid-Mounted (Ball D) PI NA 9.6 LJ OBJ 8.7.1.5 (N2 Check) NA 2 AC 0.25 CK SA 0 CT OBJ 8.7.1.5 C STP-04 Skid-Mounted LJ OBJ 8.7.1.5 301A NA 2 D 0.25 SH EX 0 EX 10Y 3.M.2-5.6.8 C STP-04 Skid-Mounted (Shear A) 301B NA 2 D 0.25 SH EX 0 EX 10Y 3.M.2-5.6.8 C STP-04 Skid-Mounted (Shear B) 301C NA 2 D 0.25 SH EX 0 EX 10Y 3.M.2-5.6.8 C STP-04 Skid-Mounted (Shear C) 301D NA 2 D 0.25 SH EX 0 EX 10Y 3.M.2-5.6.8 C STP-04 Skid-Mounted (Shear D)

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 94 of 174 7.0 PROGRAM JUSTIFICATIONS AND RELIEF REQUESTS 7.1 PUMP TESTING PROGRAM RELIEF REQUESTS Pump Relief Requests (PR) are provided for conditions in which compliance to ASME OM Code Subsection ISTB test requirements cannot practically be satisfied. Each Relief Request identifies:

pump(s) involved, test requirement(s) of noncompliance, basis for relief, and alternate testing.

[1] PUMP RELIEF REQUEST PR-01 Retracted

[2] PUMP RELIEF REQUEST PR-02 Retracted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 95 of 174

[3] PUMP RELIEF REQUEST PR-03 (Revision 4)

Information to Support NRC Approval of a Revised 10CFR50.55a Request (PR-03 Revision 4), which is Similar to a Previously Approved Request (PR-03 Revision 3),

for Use During a New 10-Year Interval Inservice Testing Program Pump: P-205 (Main/Booster)

System: High Pressure Coolant Injection (HPCI)

Class: 2 Function:

Provides emergency core cooling subsequent to a small break LOCA.

Test Requirements:

ASME OM Code OMb-2006, ISTB-5123, Comprehensive Test ISTB-5123(e): All deviations from the reference values shall be compared with the ranges of Table ISTB-5121-1 and corrective action taken as specified in paragraph ISTB-6200. The vibration measurements shall be compared to the relative and absolute criteria shown in the Alert and Required Action Ranges of Table ISTB-5121-1. For example, if vibration exceeds either 6 Vr or 0.7 in./sec, the pump is in the Required Action Range.

Relief Requested:

Relief is requested from the ASME OMb-2006, ISTB-5123(e) required vibration ranges for Acceptable, Alert, and Required Action as specified in Table ISTB-5121-1.

PNPS requests relief from the Acceptance Range Code requirements of paragraph ISTB-5123(e) for the HPCI Main and Booster pumps, specifically from the vibration velocity (V,)

Acceptance criteria specified in Table ISTB-5121-1 for all Main pump and Booster pump vibration points except for the Main pump outboard vertical vibration point (P4V) and Booster pump outboard horizontal axial vibration point (P8A). PNPS proposes to expand the Acceptable Range (and corresponding Alert Low criteria) identified in Table ISTB-5121-1 for the Code specified Biennial Comprehensive pump vibration monitoring.

PNPS requests relief from the High Alert Range (and corresponding Required Action Limit)

Code requirements of paragraph ISTB-5123(e) for the HPCI Main pump, specifically from the vibration velocity (Vv) High Alert criteria specified in Table ISTB-5121-1 for Main pump inboard (turbine side) bearing horizontal point (P3H) and Main pump outboard (gearbox side) bearing horizontal point (P4H). PNPS proposes to expand the High Alert criteria (and corresponding Required Action Limit) identified in Table ISTB-5121-1 for the Code specified Biennial Comprehensive pump vibration monitoring.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 96 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

Relief Requested:

Relief from the referenced Code requirements is based on the determination that the proposed alternative testing coupled with supplemental pump vibration monitoring, performance monitoring and maintenance activities provide reasonable assurance of operational readiness because PNPS will maintain consistent (alternative) vibration alert ranges and action ranges. The proposed alternative to the OM Code requirements regarding raising the allowable vibration level for the HPCI main and booster pump bearings is pursuant to 10CFR50.55a(a)(3)(ii) on the basis that compliance with the specified requirements results in an undo hardship without a compensating increase of the level of quality and safety.

Historic testing and analysis performed on the HPCI System by PNPS (and the pump manufacturer) have consistently revealed characteristic pump vibration levels that exceed the acceptance criteria stated in Table ISTB-5121-1. High vibration appears on the Main pump bearing housings at approximately 2x RPM in the horizontal direction, which is caused by Booster pump excitation (at 4x RPM of the Booster pump). Under normal circumstances at 4000 RPM, the vibration amplitude at the Main pump bearings in the horizontal direction exceeds the OM Code absolute vibration Required Action Range of > 0.7 in./sec.

Additionally, under the same conditions, all of the remaining HPCI Main and Booster pump vibration monitoring points, except for three, typically exceed the OM Code absolute acceptable range upper value of 0.325 in./sec.

Basis for Relief:

The vibration characteristics of the HPCI pump are predominantly a function of the pump design and should be identified as such rather than attributed to pump degradation. The high vibration has been present to the same order of magnitude since the pump was new.

Although existing vibration levels of the HPCI pump are higher than the acceptance criteria provided in Table ISTB-5121-1, they reflect the unique operating characteristics of the HPCI pump design configuration. The historical data shows that there are no inherent vibrational concerns that would result in pump degradation or would prevent the HPCI pump from performing its design safety function for an extended period of operation.

The purpose of the Code required testing is to demonstrate the operational readiness of the HPCI pump by monitoring pump vibrations for degradation and taking corrective actions when those vibration levels exceed the Code specified values. The Code specifies in ISTB-3300(g) and ISTB-6400 (including footnote) that the reference vibration measurements should be representative of the pump and that the measured vibration will not prevent the pump from fulfilling its function. Accordingly, PNPS is proposing to perform supplemental performance monitoring to support expansion of the Code specified test Acceptance (for Main & Booster pump vibration points P3H & P4H P3V, P3A, P7H, P8H, P7V, and P8V) and High Alert vibration limits (for Main pump vibration points P3H & P4H) that will be used to demonstrate the operational readiness, which take into consideration the vibration measurements representative of the as-built configuration of the HPCI pump.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 97 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

Proposed Supplemental Testing, Performance Monitoring, and Preventative Maintenance Activities to ASME OMb-2006 Code Pump Testing:

PNPS will perform supplemental testing and performance monitoring activities as follows:

1. A supplemental vibration monitoring activity will be performed to remove the 4x Booster pump RPM frequency component (discrete peak) from the vibration spectrum of the Main pump since its amplitude is not related to the physical condition or rotating dynamics of the Main pump rotor or bearing system. The Main pump vibration spectrum, with this single 4x Booster pump RPM frequency component removed, has been shown to be stable and more useful for monitoring actual pump condition. When this vibration frequency component at 4x Booster pump RPM is subtracted from the Main pump vibration spectrum, the remaining vibration, which is attributed to the Main pump, is below the OM Code Required Action Range. This modified (corrected) vibration level provides a more representative measurement of the pump condition, and thus, will be reviewed and trended as a supplemental vibration performance monitoring activity. Adverse trends and/or vibration anomalies will be evaluated and dispositioned, as necessary, using the PNPS Corrective Action Program.

2. As part of a vibration frequency spectrum review, all other discrete vibration peaks observed at the Main pump horizontal vibration points will be evaluated following each pump vibration test and will be reviewed for adverse trends and abnormalities.

The reviews of the frequency spectrum data ensure that any significant change in the vibration signature at these points will be noted, and when necessary, dispositioned within the PNPS Condition Monitoring Program regardless of whether the severity causes the overall vibration level to exceed its criteria.

3. PNPS will increase the ASME OMb-2006, ISTB-3400 required frequency for vibration monitoring (that is part of the comprehensive testing) from once/2 years to once/year.

The Code required comprehensive test for flow rates would continue to be once/2 years. Given that the HPCI vibration will normally exceed the OM Code limiting Alert Range of > 0.325 in./sec, the once/year frequency will be doubled to twice/year. The twice/year frequency will be the commitment frequency. However, the normal PNPS practice will be to monitor vibration in the same manner during each of the Quarterly Group B Hydraulic Tests, whenever practicable. Thus, vibration monitoring will be performed up to 8 times in 2 years as part of the Group B Hydraulic Tests instead of once/2 years as part of the Comprehensive pump tests

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 98 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

Proposed Supplemental Testing, Performance Monitoring, and Preventative Maintenance Activities to ASME OMb-2006 Code Pump Testing:

4. As normal practice, PNPS will continue to monitor vibration of HPCI pump during each of the Quarterly-Group B Hydraulic Tests in the same manner as required by the OM Code. The preventive maintenance (PM) procedure will also typically be performed, which provides for vibration monitoring of specific pumps for preventive maintenance and balancing, and includes vibration monitoring and trending of the HPCI pump to detect and monitor changes in equipment conditions. As shown in the HPCI pump configuration figure, vibration monitoring is performed at locations required by the OM Code and at additional locations within the scope of the PM procedure (perpendicular to the shaft in the horizontal and vertical positions at each bearing location and at axial direction to the shaft). Vibration monitoring is thereby routinely performed for the Main pump, Booster pump, Speed Reduction Gearbox, and Steam Turbine. Using the vibration data collected at these points, when changes to PM program overall vibration occurs, accurate diagnosis may be conducted by analyzing the related vibration point spectrums. Planned maintenance may then be determined and initiated to prevent failures. Thus, HPCI pump vibration monitoring will be performed up to 8 times in 2 years as part of Group B Hydraulic Tests and preventive or corrective maintenance will be implemented as necessary to prevent failures. Enclosures 1 and 2 (of Entergy Outgoing NRC Letter 02.08.007) provide HPCI pump vibration spectrum at locations required by the OM Code procedure.

5. PNPS will continue current HPCI pump and turbine monitoring and maintenance activities, with changes as conditions warrant, as follows:

" Quarterly pump and valve operability tests will be performed to ensure the HPCI pump and turbine function for the intended safety function.

Quarterly lubrication oil sampling and periodic laboratory analysis as appropriate for the pressure-fed bearings on the Turbine, Main pump, and Gear Reducer and once/cycle (2 years) sampling and analysis for the nonpressure-fed Booster pump will be performed. Lubrication oil analysis currently performed includes viscosity, acidity, residue, water content, metals by A.E. spectrometry, and ferrogram readings.

This type of monitoring will detect degradation of the turbine or pump bearings due to accelerated wear, fretting, surface fatigue, or oil contamination.

" HPCI pump and Turbine lube oil system is serviced as needed weekly. HPCI gland seal condenser hot well pump and motor bearings and HPCI auxiliary lube oil pump and motor bearings are serviced semiannually for lubrication.

HPCI Turbine/Main pump, Main pump/Reducer, and Reducer/Booster pump gear-type shaft couplings are cleaned, examined, and grease-lubricated every 2 years.

The Main pump/Reducer, and Reducer/Booster pump gear-type shaft couplings are cleaned, examined, and grease-lubricated every 4 years. These examinations detect excessive wear, fretting, heating, or fatigue due to any unusual loading conditions.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 99 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

Past monitoring and maintenance activities have shown no evidence or observations of degradation in the HPCI turbine, Main pump, Gear Reducer, or Booster pump.

Historical HPCI Main and Booster pump vibration spectra support this conclusion.

Thus, the continuation of the above periodic monitoring and maintenance activities will ensure that the HPCI pump remains in a high level of operational readiness and that degradation of HPCI pump mechanical condition, reliability, or performance will be detected and corrected in a timely manner.

Technical Justification:

PNPS has conducted an evaluation of the HPCI pump vibration characteristics. An important finding of this evaluation is that the mechanical condition of the Main pump can be monitored satisfactorily by disregarding the single frequency component caused by the excitation at 4x Booster pump RPM. The four-vane impeller of the Booster pump generates the excitation force hydraulically. This small pressure pulsation force exists at the vane passing frequency (number of vanes times RPM) for all centrifugal pumps and is usually seen as a significant but not particularly troublesome component on the frequency spectrum for vibration measurements taken at the bearing housings. For the HPCI pump, this vane passing frequency is a problem because it coincides with a hydraulic standing wave resonance in the cross-over piping from the Booster pump to the Main pump when the machine is operating at the rated speed of 4000 RPM. There is an acoustic pressure standing wave pattern, at the 4x RPM frequency, whose wavelength in water is equal to an even fraction (1/4 or 1/2) of the dimensional length inside the cross-over pipe. This is the same principle on which an organ pipe generates a pure tone pneumatic pressure standing wave.

In addition, and exacerbating the vibration resonance condition, the Main pump pedestal experiences a horizontal structural primary rocking mode of the pump pedestal at this same frequency when the Main pump is operating at the rated speed of 4000 RPM. The vibration mode is the second fundamental rocking mode, which is a torsional or twisting mode where the two end bearings move 180 degrees out of phase horizontally. The result of these coincident acoustic and structural resonances is that the Main pump exhibits high vibration in the horizontal direction at the 4x Booster pump RPM frequency. This is solely due to the excitation from the Booster pump being amplified by the coincident resonances. This level of vibration at 4x Booster pump RPM would be seen on the Main pump bearing housings even if the Main pump was not actually running (which is not possible as both pumps are on the same drive train).

The resonant vibration condition at the 4000 RPM operating speed is not detrimental and will not prevent the HPCI pump from fulfilling its function. At the 134 Hz frequency of the resonant vibration on the Main pump, caused by the excitation at 4x Booster pump RPM, the actual displacement amplitude at 0.7 in./sec peak velocity amplitude is 0.0017 inches peak-to-peak. This displacement imposes negligible alternating stresses on the pump pedestal, housings, and connected piping. The peak-to-peak displacement is also less than the Main pump fluid film journal bearing clearances and would impose negligible loading to these bearings.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 100 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

The purpose of the ASME OM Code for pump testing is to monitor pumps for degradation.

The concept of vibration monitoring is to establish baseline values for vibration when the pump is known to be in good working condition, such as after a maintenance overhaul.

From that reference point, trending is performed to monitor for degradation based on the ratio of subsequent vibration levels relative to the reference values. The OM Code also establishes absolute vibration level criteria for Alert (> 0.325 in./sec) and Required Action

(> 0.70 in./sec). In doing so, it was recognized that absolute vibration level limits (as opposed to relative change or ratio limits) are not always quantitatively linked directly with pump physical condition and the following remarks are stated in the ASME OMb Code 2006:

"Vibration measurements of pumps may be foundation, driver, and piping dependent.

Therefore, if initial vibration readingsare high and have no obvious relationshipto the pump, then vibration measurements should be taken at the driver, at the foundation, and on the piping and analyzed to ensure that the reference vibration measurements are representativeof the pump and that the measured vibration levels will not prevent the pump from fulfilling its function."

A significant conclusion of the PNPS HPCI pump vibration evaluation is that supplemental performance monitoring of the Main pump, by disregarding the single frequency component caused by the excitation at 4x Booster pump RPM, provides an important tool for assessing and trending pump mechanical condition. A single peak frequency component can be effectively isolated and deleted from a vibration spectrum using the mean-squared subtraction method; that is, the discrete component amplitude (in./sec peak) is squared and subtracted from the spectrum overall level squared, then the square root of that difference represents the overall vibration level that exists without the energy contributed by the deleted component. It has been found that when this method is used, the remaining vibration overall value is much more consistent, stable, and trendable.

