ML20132G578

Rev 18 to Inservice Testing Program Plan for Vermont Yankee Nuclear Power Station

ML20132G578
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	12/16/1996
From:	Legere D, Reid D, Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20132G567	List: BVY-96-163, Forwards Rev 18 to Inservice Testing Program Plan for Vermont Yankee Nuclear Power Station ML20132G578
References
PROC-961216, NUDOCS 9612260299
Download: ML20132G578 (200)

Text

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INSERVICE TESTING PROGRAM PLAN Cllll"l3 p o '

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4 L.L.J F- l Vermont Yankee Nuclear Power Station

M g I Commercial Service Date: November 29,1972 f

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1 C ,3 I Rev..ision 18 g  !

1 December,1996 M  !

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! Prepared By: Reviewed By:

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, %/ U /R,/n/9 1 TIS oordinator

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i i Approved By:

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i gJ. Legere J 3 Performane ngineering Manager ,

%.177./dmA 1775. 96*/50 /M3k& f(AV71 j Plant Operations Revief fR.J anczyk 4

Committee Jia tManager l _

Ikl6 b D.A. Reid

! Vice President - Operations j

a Vermont Yankee Nuclear Power Corporation 1A Governor Hunt Road Verr.on, Vermont 05354 l

1 9612260299 961219 PDR ADOCK 05000271 G Pb3

Verment Yankee Nuclear Power Station Inservice Testing Program Pevision Summary Revision Affected Summary Number Pages 13 All Third Ten-Year Interval Update.

14 5-14, 5-13 Revise valve listing for Diesel Generator Starting Air System to reflect changes made under PDCR 92-18; DCN'

1. Minor revisions.

15 Cover,1-1,1-2, 2-5, Revise to reflect Tech Spec Amendment No.138, EDCR 3-1, 3-2, 4-1 to 4-5, 4- 92-404, EDCR 93-404, PDCR 92-016, PDCR 92-018, 10,4-12,4-14,4-17, CARS 93-70 and 94-06, PRO 94-18, items identified during 4-19,4-21 to 4-23,4- revisions to implementing procedures, current status of 25,4-29,5-1,5-3 to each Relief Request per USNRC SERs [ References (s) and 5-11,5-13 to 5-20,5- (u)], and minor revisions.

22 to 5-30,5-32 to 5-34,5-36 to 5-71,5-74 te 5-84, 5-87, 5-08, 5-91 to 5-94,5-98 to 5-98b, 5-100, 5-102, 5-105, 5-107,5-113, 5-115, 5-117,5-119, 5-122 to 5-124,5-126

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16 Cover,4-3,4-7 to 4- Revise to reflect changes made due to the responses to the 10,4-30, 4-31,5-5,5- USNRC Safety Evaluation Report issued on September 3, 7 to 5-11,5-17,5-19, 1994, Alternate Cooling Water System additions, items 5-20,5-27,5-34,5-36, identified during the implementation of the program, 5-38 to 5-41,5-43,5- current status of each relief request and correction of 52, 5-53, 5-56 to 5-59, typographical errors.

5-60,5-68,5-77,5-83 to 5-84, 5-89 to 5-90,  !

5-104 to 5-112,5-107, 5-113, 5-114, 5-118, 5-119, 5-122, 5-125 I 17 Cover,5-9,5-129 & Revise to allow the use of the sampling disassembly 5-130 technique for V70-43 A and V70-43B in accordance with the guidance specified in USNRC Generic Letter 89-04,  !

Staff Position 2. l 1

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Revision 18 i

i Vermont Yankee Nuclear Power Station Inservice Testing Program Revision Summary Revision Affected Summary

. Number Pages 18 All General program revision comnutted to in LER 95-17 and BMO 95-07. Implemented changes due to IST program scope changes and IST program bases document upgrade.

Implemented changes due to Appendix J Program upgrade.

Implemented changes due to EDCR 93-405, PDCR 94-021, PDCR 94-022, EDCR 95-409, EDCR 96-410, EDCR 96-412, EDCR 96-416, PDCR 94-007 and Minor Mod. 96-34, L

96-52. Revised program format as suggested by NUREG l 1482.

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Revision 18 ii

Vermont Yankee Nuclear Power Station Inservice Testing Program Table of Contents

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1.0 Introduction .. . . . I

. . . . . . . . . . . . . . . . .Section 1 Page1 l 1.1 Purpose . . .. . . . . . . . . . . . . . . . . . . . . . .

. . ... .Section 1 Page 1 l 1.2 Discussion . . . . . . . . . .. . . . . . .. . . .

. . .Section 1 Page1 l

2.0 Program Plan Description. . ... . . . ... .. ..Section 2 Page 1 2.1 Test Deferral Justifications. . . . . . . . .

.Section 2 Page 1 2.2 Relief Requests. . . . . .. . .

.Section 2 Page 2 {

2.3 Flow Diagrams.. . . . . . . . . . . . . . .Section 2 Page 2 l Table 2-1: List ofInservice Testing Flow Diagrams. . . . . . . . . . .Section 2 Page 3 I

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3.0 References.. i

.Section 3 Page 1 4.0 Pump Inservice Testing Program Plan.. . . . . .. .. . . . . . . . . .Section 4 Page 1 I 4.1 Scope. .

. .. . .Section 4 Page 1 4.2 Pump Program Listing .. .. ... . .. . . . . . . . . . . . . . . . . .Section 4 Page 1 Table 4-1: Pump Listing . . . . . . . . . . . . .. . . . . . . . . .Section 4 Page 4 4.3 Pump Notes. .. .. .

. . . . . . . . . . . . . .. ..Section 4 Page 8 4.4 Pump Relief Requests . . .. . ... . . . . . . . . . . .. .Section 4 Page "

5.0 Valve Inservice Testing Program Plan.. . . . . . . . .. .. . .Section 5 Page i 5.1 Scope. .. . . . . . . . . . . . . . . . . ... . . . . . . . . . .Section 5 Page 1 5.2 Valve Program Listing. . . . .

. . . . . . .. . . .Section 5 Page 2 Table 5-1: Valve Listing. . . . . . . .. . . . . . .. . .Section 5 Page 8 5.3 Valve Notes... . . . . . . . . . . .. . . . . . .. ..Section 5 Page b7 5.4 Valve Cold Shutdown Justifications. ... . . . . . . . . . . . . .Section 5 Page 90 5.5 Valve Refueling Outage Justifications . . . . . . ... .. . .Section 5 Page 109 5.6 Valve Relief Requests. .

... . . . . . . . . . . . .. . . .Section 5 Page 145 O

Revision 18 iii l

Verm:nt Ycnk:e Nuclear Pcwer Stati:n Ins:rvice Testing Pr: gram 1 1.0 Introduction (m) 1.1 Purpose 1

The Third-Interval Vermont Yankee Inservice Testing (IST) Program establishes testmg i requirements to assess the operational readiness of certain Safety Class 1,2, and 3 pumps  :

and valves which are required to:

a) Shut down the reactor to the cold shutdown condition, 1 b) Maintain the reactor in the cold shutdown condition, or c) Mitigate the consequences of an accident.

The Third-Interval IST Program is applicable for the interval from September 1,1993 -

through and including August 31,2003 [ References (q) and (r)].

i The Third-Interval IST Program is part of the Vermont Yankee Component Testing Program.

1.2 Discussion p

V The Third-Interval Vermont Yankee IST Program was developed in accordance with the requirements of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, " Rules for Inservice Inspection of Nuclear Power Plant Components," and 10 CFR 50.55a [ Reference (b)]. The Third-Interval IST Program provides compliance with Vermont Yankee Technical Specifications 4.6.E.

In accordance with Articles IWP-1000 and IWV-1000 and Table IWA-1600-1 of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code and 10 CFR 50.55a(b)(2)(viii), the pump and valve testing requirements were based on  ;

the ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987, " Operation and l Maintenance ofNuclear Power Plants" The additional requirements described in 10 CFR 50.55a(b)(2)(vii) and the recommendations described in Generic Letter 89-04 and its supplements were also included as applicable. Request for use of the ASME/ ANSI OMa- l 1988 Addenda to ASME/ ANSI OM-1987, with the additional requirements, was l submitted by Vermont Yankee and approved by the USNRC [ References (e) and (f)]. I All references to the " Code" made within this IST Program Plan will refer to the ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987 unless otherwin,e specified.

The IST Program Plan identifies the scope of components (pumps and valves) included in the Third-Interval IST Program, the testing requirements for those components, and justifications or relief requests to support the scope and testing requirements.

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Revision 18 Section 1 Page 1 of 2

Vermcat Ycnkee Nucl=r Pcwcr Stati:n Inservice Testing Pr: gram 1.2 Discussion (cont.)

O The IST Program Plan provides conformance with Anicle IWA-2420 of the 1989 Edition -

of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code and 10 CFR 50.55a(a)(3), (f)(5) and (f)(6).

The USNRC Safety Evaluations for the Third-Interval Vermont Yankee IST Program are provided in References Section.

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Verm:nt Ycnkes Nucicer Pcwer St:ti::n Inservice Testing Program 2.0 Program Plan Description The IST Program Plan is comprised of two independent subprogram plans - the Pump Inservice  !

Testing Program Plan and the Valve Inservice Testing Program Plan. I Key features common to both Program Plans are: the Pump and Valve Listings that define the scope of the Third-Interval IST Program, Cold Shutdown and Refueling Outage Justifications, Relief Requests, and applicable Notes. -

Administrative and implementing procedures, reference values, test results, and other records required to define and execute the Third-Interval IST Program are retained at Vermont Yankee.

2.1 Test Deferral Justifications in accordance with Paragraphs 4.2.1.2 and 4.3.2.2 of Part 10 of the Code, certain valves are full stroke exercised during Cold Shutdown conditions when the valve cannot be exercised during Normal Operation. When the valve cannot be exercised during Cold Shutdown conditions, then the valve is full stroke exercised during Refueling Outages.

The technical justification for exercising a valve during Cold Shutdowns or Refueling Outages, rather than during Normal Operations, is provided in a Cold Shutdown or Refueling Outage Justification.

Valves tested during Cold Shutdowns or Refueling Outages shall be scheduled and tested in accordance with Paragraphs 4.2.1.2 and 4.3.2.2 of Part 10 of the Code.

Cold Shutdown and Refueling Outage Justifications are numbered in a "XXJ-VNN, Revision Z" format, where:

l XXJ: CSJ for Cold Shutdown Justifications, i ROJ for Refueling Outage Justifications.

V: for Valves. I NN: A unique sequential number, (e.g. CSJ-V03, Rev 0, would be the third Cold l Shutdown Justification for Valves).

Z: Revision Status.

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Verm=t Yankee Nucl:ar Pcwer Stati::n Inservice Testing Prcgram

l 2.2 Relief Requests O

i Specific requests for relief are included in accordance with 10 CFR 50.55a(a)(3), (f)(5) 4 and (f)(6). Where conformance with the requirements of the Code have been determined to be impracticable, alternate testing is proposed that would provide an acceptable level of quality and safety. Where conformance with the requirements of the Code would re.sult in hardship or unusual difficulty without a compensating increase in the level of i quality and safety, alternate testing is proposed that would provide useful information to assess the operational readiness of the component tested.

i The Relief Requests define the component (s) and test (s) involved, the basis for relief, the proposed alternate testing, and the status of the USNRC evaluation to the Relief Request.

If testing requirements are determined to be impracticable during the course of the interval, additional or modified Relief Requests will be submitted in accordance with 10 i CFR 50.55a(f)(4)(iv).

i Relief Requests are numbered in a "RR-YNN, Revision Z" format, where:

RR: for Relief Request.

i Y: P for Pumps, V for Valves.

NN: A unique sequential number, (e.g. RR-V09, Rev 0, would be the ninth Relief Request for Valves).

Z: Revision Status.

2.3 Flow Diagrams Table 2-1 provides a listing of those Flow Diagrams which depict the components, subsystems, and systems contained within the Third-Interval IST Program. Drawing number G-191155 explains the symbols and designations provided on the Flow Diagrams.

The Safety Classifications depicted on the Flow Diagrams are subj:ct to the limitations stated in the Vermont Yankee Safety Classification Manual.

"For Information Only" copies of the Flow Diagrams shall be provided with each USNRC Submittal. The copies shall be current as of the date of the IST Program Plan submittal.

Controlled copies of the Flow Diagrams are maintained at Vermont Yankee to ensure that all system additions and modifications are addressed in the Third-Interval IST Program.

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Verm:nt Yc kee Nuclear Pcwer Stati:n Inservice Testing Progrcm Table 2-1 LIST OF INSERVICE TESTING FLOW DIAGRAMS (sorted by flow diagram number)

Flow Diagram System Number Name G-191155 Piping and Instrument Symtols G-191157 Sh 1 Condensate Feedwater and Air Evacuation G-191159 Sh 1 Service Water System G-191159 Sh 2 Service Water System G-191159 Sh 3 RCW Cooling Water System G-131159 Sh 5 Recirculation Pump Cooling Water G-191160 Sh 3 Instrument Air System G-191160 Sh 4 Instrument Air System G-191160 Sh 7 Diesel Generator Starting Air System G-191160 Sh 8 Instrument Air System G-191162 Sh 2 Miscellaneous Systems - Fuel Oil G-191165 Sampling System G-191167 Nuclear Boiler G-191168 Core Spray System .

l G-191169 Sh 1 High Pressure Coolant Injection System 1

G-191169 Sh 2 High Pressure Coolant Injection System G-191170 Control Rod Drive Hydraulic System G-191171 Standby Liquid Control System )

G-191172 Residual Heat Removal System ll Revision 18 Section 2 Page 3 of 4

Verm:nt Ycnkee Nuclear Power Stati:n Ins:rvice Testing Prcgram Table 2-1 l i

I LIST OF INSERVICE TESTING FLOW DIAGRAMS (sorted by flow diagram number)

Flow Diagram System Number Name G-191173 Sh 1 Fuel Pool Cooling & Cleanup System G-191173 Sh 2 Fuel Pool Cooling & Cleanup System G-191174 Sh 1 Reactor Core Isolation Cooling System G-191174 Sh 2 Reactor Core Isolation Cooling System i

G-191175 Sh 1 Primary Contaiament & Atmosphere Control  ;

G-191177 Sh 1 Radwaste Systems G-191178 Sh 1 Reactor Water Clean Up System G-191237 Sh 2 HVAC - Turbine, Service & Control Room Building G-191238 HVAC - Reactor Building G-191267 Sh 1 Nuclear Boiler Vessel Instrumentation G-191267 Sh 2 Nuclear Boiler Vessel Instrumentation

VY-E-75-002 Containment Atmosphere Dilution System 5920-271 Neutron Monitoring System l l

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Vcrm: t Yankee Nuclear P wer Stati:n Ins:rvice Testing Prcgram 3.0 References a) License No. DPR-28 (Docket No. 50-271).

b) United States Code of Federal Regulations, Title 10 Chapter 1, Part 50, Section 50.55a (57FR34666, dated August 6,1992).

c) ASME Boiler and Pressure Vessel Code,Section XI, Division 1, " Rules for Inservice Inspection of Nuclear Power Plant Components," 1989 Edition.

d) ASME/ ANSI Standard OMa-1988 Addenda to ASME/ ANSI OM-1987, " Ope: ration and Maintenance of Nuclear Power Plants" I i

e) Letter, Mr. P.M. Sears, USNRC, to Mr. L.A. Tremblay, VYNPC, " Vermont Yankee Nuclear Power Station, Approval of the Use of ASME/ ANSI Standard OMa-1988 With Clarification," NVY 92-161, dated September 2,1992.

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f) Letter, Mr. J F. Pelletier, VYNPC, to Document Control Desk, USNRC, " Vermont Yankee Nuclear Power Corporation Inservice Testing Program Update," BVY 92-98,  ;

dated August 13,1992.

l g) I Letter, Mr. J.G. Partlow, USNRC, to Ali Holders of Light Water Reactor Operating i n Licenses and Construction Permits, " Supplement to Minutes of the Public Meetings on o Generic Letter 89-04", dated September 26,1991, i

h) Letter, Mr. J.G. Partlow, USNRC, to All Holders of Light Water Reactor Operating '

Licenses and Construction Permits, " Minutes of the Public Meetings on Generic Letter i 89-04", NVY 89-239, dated October 25,1989.

i) Letter, Mr. W.P. Murphy, VYNPC, to Document Control Desk, USNRC, " Response to USNRC Generic Letter 89-04: Guidance on Developing acceptable Inservice Testing Programs", BVY 89-90, dated October 3,1989.

j) Letter, Mr. S.A. Varga, USNRC, to All Holders of Light Water Reactor Operating Licenses and Construction Permits, " Guidance on Developing Acceptable Inservice Testing Programs (Generic Letter 89-04)", NVY 89-75, dated April 3,1989.

k) Letter, Mr. W.P. Murphy, VYNPC, to Dr. Thomas E. Murley, USNRC, " Response to USNRC Generic Letter 87-06: Periodic Verification of Leak-Tight Integrity of Pressure Isolation Valves", FVY 87-64, dated June 11,1987.

1) Letter, Mr. R.W. Reid, USNRC, to Mr. R.H. Groce, YAEC, "NRC Staff Guidance for Preparing Pump and Valve Testing Program Descriptions and Associated Relief Requests Pursuant to 10 CFR 50.55a(g)," dated January 10,1978.

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Verm=t Ycnkee Nuclear Pcwer Stati:n Ins:rvice Testing Program l l

J 3.0- References (cont.)

. g) t m) Letter, Mr. R.W. Reid, USNRC, to Mr. R.H. Groce, YAEC, "NRC Staff Guidance for I Complying with Certain Provisions of 10 CFR 50.5',a(g), 'hservice Inspettion l Requirements'," dated November 17,1976. l n) Vermont Yankee Final Safety Analysis Report.

o) Vermont Yankee Technical Specifications.

p) Vermont Yankee Component Testing Program Plan.

q) Letter, Mr. J.P. Pelletier, VYNPC, to Document Control Desk, USNRC, " Submittal of Vermont Yankee Nuclear Power Corporation Third-Interval Inservice Testing Progran  :

Plan and Safety Evaluation Responses," BVY 92-133, dated November 30,1992. -

I r) Letter, Mr. D.H. Dorman, USNRC, to Mr. J.P. Pelletier, VYNPC, " Relief from th:

ASME Section XI Requirement to Update (On a 120-Month Interval) the Inservic Inspection and Testing Programs at Vermont Yankee (TAC No. M85067)," NVY 93 031, dated April 6,1993.

s) Letter, Mr. D.H. Dorman, USNRC, to Mr. D.A. Reid, VYNPC, " Safety Evaluation of' the Inservice Testing Program Relief Requests for Pumps and Valves, Vermont Yankee

- Nuclear Power Station (TAC No. M85067)," NVY 93-151, dated September 3,1993.

I t) Letter, Mr. D.A. Reid, VYNPC, to Document Control Desk, USNRC, " Proposed l Change No.168 to the Vermont Yankee Technical Specifications - Auxiliary Electric l Power System Technical Specifications and Associated Revision to the Vermont Yankee Inservice Testing Program," BVY 93-30, dated August 4,1993.

u) Letter, Mr. W.R. Butler, USNRC, to Mr. D.A. Reid, VYNPC, " Issuance of Amendment i No.138 to Facility Operating License No. DPR-28, Vermont Yankee Nuclear Power i Station (TAC No. M87171)," NVY 94-45, dated March 22,1994.

. v) Letter, Mr. D.H. Dorman, USNRC to Mr. D.A. Reid, " Safety Evaluation of the Inservice i Test Program Relief Requests for Pumps and Valves, Vermont Yankee Nuclear Pov'er Station (TAC No. M85067)".

w) Letter, Mr. P.F. McKee, USNRC to Mr. D.A. Reid, " Safety Evaluation of Relief Requests and Action Item Responses for the Third Interval Pump and Valve Inservice Testing Program - Vermont Yankee Nuclear Power Station, (TAC No. M91450)," NVY 95-6d, dated June 12,1995.

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Verm:nt Ycnkee Nuclear Pcwcr Statim Inservice Testing Prcgram )

l 3.0 References (cont.)

J O x) Letter, Mr. P.F. McKee, USNRC to Mr. D. A. Reid, " Safety Evaluation of Relief Request RR-V12 for the Third Interval Pump and Valve Inservice Testing Program - Vermont 4

Yankee Nucien. Power Station (TAC No. M92018)," NVY 95 100, dated July 27,1995.

, y) USNRC NUREG 1482, dated April 1995, " Guidelines for Inservice Testing at Nuclear Power Plants".

z) USNRC NUREG/CR-6396, dated February 1996, " Examples, Clarifications, and )

Guidance on Preparing Requests for Relief from Pump and Valve Inservice Testing l Requirements".

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Verm:nt Ycnkee Nuclear P;wer Station Inservice Testing Pr: gram 4.0 Pump Inservice Testing Program 4.1 Scope f

The Third-Intental Pump Inservice Testing (IST) Program establishes testing requirements to assess the operational readiness of those Safety Class 1,'2, and 3 centrifugal and positive displacement pumps, provided with a Safety Class Electrical power source, which are required to:

a) Shut down the reactor to the cold shutdown condition, b) Maintain the reactor in the cold shutdown condition, or c) Mitigate the consequences of an accident.

Excluded from the above are:

a) - Drivers, except where the pump and driver form an integral unit and the pump

, bearings are in the driver;

. b) Pumps that are provided with a Safety Class Electrical power source solely for system design or operating convenience.

tO V 4.2 Pump Program Listing Table 4-1, " Pump Listing" lists all pumps included in the Third-Interval Pump IST Program.

This Table identifies all pumps subject to inservice testing, the inservice test parameters, testing frequency, and any applicable relief requests and/or remarks. The column headings in Table 4-1 are listed and explained below:

Pump Number The unique number that identifies the pump. The pump number corresponds to the Maintenance Planning and Control (MPAC) Database utilized at Vermont Yankee. Table 4-1 is sorted by Pump Number.

Nomenclature The common name for the pump.

Drawing The Flow Diagram which d:picts the pump. If the pump appears on multiple Flow Diagrams, then the primary Flow Diagram identifier is listed.

Revision 18 Section 4 Page 1 of 31

Vcrm:nt Ycnkee Nucl:cr Power Statira Ins:rvice Testing Prcgram 4.2 Pump Program Listing (cont.)

Dwg Coor The coordinate location (e.g., D-05) on the Flow Diagram where the pump appears.

Safety Class The safety classification of the pump, as determined in accordance with administrative  !

procedure AP 0014, " Safety Class Determination Instructions," and the Vermont Yankee l Safety Classification Manual. .

1 Pump Type  !

I The pump type:

i CENT for Centrifugal pumps.

VERT for Verticalline shaft pumps.

PD-RECIP for Positive Displacement Reciprocating pumps.

PD-ROT for Positive Displacement Rotary pumps. l l

Test Type The Inservice Test Parameters to be determined and recorded during testing performed in accordance with Paragraph 5 of Part 6 of the Code. The abbreviations correspond to those provided in Table 2 of Part 6 of the Code:

N for Speed,(ifvariable speed).

dP for differential Pressure, (for centrifugal and venical line shaft pumps).

P for discharge Pressure, (for positive displacement pumps).

Q for Flow Rate.

Yv for Vibration, velocity, peak.

Test Freq The test frequency associated with each test type.

OC for Once Per Operating Cycle Q for Quarterly ,

RO for Reactor Refueling Outage )

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Revision 18 Section 4 Page 2 of 31

1 Verm:nt Ycnkee Nuclear Pcwsr Stati:n Inservice Testing Pr: gram l

^ 4.2 Pump Program Listing (cont.) l ReliefRequest The Relief Request number associated with the subject pump and test type or test '

i frequency, where applicable. l l  !

Remarks  !

I i Clarifying comments or other remarks related to the subject pump and test type or test !

l frequency, where applicable. {

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O O O Vcrmont Yc kee Nuclear Pcwcr Stati=a Inservice Testing Program Table 4-1 Pump Listing Pump Number Nomenclature Drawing Dwg Safel Pump Speed T T Rei Remarks p

P-10-1 A RIIR(LPCI) Pump G-191172 L-05 2 CENT Fixed dP Q n/a Q Q da I Vv Q n/a P-10-1 B RIIR (LPCI) Pump G-191172 L-12 2 CENT Fixed dP Q n/a Q Q n/a j Vv Q n/a t P-10-1C RHR (LPCI) Pump G-191172 J-05 2 CENT Fixed dP Q n/a Q Q n/a VV Q n/a P-10-1D RIIR (LPCI) Pump G-191172 J-12 2 CENT Fixed dP Q n/a Q Q Wa Vv Q n/a P-19-2A Standby Fuel Pool Cooling G-191173 Sh 2 G-12 3 CENT Fixed dP Q n/a  ;

Nmp Q Q Wa  !

Vv Q n/a P-19-2B Standby Fuel Pool Coolers G-191173 Sh 2 I-12 3 CENT Fixed dP Q n/a Pump n/a Q Q Vv Q n/a P-213-1 A RCIC Gland Seal G-191174 Sh 2 K-13 3 PD-RECIP Fixed SKID Q n/a Condensate Pump P-44-1A IIPCI(Booster) Pump G-191169 Sh 2 G-10 2 CENT Vari dP Q RR-P02 RR-P03 N Q n/a Q Q RR-P02 Vv Q RR-PO4 Revision 18 Section 4 Page 4 of 31

O @ O Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 4-1 Pump Listing Pump Number Nomenclaturt Drawing Dwg Safet Pump Speed Ted Ted RelM Remarks Coor y Type Type Freq Requed P-44-1 B HPCI(High Pressure) Pump G-191169 Sh 2 G-11 2 CENT Vari dP Q RR-P02 RR-P03 N Q n/a Q Q RR-P02 Vv Q RR-PO4 P-45-I A SLC Pump G-191171 11-0 8 2 PD-RECIP Fixed P Q n/a l Q Q n/a Vv Q n/a P-45-1 B SLC hmp 0-191171 J-08 2 PD-RECIP Fixed P Q n/a Q Q n/a Vv Q n/a P-46-1 A CS Pump G-191168 J-l 1 2 CENT Fixed dP Q RR-POS Q Q n/a Vv Q n/a P-46-1 B CS Pump G-191168 J-14 2 CENT Fixed dP Q RR-POS Q Q n/a Vv Q n/a P-47-1 A RCIC Pump G-191174 Sh 2 F-08 2 CENT Vari dP Q RR-P06 N Q n/a Q Q n/a Vv Q n/a P-59-I A RBCCW Pump G-191159 Sh 3 A-07 3 CENT Fixed dP Q RR-P08 Q Q n/a Vv Q n/a P-59-1 B RBCCW Pump G-191159 Sh 3 B-07 3 CENT Fixed dP Q RR-P08 Q Q n/a Vv Q n/a Revision 18 Section 4 Page 5 of 31

v sJ Vcrmont Yc kee Nuclear Pow;r Station Inservice Testing Program Table 4-1 Pump Listing Pump Number NomencInture Drawing Dwg Safet Pump Test Test Relief Spd Remarks Coor y Type Type Freq Request P-92-1A DFOT Pump G-191162 Sh 2 E-05 3 PD-ROT Fixed P Q n/a Q Q RR-P09 Q OC RR-P09 Vv Q n/a P-92-1 B DFOT Pump G-191162 Sh 2 D-05 3 PD-ROT Fixed P Q n/a Q Q RR-P09 Q OC RR-P09 Vv Q n/a P-72-2A Dievi Generator DG-1-I A G-191162 Sh 2 F-11 3 PD-ROT Vari SKID Q n/a Engine Driven Fuel Oil Pump P-92-2B Diesel Generator DG-1-1B G-191162 Sh 2 D-11 3 PD-ROT Vari SKID Q n/a Enginc Driven Fuel Oil Pump .

SP-1 Control Room liVAC G-191237 Sh 2 E-4 3 CENT Fixed dP Q n/a Chilled Water Pump Q Q n/a Vv Q n/a Revision 18 Section 4 Page 7 of 31

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Vermrt Ycnkee Nucl:ar Pcwer Statin Inservice Testing Pr:grcm 4.3 Pump Notes O 1. In accordance with the June 12,1995 Safety Evaluation Report (Reference w) the following commitments were made relating to inservice testing of the Vermont Yankee Service Water Pumps (P7-1 A through P7-ID):

Quarterly Testing (As-found for each pump)

1. Differential Pressure Measurement (for information only)

2. Full Spectrum Vibration Signatures (for information only)

3. Overall Vibration Measurements (compared to code limits)

Once Per Operating Cycle

1. One of the four service water pumps will be disassembled, inspected and refbrbished as necessary. Additionally, in no case shall a service water pump exceed a period of 4 cycles of operation without being disassembled, inspected and refurbished as necessary.

Refueling Outage

1. Full code specified IST te.= ting will be performed at a flow rate greater than or equal to design flow. l

2. A head curve will be generated to provide information so that the performance of the pump can be compared to the degree possible with the as-found quarterly data and previous refueling outage head curve d.ata.

The approval of relief request RR-P01 is not subject to the performance of shutoff head testing or motor amperage monitoring as originally proposed in RR-P01.

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Revision 18 Section 4 Page 8 of 31

1 Verm:nt Yankee Nuclear P:wer Stati:n Inservice Testing Prcgram I

RELIEF REOUEST O

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l' Number: RR-Pol, Revision 1 (Sheet 1 of4) l

SYSTEM

Service Water I l

COMPONENTS: 1 Pump Number Safety Class Drawing Number Dwg. Coord.

I P7-1 A l 3 G-191159 Sh 1 C-02 l

P7-1B 3 G-191159 Sh 1 B-02 P7-1C 3 G-191159 Sh 1 K-02

P7-1D 3 G-191159 Sh 1 J-02 )

These pumps are the station Service Water pumps. They have the safety function to provide l cooling water to systems and equipment required to operate under accident conditions and to

provide an inexhaustible supply of w;ter for standby coolant system operation.

EXAM OR TEST CATEGORY:

Flow Rate (Q).

CODE REOUIREMENT: Part 6 Para. 5.1 " Frequency ofInservice Tests" "An inservice test shall be run on each pump, nominally every 3 months, except as

, provided in paras. 5.3, 5.4, and 5 5."

REOUEST FOR RELIEF:

Reliefis requested on the basis that compliance with the Code requirements is impracticable and that the proposed alternatives would provide an acceptable level of quality and safety.

i During normal operations, neither differential pressure nor flow rate can be fixed or directly j measured. Pump vibration levels may also vary due to the inability to establish a repeatable

reference condition.

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REOUEST FOR RELIEF (cont.h i

The four Service Water pumps are vertical, two-stage, centrifugal-type pumps which are {

submerged in and take suction from the Connecticut River. They supply all the station Service l

Water System requirements. The station Service Water System is a dual header system using two '

parallel headers each containing two pumps. The two parallel headers supply both the turbine and reactor auxiliary equipment, including the Residual Heat Removal Service Water System. A

header interconnection is provided downstream of the pumps. Normally, the valves in the i interconnecting line are open, permitting any of the pumps to supply the cooling water to both l headers and to balance system operation. In addition, a cross-tie is provided to the non-nuclear  ;

safety station Fire Protection System. The 12-inch cross-tie valve is normally closed, with a 1-inch ,

cross-tie and a restricting orifice providing pressurization of the Fire Protection System h;ader.

The Service Water System contains both automatic temperature and flow control valves used to independently rgulate the cooling provided to the various turbine and reactor auxiliary i equipmer.t. Due to seasonal variations in Connecticut River water temperature and level and i constantly changing heat loads, the system resistance and flow rate vary. The number of station  ;

Service Water pumps in operation is also varied dependent on system requirements. Due to these i variations and the need to maintain proper cooling, it is considered impracticable to establish

repeatable reference values during quarterly inservice testing. l l As a result, it is impracticable to directly measure pump flow rate on a quanerly and cold l . shutdown frequency. Sufficient straight sections of piping are required to properly measure flow I rate, through the use of either permanently or temporarily installed instrumentation, such as non- l 2 intrusive flow measurement devices. )

i The only sufficient straight sections of piping in each of the two parallel headers exist between the I intake structure and the entrance to the reactor building. Use of this piping is considered impracticable because.

i a) These sections of the two parallel headers are buried piping.

. b) As discussed above, each parallel header is commen to two pumps and is cross-connected  !

to the other header. Thus, in order to measure flow rate for one pump, the parallel pump in i the same header would have to be secured and the valves in the header interconnecting line closed. This would result in single pump operation for the portion of the station Service ,

Water System supplied by the pump being tested.  !

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Verm:nt Ycnkee Nucle:r Pcwcr Stati:n Inservice Testing Program RELIEF REOUEST Number: RR-P01, Revision 1 (Sheet 3 of 4)

REOUEST FOR RELIEF (cont.h 4

i c) All four station Senice Water pumps are required to be operating durmg power operations and cold shutdowns during approximately 7 months of the year to meet cooling load i

requirements.

i Based on the above, significant redesign and modification of the station Service Water System  ;

would be required to obtain direct measurement of pump flow. Such redesign and modification l

. would be costly and burdensome to Vermont Yankee. '

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l A review of the historical test data for these pumps indicates that theses pumps are highly reliable and have not been susceptible to frequent failures. Plant operating experience has shown that the l performance of the Senice Water pumps degrades slowly over an extended period due to normal system wear.

ALTERNATE METHOD: 1 On a quarterly basis, during plant operation and cold shutdowns an as-found test will be  !

2 performed by measuring pump differential pressure and motor vibration and the data will be compared to the degree possible with test data taken during the refueling outage head curve data.

During each refueling outage, pump flow, differential pressure and vibration will be measured at a ,

reference condition which will meet or exceed the required design conditions for the pump. A '

temporary flow test loop installed on the plant fire protection system will be utilized to directly i measure pump flow. This will provide a mechanism to assess the hydraulic condition of the pump ar.d to detect pump degradation against the code required limits. An additional reference  !

condition will be established with the pump at a dead head condition where pump vibration and i differential pressure can be measured for comparison in the event that maintenance is required to

)

be performed between refueling outages.

Additionally, a head curve will be generated each refueling outage to provide information so that the performance of the pump can be compared to the degree possible with the as-found quarterly test data and previous refueling outage head curve data. Overall peak and full spectrum vibration

]

measurements will be also be taken at each of the points used to generate the head curve to  :

provide additional operational information.

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. . RELIEF REOUES__T__

Number: RR "01, Revision 1 (Sheet 4 of 4) l l

ALTERNATE METHOD (cont.h l l

A review of the operating history of these pumps has shown that they are highly reliable and have not been susceptible to frequent failures. In order to provide additional assurance of proper pump operation and mechanical condition, an enhanced maintenance / monitoring program for these pumps will be established which will include the following:

1. Service water pump motor amperage will be monitored on a once per shift frequency during periods when the pumps are in operation.

2. Full spectrum vibration signatures of the accessible motor bearing points will be obtained i and analyzed on a quarterly basis.'

3. At least one service water pump will be partially disassembled, inspected and refurbished as required every operating cycle. Additionally, in no case shall a service water pump exceed a l period of 4 cycles of operation without being partially disassembled, inspected and I refurbished as required.

USNRC EVALUATION STATUS:

Provisional relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P01, Revision 0. Approval ofRevision 1 of Relief Request RR-P01, was granted in the June 1995

\ SER [ Reference (w)]. Granting RR-P01 Rev.1 was not subject to monitoring pump motor amperage each shift or refueling outage shutoff head testing as specified in the relief request.

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Verm:nt Yankee Nucl:ar Pcw:r Station Ins:rvice Testing Program RELIEF REOUEST O Number: RR-P02, Revision 0 (Sheet 1 of 2)

SYSTEM: High Pressure Coolant Injection COMPONENTS:

Pump Number Safety Class Drawing Number Dwg. Coord.

P44-1A 2 G-191169 Sh 2 G-11 P44-1B 2 G-191169 Sh 2 G-10 P44-1 A and P44-1B are the High Pressure Coolant Injection (HPCI) main (high pressure) and booster (low pressure) pumps, respectively. They have the safety functions to operate in series to  !

provide 1) adequate core cooling and reactor vessel depressurization following a small break loss of coolant accident, and 2) reactor pressure control during reactor shutdown and isolation.

EXAM OR TEST CATEGORY:

l Differentia! Pressure (dP).

Flow Rate (Q).

CODE REOUIREMENT: Part 6 Para. 4.6.2.2 " Differential Pressure" "When determining differential pressure across a pump, a differential pressure gauge or transmitter that provides direct measurement of pressure difference or the difference between the pressure at a point in the inlet pipe and the pressure at a point in the discharge pipe, may be used."

Para. 4.6.5 " Flow Rate Measurement" "When measuring flow rate, use a rate or quantity meter installed in the pump test circuit. If a meter does not indicate the flow rate directly, the record shall include the method used to reduce the data."

REOUEST FOR RELIEF:

Reliefis requested on the basis that the proposed alternatives would provide an acceptable level of quality and safety.

There is no means of measuring the differential pressure or the flow rate generated by each of these pumps individually. Installation ofindependent means of measurement for each pump is O considered impracticable and would be costly and burdensome to Vermont Yankee.