This method of vibration level correction has been applied to historical spectrums. The 4x Booster pump RPM component was taken out of the calculation for the Main pump overall vibration level. This data shows that when the 4x Booster pump RPM component is deleted from the Main pump vibration, the level is below the Required Action Range (> 0.7 in./sec) but still within the Alert Range (> 0.325 in./sec). It was also shown that the potential effects from the dynamic alignment of pump shaft couplings (at 2X Main pump RPM) can still be monitored effectively.

The vibration spectra derived from Inservice testing conducted in November 2005 conforms to the historical vibration spectra documented since 1994. Enclosure 1 (of Entergy Outgoing NRC Letter 02.08.007) provides the November 2005 test results and Enclosure 2 (of Entergy Outgoing NRC Letter 02.08.007) provides the historical tests results. Since comparison of the observed vibration spectra showed no change during the review time period, likewise no degradation in the established operational readiness of the HPCI pump occurred for this 11 year period. Similarly, this supplemental performance monitoring will support the Code specified testing for verifying the operational readiness of the HPCI pump in its as-built configuration as stipulated by ISTB-3300(g) and ISTB-6400, with corresponding footnote.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 101 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

Impact of Potential Modifications:

For the HPCI Main and Booster pumps, it has been determined that the vibration is foundation and piping dependent. To reduce the HPCI Main and Booster pump vibration down to levels that meet acceptable OM Code vibration criteria requires modifications to the HPCI pump, mounting components, foundation, and/or cross-over (interconnecting) piping.

As suggested in a Byron Jackson Tech Note, this vibration may be improved by modifying the interconnecting piping and the Main pump mounting pedestal. The alternative modification changes the Booster pump impeller from four to five vanes to alter the forcing function of the standing wave resonance.

The proposed Byron Jackson modifications, other than replacing the Booster pump impeller, are generally very difficult to implement successfully. Altering the natural frequency of a large pump installation requires either considerable additions of stiffening components or substantial additions of mass. Often the results of such design changes are unsuccessful or unfavorable due to the variable speed operation requirements.

Modification of the HPCI Booster pump would require replacing the current four-vane impeller with an upgraded five-vane impeller. The impeller modification, although yielding predictable results, requires extensive work to the HPCI pump at a time when such a major rebuild of this pump is not otherwise necessary or desired. The expected result would be a modest decrease in the vibration caused on the Main pump at 4000 RPM, although the vibration would remain above the 0.325 inch/sec Alert Range criteria. A small decrease in hydraulic performance is also expected when changing from a four to five-vane impeller.

The proposed major modification would cost approximately $500,000 without a compensating improvement in the pump vibration. Most HPCI pump vibration points would remain above the 0.325 in./sec Alert Criteria. Accordingly, the proposed modification would not achieve the underlying objective of performing the Code required testing without the need for Code relief.

PNPS has also concluded that none of the possible modifications that could be performed on the HPCI pump, mounting pedestal, or cross-over piping are necessary. This is primarily due to the nature of the HPCI pump service profile. The Byron Jackson Tech Note describes the following consideration in the Technical Discussion:

"Pumpingsystems in which the vane passing pressure pulsations form standing waves in the attachedpiping are not unusual, especially if the pumps have a variable speed driver. Standing waves are highly dependent upon water temperature. Thus, measured vibration amplitudes often vary from test to test."

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 102 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

The HPCI pump service is such that the pump runs for short periods of time at highly variable speeds. The pump inservice testing at PNPS is performed with the pump operating at or close to its rated speed (4000 RPM) and flow conditions (4250 GPM) that are unique to PNPS. For this particular pump configuration, this pump speed corresponds to the point where the acoustic resonant vibration is typically most pronounced. In actual service for high pressure coolant injection to the Reactor, the pump will operate at the speed that the flow controller requires to maintain Reactor water level. The flow rate of 4250 GPM is the maximum makeup flow rate for which the HPCI System was intended to be capable of maintaining Reactor water level. This flow rate is far in excess of the decay heat makeup water requirements for the Reactor in the isolated condition in the absence of a major leak.

The pump speed required is also dependent on Reactor pressure with the required speed decreasing along with Reactor pressure.

The same general HPCI pump configuration is used at other plants but often with different pump impellers, rated speeds, and plant design flow rates. For these plants the vibration characteristics at the inservice testing points are markedly different for that reason. The vibration monitoring performed (including a frequency spectral review) to date under the IST Program and the PNPS Pump Vibration Monitoring Program has shown that there has not been degradation of these HPCI pump components.

Inservice testing can be successfully performed for the PNPS HPCI pump using the methods proposed in this Relief Request, along with the supplemental monitoring and maintenance activities currently in practice. Any significant degradation of the HPCI pump components will be readily identified using the vibration spectral analysis methods and other preventive monitoring activities described in this Relief Request. In addition, Entergy believes that the proposed alternative testing and supplemental monitoring for the PNPS HPCI pump will provide an acceptable level of quality and safety. Therefore, Entergy requests authorization of the proposed alternative to the OM Code requirements regarding raising the allowable vibration level for the HPCI main and booster pump bearings pursuant to 10CFR50.55a(a)(3)(ii) on the basis that compliance with the specified requirements results in an undo hardship without a compensating increase of the level of quality and safety.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 103 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

Alternate Testing:

To allow for practicable monitoring of vibration levels on the HPCI pump, alternate vibration acceptance criteria is necessary. A full spectrum review will be performed for all IST vibration points during each proposed comprehensive test utilizing the following criteria.

The table below provides the acceptance criteria that will be applied to the overall vibration level for the Main pump. Note 1 clarifies the proposed performance monitoring that will be performed for the horizontal Main pump points, as the discrete frequency component at 4x Booster pump RPM will be extracted from the overall value using the mean-squared subtraction. The two modified (corrected) overall values (points P3H and P4H) resulting from the extraction of the discrete peak at 4x Booster pump RPM will be reviewed separately for change, and trended as a supplemental vibration performance monitoring activity.

The table box data typed in bold italics have Alert vibration range values and corresponding Required Action vibration limits that have been modified from the OM Code vibration criteria as follows:

P3H & P4H - The Alert vibration range of 1.1 Vr to 1.3 Vr (in lieu of the OM Code range of 2.5 Vr to 6 V,) has been applied as the modified OM vibration criteria (overall vibration) for points P3H and P4H. For these points the absolute limiting Low Alert values (i.e. 0.70 and 1.06) and absolute limiting High Alert values (i.e. 0.83 and 1.26) are based upon existing pump reference values. These two modified Alert ranges have been compared to historical pump vibration data. (The corresponding Required Action vibration limits were also modified to be consistent with the limiting High Alert values).

" P3V, P3A, P7H, P8H, P7V, & P8V - The Alert vibration range of 1.5 Vr to 6 Vr (in lieu of the OM Code range of 2.5 V, to 6 V,) has been applied as the modified OM vibration criteria (overall vibration) for points P3V, P3A, P7H, P8H, P7V, and P8V. For these points the absolute limiting Low Alert values (i.e., 0.400, 0.450, and 0.500) are based upon existing pump reference values, and fall between the values of 1.25 Vr and 1.5 Vr. All of the modified Alert Ranges have been compared to historical pump vibration data.

" The Table rows for P4V and P8A are in compliance with the OM Code vibration criteria and have been placed into this Relief Request for information only.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 104 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

MAIN PUMP**

Test Vibration Point Acceptable Range Alert Range Required Action Parameter Limit (Note 2)**

V, Main pump (Note 1)' _61.1 Vr > 1.1 Vr to 1.3 Vr > 1.3 Vr Horizontal but not or or Inboard (P3H) > 1.06 in.Isec > 1.06 to 1.26 > 1.26 in./sec in./sec V Main pump (Note 1)' *1.1 Vr > 1.5 Vr to 6 Vr > 1.3 Vr Horizontal but not or or Outboard (P4H) > 0.700 in.Isec > 0.700 to 0.83 > 0.83 in./sec in./sec V, Main pump _< 1.5 Vr > 1.5 Vr to 6 Vr > 6 Vr Vertical but not or or Inboard (P3V) > 0.450 in./sec > 0.450 to 0.70 > 0.70 in./sec in./sec Vv Main pump < 2.5 Vr > 2.5 Vr to 6 Vr > 6 Vr Vertical but not or or Outboard (P4V) > 0.325 in./sec > 0.325 to 0.70 > 0.70 in./sec in./sec Vv1 Main pump -<1.5 Vr > 1.5 Vr to 6 Vr > 6 Vr Axial but not or or Inboard (P3A) > 0.500 in./sec > 0.500 to 0.70 > 0.70 in./sec in./sec Notes:

1. ** For Main pump horizontal vibration points P3H and P4H, a frequency spectrum analysis will be performed following each pump vibration operability test, and the discrete peak at 4x Booster pump (4xBP) RPM will be extracted (using mean-squared subtraction method) from the vibration spectrum overall value. The modified P3H and P4H overall values, resulting from the extraction of the discrete peak at 4xBP RPM, will be reviewed and trended as a supplemental vibration performance monitoring activity.

2. ** The Required Action Range for overall vibration of points P3V, P4V, and P3A meets the Code required vibration values. The Required Action Range for overall vibration of points P3H and P4H has been increased to 1.3 Vr or 1.26 in/sec for P3H and 1.3 Vr or 0.83 in/sec for P4H.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 105 of 174

[3] PUMP RELIEF REQUEST PR-03 (CONTINUED)

BOOSTER PUMP Test Vibration Point Acceptable Alert Range Required Action Parameter Ranc e Limit V Booster pump <1.5 V, > 1.5 Vr to 6 Vr > 6 Vr Horizontal but not or or Inboard (P7H) > 0.450 in./sec > 0.450 to 0.70 > 0.70 in./sec in./sec V Booster pump *1.5 Vr > 1.5 Vr to 6 Vr > 6 Vr Horizontal but not or or Outboard (P8H) > 0.500 in./sec > 0.500 to 0.70 > 0.70 in./sec in./sec V Booster pump *1.5 Vr > 1.5 Vr to 6 Vr > 6 Vr Vertical but not or or Inboard (P7V) > 0.400 in./sec > 0.400 to 0.70 .> 0.70 in./sec in./sec V Booster pump _5 1.5 Vr > 1.5 Vr to 6 Vr > 6 Vr Vertical but not or or Outboard (P8V) > 0.500 in./sec > 0.500 to 0.70 > 0.70 in./sec in./sec Vv Booster pump *2.5 Vr >.2.5 Vr to 6 Vr > 6 Vr Axial but not or or Outboard (P8A) > 0.325 in./sec > 0.325 to 0.70 > 0.70 in./sec I in ./sec Duration of Approval for the Proposed Alternative This relief is requested for the duration of the PNPS 5 th Inservice Testing Ten-Year Interval (December 7 th, 2012 through December 6 th, 2022.

References:

1. NRC Letter, Pilgrim Nuclear Power Station - Entergy Relief Request PR-03 High Pressure Coolant Injection Pump (TAC No. MB8773), dated August 29, 2005.

2. Entergy Letter No. 02.05.042, Response to NRC Request for Additional Information Related to Pilgrim In-service Testing (IST) Relief Request PR-03 (TAC No. MB8773),

dated May 24, 2005.

3. Entergy Letter No. 02.05.012, Pilgrim Fourth Ten-year In-Service Testing Program, IST Relief Request PR-03, dated February 24, 2005.

4. Entergy Letter No. 02.08.007, Pilgrim Fourth Ten-Year In-Service Testing (IST) Program, IST Relief Request PR-03, Rev. 3, dated January 31, 2008 (including enclosures 1 &2).

5. NRC Letter, Pilgrim Nuclear Power Station - Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request for Relief No. PR-03, Revision 3 - Pilgrim Nuclear Power Station (TAC No. MD8052), dated August 27, 2008.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 106 of 174

[41 PUMP RELIEF REQUEST PR-04 Retracted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 107 of 174

[5] PUMP RELIEF REQUEST PR-05 Information to Support NRC Re-Approval of a 10CFR50.55a Request for Use During a New 10-Year Interval Inservice Testing Program Pumps: P-207A and P-207B System: Standby Liquid Control System (1101)

Class: 2 function:

Provides a method of shutting down the Reactor without use of the control rods.

Test Requirements:

ISTB-5300(a), Duration of Tests (1) For the Group A test and the comprehensive test, after pump conditions are as stable as the system permits, each pump shall be run at least 2 min. At the end of this time, at least one measurement or determination of each of the quantities required by Table ISTB-3000-1 shall be made and recorded.

Relief Requested:

Determine the Standby Liquid Control (SLC) pump hydraulic parameter (measured flow rate) by establishing the reference pump discharge pressure during the procedure initial conditions (prior to the start of the flow measurement test) in lieu of running the pump for at least 2 minutes after pump conditions are as stable as the system permits.

Basis For Relief:

The SLC pumps are tested by pumping fluid from the SLC storage tank into a test tank.

The test tank capacity does not allow operation of the pump for much longer than 3 minutes.

The present surveillance procedure has provided consistent test results and produces good repeatability.

During testing, the initial test conditions are established by starting the SLC pumps and adjusting the pump discharge test flow (throttle) valve to obtain the test reference discharge pressure. When the reference test pressure has been established, the pump is stopped and the initial test tank level is measured. Next, the pump is restarted and allowed to run for exactly 3 minutes. The test tank final level is measured and the pump flow rate is calculated.

Pump flow rate calculations meet the requirements of table ISTB-351 0-1 for measured values.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 108 of 174

[5] PUMP RELIEF REQUEST PR-05 (CONTINUED)

Alternate Testing:

The pump test procedure will establish the pump reference discharge pressure prior to conducting the 3-minute pump test run. When the initial conditions for reference discharge pressure are established, the SLC pump will be stopped and initial test tank level will be measured. Then the SLC pump will be run for a duration of exactly 3 minutes. An accurate measurement of the initial test tank level and final test tank level will be used to determine the measured test flow rate.

This alternative was requested by PNPS by letter dated October 29, 2003, and approved for the 4th Inservice Testing Ten-Year Interval as RP-05, Revision 0, on April 30, 2004 (NRC Letter from James W. Clifford, Office of Nuclear Reactor Regulation to Michael Kansler, Entergy Nuclear Operations, Dated April 30, 2004) (No TAC NO. was assigned). The proposed relief was approved pursuant to 1 OCFR50.55a(a)(3)(ii).

Changes to Applicable ASME Code Section Applicable Code and Addenda:

OM-2004 Edition through OMb-2006 Addenda

Applicable Code Requirement

ISTB 5300(a) Duration of Tests (1) "For the Group A test and the comprehensive test, after pump conditions are as stable as the system permits' each pump shall be run at least 2 min. At the end of this time, at least one measurement or determination of each of the quantities required by Table ISTB-3000-1 shall be made and recorded."

The specified Code requirement is unchanged for the ASME OM Code 1995 Edition, with the 1996 Addenda ISTB 5.6.3, comprehensive test requirement:

ISTB 5.6.3, Comprehensive A Test: "After pump conditions are as stable as the system permits' each pump shall be run at least 2 min. At the end of this time, at least one measurement or determination of each of the quantities required by table ISTB 4.1-1 shall be made and recorded."

Duration of Re-Approval for Proposed Alternative:

This relief is requested for the duration of the PNPS 5th Inservice Testing Ten-Year Interval (December 7th, 2012 through December 6th, 2022.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 109 of 174 7.2 VALVE TESTING PROGRAM COLD SHUTDOWN JUSTIFICATIONS Valve cold shutdown Justifications (CS) are provided for conditions where compliance to ASME OM Code, Subsection ISTC test requirements are satisfied but conditions exist that necessitate a test frequency of "cold shutdown" in lieu of "quarterly". Each justification identifies: valve(s) involved, compliance test requirement(s), basis for justification, and an alternate testing frequency of cold shutdown.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 110 of 174

[1] COLD SHUTDOWN JUSTIFICATION CS-01 System: Residual Heat Removal System (1001)

Valves: 47, 50 Cateqory: A Class: 1 Function:

The 47 and 50 valves open to provide the RHR shutdown cooling suction path. They also function as primary containment and pressure isolation valves.

Test Requirements:

ISTC-3521(c): If exercising is not practicable during operation at power, it may be limited to full-stroke exercising during cold shutdowns.

ISTC-3521(f): Valves full-stroke exercised at cold shutdowns shall be exercised during each cold shutdown, except as specified in paragraph ISTC-3521(g). Such exercise is not required if the time period since the previous full-stroke exercise is less than 3 months.

During extended shutdowns, valves that are required to perform their intended function shall be exercised every 3 months, if practicable.

Basis for Justification:

The valves are interlocked to prevent opening when Reactor pressure is greater than 70 psig. Each pressure isolation motor operated valve maintains one of the two high pressure barriers during plant operation. To exercise these valves during plant operation would involve a loss of one pressure isolation barrier between the high pressure Reactor coolant system and low pressure RHR system.

Alternate Testingq:

Exercise valves during cold shutdown.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 111 of 174

[2] COLD SHUTDOWN JUSTIFICATION CS-02 System: Nuclear Boiler - Main Steam, Vent, Drain, & Sampling System (220)

Valves: 46, 47 Catecqory: B Class: 1 Function:

These valves are used to vent the Reactor Vessel head and Main Steam Line "A" during startup. During power operation they function as Reactor pressure boundary valves.

Test Requirements:

ISTC-3521(c): If exercising is not practicable during operation at power, it may be limited to full-stroke exercising during cold shutdowns.

ISTC-3521(f): Valves full-stroke exercised at cold shutdowns shall be exercised during each cold shutdown, except as specified in paragraph ISTC-3521(g). Such exercise is not required if the time period since the previous full-stroke exercise is less than 3 months.

During extended shutdowns, valves that are required to perform their intended function shall be exercised every 3 months, if practicable.

Basis for Justification:

Exercising one of these valves during normal operation leaves the other valve as the only barrier between the Reactor Vessel and the Drywell sump. Any seat leakage through the closed valve could potentially pressurize the Drywell, which is an unacceptable risk for the sole purpose of testing a valve. Finally, PNPS operating Procedures prohibit operation of these valves during power operation.

Alternate Testingq:

Exercise valves during cold shutdown.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 112 of 174

[3] COLD SHUTDOWN JUSTIFICATION CS-03 System: Residual Heat Removal System (1001)

Core Spray System (1400)

High Pressure Coolant Injection System (2301)

Reactor Core Isolation Cooling System (1301)

Valves: 1001-29A, 1001-29B, 1400-25A, 1400-25B, 2301-8, 1301-49 Category: A Class: 1 Function:

These valves provide the pressure isolation function from the high pressure Reactor coolant system.

Test Requirements:

ISTC-3521(c): If exercising is not practicable during operation at power, it may be limited to full-stroke exercising during cold shutdowns.

ISTC-3521(f): Valves full-stroke exercised at cold shutdowns shall be exercised during each cold shutdown, except as specified in paragraph ISTC-3521(g). Such exercise is not required if the time period since the previous full-stroke exercise is less than 3 months.

During extended shutdowns, valves that are required to perform their intended function shall be exercised every 3 months, if practicable.

Basis for Justification:

These pressure isolation motor-operated valves (PIVs) maintain one of the two high to low pressure barriers during plant operation. The other pressure isolation barrier is a check valve. To exercise these valves during plant operation results in a loss of one isolation barrier. Exercising these motor-operated valves quarterly with the Reactor Vessel at pressure significantly increases the occurrence probabilities of an intersystem loss of coolant accident.

Alternate Testing:

Exercise valves during cold shutdown.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 113 of 174

[4] COLD SHUTDOWN JUSTIFICATION CS-04 System: Salt Service Water System (29)

Valves: 3801, 3805 Category: B Class: 3 Function:

These valves control cooling water flow for each of the Turbine Building Closed Cooling Water (TBCCW) heat exchangers and, therefore, the amount of cooling available for TBCCW equipment. The safety function of these valves is to close on LOCA signal in order to direct maximum cooling water flow to the RBCCW heat exchanger.

Test Requirements:

ISTC-3521(b): If full-stroke exercising during operation at power is not practicable, it may be limited to part-stroke during operation at power and full-stroke during cold shutdowns.

ISTC-3521(f): Valves full-stroke exercised at cold shutdowns shall be exercised during each cold shutdown, except as specified in paragraph ISTC-3521(g). Such exercise is not required if the time period since the previous full-stroke exercise is less than 3 months.

During extended shutdowns, valves that are required to perform their intended function shall be exercised every 3 months, if practicable.

Basis for Justification:

Since critical power generation equipment such as the turbine lube oil coolers, reactor feed pump coolers, and generator hydrogen and stator coolers are serviced by TBCCW, a complete stroke closure of the valves would cause an undesirable disruption of cooling in the system. This cooling disruption could lead to power generation equipment damage, possible plant forced power reduction and/or Scram (from the generator protective functions).

Alternate Testing:

Partial stroking of valves quarterly. Exercise valves during cold shutdown.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 114 of 174

[5] COLD SHUTDOWN JUSTIFICATION CS-05 System: Main Steam (203)

Valves: 1A, 1B, 1C, 1D, 2A, 2B, 2C, 2D Category: A Class: 1 Function:

The Main Steam Isolation Valves (MSIVs) must close to quickly terminate Reactor steam flow to the Turbine Building during specified plant conditions, and/or close to isolate the main steam line (primary containment) penetrations.

Test Requirements:

ISTC-3521(b): If full-stroke exercising during operation at power is not practicable, it may be limited to part-stroke during operation at power and full-stroke during cold shutdowns.

ISTC-3521(f): Valves full-stroke exercised at cold shutdowns shall be exercised during each cold shutdown, except as specified in paragraph ISTC-3521(g). Such exercise is not required if the time period since the previous full-stroke exercise is less than 3 months.

During extended shutdowns, valves that are required to perform their intended function shall be exercised every 3 months, if practicable.

Basis for Justification:

Full stroke testing [full stroke exercise (FE) and stroke timing (ST)] the MSIVs requires a reduction in power to approximately 60% or less in order to lower main steam line flow to acceptable levels for valve stroking. Attempting to full stroke an MSIV above 60% power may result in a plant trip as the MSIV goes closed due to a steam flow increase through the other (nontested) lines. Conducting power reductions for the purpose of performing MSIV quarterly exercising is costly and burdensome to the utility.

The MSIVs are designed to receive a partial stroke exercise at 100% power. Testing is accomplished by bleeding down the air from the actuator underpiston area (which holds this valve in the open position) and allowing the valve's springs to slowly move the poppet to the 10% closed (90% full open) position.

A roller arm limit switch assembly activates the MSIV closed position light for full stroke exercise (FE). This limit switch assembly was not designed to provide precise valve stem position. The PNPS fail-safe closed verification no longer relies upon the closed limit switch mechanism. This is because valve full stroke stem travel may not always be accurately indicated by the activation of the close limit switch mechanism (and light indication).

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 115 of 174

[5] COLD SHUTDOWN JUSTIFICATION CS-05 (CONTINUED)

Basis for Justification:

Performing MSIV full stroke testing on a strict quarterly frequency during power operation will require periodic power reductions for testing, which creates an undue hardship without a compensating increase in the level of quality and safety. Fail-safe "springs-only" testing the outboard MSIVs is initiated by push button manipulation from the Control Room. Each valve's air control system has a special test circuit that enables a closure test to be conducted without using air assist from the valve actuator (closing) over-piston. The Control Room push button actuates a two-way solenoid valve that slowly vents (through a bleedoff port) the valve actuator's underpiston (opening) air supply. Spring force and the (pre-established) air bleedoff rate determine the resulting MSIV closure speed.

NUREG-1 482, Guidelines for Inservice Testing at Nuclear Power Plants, Section 4, "Supplemental Guidance on Inservice Testing of Valves", provides guidance related to MSIV fail-safe closure verification. Section 4.2.4, "Main Steam Isolation Valves", endorses the General Electric Co. recommended practice (reference SIL477) for conducting a "springs only" full stroke closure test and states "by monitoring position indicators alone, the utility could not determine that the valve is fully closed" and that testing "would necessitate measurement of the actual valve stem travel because the final 10-percent of stem travel coincides with the weakest spring force." Because PNPS has adopted the NUREG 1482 recommended practice, fail-safe testing requires: 1) valve access for marking stem position,

2) closure of the MSIV by normal means (which allows full closed stem position marking),

and 3) "springs-only" closure verification using the field observed stem position.

To conduct periodic fail-safe testing to the outboard MSIVs (requiring access to the Steam Tunnel) during power operations is not consistent with PNPS ALARA practices. To perform the necessary preparations and obtain the required verifications during valve stroking requires personnel to spend an extended time in the Steam Tunnel during power operation and is beyond the reasonable task scope complexity utilized for quarterly valve stroke testing. Therefore, the outboard MSIVs (AO-203-2A through 2D) will be fail-safe tested at a cold shutdown frequency.

Access restrictions for fail-safe testing of the inboard MSIVs (AO-203-1A through 1D, located within primary containment) differ and are addressed within the appropriate Refuel Outage Justification.

Alternate Testingq:

Perform fail-safe testing of the outboard MSIVs during cold shutdowns.

Perform full stroke exercising and stroke timing of the MSIVs when practical during scheduled plant power reductions (not more frequently than quarterly testing) which are below 60% power. For those cases when full stroke testing cannot practically be conducted during a plant downpower, valve stroking will be attempted again during the next scheduled plant downpower.

Full stroke testing of the MSIVs will also be performed during cold shutdowns when the quarterly test interval has been exceeded.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 116 of 174 7.3 VALVE TESTING PROGRAM REFUEL OUTAGE JUSTIFICATIONS Valve Refuel Outage Justifications (RJ) are provided for conditions in which compliance to ASME OM Code, Subsection ISTC test requirements are satisfied but conditions exist that necessitate a test frequency of "Refueling Outage" in lieu of "quarterly or cold shutdown". Each justification identifies: valve(s) involved, test requirement(s) of compliance, basis for justification, alternate testing, and frequency of refuel interval.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 117 of 174

[1] REFUEL OUTAGE JUSTIFICATION RJ-01 System: Reactor Building Closed Cooling Water System (30)

Valves: 432 Cateqory: AC Class: 2 Function:

This valve provides isolation to non-seismic Drywell components that are supplied cooling water from RBCCW. These components require isolation for primary containment criteria and maintenance.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

This check valve is the outboard primary containment isolation valve for a system considered in-service during plant operation. The exercise open test is performed by verifying system flow during normal operation and may precede or follow the exercise close test.

Performance of a check valve closure test during plant operation would result in the isolation of "A" and "B" Loop Drywell Coolers and "A" and "B" Loop Recirculation Pump Seal Water and Motor Lube Oil Coolers. Isolation of those coolers associated with the recirculation pumps will lead to the accelerated degradation or failure of recirculation pump seals and may damage recirculation pump motors due to overheating. Additionally, isolation of the Drywell coolers during operation would impact the equipment environmental qualification requirements due to Drywell heatup.

Decay heat removal is a key safety function as defined by PNPS during cold shutdown. Due to this concern, PNPS Procedures often require that at least one recirculation pump remains in operation during cold shutdown to compensate for a potential loss of decay heat removal.

The required isolation to test valve 30-CK-432 would isolate cooling to both recirculation subsystems causing the same failure concerns addressed above. Additionally, isolation of Drywell coolers during cold shutdown will expose the Drywell to repeated unnecessary heatup with potential to adversely affect component operating life and habitability requirements.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 118 of 174

[1] REFUEL OUTAGE JUSTIFICATION RJ-01 (CONTINUED)

Basis for Justification:

Permanent plant-installed, nonintrusive test equipment does not exist for this valve. Other methods using portable equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which can be used to verify an open or close test on this valve require valve closure for extended periods and, as previously stated, are not practical for quarterly or cold shutdown testing.

Seat leak testing (normally performed during a refuel outage) in accordance with 10CFR50 Appendix J ensures the close test and establishes plant conditions to avoid recirculation pump degradation and unnecessary Drywell heatup. During refueling intervals when Appendix J leak testing will not be conducted, the close test will be performed using other methods. The alternate methods will either utilize: 1) leak testing using water to verify valve closure or 2) specialized testing that manipulates system flow and uses portable nonintrusive test equipment to document valve closure.

Alternate Testinq:

Perform the exercise close test of this check valve during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 119 of 174

[2] REFUEL OUTAGE JUSTIFICATION RJ-02 System: Nuclear Boiler - Feedwater System (6)

Valves: 58A, 58B, 62A, 62B Category: AC Class: 2 Function:

These valves serve as feedwater inlet check valves.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These check valves are the inboard and outboard primary containment isolation valves for a system considered in-service during plant operation. Verification of the closure of these check valves during plant operation would require isolation of all feedwater flow to the vessel (the feedwater trains are crosstied). Such an evolution creates an adverse operating condition which would cause automatic plant shutdown due to loss of feedwater flow.

Performing an exercise close test during cold shutdowns would create prolonged periods of system testing and unreasonable increases in radiation exposure. Permanent plant-installed, nonintrusive test equipment does not exist for verifying disc position of these valves. Other methods using portable test equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which may be used in performing an exercise close test for these check valves during a plant cold shutdown have been reviewed and determined not to be practical.

Portable methods require special test equipment to be installed and benchmarked prior to gathering final test data. This benchmarking also necessitates repeated feedwater system startup and shutdown operations which require excessive system manipulation during plant cold shutdown conditions. These alternate testing methods either require extended system testing to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for cold shutdown testing. Therefore, the only practical method available to verify check valve closure is during the performance of leak testing.

Verifying a close test more frequently than that used for seat leak testing would create a hardship without a compensating increase in the level of safety. Leak testing (in accordance with 10CFR50 Appendix J) ensures the valve is in the closed position.

The exercise open test is performed by verifying system feedwater flow during plant operation and may precede or follow the exercise close test (using a leak test).

Alternate Testing:

Perform the exercise close test of these check valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 120 of 174

[3] REFUEL OUTAGE JUSTIFICATION RJ-03 System: Recirculation Pump Seal Water System (262)

Valves: F013A, F013B, F017A, F017B Category: AC Class: 2 Function:

These valves serve as the recirculation pump seal water inboard and outboard containment isolation check valves.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These check valves are the primary containment isolation valves for a system considered in-service during plant operation. The open function of these valves is non-safety related. The exercise open test is performed by verifying system flow during normal system operation and may precede or follow the exercise close test.

The performance of an exercise close test during plant operation would require isolation of recirculation pump seal purge lines. Operation of the recirculation pumps without seal purge has been shown to lead to accelerated degradation of pump seals. Additionally, leak testing to verify valve closure would require opening a connection on the seal purge line outside of the Drywell. This may introduce air into the seal purge system requiring venting to avoid rapid degradation of recirculation pump seals. Venting of the seal purge line following testing would require Drywell access which is only accessible during cold shutdown.