Revision 18 Section 4 Page 13 of 31

Vcrm nt Ycnkee Nuclear Pswsr Stati:n Ins:rvice Testing Program RELIEF REOUEST Number: RR-P02, Revision 0 (Sheet 2 of 2) '

ALTERNATE METHOD:

l Differential pressure and flow rate parameters will be measured across both HPCI pumps as an integral unit. Vibration monitoring will be performed for each pump individually. Both pumps will be inspected and repaired as necessary if any abnormal conditions in differential pressure or flow rate occur.

USNRC EVALUATION STATUS:

Relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P02, Revision 0.

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1 Verm:nt Yankee Nucl:ar Pcwer Stati:n Ins rvica Testing Program RELIEF REOUEST Number: RR-P03, Revision 0 (Sheet 1 of 2) 1 SYSTEM; High Pressure Coolant Injection COMPONENTS:

Pump Number Safety Class Drawing Number Dwg. Coord.

P44-1A 2 G-191169 Sh 2 G-11 I P44-1B 2 G-191169 Sh 2 G-10 l P44-1A and P44-1B are the High Pressure Coolant Injection (HPCI) main (high pressure) and ,

booster (Iow pressure) pumps, respectively. They have the safety functions to operate in series to l provide 1) adequate core cooling and reactor vessel depressurization following a small break loss of coolant accident, and 2) reactor pressure control during reactor shutdown and isolation.

EXAM OR TEST CATEGORY:

Differential Pressure (dP).

CODE REOUIREMENT: Part 6 O Para. 4.6.1.2(a) " Range" "The full-scale range of each analog instrument shall not be greater than three times the reference value."

i d

Para. 4.6.2.2 " Differential Pressure"

! "When determining differential pressure across a pump, a differential pressure i

gauge or transmitter that provides direct measurement of pressure difference or the

{

difference between the pressure at a point in the inlet pipe and the pressure at a I point in the discharge pipe, may be used."

REOU'EST FOR RELIEF:

Reliefis requested on the basis that the proposed alternatives would provide an acceptable level 2

of quality and safety.

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• U Revision 18 Section 4 Page 15 of 31

Verm:nt Yankee Nucl:ar Pcwer Stati:n Inservice Testing Pr: gram l l

q RELIEF REOUEST Number: RR-P03, Revision 0 (Sheet 2 of 2)

REOUEST FOR RELIEF (cont.):

Differential pressure across the HPCI pumps is determined by the difference between pressure measurements taken at a point in the inlet pipe and at a point in the discharge pipe as allowed by Paragraph 4.6.2.2 of Part 6 of the Code. The installed HPCI pump inlet pressure indicators are

designed to provide adequate inlet pressure indication during all expected operatin; and post accident conditions. The full scale range, 85 psig, is sufficient for a post accident condition when the suppression chamber is at the maximum pressure. This, however, exceeds the full-scale range limit of three times the suction pressure reference value as required by Paragraph 4.6.1.2(a) of i Part 6 of the Code (Value = approximately 26 psig, Limit = 78 psig).

The suction pressure measurement is used to verify prescribed NPSH requirements and to determine pump differential pressure. The installed gauges are calibrated to within +/- 1.17% l accuracy (FS), thus the maximum variation m measured suction pressure due to inaccuracy would be +/- 0.99 psi. This is considered to be suitable for determining that adequate NPSH is available I for HPCI pump operation. I Pump discharge pressure during testing is approximately 1170 psig, which results in a calculated l

} differential pressure of approximately 1144 psig. The resulting inlet pressure inaccuracy of +/-

U 0.99 psi represents an error in differential pressure measurement of +/- 0.08% (0.99 psi /l144 psid i'

= 0.00086). This is consistent with Table 1 of Part 6 of the Code, which requires that instrument accuracy for differential pressure be better than 2% of full-scale.

ALTERNATE METHOD: l Differential pressure will be measured using the existing station system installed inlet pressure

! indicators.

USNRC EVALUATION STATUS:

Relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P03, Revision 0.

O Revision 18 Section 4 Page 16 of 31

Verm:nt Yankee Nuclear P wer Stati:n Ins:rvice Testing Prrgram RELIEF REOUEST O Number: RR-PO4, Revision 0 (Sheet 1 of 3)

SYSTEM: High Pressure Coolant Injection COMPONENTS:

Pump Number Safety Class Drawing Number Dwg. Coord.

P44-1A 2 G-191169 Sh 2 G-11 P44-1 A is the High Panure Coolant Injection (HPCI) main (high pressure) pump. The main pump has the safety function to operate in series with the booster pump, P44-1B, to provide 1) adequate core cooling and reactor vessel depressurization following a small break loss of coolant accident, and 2) reactor pressure control during reactor shutdown and isolation.

i EXAM OR TEST CATEGORY:

Vibration Velocity (Vv).

CODE REOUIREMENT: Part 6 a Para. 5.2(d) " Test Procedure"

i " Pressure, flow rate, and vibration (displacement or velocity) shall be determined and compared with corresponding reference values. All deviations from the reference values shall be compared with the limits given in Table 3 and corrective action taken as specified in para. 6.1."

REOUEST FOR RELIEF:

Reliefis requested on the basis that the proposed alternatives would provide an acceptable level of quality and safety.

Past testing and analysis performed on the HPCI System by Vermont Yankee, the pump manufacturer, and by independent vibration consultants has revealed characteristic pump vibration levels which exceed the acceptance criteria stated in Table 3 of Part 6 of the Code. This testing and analysis meets the intent of Paragraph 4.3 and footnote 1 of Part 6 of the Code.

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V Revision 18 Section 4 Page 17 of 31

Verm::nt Yankee Nuclear Pcwer Stati::n Inservice Testing Prtgram n RELIEF REOUEST U Number: RR-PO4, Revision 0 (Sheet 2 of 3)

REOUEST FOR RELIEF (cont.h The root causes of the higher vibration levels have been determined to be:

a) An acoustical resonance in the piping connecting the low pressure (LP) and high pressure (HP) pumps, and b) The presence of a structural resonance in the horizontal direction on the HP pump.

These resonance conditions are design related and have existed since initial pump installation.

They have been documented over a number ofyears of operating experience. '

An additional past contributor to the higher vibration levels was the excitation resulting from the blade pass frequency from the previously installed four vane impeller in the low pressure (LP) pump. In an effort to reduce /elindnate this effect, the four vane impeller was replaced with a five vane impeller daring the 1989 refueling outage. This replacement significantly reduced vibration levels in both the LP and HP pumps. However, due to the resonance effects referenced above, the HP pump vibration levels remain higher than the acceptance criteria stated in Table 3 of Part 6 of the Code.

Although existing vibration levels in the HP pump are higher than standard acceptance criteria, they are acceptable and reflect the unique operating characteristics of the HPCI pump. It has been concluded that there are no major vibrational concerns that would prevent the HPCI pump from performing its intended function.

ALTERNATE METHOD:

To allow for practicable vibration monitoring of the HPCI HP pump, alternate vibration acceptance criteria are required. Fud spectrum vibrational monitoring will be performed during each quarterly test and the following criteria will be used for the HP pump:

Iest Parameter Acceptable Range Alert Range Required Action Range V, 5; 2.5 Vr > 2.5 Vr to and > 6 Vr but not including 6 Vr or

> 0.675 in/sec. but not > 0.70 in/sec.

> 0.70 in/sec.

In addition, the resonance peaks will be evaluated during each test and will have an Acceptable Range upper limit of 1.05 Vr and an Alert Range upper limit of 1.3 Vr.

The standard acceptance criteria of Table 3 of Part 6 will be applied to the LP pump.

Revision 18 Section 4 Page 18 of 31

Vcrmr.t Ycnkee Nuclear Pcwer Stati:a Ins:rvice Testing Prcgram l

RELIEF REOUEST Number: RR-PO4, Revision 0 (Sheet 3 of 3)

. USNRC EVALUATION STATUS:

Relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P04, l Revision 0. <

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Vcrm:nt Ycnkee Nucl:nr Pcwsr Stati:n Ins:rvice Testing Program i

l RELIEF REOUEST l Number: RR-POS, Revision 0 (Sheet 1 of 2) 1 SYSTEM: Core Spray

COMPONENTS

Pump Number Safety Class Drawing Number Dwg. Coord.

i P46-1A 2 G-191168 J-11 i P46-1B 2 G-191168 J-14 i

q P46-1 A & B are the low pressure Core Spray pumps. They have the safety function to operate to provide adequate core cooling following a loss ofcoolant accident and reactor depressurization.

EXAM OR TEST CATEGORY:

Differential Pressure (dP).

CODE REOUIREMENT: Part 6 i Para. 4.6.1.2(a) " Range" O

"The full-scale range of each analog instrument shall not be greater than three l times the reference value."

Para. 4.6.2.2 " Differential Pressure" '

4 "When determining differential pressure across a pump, a differential pressure j gauge or transmitter that provides direct measurement of pressure difference or the difference between the pressure at a point in the inlet pipe and the pressure at a point in the discharge pipe, may be used."

REOUEST FOR RELIEF:

Reliefis requested on the basis that the proposed alternatives would provide an acceptable level i

of quality and safety. i 1

Differential pressure across the Core Spray pumps is determined by the difference between pressure measurements taken at a point in the inlet pipe and at a point in the discharge pipe as ,

allowed by Paragraph 4.6.2.2 of Part 6 of the Code. The installed Core Spray pump inlet pressure 4

indicators are designed to provide adequate inlet pressure indication during all expected operating and post accident conditions. The full scale range, 60 psig, is sufficient for a post accident condition when the suppression chamber is at the maximum pressure. This, however, exceeds the s full-scale range limit of three times the suction pressure reference value as required by Puagraph 4.6.1.2(a) ofPart 6 of the Code (Value = approximately 7.5 psig, Limit = 22 psig).

. Revision 18 Section 4 Page 20 of 31

Ve met Yankee Nuclear Pcwer Stati:n Ins:rvice Testing Pr: gram RELlif REOUEST Number: RR-POS, Revision 0 (Sheet 2 of 2)

REOUEST FOR RELIEF:

The suction pressure measurement is used to vedfy prescribed NPSH requirements and to determine pump differential pressure. The installed gauges are calibrated to within +/- 1.6%

accuracy (FS), thus the maximum vadation in measured suction pressure due to inaccuracy would be +/- 0.96 psi. This is considered to be suitable for determining that adequate NPSH is available for Core Spray pump operation.

Pump diccharge pressure during testing is approximately 240 psig, which results in a calculated differential pressure of approximately 232.5 psig. The resulting inlet pressure inaccuracy of +/-

0.96 psi represents an error in differential pressure measurement of +/- 0.41% (0.96 psi /232.5 psid = 0.0041). This is consistent with Table 1 of Part 6 of the Code, which requires that instrument accuracy for differential pressure be better than 2% of full-scale.

ALTERNATE METHOD:

Differential pressure will be measured using the existing station system installed inlet pressure indicators.

USNRC EVALUATION STATUS:

Relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P05, Revision 0.

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Revision 18 Section 4 Page 21 of 31 l

Verm:nt Yenkee Nuclear Pcwer Statirn Inservica Testing Pr:gr:m RELIEF REOUEST O

v  !

Number: RR-P06, Revision 0 (Sheet 1 of 2)

SYSTEM: Reactor Core Isolation Cooling  :

l

[QMPONENTS: i i

Pump Number Safety Class Drawing Number Dwg. Coord.

P47-1A 2 G-191174 Sh 2 F-08 P47-1 A is the Reactor Core Isolation Cooling (RCIC) pump. It has the safety function to operate to provide makeup water to the reactor vessel during shutdown and isolation in order to prevent  ;

the release of radioactive materials to the environs as a result ofinadequate core cooling. l EXAM OR TEST CATEGORY: )

Differential Pressure (dP).

CODE REOUIREMENT: Part 6 Para. 4.6.1.2(a) " Range" "The full-scale range of each analog instmment shall not be greater than three times the reference value."

Para. 4.6.2.2 " Differential Pressure" -

"When determining differential pressure across a pump, a differential pressure gauge or transmitter that provides direct measurement of pressure difference or the difference between the pressure at a point in the inlet pipe and the pressure at a point in the discharge pipe, may be used."

REOUEST FOR RELIEF:

Reliefis requested on the basis that the proposed alternatives would provide an acceptable level j ofquality and safety.

Differential pressure across the RCIC pump is determined by the difference between pressure measurements taken at a point in the inlet pipe and at a point in the discharge pipe as allowed by Paragraph 4.6.2.2 of Part 6 of the Code. The installed RCIC pump inlet pressure indicators are designed to provide adequate inlet pressure indication during all expected operating and post i accident conditions. The full scale range, 85 psig, is sufficient for a post accident condition when  !

the suppression chamber is at the maximum pressure. This, however, exceeds the full-scale range limit c f three times the suction pressure reference value as required by Paragraph 4.6.1.2(a) of Part 6 of the Code (Value = approximately 20 psig, Limit = 60 psig).

Revision 18 Section 4 Page 22 of 31

Verm:nt Ycnkee Nuclear Pcwer Stati:n Inservice Testing Pr:gr m RELIEF REOUEST Number: RR-P06, Revision 0 (Sheet 2 of 2) 1 REOUEST FOR RELIEF:

The suction pressure measurement is used to verify prescribed NPSH requirements and to determine pump differential pressure. The installed gauges are calibrated to within +/- 1.17%

accuracy (FS), thus the maximum variation in measured suction pressure due to inaccuracy would be +/- 0.99 psi. This is considered to be suitable for determining that adequate NPSH is available for RCIC pump operation.

Pump discharge pressure during testing is approximately 1130 psig, which results in a calculated l differential pressure of approximately 1110 psig. The resulting inlet pressure inaccuracy of +/- ,

0.99 psi represents an error in differential pressure measurement of +/- 0.09% (0.99 psi /1110 psid

= 0.00089). This is consistent with Table 1 of Part 6 of the Code, which requires that instrument i accuracy for differential pressure be better than 2% of full-scale.  !

ALTERNATE METHOD:

Differential pressure will be measured using the existing station system installed inlet pressure indicators.

USNRC EVALUATION STATUS:

Relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P06, Revision 0.

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Verm:nt Ycnkee Nuclear Pcwcr Stati:n Inservice Testing Program RELIEF REOUEST Number: RR-P07, Revision 0 (Sheet 1 of1)

THIS RELIEF REQUEST WAS DELETED IN REVISION 15 OF THE IST PROGRAM.

1 This ER number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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O Revision 18 Section 4 Page 24 of 31

Vcrm:nt Ycnkee Nuclear Powcr Stati:n Inservice Testing Prcgram RELIEF REOUEST Number: RR-P08, Revision 0 (Sheet 1 of 2)

I SYSTEM: Reactor Building Closed Cooling Water COMPONENTS: i Pump Number Safety Class Drawing Number Dwg. Coord.

P59-1A 3 G-191159 Sh 3 A-07 l P59-1B 3 G-191159 Sh 3 B-07 These pumps are the Reactor Building Closed Cooling Water (RBCCW) pumps. They have a safety function to provide cooling water to the safety related cooling loads served by the RBCCW system.

EXAM OR TEST CATEGORY:

Differential Pressure (dP).

l CODE REOUIREMENT: Part 6 Para. 4.6.1.2(a) " Range" "The full-scale range of each analog instrument shall not be greater than three times the reference value."

Para. 4.6.2.2 " Differential Pressure" "When determining differential pressure across a pump, a differential pressure i gauge or transmitter that provides direct measurement of pressure difference or the difference between the pressure at a point in the inlet pipe and the pressure at a point in the discharge pipe, may be used."

REOUEST FOR RELIEF: ,

Reliefis requested on the basis that the proposed alternatives would provide an acceptable level of quality and safety.

Differential pressure across the RBCCW pumps is determined by the difference between pressure measurements taken at a point in the inlet pipe and at a point in the discharge pipe as allowed by Paragraph 4.6.2.2 of Part 6 of the Code. The installed RBCCW pump inlet pressure indicators, with a full scale range of 30 psig, are designed to provide adequate inlet pressure indication n during all expected operating and post accident conditions. This, however, exceeds the full-scale V range limit of three times the suction pressure reference value as required by Paragraph 4.6.1.2(a) of Part 6 of the Code (Value = approximately 6.5 psig, Limit = 19 psig).

Revision 18 Section 4 Page 25 of 31

Verm=t Yankee Nucirr Powcr Statire Ins

rvice Testing Prcgram RELIEF REOUEST 4

Number

RR-P08, Revision 0 (Sheet 2 of 2)

REOUEST FOR RELIEF (cont.): l The suction pressure measurement is used to verify prescribed NPSH requirements and to determine pump differential pressure. The installed gauges are calibrated to within +/- 0.4%

accuracy (FS), thus the maximum variation in measured suction pressure due to inaccuracy would be +/ .12 psi. This is considered to be suitable for determining that adequate NPSH is available 4

for RBCCW pump operation.

Pump discharge pressure during testing is approximately 80 psig, which results in a calculated differential pressure of approximately 73.5 psig. The resulting inlet pressure inaccuracy of +/ .12 psi represents an error in differential pressure measurement of +/- 0.16% (.12 psi /73.5 psid =

0.0016). This is consistent with Table 1 of Part 6 of the Code, which requires that instrument accuracy for differential pressure be better than 2% of full-scale.

ALTERNATE METHOD: i l

l Differential pressure will be measured using the existing station system installed inlet pressure j indicators.

% I i USNRC EVALUATION STATUS:

Relief was granted in the September 1993 SER [ Reference (s)) for Relief Request RR-P08, Revision 0.

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Verm:nt Ycnkee Nucl:ar Pcw:r Stati:n Inservice Testing Program RELIEF REOUEST  !

o b Number: RR-P09, Revision 0 (Sheet 1 of 4)  ;

SYSTEM: Diesel Fuel Oil COMPONENTS:

Pump Number Safety Class Drawing Number Dwg. Coord.  !

P92-1A 3 G-191162 Sh 3 E-05 P92-1B 3 G-191162 Sh 3 D-05 These pumps are the Diesel Fuel Oil Transfer pumps. They have a safety function to provide diesel fuel oil to the diesel oil day tank during Emergency Diesel Generator gerations.

EXAM OR TEST CATEGORY:

Flow Rate (Q).

CODE REOUIREMENT: Part 6 Para. 4.6.5 "now Rate Measurement" "When measuring flow rate, use a rate or quantity meter installed in the pump test circuit. If a meter does not indicate the flow rate directly, the record shall include the method used to reduce the data."

Para. 5.1 " Frequency ofInservice Tests" "An inservice test shall be run on each pump, nominally every 3 months, except as provided in paras. 5.3, 5.4, and 5.5."

REOUEST FOR RELIEF:

l Relief is requested on the basis that compliance with the Code requirements would result m  ;

hardship or unusual difficulty without a compensating increase in the level of quality and safety l and that the proposed alternatives would govide an acceptable level of quality and safety. l During quarterly inservice testing, pump flow rate cannot be directly measured.

Revision 18 Section 4 Page 27 of 31

Verm::nt Yankee Nucirr Power St ti:n Inservice Testing Prcgram i

, RELIEF REOUEST Number: RR-P09, Revision 0 (Sheet 2 of 4)

REOUEST FOR RELIEF (cont.h The Eme:gency Diesel Generator fuel oil supply system cor.sists of two parallel trains, one for 4 each diesel. Fuel oil is supplied directly to the diesel fuel block from the 800-gallon day tank.

Makeup to each diesel day tank is accomplished automatically from the 75,000-gallon storage tank by operation of the respective Diesel Fuel Oil Transfer pump. The diesel day tank is sized for three (3) hours of continuous full load operation, based on a diesel fuel oil consumption rate of approximately 3.4 gpm. The Diesel Fuel Oil Transfer pumps are positive displacement pumps with a design capacity of approximately 8.7 gpm. l It is considered impracticable to directly measure pump flow rate on a quarterly basis. There is no flow rate instrumentation installed in the Fuel Oil Transfer system. Sufficient straight sections of piping are required to properly measure flow rate, through the use of either permanently or temporarily installed instmmentation, such as non-intrusive flow measurement devices. The only sufficient straight sections of piping exist in the buried sections of the supply headers. Installation of flow rate instmmentation or a pump test loop would require significant system redesign and modification, which would be costly and burdensome to Vermont Yankee.

r' Diesel Fuel Oil Transfer pump flow rate can be determined indirectly by measuring the level U] change in the diesel day tank versus the pump operating time required to make that change.

However, in order to allow for evaluation of the test results against the acceptance criteria of Table 3 of Part 6 of the Code, the test must be performed with the respective Emergency Diesel Generator secured. This eliminates the unknown variability of the diesel fuel oil consumption rate.

In addition, in order to provide measurement accuracy comparable with Table 1 of Part 6 of the Code, the automatic pump start feature on low diesel day tank level must be disabled and the diesel day tank volume reduced prior to the test through operation of the respective Emergency ,

Diesel Generator. I Disabling the automatic start feature of the Diesel Fuel Oil Transfer pump on low diesel day tank I level lessens the ability of the Emergency Diesel Generator to operate automatically without operator assistance, reduces the availability of an engineered safety system, and requires entry into a Vermont Yankee Technical Specifications Limiting Condition of Operation, with the required .

alternate testing requirements. l Revision 18 Section 4 Page 28 of 31

Vcrm:nt Yrnkee Nuclear Pcwer Stati:n Ins:rvice Testing Program q BELlEF REOUEST V

, Number: RR-P09, Revision 0 (Sheet 3 of 4)

ALTERNATE METHOD:

During quarterly inservice testing of each Diesel Fuel Oil Transfer pump, it will be verified that

' the pump is capable of supplying fuel oil to the respective diesel day tank at a flow rate greater than that required by the operating Emergency Diesel Generator. This is verified by an increase in diesel day tank level during the diesel surveillance testing. In addition, full spectrum vibrational monitoring and measurement of pump discharge pressure will be performed with the results evaluated against the acceptance criteria of Table 3 of Part 6 of the Code.

Once each operating cycle the flow rate of each Diesel Fuel Oil Transfer pump will be determined indirectly by measuring the level change in the diesel day tank versus the pump opera,ing time required to make that change. This will be performed with the respective Emergency Diesel Generator secured, the automatic pump start feature on low diesel day tank level disabled, and the diesel day tank volume reduced prior to the test through operation of the respective Emergency 1

Diesel Generator. This testing will provide measurement accuracy comparable with Table 1 of Pan 6 of the Code and the results will be evaluated against the acceptance criteria of Table 3 of Part 6 of the Code. As with the quarterly testing, full spectmm vibrational monitoring and measurement of pump discharge pressure will be performed with the results evaluated against the

acceptance criteria of Table 3 ofPart 6 of the Code.

Such testing is considered commensurate with the pump type and service and provides an acceptable level of quality and safety, based on the following:

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I Revision 18 Section 4 Page 29 of 31 j j

Vcrm:nt Ycnkee Nuclear Pawer Stati:n Inservice Testing Program RELIEF REOUEST

,1 Number: RR-P09, Revision 0 (Sheet 4 of 4)

ALTERNATE METHOD (cont.h a) A review of the pump design flow rate versus the diesel fuel oil consumption rate indicates an excess capacity of approximately 60 percent. As such, operational readiness of the pumps  ;

is still assured with up to 60 percent degradation, provided that pump bearing vibration is not excessive. Assurance of acceptable pump bearing vibration levels is provided through the full spectrum vibrational monitoring.

b) A review of Vermont Yankee maintenance records and industry experience, as documented in NPRDS, indicates that the pumps are highly reliable and that the above testing methods are acceptable for assessing pump operational readiness and determining potential i degradation.

At Vermont Yankee, four (4) failures have occurred in twenty (20) years of plant operations. Of these failures, three (3) were related to electrical components and one (1) was related to high bearing vibrations. In addition, minor shaft seal leakage has been noted and corrected.

p\ The industry experience is consistent with Vermont Yankee. For similar pumps in similar applications, fifteen (15) failures have been reported via NPRDS. Of these failures, nine (9) were related to excessive seal leakage, four (4) were related to electrical components and two (2) were related to high bearing vibrations.

Each of the above failure modes is adequately monitored during the quarterly inservice testing through visual inspection of the pump seals, proper starting and operation of the pump upon low diesel day tank level, and full spectmm vibrational monitoring.

J USNRC EUALUATION STATUS: l Relief was granted in the September 1993 SER [ Reference (s)] for Relief Request RR-P09, i Revision 0.

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' Revision 18 Section 4 Page 30 of 31

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Verm=t Ycnkee Nuclear Po .>er Stati::n Inservice Testing Program I

..n RELIEF REOUEST b

Number: RR-PIO, Revision 1 (Sheet 1 of1)

THIS RELIEF REQUEST WAS DELETED IN REVISION 18 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's. j i

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4 Revision 18 Section 4 Page 31 of 31

Vermont Ycnkee Nuclear Pcwer Statia Inservice Testing Progr:m 5.0 Valve Inservice Testing Program

<x/

5.1 Scope J

The Third-Interval Valve Inservice Testing (IST) Program establishes testing ,

requirements to assess the operational readiness of certain Safety Class 1, 2, and 3 valves and pressure relief devices, including their actuating and position indicating systems.

4 The active or passive Safety Class 1, 2, and 3 valves included are those which are 4

reqmred to perform a specific function in: l 1

l a) Shutting down the reactor to the cold shutdown condition,  !

L 2

b) Maintaining the reactor in the cold shutdown condition, or

] c) Mitigating the consequences of an accident. l

i The Safety Class 1,2, and 3 pressure relief devices included are those for protecting l systems or portion of systems which perform a required function in

j a) Shutting down the reactor to the cold shutdown condition,  !

1  :

i b) Maintaining the reactor in the cold shutdown condition, or i c) Mitigating the consequences of an accident.  ;

l

) The following are excluded from the above, provided that they are not required to ,

perform a specific function as specified above:

l a) Valves used only for operating convenience such as vent, drain, instrument, and test valves; b) Valves used only for system control, such as pressure regulating valves;

)

c) Valves used only for system or component maintenance.

External control and protection systems responsible for sensing plant conditions and providing signals for valve operation are also excluded from the above.

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Revision 18 Section 5 Page 1 of 160 l

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Verm:nt Yenkee N: clear Pcwer Statin Inservice Testing Program l

5.2 Valve Program Listing Table 5-1, " Valve Listing" lists all valves and pressure reliefdevices included in the Third-Interval Valve IST Program.

This Table identifies all valves and pressure relief devices subject to inservice testing, the inservice test requirements, testing frequency, and any applicable relief requests and/or i remarks. Table 5-1 is sorted by Drawing. The column headings in Table 5-1 are listed and j explained below: i Drawing The Flow Diagram which depicts the valve or pressure relief device. If the valve or  !

pressure relief device appears on multiple Flow Diagrams, then the primary Flow Diagram identifierislisted.

Drawing Title The title of the Flow Diagram. '

l Valve Number l f( The unique number that identifies the valve or pressure relief device. The valve number corresponds to the MPAC Database utilized at Vermont Yankee.

I Nomenclature The common name for the valve or pressure relief device. )

Dwg Coor The coordinate location (e.g., D-05) on the Flow Diagram where the valve or pressure reliefdevice appears.

Safety Class The safety class of the valve or pressure relief device as it appears on the Flow Diagram.

The safety classification of the valve or pressure relief device shown on the Flow Diagrams were determined in accordance with the Vermont Yankee Safety Classi6 cation Manual.

O Revision 18 Section 5 Page 2 of 160

4 Vcrm:nt Ycckee Nucle:r P wer St:ti:n Inservice Testing Program y 5.2 Valve Program Listing (cont.)

OM Cat The valve category as defined by Paragraph 1.4 of Part 10 of the Code. All categories are identified for those valves which have more than one applicable category (i.e. A/C for .

category C valves that also have leakage requirements).

I A: Valves for which seat leakage is limited to a specific maximum amount in the closed '

positica for fulfillment of their required function (s).

B: Valves for which seat leakage in the closed position is inconsequential for  !

fulfillment of their required function (s).  ;

C: Valves which are self-actuating in response to some system characteristic, such as  !

pressure (relief valves) or flow direction (check valves) for fulfillment of their  ;

required function (s).  :

D: Valves which are actuated by an energy source capable of only one operation, such l as mpture disks or explosively actuated valves.

4 Active / Passive i

The valve classification as defined by Paragraph 1.1 of Part 10 of the Code.

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Act Those valves which are required to change obturator position to accomplish j their required function.

Pass Those valves which maintain obturator position and are not required to l change obturator position to accomplish their required function. i r

Size (Inch) j The nominal pipe size of the valve, in inches.

Revision 18 Section 5 Page 3 of 160

l Verm=t Ycnkee Nuclear Pcwer Statia Itservice Testing Program Body The valve body style designator where:

)

3-WAY for 3-Way Valves I BL for Ball Valves )

BTF for Butterfly Valves l CK for simple Check Valves  ;

EFC for Excess Flow Check Valves j GA for Gate Valves '

GL for Globe Valves  ;

GSC for Globe Stop Check Valves l RD for Rupture Discs RV for Relief Valves SQUIB for explosively actuated valves l SRV for Safety / Relief Valves l Actuator i i

The type of actuator provided with the valve body where:

AO for Air Operated 6D'

'x' r r " v c" r8- -

HO for Hydraulically Operated l MAN for Manually Operated MO-AC for Motor Operated, AC powered motor  !

MO-DC for Motor Operated, DC pcwered motor SA for Self Actuated (i.e. Relief Valves, etc.)  ;

SO for Solenoid Operated Normal Position This field details the valve's position during normal plant operation (if the subject system is required to be in a standby condition during normal power operation, the valve's normal position is the position of the valve when the system is in it's standby condition), where:

C for Closed LC for Locked Closed O for Open LO for Locked Open O/KLfor Open, Key Locked O/C for Open or Closed (dependent on system demand)

T for Throttled O

Revision 18 Section 5 Page 4 of 160

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l Vcrm:nt Yc kee N: clear Pcwer Stati:n Inservice Testing Program Safety Position The position a valve or pressure-relief device is required to function in shutting down the reactor to the cold shutdown condition, maintaining the reactor in the cold shutdown condition, or mitigating the consequences of an accident.

i C for Closed function l O for Open function ,

O/C for both Open or Closed functions l

W Test Type (and Stroke Direction)

The test type is an abbreviation of the type of test required to be performed and the r stroke direction in which the test is performed.

LJ OM Category A Valves, which are containment isolation valves tested in accordance with the requirements of OM-10 subsection 4.2.2.2 and 10 CFR 50 Appendix J. Analysis of leakage rate and corrective action is in accordance with OM-10 subsection 4.2.2.3 (e) and (f). l l

LT OM Category A Valves, which perform a function to limit leakage to a specific amount and are tested in accordance with the requirements of OM-10 subsection 4.2.2.3.

l LEF Category A Excess Flow Check Valves, which perform a function to limit '

leakage to a specific amount and are tested in accordance with the  :

requirements of OM-10 subsection 4.2.2.3. Satisfactory completion of excess I flow check valve leakage testing also satisfies both the open and close valve  ;

exercise requirements of OM-10 subsection 4.3.2.2. )

l STC Active Category A and B Power-Operated Valves, which are full-stroked tested in the closed direction and the stroke times are measured and evaluated in accordance with Paragraph 4.2.1 of Part 10 of the Code.

STO Active Category A and B Power-Operated Valves, which are full-stroked tested in the open direction and the stroke times are measured and evaluated in accordance with Paragraph 4.2.1 of Part 10 of the Code.

SC Active Category A and B Non-Power-Operated Valves and Category C Check Valves, which are full-stroked tested in the closed direction in accordance with Paragraphs 4.2.1 and 4.3.2 of Part 10 of the Code.

SO Active Category A and B Non-Power-Operated Valves and Category C  !

Check Valves, which are full-stroked tested in the open direction in I accordance with Paragraphs 4.2.1 and 4.3.2 of Part 10 of the Code.

Revision 18 Section 5 Page 5 of 160

Vermut Yankee Nrclear Pawer Stati:n Inservice Testing Prcgr:m o Test Type (cont.)

i V

, PSC Active Category A and B Valves and Category C Check Valves, which are partial-stroked tested in the closed direction in accordance with Paragraphs 4.2.1 and 4.3.2 of Part 10 of the Code.

PSO Active Category A and B Valves and Category C Check Valves, which are l panial-stroked tested in the open direction in accordance with Paragraphs l 4.2.1 and 4.3.2 of Part 10 of the Code.

PIT Valves with remote position indicators are verified in accordance with 4 Paragraph 4.1 ofPart 10 of the Code. {

l FST Active Category A and B Valves with fail-safe actuators, which are tested in accordance with Paragraph 4.2.1.6 of Part 10 of the Code.

ET OM Category D Explosively Actuated Valves, which are tested in accordance ,

with Paragraph 4.4.1 of Part 10 of the Code.  !

RD OM Category D Rupture Discs, which are replaced in accordance with Paragraph 4.4.2 of Part 10 of the Code.

CD OM Category C Check Valves, which are disassembled in accordance with paragraph 4.3.2.4(c) of Pan 10 of the Code.

SP OM Category C Relief and Safety Relief Valves are tested in accordance with Paragraph 4.3.1 ofPart 10 of the Code.

SKIDActive Category A, B and C skid-mounted components or component subassemblies in which testing of the major component is an acceptable means for verifying the operational readiness of the skid-mounted and component subassemblies in accordance with NUREG 1482, Paragraph 3.4. -

N/A For passive valves with no testing required.

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i Revision 18 Section 5 Page 6 of 160

Vcrmnt Yankee Nucl:ar Pcwer Stati:::t Inservice Testing Przgram Test Freq O The test performance frequency associated with each test type will be specified where:

IM for every 1 Month 6M for every 6 Months 1Y for every 1 Year 2Y for every 2 Years 5Y for every 5 Years 10Y for every 10 Years CS for Cold Shutdown OC for Once Per Operating Cycle Q for Quarterly RO for Reactor Refueling Outage N/A forNot Applicable CSJ/ROJ/RR The Cold Shutdown Justification / Refuel Outage Justification or Relief Request number associated with the subject valve or pressure relief device and test type or test frequency, where applicable.

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Remarks When appropriate, clarifying comments or other remarks will be included for the affected valve, pressure relief device, test type or test frequency.

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O Revision 18 Section 5 Page 7 of 160

O O O Vermont Yankee Nuclear Power Station Insenice Testing Program Table 5-1 Valve Listing Drawing : 5920-271 Drawing

Title:

Neutron Monitoring System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch) Pos Pos Pos Type Freq RR CV-7-1 Neutron Monitoring C-07 2 A Act .375 BL SO C C FC FST Q n/a Note 6 System TIP Tube Ball U 2Y n/a Vaht PIT 2Y n/a STC Q n/a CV-7-2 Neutron Monitoring C-07 2 A Act .375 BL SO C C FC FST Q n/a Note 6 System TIP Tube Ball U 2Y n/a Valve PIT 2Y n/a STC Q n/a CV-7-3 Neutron Monitoring C-07 2 A Act .375 BL SO C C FC FST Q n/a Note 6 System Tip Tube Hall Vahr U 2Y n/a PIT 2Y n/a STC Q n/a S-7-1 Neutron Monitoring C-07 2 D Act .375 SQUIB EXP O C n/a ET RO n/a System Tip Tube Shear Vahe S-7-2 Neutron Monitoring C-07 2 D Act .375 SQUID EXP O C n/a ET RO n/a System Tip Tube Shear Vahc S-7-3 Neutron Monitoring C-07 2 D Act .375 SQUIB EXP O C n/a ET RO n/a System Tip Tube Shear Vaht V7-1 Neutron Monitoring C-08 2 A/C Act .5 CK SA O/C C n/a U 2Y n/a System Tip Purge Iso. SC RO ROJ-VII Check Vahr V7-2 Neutron Monitoring D-08 2 A/C Act .5 CK SA O/C C n/a U 2Y n/a System Tip Purge Iso. SC RO ROJ-VII Check Vaht Revision 18 Drawing : 5920-271 Section 5 Page 8 of 166

• O Vcrmont Yc kee Nuclear Pc.wer Siction O O Irservica Testing Program Table 5-1 Valve Listing Drawing : G-191159 Sh I Drawing

Title:

Service Water System Valve Number Nomenclature Dwg Safet OM Act / Sire Body Norm Safety Fail Test Test CSJ/ROJ Act Remarks '

Coor y Cat Pass (inch Pos Pos Pos Type Freq RR FCV-104-17A SW Supply to Screen Wash D-04 3 D Act 4 GA IIO C C FC FST Q n/a Note 8 Flow Control Vahr STC Q n/a FCV-104-17B SW Supply to Screen Wash E-04 3 B Act 4 GA 11 0 C C FC FST Q n/a Note 8 Flow Control Vahr STC Q n/a FCV-104-17C SW Supply to Screen Wash F-04 3 B Act 4 GA 11 0 C C FC FST Q n/a Note 8 Flow Control Vaht STC Q n/a FCV-104-17D SW Supply to Screen Wash F-04 3 B Act 3 GA HO C C FC FST Q n/a Note 8 Flow Control Vahr STC Q n/a FCV-104-17E SW Supply to Screen Wash G-04 3 B Act 3 GA 11 0 C C FC FST Q n/a Note 8 Flow Control Vaht STC Q n/a PCV-19449A RIIRSW PCV for SW to A- J-13 3 B Act 8 GL AO C O FO FST Q n/a Note 8 EDG and RRU's 5,7 STO n/a Q _

PCV-194498 RIIR PCV for SW to B- B-13 3 B Act 8 GL AO C O FO FST Q n/a Note 8 EDG AND RRU's 6,8 STO Q n/a SE-70-4A RIIRSW Pump Motor K-10 3 B Act 1 GA SO C O FO FST Q n/a Cooling Coil Supply Vahr STO Q n/a RIIRSW Pump Motor K-10 SE-70-4 B 3 B Act 1 GA SO C O FO FST Q n/a Cooling Coil Supply Vahr STO Q n/a SE-70-4C RIIRSW Pump Motor K-10 3 B Act 1 GA SO C O FO FST Q n/a Cooling Coil Supply Vahe STO Q n/a SE-70-4 D RIIRSW Pump Motor K-10 3 B Act I GA SO C O FO FST Q n/a Cooling Coil Supply Valve STO Q n/a SP-70-3A A Service Water Strainer 11-0 5 3 B Act 2 BL AO C C FC FST Q n/a Note 8 Backwash Vahe STC Q n/a SP-70-3B B Service Water Strainer C-05 3 B Act 2 BL AO C C FC FST Q n/a- Nete8 Backwash Valve STC Q n/a -

SR-70-13A RIIRSW Pump Motor L-09 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Cooling Relief Valve Revision 18 Drawing : G-191159 Sh 1 Section 5 Page 9 of 160

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Vcrmont Yc kee Nucirr P w r Str.ti:n Inservice Testi:g Program Table 5-1 Valve Listing Drawing: G-191159 Sh 1 Drawing

Title:

Senice Water System Valve Number Nomenclature Dwg Safet OM Act / Size Bcdy Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR SR-70-138 RIIRSW Pump Motor L-09 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Cooling Relief Valve SR-70-13C RilRSW Pump Motor L-09 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Cooling Relief Vahe SR-70-13D RIIRSW Pump Motor L-09 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Cooling Relief Vahr SR-70-2A SW Relief Vahr 1-08 3 C Act 1 RV SA C O/C n/a SP 10Y n/a S R-70-2B SW Relief Vaht C-08 3 C Act 1 RV SA C O/C n/a SP 10Y n/a V70-130A CW Pump Cooling SW C-05 3 B Act 3 GA MAN O C n/a SC Q n/a Note 8  !

Supply V70-130B CW Pump Cooling SW 1-05 3 B Act 3 GA MAN O C n/a SC Q n/a Note 8 Supply V70-13A SW Supply to the A J-07 3 B Act 24 GA MAN O O/C n/a SC RO ROJ-V16 Note 8 IIcader Isolation V70-13B SW Supply to the B IIcader B-07 3 B Act 24 GA MAN O O/C n/a SC RO ROJ-V16 Note 8 Isolation i V70-15A SW Supply to the A J-08 3 B Act 12 GA MAN O C n/a SC RO ROJ-Vl7 Note 8 RBCCW IIcat Exchanger V70-15B SW Supply to the B B-08 3 D Act 12 GA MAN O C n/a SC RO ROJ-V17 Note 8 RBCCW IIcat Exchanger V70-16B Alternate Cooling to SW C-09 3 B Act 24 GA MAN C O/C n/a SO RO ROJ-V21 Note 8 Supply V70-184 Alt Cooling Water Line C-09 3 D Act 1 GA MAN C O/C n/a SC RO ROJ-V21 Note 8 Vent SO RO ROJ-V21 V70-187A Gland Seal Supply to SW J-06 3 B Act 1.25 GA MAN O C n/a SC RO ROJ-VIS Note 8 Pumps A and B V70-187B Gland Seal Supply to SW B-06 3 B Act 1.25 GA MAN O C n/a SC RO ROJ-V18 Note 8 Pumps C and D Revision 18 Drawing : G-191159 Sh 1 Section 5 Page 10 of 160

O O O Vcrmont Yankee Nuclear Powar St: tion Inservice Testi:g Program Table 5-1 Valve Listing Drawing : G-191159 Sh 1 Drawing

Title:

Senice Water System Dwg Safet OM Act / Size Valve Number Nomenclatuit Body Act Norm Safety Fail Test Test CSJ/ROJ Resnants Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V70-19A SW SupplyIIcader 11-0 9 3 B Act 24 GA MO-AC O O/C FAI FIT 2Y n/a Crossconnect Vaht STC Q n/a STO Q n/a V70-19B SW Supply Header C-09 3 B Act 24 GA MO-AC O O/C FAI PIT 2Y n/a Crossconnect Vaht STC Q n/a STO Q n/a V70-1A Service Water Pump C-02 3 C Act 14 CK SA O/C O/C n/a PSO Q ROJ-V14 Note 10 Discharge Check Vaht SC RO ROJ-V14 SO RO ROJ-V14 -

V70-1B Senice Water Pump B-02 3 C Act 14 CK SA O/C O/C n/a PSO Q ROJ-V14 Note 10 Discharge Check Vaht SC RO ROJ V14 SO RO ROJ-V14 V70-1C Service Water Pump K-02 3 C Act 14 CK SA O/C O/C n/a PSO Q ROJ-V14 Note 10 Discharge Check Vahr SC RO ROJ-V14 SO RO ROJ-V14 V70-1D Senice Water Pump J-02 3 C Act 14 CK SA O/C O/C n/a PSO Q ROJ-V14 Note 10 Discharge Check Vahr SC RO ROI-V14 SO RO ROJ-V14 V70-20 SW Supply Vahe to H-11 3 B Act 20 GA MO-AC O C FAI PIT 2Y n/a i Turbine Building Cooling STC Q n/a Loads V70-24B Alternate Cooling Supply C-12 3 B Act 2.5 GA MAN C O n/a SO n/a Q Note 22. Vaht also appears to RBCCW on U-191159 Sh 3 i V70-27 SW to Fire Water D-02 3 B Act 1.5 GL MAN LO C n/a SC n/a Note 8 Q ,

Pressurizing Line '

V70-281A SW System Vacuum J-09 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Valve SO RO ROJ-V28 SP 10Y n/a Revision 18 Drawing : G-191159 Sh I Section 5 Page 11 of 160

O Vcrmont Yc kee Nuclear Pcw:r Station O O Inservice Testi:g Program Table 5-1 Valve Listing Crawing : G-191159 Sh 1 Drawing

Title:

Service Water System Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V70-281B SW System Vacuum B-09 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Valve SO RO ROJ-V28 SP 10Y n/a V70-281C SW System Vacuum J-09 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Vahe SO RO ROJ-V28 SP 10Y n/a V70-281D SW System Vacuum C-09 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Vahr SO RO ROJ-V28 SP 10Y n/a V70-29 Alternate Cooling to SW D-13 3 B Act 3 GA MAN C O n/a SO Q n/a Note 22. Vahe also appears Supply on G-191159 Sh 3 V70-29A Alternate Cooling to SW D-13 3 B Act 3 GA MAN C O n/a SO Q n/a Note 22. Vahr also appears Supply on G-191159 Sh 3 V70-2A SW Pump Discharge C-03 3 B Act 14 GA MAN O O/C n/a SC RO ROJ-V16 Note 8 Isolation Vaht V70-28 SW Pump Discharge B-03 3 B Act 14 GA MAN O O/C n/a SC RO ROJ-V16 Note 8 Isolation Vahe V70-2C SW Pump Discharge J-03 3 B Act 14 GA MAN O O/C n/a SC RO ROJ-V16 Note 8 Isolation Vaht V70-2D SW Pump Discharge J-03 3 B Act 14 GA MAN O O/C n/a SC RO ROJ-V16 Note 8 Isolation Vahe V70-32B Alternate Cooling Supply C-12 3 B Act 2.5 GA MAN C O n'a SO Q n/a Note 22. Vahr also appears to RBCCW on G-191159 Sh 3 V70-36A Alternate Cooling Supply C-13 3 B Act 3 GA MAN C O n/a SO Q n/a Note 22. Vahe also appears to RBCCW on G-191159 Sh 3

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V70-36B Alternate Cooling Supply C-13 B Act 3 GA MAN C O n/a SO Q n/a Note 22. Vahe also appears to RBCCW on G-191159 Sh 3 V70-38A RHRSW Pump Discharge K-12 3 C Act 12 CK SA C O/C n/a SC Q n/a Cneck Vahr SO Q n/a Revision 18 Drawing : G-191159 Sh 1 Section 5 Page 12 of 160

O Vermont Yc kee Nuch:r Pcw:r Station O Inservics Testirg Program O

Table 5-1 Valve Listing Drawing : G-191159 Sh I Drawing

Title:

Service Water System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V70-38B RIIRSW Pump Discharge B-12 3 C Act 12 CK SA C O/C n/a SC Q n/a Check Valve SO Q n/a V70-40A RIIRSW Pump Discharge K-12 3 C Act 12 CK SA C O/C n/a SC Q n/a Check Vahr SO Q n/a V70-40B RHRSW Pump Discharge A-12 3 C Act 12 CK SA C O/C n/a SC Q n/a Check Vaht SO Q n/a V70-42A RIIRSW Alternate Cooling K-13 3 B Act 8 GA MAN C O n/a SO Q n/a Note 8 to SW Supply V70-42B RilRSW Alternate Cooling B-13 3 B Act 8 GA MAN C O n/a SO Q n/a Note 8 to SW Supply V70-43A SW Discharge Loop J-12 3 C Act 8 CK SA O O/C n/a CD RO RR-V12 licader Check Vahr PSO Q n/a V70-43B SW Discharge Imop b .' 2 3 C Act 8 CK SA O O/C n/a CD RO RR-V12 Header Check Vahe PSO Q n/a V70-504 Alternate Cooling to SW D-12 3 B Pass 2 GA MAN C C n/a n/a n/a n/a Supply V70-511B SW System Vacuum 11-12 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Vahr SO RO ROJ-V28 SP 10Y n/a V70-511C SW System Vacuum 11-12 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Vahr SO RO ROJ-V28 SP 10Y n/a V70-5A SW X-connection to Fire J-03 3 B Pass 20 GA MAN O O n/a n/a n/a n/a Sysem V70-5B SW X-connection to Fire C-03 3 B Pass 20 GA MAN O O n/a n/a n/a n/a Sysem -

V70-6 SW Screen Wash Alternate F-03 3 B Act 8 GA MAN O C n/a SC Q n/a Note 8 Cooling / Isolation Vaht Revision 18 Drawing : G-191159 Sh 1 Section 5 Page 13 of 160

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Vcrmont Ycnkee Nuclear Pcwcr St; tion I scryice Testing Program Table 5-1 Valve Listing Drawing: G-191159 Sh 2 Drawing

Title:

Senice Water System Valyc Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR FCV-104-28A Diesel Generator A 11-0 5 3 B Act 8 GL AO C O FO SKID Q n/a Note 12 Cooling Water Inlet FCV-104-28B Diesel Generator B D-05 3 B Act 8 GL AO C O FO SKID Q n/a Note 12 Cooling Water Inlet PCV-104-73A Diesel Generator A Service J-04 3 B Act 8 GL AO O O FO SKID Q n/a Note 12 Water Pump Pascharge PCV-104-73B Diesel Generator B Senice B-04 3 B Act 8 GL AO O O FO SKID Q n/a Note 12 Water Pump Discharge SB-70-1 SW To hiain Condensor D-09 NNS B Act 24 BTF hiAN O/C O/C n/a SC CS CSJ-V16 Note 8 Discharge Block Vahr SO CS CSJ-V16 SR-70-16A Diesel Generator Jacket I-06 3 C Act i RV SA C O/C n/a SP 10Y n/a Wtr Cooler Safety Relief Valve SR-70-16B Diesel Generator Jacket C-05 3 C Act i RV SA C O/C n/a SP 10Y n/a Wtr Cooler Safety Relief Vaht V10-89A

• RJiR HX Senice Water * * * * * * * * * * *

• 11-0 1

• See G-191172 for vahr Outlet Vahe info and test requirements V10-89B

• RIIRIIX Senice Water E-01 * * * * * * * * * * * *

• See G-191172 for valve Outlet Vahe info and test requirements V70-11 SW Discharge to Cooling D-07 3 B Act 14 GA h1AN O/C O/C n/a SC CS CSJ-V16 Note 8 Tower Basin Isolation SO CS CSJ-V16 Vaht V70-16A Alt. Cing Deep Basin to D-07 3 B Act 24 GA hiAN C O/C n/a SO RO ROJ-V21 Note 8 RIIRSW Pump Suct. Iso.

Vaht V70-17 Alternate Cooling to SW D-06 3 B Act 20 GA h1AN O/C O n/a SO CS CSJ-V16 Note 8 Supply V70-17A #1 West Cing Tower SW D-07 3 B Pass 6 GL MAN T O n/a SC RO n/a Notes 9 and 22 Distribution Tray Vahr SO RO n/a Revision 18 Drawing : G-191159 Sh 2 Section 5 Page 14 of 160

O O O Vcrmont Ycnkee N: clear Pcwsr Station Icservice Testing Program Table 5-1 Valve Listing Drawing : G-191159 Sh 2 Drawing

Title:

Service Water System Valve Number Nomenclature Dwg Safet OM Act / Size Act Norm Safety Fail Test Test ' CSJ/ROJ Body Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V70-17B #1 West Cing Tower SW D-07 3 B Pass 6 GL MAN T O n/a SC RO n/a Notes 9 and 22 Distribution Tray Valve SO RO n/a i V70-17C #1 West Cing Tower SW D-07 3 B Pass 6 GL MAN T O n/a SC RO n/a Notes 9 and 22 Distribution Tray Valve SO RO n/a V70-17D #1 West Cing Tower SW D-07 3 B Pass 6 GL MAN T O n/a SC RO n/a Notes 9 and 22 Distribution Tray Vahr SO RO n/a V70-18 SW Discharge Header E-07 3 D Act 20 GA MAN O O/C n/a SC CS CSJ-V16 Note 8 Isolation Vaht

~

V70-203 SW Supply to RRU-9 B-03 3 B Act 1.25 GA MAN O C n/a SC Q n/a Note 8 '

Isolation Vaht V70-206 SW to RRIJs 17A and J-03 3 B Act 2.5 GA MAN O C n/a SC Q n/a Note 8 17B, Isolation Vahr ,

V70-252B SW System Vacuum G-05 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Vahe SO RO ROJ-V28 SP 10Y n/a V70-252C SW System Vacuum G-05 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Vahr SO RO ROJ-V28 SP 10Y n/a V70-414 SW to Rad Monitor from D-08 NNS B Act .75 GA MAN O/C C n/a SC Q n/a Note 8 SW Discharge to Deep ,

Basin Revision 18 Drawing : G-191159 Sh 2 Section 5 Page 15 of 160 *

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Vcrmont Yc;kee Nucl=r Pcw2r St: tion Iaervica Testing Program Table 5-1 Valve Listing Drawing : G-191159 Sh 3 Drawing

Title:

RCW Cooling Water System Vahe Number Nomenclature Dwg Safet OM Act / Sire Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR RV-70-117A RBCCW Drywell Piping M-10 2 C Act .75 RV SA LC O/C n/a SP 10Y n/a Thermal Relief Vaht SR-70-1 A SW Relief Valve B-04 3 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-70-1 B SW Relief Valve C-04 3 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-704A SW Relief Vahr J-05 3 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-704B SW Relief Vahr 1-05 3 C Act i RV SA C O/C n/a SP 10Y n/a V70-106 Rad Waste Building Inlet P-02 3 B Act 2 GA MAN O C n/a SC Q n/a Note 8 V70-107 Rad Waste Building Outlet P-05 3 B Act 2 GL MAN O C n/a SC Q n/a Note 8 V70-113 Primary Containment K-08 2 A/C Act 8 CK SA O C n/a LJ 2Y n/a RCW Supply isolation SC RO ROJ-V24 Vahe V70-113D Test Connection for L-08 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a Penetration X-23 V70-117 Primary Containment L-12 2 A Act 8 GA MO-AC O C FAI LJ 2Y n/a -

RBCCW Return Vahe PIT 2Y n/a STC CS CSJ-VOI V70-117D Test Connection for M-12 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a Penetration X-24 V70-24A RBCCW cooling to CRD N-02 3 B Act 2.5 GA MAN O O/C n/a SC Q n/a Note 8 and B/D RIIR Pumps V70-24B

• Alternate Cooling Supply 0-03 * * * * * * * * * * * *

• See G-191159 Sh ! for to RBCCW vahe info and test requirements V70-28 RBCCW cooling to CRD O-10 3 B Pass 3 GA MAN C C n/a n/a n/a n/a Note 8 and RHR Pumps V70-28A RBCCW cooling to CRD O-10 3 B Act 3 GA MAN O O/C n/a SC Q n/a Note 8 and RHR Pumps Revision 18 Drawing : G-191159 Sh 3 Section 5 Page 16 of 160

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Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drr. wing : G-191159 Sh 3 Drawing

Title:

RCW Cooling Water System .

' Dwg Safet OM Act / Size Body Norm Safety Fail Test Test CSJ/ROJ Remarks Nomenclature Act l Vcive Number Coor y Cat Pass (inch Pos Pos Pos Type Freq RR

• * * * * * * * * * * *

• Sec G-19I159 Sh I for V70-29

• Alternate Cooling Return Q-10 from RBCCW valve info and test requirements

• * * * * * * * * * * *

• Sec G-191159 Sh I for V70-29A

• Alternate Cooling Return P-10 from RBCCW valve info and test requirements V73-32A RBCCW cooling to P-02 3 B Act 2.5 GA MAN O O/C n/a SC Q n/a Note 8 Radwaste and A/C RIIR Pumps

• * * * * * * * * * * *

• See G-191159 Sh I for V70-32B

• Alternate Cooling Supply P-03 to RBCCW valve info and test requirements

• * * * * * *

• See G-191159 Sh I for V70-36A

• Alternate Cooling Supply 0-02 * * * *

• l toRBCCW valve info and test requirements ,

• Sec G-191159 Sh I for V70-36B

• Alternate Cooling Supply P-02 to RBCCW valve info and icst requirements V70-95A RBCCW Pump Discharge A-07 3 C Act 12 CK SA O/C O/C n/a SC Q n/a Check Valve SO Q n/a V70-958 RBCCW Pump Discharge B-07 3 C Act 12 CK SA O/C O/C n/a SC Q n/a Check Valve - SO Q n/a s.

Itevision 18 Drawing : G-191159 Sh 3 Section 5 Page 17 of 160

O Vcrmont Yc kee Nuclear Pow;r St: tion

@ O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191159 Sh 5 Drawing

Title:

Recirculation Pump Cooling Water Vaht Numher Nomenclature Dwg Safet ' OM Act / Sire Norm Safety Fail Ter,t Test CSJ/ROJ Body Act hads Coor y Cat Pass (inch Pos Pos Pos Type Freq RR Sb2-2-7A Recirculation Pump Inst G42 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr Sb2-2-7B Recirculation Pump Inst G-02 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V .1 Excess Flow Check Valve SL2-2-8A Recirculation Pump Inst G-02 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahe S L 2-2-8 B Recirculation Pump Inst G-02 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vaht Revision 18 Drawing : G-191159 Sh 5 Section 5 Page 18 of 160

O Vcrmont Yc kee Nuclear Pcwsr Station O O Inservice Testi:g Program Table 5-1 Valve Listing Drawing : G-191160 Sh 3 Drawing

Title:

Instrument Air System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V72-151 Compressor Suction Test E-17 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a Connection V72-152 Drainfrest Connection L-18 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V72-152C Drainfrest Connection L-18 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V72-153A Check Vaht Test L-15 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a Connection V72-28A Isolation to Outboard L-15 NNS B Act 1 GL MAN O C n/a SC CS CSJ-V17 MSIVs V72-28B Isolation to Outboard L-16 NNS B Act i GL MAN O C n/a SC CS CSJ-V17 MSIVs V72-28D isolation to Outboard K-16 NNS B Act 1 GL MAN O C n/a SO CS CSJ-V17 MSIVs V72-28E Isolation to Outboard K-16 NNS B Act 1 GL MAN O C n/a SO CS CSJ-Vl?

MSIVs V72-38A PCAC Compressor Inlet E-17 2 A Act 2 GL AO O C FC FST Q n/a Containment Isolation U 2Y n/a Vahr PIT 2Y n/a STC Q nh V72-38B PCAC Compressor Inlet F-17 2 A Act 2 GL AO O C FC FST Q n/a Containment Isolation U 2Y n/a Vahr PIT 2Y n/a STC Q n/a V72-89B DrywellInstrument Air K-14 2 A/C Act 2 CK SA O C n/a U 2Y n/a Supply Check Valve SC RO ROJ-V22 V72-89C DrywellInstrument Air K-15 2 A/C Act 2 CK SA O C n/a U 2Y n/a Supply Check Valve SC RO ROJ-V22 Revision 18 Drawing : G-191160 Sh 3 Section 5 Page 19 of 160

O O O Vcrmont Yc kee N: clear Pcw:r Station Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191160 Sh 7 Drawing Titic: Diesel Generator Starting Air System Vaht Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remads Coor y Cat Pass (inch Pos Pos Pos Type Freq RR AS-24-1A DG Starting Air Engine B-16 3 B Act 1.5 GL SO C O FC PSO 6M RR-V03 Inlet Valve PSO 1M RR-V03 STO OC RR-V03 AS-24-1 B DG Starting Air Engine B-03 3 B Act 1.5 GL SO C O FC PSO 6M RR-V03 Inlet Valve PSO IM RR-V03 STO OC RR-V03 AS-24-2A DG Startinc Air Engine B-17 3 B Act 1.5 GL SO C O FC PSO 6M RR-V03 Inlet Vaht PSO IM RR-V03 STO OC RR-V03 AS-24-2 B DG Starting Air Engine B-03 3 B Act 1.5 GL SO C O FC PSO 6M RR-V03 Inlet Vahe PSO 1M RR-V03 STO OC RR-V03 AV-24-1 A DG Starting Air Engine B-17 3 B Act .1875 GA SO O C FO SC IM RR-V03 Inlet Vent Vahr STC 6M RR-V03 AV-24-1 B DG Starting Air Engine B-03 3 D Act .1875 GA SO O C FO SC IM RR-V03 Inlet Vent Valve STC 6M RR-V03 SR-724A DG Starting Air Receiver E-15 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Relief Vaht SR-724B DG Starting Air Receiver E-13 3 C Act .75 RV SA C O/C n/a SP 10Y n/a ReliefVahr SR-724C DG Starting Air Reciever E-07 3 C Act .75 RV SA C O/C n/a SP 10Y n/a ReliefVahr SR-724D DG Starting Air Reciever E-05 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Relief Vahr SR-72-7A DG Starting Air 11-12 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Compressor Relief Vahr SR-72-7B DG Starting Air 11-0 8 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Compressor Relief Vahc Revision 18 Drawing : G-191160 Sh 7 Section 5 Page 20 of 160

O Vcrmont Yc;kee Nuclear Pow;r Str tion O O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191160 Sh 7 Drawing

Title:

Diesel Generator Starting Air Sysem Valve Number Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Nomenclature Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V72-76A DG Starting Air 11-12 3 C Act .75 CK SA C O n/a PSO Q ROJ-V30 Compressor Discharge SO RO ROJ-V30 ,

Check Vaht V72-76B DG Starting Air h-08 3 C Act .75 CK SA C O n/a PSO Q ROJ-V30 Compressor Discharge SO RO ROJ-V30 Check Vaht V72-82AX Senice Air to DG Starting D-14 3 B Pass 2 GA MAN LC C n/a n/a n/a n/a Air Cross Connect Vaht V72-82BX Senice Air to DG Starting D4)6 3 D Pass 2 GA MAN LC C n/a n/a n/a n/a Air Cross Connect Vahr Revision 18 Drawing : G-191160 Sh 7 Section 5 Page 21 of 160

O Vcrmont Ycckee Nuclear Pow;r Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191160 Sh 8 Drawing

Title:

Instrument Air Sysem Valve Number Nomenclature Dwg l Safet OM Act / Size My Norm Safety Fail Test Test CSJ/ROJ Act  %,g Coorl y Cat Pass (inch Pos Pos Pos Type Freq RR SE-105-2A N2 Supply 3 Way Vent C-06 3 B Pass .75 3-WAY SO O O FO n/a n/a n/a Vaht SE-105-2B N2 Supply 3-Way Vent C-12 3 B Pass .75 3-WAY SO O O FO n/a n/a n/a Vaht SR-72-10A N2 Supply Relief Vaht C-05 3 C Act .75 RV SA C O/C n/a SP 10Y n/a SR-72-10B N2 Supply Relief Vaht C-04 3 C Act .75 RV SA C O/C n/a SP 10Y n/a SR-72-9A N2 Supply Relief Vaht F-05 3 C Act .75 RV SA C O/C n/a SP 10Y n/a SR-72-9B N2 Supply Relief Vaht F-04 3 C Act .75 RV SA C O/C n/a SP 10Y n/a V72-609 Instrument Air to Reacter E-04 3 A/C Act .375 CK SA O C n/a LT 2Y n/a Building Railroad Airlock SC n/a Q

V72-610 Instrument Air to Reactor E-04 3 A/C Act .375 CK SA O C n/a LT 2Y n/a Building Railroad Airlock SC Q n/a Revision 18 Drawing : G-191160 Sh 8 Section 5 Page 22 of' 160

O Vcrmont Yazkee Nuclear Power Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191162 Sh 2 Drawing

Title:

Fuel Oil- Miscellaneous S 3stems Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR LCV-198-2A DieselOil DayTank Level E-08 3 B Act 1.5 GA AO C O/C FC FST Q n/a Control Vahr STC Q n/a STO Q n/a l LCV-108-2B DieselOil Day Tank Level C-08 3 B Act 1.5 GA AO C O/C FC FST Q rt'a Control Vaht STC Q n/a i STO Q n/a SR-78-3A Diesel Fuel Oil Transfer E-06 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Pump Discharge Relief Vahc SR-78-3B Diesel Fuel Oil Transfer D-06 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Pump Discharge Relief Vaht V78-10A LCV-3 Inlet FO Storage E-07 3 D Pass 1.5 GA MAN LC C n/a n/a n/a n/a Building FO Pump Room V78-11 LCV-3 Bypass FO Storage F-07 3 B Pass 1.5 GA MAN C C n/a n/a n/a n/a Building FO Pump Room ,

V78-2 Fuel Oil Storage Tank Fill E-02 3 C Act 4 CK SA C C n/a CD RO n/a Line Check Vaht V78-28A Diesel Fuel OilTransfer E-06 3 C Act 1.25 CK SA C O/C n/a SC Q n/a Pump Discharge Check SO Q n/a Vaht V78-28B Diesel Fuel Oil Transfer D-06 3 C Act 1.25 CK SA C O/C n/a SC Q n/a Pump Discharge Check SO n/a Q

Vaht V78-41 A D/G lland Priming Pump G-11 3 C Act CK SA C C n/a SKID 1M n/a Note 12 Discharge Check V78-41B D/Giland Priming Pump E-Il 3 C Act CK SA C C n/a SKID IM n/a Note l2 Discharge Check V78-42A D/G Engine Driven Fuci G-12 3 C Act CK SA C O/C n/a SKID IM n/a Note 12 Pump Discharge Check Revision 18 Drawing : G-191162 Sh 2 Section 5 Page 23 of 160

p '

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O Vcrmont Yc kee Nuclear Pow:r St: tion Inservice Testing Program ,

Table 5-1 Valve Listing Drawing : G-191162 Sh 2 Drawing

Title:

Fuel Oil- Miscellaneous Systems Valve Number Menenclature Dwg Safet OM Act / Sire Body Act Norm Safety Fsil Test Test CSJ/ROJ g,4, Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V78-42B D/G Engine Driven Fuct D-12 3 C Act CK SA C O/C n/a SKID IM n/a Note 12 Pump Discharge Check V78-9 Fuel OilTransfer Pump E-06 3 B Pass 2 GA MAN LC C n/a n/a n/a n/a Recirc.

i Revision 18 Drawing : G-191162 Sh 2 Section 5 Page 24 of 160

O Vcrmont Ycnkee Nuclear Pcwcr Stction

@ O Inservice Testi:g Program Table 5-1 Valve Listing Drawing : G-191165 Drawing

Title:

Sampling System Valve Number Nomendatuit Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR G-16 * * * * * * * * * * *

• FCV-10-160

• RIIR Return Line Process

• See G-191172 for vaht Sampling Vahr info and test requirements FSO-199-75A-1

• CAD Sample Torus D-17 * * * * *

• See VY-E-75-002 for Inboard Isolation Valve vahr info and test requirements FSO-199-75A-2

• CAD Sampic Torus D-17 * * * * * * * * * * * *

• See VY-E-75-002 for Outboard Isolation Vahr valve info and test requirements FSO-109-75A-3

• CAD Sample Torus D-17 * * * * * * * * * * * *

• See VY-E-75-002 for Inboard Isolation Vahe vahr info and test requirements FSO-109-75A-4

• CAD Sample Torus D-17 * * * * * * * * * * * *

• See VY-E-75-002 for Outboard Isolation Vahr valve info and test requirements FSO-109-75B-1

• CAD Sample Lower D-16 * * * * * * * * * * * *

• See VY-E-75-002 for Drywell Inboard Isolation vaht info and test Vaht requirements FSO-109-75B-2

• CAD Sampic Lower D-16 * * * * * * * * * * * *

• See VY-E-75-002 for Drywell Outboard Isolation vahe info and test Vaht requirements FSO-109-75C-1

• CAD Sample Mid Drpvell C-15 * * * * * * * * * * * *

• See VY-E-75-002 for Inboard Isolation Vahr vahe info and test requirements FSO-109-75C-2

• CAD Sample Mid Drywell C-15 * * * * * * * * * * * *

• See VY-E-75-002 for Outboard Isolation Vaht vahr info and test requirements FSO-109-75D-1

• CAD Sample Upper B-15 * * * * * * * * * * * *

• See VY-E-75-002 for DrywellInboard Isolation vahr info and test Vahe requirements Revision 18 Drawing : G-191165 Section 5 Page 25 of 160

p q p

\

v v Vcrmont Yc:kee Nuclear Pow r Station Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191165 Drawing

Title:

Sampling System Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR FSO-109-75D-2

• CAD Sample Upper B-15 * * * * * * * * * * * *

• See VY-E-75-002 for Drywell Outboard Isolation vahr info and test Vaht requirements FSO-199-76A

• CAD CAM Return F-19 * * * * * * * * * * * *

• See VY-E-75-002 for ContainmeiiIsolation vahr info and test Vaht requirements F30-109-76B

• CAD CAM Return G-19 * * * * * * * * * * * *

• See VY-E-75-002 for Containment Isolation vahc info and test Valve requirements PAS-101 Post Accident Samping " 14 2 A Pass .375 GL AO C C FC U 2Y n/a Note 13 System Jet Pump Sample Vahe PAS-102 Post Accident Sampling K-16 2 A Pass .375 GL AO C C FC U 2Y n/a Note 13 System Jet Pump Sample Vahe PAS-103 Post Accident Sampling L-14 2 A Pass .375 GL AO C C FC U 2Y n/a Note 13 System Jet Pump Sample Vahe PAS-104 Post Accident Sampling K-16 2 A Pass .375 GL AO C C FC U 2Y n/a Note 13 System Jet Pump Sample Vahe PAS-106 Post Accident Sampling L-16 2 A Pass .375 GL AO C C FC LT 2Y n/a Note 13 Vahe also appears System Purge Vahr on G-191169 Sh I PAS-108C Post Accident Sampling K-15 2 A/C Pass .375 CK SA C C FC U 2Y n/a Note 13 System Jet Pump Sample Check Viv PAS-109C Post Accident Sampling L-15 2 A/C Pass .375 CK SA C C FC U 2Y n/a Note 13 System Jet Pump Sample Check Vlv PAS-114 Drain Vaht Isolation L-15 2 B Pass .375 GL MAN LC C n/a n/a n/a n/a Revision 18 Drawing : G-191165 Section 5 Page 26 of 160

O Vcrmont Ycnkee Nuclear Pow r Station O O-Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191165 Drawing

Title:

Sampling System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR PAS-115 Drain Line Isolation L-15 2 B Pass .375 GL MAN LC C n/a n/a n/a n/a PAS-117 Jet Pump i Sample Line I-13 2 B Pass .375 GL MAN LC C n/a n/a n/a n/a Test Connection PAS-118 Jet Pur,p 1 Sample Line I-13 2 B Pass .375 GL MAN LC C n/a n/a n/a n/a Test Omnection PAS-120 Jet Puntp 2 Sample Line I-12 2 B Pass .375 GL MAN LC C n/a n/a n/a n/a Test Comnection PAS-121 Jet Pump 2 Sample Line I-12 2 B Pass .375 GL MAN LC C n/a n/a n/a n/a Test Connection S1,2-3-21B Nuclear Boiler Vessel Inst C-13 1 A/C Act i EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahe sir 2-3-21D Nuclear Boiler Vessel Inst C-13 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr VG-23

• CAD Rad Monitor Supply G-16 * * * * * * * * * * * *

• See VY-E-75-002 for Containment Isolation vahm info and test Vahr requirements VG-26

• CAD Rad Monitor Supply G-15 * * * * * * * * * * * *

• See VY-E-75-002 for Containment Isolation vahr info and test Vaht requirements 1

Revision 18 Drawing : G-191165 Section 5 Page 27 of 160

O Vcrm: t YCckee NucIcar Paw;r Station O O I sersice Testi g Program Table 5-1 Valve Listing Drawing : G-191167 Drawing

Title:

Nuclear Boiler Valve Number Nomenclature Dwg Safet OM Act / Sire Nonn Safety Fail Test Test CSJ/ROJ Body Act g,,,4, Coor y Cat Pass (inch Pos Pos Pos Type Freq RR FCV-2-17 Rx VesselIIcad Steam B-06 2 B Pass 1 GL AO C C FC P!T 2Y n/a Note 14 Space Vent Flow Control Valve FCV-2-18 Rx VesselIIcad Steam B-05 1 B Pass I GL AO C C FC PIT 2Y n/a Note 14 Space Vent Flow Control Valve FCV-2-20 Rx Vessel Head Seal C-06 i B Pass 1 GL AO O O FO PIT 2Y n/a Note 14 Leakage Flow Control Valve FCV-2-21 Rx Vessel Head Seal C-05 2 B Pass 1 GL AO C C FC PIT 2Y n/a Note 14 Izakage FlowControl Valve FCV-2-39 Rx Recirc Sample Line L-04 2 A Act .75 GL AO C C FC FST Q n/a Note 20 Flow Control / Isolation LJ 2Y n/a Valve STC Q n/a ,

FCV-2-40 Rx Recirc Sample Line L-03 2 A Act .75 GL AO C C FC FST Q n/a Note 20 Flow Control / Isolation LJ 2Y n/a Valve STC Q n/a RV-2-71A Nuclear Boiler D-08 1 C Act 6X10 SRV SA C O/C n/a SP SY ROJ-V02 Note 7 Safety /ReliefVaht RV-2-71B Nuclear Boiler G-08 1 C Act 6X10 SRV SA C O/C n/a SP 5Y ROJ-V02 Note 7 Safety /ReliefVaht RV-2-71C Nuclear Boiler G-08 i C Act 6X10 SRV SA C O/C n/a SP SY ROJ-V02 Note 7 Safety /ReliefVaht RV-2-71 D Nuclear Boiler 11-0 8 I C Act 6X10 SRV SA C O/C n/a SP SY ROJ-V02 Note 7 Sarcty/ Relief Vahr SL-2-23 NB Flow Limiting Valve C-03 2 C Pass 1 EFC SA O O n/a n/a n/a n/a SL-2-3MA NB Flow Limiting Vahr E-04 2 C Pass 1 EFC SA O O n/a n/a n/a n/a SL-2-301B NB Flow Limiting Vahr D-04 2 C Pass 1 EFC SA O O n/a n/a n/a n/a Revision 18 Drawing : G-191167 Section 5 Page 28 of 160  ;

P O O O Vcrmont Yc kee Nuclear Pcwsr St ti:n Inservice Testing Program Table 5-1 Valve Listing Drawing : G-19 t I67 Drawing

Title:

Nuclear Boiler Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hads Coor y Cat Pass (inch Pos Pos Pos Type Freq RR Str2-301E NB Flow Limiting Valve E-04 2 C Pass 1 EFC SA O O n/a n/a n/a n/a Str2-301F NB Flow Limiting Valve D-04 2 C Pass 1 EFC SA O O n/a n/a n/a n/a Ste2-305A Nuclear Boiler VesselInst M-02 2 NC Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve Ste2-305B Nuclear Boiler VesselInst M-02 2 NC Act 1 EFC SA O O/C n/a LEF RO ROJ-VOI Excess Flow Check Valve Ste242A Nuclear Doiler Vessel Inst I-13 2 A/C Act i EFC SA O C n/a LEF RO ROJ-V01 Excess Flow Check Valve Str242B Nuclear Boiler Vessel Inst I-13 3 NC Act i EFC SA O C n/a LEF RO ROJ-Vol Excess Flow Check Valve Str242C Nuclear Boiler VesselInst J-13 2 NC Act i EFC SA O C n/a LEF RO ROJ-V01 Excess Flow Check Valve i I

Str242D Nuclear Boiler Vessel Inst J-13 2 A/C Act i EFC SA O C n/a LEF RO ROJ-Vol +

Excess Flow Check Valve Sir 244A Nuclear Boiler VesselInst K-13 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-Vol  ;

Excess Flow Check Valve  :

Ste244B Nuclear Boiler VesselInst K-13 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve Ste244C Nuclear Boiler VesselInst K-13 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve Str244D Nuclear Boiler Vessel Inst L-13 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve Str2-73/. Nuclear Boiler Vessel Inst F-12 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 f Excess Flow Check Vahe Str2-73B Nuclear Boiler Vessel Inst F-12 2' A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-VOI i Excess Flow Check Valve

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Ste2-73C Nuclear Boiler Vessel Inst G-12 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve i Revision 18 Drawing : G-191167 Section 5 Page 29 of' 160 t

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Vcrmont Yc kee N: clear Pow::r Station I; service Testing Program Table 5-1 Valve Listing Drawing : G-191167 Drawing

Title:

Nuclear Boiler Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR SL-2-73D Nucicar Boiler Vessel last G-12 2 A/C Act EFC O 1 SA O/C n/a LEF RO ROJ-Vol Excess Flow Check Vaht SL-2-73E Nuclear Boiler Vessel Inst G-12 2 A/C Act EFC SA 1 O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vaht SL-2-73F Nuclear Boiler Vessel Inst G-12 2 A/C Act EFC SA O O/C 1

LEF RO ROJ-V01 ,

Excess Flow Check Vaht SL-2-73G Nuclear Boiler VesselInst H-12 2 A/C Act EFC SA O 1

O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr SL-2-7311 Nuclear Boiler Vessci inst 11-12 2 A/C Act EFC SA 1 O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr i

SR-2-14 A SRV Exhaust Line M-12 NNS C Act 3 CK SA C C n/a SC RO ROJ-V27 Vacuum Breaker ,

SR-2-14 B SRV Exhaust Line M-12 NNS C Act 3 CK SA C C n/a SC RO ROJ-V27 Vacuum Breaker SR-2-14C SRV Exhaust Line M-12 NNS C Act CK 3 SA C O n/a SC RO ROJ-V27 Vacuum Breaker SR-2-14D SRV Exhaust Line M-12 NNS C Act 3 CK SA C C n/a SC RO ROJ-V27 Vacuum Breaker SR-2-14E SRV Discharge Line L-12 1 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum Relief Vahe SR-2-14F SRV Discharge Line L-12 3 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum Relief Vahe SR-2-14G SRV Discharge Line L-12 3 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum Relief Vahe SR-2-1411 SRV Discharge Line L-12 3 C Act 10 CK SA C ^

O/C u/a SP 10Y n/a Vacuum Relief Vahr SR-2-141 SRV Discharge Line L-12 3 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum Relief Vahe SR-2-14J SRV Discharge Line L-12 3 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum RelicIVahe Revision 18 Drawing : G-191167 Section 5 Page 30 of 160

I .