The required isolation to test these check valves during plant cold shutdowns would result in the isolation of the recirculation seal purge lines, causing the same failure concerns addressed above. Decay heat removal is a key safety function as defined by PNPS during cold shutdown. Due to this concern, PNPS Procedures often require that at least one recirculation pump remains in operation during cold shutdown to compensate for a potential loss of decay heat removal. Permanent, plant-installed, nonintrusive test equipment does not exist for verifying disc position of this valve. Other methods using portable equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which may be used in testing those check valves associated with the secured recirculation pump have been reviewed for use in performing a close test and have been determined to be not practical. Portable methods require special test equipment to be installed and benchmarked prior to gathering final test data. These alternate testing methods either require extended system outages to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for cold shutdown testing.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 121 of 174

[3] REFUEL OUTAGE JUSTIFICATION RJ-03 (CONTINUED)

Basis for Justification:

Primary containment leak testing or closure verification using a water leak test method each refueling interval constitutes the most prudent method for the exercise close test. The local leak rate tests require system draining and/or venting with entrance into the plant Drywell.

Therefore, verifying a close test more frequently than that used for seat leak testing would create a hardship without a compensating increase in the level of safety. Leak testing verifies the valve exercise close test.

Alternate Testinq:

Perform the exercise close test of these check valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 122 of 174

[4] REFUEL OUTAGE JUSTIFICATION RJ-04 Deleted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 123 of 174

[5] REFUEL OUTAGE JUSTIFICATION RJ-05 System: High Pressure Coolant Injection (23)

Valves: 232, 233 Catecqory: C Class: 2 Function:

These valves provide a vent path to relieve the vacuum which is created within the HPCI exhaust line after turbine operation. These valves close to prevent HPCI exhaust steam from bypassing the exhaust steam line submerged pipe header (which provides suppression for the exhaust steam).

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These normally closed valves are exercise open tested by performing a specialized flow test.

The exercise close test of each valve is verified concurrently with the exercise open test by performing an air leakage test.

Testing these valves at power requires isolation of the HPCI turbine for safety reasons. This requires HPCI to be inoperable during preparation, testing, and restoration of this system.

To conduct the exercise open test, an air test flow instrument with a pressure gauge is connected to the vacuum breaker system using a test hose and a system vent path is established. Then the desired air flow rate is achieved through these check valves while the test cart air makeup pressure is monitored. The test is satisfactory when a specified rate of flow is established without exceeding a maximum backpressure. To perform the exercise close test of this valve, a leak rate monitor with an installed pressure gauge is connected to the vacuum breaker system using a test hose and a system vent path is established. Two separate air leak rate tests are conducted to ensure that each valve returns to its closed position. These tests are satisfactory when a specified leakage rate which is indicative of full valve closure is achieved.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 124 of 174

[5] REFUEL OUTAGE JUSTIFICATION RJ-05 (CONTINUED)

Basis for Justification:

Performing an exercise open test during power operations or cold shutdowns is impractical because it is intrusive and renders the HPCI System inoperable for an extended period of time. To verify this valve open using flow, it is necessary to take the HPCI System out of service, set up special test equipment, establish a vent path, and perform an exercise open test. The system must be then realigned, leak rate test equipment installed, two leak rate tests conducted to verify valve closure, test equipment removed, and the system restored to operable status. Even during cold shutdowns, to create the system conditions and conduct testing for these check valves, the HPCI System must be made inoperable for an extended period of time such that it could delay returning the plant to power.

Permanent, plant-installed, nonintrusive test equipment does not exist for this valve. Other methods using portable equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which could be used in verifying an open test would still require a temporary test circuit to establish the conditions for conducting an exercise open test. Also, because this is a vacuum relief line (and contains air), ultrasonic methods for verifying valve movement will not work.

All methods mentioned above for conducting an exercise open test either require extended system outages to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for cold shutdown testing.

Alternate Testingq:

Perform the exercise test (both open and close tests) during a refueling outage (not to exceed a refueling interval) and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 125 of 174

[6] REFUEL OUTAGE JUSTIFICATION RJ-06 Deleted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 126 of 174

[7] REFUEL OUTAGE JUSTIFICATION RJ-07 System: High Pressure Coolant Injection (2301)

Valves: 64 Category: B Class: 2 Function:

This valve opens to create a flow path that enables condensate from the HPCI Gland Seal Condenser (GSC) to be pumped to Clean Radwaste when the HPCI System is in standby mode. This valve isolates while HPCI is in operation to prevent the HPCI GSC pump discharge piping from communicating with Clean Radwaste.

Test Requirements:

ISTC-3521 (e): If exercising is not practicable during operation at power or cold shutdowns, it may be limited to full-stroke during refueling outages.

ISTC-3521(h): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

This air-operated valve opens to create a flow path that enables condensate from the HPCI Gland Seal Condenser (GSC) to be pumped to Clean Radwaste when the HPCI System is in standby mode. This valve isolates while HPCI is in operation to prevent the HPCI GSC pump discharge piping from communicating with Clean Radwaste.

This normally closed valve cannot be stroked when either of the following isolating conditions is present: 1) isolation signal when the HPCI System is initiated either manually or through an automatic signal or 2) isolating condition when HPCI GSC level is below the HPCI GSC pump low level permissive setpoint. If either of the above conditions occurs, then this valve will not respond to handswitch operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 127 of 174

[7] REFUEL OUTAGE JUSTIFICATION RJ-07 (CONTINUED)

Basis for Justification:

On some occasions immediately following a HPCI pump run, there is sufficient level in the GSC to permit valve stroke testing. However, if the GSC pump begins to pump down within 2 to 3 minutes before the HPCI pump is tripped or while the HPCI System is being secured following pump trip, then the GSC low level permissive will not be clear and the valve cannot be stroked using the control switch. In this case, valve stroking may not be possible for extended periods of time. Restarting HPCI to obtain another chance for valve stroking would place undue wear and tear on the HPCI turbine/pump system and would be burdensome.

A review of plant on-line computer data for the GSC system shows that when the HPCI System is in the standby mode a low level permissive may not clear for weeks or months to allow stroking the AO-2301-64 valve. This is due to zero or very low steam leakage past the HPCI steam admission valve (a new design valve was installed during RFO #10 which greatly diminished steam leakage).

Testing this valve on a quarterly basis would cause an undue hardship. To conduct a full stroke exercise (FE) and stroke time (ST) consistently on a quarterly basis, one of the following measures will need to be implemented: 1) water must be manually introduced into the HPCI GSC shell side by installing a temporary hose that supplies water from either the HPCI pump suction (using CST static head pressure) or the plant demineralized water system or 2) the low GSC water level permissive that prevents stoking this valve using the control switch must be defeated by installing an electrical jumper.

Stroking the AO-2301-64 valve on a strict quarterly frequency during power operation will require conduct of one of the above listed special provisions and renders the HPCI System inoperable for an extended period of time (several hours). Even during cold shutdowns, to create the system conditions and conduct testing of the air-operated valve, the HPCI System must be made inoperable for an extended period of time such that it could delay returning the plant to power. To create the conditions that will allow stroke testing quarterly or during cold shutdown is burdensome and creates an undue hardship without a compensating increase in the level of quality and safety.

Alternate Testing:

Conduct valve stroking on a refueling interval frequency and complete testing prior to returning the plant to power operation.

PNPS will administratively attempt to perform stroke timing for this valve following each quarterly IST HPCI pump test.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 128 of 174

[8] REFUEL OUTAGE JUSTIFICATION RJ-08 System: As Applicable Valves: Containment Isolation Check valves requiring exercise close test as listed within this justification.

Cateqory: AC Class: 1, 2, and 3 Function:

These valves are required to return to their normally closed position to perform their function or minimize seat leakage for containment isolation purposes.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These check valves are required to return to their normal closed position to perform their function and/or limit seat leakage for primary containment isolation. Obtaining a satisfactory seat leakage test (normally performed during a refuel outage) ensures that each valve's obturator has returned to the normal closed position. Attempting to verify their closed position by "other positive means" or by performing seat leakage tests on a quarterly or cold shutdown basis is not practical and would place undue hardship on the plant.

Permanent, plant-installed, nonintrusive test equipment does not exist for verifying disc position for any of these listed valves. Other methods using portable equipment (e.g.,

acoustic, ultrasonic, magnetic, and radiography) which may be used in performing a closure verification on a quarterly or cold shutdown frequency have been reviewed and determined to be not practical. Portable methods require special test equipment to be installed and benchmarked prior to gathering final test data. These alternate testing methods either require extended system outages to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for quarterly or cold shutdown testing.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 129 of 174

[8] REFUEL OUTAGE JUSTIFICATION RJ-08 (CONTINUED)

Basis for Justification:

Seat leak testing (normally performed during a refuel outage) in accordance with 10CFR50 Appendix J will be used to verify the normal closed position of these valves. During refueling intervals when Appendix J leak testing will not be conducted, the closure verification will be performed using other methods. The alternate methods will either utilize: 1) leak testing using water or air to verify valve closure or 2) specialized testing that manipulates system flow and uses portable nonintrusive test equipment to document valve closure. These valves are tabulated as part of this justification with a brief explanation of the unique hardship or impracticality of testing.

Alternate Testing:

Perform the exercise close test of these check valves during a refueling outage (not to exceed a refueling interval) and complete testing prior to returning the plant to operation.

Valve No.(s) Description 9-CK-340 These valves are the inboard containment isolation valves (Type C tested) 9-CK-341 for the Drywell and Torus Nitrogen Makeup System. This makeup system is part of the Primary Containment Atmospheric Control System and ensures a source of nitrogen to maintain the inerted condition of containment. These valves are exercised periodically in the open direction. The most practical method of the exercise close test is during an Appendix J seat leakage test or simplified closure test which also measures seat leakage. These tests require isolating the line to this penetration (which would make Containment Atmospheric Dilution System components unavailable) and installing test equipment. Therefore, testing to verify normal closed position more frequently than once each refueling interval results in a hardship and is not practicable due to an increase in personnel radiation exposure, loss of system availability, and manpower constraints.

31 -CK-434 This is an inboard containment isolation valve (Type C tested) in the air/nitrogen line which supplies a remote air-operated exerciser for exercising the Torus to Reactor Building Vacuum Breakers. This check valve is opened quarterly during vacuum breaker testing. The only practical method of the exercise close test is during an Appendix J seat leakage or simplified closure test which also measures seat leakage. These tests require climbing to the top of the Torus to allow valve lineup and test equipment installation. This location is a high radiation area and difficult to access. Testing to verify closed position more frequently than once each refueling interval results in a hardship and is not practicable because of the increase in personnel radiation exposure and the safety risks associated with working on top of the Torus.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 130 of 174

[8] REFUEL OUTAGE JUSTIFICATION RJ-08 (CONTINUED)

Alternate Testing:

Valve No.(s) Description 31-CK-167 The exercise close test can only be performed by entering the containment (Drywell) environment, isolating components supplied by Drywell instrument air, and performing a closure test using leak test methods.

Drywell components which will be isolated include: inboard MSIVs and Main Steam Relief Valve Accumulators, equipment and floor drain sump instruments, and Drywell air-operated valves (i.e., Reactor head vent, Reactor sampling, and Reactor seal drain). To conduct an exercise close test other than during a refueling interval would result in a hardship and is not practicable due to excessive Drywell component isolation and an increase in personal radiation exposure CK-2301-45 These valves are containment isolation valves (Type C tested) between the CK-2301-74 HPCI turbine exhaust header and the suppression pool (Torus). The most practical method available to verify each valve's exercise close test is to perform an Appendix J seat leakage test or closure verification using leak test methods. This testing requires erection of scaffolding, isolation of the penetration, and installation of test equipment. Verifying the closed position more frequently than once each refueling interval would be a hardship and is not practicable due to increased personnel radiation exposure and prolonged periods of safety system inoperability.

CK-1301-41

These valves are water-tested containment isolation valves for the Core CK-1301-59 and Suppression Pool Cooling and Reactor Vessel Level Makeup systems.

CK-1301-64 These valves require testing to assure that the seal-water (Torus) fluid CK-2301-34* inventory is sufficient to maintain the sealing function for at least 30 days.

CK-2301-217 The most practical method available to verify each valve's exercise close test is to perform an Appendix J seat leakage test or closure verification using water leak test methods. Testing requires isolating the penetration, in some cases the erection of scaffolding, and installation of test equipment.

Testing often requires personnel to enter high radiation areas. Verifying the closed position more frequently than once each refueling interval would be a hardship without any compensating increase in the level of safety due to increased personnel radiation exposure and prolonged periods of safety system inoperability. The asterisked (*) valves are not required to be designated category "AC", but their exercise close test is normally performed (using leak test methods) in conjunction with the other valve seat leakage tests.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 131 of 174

[9] REFUEL OUTAGE JUSTIFICATION RJ-09 System: High Pressure Coolant Injection System (2301)

Valves: 34, 217 Cateqory: C and AC Class: 2 Function:

These check valves are the HPCI Turbine Exhaust Drain Line Check Valves. They open to allow the Turbine exhaust drain system condensate to be directed via steam traps to the Torus. These valves close to prevent Torus water backflow during Torus accident pressurization when HPCI is not in operation.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These normally closed check valves (installed in series) are exercise open tested by the performance of a specialized flow test. The exercise close test is verified once per refueling interval by using seat leakage test methods.

Performing this test at power requires isolation of the HPCI turbine for personnel safety reasons. This requires HPCI to be inoperable during preparation, testing, and restoration of this system.

To conduct the exercise open test, a water flow rate monitor with an installed test gauge is connected to the HPCI Exhaust Drain Line using a test hose. The exercise open test is achieved by observing flow through both check valves to the Torus.

Performing an exercise open test during power operation or cold shutdowns is impractical because it is intrusive and renders the HPCI System inoperable for an extended period of time. To verify these valves open using flow, it is necessary to take the HPCI System out of service, set up a flow meter with a hydraulic test supply, install the test flow meter, perform an exercise open test and restore the system to operable status. Even during cold shutdowns, to create the system conditions and conduct testing for this check valve exercise, the HPCI System must be made inoperable for an extended period of time (approximately 10 to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) such that it could delay returning the plant to power.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 132 of 174

[9] REFUEL OUTAGE JUSTIFICATION RJ-09 (CONTINUED)

Basis for Justification:

Permanent, plant-installed, nonintrusive test equipment does not exist for these valves.

Other methods using portable equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which could be used in verifying an open test would still require a temporary test circuit to establish the conditions for conducting an open test. Also, because this is a steam exhaust drain line, it is difficult to ensure the drain piping is completely free of air voids during testing, which will render some methods unreliable.

All methods mentioned above for conducting an exercise open test require extended system outages to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for cold shutdown testing.

Alternate Testing:

Perform the exercise open test of these check valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 133 of 174

[10] REFUEL OUTAGE JUSTIFICATION RJ-10 Deleted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 134 of 174

[11] REFUEL OUTAGE JUSTIFICATION RJ-1 1 System: Main Steam (203)

Valves: 1A, 1B, 1C, 1D Category: A Class: 1 Function:

The Inboard Main Steam Isolation Valves (MSIVs) are used to isolate the main steam line penetrations.

Test Requirements:

ISTC-3560, Fail-Safe Valves. Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of paragraph ISTC-3510.

Basis for Justification:

Fail-safe "springs-only" testing of the inboard MSIVs is initiated by push button (switch) manipulation from the Control Room. Each valve's air control system has a special test circuit that enables a closure test to be conducted without using air assist from the valve's over-piston (closing) actuator. The Control Room push button actuates a two-way solenoid valve that slowly vents (through a bleedoff port) the valve actuator's underpiston (opening) air supply. Spring force and the (pre-established) air bleedoff rate determine the resulting MSIV closure speed.

A roller arm limit switch assembly activates the MSIV closed position light for full stroke exercise (FE). This limit switch assembly was not designed to provide precise valve stem position. The PNPS fail-safe closed verification no longer relies upon the closed limit switch mechanism. This is because valve full stroke stem travel may not always be accurately indicated by the activation of the close limit switch mechanism (and light indication).

NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, Revision 1, Section 4, "Supplemental Guidance on Inservice Testing of Valves", provides guidance related to MSIV fail-safe closure verification. Section 4.2.6, "Main Steam Isolation Valves", endorses the General Electric Co. recommended practice (reference SIL477) for conducting a "springs only" full stroke closure test and states "by monitoring position indicators alone, the utility could not determine that the valve is fully closed" and that testing "would necessitate measurement of the actual valve stem travel because the final 10-percent of stem travel coincides with the weakest spring force." Because PNPS has adopted the NUREG 1482 recommended practice, fail-safe testing requires: 1) valve access for marking stem position,

2) closure of the MSIV by normal means (which allows full closed stem position marking),

and 3) "springs-only" closure verification using the field observed stem position.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 135 of 174

[11] REFUEL OUTAGE JUSTIFICATION RJ-11 (CONTINUED)

Basis for Justification:

The inboard MSIVs are located in the Drywell (primary containment) and are not accessible during normal plant operations. The Drywell is normally inerted with nitrogen during power operations and often remains de-inerted during plant cold shutdowns. The de-inerted Drywell is considered a hazardous environment from a personnel safety perspective.

Therefore, the inboard MSIVs (AO-203-1A through 1 D) will be fail-safe tested at a cold shutdown frequency when the Drywell is de-inerted.

Alternate TestinQ:

Perform fail-safe tests of the inboard MSIVs during extended cold shutdowns (when Drywell is de-inerted) or during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 136 of 174

[12] REFUEL OUTAGE JUSTIFICATION RJ-12 System: Reactor Core Isolation Cooling System (1301)

Valves: 59 Catecqory: C Class: 2 Function:

This valve is the inboard RCIC Vacuum Pump Discharge Check Valve. The valve opens to allow the vacuum pump to discharge air and noncondensable gases from the RCIC (Barometric Condenser) Vacuum Tank to the Torus. This valve closes to prevent Torus water backflow into the RCIC Vacuum Tank during Torus (accident conditions) pressurization.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

This normally closed valve must be exercise open tested by performing a specialized flow test and the exercise close test is conducted by performing an air leakage test.

This testing during normal operation or cold shutdown would cause undue hardship.

Performing these tests at power would require isolation of the RCIC turbine for safety reasons. This will require RCIC to be inoperable during preparation, testing, and restoration of this system.

To conduct the exercise open test, an air test flow instrument with a pressure gauge is connected to the vacuum pump exhaust line using a test hose and a system vent path to the Torus is verified. Then the desired air flow rate is achieved through the check valve while the test cart air makeup pressure is monitored. The test is satisfactory when a specified rate of flow is established without exceeding a maximum backpressure.

Performing an exercise open test during power operations or cold shutdowns is impractical because it is intrusive and renders the RCIC System inoperable for an extended period of time. To verify this valve open using flow, it is necessary to take the RCIC System out of service, set up special test equipment, establish a vent path, and perform exercise open testing. Then the system must be realigned, test equipment removed, and the system restored to operable status. Even during cold shutdowns, to create the system conditions and conduct testing for this check valve exercise open test, the RCIC System must be made inoperable for an extended period of time such that it could delay returning the plant to power.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 137 of 174

[12] REFUEL OUTAGE JUSTIFICATION RJ-12 (CONTINUED)

Basis for Justification:

Permanent, plant-installed, nonintrusive test (NIT) equipment does not exist for this valve.

Other methods using portable equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which could be used in verifying an open test would still require a temporary test circuit to establish the conditions for conducting an open test. To obtain access to this valve to install test equipment for NIT requires scaffolding to be built. Also, because this is a vacuum pump exhaust line (contains air and noncondensables), ultrasonic methods for verifying valve movement will not work.

All methods mentioned above for conducting an exercise open test either require extended system outages to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for cold shutdown testing.

Alternate Testing:

Perform the exercise open test of this check valve during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 138 of 174

[13] REFUEL OUTAGE JUSTIFICATION RJ-13 Deleted

[14] REFUEL OUTAGE JUSTIFICATION RJ-14 Deleted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 139 of 174

[15] REFUEL OUTAGE JUSTIFICATION RJ-15 System: Standby Liquid Control System (1101)

Valves: 15 Category: AC Class: 1 Function:

This valve is the Inboard SLC injection check valve.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

To verify forward flow during normal operation or cold shutdown would require firing a squib valve and injecting water into the Reactor Vessel using the SLC pumps. Injecting water during operation could result in adverse plant conditions such as changes in reactivity, power transient, thermal shock-induced cracking, and possible plant trip. Injecting water during a cold shutdown can result in cyclical thermal shock-induced cracking as cold water enters the Reactor Vessel, which is at an elevated temperature due to decay heat.

Injection of the SLC System during cold shutdowns using demineralized water requires a lengthy flushing operation to remove boron from the system. Even after extensive flushing, boron remains present in system dead legs. Some of this boron propagates into the Reactor Vessel during injection and is difficult to remove from the system. The presence of boron in Reactor water impedes the plant ability to achieve criticality for plant startup.

Verify the exercise open test during refueling while performing the Standby Liquid Control System injection test, which requires pumping demineralized water into the Reactor Vessel after firing at least one squib valve.

Alternate Testing:

Perform the check valve exercise open test during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 140 of 174

[16] REFUEL OUTAGE JUSTIFICATION RJ-16 System: Reactor Water Cleanup System (1,201)

Valves: 81 Catecqory: C Class: 1 Function:

This valve provides the path for RWCU return flow to the Reactor Vessel.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

The Reactor Water Cleanup return check valve shall be exercise close tested. This normally open check valve is verified in the closed direction by isolating the RWCU return header, pressurizing the feedwater header, and venting the piping on the upstream side of this check valve to verify restricted flow. Controlled pressurization of the feedwater header to perform this test occurs during hydrodynamic leakage testing of the RCIC pressure isolation valves each refueling interval.

During cold shutdown conditions, this system is required to provide functional operation in order to control Reactor water level and primary system chemistry. Pressurization of the feedwater header for hydrodynamic leak testing requires extensive valve alignment to the feedwater system and renders both the feedwater and RWCU Systems inoperable for an extended period of time.

Alternate Testing:

Perform the exercise close test of this check valve during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 141 of 174

[17] REFUEL OUTAGE JUSTIFICATION RJ-17 System: Control Rod Drive Hydraulic System (302)

Valves: 21A, 21B, 22A, 22B, 23A, 23B, 24A, 24B Category: A Class: 2 Function:

These valves close to isolate Reactor coolant flow past the control rod drive seals during a Scram condition. They open to drain the Scram Discharge Volume Tank and allow a Scram condition to be reset.

Test Requirements:

ISTC-3521(d): If exercising is not practicable during operation at power and full-stroke during cold shutdowns is also not practicable, it may be limited to part-stroke during cold shutdowns and full-stroke during refueling outages.

ISTC-3521(e): If exercising is not practicable during operation at power or cold shutdowns, it may be limited to full-stroke during refueling outages.

ISTC-3521(h): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These air-operated valves are stroked closed quarterly using a separate testing air vent (bleed) circuit. Stroke times using the test circuit can only verify the valve's closed exercise since the closing stroke times are very erratic. Timing of the valve's close stroke using the test circuit quarterly does not provide representative or reliable data for tracking valve performance. The only method available to close stroke time these valves is by initiating a Reactor Scram. Trending these valves in the closed direction by Scramming the Reactor quarterly is impractical. Valve open stroke times are measured and trended quarterly to monitor for valve degradation.

Once per refueling outage, in accordance with Technical Specifications, a full Reactor Scram is initiated which utilizes the normal vent circuit for these valves. Since the Technical Specifications surveillance requirement is being satisfied, then the test frequency assures that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits, and that the limiting condition for operation will be met.

Alternate Testing:

Measure close stroke times by inserting a full Reactor Scram in accordance with Technical Specifications during a refueling outage not to exceed a refueling interval.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 142 of 174

[18] REFUEL OUTAGE JUSTIFICATION RJ-18 System: Reactor Building Closed Cooling Water System (30)

Valves: 4009A, 4009B, 4002 Catecqory: A, B Class: 2, 3 Function:

These valves provide isolation to Drywell components cooled by RBCCW. The components would require isolation for primary containment criteria and maintenance.

Test Requirements:

ISTC-3521(d): If exercising is not practicable during operation at power and full-stroke during cold shutdowns is also not practicable, it may be limited to part-stroke during cold shutdowns and full-stroke during refueling outages.

ISTC-3521(e): If exercising is not practicable during operation at power or cold shutdowns, it may be limited to full-stroke during refueling outages.

ISTC-3521(h): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

The testing of these valves requires isolation of the following components: Drywell Area Coolers, Reactor Recirculation Pump Seal Coolers, Reactor Recirculation Pump Lube Oil Coolers. Additionally, for testing 4009A and 4009B, the Reactor Water Cleanup (RWCU) nonregenerative heat exchanger, B Fuel Pool Cooling Heat Exchanger, RWCU Pump Cooling System Coolers, Control Rod Drive (CRD) Pump Area Cooling, and CRD Pump Thrust Bearing Coolers must also be isolated. The listed components supply numerous plant systems required for safe plant operation. The recirculation pumps and Drywell coolers may be required to support the plant during cold shutdown conditions to prevent water stratification in the vicinity of Reactor Vessel lower head and overheating of Drywell components.

Exercising these valves quarterly during power operation is impractical because the resulting flow interruption could cause equipment damage. It also is impractical to exercise these valves during cold shutdown when Drywell cooling loads are high or when a Reactor recirculation pump is operating. Stopping of Reactor recirculation pumps during each cold shutdown to allow exercising these valves could result in extending the cold shutdown, which would be costly and burdensome to the plant.

Alternate Testing:

Exercise valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 143 of 174

[19] REFUEL OUTAGE JUSTIFICATION RJ-19 System: Nuclear Boiler - Reactor Recirculation System (202)

Valves: 5A, 5B Category: B Class: 1 Function:

These valves are the Reactor recirculation pump discharge valves and function to close upon LOCA signals. (This is the remaining function of LPCI Loop Selection Logic.)

Test Requirements:

ISTC-3521(d): If exercising is not practicable during operation at power and full-stroke during cold shutdowns is also not practicable, it may be limited to part-stroke during cold shutdowns and full-stroke during refueling outages.

ISTC-3521 (e): If exercising is not practicable during operation at power or cold shutdowns, it may be limited to full-stroke during refueling outages.

ISTC-3521(h): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

Closure of these valves during normal operation will result in loss of forced circulation to the Reactor, prohibited by PNPS license.

Closure of these valves during cold shutdown necessitates securing operation of the Reactor recirculation pumps. This is detrimental because even though the moderator temperature is less than 212 0 F, the recirculating system is usually kept in operation during cold shutdown to provide Reactor coolant mixing to prevent Reactor Vessel temperature stratification. The Reactor Vessel temperature profile takes on an increasing temperature gradient between the bottom vessel head and the shutdown core when mixing (forced circulation) is stopped.

Additionally, the water in the idle recirculation loops cools down. This stratification can have the following adverse effects: Reactor Vessel temperatures become greater between the vessel bottom and top resulting in unnecessary thermal cycling, startup of the shutdown recirculation pump can cause a cold water intrusion affecting Reactor Vessel metal temperatures. Deliberate stopping and starting of the recirculation pumps 1) creates unnecessary cycling wear on major equipment important to plant reliability and 2) could result in extending the shutdown, both of which would be costly and burdensome to the plant. Therefore, performing this testing on a cold shutdown or quarterly frequency is not practicable.

Alternate Testing:

Exercise valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 144 of 174

[20] REFUEL OUTAGE JUSTIFICATION RJ-20 System: Reactor Building Closed Cooling Water (30)

Valves: 4085A, 4085B Cateqory: B Class: 3 Function:

Valves 4085A and 4085B are RBCCW Loop A Isolation valves in supply lines to non-safety related components.

Test Requirements:

ISTC-3521(d): If exercising is not practicable during operation at power and full-stroke during cold shutdowns is also not practicable, it may be limited to part-stroke during cold shutdowns and full-stroke during refueling outages.

ISTC-3521(e): If exercising is not practicable during operation at power or cold shutdowns, it may be limited to full-stroke during refueling outages.

ISTC-3521(h): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

Valves 4085A and 4085B are the non-safety related component isolation valves for RBCCW Loop A. Components cooled by this RBCCW branch include the Reactor recirculation pump motor-generator set fluid coupling oil and bearing coolers.

Stroke testing quarterly during power operation could result in loss of cooling to the recirculation pump motor-generator set fluid coupling oil and bearing coolers with consequent loss of forced circulation to the Reactor, requiring plant shutdown.

Stroke testing at cold shutdown could result in loss of the recirculation pump operation due to interruption of cooling to the recirculation pump motor-generator set fluid coupling oil and bearing coolers. This is detrimental because even though the moderator temperature is less than 212 0 F, the recirculation system is kept in operation during cold shutdown to provide mixing of the Reactor coolant to prevent Reactor Vessel temperature stratification.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 145 of 174

[20] REFUEL OUTAGE JUSTIFICATION RJ-20 (CONTINUED)

Basis for Justification:

The Reactor Vessel temperature profile takes on an increasing temperature gradient between the bottom vessel head and the core when mixing (forced circulation) is stopped, plus the water in the idle recirculation loops cools down. This stratification can have the following adverse effect: Reactor Vessel metal temperature differences become greater between Reactor Vessel bottom and top resulting in unwanted thermal cycling. Startup of the shutdown recirculation pumps causes a cold water intrusion which affects Reactor Vessel metal temperatures and causes thermal cycling of the Reactor Vessel.

Alternate Testinq:

Exercise valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 146 of 174

[211 REFUEL OUTAGE JUSTIFICATION RJ-21 System: As Applicable Valves: Check valves, except Containment Isolation Check Valves, requiring seat leakage measurement for the performance of the closed test are as listed within this justification.

Catecqory: AC Class: 1 and 2 Function:

These valves are required to return to their normally closed position to minimize seat leakage.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These check valves are required to return to their normal closed position to perform a safety function and limit seat leakage to a specific amount. Conducting a satisfactory seat leakage measurement assures the valve's obturator has returned to the normal closed position.

Permanent, plant-installed, nonintrusive test equipment does not exist for verifying disc position for any of these listed valves. Other methods using portable equipment (e.g.,

acoustic, ultrasonic, magnetic, and radiography) which may be used in performing a closure verification on a quarterly or cold shutdown frequency have been reviewed and determined to be not practical. Portable methods require special test equipment to be installed and benchmarked prior to gathering final test data. These alternate testing methods either require extended system outages to perform and/or cannot consistently provide accurate reliable results when utilized within the time constraints necessary for quarterly or cold shutdown testing.

Therefore, attempting to verify the closed position by other positive means or by performance of seat leakage tests on a quarterly or cold shutdown basis is not practical and would place undue hardship on the plant. These valves are tabulated as part of this relief with a brief explanation of the unique hardship or impracticality of testing.