Gw Q- @a Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Crawing : G-191167 Drawing

Title:

Nucicar Boiler Valve Number Dwg Safet OM Act / Size Body Nomi Safety Fail Test Test CSJ/ROJ Nomenclature Act  %,g Coor y Cat Pass (inch Pos Pos Pos Type Freq RR S R-2-14 K SRV Discharge Line L-12 3 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum Relief Vahr SR-2-14L SRV Discharge Line L-12 3 C Act 10 CK SA C O/C n/a SP 10Y n/a Vacuum Relief Vahr SV-2-70A Nuclear Boiler D-08 1 C Act 6X8 SRV SA C O/C n/a SP SY Safety /ReliefVahr SV-2-70B Nucicar Boiler 11-0 8 1 C Act 6X8 SRV SA C O/C n/a SP SY n/a Safety / Relief Vahr V2-27A Reactor Fecdwater F-03 1 NC Act 16 CK SA O O/C n/a LJ 2Y ria Injection Line Check Vahr SC RO ROJ-V03 SO Q ROJ-V03 V2-27B Reactor Feedwater 11-0 2 2 C Act 16 CK SA O C n/a SC RO ROJ-V03 Injection Line Check Vahr V2-28A Reactor Feednter F-05 I A/C Act 16 CK SA O O/C n/a LJ 2Y n/a injection Line Check Valve SC RO ROJ-V03 SO Q ROJ-V03 V2-28B Reactor Fecdwater 11-0 5 i A/C Act 16 CK SA O O/C n/a LJ 2Y n/a injection Line Check Valve SC RO ROJ-V03 SO Q ROJ-V03 V2-300G NBS Globe Vahr E-04 2 B Pass 1.00 GL MAN C C n/a n/a n/a n/a V2-300ll NBS Globe Valve E-04 2 B Pass 1.00 GL MAN C C n/a n/a n/a n/a V2-37A MSRV Actuator Nitrogen B-08 2 A/C Act 1 CK SA O/C C n/a LT 2Y n/a Supply Check Vahr SC RO ROJ-V04 V2-37B MSRV Actuator Nitrogen G-09 2 A/C Act 1 CK SA O/C C n/a LT 2Y n/a Supply Check Vahr SC RO ROJ-V04 V2-37C MSRV Actuator Nitrogen G-09 2 A/C Act 1 CK SA O/C C n/a LT 2Y n/a Supply Check Vahr SC RO ROJ-V04 Revision 18 Drawing : G-191167 Section 5 Page 31 of 160

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191167 Drawing Titic: Nuclear Boiler Vaht Number Nomenclature Dwg Sarct OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch) Pos Pos Pos Type Freq RR V2-37D MSRV Actuator Nitrogen 11-0 9 2 A/C Act 1 CK SA O/C C n/a LT 2Y n/a  !

Supply Check Vahr SC RO ROJ-V04 V2-39AA Penetration X-4i Hermal L-04 2 A/C Act .5 CK SA C O/C n/a LJ 2Y n/a ReliefVaht SC RO ROJ-V29 SO RO ROJ-V29 V2-41 Recirc Sampic Test Line M-03 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Test Connection >

V2-53A Rx Recirculation Pump L-09 1 B Act 28 GA MO-AC O C FAI PIT 2Y n/a Discharge Isolation Vaht STC CS CSJ-V04 V2-53B Rx Recirculation Pump L-09 1 B Act 28 GA MO-AC O C FAI PIT 2Y n/a Discharge Isolation Valve STC CS CSJ-V04 V2-54A Rx Recirculation Pump L-09 i B Act 4 GA MO-AC O C FAI PIT 2Y n/a Discharge Bypass Vahr STC CS CSJ-V04  !

V2-54B Rx Recirculation Pump L-09 i B Act 4 GA MO-AC O C FAI PIT 2Y n/a Discharge Bypass Vahr STC CS CSJ-V04 V2-74 Main Steam Line Drain E-10 1 A Act 3 GA MO-AC C C FAI LJ 2Y n/a i Containment Isolation PIT 2Y n/a Vahr STC Q n/a V2-74A Penetration X-8 Therifial D-10 2 A/C Act .5 CK SA C O/C n/a LJ 2Y n/a Relief Vaht SC RO ROJ-V29 SO RO ROJ-V29 V2-75 Test Connection Between F-12 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Vahts V2-74 and 77 V2-77 Main Steam Line Drain E-13 1 A Act 3 GA MO-DC C C FAI LJ 2Y n/a Containment Isolation PIT 2Y n/a Vahr STC Q n/a Revision 18 Drawing : G-191167 Section 5 Page 32 er 160

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hv Ov ($b Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191167 Drawing

Title:

Nuclear Boiler Vaive Number Nomenclature Dwg Safet OM Act / Sire Nonn Safety Fail Test Test CSJ/ROJ Body Act  %,g Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V2-80A Main Steam Isolation Valve D-10 1 A Act 18 GL AO O C FC FST Q n/a Note 1 W 2Y n/a PIT 2Y n/a STC Q n/a V2-80B Main Steam Isolation Vahr G-12 1 A Act 18 GL AO O C FC FST n/a Q Note 1 W 2Y n/a PIT 2Y n/a STC Q n/a V2-80C Main Steam Isolation Vahr G-12 1 A Act 18 GL AO O C FC FST Q n/a Note 1 W 2Y n/a PIT 2Y n/a STC Q n/a V2-80D Main Steam Isolation Vahr 11-12 1 A Act 18 GL AO O C FC FST n/a Q Note 1 U 2Y n/a PIT 2Y n/a STC Q n/a V2-82A MSIV Actuator Nitrogen B-lO 2 A/C Act I CK SA O/C C n/a LT 2Y n/a

, Supply Check Vahr SC RO ROJ-V05 V2-82B MSIV Actuator Nitrogen B-lO 2 A/C Act i CK SA O/C C n/a LT 2Y n/a

• Supply Check Vaht SC RO ROJ-V05 V2-82C MSIV Actuator Nitrogen B-lO 2 NC Act I CK SA O/C C n/a LT 2Y n/a Supply Check Valve SC RO ROJ-VOS V2-82D MSIV Actuator Nitrogen B-lO 2 NC Act I CK SA O/C C n/a LT 2Y n/a Supply Check Vahr SC RO ROJ-V05 V2-83A Steam Line A Test D-12 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection V2-83B Steam Line B Test G-12 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection Revision 18 Drawing : G-191167 Section 5 Page 33 of 160

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191167 Drawing

Title:

Nuclear Boiler Valre Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act g,g, Coor y Cat Pass (inch Pos Pos Pos Type Farq RR V2-83C Steam Line C Test 11-1 2 2 B Pass .75 GL MAN LC C n/a n/a + /a Connection V243D Steam Line D Test 1-12 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection V246A Main Steam Isolation Vahe D-13 1 A Act 18 GL AO O C FC FST Q n/a Note 1 LJ 2Y n/a PIT 2Y n/a STC Q n/a V246B Main Steam Isolation Vahr G-12 1 A Act 18 GL AO O C FC FST Q n/a Note 1 LJ 2Y n/a PIT 2Y n/a STC Q n/a V2-86C Main Steam Isolation Vahr G-12 1 A Act 18 GL AO O C FC FST Q n/a Note 1 LJ 2Y n/a PIT 2Y n/a STC Q n/a V2-86D Main Steam Isolation Vahr 11-12 A Act GL- AO 1 18 O C FC FST Q n/a Note 1 LJ 2Y n/a PIT 2Y n/a STC Q n/a V2-87A MSIV Actuator Air Supply B-13 2 A/C Act 1 CK SA O/C C n/a LT 2Y n/a Check Vcht SC RO ROJ-VOS V2-87B MSIV Actuator Air Supply B-13 2 A/C Act CK SA O/C C 1 n/a LT 2Y n/a Check Vaht SC RO ROJ-V05 V247C MSIV Actuator Air Supply B-13 2 A/C Act I CK SA O/C C n/a LT 2Y n/a Check Vahr SC RO ROJ-V05 V2-87D MSIV Actuator Air Supply B-13 2 A/C Act I CK SA O/C C n/a LT 2Y n/a Check Vahr SC RO ROJ-V05 Revision 18 Drawing : G-191167 Section 5 Page 34 of 160

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191167 Drawing

Title:

Nuclear Boiler Valve Nuneber Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hah Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V2-94A Feed Water Line A Test F-04 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection V2-948 Fced Water B Test 11-0 4 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection V2-%A Reactor Feedwater 11-03 1 A/C Act 16 CK SA O O/C n/a IJ 2Y n/a Injection Line Check Vaht SC RO ROJ-V03 SO Q ROJ-V03 V2-%B Reactor Fecdwater F-03 2 C Act 16 CK SA O C n/a SC RO ROJ-V03 Injection Line Check Vahr V2-98A NBS Globe Vahc N-03 2 B Pass 1.00 GL MAN C C n/a n/a n/a n/a V2-988 NBS Globe Vahr N-12 2 B Pass 1.00 GL MAN C C n/a n/a n/a n/a Revision 18 Drawing : G-191167 Section 5 Page 35 of 160

Qt; O th v to Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191168 Drawing

Title:

Core Spray System Vaire Number Nomenclature Dwg Safet OM Act / Sire Nom Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat rass (inch Pos Pos Pos Type Freq RR Stel4-31 A Core Spray Inst Excess F-07 2 A/C Act i EFC SA O O/C n/a LEF RO POJ-Vol Flow Check Vahr S1-14-31B Core Spray Inst Excess E-07 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-Vol Flow Check Vahr SR-14-20A Core Spray Pump F-11 2 C Act 1.5 RV SA C O/C n/a SP 10Y n/a Discharge Relief Vahr SR-14-20B Core Spray Pump D-12 2 C Act 1.5 RV SA C O/C n/a SP 10Y n/a Discharge Relief Vaht V14-10A Core Spray Pump I-l1 2 C Act 10 CK SA C O/C n/a SC Q n/a Discharge Check Vaht SO Q n/a V14-10B Core Spray Pump I-14 2 C Act 10 CK SA C O/C n/a SC Q n/a Discharge Check Vaht SO n/a Q

v!411A Core Spray Injection Vaht G-09 2 B Act 8 GA MO-AC O O FAI PIT 2Y n/a STO Q n/a V14-IIB Core Spray Injection Vahc C-09 2 B Act 8 GA MO-AC O O FAI PIT 2Y n/a STO Q n/a V14-12A Core Spray Injection G-08 1 A Act 8 GA MO-AC C O/C FAI LJ 2Y n/a Note 21 Isolation Vaht LT 2Y n/a PIT 2Y n/a STC Q n/a STO Q n/a V14-12B Core Spray Injection C-08 I A Act 8 GA MO-AC C O/C FAI LJ 2Y n/a Note 21 Isolation Vahc LT 2Y n/a PIT 2Y n/a STC Q n/a STD Q n/a V14-13A Core Spray injection Check F-06 i A/C Act 8 CK SA C O/C n/a LT 2Y n/a Note 21 Vahr SC Q n/a SO CS CSJ-VOS Revision 18 Drawing : G-191168 Section 5 Page 36 of 160

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191168 Drawing

Title:

Core Spray System Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ '

Vaht Number Nomenclature Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V14-13B Core Spray injection Check D-06  : A/C Act 8 CK SA C O/C n/a LT 2Y n/a Note 21 Vaht SC Q n/a SO CS CSJ-V05 V14-14A Core Spray Injection Line F-05 1 B Pass 8 GA MAN LO O n/a PIT 2Y n/a Note 14 Vahr V14-14B Core Spray injection Line E-05 1 B Pass 8 GA MAN LO O n/a PIT 2Y n/a Note 14 Valve V14-16A Core Spray to RIIR System L-11 2 B Pass 4 GA MAN C C n/a n/a n/a n/a Isolation Vahc 1 V14-16B Core Spray to RilR System M-11 2 B Pass 4 GA MAN C C n/a n/a n/a n/a Isolation Vahc V14-22A Condensate Transfer Flush I-09 '2 C Act 2 CK SA C C n/a SC Q n/a Note 11 Check Vaht V14-22B Condensate Transfer Flush B-10 2 C Act 2 CK SA C C n/a SC Q n/a Note ll Check Vaht V14-23A Condensate Transfer Flush I-09 2 C Act 2 CK SA C C n/a SC Q n/a Note ll Check Vaht V14-23B Condensate Transfer Flush B-09 2 C Act 2 CK SA C C n/a SC Q n/a Note 11 Check Vaht V14-26A Core Spray Pump Test F-10 2 B Act 8 GL MO-AC C C FAI PIT 2Y n/a Line Isolation Valve STC Q n/a V14-26B Core Spray Pump Test E-10 2 B Act 8 GL MO-AC C C FAI PIT 2Y n/a Line Isolation Vahr STC Q n/a V14-30A Core Spray Injection Check F-07 2 A Pass .75 GA MAN C C n/a LT 2Y n/a Note 21 Vaht Bypass Vahr V14-30B Core Spray Injection Check D-07 2 A Pass .75 GA MAN C C n/a LT 2Y n/a Note 21 Valve Bypass Vaht Revision 18 Drawing : G-191168 Section 5 Page 37 of 165  ;

Vermont Yankee Nuclear Power Station O @

Inservice Testing Program Table 5-1 Valve Listing Crawing : G-191168 Drawing

Title:

Core Spray System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V14-33A Corc Spray Discharge 11-11 2 C Act i CK SA O/C C n/a SC Q n/a Note ll Pressurization Line Check Vaht V14-33B Core Spray Discharge 11-15 2 C Act 1 CK SA O/C C n/a SC Q n/a Note 11 Pressurization Line Check Vahr V14-5A Core Spray Pump 11-1 2 2 B Act 3 GA MO-AC O O/C FAI PIT 2Y n/a Minimum Flow Vahr STC Q n/a ,

STO Q n/a V14-5B Core Spray Pump 11-13 2 B Act 3 GA MO-AC O O/C FAI PIT 2Y n/a Minimum Flow Vahr STC Q n/a STO Q n/a V14-6A Test Connection K-04 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V14-6B Test Connection L-04 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V14-7A Core Spray Pump Suction K-11 2 B Act 12 GA MO-AC O/KL O/C FAI PIT 2Y n/a Vahr STC Q n/a V14-7B Core Spray Pump Suction L-11 2 B Act 12 GA MO-AC O/KL O/C FAI PIT 2Y n/a Vaht STC Q n/a V14-843A Test Connection 11-0 7 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V14-843B Test Connection C-07 2 B Pass .75 GL MAN , LC C n/a n/a n/a n/a V14-8A Core Spray Supply From K-12 2 B Pass 12 GA MAN LC C n/a n/a n/a n/a CSTIsolation Vahr V14-8B Core Spray Supply From L-14 2 B Pass 12 GA MAN LC C n/a n/a n/a n/a CST isolation Vaht Revision 18 Drawing : G-191168 Section 5 Page 38 of 160

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing .

Crawing : G-191169 Sh I Drawing

Title:

liigh Pressure Coolant Injection System Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Ted Ted CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch) Pos Pos Pos Type Freq RR i.

FCV-23-42 IIPCI Steam Line Drain K-13 NNS B Act 1 GL AO O O/C FC FST Q n/a Pot Isolation vaht PIT 2Y n/a STC Q n/a STO Q n/a FCV-23-43 IIPCI Steam Line Drain K-14 NNS B Act i GL AO O O/C FC FST Q n/a Pot Isolation valve PIT 2Y n/a i STC Q n/a STO Q n/a LCV-23-53 Mn Stm Drain Trap Bypass 11-14 2 B Act 1 GL AO C O FC FST Q n/a

& Strainer BlowofITrap PIT 2Y n/a '

Bypass STC Q n/a STO Q n/a PAS-106* Post Accident Sampling J-05 * * * * * * * * * * * *

• See G-191165 for vaht System Purge Vahr info and test requirements S-23-6 IIPCI Turbine Exhaust I 06 2 D Act 16 RD SA C O/C n/a RD SY n/a Line Rupture Disc S-23-7 HPCI Turbine Exhaust R06 NNS D Act 16 RD SA C O/C n/a RD SY n/a Line Rupture Disc '

sir 23-37A Nuclear Boiler Vessel Inst F-05 2 A/C Act 75 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe SL-23-37B Nuclear Boiler Vessel Inst F-05 2 A/C Act .75 EFC SA O O/C FC LEF RO ROJ-Vol Excess Flow Check Vahr Sle23-37C Nuclear Boiler Vessel Inst F-05 2 A/C Act .75 EFC SA O O/C FC LEF RO ROJ-Vol Excess Flow Check Vahr SL-23-37D Nuclear Boilcr VesselInst F-05 2 A/C Act .75 EFC SA O O/C FC , LEF RO ROJ-V01 Excess Flow Check Vahr SSC-23-12 IIPCI Turbine Exhaust I-03 2 C Act 20 GSC SA C O n/a SO Q n/a Stop Check Vahe SSC-23-13 IIPCI Turbine Exhaust J-04 2 C Act 2 GSC SA C O n/a CD RO n/a Drain Stop Check Vahr PSO Q n/a Revision 18 Drawing : G-191169 Sh 1 Section 5 Page 39 of 160

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Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191169 Sh I Drawing

Title:

High Pressure Coolant Injection System Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ I

Valve Namber Nomenclaterr hads Coor y Cat Pass (inch) Pos Pos Pos Type Freg RR  ;

V23-14 HPCI Steam Supply Vaht F-13 2 B Act 10 GA MO-DC C O FAI PIT 2Y n/a STO Q n/a V23-15 HPCI Steam Supply D-04 1 A Act 10 GA MO-AC O O/C FAI LJ 2Y n/a inboard Containment PIT 2Y n/a -

Isolation Vaht STC Q 2/a STO Q n/a V23-153A HPCI Torus Supply Line L-03 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Drain Valve V23-16 HPCI Steam Supply D 05 1 A Act 10 GA MO-DC O O/C FAI LJ 2Y i. a inboard Containment PIT 2Y n/a Isolation Valve STC Q n/a STO Q n/a i V23-160A HPCI Steam Supply Drai:n E-05 2 A Pass 1.75 GL MAN LC C n/a LJ 2Y n/a Vahc V23-17 HPCI Pump Suction Vaht D-11 2 B Act 14 GA MO-DC O O/C FAI PIT 2Y n/a from Condensate Storage STC Q n/a Tank STO Q n/a V23-18 IIPCI Pump Discharge to G-05 2 C Act 14 CK SA C O/C n/a SC CS CSJ-V06 Feedwater Check Vahc -

SO CS CSJ-V06 V23-19 IIPCI Discharge to G-06 2 B Act 14 GA MO-DC C O FAI PIT 2Y n/a Feedwater Isolation Vaht STO Q n/a V23-20 HPCI Pump Discharge G-08 2 B Act 14 GA MO-DC O O FAI PIT 2Y n/a Vahr STO Q n/a V23-20B Pressurizing and Vent G-07 2 C Act 1 CK SA C C n/a SC Q n/a Note 11 Vahrs HPCI to RX Keep Fill V23-21 IIPCI Pump Test Return E-07 2 B Act 10 GL MO-DC C C FAI PIT 2Y n/a Valve to Condensate STC Q n/a Storage Tank ,

L Revision 18 Drawing : G-191169 Sh 1 Section 5 Page 40 of 160 ,

Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191169 Sh 1 Drawing

Title:

liigh Pressure Coolant Injection System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V23-25 IIPCI Pump Minimum F-09 2 D Act 4 GL MO-DC C O/C FAI PIT 2Y n/a Recire Valve STC Q n/a STO Q n/a V23-26B IIPCI Exhaust Drain to J44 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Torus Test Connection Vaht V23-27A IIPCI Test Connection E-04 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V23-32 IIPCI Pump Suction Check E-l1 2 C Act 14 CK SA O/C O/C n/a SC Q n/a Note 11 Vahr from Cond. Storage SO Q n/a Tank V23-56 IIPCI Turbine Exhaust J-05 2 C Act 2 CK SA O/C C n/a CD RO n/a Drain Check Vaht LT 2Y n/a PSO Q n/a SC RO ROJ-V25 V23-57 IIPCI Torus Suction F-10 2 B Act 16 .GA MO-DC C O/C FAI PIT 2Y n/a Containment Isol. Vah c - STC Q n/a Outbd STO n/a Q

V23-58 IIPCI Torus Suction L-04 2 B Act 16 GA MO-DC C O/C FAI PIT 2Y n/a Containment Isolation STC Q n/a Vaht STO Q n/a V23-59A IIPCI Torus Supply Test L-05 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Connection Vaht V23-61 IIPCI Tonis Suction Check L-06 2 C Act 16 CK SA C O n/a CD RO n/a Valve.

V23-62 IIPCI Pump Minimum J-10 2 C Act 4 CK SA C O/C n/a CD RO n/a Recire Check Vaht PSO Q n/a V23-63A IIPCI Test Connection on I-03 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Torus Exhaust Revision 18 Drawing : G-191169 Sh 1 Section 5 Page 41 of 160

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Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191169 Sh I Drawing

Title:

Iligh Pressure Coolant injection System Valve Number No:nenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hah Coor y Cat Pass (inch) g Pos Pos Pos Type Freq RR V23-65 IIPCI Turbine Exhaust 14)3 2 C Act 20 CK SA C O/C n/a SC RO ROJ-V25 Note 2 Check Vahc SO Q n/a V23-842 IIPCI Gland Seal 1-02 2 C Act 1 GSC SA C O n/a CD RO n/a Exhauster Return Stop PSO n/a Q

Check Vaht V23-843 IIPCI Gland Seal 1-03 2 C Act 1 CK SA C O/C n/a CD RO n/a Exhauster Return Check PSO n/a Q

Vahc t

Revision 18 Drawing : G-191169 Sh 1 Section 5 Page 42 of 160

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-s hv b(iv Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191169 Sh 2 Drawing

Title:

liigh Pressure Coolant Injection System Vaht Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act g,g Coor y Cat Pass (inch Pos Pos Pos Type Freq RR HPCI- IIPCI Steam Supply F-13 2 B Act 10 GL 11 0 O O FC PIT 2Y n/a Note 12 CONTROL Turbine Control Valve SKID Q n/a HPCI-STOP IIPCI Turbine Stop Vahe F-15 2 B Act 10 GA 11 0 C O/C FC PIT 2Y n/a Note 12 SKID Q n/a LCV-23-39 IIPCI Gland Seal Return M-11 3 B Act 1 GL AO C O/C FC FST Q n/a Level Cont Drn Vahr to PIT 2Y n/a Radwaste STC Q n/a STO Q n/a LCV-23-40 IIPCI Gland Seal Return M-l 1 3 B Act 1 GL AO C O/C FC FST n/a Q

Level Cont Drn Vahr to PIT 2Y n/a Radwaste STC Q n/a STO Q n/a i

LCV-23-54 IIPCI Drain to Gland Scal 1-12 2 B Act i GL AO C O FC FST Q n/a Condenser PIT 2Y n/a STC Q n/a STO Q n/a PCV-23-50 Pressure Control Vahr J-12 2 B Act 2 GL AO O O FO SKID Q n/a Cooling Water to Lube Oil Cooler SR-23-34 HPCI Safety Valve D-09 2 C Act 1.5 RV SA C O/C n/a SP 10Y n/a SR-23-66 IIPCI Booster Pump Gland J-12 2 C Act 1.5 RV SA C O/C n/a SP 10Y n/a t Seal Cond. Line Relief Vaht V23-130 IIPCI Gland Seal Tube L-11 2 C Act 2 CK SA C O n/a SKID Q n/a Note 12 l Side Return Check Valve V23-131 IIPCI Gland Seal Cond. L-11 2 C Act 2 CK SA C O/C n/a SKID Q n/a Note 12 Shell Side Return Check Vahr V23-148A IIPCI Steam Ring / Chest G-14 3 C Act I CK SA O/C O n/a SKID Q n/a Note 12 Drain Check Vaht Revision 18 Drawing : G-191169 Sh 2 Section 5 Page 43 of 160 l

. _ - _ - -_ _ _ ~__ - __ _ _ _ - _-_ _ _ _ _ - _ _ _ __- __ _ _ _ _ _ __- - - _ _- -__ - __ _ _ _ - _ - _ - _ _ -

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Vcrmont Yankee Nuclear Power Station. Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191169 Sh 2 Drawing

Title:

High Pressure Coolant injection System Valve Number Nomenclature Dwg Safet OM Act / Size Nom Safety Fail Test Test CSJ/ROJ Body Act  %,g Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V23-148B llPCI Steam Ring / Chest G-14 3 C Act I CK SA O/C O n/a SKID Q n/a Note 12 Drain Check Vahr V23-50A IIPCI Gland Scal Cond. J-10 2 B Act 2 GA AO C O FO PIT 2Y n/a Note 12 Supply Vahr SKID Q n/a F

Revision 18 Drawing : G-191169 F 2 Section 5 Page 44 of 160 l

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Draning : G-191170 Drawing Titic: Control Rod Drive Ilydraulic System Vaht Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ '

Body Act hah Coor y Cat Pass (inch Pos Pos Pos Type Freg RR CV-3-126 Control Rod Drive Scram C-16 2 B Act 1 GL AO C O FO FST RO ROJ-V06 Note 16 Vahr (Typical of 89) PIT 2Y n/a  ;

STO RO ROJ-V06 CV-3-127 Control Rod Drive Scram C-18 2 B Act .75 GL AO C O FO FST RO ROJ-V06 Note 16 Vahe (Typical of 89) PIT 2Y n/a STO RO ROJ-V06 ,

r LCV-3-33A Scram Discharge Volume D-07 2 A Act 2 GL AO O C FC FST Q n/a Drain Level Control Vaht LT 2Y n/a PIT 2Y n/a STC Q n/a LCV-3-33B Scram Discharge Volume D-03 2 A Act 2 GL AO O C FC FST Q n/a Drain Level Control Vaht LT 2Y n/a PIT 2Y n/a STC Q n/a  !

LCV-3-33C Scram Discharge Volume D-08 2 A Act 2 GL AO O C FC FST Q n/a ,

Drain Level Control Vaht LT 2Y n/a PIT 2Y n/a STC Q n/a LCV-3-33D Scram Discharge Volume D-03 2 A Act 2 GL AO O C FC FST Q n/a Drain Level Control Vaht LT 2Y n/a PIT 2Y n/a STC Q n/a PCV-3-32A Scram Discharge Volume A-09 2 A Act i GL AO O C FC FST Q n/a Vent Pressure Control LT 2Y n/a Vaht PIT 2Y n/a STC Q n/a PCV-3-32B Scram Discharge Volume A-01 2 A Act 1 GL AO O C FC FST Q n/a ~

Vent Pressure Control LT 2Y n/a Vahe PIT 2Y n/a STC Q n/a Revision 18 Drawing : G-191170 Section 5 Page 45 of 160 i

b C i.)

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Crawing : G-191170 Drawing

Title:

Control Rod Drive Ilydraulic System Vahe Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hah Coor y Cat Pass (inch Pos Pos Pos Type Freq RR S-3-132 HCU Rupture Disk (typ of D-16 2 D Act .5 RD SA C O/C n/a RD SY n/a 89) 5 0-3-120 CRD Manual Control D-17 2 B Act .5 GA SO C C FC FST RO ROJ-V06 Note 17 Exhaust Vaht (Typical of STC RO ROJ-V06 89) 5 0-3-121 CRD Manual Control D-18 2 B Act .5 GA SO C C FC FST RO ROJ-V06 Note 17 Exhaust Vaht (Typical of STC RO ROJ-V06 89) 5 0-3-122 CRD Manual Control C-18 2 B Act .5 GA SO C C FC FST RO ROJ-V06 Note 17 Withdrawal Vaht (Typical STC RO ROJ-V06 of89)

S 0-3-123 CRD Manual Control C-17 2 B Act .5 GA SO C C 1C FST RO ROJ-V06 Note 17 Insertion Vaht (Typical of STC RO ROJ-V06 89)

V3-114 Scram Exhaust To C-18 2 C Act .75 CK SA C O/C n/a SC RO ROJ-V06 Note 18 Discharge Volume Check SO RO ROJ-V06 (Typical of 89)

V3-115 CRD Charging Water To B-16 2 A/t Act .5 CK SA O/C C n/a LT RO ROJ-V06 Note 19 Accumulator Check SC RO ROJ-V06 (Typical of 89)

V3-130 Drive Water Insert High B-14 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Point Vent (Typical of 89)

V3-132 Drhc Water Withdrawal A-14 2 B Pass .75 GL MAN -C C n/a n/a n/a n/a High Point Vent (Typical of89)

V3-137 Drive Water Supply Check B-17 2 C Act .5 CK SA O/C C n/a SC RO ROJ-V06 Valve (Typical of 89)

V3-138 CRD Cooling Water B-16 2 C Act .5 CK SA O/C C n/a SC n/a Q Note 3 Supply Check Vaht (Typical of 89)

R. vision 18 Drawing : G-191170 Section 5 Page 46 of 160

Vermont Yankee Nuclear Power Station O @

Inservice Testing Program Table 5-1 Valve Listing '

Drawing : G-191170 Drawing

Title:

Control Rod Drive flydraulic System Valve Number Nomenclature Dwg Saftt OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Rema h Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V3-162A Scram Discharge Volume A-09 2 NC Act 1 CK SA O C n/a LT 2Y n/a Vent Check Valve SC RO ROJ-V12 V3-162B Scram Discharge Volume A-01 2 NC Act i CK SA O C n/a LT 2Y n/a Vent Check Vahr SC RO ROJ-V12 V3-408A Recirc Pump Seal Purge F-12 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Vahts V3-408B Recirc Pump Seal Purge F-12 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Vaht V3-410A Recirc Pump Scal Purge F-13 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Vaht V3-410B Recirc Pump Seal Purge F-13 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Vaht V3-412A Recirculation Pump Scal E-12 2 NC Act .75 CK SA O/C C n/a LJ 2Y n/a Purge Supply Check Vaht SC RO ROJ-V07 V3-412B Recirculation Pump Seal E-12 2 NC Act .75 CK SA O/C C n/a LJ 2Y n/a Purge Supply Check Vahe SC RO ROJ-V07 V3-413A Recirculation Pump Seal E-12 2 NC Act .75 CK SA O/C C n/a LJ 2Y n/a Purge Supply Check Vaht SC RO ROJ-V07 V3-413B Recirculation Pump Seal E-12 7 NC Act .75 CK SA O/C C n/a LJ 2Y n/a Purge Supply Check Valve SC RO ROJ-V07 e

t Revision 18 Drawing : G-191170 Section 5 Page 47 of 160

1 Vermont Yankee Nuclear Power Station

@ O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191171 Drawing

Title:

Standby Liquid Control System Vahe Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR S-11-14 A SLC Injection (Squib) G-03 2 D Act 1.5 SQUlB EXP C O n/a ET 2Y n/a Note 4 Vahr S-11-14B SLC Injection (Squib) 11-03 2 D Act 1.5 SQUIB EXP C O n/a ET 2Y n/a Note 4 Vahr SR-11-39A SLC Discharge Relief Vahr G-08 2 C Act 1 RV SA C O/C n/a SP OC RR-V08 SR-11-39B SLC Discharge Relief Vahr K-08 2 C Act i RV SA C O/C n/a SP OC RR-V08 V11-16 SLC Injection Piping 11-0 3 I C Act 1.5 CK SA C O/C n/a SC RO ROJ-V08 Check Vahr SO RO ROJ-V08 Vil-17 SLC Injection Piping 1-02 1 C Act 1.5 CK SA C O/C n/a SC RO ROJ-V08 Check Vaht SO RO ROJ-V08 VII-18 SLC Injection Vahr 1-02 i B Pass 1.5 GA MAN LO O n/a PIT 2Y n/a Note 14 VII-36 SLC Test Connection 1-03 2 D Pass .75 GL MAN LC C n/a n/a n/a n/a V11-43A SLC Pump Discharge 11-0 7 2 C Act 1.5 CK SA C O/C n/a SC Q n/a Check Vahr SO Q n/a V11-43B SLC Pump Discharge J-07 2 C Act 1.5 CK SA C O/C n/a SC Q n/a Check Vahr SO Q n/a Revision 18 Drawing : G-191171 Section 5 Page 48 of 160

b v bv he Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Crawing : G-191172 Drawing

Title:

Residual lleat Removal System Dwg Safet OM Act / Size l Norm Safety Fail Test Test CSJ/ROJ Valve Number Nomenclature g h4 Coor y Cat Pass (inch y l Act Pos Pos Pos Type Freq RR FCV-10-160 RilR Return Line Process F-15 2 B Act .75 GL AO C C FC FST Q n/a Vahr also appears on G-Sampling Vaht PIT 2Y n/a 191165 STC Q n/a RV-10-210A ReliefVahr E-12 1 A/C Pass .5 RV SA C C n/a LJ 2Y n/a RV-10-210B ReliefVahr E-06 i A/C Pass .5 RV SA C C n/a LJ 2Y n/a SR-10-35A RilR Pump Discharge D-02 2 C Act i RV SA C O/C n/a SP 10Y - n/a ReliefVahr SR-10-35B RilR Pump Discharge D-15 2 C Act 1 RV SA C O/C n/a SP 10Y n/a ReliefVahe SR-10-40 RilR Shutdown Cooling G-07 2 C Act 1 RV SA C O/C n/a SP 10Y n/a Suction Line Relief Vahc SR-10-72A RilR Relief Vaht L-07 2 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-10-72B RIIR Relief Vaht L-l1 2 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-10-72C RilR Relief Vaht J-07 2 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-10-72D RilR Relief Vaht J-10 2 C Act 1 RV SA C O/C n/a SP 10Y n/a SR-19-80A RilR IIcat Exchanger Tube L-03 2 C Act 4 RV SA C O/C n/a SP 10Y n/a Side Relief Vahr SR-10-80B RIIR llcat Exchanger Tube L-14 2 C Act 4 RV SA C O/C n/a SP 10Y n/a Side Relief Vahr SR-10-86A RilR IIcat Exchanger ShcIl L-03 2 C Act i RV SA C O/C n/a SP 10Y n/a Side Relief Vahr SR-10-86B RilR llcal Exchanger Shell L-14' 2 C Act 1 RV SA C O/C n/a SP 10Y n/a Side Relief Vaht V10-13A Suppression Pool RilR L-08 2 B Act 20 GA MO-AC O/KL O/C FAI PIT 2Y n/a Pump Suction Vahr STC Q n/a STO Q n/a Revision 18 Drawing : G-191172 Section 5 Page 49 of 160

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Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191172 Drawing

Title:

Residual IIcat Removal System Valve Number Nomenclature Dwg Safet OM Act / Size Nom Safetybail Test Test CSJ/ROJ Body Act g,,,g Coor y Cat Pass (inch Pos Pcs l Pos Type Freq RR V10-13B Suppression Pool RIIR L-09 2 B Act 20 GA MO-AC O/KL O/C FAI PIT 2Y n/a Pump Suction Vahe '

STC Q n/a STO Q n/a V10-13C Suppression Pool RIIR J-08 2 B Act 20 GA MO-AC O/KL O/C FAI PIT 2Y n/a Pump Suction Vaht STC Q n/a STO Q n/a V10-13D Suppression Pool RIIR J-09 2 B Act 20 GA MO-AC O/KL O/C FAI PIT 2Y n/a Pump Suction Vahr STC Q n/a STO Q n/a V10-15A Recirculation Loop Supply K-07 2 B Act 20 GA MO-AC C O/C FAI PIT 2Y n/a Valve to RIIR Pump STC n/a Q

Suction STO n/a Q

V10-15B Recirculation Loop Supply K-10 2 B Act 20 GA MO-AC C O/C FAI PIT 2Y n/a Vahr to RIIR Pump STC Q n/a Suction STO n/a Q

V10-15C Recirculation Loop Supply J-07 2 B Act 20 GA MO-AC C O/C FAI PIT 2Y n/a Vahr to RIIR Pump STC Q n/a Suction STO n/a Q

V10-15D Recirculation Loop Supply J-10 2 B Act 20 GA MO-AC C O/C ' FAI PIT 2Y n/a '

Vaht to RIIR Pump STC Q n/a Suction STO Q n/a V10-16A RilR Pump Discharge I-05 2 B Act 4 GA MO-AC O O/C FAI PIT 2Y n/a Mini Flow Return to STC Q n/a Suppression Pool STO Q n/a V10-16B RIIR Pump Discharge 1-12 2 B Act 4 GA MO-AC O O/C FAI PIT 2Y n/a Mini Flow Return to STC Q n/a Suppression Pool STO Q n/a Revision 18 Drawing : G-191172 Section 5 Page 50 of 160

h Vermont Yankee Nuclear Power Station h h Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191172 Drawing

Title:

ResidualIIcat Removal Sysem Dwg Safet OM Act / Size Valve Number Nomenclature Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y i Cat Pass (inch Pos Pos Pos Type Freq RR V10-17 Recire Supply to RIIR G-08 1 A At.t 20 GA MO-DC C O/C FAI U 2Y n/a Note 21 Pump Suct. Outbd. LT 2Y n/a isolation Vaht PIT 2Y n/a STC CS CSJ-V08 STO CS CSJ-V08 V10-17Al V10-17 Bonnet Vent Vaht G-08 2 B Act .5 GL MAN C O/C n/a SC CS CSJ-V18 V10-18 Recire Supply to RilR F-08 I A Act 20 GA MO-AC C O/C FAI U 2Y n/a Note 21 Pump Suction Inbd LT 2Y n/a Isolation Vaht PIT 2Y n/a STC CS CSI-V08 STD CS CSJ-V08 V10-182 RilRSW to RilR I-02 2 C Pass 10 CK SA C n/a n/a n/a n/a n/a Emergency Fill Check Vaht V10-183 RilRSW to RilR J-02 2 B Pass 10 GA MO-AC C C FAI PIT 2Y n/a Emugency Fill Isolation Vahc V10-184 RilRSW to RilR J-02 2 B Pass 10 GA MO-AC C C FAI PIT 2Y n/a Emergency Fill Isolation Valve V10-18A Penetration X-12 Thermal F-08 2 A/C Act .50 CK SA C O/C n/a U 2Y n/a Note 21 Relief Vaht LT 2Y n/a SC RO ROJ-V29 SO RO ROJ-V29 V10-18A5 V10-18 Bannet Relief F-08 2 A/C Act .50 CK SA C O/C n/a U 2Y n/a Note 21 Vaht LT 2Y n/a SC RO ROJ-V29 SO RO ROJ-V29 V10-1988 Process Sampling Vahr E-14 2 A Pass .75 GL MAN LC C n/a U 2Y n/a Revision 18 Drawing : G-191172 Section 5 Page 51 of 160

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Vermont Yankee Nucicar Power Station Inservice Testing Program Table 5-1 Valve Listing Drewing : G-191172 Drawing Titic: Residualllcat Removal System Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Valve Number Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V10-19A RiiR Minimum Flow L-05 2 C Act 3 CK SA O/C O/C n/a CD RO n/a Check Valve PSO Q n/a t

V10-19B RilR Minimum Flow L-12 2 C Act 3 CK SA O/C O/C n/a CD RO n/a Check Valve PSO Q n/a Vi>19C RilR Minimum Flow J-05 2 C Act 3 CK SA O/C O/C n/a CD RO n/a Check Vahr PSO Q n/a V10-19D RIIR Minimum Flow J-12 2 C Act 3 CK SA O/C O/C n/a CD RO n/a Check Vahr PSO Q n/a V10-20 RilR Loop Crossconnect 114)5 2 B Pass 20 GA MO-AC C C FAI PIT 2Y n/a  ;

Vaht V10-206A RIIR Suppression Pool I-08 2 B Pass .75 GA MAN C C n/a n/a n/a n/a  !