Alternate Testing:

Perform the exercise close test of these check valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 147 of 174

[21] REFUEL OUTAGE JUSTIFICATION RJ-21 (CONTINUED)

Alternate Testing:

Valve No.(s) Description CK-1001-68A These valves are part of the Reactor coolant pressure boundary for the CK-1001-68B Core and Suppression Pool Cooling and Reactor Vessel Level Makeup systems. Each valve performs a pressure isolation function between the Reactor coolant pressure boundary and the associated low pressure system. A seat leakage test is the only practical method for verifying the normally closed position. Testing these valves requires entry into the Drywell (e.g., Drywell must be de-inerted) for system isolation, installation of test equipment, and sometimes partial draindown of the system. Therefore, verifying the exercise close more frequently than the interval used for seat leakage testing would create a hardship and is not practicable due to an increase in personnel radiation exposure and prolonged periods of safety system inoperability.

CK-1 101-15 This valve is part of the Reactor coolant pressure boundary for the Standby Liquid (reactivity) Control (SLC) System. This check valve maintains the Reactor coolant pressure boundary for this high pressure system. A seat leakage test is the only practical method for verifying the normally closed position. Since there is no test connection upstream from this valve to enable pressurization, the closed position can only be practically verified by performing a leakage test during the Class 1 Reactor Pressure Vessel System Leakage Test. This system leakage test is only conducted near the end of each refueling outage. Because this valve is located inside the Drywell and conducting an exercise requires initiating the SLC System and injecting into the Reactor Vessel nonintrusive test methods for verifying valve closure are not practical.

Therefore, verifying the exercise close test more frequently than the interval used for leakage testing would create a hardship without a compensating increase in the level of safety.

CK-1001-362B These valves are the seismic boundary isolation between the Core Spray CK-1001-363A or LPCI suppression pool cooling systems and their keepfill supplies. Loss of CK-1400-212A keepfill for maintenance or testing creates a potential for air intrusion which CK-1400-212B would jeopardize the affected cooling system's operability. The only reasonable method to verify the valves' exercise close is by performing a seat leakage test. Leak testing of these valves requires isolation of the keepfill supply and entry into a radiation area to install test equipment and conduct testing. Therefore, verifying the closed position more frequently than the interval used for the seat leak testing would be a hardship and is not practicable due to increased personnel radiation exposure and prolonging the duration of a safety system's inoperability.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 148 of 174

[22] REFUEL OUTAGE JUSTIFICATION RJ-22 Deleted

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 149 of 174

[23] REFUEL OUTAGE JUSTIFICATION RJ-23 System: Residual Heat Removal System (1001)

Valves: 68A, 68B Catecqory: AC Class: 1 Function:

These valves are the Low Pressure Coolant Injection (LPCI) pressure isolation check valves which provide a flow path for Shutdown Cooling.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

These check valves are the pressure isolation valves for the Low Coolant Injection (LPCI)

System and remain closed during normal plant operation. Each valve is an integral part of its respective RHR Shutdown Cooling loop flow path. One loop of RHR Shutdown Cooling is necessary for decay heat removal during a cold shutdown.

PNPS practice is to select an RHR Shutdown Cooling loop on a staggered basis each shutdown and remain exclusively in the selected loop for shutdown duration, unless changing plant conditions or maintenance activities necessitate shifting to the other loop.

Swapping from one RHR loop to another creates a "higher risk evolution" because of 1) excessive manpower loading, 2) deviations from normal system configurations,

3) complexity of this task (i.e., high susceptibility to events causing the loss of key safety functions), and 4) large dose accumulations.

This method of devoting one loop to Shutdown Cooling for the shutdown duration is supported by the conclusions of NUMARC 91-06, Guidelines For Industry Actions to Assess Shutdown Management. This document references numerous NRC lENs and IEBs in which a loss of "key safety functions" (i.e., decay heat removal capability and inventory control) has occurred during "higher risk evolutions". Because of task complexity, swapping RHR Shutdown Cooling loops for the purpose of partial exercising these injection check valves creates a high risk evolution and should be avoided. For the case of mid-cycle and refueling outages, plant conditions/activities usually require swapping from one RHR loop to the other.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 150 of 174

[23] REFUEL OUTAGE JUSTIFICATION RJ-23 (CONTINUED)

Basis for Justification:

Exercise open testing an injection check valve at the maximum required accident flow rate is only obtainable by operating three RHR pumps. Normal plant limitations do not allow the operation of more than two RHR pumps within a loop. An exercise open test can be verified by performing diagnostic testing while two pumps pass flow through a Shutdown Cooling loop. This testing requires entry into primary containment (Drywell) and operation of special test equipment in a high radiation area.

Performing diagnostic nonintrusive monitoring of these valves on a refueling interval demonstrates that the full-stroke capability of each valve is acceptable.

Alternate Testingq:

Perform the exercise open test of these check valves during a refueling outage not to exceed a refueling interval and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 151 of 174

[24] REFUEL OUTAGE JUSTIFICATION RJ-24 System: High Pressure Coolant Injection System (2301)

Reactor Core Isolation Cooling System (1301)

Valves: 2301-7, 1301-50 Cateqory: AC Class: 1 Function:

These valves are the HPCI and RCIC injection pressure isolation check valves.

Test Requirements:

ISTC-3522(c): If exercising is not practicable during plant operation at power and cold shutdowns, it shall be performed during refueling outages.

ISTC-3522(f): All valve testing required to be performed during a refueling outage shall be completed before returning the plant to operation at power.

Basis for Justification:

Testing these valves with flow during normal operation would require: 1) injecting cold water into the Reactor Vessel using the HPCI (or RCIC) System which would result in both a reactivity excursion and thermal shock to the feedwater nozzle and piping and 2) loss of one pressure isolation barrier between the high pressure Reactor coolant system and low pressure HPCI (or RCIC) System.

Performing the OM Code exercise test (both open or close tests) during cold shutdowns is typically impaired due to differential pressure experienced across the valve disc. Valve disc locking occurs due to residual feedwater system pressure and/or negative pressure on the upstream side of the HPCI supply piping. Additionally, access to the CK-2301-7 valve exercise adapter requires installation and removal of temporary scaffold planking (personal safety concern) thus increasing the test preparation and demobilization duration.

These check valves are manually stroke testable. Since a maximum allowable torque on the mechanical test lever of 120 ft-lb for valve 2301-7 and 75 ft-lb for valve 1301-50 are specified by the manufacturer to prevent damage, the exercise open test is performed by use of the manual exerciser and applying normal force to move the obturator from the close seat to the full open position.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 152 of 174

[24] REFUEL OUTAGE JUSTIFICATION RJ-24 (CONTINUED)

Basis for Justification:

The CK-2301-7 and 1301-50 exercise close test is conducted by repositioning the manual exerciser to the close position, which releases the obturator, allowing the disc to move to the valve (close) seat. The valve disc weight and gravity return the obturator to the closed position. Full valve closure is verified by repositioning the manual exerciser to the verified-close position, which demonstrates (by position and force increase) that the disk has returned to its seat. The manual exerciser is then positioned to the neutral (normal operation) position, ensuring that the exercise mechanism does not interfere with valve open stroke operation.

Permanent, plant-installed, nonintrusive test equipment does not exist for this valve. Other methods using portable equipment (e.g., acoustic, ultrasonic, magnetic, and radiography) which can be used to verify an open or close test on this valve require valve cycling and, as previously stated, are not practical for quarterly or cold shutdown testing.

Alternate Testinq:

Perform the exercise test (both open and close tests) during a refueling outage (not to exceed a refueling interval) and complete testing prior to returning the plant to operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 153 of 174 7.4 VALVE DISASSEMBLY EXAMINATION JUSTIFICATIONS Valve Disassembly Examination Justifications are provided for conditions in which compliance to ASME OM Code, Subsection ISTC test requirements exist that make it impractical to demonstrate the check valve exercise open or close test. Check valve obturator movement will be verified by a sample disassembly examination program. Each justification identifies: the valve(s) involved, basis for justification, and alternate testing.

NONE

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 154 of 174 7.5 VALVE TESTING PROGRAM RELIEF REQUESTS Valve Relief Requests (VR) are provided for conditions in which compliance to ASME OM Code, Subsection ISTC test requirements cannot practically be satisfied. Each Relief Request identifies:

the valve(s) involved, test requirements(s) of noncompliance, basis for relief, and alternate testing.

NONE

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 155 of 174 7.6 VALVE TESTING PROGRAM TECHNICAL POSITIONS Valve Technical Positions are provided to define and clarify a technical position that has been incorporated into the PNPS Inservice Testing Program. Technical positions are consistent with OM Code requirements and do not provide justification for extending Code testing frequencies or provide justification for deviating from the OM Code test requirements.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 156 of 174

[1] VALVE TECHNICAL POSITION VTP-01 System: Core Spray System (1400)

Low Pressure Safety Injection System (1001)

Standby Liquid Control System (1101)

Control Rod Drive Hydraulic System (302)

Valves: 1400-2A/2B (Class 2)

MO-1 001-21(Class 2), MO-1 001-32 (Class 2)

CK-1 101-15 (Class 1), CK-1 101 -43A/43B (Class 2)

CV-302-21A/21B (Class 2), CV-302-22A/22B (Class 2)

CV-302-23A/23B (Class 2), CV-302-24A/24B (Class 2)

Category: B and C Class: 1 and 2 Function:

Valves that perform a system boundary isolation function, in the closed position. These valves are grouped by valve type, system application, and safety function, and are listed as follows: 1) Parallel Pump Train Bypass Flow Isolation, 2) Safety-related ("Q" listed/seismic)

System to Nonseismic System Boundary Isolation, and 3) Reactor Coolant System (High Pressure System) to Standby Liquid Control (High Pressure System) Boundary Isolation.

PNPS Position:

The above listed Category B and C valves perform a system boundary isolation function and will have their full closure/isolation function verified by performing an augmented seat leakage test. The seat leakage to verify system boundary isolation (LSBI) test for these valves will be performed on a frequency as defined by the Performance-Based Testing (PBT) methods described within this Valve Technical Position.

Basis:

A Performance-Based Testing Plan (PBT Plan) has been developed to control the test frequencies for periodic testing of system boundary isolation valves that have their full valve closure function verified. The PBT Plan assigns the augmented seat leakage test intervals (which verify satisfactory valve full closure) based on system service and component performance. Through a screening process, the PBT test frequency criteria have only been applied to valves that exhibit a high degree of seat leakage reliability when the extended test interval is applied.

The PBT Plan for valves that have their system boundary isolation (SBI) function verified through the performance of augmented LSBI leakage testing adopts the same practices for establishing test frequency as the Option B Program for Appendix J tested valves, which was approved by amendment of the Code of Federal Regulations on October 26, 1995. The augmented LSBI leakage testing will administratively use the test method guidance identified within the ASME OM Code, Subsection ISTC.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 157 of 174

[1] VALVE TECHNICAL POSITION VTP-01 (CONTINUED)

Basis:

A test history review was conducted for each valve to determine long-term valve closure/seat leakage performance to screen out "suspect valves" that exhibit inconsistent or erratic valve seating performance. In this way, the PBT test frequency criterion is only being applied to valves that exhibit a high confidence level that seat leakage reliability will not be impacted when an extended test interval is applied.

Valves included within this Technical Position that exhibit normal operation behavior, pass a minimum of two consecutive tests, and have not been flagged as "suspect valves" will be placed on an extended test interval of up to 5 years or two refueling intervals, whichever is longer (not including a 15-month grace period). Any valve not meeting the minimum threshold test performance requirement will be placed on a 2-year test interval. When a seat leakage test failure occurs on an extended interval valve, the default test frequency of 2 years must be re-established until an assessment of the valve demonstrates that extending the test interval will not impact valve seat leakage performance, and two periodic consecutive seat leakage tests are acceptable.

Technical Position TestinQ:

IST Category B and C valves that receive augmented seat leakage testing to verify their system boundary isolation function AND are included within this Valve Technical Position will have their seat leakage testing frequency established in accordance with the methods described within the PBT Plan (described at the end of this document). PNPS 8.7.1.3.1, "Performance-Based Leakage Testing of the Primary Containment", will be used as a guideline for controlling the augmented LSBI leakage test frequencies. (Test frequency grace limitations identified within PNPS 8.7.1.3.1 will not be applied to the LSBI leak test intervals.)

Valves that exhibit normal operation behavior and pass a minimum of two periodic consecutive seat leakage tests will be permitted to be tested once every 5 years or two refueling intervals, whichever is longer (not including grace period).

Valves which fail their augmented seat leakage acceptance criterion will be placed on a 2-year test frequency. Before the extended test interval may be reinstated, an assessment must be performed which demonstrates that extending the test interval will not impact valve full seating and system boundary isolation performance. In addition, each valve must pass a minimum of two periodic consecutive seat leakage tests.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 158 of 174

[1] VALVE TECHNICAL POSITION VTP-01 (CONTINUED)

Performance-Based Testing Plan:

The valves included within this PBT Plan are categorized by the valve type, system application, and design function requiring system boundary isolation. These valves meet the applicable guidelines of the Nuclear Energy Institute (NEI-94-01) Industry Guidelines for Implementing Performance-Based Option of 10CFR Part 50, Appendix J for a performance-based testing program. Using this guideline, valves that have passed a minimum of two consecutive LSBI leak tests may be placed on an extended testing interval. All valves placed on an extended testing interval for seat leakage will still have all other associated ASME OM Code testing (i.e., exercising and position verification) performed at the required frequency as prescribed by the Inservice Testing Program. Valves that have not passed the minimum of two consecutive tests will continue to be tested once every 2 years (default frequency) until their LSBI test performance permits using the extended testing interval.

The PBT Plan for establishing augmented seat leakage test frequencies to verify full valve closure and system boundary isolation capability utilizes at least a decade of valve seat leakage test data for each valve included within this scope.

A test history review was conducted for each valve to determine long-term valve closure/seat leakage performance to screen out "suspect valves" that exhibit inconsistent or erratic valve seating performance. The screening process used Condition Monitoring methods to ensure that valve seating performance is maintained at a high confidence level. In this way, the PBT test frequency criterion is only being applied to valves that exhibit a high confidence level (through seat leakage trend data, maintenance history data, and valve failure analysis, when applicable) that seat leakage reliability will not be impacted when an extended test interval is applied.

Valves that exhibit normal operation behavior, pass a minimum of two consecutive tests, and have not been flagged as "suspect valves" will be placed on an extended test interval of up to 5 years or two refueling intervals, whichever is longer (not including a 15-month grace period). Any valve not meeting the minimum threshold test performance requirement will be placed on a 2-year test interval. When a seat leakage test failure occurs on any extended interval valve, the default test frequency of 2 years must be re-established until an assessment of the valve demonstrates that extending the test interval will not impact valve LSBI performance, and two consecutive periodic seat leakage tests are acceptable.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 159 of 174

[1] VALVE TECHNICAL POSITION VTP-01 (CONTINUED)

The following sections contain the assessment for each valve group.

REACTOR COOLANT SYSTEM TO HIGH PRESSURE SYSTEM (STANDBY LIQUID CONTROL) BOUNDARY ISOLATION Standby Liquid Control (SLC) Injection Line:

SLC Inboard Injection Check Valve (CK-1101-15) provides a pressure isolation barrier between the RCS and the high pressure portion of the SLC System. The system RCS isolations include an inboard injection check valve, an outboard injection check valve, and squib valves. Upon firing of the squib valves with the RCS at operating pressure, this check valve becomes one of two RCS isolation barriers. Therefore, the valve performs an isolating barrier between the high pressure reactor system and the high pressure SLC safety system interface. The SLC pump discharge check valves and the SLC reciprocating pump internal check valves provide two additional high pressure barriers between the RCS and the low pressure SLC suction. Maintaining valve seat leakage below the assigned LSBI limit ensures satisfactory valve closure/system boundary isolation function.