Suction Drain V1D-206B RilR Suppression Pool I-09 2 B Pass .75 GA MAN C C n/a n/a ria n/a Suction Linc Drain V10-25A RIIR to Recirc Loop E-06 1 B Act 24 GA MO-AC O O FAI PIT 2Y n/a Isolation Vaht STO Q n/a VI*e-25B RilR to Recirc Loop E-12 i B Act 24 GA MO-AC O O FAI PIT 2Y n/a Isolation Vahr STD Q n/a V10-26A RIIR To Containment C-07 2 A Act 12 GA MO-AC C O/C FAI LJ 2Y n/a Spray Isolation Vaht PIT 2Y n/a STC Q n/a STO Q n/a V10-26B RilRTo Containment C-10 2 A Act 12 GA MO-AC C O/C FAI LJ 2Y n/a ,

Spray Isolation Vaht PIT 2Y n/a STC Q n/a STD Q n/a Revision 18 Drawing : G-191172 Section 5 Page 52 of 160

0 huf f-)

a hu Vermont Yankee Nuclear Power Station Insenice Testing Program Table 5-1 Valve Listing Drawing : G-191172 Drawing Titic: Residual Heat Removal System Valve Number Nomenclature Dwg Safet OM Act / Size Body Norm Safety Fail Test Test CSJ/ROJ Act hah Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V10-27A RIIR to Recire Loop D-06 1 A Act 24 GL h10-AC C O/C FAI U 2Y n/a Note 21 Isolation Valve LT 2Y n/a PIT 2Y n/a STC Q n/a STD Q n/a V10-27B RilR to Recire Loop D-12 1 A Act 24 GL MO-AC C O/C FAI LJ 2Y n/a Note 21 Isolation Valve LT 2Y n/a PIT 2Y n/a STC Q n/a STO Q n/a V10-31A RIIR To Containment C-08 2 A Act 12 GA MO-AC C O/C FAI U 2Y n/a Spray isolation Vahr PIT 2Y n/a STC Q n/a STO Q n/a V10-31B RIIR To Containment C-10 2 A Act 12 GA MO-AC C O/C FAI U 2Y n/a Spray isolation Vahr PIT 2Y n/a STC Q n/a STO Q n/a V10-34A RIIR Suppression Pool E-03 2 A Act 10 GL MO-AC C O/C FAI U 2Y n/a Cooling Supply Vaht PIT 2Y n/a STC Q n/a STO Q n/a V10-34B RIIR Suppression Pool E-14 2 A Act 10 GL MO-AC C O/C FAI LJ 2Y n/a Cooling Supply Vaht PIT 2Y n/a STC Q n/a STO Q n/a V10-36A RilR Pump Discharge J-03 2 C Act 1 CK SA O/C C n/a SC Q n/a Note 11 Pressurizing Line Check Valve Revision 18 Drawing : G-191172 Section 5 Page 53 of 160

Vcrmont Yankee Nuclear Power Station

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Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191172 Drawing

Title:

Residual Heat Removal System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ  %,g Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V10-36B RHR Pump Discharge J-14 2 C Act 1 CK SA O/C C p!a SC Q n/a Note 11 Pressurizing Line Check Vaht V10-38A Suppression Chamber RIIR E-04 2 A Act 4 GL MO-AC C O/C FAI U 2Y n/a Spray Ring Supply Vaht PIT 2Y n/a STC Q n/a STO Q n/a V10-38B Suppression Chamber RHR F-13 2 A Act 4 GL MO-AC C O/C FAI U 2Y n/a Spray Ring Supply Vahe PIT 2Y n/a STC Q n/a STO Q n/a V10-39A RHR Cont D4)4 2 A Act 12 GA MO-AC C O/C FAI U 2Y n/a Spray / Suppression Pool PIT 2Y n/a Cooling Supply Valve STC Q n/a STO Q n/a V10-39B RHR Cont D-13 2 A Act 12 GA MO-AC C O/C FAI U 2Y n/a Spray / Suppression Pool PIT 2Y n/a Cooling Supply Vaht ,

STC Q n/a STO Q n/a V10-46A RHR to Recirc injection E-07 1 A/C Act 24 CK SA C O/C n/a LT 2Y n/a Note 21 Check Vaht dC Q n/a SO CS CSJ-V14 V10-46B RHR to Recirc injection E-10 1 A/C Act 24 CK SA C O/C n/a LT 2Y n/a Note 21 Check Vahe SC Q n/a SO CS CSJ-V14 V10-48A RHR Pump Discharge L-05 2 C Act 16 CK SA O/C O/C n/a PSO Q CSJ-VIS Check Valve SC Q n/a SO CS CSJ-VIS Revision 18 Drawing : G-191172 Section 5 Page 54 of 160

Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing .

Drawing : G-191172 Drawing

Title:

Residual Heat Removal System .

Valve Number Nonsenclature Dwg Safet OM Act / Size None Safety Fail Test Test CSJ/ROJ Body Act %g Coor y Cat Pass (inch Pos Pos Pos Type Farq RR V10-48B RHR Pump Discharge L-13 '2 C Act 16 CK SA O/C O/C n/a PSO Q CSJ-V15 1 Check Valve SC n/a Q

SO CS CSJ-VIS V10-48C RHR Pump Discharge J-05 2 C Act 16 CK SA O/C O/C n/a PSO Q CSJ-VIS Check Valve SC n/a Q ,

SO CS CSJ-VIS i

V10-48D RHR Pump Discharge J-13 2 C Act 16 CK SA O/C O/C n/a PSO Q CSJ-VIS  :'

Check Valve SC n/a Q

SO CS CSJ-VIS V10-52A RHR Containment Spray C-08 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Loop A Test Connection V10-52B RHR Containment Spray C-10 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Loop B Test Connection V10-57 RHR Loop Crossconnect H-13 2 A Act 4 GA MO-DC C C FAI LT 2Y n/a To RadWaste Isolation PIT 2Y n/a i Vahr STC Q n/a V10-65A RHR HX Bypass Vahr J-03 2 B Act 20 GL MO-AC O O/C FAI PIT 2Y n/a STC Q n/a l STO Q n/a V10-65B RHR HX Bypass Vahr K-15 2 B Act 20 GL MO-AC O O/C FAI PIT 2Y n/a STC Q n/a STO Q n/a  !

V10-66 RHR Loop Crossconnect H-13 2 A Act 4 GA MO-AC C C FAI LT 2Y n/a To RadWaste Isolation PIT 2Y n/a Vahr STC Q n/a V10-78A RHR "A" Injection Line F 06 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Test Connection V10-78B RHR "B" Injection Line F-12 2 B Pass .75 GL MAN C C n/a n/a n/a n/a

. Test Connection Revision 18 Drawing : G-191172 ' Section 5 Page 55 of 160 '

L Vcrmont Yankee Nuclear Power Station

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Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191172 Drawing

Title:

Residuallicat Removal System Vaht Number Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ i Nomenclature Body Act  %,g Coor y Cat Pass (inch Pos Pos Pos Type Freq RR ~

V12-81A RIIR to Recite Loop E-07 2 B Pass 24 GA MAN LO O n/a PIT 2Y n/a Note 14 '

Injection Vahr j V13-8 t B RIIR io Recirc Loop E-10 2 B Pass 24 GA MAN LO O n/a PIT 2Y n/a Note l4

• Injection Vaht t VIO-84 Test Connection Between G-09 2 B Pass .75 GL MAN C C n/a n/a n/a n/a f V10-17 and 18 VI*.-89A . RIIRllX Senice Water M-UI 3 B Act 12 GL MO-AC C O/C FAI PIT 2Y n/a Vahr also appears on G-Outlet Vahr STC Q n/a 191159 Sh 2 STO Q n/a VI'A19B RIIRIIX Scnice Water M-17 3 B Act 12 GL MO-AC C O/C FAI PIT 2Y n/a Valve also appears on G-Outlet Vahr STC Q n/a 191159 Sh 2 STO Q n/a V10-91A Test Connection E-07 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a i Vle-9tB Test Connection E-Il 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V10-95A Test Connection D-04 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V10-95B Test Connection D-14 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V10-%A Test Connection C-08 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a V10-%E Test Connection C-10 2 B Pass .75 GA MAN LC C n/a n/a n/a n/a i

t r

Revision 18 Drawing : G-191172 Section 5 Page 56 of 160

9 Vermont Yankee Nuclear Power Station O @

Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191173 Sh 1 Drawing

Title:

Fuct Pool Cooling & Cleanup System Valve Namber Nomenclaterr Dwg Safet OM Act / Size Nonn Safety Fail Test Test CSJ/ROJ Body Act h at'a Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V19-18 Fuel Pool Dcmineralizer 1-13 3 C Act 6 CK SA O C n/a SC Q n/a Return Check Vahr V19-21 A Fuel Pool Cooling Return A-07 2 C Act 6 CK SA O O n/a SO Q n/a to Fuel Pool Check Vahr Vl9-21B Fuel Pool Cooling Return A-07 2 C Act 6 CK SA O O n/a SO Q n/a to Fuel Pool Check Vahr V19-220 Normal Fuct Pool Cooling G-07 3 D Act 8 GA MO-AC O C FAI PIT 2Y n/a Subsystem isolation Vahr STC Q n/a V19-221 Normal Fuel Pool Cooling G-08 3 B Act 8 GA MO-AC O C FAI PIT 2Y n/a Subsystem Isolation Vahr STC Q n/a V19-224 Fuel Pool Demineralizer I-12 3 C Act 6 CK SA O C n/a SC RO ROJ-V23

• Return Check Vahr V19-46 U Gate Vahr F/D Bypass K-06 3 B Pass 4 GA MAN C C n/a ' n/a n/a n/a V19-53 U Gate Vahr RX WellL H-05 3 B Pass 6 GA MAN C C n/a n/a n/a n/a Recire Vahr i

Revision 18 Drawing : G-191173 Sh 1 Section 5 Page 57 of 160

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Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191173 Sh 2 Drawing

Title:

Fuel Pool Cooling & Cleanup System Vaht Number Ncmenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hah Coor y Cat Pass (inch Pos Pos Pos Type Freq RR RVl9-232A Standby Fuel Pool 11 eat C-07 3 C Act 1 RV SA C O/C n/a SP 10Y n/a Exchanger FPC Side Relief Valve RVl9-232B Standby Fuct PoolIIcat F-07 3 C Act 1 RV SA C O/C n/a SP 10Y n/a Exchanger FPC Side Relief Vahr i RV70-260A Senice Water Relief Valve D-08 3 C Act .75 RV SA C O/C n/a SP 10Y n/a i

RV70-260B Senice Water Relief Vaht F-08 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Vl9-226 Standby FPCS Manual C-13 3 B Pass 6 GA MAN C C n/a n/a n/a n/a Isolation - ESS Pump I Cross-tic Vl9-230A Standby Fuel Pool Cooling G-10 3 C Act 6 CK SA C O/C n/a SC Q n/a Pump Discharge Check SO Q n/a  !

Vaht Vl9-230B Standby Fuel Pool Cooling 11-1 0 3 C Act 6 CK SA C O/C n/a SC Q n/a Pump Discharge Check SO Q n/a Vaht '

V70-244A SW System Vacuum B-10 3 C Act 2 CK SA C O/C n/a SC RO ROJ-V28 Breaker Check Valve 50 RO ROJ-V28 ,

SP 10Y n/a V70-244 B SW System Vacuum B-10 3 C Act 2 CK SA C O/C SC RO ROJ-V28 Breaker Check Vahr SO RO ROJ-V28 '

SP 10Y n/a V70-257A SFPC IIx Scnice Water C-10 3 B Act 4 GL MO-AC C O FAI PIT 2Y n/a Outlet Throttling / isolation STC Q n/a ,

me STO Q n/a V70-257B SFPC Hx Service Water F-10 3 B Act 4 GL MO-AC C O FAI PIT 2Y n/a Outlet Throttling / isolation STC Q n/a Vaht S70 Q n/a Revision 18 Drawing : G-191173 Sh 2 Section 5 Page 58 of 160

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Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing .

Drawing : G-191174 Sh 1 Drawing Titic. Reactor Core Isolation Cooling System Valte Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hah Coor y Cat Pass (inch Pos Pos Pos Type Freq RR FCV-13-34 RCIC Main Steam Drain to K-17 NNS B Act i GL AO O O/C FC FST Q n/a Condenser Valve PIT 2Y n/a STC Q n/a STO Q n/a FCV-13-35 RCIC Main Steam Drain to L-17 NNS B Act 1 GL AO O O/C FC FST Q n/a Condenser Vaht PIT 2Y n/a STC Q n/a STO Q n/a LCV-13-32 RCIC Steam Drain Trap 11-17 2 B Act .75 GL AO C O/C FC FST Q n/a Dypass Vaht PIT 2Y n/a STC Q n/a STO Q n/a SL-13-55A RCIC Instmmentation B-08 2 NC Act .75 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr SL-13-55B RCIC Instrumentgion C-08 2 NC Act .75 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check hhv SL-13-55C RCIC Instrumentation B-08 2 NC Act .75 EFC SA O O/C n/a LEF RO ROJ-VOI Excess Flow Check Vahr SL-13-55D RCIC Instrumentation C-08 2 NC Act .75 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr SSC-13-10 RCIC Barometric Cond K-09 2 C Act 2 GSC SA C O n/a SO Q n/a Vacuum Pump Disch Stop Chert Vahr SSC-13 9 r.CIC Steam Exhaust Stop K-08 2 C Act 8 GSC SA C O n/a SO Q n/a Check Vaht V13-143A RCIC Pump Suction Test N-09 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection ,

Revision 18 Drawing : G-191174 Sh 1 Section 5 Page 59 of 160 i

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Title:

Reactor Core Isolation Cooling System Dwg Safet OM Act / Size Vaht Number Nomenclature Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch) Pos Pos Pos T3pe Freq RR V13-15 RCIC Steam Supply E-07 i A Act 3 GA MO-AC O O/C FAI LJ 2Y n/a Inboard Containment PIT 2Y n/a Isolation Vaht STC Q n/a STO Q n/a V13-150A Steam Line Drain Root E-09 2 A Pass 1 GL MAN LC C n/a LJ 2Y n/a Vaht V13-16 RCIC Steam Supply E-09 1 A Act 3 GA MO-DC O O/C FAI LJ 2Y n/a Outboard Containment PIT 2Y n/a isolation Vaht STC Q n/a STO Q n/a V13-18 Condensate Supply to E-14 2 B Act 6 GA MO-DC O O/C FAI PIT 2Y n/a RCIC Pump Suction Vahe STC Q n/a STO Q n/a V13-19 Condensate Supply to E-14 2 C Act 6 CK SA O/C O/C n/a SC Q n/a Note 1I RCIC Pump Suction Check SO Q n/a ,

Vaht V13-20 RCIC Pump Discharge G-10 2 B Act 4 GA MO-DC O O FAI PIT 2Y n/a Isolation Vaht STO Q n/a V13 20B Pressurizing Vaht Keep G-ll 2 C Act 1 CK SA C C n/a SC Q n/a Note 11 Fill System V13-21 RCIC Injection Isolation G-09 2 B Act 4 GA MO-DC C O FAI PIT 2Y n/a Vahr STO Q n/a V13-22 RCIC Injection Check G-09 2 C Act 4 CK SA C O/C n/a SC CS CSJ-V10 -

Vaht SO CS CSJ-V10 V13-27 RCIC Pump Discharge I-l1 2 B Act 2 GL MO-DC C O/C FAI PIT 2Y n/a Minimum Flow Vaht STC Q n/a STO Q n/a Revision 18 Drawing : G-191174 Sh i Section 5 Page 60 of 160

h fu;) u

() (2d Vermont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191174 Sh 1 Drawing

Title:

Reactor Core Isolation Cooling System Vaht Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch) Pos Pos Pos Type Freq RR V13-29 RCIC Pump Discharge J-10 2 C Act 2 CK SA C O/C n/a CD RO n/a Minimum Flow Check PSO Q n/a Valve V13-30 RCIC Pump Discharge E-10 2 B Act 4 GL MO-DC C C FAI PIT 2Y n/a Return to Condensate STC Q n/a Storage Tank V13-38 RCIC Vacuum Pump K-11 2 C Act 2 CK SA C O/C n/a SC RO ROJ-V26 Discharge Check Valve SO Q n/a V13-39 Supp Pool Supply to RCIC F-13 2 B Act 6 GA MO-DC C O/C FAI PIT 2Y n/a Pump SuctInbd Isol Vahr STC Q n/a >

STO Q n/a V13-40 Suppression Pool Supply to N-11 2 C Act 6 CK SA C O n/a CD RO n/a RCIC Pump Suction Check VIV.

V13-41 Supp Fool Supply to RCIC N-10 2 B Act 6 GA MO-DC C O/C FAI PIT 2Y n/a Pump Suct Inbd Isol Valve STC Q n/a STO Q n/a V13-43A RCIC Pump Suction Test N-11 2 B Pass .75 GL MAN C C n/a n/a n/a n/a Connection  !

V13-46A RCIC Main Steam Line E-08 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Test Connection V13-50 RCIC Turbine Exhaust K-08 2 C Act 8 CK SA C O/C n/a SC RO ROJ-V26 Check Vahr SO Q n/a V13-528 RCIC Pump Discharge K-09 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Test Connection VI3-53A RCIC Turbine Exhaust K-08 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Test Connection V13-817 RCIC Turbine Exhuast K-07 2 C Act i GSC SA C O n/a CD RO n/a Vacuum Breaker Stop PSO n/a Q  !

Check Vahr

• Revision 18 Drawing : G-191174 Sh 1 Sectioit 5 Page 61 of 160 i

Vcrmont Yankee Nuclear Power Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191174 Sh 1 Drawing

Title:

Reactor Core Isolation Cooling System Valve Number Nomenclature Dwg Safet OM Act / Site Norm Safety Fail Test Test CSJ/ROJ gg, ,,,g, Body Act

, Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V13418 RCIC Tuibine Exhuast K-07 2 C Act 1 CK SA C O/C n/a CD RO n/a Vacuum Breaker Check PSO Q n/a Valve Revision 18 Drawing : G-191174 Sh 1 Section 5 Page 62 of 160

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Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191174 Sh 2 Drawing

Title:

Reactor Core Isolation Cooling System Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remah Coor y Cat Pass (inch Pos Pos Pos Type Freq RR LCV-13-12 RCIC Baro Cond L-10 3 B Act i GA AO C O/C FC FST Q n/a Condensate Pump Disch PIT 2Y n/a Drain LCV STC Q n/a STO Q n/a LCV-13-13 RCIC Baro Cond L-10 3 B Act 1 GA AO C O/C FC FST Q n/a Condensate Pump Disch PIT 2Y n/a Drain LCV STC Q n/a STD Q n/a PCV-13-23 Cooling Water to 11-0 6 2 B Act 2 GL AO O O FO SKID Q n/a Note 12 Barometric Condenser RCIC- RCIC Turbine Governing D-12 2 B Act 2 GL 11 0 O O/C FAI PIT 2Y n/a Note 12 CONTROL Vahr SKID Q n/a 5-13-3 RCIC Turbine Steam E-11 2 D Act 8 RD SA C O/C n/a RD SY n/a Exhaust Rupture Disc S-13-4 RCIC Turbine Steam E-09 NNS D Act 8 RD SA C O/C n/a RD SY n/a Exhaust Rupture Disc SR-13-25 RCIC Pump Suction Relief C-07 2 C Act 1 RV SA C O/C n/a SP 10Y n/a Vaht SR-13-26 RCIC Pump Discharge 1-07 2 C Act 1 RV SA C O/C n/a SP 10Y n/a Line to Barometric Cond Relief Viv SR-13-27 RCIC Barometric J-14 3 C Act 1.5 RV SA C O/C n/a SP 10Y n/a Condenser Relief Vaht VI3-1 RCIC Turbine Trip & D-14 2 B Act 2 GL MO-DC O O/C FAI PIT 2Y n/a Note 12 Throttle Vaht SKID Q n/a V13-131 RCIC Steam Supply Vaht E-15 2 B Act 3 GL MO-DC C O/C FAI PIT 2Y n/a STC Q n/a S1D Q n/a Revision 18 Drawing : G-191174 Sh 2 Section 5 Page 63 of 160

I O w /2) w @w Vcrmont Yankee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191174 Sh 2 Drawing

Title:

Reactor Core Isolation Cooling S3 5cm Valve Number Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Nomenclature Body Act Remarks Coor y Cat Pass (inch) Pos Pos Pos Type Freq RR V13-132 RCIC Pump Discharge G-06 2 B Act 2 GL MO-DC C O/C FAI PIT 2Y n/a Valve to Barometric STC Q n/a Condenser STO Q n/a V13-133 RCIC Barometric Cond K-10 2 C Act 2 CK SA C O/C n/a SKID Q n/a Note 12 Condensate Pump Disch Check Valve  ;

V13-70 RCIC Baro Cond K-13 3 C Act 1.25 CK SA C O n/a SKID Q n/a Note 12 Condensate Pump Disch Check Valve Revision 18 Drawing : G-191174 Sh 2 Section 5 Page 64 of 160

O O O Vcrm:nt Ycckee N:clacr Pow r St: tion Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191175 Sh 1 Drawing Titic: Primary Containment & Atmosphere Control Valve Number Nomenclature Dwg Safet OM Act / Size Body Norm Safety Fail Test Test CSJ/ROJ

~ Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Fmq RR S16-19-1 Torus Vent Rupture Disc J-8 2 D Act 8 RD SA C O/C n/a RD SY n/a S 8-16-19-10 PCAC Supp Chamber J-13 2 A Act i8 BTF AO C C FC FST Q n/a Purge Supply Cont Isol u 2Y n/a Vaht PIT 2Y n/a STC Q n/a 50-16-19-11 A PCAC Vacuum Relief K-15 2 A/C Act 20 BTF AO C O/C FO FST Q n/a Note 15 From Secondary U 2Y n/a Containment Iso Vaht PIT 2Y n/a SP 6M n/a STC Q n/a STO Q n/a S~A16-19-11B PCAC Vacuum Relief M-15 2 A/C Act 20 BTF AO C O/C FO FST Q n/a Note 15 From Secondary U 2Y n/a Containment Iso Vaht PIT 2Y n/a SP 6M n/a

^

STC Q n/a STO Q n/a S 0-16-19-23 PCAC Containment Purge 11-1 5 2 A Act 6 BTF AO C C FC FST Q n/a  !

Supply Containment Iso U 2Y n/a Vaht PIT 2Y n/a STC Q n/a S0-16-19-6 PCAC to SBGT D-01 2 A Act 8 UTF AO O C FC FST Q n/a Vahr also appears on G-Containment Isolation U 2Y n/a 191238 Vaht PIT 2Y n/a STC Q n/a 50-16-19-6A PCAC Drywell Purge & F-02 2 A Act 3 BTF AO C C FC FST Q n/a Vent Outlet Bypass Cont U 2Y n/a Isol Vaht PIT 2Y n/a STC Q n/a Revision 18 Drawing : G-191175 Sh 1 Section 5 Page 65 of 160 t

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Vcrmont Yc kee Naclear Pow;r Station O O Inservica Testing Program Table 5-1 Valve Listing Drawing : G-191175 Sh 1 Drawing

Title:

Primary Containment & Atmosphere Control Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act %g, Coor y Cat Pass (inch Pos Pos Pos Type Freq RR SD-16-19-6B PCAC Supp Chb Prg & L-03 2 A Act 3 BTF AO O C r'C FST Q n/a Vent Line Bypass Inbd U 2Y n/a Cont Isol Viv PIT 2Y n/a STC Q n/a 50-16-19-7 PC/ C Purge & Vent D-01 2 A Act I8 BTF AO C C FC FST Q n/a Exhaust Containment U 2Y n/a Isolation Valve PIT 2Y n/a STC Q n/a S0-16-19-7A PCAC Drywril Purge & E-02 2 A Act 18 BTF AO C C FC FST Q n/a Vent Outlet Cont Isol Vaht U 2Y n/a PIT 2Y n/a STC Q n/a S3-16-19-7B PCAC Supp Chb Purge & K-03 2 A Act 18 BTF AO C C FC FST Q n/a Vent Outlet Inbd Cont Isol U 2Y n/a Vaht PIT 2Y n/a STC Q n/a S*s16-19-8 PCAC Drywell Air Purge I-12 2 A Act 18 BTF AO C C FC FST Q n/a Inlet Containment Isol U 2Y n/a Vaht PIT 2Y n/a STC Q n/a S 0-16-19-9 PCAC Air Purge Supply I-16 2 A Act 18 BTF AO C C FC FST Q n/a From Rx Bldg Cont Isol U 2Y n/a Vaht PIT 2Y n/a STC Q n/a VI6-19-12A PCAC Vacuum Relief K-16 2 A/C Act 20 CK SA C O/C n/a iJ 2Y n/a From Secondary SC Q n/a Containment Check Viv SO n/a Q

SP 6M n/a Revision 18 Drawing : G-191175 Sh 1 Section 5 Page 66 ol' 160

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\

m

(

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Vcrmont Ycrkee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191175 Sh I Drawing

Title:

Primary Containment & Atmosphere Control Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safd, ' Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V16-19-12B PCAC Vacuum Relief M-16 2 NC Act 20 CK SA C O/C n/a LJ 2Y n/a From Secondary SC Q n/a Containment Check Viv SO n/a Q

SP 6M n/a V16-19-13A Drywell Air Lock 11-06 2 A Pass 1.5 BL MAN LC C n/a LJ 2Y n/a Equalizing Vent

'16-19-138 Drywell Airlock Vent H-06 2 A Pass 1.5 BL MAN LC C n/a LJ 2Y n/a V16-19-19B PCAC Globe Valve M-16 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V16-19-20A PCAC Globe Vahr K-16 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V16-19-21 A PCAC Test Connection K-01 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V16-19-22B PCAC Test Connection 1-13 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a V16-19-56 Drywell Air 14ck 1-06 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Pressurization Vahr V16-19-5A Drywell To Suppression J-08 2 NC Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC n/a Q

Vacuum Breaker SO Q n/a SP 6M n/a V16-19-5B Drywell To Suppression J-08 2 NC Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC n/a Q

Vacuum Breaker SO n/a Q

SP 6M n/a V16-19-5C Drywell To Suppression J-08 2 NC Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Dmyncomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-19-5D Drywell To Suppression K-08 2 NC Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Dmyncomer SC Q n/a Vacuum Breaker SO n/a Q

SP 6M n/a Revision 18 Drawing : G-191175 Sh 1 Section 5 Page 67 of 160 ,

O Vcrmont Yc;kee Nuclear Pow:;r Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191175 Sh 1 Drawing

Title:

Primary Containment & Atmosphere Control Valve Number Nomenclature Dwg Safet OM Act / Size Body Norm Safety Fail Test Test CSJ/ROJ Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V16-19-5E DrywellTo Suppression J 08 2 A/C Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-19-5F Drywell To Suppression J-08 2 A/C Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-19-5G Drywell To Suppression J-08 2 A/C Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-19-5H Drywell To Suppression J-08 2 A/C Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-19-51 Drywell To Suppression K-08 2 A/C Act I8 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-19-5J Drywell To Suppression J-08 2 A/C Act 18 CK SA C O/C n/a LT 2Y n/a Chamber Downcomer SC Q n/a Vacuum Breaker SO Q n/a SP 6M n/a V16-20-20 PCAC Containment Purge 11-12 2 A Act i GA SO O C FC FST Q n/a Makeup Outbd Cont Isol LJ 2Y n/a Vah,c PIT 2Y n/a STC Q n/a Revision 18 Drawing : G-191175 Sh 1 Section 5 Page 68 of 160

O O O Vcrmont Yc kee Nuclear Pcwcr Station I: service Testing Program Table 5-1 Valve Listing Drawing: G-191175 Sh 1 Drawing

Title:

Primary Containment & Atmosphere Control Dwg Safet OM Act / Size Valve Number Nomenclature Body Act - Norm Safety Fail Test Test CSJ/RW Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V16-20-22A PCAC Containment Purge H-11 2 A Act 1 GL SO C C FC FST Q n/a Makeup Cont Isol Vahe U 2Y n/a PIT 2Y n/a STC Q n/a V16-20-228 PCAC Containment Purge H-12 2 A Act 1 GL SO O C FC FST Q n/a Makeup Cont Isol Vahe W 2Y n/a PIT 2Y n/a STC Q n/a V16-20-28 Torus Makeup Test I-l1 2 B Pass .75 GA MAN C C n/a n/a n/a n/a Connection V16-20-30 Drywell Makeup Test I-12 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Connection V20-400A Torus Drain Vahe N-09 2 B Pass 2 GL MAN LC C n/a n/a n/a n/a V20-4008 Torus Drain Vahe N-07 2 B Pass 2 GL MAN LC C n/a n/a n/a n/a Revision 18 Drawing : G-191175 Sh 1 Section 5 Page 69 of 160

O Vcrmont Yc1kee Nuclear Pcwcr St-tion O O I;servic2 Testing Program Table 5-1 Valve Listing Drawing : G-191177 Sh 1 Drawing

Title:

Radwaste Systems Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR RV-20-82A Pen. X-18 Thermal Relief A-03 2 C Act .75 RV SA C O/C n/a SP 10Y n/a Vaht RV-20-94A Pen. X-19 Thermal Relief F-03 2 C Act .75 RV SA C O/C n/a SP 10Y n/a Vaht V20-319A Rx Bldg Floor Drain Pump D-06 2 B Pass 1.5 GA MAN LC C n/a n/a n/a n/a To Suppression Poolim Vaht V20-319B Rx Bldg Floor Dmin Pump D-06 2 B Pass 1.5 GA MAN LC C n/a n/a n/a n/a To Suppression Pool Iso Vaht V20-319C Rx Bldg Floor Drain Pump C-08 2 B Pass 1.5 GA MAN LC C n/a n/a n/a n/a To Suppression PoolIso Vaht V20-319D Rx Bldg Floor Drain Pump C-08 2 B Pass 1.5 GA MAN LC C n/a n/a n/a n/a To Suppression Pool Iso Valve V20-78A Test Connection B-04 2 B Pass .75 GA MAN C C n/a n/a n/a n/a V20-79A Test Connection B-04 2 B Pass .75 GA MAN C C n/a n/a n/a n/a V20-82 Drywell Floor Drain Sump B-04 2 A Act 3 BL AO O C FC FST Q n/a Containment Isolation LJ 2Y n/a Vahe PIT 2Y n/a STC Q n/a V20-83 Drywell Floor Drain Sump B-05 2 A Act 3 BL AO O C FC FST Q n/a Containment Isolation LJ 2Y n/a Vaht PIT 2Y n/a STC Q n/a V20-92A Test Connection G-04 2 B Pass .75 GA MAN C C n/a n/a n/a n/a V20-93A Test Connection G-04 2 B Pass .75 GA MAN C C n/a n/a n/a n/a Revision 18 Drawing : G-191177 Sh 1 Section 5 Page 70 of 160

O Vcrmont Yc kee Nuclear Pow;r St tion O O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191177 Sh 1 Drawing

Title:

Radwaste S 35 ems Vale Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act hads Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V20-94 Drywell Equipment Drain G-04 2 A Act 3 BL AO O C FC FST Q n/a Sump Containment U 2Y n/a Isolation Viv PIT 2Y n/a STC Q n/a V20-95 Drywell Equipment Drain G-05 2 A Act 3 BL AO O C FC FST Q n/a i Sump Containment U 2Y n/a Isolation Viv PIT 2Y n/a STC Q n/a k

Revision 18 Drawing : G-191177 Sh 1 Section 5 Page 71 of 160

Vcrmont Yc:kee Nuclear Powcr St-tion O O O Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191178 Sh I Drawing

Title:

Reactor Water Clean Up System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remds Coor y Cat Pass (inch Pos Pos Pos Type Freq RR V12-15 Reactor Water Cleanup D-02 1 A Act 4 GA MO-AC O C FAI IJ 2Y n/a Containment Isolation PIT 2Y n/a Valve STC Q n/a V12-16 RWCU Drainfrest D-03 1 B Pass .75 GL MAN C C n/a n/a n/a n/a Connection V12-18 Reactor Water Cleanup D-03 1 A Act 4 GA MO-DC O C FAI IJ 2Y n/a Containment Isolation PIT 2Y n/a Valve STC Q n/a V12-28A RWCU Pump Discharge D418 3 C Act 3 CK SA O/C C n/a CD RO n/a Check Vaht V12-28B RWCU Pump Discharge G-08 3 C Act 3 CK SA O/C C n/a CD RO n/a Check Vahe V12-62 RWCU Flow to Feedwater B-05 2 C Act 4 CK SA O/C C n/a CD RO n/a System Isolation Check Vaht V12-62A RWCU Flow to Feedwat:r B-05 3 C Act 4 CK SA O/C C n/a CD RO n/a System Isolation Check Valve i

a Revision 18 Drawing : G-191178 Sh 1 Section 5 Page 72 of 160

O Vcrmont Ycckee Nuclear Pow:r Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191237 Sh 2 Drawing

Title:

IIVAC - Tmbine, Service & Control Room Building Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Ssfety Fail Test Test CSJ/ROJ  %,g Coor y Cat Pass (inch Pos Pos Pos Type Freg RR SACC-1A- Cont Room HVAC Chiller 1-03 3 C Act 1.375 CK SA O/C O n/a SKID Q n'a CHECK Refrigerant Piping Check Vahe SACC-1B- Cent Room IIVAC Chiller I-04 3 C Act 1.375 CK SA O/C O n/a SKID Q n/a CHECK Refrigerant Piping Cluck Vahr SCW-46A SAC-1 Temperature B-6 3 B Act 2 3-WAY AO O O FO ' FST Q n/a Control Vaht Heating and STO Q n/a Vent Room SCW45A HVAC Expansion Tank D-04 3 B Act .75 GL MAN C C n/a SC Q n/a Isolation Vahr -

SCW-8A Control Room HVAC E-05 3 C Act 3 CK SA O O n/a SO Q n/a Chilled Water Pump Disch Check Vaht '

S~aSACC-1A-1 Cont Room HVAC Chiller H-03 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Refrigerant Piping Relief Vaht SR-SACC-IA-2 Cont Room HVAC Chiller H 03 3 C Act .75 RV SA C O/C n/a SP 10Y n/a Refrigerant Piping Relief Vahr SR-SACC-1B-1 Cont Room HVAC Chiller H-04 3 C Act 1.25 RV SA C O/C n/a SP 10Y n/a I

Refrigerant Piping Relief Vaht i

t Revision 18 Drawing : G-191237 Sh 2 Section S Page 73 of 160  ;

q

O Vcrmont Yc kee N ciz.:r P .w;r St tion O O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191238 Drawing

Title:

IIVAC - Reactor Building Valve Nnmber NomencIMure Dwg Safet OM Act / Size Body Norm Safety Fail Test Test CSJJROJ Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq FR SB-1-125-1 A Standby Gas Treatment G418 2 B Act 12 BTF AO C O FO FSY Q n/a Train Inlet Vahr PIT 2Y n/a STO Q n/a SB-1-125-1B Standby Gas Treatment G-07 2 B Act 12 BTF AO C O FO FST Q n/a Train Inlet Vahe PIT 2Y n/a STO Q n/a SB-1-125-2A Standby Gas Treatment H-06 2 B Act 12 BTF AO O O/C FO FST Q n/a Train Inlet Vah'e PIT 2Y n/a STC Q n/a STO Q n/a SB-1-125-2B Standby Gas Treatment 11-06 2 B Act 12 BTF AO C O/C FO FST Q n/a Train Inlet Vahr PIT 2Y n/a STC Q n/a STO Q n/a SB-1-125-3A Standby Gas Treatment G-06 2 B Act 12 BTF AO O O FO FST n/a Q

Train Discharge Vahc PIT 2Y n/a STO Q n/a SB-1-125-3 B Standby Gas Treatment G-06 2 B Act 12 BTF AO C O FO FST Q n/a Train Discharge Vahr PIT 2Y n/a STO Q n/a SB-1-125-4A Standby Gas Treatment H-07 2 B Act 12 BTF AO C O/C FC FST n/a Q

Train Dilution Vahr PIT 2Y n/a STC Q n/a STO Q n/a SB-1-125-4B Standby Gas Treatment 11-0 6 2 B Act 12 BTF AO C O/C FC FST Q n/a Train Dilution Vahr PIT 2Y n/a STC Q n/a STO Q n/a Revision 18 Drawing : G-191238 Section 5 Page 74 of 160

O Vcrmont Yc kee Nuclear Power Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing : G-191238 Drawing

Title:

IIVAC - Reactor Building Valve Number Nomenclature Dwg Safet OM Act/ Size Norm Safety Fail Test Test CSJ/ROJ Boely Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Frrq RR SB-1-125-5 Standby Gas Treatment G-06 2 B Act 4 BTF AO C O/C FC FST Q n/a Train Xconn Vahr Pirr 2Y n/a STC Q n/a STO Q n/a SB-10 RX Building Supply 11-0 8 2 B Act 54 BTF AO O C FC FST Q n/a isolation Vaht PIT 2Y n/a STC Q n/a SB-11 Rx Bldg Exhaust to Stack G-08 2 B Act 54 BTF AO O C 7C FST Q n/a Sec Cond Isolation Valve FIT 2Y n/a STC Q n/a SB-12 Rx Bldg Exhaust to Stack G-09 2 B Act 54 BTF AO O C FC FST Q n/a Sec Cond Isolation Valve PIT 2Y n/a STC Q n/a SB-16-19-6

• PCAC to SBGT F-06 * * * * * * * * * * * *

• See G-191175 for vaht Containment Isolation info and test requirements Vahe SB-9 RX Building Supply 11-0 9 2 B Act 54 BTF AO O C FC IST Q n/a Isolation Vahe PIT 2Y n/a STC Q n/a SGT-7A Standby Gas Treatment F-07 2 C Act 12 CK SA C O/C n/a SC Q n/a Train Discharge Check SO n/a Q

Vahe SGT-7B Standby Gas Treatment F-06 2 C Act 12 Ci' SA C O/C n/a SC Q n/a Train Discharge Check SO n/a Q

Vaht Revision 18 Drawing : G-191238 Section 5 Page 75 of 16d

O Vcrmont Yc kee Nuclear Powar Station O Icservic2 Testing Program O

Table 5-1 Valve Listing Drawing : G-191267 Sh 1 Drawing

Title:

Nuclear Boiler Vessel Instrumentation Valve Number Nomenclature Dwg Safet OM Act / Sire Body Act Norm Safety Fail Test Test CSJ/ROJ harks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR Sle2-3-11 Nuclear Boiler Vessel Inst C4 2 NC Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahe Ste2-3-13A Nuclear Boiler VesselInst DM 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe Str2-3-13B Nuclear Boiler Vessel Inst C-12 2 NC Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr sir 2-3-15A Nuclear Boiler Vessel Inst D-06 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahe Str2-3-15B Nuclear Beiler Vessel Inst D-12 2 NC Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe sir 2-3-17A Nuclear Boiler Vessel Inst E-06 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe Sle2-3-17B Nuclear Boiler Vessel Inst E-12 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve Str2-3-19A Nuclear Boiler Vessel Inst F-06 2 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr sir 2-3-19B Nuclear Boiler Vessel Inst F-12 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr Sle2-3-2 t A Nuclear Boiler Vessel Inst 11-06 1 A/C Act i EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahr Str2-3-21B

• Nuclear Boiler Vessel Inst 11-0 6 * * * * * * * * *

• See G-191165 for vahr Excess Flow Check Vahr info and test requirements sir 2-3-21C Nuclear Boiler VesselInst 11-12 1 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe Str2-3-21D

• Nuclear Boiler Vessel Inst 11-1 2 * * * * * * * * *

• See G-191165 for vahe Excess Flow Check Vahr info and test requiren -.::3 Str2-3-23A Nuclear Boiler Vessel Inst G-06 i A/C Act i EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahr Str2-3-23B Nuclear Boiler VesselInst G-06 i A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr Revision 18 Drawing : G-191267 Sh 1 Section 5 Page 76 of 160

O Vcrmont Yc kee Nuclear Pow;r St: tion O O E. service Testing Program Tabic 5-1 Valve Listing Drawing : G-191267 Sh 1 Drawing (itle: Nuclear Boiler Vessel Instrumentation Valve Number Nomenclature Dwg Safet OM Act / Size Nom Safety Fail Test Test CSJ/ROJ Body Act hds Coor y Cat Pass (inch Pos Pos Pos Type Freq RR Sle2-3-23C Nuclear Boiler VesselInst G-12 1 NC Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe sir 2-3-23D Nuclear Boiler Vessel Inst G-12 1 NC Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe sir 2-3-25 Nuclear Boiler Vessel Inst 1-06 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahr i sir 2-3-27 Nuclear Boiler Vessel Inst I-06 2 NC Act 1 EFC SA O O/C n/a LEF RO ROJ-VOI Excess Flow Check Vahe Str2-3-3 t A Nuclear Boiler Vessel Inst F-06 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr sir 2-3-31B Nuclear Boiler VesselInst F-06 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe S1,2-3-31 C Nuclear Boiler VesselInst F-06 i A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Valve sir 2-3-31D Nuclear Boiler Vessel Inst F-06 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr sir 2-3-31E Nuclear Boiler Vessel Inst F-06 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe S1,2-3-31 F Nuclear Boiler Vessel Ins. F-06 i NC Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahr sir 2-3-31G Nuclear Boiler VesselInst F-06 1 NC Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr Sir 2-3-3111 Nuclear Boiler Vessel Inst F-06 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr sir 2-3-311 Nuclear Boiler VesselInst F-12 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vche Sir 2-3-31J Nuclear Boiler Vessel Inst F-12 1 NC Act i EFC SA O O/C n/a LEF RO ROJ-Vol '

Excess Flow Check Vaht Ste2-3-31K Nuclear Boiler Vessel Inst F-12 1 A/C Act i EFC SA O O/C n/a LEF RO ROJ.V01 Excess Flow Check Vaht Revision 18 Drawing : G-191267 Sh 1 Section 5 Page 77 of' 160

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\)

Vcrmont Yc:kee Nuclear Pow;r Station

% ./ 0 Inservice Testing Pawgram Table 5-1 Valve Listing Drawing : G-191267 Sh I Drawing

Title:

Nuclear Boiler Vessel Instrumentation Valve Number NomencInture Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR SL-2-3-31 L Nuclear Boiler VesselInst F-12 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr 5L-2-3-31M Nuclear Boiler VesselInst F-12 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr SL-2-3-31N Nuclear Boiler VesselInst F-12 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-Vol Excess Flow Check Vahr 50-2-3-31P Nuclear Boiler Vessel Inst F-12 1 A/C Act i EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vaht SL-2-3-31Q Nuclear Boiler VesselInst F-12 1 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahe 5 0-2-3-33 Nuclear Boiler Vessel Inst I-12 2 A/C Act 1 EFC SA O O/C n/a LEF RO ROJ-V01 Excess Flow Check Vahr SL-2-3-35 Nuclear Boiler VesselInst J-06 2 A/C Act 1 EFC SA O C n/a LEF RO ROJ-V01 Excess Flow Check Vaht V2-3-28A NBVI Globe Vaht F-06 2 B Pass 1.00 GL MAN C C n/a n/a n/a n/a V2-3-28B NBVI Globe Vaht F-11 2 D Pass 1.00 GL MAN C C n/a n/a n/a n/a Revision 18 Drawing : G-191267 Sh 1 Section 5 Page 78 of 160

O O O Vcrmont Yc kee Nuclear Pow r St; tion Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191267 Sh 2 Drawing

Title:

Nuclear Boiler Vessel Instrumentation Valve Number Nomenclature Dwg Safet OM Act / Sire Body Act Norm Safety Fail Test Test CSJ/ROJ harks Coor y Cat Pass (inch Pos Pos Pos Type Freg RR V2-3-430A Reference Leg Back Fill C-09 2 A/C Act .375 CK SA C C n/a U 2Y n/a Inlet Check Vahr SC RO ROJ-V15 V2-3-430B Reference Leg Back Fill E4)9 2 A/C Act .375 CK SA C C n/a U 2Y n/a Inlet Check Vahe SC RO ROJ-VIS ,

V2-3-432A Reference Leg Back Fill C-10 2 A/C Act .375 CK SA C C n/a U 2Y n/a Check Valve SC RO ROJ-V15 V2-3-432B Reference Leg Back Fill E-10 2 A/C Act .375 CK SA C C n/a U 2Y n/a Check Vahe SC RO ROJ-V15 V2-3-433A Reference Leg Back Fill F-09 2 A/C Act .375 CK SA C C n/a U 2Y n/a Inlet Check Vaht SC RO ROJ-VIS V2-3-4338 Reference Leg Back Fill H-09 2 A/C Act .375 CK SA C C n/a U 2Y n/a Inlet Check Valve SC RO ROJ-VIS i V2-3-435A Referenceleg Back Fill F-10 2 A/C Act .375 CK SA C C n/a U 2Y n/a  !

Inlet Check Vaht SC RO ROJ-VIS V2-3-435B Reference Leg Back Fill 11-1 0 2 A/C Act .375 CK SA C C n/a U 2Y n/a inlet Check Vaht SC RO ROJ-VI5 V2-3-444A Penetration X-28A Test C-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection V2-3-444 B Penetration X-28D Test E-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection V2-3-447A Penetration X-29A Test F-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection V2-3-447B Penetration X-29D Test II-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection V2-3-448A Penetration X-28A Test C-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection V2-3-448B Penetration X-28D Test E-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a '

Connection Revision 18 Drawing : G-191267 Sh 2 Section 5 Page 79 of 160 i

b N [vl v)

Vcrmont Ycckee Nuclear Power Station Inservice Testing Program Table 5-1 Valve Listing Drawing: G-191267 Sh 2 Drawing

Title:

Nuclear Boiler Vessel Instrunwntation Nomenclaturt Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Valve Number Body Act hads Coor y Cat Pass (inch Pos Pos Pos Type Freq - RR V2-3-451A Penetration X-29A Test F-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection V2-3-451B Penetration X-29D Test 11-10 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Connection Revision 18 Drawing : G-191267 Sh 2 Section 5 Page 80 of 160

O O O Vcrmont Ycnkee Nuclear Powcr Station Inservice Testing Program Table 5-1 Valve Listing Drawing : VY-E-75-002 Drawing

Title:

Containment Atmosphere Dilution System Valve Number Nomenclature Dwg Safet OM Act / Size Norm Safety Fail Test Test CSJ/ROJ Body Act Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR FSO-199-75A-1 CAD Sample Torus J-13 2 B Act I GA SO C O/C FC FST Q n/a Note 5. Valve also appears Inboard Isolation Valve PIT 2Y n/a on G-191165 STC Q n/a STO Q n/a FSO-199-75A-2 CAD Sample Torus J-14 2 B Act 1 GA SO C O/C FC FST Q n/a Note 5. Vahe also appears Outboard Isolation Vahr PIT 2Y n/a on G-191165 STC Q n/a STO Q n/a FSO-109-75A-3 CAD Supply Torus Inboard J-14 2 B Act 1 GL SO C O/C FC FST Q n/a Note 5. Vahe also appears Isolation Vaht PIT 2Y n/a en G-191165 STC Q n/a STO Q n/a FSO-109-75A-4 CAD Supply Torus J-14 2 B Act I GL SO C O/C FC FST Q n/a Note 5. Valve also appears Outboard isolation Vahe PIT 2Y n/a on G-191165 STC Q n/a STO Q n/a FSO-109-75B-1 CAD Sample Lower I-12 2 B Act 1 GA SO O O/C FC FST Q n/a Note 5. Vahr also appears Drywell Inboard Isolation PIT 2Y n/a on G-191165 Vahe STC n/a Q

STO Q n/a FSO-109-75B-2 CAD Sample Lower I-12 2 B Act 1 GA SO O O/C FC FST Q n/a Note 5. Vahr also appears Drywell Outboard Isolation PT 2Y n/a on G-191165 Vaht STC Q n/a STO Q n/a FSO-109-75C-1 CAD Sample Mid Drywell G-12 2 B Act 1 GA SO O O/C FC FST Q n/a Note 5. Vahe also appears Inboard Isolation Vaht PIT 2Y n/a on G-191165 STC Q n/a STO Q n/a Revision 18 Drawing : VY-E-75-002 Section 5 Page 81 of 160

_ . _ . _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ . _ . . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _- __----_1 _ . - _ _ _ __.____________.____.______.____m__ _ _ _ _

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Vcrm;;t Yc kee N: clear Paw 2r Station I service Testing Program Table 5-1 Valve Listing Drawing : VY-E-75-002 Drawing

Title:

Containment Atmosphere Dilution System Vaht Number Nomenclature Dwg Safet OM Act / Si'" Norm Safety Fail Test Test CSJ/ROJ Body Act Remah Coor y Cat Pass (3:,ch Pos Pos Pos Type Freg RR FSO-109-75C-2 CAD Sample Mid Drywel! G-12 2 B Act 1 GA SO O O/C FC FST Q n/a Note 5. Vaht also appears Outboard Isolation Vaht PIT 2Y n/a on G-191165 STC Q n/a STO Q n/a FSO-109-75D-1 CAD Sample Upper G-12 2 B Act i GA SO O O/C FC FST Q n/a Note 5. Vahr also appears DrywellInboard Isolation PIT 2Y n/a on G-191165 Valve . STC Q n/a STO Q n/a FSO-109-75D-2 CAD Sample Upper G-12 2 B Act I GA SO O O/C FC FST Q n/a Note 5. Vahr also appears Drywell Outboard Isolation PIT 2Y n/a on G-191165 Vahr STC Q n/a STO Q n/a FSO-109 76A CAD CAM Return L-13 2 A Act i GA SO O O/C FC FST Q n/a Vahe also appears on G-Containment Isolation U 2Y n/a 191165 Vaht PIT 2Y n/a STC Q n/a STO Q n/a FSO-109-76B CAD CAM Return L-13 2 A Act i GA SO O O/C FC FST Q n/a Vahr also appears on G-Containment Isolation U 2Y n/a 191165 Vaht PIT 2Y n/a STC Q n/a STO Q n/a NG-il A CAD N2 Supply 11-0 9 2 A Act 1 GL SO C O/C FC FST Q n/a Containment isolation U 2Y n/a Vaht PIT 2Y n/a STC Q n/a STO Q n/a

~

Revision 1. Drawing : VY-E-75-002 Section 5 Page 82 of 160

Vcrmont Yckee N:clar Powar Station Inservica Testirg Program Table 5-1 Valve Listing Drawing: VY-E-75-002 Drawing

Title:

Containment Atmosphere Dilution System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ hads Coor y Cat Pass (inch Pos Pos Pos Type Freq RR NG-11B CAD N2 Supply 11-08 2 A Ac* I GL SO C O/C FC FST Q n/a Containment isolation LJ 2Y n/a Valve PIT 2Y n/a STC Q n/a STO Q n/a NG-12A CAD N2 Supply J-09 2 A Act 1 GL SO C O/C FC FST Q n/a Containment Isolation LJ 2Y n/a Vaht PIT 2Y n/a STC Q n/a STO Q n/a NG-12B CAD N2 Supply J-08 2 A Act 1 GL SO C O/C FC FST Q n/a Containment Isolation LJ 2Y n/a Vaht PIT 2Y n/a STC Q n/a STO Q n/a NG-13A CAD N2 Supply I-09 2 A Act 1 GL SO C O/C FC FST Q n/a Containment Isolation LJ 2Y n/a Vahe PIT 2Y n/a STC Q n/a STO Q n/a NG-13B CAD N2 Supply H-09 2 A Act 1 GL SO C O/C FC FST Q n/a Containment Isolaticn LJ 2Y n/a Vaht PIT 2Y n/a STC Q n/a STO Q n/a NG-19 Manual Valve 11-10 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a NG-20 Manual Valve J-08 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a PSV-NG-34A Pressure Relief Vahe B-03 2 C Act .5 RV SA C O/C n/a SP 10Y n/a PSV-NG-34B Pressure Relief Vaht G-03 2 C Act .5 RV SA C O/C n/a SP 10Y n/a Revision 18 Drawing : VY-E-75-002 Section 5 Page 83 of 160 1

_ - . - - _ . - _ _ - . - - - _ - - . - - - _ _ - _ - . _ - - - - _ _ - . _ _ - . - - - _ _ _ - - _ _ < , _ - _ - _ _ - . _ _ - _ _ _ . - _ _ . _ .- _ - _ _ _ _ - - - _ - - - _ . _ _ _ ~ - . _ - _ _ _ - - _ - _ _ . _ - - - - _ _ _ _ . - _ . - - _ . _ - . _ _ _ _ . . _ - _ - _ _ - . _

V O O O Inservice Testing Program

_crmont Ycckee N clear Powcr St tion Table 5-1 Valve Listing Drawing : VY-E-75-002 Drawing

Title:

Containment Atmosphere Dilution System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ '"' '

Coor y Cat Pass (inch Pos Pos Pos Type Fatg RR SS-78A Isolation Vaht L-14 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a VG-12A VG-8A Test Conncction H-12 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a VG-12B VG-8B Test Connection 1-14 2 B Pass 1 GL MAA C C n/a n/a n/a n/a VG-13 Torus Root Isolation J-13 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a Sepply to 112/02 Monitor VG-17A Test Connection C-14 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a VG-17B Torus CAD Vent Line Test F-14 2 B Pass 1 GL MAN C C n/a n/a n/a n/a Connection VG-19 Torns/ Cam /il2/02 J-14 2 B Pass I GL MAN LC C n/a n/a n/a n/a Analyzer Test Connection VG-21 CAM sa:nple Line From G-18 2 B Pass 1 GL MAN C C n/a n/a n/a n/a Drywell/ Toms Atmos VG-22A CAD To SBGT System B-16 2 A Act 1 GL MO-AC C O/C FAI LJ 2Y n/a Containment Isolation PIT 2Y n/a Vahe STC Q n/a STO Q n/a VG-22B CAD To SBGT System E-16 2 A Act 1 GL MO-AC C O/C FAI LJ 2Y n/a Containment Isolation PIT 2Y n/a Vaht STC Q n/a STO Q n/a VG-23 CAD Rad Monitor Supply J-19 2 A Act 1 GL SO O O/C FC FST Q n/a Vahe also appears on G-Containment Isolation LJ 2Y n/a 191165-Vahe PIT 2Y n/a STC Q n/a STO Q n/a VG-24 CAD 11202 Analyzer L-11 2 B Act 1 GL SO O O/C FC FST Q n/a Note 5 Return Isolation Vaht PIT 2Y n/a STC Q n/a STO Q n/a Revision 18 Drawing : VY-E-75-002 Section 5 Page 84 of 160

O Vcrmont Yc~kee Nacie r Pow 2r Station

@ O Inservice Testing Program t Table 5-1 Valve Listing Drawing: VY-E-75-002 Drawing

Title:

Containment Atmosphere Dilution Sysem Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Coor y hads Cat Pass (inch Pos Pos Pos Type Freq RR VG-25 CAD H2O2 Analyzer L-11 2 B Act 1 GL SO O O/C FC FST Q n/a Note 5 Return Isolation Valve PIT 2Y n/a STC Q n/a STO Q n/a VG-26 CAD Rad Monitor Supply J-19 2 A Act 1 GL SO O O/C FC FST Q n/a Vahe also appears on G-Containment Isolation LJ 2Y n/a 191165 Valve PIT 2Y n/a STC Q n/a STO Q n/a ,

VG-32 Oneck Vahe M-15 2 C Pass 75 CK SA C C n/a n/a n/a n/a VG-33 CAD H2O2 Analyzer L-11 2 B Act 1 GL SO O O/C FC FST Q n/a Note 5 Return Isolation Vahe PIT 2Y n/a STC Q n/a STO Q n/a VG-34 CAD H2O2 Analyzer L-11 2 B Act 1 GL SO O O/C FC FST Q n/a Note 5 Return Isolation Vahe PIT 2Y n/a STC Q n/a STO Q n/a VG-35 VG-18 Test Connection K 12 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a VG-37 Sample Station Inlet K-17 2 B Pass .375 GL MAN C C n/a n/a n/a n/a Isolation. ,

i VG-38 Sample Station Outlet L-17 2 B Pass .375 GL MAN C C n/a n/a n/a n/a isolation i VG-39 Outlet Isolation L-18 2 B Pass .375 GL MAN C C n/a n/a n/a n/a VG-41 H2/02 Monitor Return to L-16 2 B Pass 1 GL MAN LC C n/a n/a n/a n/a RB Ventilation System VG-42 H2/02 Monitor Return to K-16 NNS B Pass 1 GL MAN LC C n/a n/a n/a n/a RB Ventilation System VG-7s Test Connection G-17 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a Revision 18 Drawing : VY-E-75-002 Section 5 Page 85 of' 160

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O Vcrmont Ycrkee Nuclear Pow;r Station O O Inservice Testing Program Table 5-1 Valve Listing Drawing : VY-E-75-002 Drawing

Title:

Containment Atmosphere Dilution System Valve Number Nomenclature Dwg Safet OM Act / Size Body Act Norm Safety Fail Test Test CSJ/ROJ Remarks Coor y Cat Pass (inch Pos Pos Pos Type Freq RR VG-79 112/02 Mon. Return Line H-15 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a to Torus /RB Vent. Sys.

Test Conn VG-80 K-17 2 B Pass .75 GL MAN LC C n/a n/a n/a n/a VG-9A CAD To SBGT System 11-1 2 2 A Act 1 GL SO C O/C FC FST Q n/a Containment Isolation U 2Y n/a Vaht PIT 2Y n/a STC Q n/a STO Q n/a VG-9B CAD To SBGT System 1-14 2 A Act 1 GL SO C O/C FC FST Q n/a Containment Isolation U 2Y n/a Vaht PIT 2Y n/a STC Q n/a STO Q n/a Revision 18 Dra wing : VY-E-75-002 Section 5 Page 86 of 160

Vcrnent Ycnkee Nuclear P;wer Stati:n Inservice Testing Program n 5.3 Valve Notes

(") 1.

Full stroke exercising and timing tests of the MSIVs are accomplished quarterly when i reactor power is decreased to less than 75 percent. At least twice a week, the MSIVs are also exercised by partial closure and subsequent reopening.

2. The RCIC and HPCI turbine exhaust check valves, V23-65 and V13-50, will also be  !

disassembled and inspected at least once every 10 years. (LER 87-18 and NRC l Inspection Repon 50-271/87 21) This additional testing is not required by 10 CFR 50.55a. ,

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3. Testing of CRD Cooling Water Supply Check valve, V3-138, is not required when the l corresponding control rod has been declared inoperable and its directional control valves I disarmed in accordance with the provisions of Technical Specifications 3.3.A.2.

4. At least once during each operating cvcle, the Standby Liquid Control system shall be i verified operable by: a) Initiating one of the SLC loops, excluding the primer chamber and inlet fitting, and verifying a flow path from the pump to the reactor vessel. Both loops shall be tested over the course of two operating cycles. (T.S. 4.4.A.2) b) Testing ofboth trigger assemblies removed from the system (each refueling outage) by installing in the test block to verify operability, c) Testing of the replacement trigger assemblies by installing one of the assemblies in the test block and firing it using the installed O circuitry. Replacement triggers shall be from the same batch as the test assembly. The U unfired replacement triggers, taken from the same batch shall be installed into the explosive valves. (T.S. 4.4.A.3 and 4.6.E)  ;

5. Only one valve in each CAD sample line or H202 analyzer line is required to be operable in accordance with Technical Specifications Table 4.7.2.b.

6. Positive verification of full opening of the Transversing In-Core (TIP) Ball Valves, BV-7-1 through 3 is shown by successful insertion of the TIP Probe through the valve.

l Positive verification of full closing of the valve is shown during each Refueling Outage l by successful leak testing (U). This testing is in compliance with Paragraph 4.1 of Pan ,

10 of the Code which states "Where local observation is not possible, other indications l shall be used for verification." Local observation of valve position would require  !

disassembly of the valve enclosures as the limit switches are contained within the i enclosures. I

7. The Nuclear Boiler Safety / Relief Valves, SV-2-70A & B and RV-2-71 A through D will ,

be tested in accordance with the requirements of OM-1 only.

8. These valves form the Alternate Cooling System boundary alignment and are othenvise included in the IST Program per based on a commitment to NRC.

Reference:

OP-2181; NRC Inspection Report 94-03, dated 3/4/94.

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Revision 18 Section 5 Page 87 of 160

I Vcrmrt Ycnkee Nucl:ar Pcwer Statia Inservice Testing Progrcm

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p 5.3 Valve Notes (Con't)

V 9. The Cooling Tower SW Distribution Tray valves, V70-17A through V70-17D shall be cycled once each refueling as part of augmented testing to the 1994 Service Water  ;

System Self Assessment Report. These valves are not required to be exercised m  !

accordance with the requirements of 10 CFR 50.55a. l l

10. The Service Water Pump Discharge Check Valves, V70-1A through ID, are part- l I

stroke exercised open on a quarterly basis during plant operation during the regularly scheduled quarterly service water pumps tests. These valves are full stroke exercised open during refueling outages.

I1. The RHR and Core Spray Keep-Fill Check Valves (V14-33A, V14-33B, V10-36A and V10-36B) are verified closed on a quarterly basis utihzing a non-intmsive technique.

The Core Spray Discharge Flushing Line Check Valves (V14-22A, V14-22B, V14-23 A and V14-23B) are verified closed on a quarterly basis utilizing a non-intrusive technique.

The HPCI and RCIC Keep-Fill Check Valves (V23-20B and V13-20B) are verified closed on a quarterly basis utilizing a non-intmsive technique.

The HPCI and RCIC Pump Suction Valves (V23-32 and V13.-19 are verified closed on a quarterly basis utilizing a non-intrusive technique.

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12. These valves are skid-mounted. NUREG 1482, Section 3.4 states that the testing of the major component is an acceptable means for verifying the operational readiness of the skid-mounted component and component subassemblies if the licensee documents this approach in the IST program. The scope of skid-mounted components additionally includes components that are not mounted on the skid, but which function much the same as skid mounted components.

13. These valves are post-accident sampling (PASS) valves. NUREG 1482, Section 4.4.2 states that PASS valves that perform a contamment isolation function are required to be includedin theist Program.

14. These valves are Category B passive valves. NUREG 1482, Section 4.2.6 states that the code does not restrict the verification of remote position indication to only active valves.

OM-10, Table 1, indicates that the licensee must also verify the position for Category B passive valves.

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Revision 18 Section 5 Page 88 of 160

Vcrm::t Yc:kee Ntclear Pcwer Stati:n InserviceTesting Program 5.3 Valve Notes (Con't) 15 The pdau-nce of pressure sensing instrumentation testing for PCAC Vacuum Relief from k+ Sry Containment Iso Valves, SB-16-19-IIA and 11B, is performed (OP 4376) as required by OM-1 subsection 3.3.2.3 and Technical Specification 4.7.A.5.a.

16. The Control Rod Drive Scram Inlet and Outlet Valves, CV-3-126 and CV-3-127, will be tested in accordance with the requirements of Tech. Spec Surveillance 4.3.C.1 after refueling outages and prior to operation above 30% power.

17. The CRD Manual Control Insertion, Withdrawal and Exhaust Valves, SO-3-120,121, 122 and 123 are veri 6ed operable in accordance with the weekly rod testing performed in accordance Tech. Spec. Surveillance 4.3.A.2 during power operations. The closure operation of these valves and the control rod dnves are verified after refueling outages and prior to operation above 30% power per Tech. Spec. Surveillance 4.3.C. I.

18. The Scram Exhaust To Discharge Volume Check Valves, V3-114, will be tested in accordance with the requirements of Tech. Spec Surveillance 4.3.C.1 after reflieling outages and prior to operation above 30% power.

19. The CRD Charging Water to Accumulator Check Valves, V3-115, closure verification will be performed by the accumulator pressure decay test (OP 4111) during refueling outages.

20. The Rx Recirc Sample Line Flow Control / Isolation valves, FCV-2-39 and FCV-2-40, position indication testmg is planned to commence after a design change to the valve position indication circuitry in 1998. (Commitment BMO9507Rl_02)

21. These valves are PIV's, they are subject to leak testing commensurate with nominal operating reactor vessel pressure.

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Revision 13 Section 5 Page 89 of 160 i

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Verm:1t Ycnkee Nuclear Pcwcr Stati:n Ins:rvice Testing Progran COLD SHUTDOWN JUSTIFICATION Number: CSJ-V01, Revision 2 (Sheet 1 of1)

SYSTEM: Reactor Building Closed Cooling Water COMPONENTS: l Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord. i V70-117 A 2 G-191159 Sh 3 L-12 I Valve V70-117 is the reactor building closed cooling water return containment isolation valve. This valve has a safety function in the closed position to provide primary containment isolation. .

JUSTIFICATION:

This valve cannot be full-stroke exercised closed during normal (power) operation since shutting this valve I would stop cooling water flow to vital primary containment equipment, including the primary containment i air coolers and the reactor recirculation pumps.

This valve will be stroke timed closed during Cold Shutdowns in accordance with Paragraphs 4.2.1.2(c), '

(f), (g) and (h) of the code.

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5 vision 18 Section 5 Page 90 of 160

. . . _ - . = ..- . .

> 1 Verm::nt Ycnkee Nuclear Power Stati:n Inservice Testing Pr: gram COLD SHUTDOWN JUSTTFICATION l I

Number: CSJ-V02, Revision 1 (Sheet 1 of1)  ;

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THIS COLD SHUTDOWN JUSTIFICATION WAS DELETED IN REVISION 18 OF THE IST PROGRAM.

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This CSJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's. I l

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O Revision 18 Section 5 Page 91 of 160

Verm:nt Ycnkee Nrcirr Pcwer Station Inservice Testing Program COLD SHUTDOWN JUSTIFICATION Number: CSJ-V03, Revision 1 (Sheet 1 of1) 4 d

THIS COLD SHUTDOWN JUSTIFICATION WAS DELETED IN REVISION 15 OF THE IST PROGRAM.

This CSJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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V Revision 18 Section 5 Page 92 of 160

Vermrt Ycnkee Nuclear Pow:.r Stati:n Inservice Testing Program  ;

4 COLD SHUTDOWN JUSTIFICATION Number: CSJ-V04, Revision 2 (Sheet 1 of1) i i

SYSTEM: NuclearBoiler COMPONENTS:  ;

i Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.  ;

V2-53A B 1 G-191167 L-09 i V2-53B B 1 G-191167 L-09 l V2-54A B 1 G-191167 L-09 i V2-54B B 1 G-191167 L-09 l l

l These valves are the reactor recirculation pump discharge isolation and bypass valves. They have a safety ,

function in the closed position to limit primary system coolant loss following a LOCA and to ensure low l pressure coolant injection flow is properly directed to the reactor.

JUSTIFICATION

These valves cannot be exercised closed during reactor power operation since cycling these valves would  !

,_ result in a reactor recirculation pump trip. I V These valves will be stroke timed closed dudng Cold Shutdowns in accordance with Paragraphs 4.2.1.2(c),

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(f),(g) and (h) ofPart 10 ofthe Code. l i

Revision 18 Section 5 Page 93 of 160

Verm:nt Yankee Nuclear P;wer Stati a Inservice Testing Program COLD SHUTDOWN JUSTIFICATION O

Number: CSJ-V05, Revision 2 (Sheet 1 of1)

SYSTEM: Core Spray I

COMPONENTS: '

i ,

Valve Number OM Cat. Safety Class Drawing Number

' Dwg. Coord.

V14-13A A/C 1 G-191168 F-06 j V14-13B A/C i

1 G-191168 D-06 4

These valves are the Core Spray injection check valves. They have a safety function in the open pos pass core spray injection flow to the reactor, and in the closed position for primary containment and pressureisolation.

JUSTIFICATION:

i 4 j These valves cannot be exercised open during normal (power) operation since core spray pump d

' cannot overcome reactor pressure. Manual exercising is precluded dudng power operation since the valves are inside the inerted drywell.

/ These valves will be manually exercised open during Cold Shutdowns when the drywell is accessible in accordance with 4.3.2.2(c), (f), (g) and (h) ofPan 10 of the Code.

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f Revision 18 Section 5 Page 94 of 160

Verment Ycnkte Nucl ar Pcwer St:ti:n Inservice Testing Pragram

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l COLD SHUTDOWN JUSTIFICATION Number: CSJ-V06, Revision 1 (Sheet 1 of1)

SYSTEM: High Pressure Coolant Injection ,

I COMPONENTS: l

)

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

l V23-18 C 2 G-191169 Sh 1 G-05 This valve is Ge High Pressure Coolant Injection (HPCI) discharge to Feedwater Line "A" check valve.