Seat leakage measurement is performed by collecting leakage at the upstream test connection during the Reactor Pressure Vessel Leakage Test. A review of seat leakage testing data from 1987 to present (most recent testing in 2009) shows that CK-1 101-15 valve has never had a test failure. There are no adverse seat leakage trends that point to degradation of the valve seating characteristics, nor is there an erratic seat leakage test history.

The CK-1 101-15 valve has not been incorporated into the PNPS IST Program Condition Monitoring Program, and therefore, will continue to be leak tested for the LSBI function on an administrative 2 year frequency. The Check Valve Condition Monitoring Program, which is governed by the OM Code Subsection ISTC, Appendix II, improves valve performance and optimizes maintenance, performance monitoring, and testing activities to ensure continued acceptable check valve performance.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 160 of 174

[1] VALVE TECHNICAL POSITION VTP-01 (CONTINUED)

SAFETY RELATED ("Q" LISTED/SIESMIC) SYSTEM TO NONSEISMIC SYSTEM BOUNDARY ISOLATION Safety related systems that require an active/passive isolation between an ASME Code Class piping system and a nonseismic/non-"Q" listed piping system. The system isolation ensures that safety-related system inventory will be maintained and that post accident system flow (with potentially contaminated water) outside the "Q" listed/seismic boundary is minimized. These valves remain closed during normal plant operation and have their system isolation function verified through the periodic performance of the LSBI leak test.

Maintaining the valve seat leakage below the assigned LSBI limit ensures satisfactory valve closure/system isolation function.

Residual Heat Removal to Radwaste Crosstie:

RHR System Discharge to Radwaste Flow Control Valve (MO-1001-21) and RHR System Discharge to Radwaste Block Valve (MO-1001-32) provide the system boundary isolation between the RHR System and the Radwaste System. These valves receive periodic diagnostic testing and performance monitoring as prescribed through the PNPS MOV Program. The MOV Program also implements a rigorous preventive maintenance plan which includes periodic inspections of the valve actuator.

Augmented seat leakage (LSBI) testing verifies that the valve full closure/isolation function is satisfactory, which ensures that secondary containment bypass flow will be minimized.

The augmented seat leakage testing is performed by measuring the feed rate required to maintain test pressure in the test volume. A review of seat leakage testing data for MO-1001-21 and MO-1 001-32 from 1987 to present (most recent testing in 2009) shows that these valves have had no test failures. There are no observed adverse seat leakage trends that point to degradation of the valve seating characteristics, nor is there an erratic seat leakage test history.

Core Spray to Condensate Storage Tank Crosstie:

Core Spray Pump A and B Manual Suction Valves from Condensate Storage Tank (1400-2A and 2B) provide the system boundary isolation between the Core Spray System and the Condensate Storage Tank. Maintaining valve seat leakage below the assigned LSBI limit ensures the secondary containment bypass leakage is minimized. Seat leakage is currently being measured by either the feed rate required to maintain test pressure in the test volume or by measuring leakage through a downstream telltale connection while maintaining test pressure on one side of the valve. A review of periodic seat leakage testing data for 1400-2A and 2B from 1993 to present (most recent testing in 2009) shows that these valves have no historical test failures. There are no observed adverse seat leakage trends that point to degradation of valve seating characteristics, nor is there an erratic seat leakage test history.

NOTE: The 1400-2A and 2B are normally closed hand-operated gate valves that are designated as being passive within the IST Program. These valves have been included (as active valves) within this technical position, because typically one valve is opened once every refueling outage (each valve is cycled once every 4 years or two refueling outages) to provide the Core Spray System injection test flow path from the Condensate Storage Tank to the Reactor Vessel.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 161 of 174

[1] VALVE TECHNICAL POSITION VTP-01 (CONTINUED)

SDV Vent and Drain Valves to Reactor Buildinq Sump:

Scram Discharge Volume Vent Valves (CV-302-21A/B and CV-302-23A/B) and Scram Discharge Volume Drain Valves (CV-302-22A/B and CV-302-24A/B) provide the boundary isolation between the SDV and the Reactor Building Sump. These valves receive periodic diagnostic testing and performance monitoring as prescribed through the PNPS AOV Program. Maintaining valve seat leakage below the assigned LSBI limit ensures the system isolation from the Reactor Coolant Pressure Boundary (RCPB) during a Scram is maintained. Seat leakage is currently being measured by the feed rate required to maintain test pressure in the test volume. A review of seat leakage testing data from 1991 to present (most recent testing in 2009) shows that these valves have never experienced a seat leakage test failure.

One postmaintenance seat leakage test failure was observed following improper limit switch bracket adjustment. During this maintenance activity, improper alignment of a limit switch bracket prevented the valve from fully closing. The historical test data for all valves shows no observed adverse seat leakage trends that point to degradation of the valve seating characteristics, nor has there been an erratic seat leakage test history.

PARALLEL PUMP TRAIN BYPASS FLOW Standby Liquid Control (SLC) System Parallel Pump Bypass Flow:

SLC Pump A and Pump B Discharge Check Valves (CK-1 101-43A/B) allow flow of borated coolant to the Reactor Vessel upon SLC activation, and close to prevent pump bypass flow back through the idle parallel pump loop. In the event that a pump's discharge relief valve fails to close (single failure), the associated check valve within that loop will need to fully close and provide system boundary isolation. Maintaining check valve seat leakage below the assigned LSBI limit ensures pump bypass from the parallel pump system is minimized.

Seat leakage is performed by measuring the feed rate required to maintain test pressure in the test volume. A review of historical seat leakage testing data for CK-1 101-43A and 43B from 1987 to present (most recent testing in 2009) shows that these valves have never experienced a test failure. There are no adverse seat leakage trends that point to degradation of the valve seating characteristics, nor has there been an erratic seat leakage test history.

The SLC Pump Discharge Check Valves CK-1 101-43A and 43B have been included within the PNPS IST Program Condition Monitoring Program. The Check Valve Condition Monitoring Program, which is governed by the OM Code Subsection ISTC, Appendix II, improves valve performance and optimizes maintenance, performance monitoring, and testing activities to ensure continued acceptable check valve performance.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 162 of 174 7.7 SKID-MOUNT COMPONENT TECHNICAL POSITIONS Skid-Mount Component Technical Positions (STP) are provided to identify skid-mounted pump and valve information used to determine that the skid-mounted component is justified to be adequately tested. Discussion of supplemental testing activities, which may be performed in addition to the main component testing, have also been included.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 163 of 174

[1] SKID COMPONENT TECHNICAL POSITION STP-01

Subject:

Testing for the Skid-Mounted Control Rod Drive (CRD) valves included in the IST Program.

System:

Control Rod Drive (CRD) Hydraulic Units (302, 305).

Valves:

305-114, CRD Scram Header Discharge Check Valves 305-115, CRD Charging Water Check Valves CV-305-126, CRD Scram Inlet Valves CV-305-127, CRD Scram Outlet Valves 305-138, CRD Cooling Water Supply Check Valves 302-120, 302-121, 302-122, and 301-123, CRD Hydraulic Control Unit (HCU)

Directional Control Valves Catecqory: B and C Class: 2 Function:

All valves indentified in this Skid Technical Position (STP) are located on the CRD hydraulic control units (HCUs) for the control rod drives. All valves, except for the HCU Directional Control Valves (302-120, 302-121, 302-122, and 301-123) are required to change position in support of a control rod scam. The directional control valves have a passive function to remain closed during the scram. Upon receipt of a scram signal from the reactor protection system, the active skid valves indentified in this STP are required to function (change position) for rapid insertion of the control rods into the reactor core to facilitate safe and orderly shutdown of the reactor.

Discussion:

CRD Scram Inlet and Outlet Valves The skid-mounted CRD Scram Valves CV-305-126 and CV-305-127 valves have a safety function to open to provide the supply (charge) flow path from the CRD HCU accumulator to the control rod drive mechanism (CRDM), and provide a discharge flow path from CRDM to the associated scram discharge instrument volume (SDIV) header. These valves isolate the scram charge and discharge lines prior to scram initiation, during normal operation, and following scram reset. When a control rod is not being moved manually, the CV-305-126 and CV-305-127 are closed to ensure the ability of the HCU to properly execute a scram is maintained.

The CV-305-126 and CV-305-127 valves are functionally tested during the individual rod scram insertion time testing and evaluation required by Technical Specifications Sections 4.3.C.1, 4.3.C.2, 4.3.C.3 and 4.3.C.4. The functional testing is verified during scram of the associated rod, these valves receive full stroke exercising (FE) and fail-safe (FS) testing.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 164 of 174

[1] SKID COMPONENT TECHNICAL POSITION STP-01 (CONTINUED)

Discussion:

CRD Cooling Water Supply Check Valves The skid-mounted cooling water supply check valves 305-138 adequately demonstrates proper functioning for an exercise closed test (CT) during the individual control rod exercising as required by PNPS Technical Specifications. The check valves are positioned in the closed direction when notching the associated control rod in. Control rod notching satisfactorily verifies check valve closure. An exercise open test (OT) is not performed since valve opening capability is not required to support the scram function or any other safety function. The 305-138 valves are functionally tested open during normal operation, by the observing normal cooling water flow, which is passed through each valve to the Control Rod Drive Mechanism.

Background:

The 305-138 check valve is located in the cooling water line that supplies each control rod mechanism with a continuous flow of cooling water past the various mechanism seals and into the reactor vessel. This valve has an operational function to open to provide a supply flow path for the CRDM cooling. Check valve 305-138 closes to ensure the ability of each HCU to perform a reactor scram despite failure of the cooling water portion of this system.

Experience has demonstrated that prolonged exposure of one of these seals to reactor coolant temperatures can shorten the expected life of the seal. Loss of cooling water, however, does not prevent a reactor scram or threaten the integrity of the reactor coolant pressure boundary.

CRD Scram Header Dischargeand Charging Water Check Valves The skid-mounted scram discharge header check valves, 305-114 and charging water check valves, 305-115 adequately demonstrate proper functioning for an exercise open test (OT) during the individual rod scram insertion time evaluation required by Technical Specifications Sections 4.3.C.1, 4.3.C.2, 4.3.C.3 and 4.3.C.4.

Supplemental testing is performed on skid-mounted charging water check valves, 305-115 which adequately demonstrates proper functioning for an exercise closed test (CT). The closed test is considered acceptable when either no Accumulator High Level/Low Pressure Annunciator energizes or HCU accumulator pressure of > 700 psig is observed following a 30-second hold period after shutting off the operating CRD pumps. Skid-mounted scram discharge header check valves, the 305-114 valves do not receive an exercise close test (CT) since they are not required to close to support the scram function.

Background:

The 305-114 has no safety function in the closed direction. By design, the 305-114 valves were installed to prevent communication between the scram discharge volume and the drive mechanisms. However, this can only occur during the brief moment the control rod is settling to its latched position. Typically, flow from the overpiston area of the drive to the scram discharge volume occurs throughout the scram of the control rod by pressure exerted to the underpiston area from the accumulator or reactor pressure, and continues until the volume pressure equals reactor vessel pressure with the control rod latched in its fully inserted position.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 165 of 174

[1] SKID COMPONENT TECHNICAL POSITION STP-01 (CONTINUED)

In the unlikely event that reverse flow attempts to return back to the control rod drive (past the 305-114 check valve), it will not be capable to provide significant flow in the direction against the latched control rod (pressure in the withdraw direction), because the boundary valves on the underpiston side of the CRDM (305-115, 305-138, SV-305-120, & SV-305-123) all still remain closed, causing the underpiston piping section to become pressure locked. In addition, during the same scenario a portion of the reverse flow/pressure will dissipate as it passes through the CRDM seals back to the vessel. Therefore, failure of the scram outlet valve to close would not prevent the CRD system from performing its safety function.

CRD Hydraulic Control Unit (HCU)DirectionalControl Valves Control rod directional control valves SV-305-120, 121, 122 and 123 are considered skid-mounted but do not perform an active function in scramming the control rods to rapidly shutdown the reactor. The control rod directional control valves change position to provide the mechanism for normal rod movement, which is not safety related. These valves are not required to cycle open or closed to achieve a successful reactor scram. The valves are excluded from the IST program scope, however they are listed within the program valve table as skid-mounted valves because they close following control rod exercising, and remain in the closed position to ensure the ability of the HCU to properly execute a reactor scram.

The performance of control rod exercising within PNPS 8.3.2 satisfies the functional testing requirements for the associated skid-mounted CRD Hydraulic Control Unit (HCU) Directional Control Valves (302-120, 302-121, 302-122, and 301-123). The demonstration of satisfactory valve exercise to the alternate position and returning back to the normal position is verified through the satisfactory control rod notching in and out during control rod exercise testing.

CRD Scram Accumulator Rupture Discs CRD scram accumulator rupture discs, 305-132 are excluded from the requirements of IST.

OM Code ISTC-1200(c) states; "Non-reclosing pressure relief devices (rupture disks) used in BWR Scram Accumulators are excluded from the requirements of this Subsection" meaning ISTC.

I

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 166 of 174

[2] SKID COMPONENT TECHNICAL POSITION STP-02 Subiect:

Testing for RCIC Pump/Turbine Skid-Mounted pumps and valves included in the IST Program.

System:

Reactor Core Isolation Cooling (RCIC) System (1301)

Valves:

P-221, RCIC Gland Seal Condensate Pump CK-1301-24, RCIC Condensate Pump P-221 Discharge Check Valve CK-1 301-40, RCIC Vacuum Pump Discharge to Torus Check Valve CK-1 301-63, RCIC Vacuum Tank Condensate Return Check Valve AO-1301-12, RCIC Condensate Pump Discharge Valve to CRW PCV-1301-43, RCIC Cooling Water Supply Line Pressure Regulator PSV-1301-70, RCIC Vacuum Tank Relief Valve VRV-1 301-9067, RCIC Turbine Exhaust Line Vacuum Relief Valve SV-1 301-1, RCIC Turbine Steam Inlet Trip-Throttle Valve HO-1 301-159, RCIC Turbine Governor Valve Catecqory: B, AC and C Class: 2 Function:

All components identified in this Technical Position are required to function in support of startup and satisfactory operation of the turbine driven Reactor Core Isolation Cooling (RCIC) Pump. The ability for the RCIC pump to satisfy its flow requirements when periodically tested demonstrates satisfactory operation of the listed components. The following position discussion will support justification of adequate testing for the classification of these components as skid-mounted, as recognized by the OM Code Discussion:

The skid-mounted RCIC vacuum tank condensate pump, P-221, adequately demonstrates proper functioning when the RCIC vacuum tank water level is maintained within the normal expected range during RCIC turbine operation Normal expected range is defined as vacuum tank high level annunciator from LS-1301-15 does not alarm and fail to clear.

Likewise, skid-mounted vacuum tank condensate pump to RCIC pump suction check, CK-1301-63, and vacuum tank condensate pump discharge check, CK-1301-24, are functionally exercised open (OT) by this same activity of vacuum tank level control.

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[2] SKID COMPONENT TECHNICAL POSITION STP-02 (CONTINUED)

Discussion:

Normal function to close of CK-1301-63 is apparent following restoration from testing of the RCIC pump turbine to the normal standby condition, as failure of this check valve to close would result in inability to restore the system to a standby condition. Supplemental testing performed every 2 years demonstrates a functional closed test (CT) of CK-1301-63 by observing vacuum tank condensate pump, P-221 does not start more than once during a 10 minute period. Closure verification of CK-1301-24 is not performed since the valve is not required to close in support of RCIC turbine operation.