The valve has a safety function in the open position to pass HPCI flow to the reactor and in the closed position to prevent backflow feedwater into HPCI piping. j JUSTIFICATION:

This valve cannot be exercised open dudng normal (power) operation since flow through this valve must i be inje:ted into the reactor coolant system. This would thermally shock the reactor nozzles and cause a l positive reactivity excursion. Manual operation of the valve is not possible since the valve is located in the l steam tunnel which is inaccessible during power operations. J This valve will be manually exercised open and closed during Cold Shutdowns when the steam tunnel is accessible in accordance with 4.3.2.2(c), (f), (g) and (h) ofPart 10 of the Code. i Revision 18 Section 5 Page 95 of 160

Vcrm:nt Yenkee Nucirr Pcwzr Stati:n Ins:rvice Testing F mgram

- COLD SHUTDOWN JUSTIFICATION Number: CSJ-V07, Revision 0 (Sheet 1 of1)

THIS COLD SHUTDOWN JUSTIFICATION WAS DELETED IN REVISION 18 OF THE IST PROGRAM.

This CSJ numbrs is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Revision 18 Section 5 Page 96 of 160

Verm:nt Ycnkee Nucl:ar Pcw:r Stati:n Inservic2 Testing Prcgram  !

q COLD SHUTDOWN JUSTWICATION V Number: CSJ-V08, Revision 2 (Sheet 1 of1)

SYSTEM: Residual Heat Removal 1

i COMPONENTS:

l Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V10-17 A 2 G-191172 G-08 V10-18 A 2 G-191172 F-08 These valves are the Residual Heat Removal (RHR) shutdown cooling supply isolation valves. They have a safety function in the closed position to provide primary contairunent and pressure isolation, and in the l open position to provide RHR pump suction during shutdown cooling operation.

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JUSTIFICATION:

These valves cannot be stroke timed open during reactor power operation since there is a 100 psig interlock that prevents opening these valves during power operation. This interlock is required to prevent overpressurization of the lower pressure rated RHR shutdown cooling subsystem.

These valves will be stroke timed open and stroke timed closed during Cold Shutdowns in accordance with Paragraphs 4.2.1.2(c), (f), (g) and (h) ofPart 10 of the Code.  ;

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Revision 18 Section 5 Page 97 of 160

1 Verm::nt Yenkee Nucle:r Pcwer Stati:n Inservice Testing Prcgram i COLD SHUTDOWN JUSTIFICATION i O Number: CSJ-V09, Revision 1 (Sheet 1 of1) l THIS COLD SHUTDOWN JUSTIFICATION WAS DELETED IN REVISION 15 OF THE IST PROGRAM. I 1

This CSJ number is being maintained for traceability to the Third Interval IST Program Submittal l SER's.  ;

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Revision 18 Section 5 Page 98 of 160

Verm:nt Ycnkee Ncci:ar Pcwer Stati:3 Inservice Testing Pr:grt:m q COLD SHUTDOWN JUSTIFICATION V

Number: CSJ-V10, Revision 2 (Sheet 1 of1)

SYSTEM: Reactor Core Isolation Cooling COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V13-22 C 2 G-191174 Sh 1 G-09 This valve is the Reactor Core Isolation Cooling (RCIC) discharge to Feedwater Line "B" check valve.

The valve has a safety function in the open position to pass RCIC flow to the reactor and in the closed position to prevent backflow into the RCIC system.

JUSTIFICATION:

This valve cannot be exercised open dwing normal (power) operation since flow through this valve must be injected into the reactor coolant system. This would thermally shock the reactor nozzles and cause a positive reactivity excursion.

This valve is located in the steam tunnel which is inaccessible during power operations. Additionally, there O

wJ is no means to manually exercise this valve.

This valve will be exercised open and closed during Cold Shutdowns in accordance with Paragraphs 4.3.2.2(c), (f), (g) and (h) ofPart 10 ofthe Code.

O Revision 18 Section 5 Page 99 of 160

Verm:nt Ycnkee Nucl=r Pcwer Stati:n Ins:rvice Testing Prcgram COLD SHUTDOWN JUSIP' CATION

. b Number: CSJ-Vil, Revision 1 (Sheet 1 of1)

THIS COLD SHUTDOWN JUSTIFICATION WAS DELETED IN REVISION 18 OF THE IST PROGRAM.

This CSJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's. l l

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O Revision 18 Section 5 Page 100 of 160

4 Vcrm::t Ycnkee N:cle:r Pcwcr Stati:n Inservice Testing Prcgram i COLD SHUTDOWN JUSTIFICATION O Number: CSJ-V12, Revision 1(Sheet 1 of1)

THIS COLD SHUTDOWN JUSTIFICATION DELETED IN REVISION 18 OF THE IST ,

PROGRAM.

This CSJ number is being maintained for traceability to the Third Interval IST Program Submittal  :

SER's.

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I Revision 18 Section 5 Page 101 of 160 1

Verm:nt Yrnkee N: clear Pcwer Stati:n Ins:rvice Testing Pr: gram '

COLLSHUTDOWN JUSTIFICATION I Numbec CSJ-V13, Revision 1 (Sheet 1 of1)

THIS COLD SHUTDOWN JUSTIFICATION DELETED IN REVISION 18 OF THE IST PROGRAM. '

This CSJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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O Revision 18 Section 5 Page 102 of 160

Verm:nt Ycnkee Nuclear Pcwer Stati:n Ins rvice Tcsting Program COLD SHUTDOWN JUSTIFICATION U

Number: CSJ-V14 Revision 1 (Sheet 1 of1)

SYSTEM: ResidualHeatRemoval COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V10-46A A/C 1 G-191172 E-07  ;

V10-46B A/C 1 G-191172 E-10 These valves are the Low Pressure Coolant Injection (LPCI) injection check valves. They have a safety function in the open position to pass LPCI flow to the reactor, and in the closed position for primary containment and pressure isolation.

JUSTIFICATION:

These valves cannot be exercised open during nonnal (power) operation since the Residual Heat Removal (RHR) pump discharge cannot overcome reactor pressure. Manual exercising is not possible during plant operation because the valves are located inside the inertal drywell and are not accessible during plant ;

operation.

i These valves will be exercised open with flow during Cold Shutdown in accordt.nce with Paragraphs 4.3.2.2(c), (f), (g) and (h) ofPart 10 of the Code.

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Revision 18 Section 5 Page 103 of 160

Verm:nt Yarkee Nucisar P:wer Stati:n Inservice Testing Program COLD SHUTDOWN JUSTIFICATION Number: CSJ-V15 Revision 1 (Sheet 1 of1) i SYSTEM: ResidualHeatRemoval COMPONENTS:

l Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord. {

V10-48A C 2 G-191172 L-05 V10-48B C 2 G-191172 L-13 V10-48C C 2 G-191172 J-05 V10-48D C 2 G-191172 J-13 l These valves are the RHR pump discharge check valves. These valves have a safety function to open to pass RHR pump discharge flow for all modes of RHR/LPCI system operation. These valves also have a safety function to close to prevent the backflow ofwater through an idle RHR pump.

Justification:

It is not practical to perform a full-stroke open exercise of these valves on a quarterly basis during plant operation. These valves are exercised during the performance of the quarterly RHR pump surveillance.

During plant operation, the only practical Bow path for these pumps is in the torus to torus flow path. The RHR pumps do not have sufficient head to flow to the recirculation loops with the reactor coolant system at normal operating pressure. Typical RHR pump flowrates in the torus to torus flow configuration are approximately 6,500 gpm for the RHR pump test. The required design accident condition flow rate to perform a full-stroke exercise is 7,300 gpm which can only be achieved when the RHR system is in the vessel to vessel flow configuration.

These valves will be part-stroke open exercised quarterly during plant operation and will be full-stroke open exercised during cold shutdowns in accordance with Paragraphs 4.3.2.2(b), (f), (g) and (h) of Part 10 of the Code.

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O Revision 18 Section 5 Page 104 r,f 160

Verm::t Ycckee Nucl=r P;w:r St:tinn Irservice Tcsting Program q COLD SHUTDOWN JUSTIFICATION V

Number: CSJ-V16, Revision 0 (Sheet 1 of 2)

SYSTEM: Service Water COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-11 B 3 G-191159 Sh 2 D-07 V70-17 B 3 G-191159 Sh 2 D-06 SB-70-1 B NNS G-191159 Sh 2 D-09 V70-18 B 3 G-191159 Sh 2 E-07 Valve V70-11 is the SW Discharge to Cooling Tower Basin Isolation Valve. The safety function of this valve is to open to provide water to the Cooling Tower Basin to prevent the development ofice in the basin.

Valve V70-17 is the Altemate Cooling to SW Supply - /e. The safety function of this valve is to open to provide flow to the Cooling Tower spray header to ovide an altemate cooling path for the Service Water System.

b v

Valve SB-70-1 is the SW Discharge to Main Condenser Discharge Block Valve. The safety function of this valve is to open to provide a flow path to the river. These valves have an active function to change position in order to provide a path for flow to the Cooling Tower when the Service Water System is required to be in the Altemate Cooling mode ofoperation.

Valve V70-18 is the Service Water Header Discharge Isolation Valve. The safety function of this valve is to close to maintain inventory of vital cooling water when the Service Water System is in the Altemate Cooling mode ofoperation.

JUSTIFICATION:

It is not practicable to full or part stroke exercise these valves quarterly during normal (power) operation for the following reasons.

l The Service Water system is designed with two discharge flow paths - main condenser discharge )

block and cooling tower. The discharge flow path is selected based on environmental conditions.

Service Water is discharged to the cooling tower deep basin when Connecticut River temperature is less than 45 F to keep the deep basin from freezing. The deep basin cannot be allowed to freeze, because it contains water inventory necessary for alternate cooling mode operation of Service Water system. Service Water is discharged to the condenser discharge block when Connecticut River temperature is greater than or equal to 45 F. This discharge path causes less back pressure in the O discherseiine di creesestiewtare satne smerse cr oiesei oe erteriectet Weterceeiers.

The flow increase is necessary to compensate for higher river water temperature.

I Revision 18 Section 5 Page 105 of 160

Vcrmrt Ycnkee Nuclear Pcwer Statin Iaervice Testing Program f

t COLD SHUTDOWN JUSTIFICATION Number: CSJ-V16, Revision 0 (Sheet 2 of 2)

V70-11, V70-17, SB-70-1 and V70-18 are large (20" and 24") manual valves and must be operated as a set because their positions are interdependent. For example, allowing V70-11 and ,.< I to be open at the same time, except for the minimum time required for system lineup changes, cod.d drain the deep basin inventory. Closing V70-18 with Service Water discharge lined-up to the main condenser discharge block will isolate Service Water from the safety-related cooling loads and closing V70-18 with Service Water discharge lined-up to the cooling tower will isolate turune building cooling loads.

These valves will be full stroked open and close during cold shutdown in accordance with Paragraph 4.2.1.2(c), (f) and (g) when Service Water cooling loads are more managable.

I Revision 18 Section 5 Page 106 of 160

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Verm:nt Ycnkee N: clear Pcwcr Statia Ins:rvice Testing Program l l

O COLD SHUTDOWN JUSTIFICATION '

V l

Number: CSJ-V17 Revision 0 (Sheet 1 of1) l SYSTEM: Instrument Air  !

COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V72-28A B NNS G-191160 Sh 3 L-15 V72-28B B NNS G-191160 Sh 3 L-16  !

V72-28D B NNS G-191160 Sh 3 K-16 l V72-28E B NNS G-191160 Sh 3 K-16 l

V72-28A and V72-28B have a safety function to close to isolate the Instrument Air (IA) system from the ,

outboard MSIVs. Following an Appendix R fire event in the control room and cable spreading room, V72- l' 28A and V72-28B are closed to isolate the air supply and V72-28D and V72-28E are opened to vent the residual air contained in the piping from the air supply isolation valves to the MSIV accumulators. This action prevents the MSIVs from inadvenently re-opening if a hot shon were to occur in the MSIV solenoid circuitry.

Reference:

Minor Modification 96-34.

V72-28D and V72-28E have a safety function to open to vent the residual air contained in the IA system r3 A

seggiv 1e the ee1 beard MS1vs <v2-8ei. v2-8ee. v2-8eC and v2 8eD>. Pe1>ew1ns an Aggendix R 11re event in the control room and cable spreading room, V72-28A and V72-28B are closed to isolate the air supply and V72-28D and V72-28E are opened to vent the residual air contained in the piping from air supply isolation valves to the MSIV accumulators. This action prevents the MSIVs from inadvertently re-opening if a hot shon were to occur in the MSIV solenoid circuitpf.

Reference:

Minor Modification 96-34. 1 I

Justificatmn:

l V72-28A and V72-28B cannot be exercised closed and cannot be exercised open during power operation on a quanerly basis exercising these valves would isolate / vent instrument air to the MSIV accumulators.

Instmment Air pressure is required to maintain the MSIVs in the open position. The loss ofinstrument air pressure would cause the MSIVs to go to the closed position and result in a reactor power transient.

I V72-28A and V72-28B will be full-stroke closed and V72-28D and V72-28E will be full-stroke opened during Cold Shutdowns in accordance with Paragraphs 4.2.1.2(f) & (g) ofPart 10 of the Code.  ;

O Revision 18 Section 5 Page 107 of 160

Vcrmut Yc kee Nuclear Pcwer Stati:n Inservice Testing Progr m COLD SHUTDOWN JUSTIFICATION Number: CSJ-V18, Revision 0 (Sheet I of1) 4 SYSTEM: Residual Heat Removal COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V10-17Al B 2 G-191172 G-08 i

Valve V10-17Al is the V10-17 outboard containment isolation valve bonnet pressure locking relief r device. This is a manual globe type valve. The safety function of this valve is to open to alleviate ;

bonnet pressure locking of V10-17 in an Appendix R scenario.

JUSTIFICATION:

i It is not practical to full or part-stroke open V10-17Al during power operation. Opening V10-17Al is a potential personnel / contamination hazard due to the release of high pressure (1000 psig) contaminated fluid. During cold shutdown, V10-17 bonnet pressure will equalize with reactor l,f pressure and opening V10-17Al is not a personnel hazard.

2 V10-17Al will be full-stroke opened during Cold Shutdowns in accordance with Paragraphs 4.2.1.2(f) &

(g) ofPart 10 of the Code.

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Revision 18 Section 5 Page 108 of 160

1 Verm:nt Yenkee Nucl cr Pcwcr Stati::n Inservice Testing Pr: gram p REFUELING OUTAGE JUSTIFICATION l

Number: ROJ-V01, Revision 2 (Sheet 1 of3)  !

SYSTEM: Nuclear Boiler ,

High Pressure CoolantInjection j Reactor CoreIsolation Cooling i

Recirculation Pump Cooling Water )

COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord. j

, SL-13-55A A/C 2 G-191174 Sh 1 B-08  !

SL-13-55B A/C 2 G-191174 Sh 1 C-08

, SL-13-55C A/C 2 G-191174 Sh 1 B-08 l SL-13-55D A/C 2 G-191174 Sh 1 C-08 SL-14-31 A A/C 2 G-191168 F-07 i

SL-14-31B A/C 2 G-191168 E-07 SL-2-62A A/C 2 G-191167 I-13 SL-2-62B A/C 2 G-191167 I-13 i SL-2-62C A/C 2 G-191167 J-13 i SL-2-62D A/C 2 G-191167 J-13 SL-2-64A NC 2 G-191167 K-13 SL-2-64B A/C 2 G-191167 K-13 i SL-2-64C A/C 2 G-191167 K-13 j SL-2-64D NC 2 G-191167 L-13 SL-2-73A A/C 2 G-191167 F-12 SL-2-73B NC 2 G-191167 F-12 SL-2-73C A/C 2 G-191167 G-12 l SL-2-73D NC 2 G-191167 G-12 SL-2-73E A/C 2 G-191167 G-12

, SL-2-73F A/C 2 G-191167 G-12 SL-2-73G NC 2 G.191167 H-12 SL-2-73H NC 2 G-191167 H-12 SL-2-2-7A A/C 2 G-191159 Sh 5 G-02 I SL-2-2-7B A/C 2 G-191159 Sh 5 G-02 SL-2-2-8A A/C 2 G-191159 Sh 5 G-02 SL-2-2-8B A/C 2 G-191159 Sh 5 G-02 3

SL-2-3-11 NC 2 G-191267 Sh 1 C-06 SL-2-3-13 A A/C 2 G-191267 Sh 1 D-06 SL-2-3-13B NC 2 G-191267 Sh 1 C-12 SL-2-3-15 A A/C 2 G-191267 Sh 1 D-06 O

Revision 18 Section 5 Page 109 of 160

Vermrt Ycnkee Nucirr P wer Stati:n Ins:rvice Testing Prsgram REFUELING OUTAGE JUSTIFICATION i

O Number: ROJ-V01, Revision 2 (Sheet 2 of 3) l Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

SL-2-3-15B NC 2 G-191267 Sh 1 D-12 SL-2-3-17A NC 2 G-191267 Sh 1 E-06 SL-2-3-17B A/C 2 G-191267 Sh 1 E-12 SL-2-3-19A NC 2 G-191267 Sh 1 F-06 SL-2-3-19B A/C 2 G-191267 Sh 1 F-12 SL-2-3-21 A NC 1 G-191267 Sh 1 H-06 SL-2-3-21B NC 1 G-191165 C-13 SL-2-3-21C A/C 1 G-191267 Sh 1 I-12 SL-2-3-21D NC 1 G-191165 C-13

. SL-2-3-23A A/C 1 G-191267 Sh 1 G-06

! SL-2-3-23B A/C 1 G-191267 Sh 1 G-06 SL-2-3-23C A/C 1 G-191267 Sh 1 G-12 SL-2-3-23D NC 1 G-191267 Sh 1 G-12

! SL-2-3-25 NC 2 G-191267 Sh 1 I-04 SL-2-3-27 NC 2 G-191267 Sh 1 I-06 SL-2-3-31 A. NC 1 G-191267 Sh 1 F-06 Q

V SL-2-3-31B SL-2-3-31C NC A/C 1

1 G-191267 Sh 1 G-191267 Sh 1 F-06 F-06 j _

SL-2-3-31D A/C 1 G-191267 Sh 1 F-06 SL-2-3-31E A/C 1 G-191267 Sh 1 F-06 J

SL-2-3-31F A/C 1 G-191267 Sh 1 F-06

SL-2-3-31G A/C 1 G-191267 Sh 1 F-06 SL-2-3-31H A/C 1 G-191267 Sh 1 F-06 SL-2-3-3 II NC 1 G-191267 Sh 1 F-12 SL-2-3-31J A/C 1 G-191267 Sh 1 F-12 SL-2-3-31K A/C 1 G-191267 Sh 1 F-12 SL-2-3-31L NC 1 G-191267 Sh 1 F-12

, SL-2-3-31M A/C 1 G-191267 Sh 1 F-12 SL-2-3-31N NC 1 G-191267 Sh 1 F-12 SL-2-3-31P A/C 1 G-191267 Sh 1 F-12

_SL-2-3-31Q A/C 1 G-191267 Sh 1 F-12 SL-2-3-33 NC 2 G-191267 Sh 1 I-12 a SL-2-3-35 A/C 2 G-191267 Sh 1 J-06 r

Revision 18 Section 5 Page 110 of 160 1

i Verm:nt Yankee Nuchar Power Stati:n Inservice Testing Progrcm REFUELING OUTAGE JUSTIFICATIOIS l Number: ROJ-V01, Revision 2 (Sheet 3 of 3) l 4

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

SL-2-305A A/C 2 G-191167 M-02 SL-2-305B A/C 2 G-191167 M-02 SL-23-37A A/C 2 G-191169 Sh 1 F-05 SL-23-37B A/C 2 G-191169 Sh 1 F-05 SL-23-37C A/C 2 G-191169 Sh 1 F-05 SL-23-37D A/C 2 G-191169 Sh 1 F-05 These valves are instrumentation line excess flow check valves. They are required to be verified operable in accordance with Vermont Yankee Technical Specification 3.7.D.l. I JUSTIFICATION:

These valves cannot be exercised closed during normal (power) operation since closing these valves would isolate instrumentation required for power operation. These valves can only be verified to closed by leak testing performed during the primary system inservice pressure test performed each refueling outage. This test cannot be repeated during each Cold Shutdown since the reactor vessel is not pressurized during Cold Shutdowns.

O d These valves will be exercised open and closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.3 and 4.3.2.2(e), (h) ofPart 10 of the Code. l l

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Revision 18 Section 5 Page 111 of 160

Vermrt Yankee Ncclear Pcw r Stati:n I: service Testing Program O REFUEI,ING OUTAGE JUSTIFICATION V

Number: ROJ-V02, Revision 1 (Sheet 1 of1)

SyjTJM: NuclearBoiler COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

RV-2-71 A C 1 G-191167 D-08 RV-2-71B C 1 G-191167 G-08 RV-2-71C C 1 G-191167 G-08 RV-2-71D C 1 G-191167 H-08 Valves RV-2-71 A through D are the nuclear boiler main steam srfety/ relief valves. They have a safety function in the open position for Automatic Depressunzation System (ADS) operation and for overpressure protection, and in the closed position to maintain reactor coolant inventory.

JUSTIFICATION:

Stroke testing of these valves quarterly during power operation is not recommended. As recommended by g". NUREG-0737 and the corresponding study by the BWR Owners Group (BWR Owners Group Evaluation of NUREG-0737, Item II.K.3.16, " Reduction of Challenges and Failures of Relief Valves"), the exercising of these valves should be minimized to reduce the number of challenges to safety /reliefvalves.

The failure of any relief valve to close during stroking at power operation would cause an uncontrolled rapid depressurization of the primary system (stuck open relief valve transient) along with an undesired positive reactivity excursion.

These valves can only be tested with primary system pressure greater than 100 psig, therefore, they cannot be exercised during cold shutdowns or during refueling outages. Exercising these valves at each start-up from cold shutdown (or quarterly) would produce additional stress cycles on the nuclear boiler system which could lead to a low cycle fatigue failure. These valves will be exercised during plant startup aller each refueling outage in accordance with Paragraph 4.2.1.2(h) of Part 10 of the Code and Paragraph 3.4.1.l(d) ofPart 1 of the Code.

This Refueling Outage JustiScation is for clarification purposes only.

O Revision 18 Section 5 Page 112 of 160

Vermxt Ycnkee Nuclear Peer Statin Inservice Testing Program REFUELING OUTAGE JUSTIFICATION Number: ROJ-V03, Revision 2 (Sheet 1 of 2)

SYSTEM: NuclearBoiler COMPONENTS:

4 Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V2-27A NC 2

{

G-191167 F-03 i V2-27B C 2 G-191167 H-02

, V2-28A NC 2 G-191167 F-05 V2-28B NC 2 G-191167 H-05 ,

V2-96A NC 2 G-191167 H-03 i V2-96B C 2 G-191167 F-03 i

' Valves V2-27A and V2-28A are the outboard and inboard containment isolation check valves for the "A"  !

Feedwater Line, respectively. They have a safety function in the closed position to provide primary containment isolation, and in the open position to pass High Pressure Coolant Injection flow to the reactor.

, Valves V2-96A and V2-28B are the outboard and inboard containment isolation check valves for the "B" Feedwater Line, respectively. They have a safety function in the closed position to provide primary containment isolation, and in the open position to pass Reactor Core Isolation Cooling flow to the reactor.

Valve V2-96B is the second outboard check valve in the "A" Feedwater Line. The valve has a safety function in the closed position to prevent diversion of High Pressure Coolant Injection flow in the "A" ,

Feedwater Line.

i Valve V2-27B is the second outboard check valve in the "B" Feedwater Line. The valve has a safety function in the closed position to prevent diversion of Reactor Core Isolation Cooling flow in the "B" Feedwater Line.

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l Verm::t Yankee Nucl:ar Psw:r Statin Inservice Testing Pr:grcm REFUELING OUTAGE JUSTIFICATION Number: ROJ-V03, Revision 2 (Sheet 2 of 2) 4 JUSTIFICATION:

These valves cannot be exercised closed during normal (power) operation because the feedwater system is required to maintain reactor vessel water level. Interruption of feedwater to perform the exercise test of these valves would result in a reactor scram.

These valves can only be stroke close tested via a leak type or non-intmsive test. Testing durin; Cold  :

Shutdowns is impracticable due to the significant system and test equipment configurations required.  !

Additionally, valves V2-27B, V2-28B and V2-96A cannot be exercised during Cold Shutdowns since this would require removing the only mechanism of reactor vessel level control (via the Reactor Water Cleanup System).

l' V2-27A, V2-28A, V2-28B and V2-96A will be stroke close tested each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.2 and 4.3.2.2(e), (h) ofPan 10 of the Code.

V2-27B and V2-96B will be stroke close tested each Refueling Outage using non-intrusive testing or disassembly in accordance with Paragraphs 4.3.2.2(e), (h) and 4.3.2.4(a) or (c) ofPan 10 of the Code.

t' V2-27B and V2-96B were replaced during the 1996 Refueling Outage. Valve disassembly was used instead of a non-intrusive test for this refuel outage because-

. These valves were disassembled to facilitate installation.

• These valves were manually stroked after installation to verify freedom of movement and blue checked.

These action satisfy the OM-10 stroke close requirement intent. l I

l Revision la Section 5 Page 114 of 160

Verm=t Yankee Nuclear Psw r Stati:n Ins:rvice Testing Prcgram REFUELING OUTAGE JUSTIFICATION Number: ROJ-V04, Revision 1 (Sheet 1 of1)

SYSTEM: NuclearBoiler l

COMPONENTS:

! Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

I V2-37A A/C 2 G-191167 B-08 V2-37B A/C 2 G-191167 G-09 i V2-37C A/C 2 G-191167 G-09 V2-37D A/C 2 G-191167 H-09

~

These valves are the Main Steam Relief Valve (MSRV) actuator air supply check valves. They have a safety function in the closed position to ensure automatic depressunzation system (ADS) capability is l maintained via the MSRV accumulators on a loss of the instmment nitrogen supply. l JUSTIFICATION:

These valves are located in the drywell and thus cannot be stroke close exercised during normal power .

operations or Cold Shutdowns when the drywell is inerted. These valves can only be stroke close tested O

(V via a leak type or non-intmsive test which would require isolating the instrument nitrogen supply to the MSRVs. Testing during Cold Shutdowns is impracticable due to the significant system and test equipment configurations required.

These valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.3 and 4.3.2.2(e), (h) ofPart 10 of the Code.  ;

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1 Revision 18 Section 5 Page 115 of 160 J

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Verm:ct Yc:kee Nuclear Pcwer Stati:n Inservice Testing Program REFUELING OUTAGE JUSTIFICATION I (m1

%)  !

Number: ROJ-V05, Revision 1 (Sheet 1 of1)

SYSTEM: NuclearBoiler I

COMPONENTS:  !

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V2-82A A/C 2 G-191167 B-10 l V2-82B A/C 2 G-191167 B-10 V2-82C A/C 2 G-191167 B-10 V2-82D A/C 2 G-191167 B-10 V2-87A A/C 2 G-191167 B-13 V2-87B A/C 2 G-191167 B-13 V2-87C A/C 2 G-191167 B-13 i V2-87D A/C 2 G-191167 B-13 l l

i These valves are the Main Steam Isolation Valve (MSIV) actuator air supply check valves. They have a safety function in the closed position to ensure MSIV capability is maintained via the MSIV accumulators on a loss ofinstrument air or nitrogen supply.

JUSTIFICATION:

Valves V2-82A through D are located in the drywell and thus cannot be stroke close tested during normal power operations or Cold Shutdowns when the drywell is inerted.

Valves V2-87A through D are located in the main steam tunnel which is inaccessible during power operations, thus, these valves cannot be stroke close tested during normal power operations.

l These valves can only be stroke close tested via a leak type or non-intmsive test which would require isolating the instrument air or nitrogen to the MSIVs. Testing during Cold Shutdowns is impracticable due to the signmcant system and test equipment configurations required.

Tir:se valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.3 and 4.3.2.2(e), (h) ofPart 10 of the Code.

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Revision 18 Section 5 Page 116 of 160

Verm=t Yc:kee Nuclear Pcwer Station Ins:rvice Testing Pr: gram REFUELING OUTAGE JUSTIFICATION Number: ROJ-V06, Revision 3 (Sheet 1 of2)

SYSTEM: Control Rod Drive Hydraulic COMPONENTS: (Typical of 89 each)

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

CV-3-126 B 2 G-191170 C-16 CV-3-127 B 2 G-191170 C-18 S 0-3-120 B 2 G-191170 D-17 S 0-3-121 B 2 G-191170 D-18 S 0-3-122 B 2 G-191170 C-18 SO-3-123 B 2 G-191170 C-17 V3-114 C 2 G-191170 C-18 V3-115 A/C 2 G-191170 B-16 V3-137 C 2 G-191170 B-17 Valves CV-3-126 & 127 are the Control Rod Drive (CRD) scram valves. These valves have a safety function in the open position to pass scram accumulator discharge to the control rod drives for a reactor g scram.

O Valves S0-3-120 & 123 are the manual control exhaust and insertion valves, respectively, for the CRD under piston area. They have a safety function in the closed position to prevent diversion of scram accumulator discharge into the exhaust water header or the drive water header.

Valves S0-3-121 & 122 are the manual control exhaust and withdrawal valves, respectively, for the CRD over piston area. These valves have a safety function in the closed position to ensure that scram exhaust flow is properly directed to the discharge volume.

Valves V3-114 are the scram exhaust to the discharge volume check valves. These valves have a safety function in the open position to pass scram exhaust flow to the discharge volume. These valves have a safety function in the closed position to prevent the scram discharge volume from operating the drive in the event that the scram discharge volume pressure should exceed reactor pressure following a scram.

Valves V3-115 are the charging water to the CRD accumulator check valves. These valves have a safety function in the closed position to prevent diversion of scram accumulator discharge into the charging header.

Valves V3-137 are the Drive Water Supply Check Valves. These valves have a safety function in the j closed position to prevent a loss ofinventory to the drive water riser i bp r

Revision 18 Section 5 Page 117 of 160 l

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I Verm:nt Yankee Nuclear Power Stati:n Ins:rvicit aang Program l REFUELING OUTAGE JUSTIFICATION Number: ROJ-V06, Revision 3 (Sheet 2 of2)

JUSTIFICATION: '

Exercising valves CV-3-126, CV-3-127, V3-114 and V3-137 during power operation would require l scrammmg the plant solely for testing purposes. Since scram insertion times are representative of valve l

operability and stroke times, testing will be performed in accordance with Technical Specifications 4.3.C.1 and 2. These sections require that all control rods be subjected to scram-time measurements on a refueling l l outage basis and that 50% of the control rods be measured for scram times every 16 to 32 weeks. An  ;

l evaluation is required that provides reasonable assurance that proper control rod drive performance is '

l being maintained. These tests will adequately verify valve operability and stroke times.

Valves S0-3-120 through 123 are verified operable at least once a week in accordance with Technical Specifications 4.3.A.2 for each partially or fully withdrawn operable control rod. For a control rod that is fully inserted, the safety function of these valves is fulfilled. Stroke testing these valves while the control rod is fully inserted will not result in an increase in safety or quality. All control rods drives are verified operable in accordance with Technical Specifications 4.3.B.1 each Refueling Outage, thus stroke testing of these valves for control rods that remained fully inserted throughout the operating cycle will occur each Refueling Outage. This alternate test frequency is consistent with NUREG 1482, Appendix A, NRC Staff Position 7 " Testing Individual Scram Valves For Control Rods in Boiling Water Reactors."

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' V Closure testing of V3-Il5 requires that the CRD pumps be stopped to depressurize the charging water header. Therefore the accumulator pressure decay test (OP 4111) will be performed during refueling outages. This altemate test frequency is consistent with NUREG 1482, Appendix A, NRC Staff Position 7

" Testing Individual Scram Valves For Control Rods in Boiling Water Reactors." '

Closure testing of V3-ll4 requires a scram signal to be in and the scram discharge volume to be pressurized. This condition is beyond operating parameters as specified in Technical Specification 3.3.,

therefore testing can only be perfonned during a refueling outage when test conditions can be met.

These valves will be exercised as a muumum each Refueling Outage in accordance with Paragraphs i 4.2.1.2(e), (h) and 4.3.2.2(e), (h) of Part 10 of the Code.  !

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l Revision 18 Section 5 Page 118 of 160

Vcrmut Ycnkee Nucle r Pcwer Stati:n Inservice Testing Program o REFUELING OUTAGE JUSTIFICATION

]

Number: ROJ-V07, Revision 2 (Sheet 1 of1)

SYSTEM: Control Rod Drive Hydraulic COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V3-412A A/C 2 G-191170 E-12 V3-412B A/C 2 G-191170 E-12 V3-413 A A/C 2 G-191170 E-12 V3-413B A/C 2 G-191170 E-12 These valves are the recirculation pump seal purge supply check valves. They have a safety function in the closed position to provide primary containment isolation.

JUSTIFICATION:

These valves can only be exercised closed via a leak type or non-intrusive test which would require isolating the seal purge to the recirculation pumps. To preclude adverse affects on seal life, the recirculation pumps would have to be secured. Testing during Cold Shutdowns is impracticable due to the significant system and test equipment configurations required.

These valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.2 and 4.3.2.2(e), (h) ofPart 10 of the Code.

e Revision 18 Section 5 Page 119 of 160 I

Vermnt Yankee Nuclear Pcwer Stati:n Inservice Testing Prrgram p REFUELING OUTAGE JUSTIFICATION Number: ROJ-V08, Revision 2 (Sheet 1 of1)

SYSTEM: Standby Liquid Control COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

VIl-16 C 1 G-191171 H-03 VIl-17 C 1 G-191171 I-02 These valves are the Standby Liquid Control (SLC) injection line check valves. They have a safety function in the open position to pass borated water into the reactor, and in the closed position to provide primary containment isolation.

JUSTTFICATION:

Exercising these valves open during power operation would require injecting borated water into the reactor coolant system. This would create a reactivity excursion and potential for reactor trip. Injection of deminerahzed water would require removing the SLC system from service to clean the borated solution from the piping and replacing the explosive actuated valves.

Full-stroke closed testing requires the removal of at least one explosive actuated valve.

These valves will be full-stroke exercised open during refueling outages when the SLC system can be tested without creating a reactivity excursion.

These valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.3.2.2(e), (h) ofPart 10 of the Code.

Revision 18 Section 5 Page 120 of 160

Verm:nt Ycnkee Ncclear Pcwer Stati:n Ins:rvice Testing Program q REFUELING OUTAGE JUSTIFICATION b

Number: ROJ-V09, Revision 0 (Sheet 1 of 1) l l

tills REFUELING OUTAGE JUSTIFICATION WAS WITHDRAWN IN REVISION 18 OF THE IST PROGRAM.

This ROJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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O i Revision 18 Section 5 Page 121 of 160

Verm:nt Ymkee Nucl:ar Pcwer Station Inservice Testing Pr: gram O. ,

REFUELING OUTAGE JUSTIFICATION Number: ROJ-V10, Revision 1 (Sheet 1 of1)

THIS REFUELING OUTAGE JUSTIFICATION WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM.

This ROJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Revision 18 Section 5 Page 122 of 160

i Vermut Yenkee Nuclear Power Stati:n Ins:;rvice Testing Pr: gram '

REFUELING OUTAGE JUSTIFICATION Number: ROJ-Vil, Revision 2 (Sheet 1 of1)

SYSTEM: Neutron Monitoring System COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.  ;

V7-1 A/C 2 5920-271 C-08  !

V7-2 A/C 2 5920-271 D-08 i These valves are the Neutron Monitoring System (NMS) Tip Purge primary containment isolation valves.

These valves have a safety function in the closed position to provide primary containment isolation.

JUSTIFICATION:

These valves can only be tested via a leak type or non-intrusive test that requires secunng of the nitroge purge to the NMS Tip system. The nitrogen purge is required dwing operation and Cold Shutdown to i prevent condensation and corrosion in the NMS Tip system. ,

These valves will be exercised closed each Refueling Outage during leakage rate testing performed in O

l accordance with Parasraphs 4.2.2.2 and 4.3.2.2(e), (h) ofPart 10 of the Code.

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I Revision 18 Section 5 Page 123 of 160 l 1 -

Verntnt Ycnkee Nuclear Pcwer Stati:n Inservice Testing Progrcm l

REFUELING OUTAGE JUSTIFICATION U

Number: ROJ-V12, Revision 1 (Sheet 1 of1) l SYSTEM: Control Rod Drive Hydraulic COMPONENTS:

i i

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V3-162A A/C 2 G-191170 A-09  ;

V3-162B A/C 2 G-191170 A-01 These valves are the Control Rod Drive (CRD) scram discharge volume vent check valves. They have a safety function in the closed position to isolate the scram discharge volume during a scram condition, thereby preventing reactor coolant inventory loss.

JUSTIFICATION:

These valves can only be exercised closed via a leak type or non-intrusive test which would require removing the CRD system from service. Testing during Cold Shutdowns is impracticable due to the significant system and test equipment configurations required.

These valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.3 and 4.3.2.2(e), (h) ofPart 10 of the Code.

O Revision 18 Section 5 Page 124 of 160

1 Verm:nt Ycnkee Nuclear Psw:r Stati:n Inservice Testing Pr: gram REFUELING OUTAGE JUSTTFICATION i

Number: ROJ-V13, Revision 1 (Sheet 1 of1)

THIS REFUELING OUTAGE JUSTIFICATION WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM.