Skid-mounted barometric condenser drain valve, AO-1 301-12 adequately demonstrates proper functioning by its ability to auto open when a high water level is detected in the barometric condenser and the system is in the standby mode. The valve's opening provides a flow path to clean radwaste. The valve auto closes on system initiation or test.

Supplemental testing is performed each quarter on the skid-mounted barometric condenser drain valve, AO-1301-12 by performing a full stroke exercise (FE), stroke timing (ST), and fail-safe test.

Skid-mounted pressure control (regulating) valve PCV-1 301-43 adequately demonstrates proper functioning during quarterly RCIC Turbine/Pump operation by ability of the system to maintain normal cooling water supply line pressure. Failure of the PCV-1301-43 to regulate pressure in the cooling water supply line would be observed as cause for low flow condition (inadequate cooling for turbine skid components) or high flow condition (excess cooling line water pressure and eventual lifting of PSV-1301-42, set pressure of 75 psig) during normal RCIC operation.

Skid-mounted vacuum pump discharge to torus check, CK-1301-40 is functionally exercised open tested (OT) when the RCIC vacuum tank pressure can be maintained within the normal expected band (vacuum tank high pressure annunciator from PS-1 301-14 does not alarm and fail to clear). The exercise closed test (CT) of CK-1301-40 is satisfactorily verified when a restricted flow rate of 5 CFM (141 SLM) or less at 25 to 30 psig is obtained in the reverse flow direction.

Skid-mounted relief valve PSV-1301-70, was GE supplied as part of the barometric condenser skid package, and is attached to the barometric condenser. The passive close function of the relief valve is apparent during quarterly RCIC Turbine operation by ability to maintain normal barometric condenser parameters. As a supplemental test, PSV-1301-70 is set pressure tested in accordance with OM Code methods once every 10 years.

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[2] SKID COMPONENT TECHNICAL POSITION STP-02 (CONTINUED)

Discussion:

RCIC Turbine Trip-Throttle Valve and Governor Valve The skid-mounted RCIC turbine trip-throttle valve, SV-1301-1 and RCIC turbine governor valve, HO-1 301-159 adequately demonstrate proper functioning and operational readiness by observation of acceptable turbine operation (i.e. obtaining pump rated flow and turbine speed during surveillance testing). This position is further supported by NUREG 1482, Rev. 1, Section 3.4. In addition, verifying pump operational time to achieve rated flow during testing is performed at least once per operating cycle (typically performed quarterly). Control valve, HO-1 301-159, degradation would be identified during this testing because the control valve performance directly impacts the time for the RCIC pump/turbine system to obtain rated flow. Likewise, trip-throttle valve SV-1301-1 must close to stop RCIC turbine.

Operation of this normally latched open valve is automatic and closes on a RCIC turbine trip signal (turbine overspeed, high turbine exhaust pressure, low pump suction pressure, RCIC isolation signal). The valve must be manually relatched in the open position subsequent to closure.

RCIC Exhaust Vacuum Breaker Line The skid-mounted RCIC exhaust vacuum breaker line check valve, VRV-1 301-9067 is functionally tested, and is verified to be satisfactorily exercised open (OT) when a flow rate of 15 CFM (425 SLM) or greater is observed through the valve with a driving pressure of no more than 25 psig. VRV-1301-9067 is functionally exercised close tested (CT) when no steam leakage is observed from the open end of the RCIC steam exhaust vacuum relief line during quarterly operation of the RCIC turbine.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 169 of 174

[3] SKID COMPONENT TECHNICAL POSITION STP-03

Subject:

Testing for HPCI Pump/Turbine Skid-Mounted pumps and valves included in the IST Program.

System:

High Pressure Coolant Injection (HPCI) System (2301)

Valves:

P-220, HPCI Condensate Pump CK-2301-34, HPCI Exhaust Drain Check AO-2301-64, HPCI Gland Seal Condenser Drain Valve to CRW PCV-1 301-43, HPCI Gland Seal Condenser/Lube Oil Cooler Cooling Water Regulating Valve CK-2301-75, HPCI Gland Seal Condenser Return Check CK-2301-76, HPCI Gland Seal Condenser Return Check CK-2301-217, HPCI Exhaust Drain Check HO-2300-23 (HO-1), HPCI Turbine Steam Inlet Stop Valve HO-2300-24 (HO-2), HPCI Turbine Steam Inlet Control Valve CV-9068A & CV-9068B, HPCI Turbine Exhaust Line Drain Pot Drain Valve Cateqory: A, B, AC and C Class: 2 Function:

All components identified in this Technical Position are required to function in support of startup and satisfactory operation of the turbine driven High Pressure Coolant Injection (HPCI) Pump. The ability for the HPCI pump to satisfy its flow requirements when periodically tested demonstrates satisfactory operation of the listed components. The following position discussion will support justification of adequate testing for the classification of these components as skid-mounted, as recognized by the OM Code Discussion:

HPCI GSC System Components The skid-mounted HPCI gland seal condenser (GSC) hotwell pump, P-220, adequately demonstrates proper functioning by its ability to control hotwell level during the HPCI turbine run. Likewise, skid-mounted HPCI GSC hotwell pump discharge check, CK-2301-76 is functionally exercised open (OT) by this same activity of hotwell level control and further verified by observing a cleared annunciator "Gland Seal Condenser Hotwell Level Hi".

Closure verification is demonstrated by the lack of cycling of the hotwell pump as a result of condensate draining to the gland seal condenser since the hotwell (spacing) pump auto starts on high water level in the gland seal condenser. Failure of this check valve to close would result in the inability to restore the system to a standby condition.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 170 of 174

[3] SKID COMPONENT TECHNICAL POSITION STP-03 (CONTINUED)

Discussion:

HPCI GSC System Components Supplemental testing is performed on the skid-mounted HPCI GSC hotwell pump discharge check, CK-2301-76 every 2 years by observing that the HPCI gland seal condenser hotwell pump, P-220 does not start more than once during a 10 minute period when the HPCI GSC system is in service and the HPCI main/booster pump suction is aligned to the CST.

Skid-mounted gland seal condenser drain block valve, AO-2301-64 adequately demonstrates proper functioning by its ability to auto open when a high water level is detected in the CSG and system is in the standby mode. The valve will not open or will auto close if a HPCI initiation signal is present. The valves' opening provides a flow path to clean radwaste.

Supplemental testing is attempted each quarter on the skid-mounted gland seal condenser drain block valve, AO-2301-64 by performing a full stroke exercise (FE), stroke timing (ST),

and fail-safe test. However the valve will not open if a low level is present in the GSC. If quarterly testing is unsuccessful, supplemental testing is performed every 2 years by installing electrical jumpers to allow valve stroking during any system condition. PI test is performed every 2 years.

Skid-mounted pressure control (regulating) valve PCV-1 301-46 adequately demonstrates proper functioning during quarterly HPCI Turbine/Pump operation by ability of the system to maintain normal cooling water supply line pressure. Failure of the PCV-1301-46 to regulate pressure in the cooling water supply line would be observed as cause for low flow condition (inadequate cooling for turbine skid components) or high flow condition (excess cooling line water pressure and eventual lifting of PSV-2301-53, set pressure of 100 psig) during normal HPCI operation.

HPCI Lube Oil Cooler Components Skid-mounted HPCI turbine lube oil cooler check valve, CK-2301-75 adequately demonstrates functionality for the exercise open test (OT) by the observation of normal GSC pressure and lube oil cooler discharge temperature during the HPCI run. CK-2301-75 does not have a safety function to close. Failure of the CK-2301-75 to close does not affect HPCI system ability to start (on initiation signal) and meet design bases function. Since the valve does not perform any function in the closed direction, an exercise closed test (CT) will not be performed or verified.

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[3] SKID COMPONENT TECHNICAL POSITION STP-03 (CONTINUED)

Discussion:

HPCI Exhaust Drain Components The skid-mounted HPCI exhaust drain outboard and inboard check valves, CK-2301-34 and CH-2301-217 adequately demonstrate proper functioning for an open exercise test (OT) by observing a cleared annunciator "Exhaust Drain Pot Level Hi" during system standby and subsequent to a HPCI run.

Supplemental testing is performed on series skid-mounted CK-2301-34 and CH-2301-217 once every 4 years (with one year grace), typically during a refueling outage, by performing a specialized flow test. Satisfactory open exercise test is demonstrated by the valves ability to pass a flow rate of at least 6 gpm. The exercise close test (CT) is demonstrated each refueling outage by performance of a seat leakage test.

HPCI Turbine Inlet Stop Valve and Control Valve The skid-mounted HPCI - Stop Valve (trip throttle valve), HO-2300-23 (HO-1), and steam inlet control valve (governor valve).HO-2300-24 (HO-2), adequately demonstrate proper functioning and operational readiness by observation of acceptable turbine operation (i.e.,

obtaining turbine rated flow and speed during surveillance testing). This position is further supported by NUREG 1482, Rev. 1, Section 3.4. In addition, verifying pump operational time to achieve rated flow during testing is performed at least once per operating cycle (typically performed quarterly). Valve degradation would be identified during this testing because stop valve and steam inlet control valve performance directly impacts the time for the HPCI pump/turbine system to start and obtain rated flow. PI test is performed once every 2 years. Likewise HO-1 must close to stop HPCI turbine. Valve operation is automatic and closes on a HPCI turbine trip signal (turbine overspeed, high turbine exhaust pressure, low pump suction pressure, HPCI isolation signal). PI test is performed once every 2 years.

HPCI Solenoid Operated Drain Pot Drain Valves The specified functional tests for the skid-mounted HPCI drain pot drain valves, are based upon the required safety functions. These valves must close to provide an isolation boundary to meet 10CFR50 Appendix J Criterion. CV-9068A & CV-9068B have no open safety-related function. They receive seat leakage testing in accordance with the requirements of 10CFR Appendix J, Option B. The specified functional testing, in addition to the Appendix J leak testing, ensure that these valves will properly close and provide an acceptable leakage boundary. The assigned testing is as follows:

Stroke Time (ST), Full Stroke Exercise (FE) and Fail Safe (FS) Test - CV-9068A & CV-9068B have a safety function to close and provide a containment isolation function. The performance of quarterly valve stroke timing (ST) and full stroke exercising (FE) also verifies the functional close fail safe test (FS) for these solenoid operated valves. This quarterly valve functional testing, combined with periodic Appendix J local leak rate testing (LLRT),

satisfactorily demonstrates the valve function to properly close during normal and fail safe conditions.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 172 of 174

[3] SKID COMPONENT TECHNICAL POSITION STP-03 (CONTINUED)

Discussion:

HPCI Solenoid OperatedDrain Pot Drain Valves Position Indication Function (PI) - The satisfactory function of the solenoid operated HPCI drain pot drain valve position indicators is demonstrated by observing the normal operation of the indicator lights during the HPCI valve quarterly switch to light stroke timing and full stroke exercising, combined with a periodic (once every 4 years with one year grace) position indication (PI) verification. The PI functional verification is typically performed during the Appendix J Local Leak Rate Test (LLRT) of each solenoid operated valve. The LLRT is performed after the corresponding solenoid operated valve has been closed by its normal means (and closed indicator light observed). The valve position indication verification is also performed as a post-maintenance testing requirement (regardless of whether the Appendix J local leak rate testing will be performed) which ensures that the remote valve position indicators will accurately reflect valve operation following applicable maintenance activities.

The satisfactory functional operation for each HPCI drain pot drain valve is demonstrated by observation of normal functional performance during routine valve operation, the performance of quarterly valve functional testing and periodic (once every four years) position indication verification, In addition, the performance of periodic 10CFR Appendix J seat leakage testing provides additional information about each valve's performance that further demonstrates satisfactory valve function and operation.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 173 of 174

[4] SKID COMPONENT TECHNICAL POSITION STP-04 Submect:

Testing for the skid-mounted Traversing In-core Probe (TIP) valves included in the IST Program.

System:

Traversing In-core Probe (45)

Valves:

SV-45-300A, A TIP Ball Valve SV-45-300B, B TIP Ball Valve SV-45-300C, C TIP Ball Valve SV-45-300D, D TIP Ball Valve TIP N2 Purge Check Valve 45-301A, A Shear Valve 45-301B, B Shear Valve 45-301 C, C Shear Valve 45-301 D, D Shear Valve Category: A, AC and D Class: 2 Function:

All valves identified in this Technical Position (TP) are associated with the Traversing In-core Probe (TIP) system. The ball valves change position during the process function of running the TIPs for measuring the axial neutron flux profile of the reactor core, which is performed routinely during power operation. All valves associated with the TIP system must be capable of closure to provide containment isolation. The shear valves are provided for rapid containment isolation, if necessary, when the associated TIP detector is inserted into the reactor core. The following position discussion will support justification of adequate testing for the classification of these components as skid-mounted, as recognized by the OM Code.

Discussion:

SV-45-300AIBICID - The skid-mounted TIP ball valves, SV-45-300A/B/C/D adequately demonstrate proper functioning for a full stroke exercise test (FE), fail-safe test (FS) and position indication (PI) testing during the normal operation and functional testing of the TIP system. As containment isolation valves, they perform a safety function in the closed position and are leak tested at the Appendix J, Option B test frequency. Supplemental testing is performed on the skid-mounted TIP ball valves which further demonstrate proper functioning by performing a quarterly full stroke exercise test (FE), a fail-safe test (FS), a stroke time test (ST). The position indication (PI) test is functionally verified once per cycle during performance of the "TIP System Operation Checkout" surveillance.

Pilgrim Nuclear Power Station SEP-PNPS-IST-001 Rev.: 2 Page 174 of 174

[4] SKID COMPONENT TECHNICAL POSITION STP-04 (CONTINUED)

Discussion:

The ball valves are considered rapid acting, but due to control system design (software and firmware) there is a system processing time delay when the ball valve control firmware "softkey" is depressed which initiates both open and close valve cycling. Therefore, due to the inherent design time delay, and valve cycle time benchmarking the satisfactory closed stroke time is designated to be < 3.5 seconds. The quarterly valve supplemental testing, seat leakage testing, once per cycle functional testing, seat leakage testing, and observation of proper TIP system operation demonstrate proper functioning and operational readiness of the TIP ball valves.

TIP N2 purge check - The skid-mounted TIP N2 purge check valve performs a safety related function in the closed position for containment isolation. The Appendix J Local Leak Rate Testing of this check valve satisfactorily verifies the exercise close test (CT) and will be performed at the Appendix J, Option B frequency. The valve open function allows nitrogen to flow into the TIP index mechanism to prevent corrosion from building up over a longer term period. The constant nitrogen flow creates a non-corrosive and mild service condition for the N2 supply check valve. The open function performs an operational purpose only and has no safety related basis. A supplemental exercise open test (OT) is not deemed necessary to ensure long term reliable valve operation. Due to the N2 check valve seat leakage performance history and mild service conditions, the Appendix J seat leak testing at the Option B test frequency has been determined to satisfactorily verify valve function at a frequency that ensures continued valve operability. The leakage test using air is a reliable performance monitoring indicator that provides the necessary information to demonstrate satisfactory long term valve condition.

45-301AIBICID - The skid-mounted shear valves 45-301A/B/C/D are considered skid since they were provided by the manufacturer as part of the major component. Although considered skid-mounted, the explosive charges shall meet the requirements of ISTC-5260 of the OM Code.

v t e Letter Sequence Request
TAC:MF0370, (Open)
Initiation Request, Request, Request Acceptance Administration Questions 1 July 2013 Answers 19 September 2013 Results Approval... Other: ML14198A157
MONTHYEAR2013-07-01 [Table View]

ML12342A386

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SUMMARY

4.4 REFERENCES

4.4 REFERENCES

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Applicable Code Requirement

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