\

This ROJ number is being maintained for traceability to the Third Interval IST Program Submittal i SER's.

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Revision la Section 5 Page 125 of 160

i Verm:nt Ytnkee Nucirr Pcwer St:ti:n Inservice Testing Pr: gram q REFUELING OUTAGE JUSTIFICATION O

Number: ROJ-V14, Revision 1(Sheet 1 of1) l SYSTEM: ServiceWater COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-1 A C 3 G-191159 Sh I C-02 V70-1B C 3 G-191159 Sh 1 B-02 V70-1C C 3 G-191159 Sh 1 K-02 V70-ID C 3 G-191159 Sh 1 J-02 ,

These valves are the station service water pump discharge check valves. They have a safety function in the open position to provide cooling water to systems and equipment required to operate under accident conditions and to provide an inexhaustible supply of water for standby coolant system operation. They have a function in the closed position to prevent the diversion of cooling water through an idle station Service Water pump JUSTIFICATION:

O O It is not practical to verify the full-stroke open function of these check valves. The service water pumps do not have installed plant instmmentation to determine pump flow rate. The design flow rate through these check valves can only be verified during the refheling outage service water pump capacity test when each service water pump is run at its design capacity (Ref. RR-P01). This testing involves installing a temporary fully instrumented test loop to determine pump flow rate.

These check valves will be part-stroke exercised open during the regularly scheduled quarterly service water pump tests. These valves will be full-stroke exercised open and closed in accordance with Paragraph 4.3.2.2(e), (h) ofPart 10 of the Code during each refueling outage service water pump capacity test.

l Revision 18 Section 5 Page 126 of 160

Verm:nt Ytnkee Nuclear Power St:ti:n Inservice Testing Program REFUELING OUTAGE JUSTTFICATION C

Number: ROJ-V15, Revision 1 (Sheet 1 of1)

SYSTEM: NuclearBoilerVesselInstrumentation COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V-2-3-430A A/C 2 G-191267 Sh 2 C-09 '

V-2-3-430B A/C 2 G-191267 Sh 2 E-09 V-2-3-432A A/C 2 G-191267 Sh 2 C-10 V-2-3-432B A/C 2 G-191267 Sh 2 E-10 V-2-3-433 A A/C 2 G-191267 Sh 2 F-09 V-2-3-433 B A/C 2 G-191267 Sh 2 H-09 V-2-3-435A A/C 2 G-191267 Sh 2 F-10 V-2-3-435B A/C 2 G-191267 Sh 2 H-10 These valves are the reactor vessel instmmentation reference leg back fill inlet check valves. These valves are required to close in order to prevent the reference legs from emptying in the event of a break in the non-safety related portion of the back-fill system and perform a containment isolation function.

JUSTIFICATION:

, These valves cannot be exercised during power operations or during cold shutdowns since shutting these valves would isolate filling water to the reference leg lines for reactor vessel pressure and level instrumentation. The function of the reference leg backfill system is to ensure that the reactor water level reference leg fluid does not become saturated with non-condensable gases. Full closure of these valves is verified during local leakage rate tests when the test boundary is drained and vented which could introduce ,

non-condensable gases to the reference legs. l

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These valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.2 and 4.3.2.2(e), (h) ofPart 10 of the Code.

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Vermnt Ycnkee Nucirr Power Statirn Inservice Testing Pr: gram p REFUELING OUTAGE JUSTIFICATION V

Number: ROJ-V16, Revision 1 (Sheet 1 of1)

SYSTEM: ServiceWater COMPONENTS:

j Valve Number OM Cat. Safety Class I Drawing Number Dwg. Coord.

! V70-13A B l 3 G-191159 Sh 1 J-07 V70-13B B 3 G-191159 Sh 1 B-07 V70-2A B 3 G-191159 Sh 1 C-03 V70-2B B 3 G-191159 Sh 1 B-03 l V70-2C B 3 G-191159 Sh 1 J-03 V70-2D B 3 G-191159 Sh 1 i

J-03 l These valves are the Service Water System Header Supply Isolation Valves. The safety function of these valves is to close in order to isolate the upstream portions of the Service Water System when transferr to the Altemate Cooling mode ofoperation.

l JUSTIFICATION:

It is impracticable to full or part-stroke close exercise these valves on a quarterly basis or during cold shutdowns. Closure of these valves could interrupt cooling flow to their respective trains of the Reactor Building Closed Cooling Water system heat exchangers (E-8-1 A or E-8-1B), the RHR heat exchangers (E14-1 A or E14-1B) and the emergency diesel generator lube oil and jacket water coolers. Interruption of flow to these components could cause damage to equipment. These valves will be full-stroke exercised i closed during refuelirig outages in accordance with Paragraphs 4.2.1.2 (e) and (h) ofPart 10 of the Code when cooling flow to these components can be isolated.

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Number: ROJ-V17, Revision 1 (Sheet 1 of1) ,

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SYSTEM: ServiceWater i COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-15A B 3 G-191159 Sh 1 J-08 V70-15B B 3 G-191159 Sh 1 B-08 These valves are the Service Water System Header Supply Isolation Valves. The safety function of these l valves is to close to eliminate Altemate Cooling Flow from bypassing the cooling tower while the Service '

Water System is in the Altemate Cooling mode of operation.

1 JUSTIFICATION:

1 It is impracticable to full or part-stroke exercise these valves on a quarterly basis or during cold shutdowns.

l Closure of either of these valves could intermpt cooling flow in their respective trains to the Reactor '

Building Closed Cooling Water system heat exchangers (E-8-1A or E-8-1B). The RBCCW systems

' provide cooling flow to components such as the RHR pump coolers, the CRD pump coolers, and the spent

( fuel pool heat exchangers (E9-1A and E9-1B) which are required for the safe operation of the plant at power and cold shutdown periods. These valves shall be full-stroke exercised during refueling outages These valves will be full-stroke exercised closed in accordance with Paragraph 4.2.1.2(e), (h) of Part 10 of the Code during each refueling outage when these cooling loads can be safely isolated.

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l Revision 18 Section 5 Page 129 of 160

Verm:nt Ycnkee Nucle r Pcwer Stati:n Inservice Testing Pr:gr !m REFUELING OUTAGE JUSTIFICATION Number: ROJ-V19, Revision 0 (Sheet 1 of1)

THIS REFUELING OUTAGE JUSTIFICATION WAS WITHDRAWN FROM THE IST PROGRAM.

This ROJ number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

O Revision 18 Section 5 Page 131 of 160

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i Verm:nt Ycnkee Nucl=r Pswer Stati::n Inssrvice Testing Prcgram q REFUELING OUTAGE JUSTIFICATION b'  ;

Number: ROJ-V20, Revision 0 (Sheet 1 of1)  !

THIS REFUELING OUTAGE JUSTIFICATION WAS WITHDRAWN FROM THE IST PROGRAM.

l This ROJ number is being maintained for traceability to the Third Interval IST Program Submittal i SER's.

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Verm:nt Yankee Nucirr Pcwer Statin Inservice Testing Prcgram REFUELING OUTAGE JUSTIFICATION Number: ROJ-V21, Revision 2 (Sheet 1 of1)

SYSTEM: ServiceWater COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-16A B 3 G-191159 Sh 2 D-07 V70-16B B 3 G-191159 Sh I C-09 V70-184 B 3 G-191159 Sh 1 C-09 _

V70-16A and V70-16B are in the RHRSW pump suction line from the cooling tower. The safety function of these is to open to supply a suction source to the RHRSW pumps from the cooling tower when the senice water system is aligned for the altemate cooling mode of operation. V70-184 is used to vent piping between V70-16A and V70-16B when altemate cooling is placed in senice.

JUSTIFICATION:

It is impracticable to full or part-stroke exercise these valves on a quarterly basis or during cold shutdowns.

The suction line from the cooling towers to valve V70-16B is chemically treated to control the growth of O' microbiologically influenced corrosion (MIC). The opening of these valves could dilute the chemical nuxture in this line. Dilution ofthe chemical mixture in this line would reduce its effectiveness.

i These valves will be full-stroke exercised closed in accordance with Paragraph 4.2.1.2(e), (h) of Pan 10 of l the Code during each refueling basis when the line can be retreated with chemicals.

Revision 18 Section 5 Page 133 of 160

Verm=t Ycnkee Nuclear Pawar Stati:n Inservice Testing Program 1

l q REFUELING OUTAGE JUSTTFICATION v ,

1 Number: ROJ-V22, Revision 1 (Sheet 1 of1)

SYSTEM: Instmment Air System l

COMPONENTS:

3 1

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord. '

V72-89B A/C 2 G-191160 Sh 3 K-14 l V72-89C A/C 2 G-191160 Sh 3 K-15  :

These valves are the Instrument Air to the Drywell Containment Isolation Valves. These valves have a i safety function to close to prevent the release of fission products from the drywell to the reactor building in the event ofan accident.

JUSTIFICATION:

It is not practical to verify the closure function of these valves on a quarterly or cold shutdown basis. The only means to verify closure of these valves is to perform a leakage type test or to utilize a non-intmsive testing method.

'.( In order to assure closure of these valves to perform a leakage rate test or non-intrusive test, it would be necessary to isolate instmment air to the main steam isolation valves inside contamment. Additionally, in order to perform a leakage type test, it would be necessary to open system drams and vents. Isolation of l this line could result in closure of the air operated main steam isolation valves and subsequent reactor scram.

] These valves will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.2 and 4.3.2.2(e), (h) ofPart 10 of the Code.

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Vcrm:nt Ycnkee Nuclear P wsr Stati:a Inservice Testing Pr: gram  !

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REFUELING OUTAGE JUSTIFICATION Number: ROJ-V23, Revision 1 (Sheet 1 of1)

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SYSTEM: StandbyFuelPoolCooling l 1

COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

Vl9-224 C 3 G-191173 Sh 1 I-12 V19-224 is normally open during plant operation to allow cooling flow to pass to the fuel storage pool.

i The safety function of this valve is to close to provide isolation capability of the non-seismic nonnal fuel pool cooling system from the standby fuel pool cooling system.

JUSTIFICATION:

It is not practical to full-stroke exercise this valve to the closed position on a quanerly basis or during cold shutdowns. This valve is normally open to allow cooling flow from the normal fuel pool cooling pumps to the fuel storage pool. Additionally this valve is located directly upstream ofcheck valve Vl9-18. The only practical means to verify the closure of this valve is to pressurize the volume between Vl9-223, Vl9-53 and Vl9-46 and Vl9-224 utihzmg a differential pressure or leakage type test or by performing a non-intrusive test when normal fuel pool cooling system flow through the demineralizer is isolated. The use of L non-intmsive techniques to verify closure on a quarterly basis or during cold shutdowns is not practical.

These valves are not located in an easily accessible location and scaffolding is required to setup the necessary equipment.

NUREG 1482, subsection 4.1.4, states, "The NRC has determined that the need to setup test equipment is adequatejustification to defer backflow testing of a check va've until a refueling outage" l

The closure function of this valve will be verified during refueling outages by utilizing a differential pressure l or leakage type test or by performing a non-intrusive test in accordance with Paragraph 4.3.2.2(e), (h) of Part 10 of the Code.

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l Vcrmrt Yc kee N:cle:;r Pcwcr Station Izservice Testing Pr: gram l l

REFUELING OUTAGE JUSTIFICATION D

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Number: ROJ-V24, Revision 2 (Page 1 of1)

SYSTEM: Instrument Air System COMPONENTS: )

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-113 A/C 2 G-191159 Sh 3 K-08 This valve is the RBCCW system drywell supply line containment isolation valve. This valve has a safety function to close to prevent the release of fission products from the drywell to the reactor building in the event of an accident.

JUSTIFICATION: l It is not practical to verify the closure function of this valve on a quarterly or cold shutdown basis. The only means to verify closure of this valve is to perfonn a leakage type test or to utilize a non-intrusive i testing method.

l In order to assure closure of this valve, it is necessary to perform a leakage rate test or non-intrusive test.

Leakage type testing would require isolation of the RBCCW system supply to the drywell and opening of l (V] system drains and vents. Isolation of this lir.e would result in a loss of cooling flow to important plant I

equipment such as the recirculation pumps, pc ssibly resulting in their failure.

This valve will be exercised closed each Refueling Outage during leakage rate testing performed in accordance with Paragraphs 4.2.2.2 and 4.3.2.2(e), (h) ofPart 10 of the Code.

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,q REFUELING OUTAGE JUSTIFICATION  :

O Number: ROJ-V25, Revision 1 (Sheet 1 of1)

SYSTEM: High Pressure Coolant Injection COMPONENTS: I Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V23-56 C 2 G 191169 Sh 1 J-05 V23-65 C 2 G-191169 Sh 1 I-03 L

i V23-56 is the HPCI Turbine Drain Condensate Exhaust Line Containment Isolation Valve. This valve is 1 normally closed during plant operation. This valve has a safety function to open to pass turbine exhat.st condensate to the toms. Additionally, this valve has a safety function to close for containment isolation.

V23-65 is the HPCI Turbine Steam Exhaust Line Containment Isolation Valve. This valve is nonnally closed dudng plant operation. This valve has a safety function to open to pass turbine exhaust steam to the toms. Additionally, this valve has a safety function to close for containment isolation. I JUSTIFICATION:

It is not practical to vedfy the closure function of these valves on a quarterly or cold shutdown basis.

V During normal operation the HPCI steam and condensate exhaust lines are required to be available to support operation of the HPCI turbine.

The only practical means to verify the closure of V23-56 is to pressurize the volume between SSC-23-13 and V23-56 utilizing a differential pressure or leakage type test or by performing a non-intmsive test. The only practical means to verify the closure of V23-65 is to pressurize the volume between SSC-23-12 and V23-65 utilizing a differential pressure or leakage type test or 's p-forming a non-intrusive test.

Additionally, the use of non-intrusive techniques to verify closure on a quarterly basis or during cold shutdowns is not practical. These valves are not located in an easily accessible location and scaffolding is required to setup the necessary equipment.

I NUREG 1482, subsection 4.1.4, states, "The NRC has determined that the need to setup test equipment is adequatejustification to defer backflow testing of a check valve until a refueling outage" These closure function of this valve will be veri 6ed during refueling outages by utihzmg a differential l pressare or leakage type test or by performing a non-intmsive test in accordance with Paragraph 4.3.2.2(e), I (h)ofPart 10 of the Code.

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Verm:nt Yarkee Nuclear Power Stati:n Ins:rvice Testing Prcgram  !

REFUELING OUTAGE JUSTIFICATION Number: ROJ-V26, Revision 1 (Sheet 1 of1) i SYSTEM: Reactor Core Isolation Cooling COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord. l V13-38 C 2 G-191174 Sh 1 K-11 l V13-50 C 2 G-191174 Sh 1 K-08 )

l V13-38 is the RCIC Turbine Vacuum Pump Exhaust Line Containment Isolation Valve. This valve is l normally closed during plant operation. This valve has a safety function to open to pass vacuum pump discharge to the toms. Additionally, this valve has a safety function to close for containment isolation.

4 V13-50 is the RCIC Turbine Steam Exhaust Line Containment Isolation Valve. This valve is normally closed during plant operation. This valve has a safety function to open to pass turbine exhaust steam to the 4

toms. Additionally, this valve has a safety function to close for containment isolation. ,

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JUSTIFICATION

It is not practical to verify the closure function of these valves on a quarterly or cold shutdown basis. ,

During normal operation the RCIC steam and vacuum pump exhaust lines are required to be available to l support operation of the RCIC turbine.

The only practical means to verify the closure of V13-38 is to pressurize the volume between SSC-13-10 and V13-38 utilizing a differential pressure or leakage type test or by performing a non-intmsive test. The only practical means to verify the closure of V13-50 is to pressurize the volume between SSC-13-12 and V13-50 utilizing a differential pressure or leakage type test or by performing a non-intrusive test.

Additionally, the use of non-intmsive techniques to verify closure on a quarterly basis or during cold  !

shutdowns is not practical. These valves are not located ir. an easily accessible location and scaffolding is i required to setup the necessary equipment.  ;

NUREG 1482, subsection 4.1.4, states, "The NRC has determined that the need to setup test equipment is adequate justification to defer backflow testing of a check valve until a refueling outage".

These closure function of this valve will be verified during refueling outages by utilizing a differential pressure or leakage type test or by performing a non-intmsive test in accordance with Paragraph 4.3.2.2(e),

(h) ofPart 10 ofthe Code.

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O V REFUELING OUTAGE JUSTIFICATION 1

Number: ROJ-V27, Revision 0 (Sheet 1 of1) l SYSTEM: Nuclear Boiler j

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COMPONENTS- I l

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

SR-2-14A C NNS G-191167 M-12 SR-2-14B C NNS G-191167 M-12 SR-2-14C C NNS G-191167 M-12 SR-2-14D C NNS G-191167 M-12 4

SR-2-14A through 14D are 3 inch Main Steam Relief Valve (MSRV) exhaust line vacuum breakers. These  ;

valves are normally closed during plant operation. These valves have a safety function to close to direct MSRV discharge flow to the suppression pool in the event of a MSRV actuation.

JUSTIFICATION:

~

It is not practical to verify the closure function of these valves on a quarterly or cold shutdown basis. These r valves are located in the drywell and are not accessible during normal power operations or Cold Shutdowns when the drywell is inerted. These valves can only be tested when the drywell is accessible and the reactor vessel is depressurized.

,1 These valves will be manually exercised closed each refuel outage in accordance with Paragraph 4.3.2.2(e),

(h) ofPart 10 ofthe Code.

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\J REFUELING OUTAGE JUSTIFICATION

Number

ROJ-V28, Revision 0 (Sheet 1 of1)

SYSTEM: FuelPool Cooling & Cleanup Service Water

COMPONENTS

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-244A C 3 G-191173 Sh 2 B-10 3 V70-244B C 3 G-191173 Sh 2 B-10 V70-281 A C 3 G-191159 Sh 1 J-09 j V70-281B C 3 G-191159 Sh 1 B-09 V70-281C C 3 G-191159 Sh 1 J-09

. V70-281D C 3 G-191159 Sh 1 C-09

. V70-51IB C 3 G-191159 Sh 1 H-12 V70-511C C 3 G-191159 Sh 1 H-12

V70-252B C 3 G-191159 Sh 2 G-05 i

V70-252C C 3 G-191159 Sh 2 G-05 These valves are the Service Water System Vacuum Breaker valves. They have a safety function in the

> 3) open position to eliminate vacuum in the service water system and minimize the potential for system water

, hammer upon service water pump restart after station blackout or Appendix R fire scenarios, and in the

closed position to maintain the service water system pressure boundary.  ;

JUSTIFICATION:

) It is not practical to full or part-stroke exercise these valves on a quarterly basis or during cold shutdowns

, due to the significant system and test equipment configurations required. These valves are located in the t

overhead and require installation of scaffolding and test equipment.

NUREG 1482, subsection 4.1.4, states, "The NRC has detemuned that the need to setup test equipment is adequate justification to defer backflow testing of a check valve until a refueling outage". The reasoning used in NUREG 1482, subsection 4.1.4 can also be applied to justify deferred testing for these valves in the ,

forward flow direction. l These opening and closure functions of these valves will be verified using a differential pressure or leakage  !

type test or by performing a non-intmsive test during refueling outages in accordance with Paragraph '

l 4.3.2.2(e), (h) of Part 10 of the Code.

Revision 18 Section 5 Page 140 of 160

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Vermnt Yc-kee Nucirr Pcer Station Inservice Testing Pr: gram i REFUELING OUTAGE JUSTIFICATION O Ntaber: ROJ-V29, Revision C (Sheet 1 of1) l' SYSTEM: Residual Heat Removal Nuclear Boiler

]

COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V10-18A A/C 2 G-191172 F-08 V10-18A5 A/C 2 G-191172 F-08 I V2-39A A/C 2 G-191167 L-04 V2-74A A/C 2 G-191167 D-10  !

V10-18A has a safety function to open to provide overpressure protection for Penetration X-12. V10-18A and associated piping are configured like an typical equalizing line across V10-18. V10-18A also has a safety function to close as one of the inboard containment isolation valves for Penetration X-12.

(Ref. EDCR# 96-416).

V10-18A5 has a safety function to open to provide V10-18 bonnet pressure locking protection. V10 18A5 also has a safety function to close as one of the inboard containment isolation valves for l Penetration X-12. (Ref. EDCR# 96 416, ECN-1).

V2-39A has a safety function to open to provide overpressure protection for Penetration X-41. V2-39A and associated piping are configured like an typical equalizing line across FCV-2-39. V2-39A is j oriented to open anytime pressure on the penetration side of FCV-2-39 is greater than reactor side of FCV-2-39. V2-39A also has a safety function to close for primary containment isolation of penetration X-41. (Ref. EDCR# 96-416).

V2-74A has a safety function to open to provide overpressure protection for Penetration X-8. V2-74A and associated piping are configured like an typical equalizing line across V2-74. V2-74A is oriented to open anytime pressure on the penetration side of V2-74 is greater than reactor side of V2-74. V2-74A also has a safety function to close for pnmary containment isolation of penetration X-8. (Ref.

EDCR# 96-416).  !

i JUSTIFICATION: '

I It is not practical to full or part-stroke exercise these valves on a quarterly basis or during cold shutdowns. These valves are located inside the drywell and are not accessible during power operation. ,

I The stroke open and close function of these valves will be verified each refueling outage in accordance with Paragraph 4.3.2.2(e), (h) of Part 10 of the Code.

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Verm:nt Yankee N: clear P wcr Stati:n Inservice Testing Program REFUELING OUTAGE JUSTIFICATION Number: ROJ-V30, Revision 0 (Sheet I of1)

SYSTEM: Diesel Generator Staning Air i

COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V72-76A C 3 G-191160 Sh 7 H-12 V72-76B C 3 G-191160 Sh 7 H-08 V72-76A and V72-76B, DG Starting Air Compressor Discharge Check Valves, have a safety function to open to provide a flow path from the diesel starting air compressors C-3-1A and C-3-1B to air receiver tanks TK-80-1A, B and TK-80-lC, D. This flow path is required to recharge the diesel generator starting air receiver tanks.

i JUSTIFICATION-It is not practical to full-stroke exercise V72-76A and V72-76B on a quarterly basis or during cold shutdowns. Full-stroke exercising of these valves requires depressurizing the air receiver tanks and i

timing the recovery to normal operating pressure. These valves will be part-stroke open exercised

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quarterly during diesel generator operability testing. The full-stroke open exercise function of these valves will be verified each refueling outage during the Diesel Generator Air Compressor Capacity Test.

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4 Verm::t Ycnkee N clear Pcwer Statim Ins:rvice Testing Program REFUELING OUTAGE JUSTIFICATION Number: ROJ-V31, Revision 0 (Sheet I of 2)  !

SYSTEM: Resdiual Heat Removal, Reactor Water Cleanup, Reactor Core Isolation Cooling and High  !

Pressure Coolant Injection COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord. ,

i V10-19A C 2 G-191172 L-05

V10-19B C 2 G-191172 L-12 V10-19C C 2 G-191172 J-05 V10-19D C 2 G-191172 J-12 i V12-28A C 3 G-191178 Sh 1 D-08 V12-28B C 3 G-191178 Sh 1 G-08 V12-62 C 2 G-191178 Sh 1 B-05 V12-62A C 3 G-191178 Sh 1 B-05 V13-29 C 2 G-191174 Sh 1 J-10 V13-40 C 2 G-191174 Sh 1 N-ll

V13-817 C 2 G-191174 Sh 1 K-07 V13-818 C 2 G-191174 Sh 1 K-07 SSC-23-13 C 2 G-191169 Sh 1 J-04 i V23-56 C 2 G-191169 Sh 1 J-05 l

V23-61 C 2 G-191169 Sh 1 L-06 V23-62 C 2 G-191169 Sh 1 J-10 V23-842 C 2 G-191169 Sh 1 I-02 V23-843 C 2 G-191169 Sh 1 I-03 i V78-2 C 3 G-191162 Sh 2 E-02 j Valves V10-19A through D are RHR pump. They have a safety function in the open position to pass RHR j flow to the suppression pool for pump protection. They have a safety function in the close position to

prevent backflow through an idle RHR pump.

Valves V12-28A and V12-28B are RWCU pump discharge check valves. They have a safety function in

! the closed position to prevent reverse flow in the event of a HELB upstream of the RWCU pumps.

Valves V12-62 and V12-62A are RWCU flow to Feedwater system isolation check valves. They have a

safety function in the closed position to prevent gross diversion of RCIC flow and prevent reverse flow in the event ofa HELB upstream. l Valve V13-29 is RCIC pump mmunum flow recirculation line check valve. It has a safety function in the closed psition to provide primary containment isolation and a safety function in the open position to pass RCIC flow to the suppression pool for pump protection.

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Vermrt Yankee N clear Pcwcr Stati:n Inservica Testing Progrcm

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, REFUELING OUTAGE JUSTIFICATION V

Number: ROJ-V31, Revision 0 (Sheet 2 of2)

Valve V13-40 is RCIC pump suction valve from the suppression pool. It has a safety function in the open position to pass flow from the suppression pool to the RCIC pump suction. This flow path is requied after the CST supply has been exhausted.

Valves V13-817 and V13-818 are RCIC turbine exhaust line vacuum breakers. They have a safety function in the closed position to provide primary containment isolation and a safety function in the open position to prevent water hammer on RCIC turbine exhaust line check SSC-13-9.

Valves SSC-23-13 and V23-56 are HPCI turbine exhaust drain check valves. They have a safety function  !

in the closed position to provide primary containment isolation and a safety function in the open position to l

pass condensate in the HPCI turbine exhaust line drain pot to the suppression pool.

1 Valve V23-61 is HPCI pump suction valve from the suppression pool. It has a safety function in the open i position to pass flow from the suppression pool to the HPCI pump suction. This flow path is requied after the CST supply has been exhausted.

l Valve V23-62 is HPCI pump miminum flow recirculat:on line check valve. It has the closed position to  !

provide primary containment isolation and a safety function in the open position to pass HPCI flow to the  !

suppression pool for pump protection.

U Valves V23-842 and V23-843 are HPCI turbine exhaust line vacuum breakers. They have a safety function in the closed position to provide pnmary contamment isolation and a safety function in the open position to prevent water hammer on HPCI turbine exhaust line check SSC-23-12.

l Valve V78-2 is the Diesel Fuel Oil Storage Tank 511 line check valve. It has a safety function in the close position to prevent loss of Diesel Fuel Oil Storage Tank contents int he event of failure of the NNS fill line.

JUSTTFICATION:

Valves V10-19A, V10-19B, V10-19C, V10-19D, V13-29, V13-40 SSC-23-13, V23-56, V23-61 and V23-62 cannot be verified to exercise full open during normal (power) operation due to the lack of instrumentation for positive test results.

Valves V12-62,V12-62A, V13-817, V13-818, V23-842, V23-843 and V78-2 cannot be indisidually exercised closed during normal (power) operation because there are two check valves in series without test connections.

Valves V12-28A and V12-28B cannot be verified to exercise closed during normal (power) operation due to the lack ofinstrumentation for positive test results.

Each of these valve will be disassembled and inspected each refueling outage in accordance with Paragraph 4.3.2.4(c) of the Par t10 ofthe Code.

Revision 18 Section 5 Page 144 of 160

Vermont Ycnkee Nuclear Pcwer Statin Inservice Testing Program REFUELING OUTAGE JUSTIFICATION Number: ROJ-V18, Revision 1 (Sheet 1 of1)

SYSTEM: ServiceWater COMPONENTS:

I Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

, V70-187A B 3 G-191159 Sh 1 J-06 V70-187B B 3 G-191159 Sh 1 B-06 These valves are the Service Water Pump Gland Seal Supply Isolation Valves. The safety function of these valves is to close in order to provide a backup leakage barrier for the Service Water Supply Header Isolation Valves (V70-13A and V70-13B) while the Service Water System is in the Alternate Cooling mode of operation.

JUSTIFICATION:

It is impracticable to full or part-stroke exercise these valves on a quarterly basis or during cold shutdowns.

The closure of either one of these valves will intermpt cooling flow to two of the Service Water Pump gland seals. Operation of the service water pumps without gland seal flow is not recommended. The Service Water Pumps are required to be operable during power operation and cold shutdowns for cooling to otherloads.

These valves will be full-stroke exercised closed in accordance with Paragraph 4.2.1.2(e), (h) of Part 10 of the Code during each refueling outage when one loop of service water (two service water pumps) can be secured.

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Verm:nt Ycnkee Nuclear Pcwer Stati:n Inservice Testing Program RELIEF REOUEST Number: RR-V01, Revision 1 (Sheet 1 of1)

SYSTEM: ServiceWater l

COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

SE-70-4A B 3 G-191159 Sh 1 K-11 SE-70-4B B 3 G-191159 Sh 1 K-11 SE-70-4C B 3 ,,

G-191159 Sh 1 K-11 SE-70-4D B 3 G-191159 Sh 1 K-11 These valves are the Residual Heat Removal Service Water (RHRSW) pump motor cooling coil supply isolation valves. They have a safety function in the open position to provide cooling for the RHRSW pump motor during operation.

EXAM OR TEST CATEGORY:

Category B.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Verm:nt Yenkee Nuclear Psw:r Stati:n _

Inservice Testing Program RELIEF PJEOUEST Number: RR-V02, Revision 1 (Sheet 1 of1)

SYSTEM: ServiceWater COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-1 A C 3 G-191159 Sh 1 C-03 V70-1B C 3 G-191159 Sh 1 B-03 V70-lC C 3 G-191159 Sh 1 I-03 V70-1D C 3 G-191159 Sh 1 H-03 These valves are the station Service Water pump discharge check valves. They have a safety function in the open position to provide cooling water to systems and equipment required to operate under accident conditions and to provide an inexhaustible supply of water for standby coolant system operation. They have a function in the closed position to prevent the diversion of cooling water through an idle station Service Water pump.

EXAM OR TEST CATEGORY:

Category C.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Vcrm:nt Ycnkee Nuclear Power Stati:n Ins::rvice Testing Prcgrcm 4

RELIEF REOUEST Number: RR-V03, Revision 1 (Sheet 1 of2)

SYSTEM: Diesel Generator Starting Air COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

l AS-24-1A B 3 G-191160 Sh 7 B-16 l AS-24-1B B 3 G-191160 Sh 7 B-03 l AS-24-2A B 3 G-191160 Sh 7 B-17 AS-24-2B B 3 G-191160 Sh 7 B-03 l AV-24-1 A B 3 G-191160 Sh 7 B-17 l AV-24-1B B 3 G-191160 Sh 7 B-03 l l

These valves are the Emergency Diesel Generator (EDG) startmg air inlet and vent valves. They have safety functions, as a set, to provide starting air to the EDG, to prevent EDG and/or piping damage after 4 the EDG starts, and to prevent inadvertent EDG starts.

EXAM OR TEST CATEGORY:

Category B.

CODE REOUIREMENT: Part 10 Para. 4.2.1.1 " Exercising Test Frequency"

" Active Category A and B valves 2 'l be tested nominally every 3 months, except as provided by paras. 4.2.1.2,4.2.1.3, r.nd 4.2.1.7."

Para. 4.2.1.3 " Valve Obturator Movement" "The necessa y valve obturator movement shall be determined by exercising the valve while observing an appropriate indicator, such as indicating lights which signal the required change of obturator position, or by observing other evidence, such as changes in system pressure, flow rate, level, or temperature, which reflect change ofobturator position."

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Verm:nt Yankee Nuclear Pcwcr Stati:n Inservice Testing Program RELIEF REOUEST

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Number: RR-V03, Revision 1 (Sheet 2 of 2)

REOUEST FOR RELIEF: I Reliefis requested on the basis that compliance with the Code requirements is impractical and that the proposed alternatives would provide an acceptable level ofgrJty and safety.

These valves do not have remote position indication. Measunng the stroke time of these valves by  ;

observing stem travel would require disassembly of the operator.

Testing of the inlet valves individually would require the lifting of the power leads to the other valve. Since the stroke timing of these valves is performed by the indirect indication of the respective EDG start time, f o lift leads each quaner and perform the necessary EDG fast starts to verify each valve's stroke time wonid -

be an undue hardship. Because excessive EDG fast stans are a known contributor to decreased EDG reliability and owing to the criticality of the EDGs as part of the ECCS system, the overall impact of testing these valves in accordance with Code requirements would be an overall decrease in plant safety.

Funhermore, since the air stan system is not totally redundant (e.g. they share common piping, i components and initiating logic), testing of these valves individually on a quanerly basis would not increase the quality and safety ofthe system. 4 ALTERNATE METHOD:

During EDG slow start testing perfonned each month, indirect indication that at least one of the two parallel air stan inlet valves opens, and the vent valve closes, will be performed by ensunng the EDG stans.

During EDG fast start testing performed every six months, indirect measurement that at least one of the two parallel air start inlet valves opens promptly, and the vent valve closes promptly, will be performed by ensunng the EDG starts within the Technical Specification limit of 13 seconds. Measunny the EDG start time gives indication of possible valve degradation (as a pair) since any significant changes in valve stroke ,

time will be identified by longer than normal EDG stan times.

In addition, to funher assess the operational readiness of each air start inlet valve, an independent l operability test is performed once per operating cycle. This test will be accomplished by altemately lifling i the power leads to one of the two air stan valves, and then measuring the EDG fast start time with the l remaining valvein operation. I USNRC EVALUATION STATUS:

Reliefwas granted in the March 1994 SER [ Reference (u)] for ReliefRequest RR-V03, Revision 1.

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1 Vermant Ycnkee Nuclear Pcwer Stati:n Inservice Testing Przgram l

9 RELIEF REOUEST

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Number: RR-V04, Revision 1 (Sheet 1 of1)

SYSTEM: FuelOilTransfer COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V78-2 C 3 G-191162 Sh 2 E-02 Valve V78-2 is the fuel oil storage tank fill line check valve. It has a safety function in the close position to ensure fuel oil system integrity upon loss of the non-safety related fill line piping.

EXAM OR TEST CATEGORD Category C.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal  ;

SER's. 1 O

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Verm:nt Ycnkee Nucl:ar Pcwcr Stati:n Inservice Testing Prcgram l

e RELIEF REOUEST I

,. I Number: RR-VOS, Revision 1(Sheet 1 of1) i SYSTEM: NuclearBoiler J COMPONENTS:

Valve Number Oidpt. Safety Class Drawing Number Dwg. Coord.

RV-2-71A B/C 1 G-191167 D-08 ,

RV-2-71B B/C G-191167 1 G-08 )

RV-2-71C B/C 1 G-191167 G-08 RV-2-70A C 1 G-191167 D-08 l RV-2-70B C 1 G-191167- D-08

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Valves SV-2-70A and SV 2-70B are the main steam safety valves. They have a safety function to prevent over-pressurization of the mactor coolant system. j Valves RV-2-71 A through D are the main steam dual function reliefvalves. They have a safety function to i prevent over-pressurizadon of the reactor coolant system and to function as part of the reactor coolant Automatic Depressuhation System (ADS).

j; EXAM OR TEST CATEGORY:

Category C

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THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's. i i

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Vcrm:nt Yrnkee Nuclear P wer Stati:n Inservice Testing Pr: gram RELIEF REOUEST Number: RR-V06, Revision 0 (Sheet 1 of1)

SYSTEM: High Pressure Coolant Injection l COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

HPCI-CONTROL B 2 G-191169 Sh 2 F-13 This valve is the High Pressure Coolant Injection (HPCI) turbine steam inlet Control valve. It has a safety function to operate to provide and regulate steam to the HPCI turbine.

EXAM OR TEST CATEGORY:

l Category B.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 18 OF THE ET PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Vcrm:ct Ycakee Nuclear Pcw;r Stati:n Inservice Testing Pr: gram 7 RELIEF REOUEST l

Number: RR-V07, Rewisi >n 2 (Sheet 1 of1)

SYSTEM: ControlRod Drive Hydraulic COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V3-162A A/C 2 G-191170 A-09 V3-162B A/C 2 G-191170 A-01 These valves are the Control Rod Drive (CRD) scram discharge volume vent check valves. They have a safety function in the closed position to isolate the scram discharge volume during a scram condition, thereby preventing reactor coolant inventory loss.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 15 OF THE IST PROGRAM.

This RR number is being maintained for tracee8ty to the Third Interval IST Program Submittal SER's.

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Verm:nt Ycnkee Nucirr Pcwcr Statin Inservice Testing Pregram q

> RELIEF REOUEST Number: RR-V08, Revision 0 (Sheet 1 of2)

SYSTEM: StandbyLiquid Control COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

SR-ll-39A C 2 G-191171 G-08 SR-11-39B C 2 G-191171 K-08 l

Valves SR-11-39A & B are the Standby Liquid Control (SLC) pump discharge relief valves. They have a safety function to operate to prevent overpressurization of the SLC system.

EXAM OR TEST CATEGORY:

Category C CODE REOUIREMENT: Part 1 l Para.1.3.4.l(b) " Subsequent 10 Year Periods" O "All valves of each type and manufacturer shall be tested within each subsequent 10 year period, with a minimum of 20% of the valves tested within 48 months. This 20% shall be previously untested valves, ifthey exist." l Para.1.3.4.1(e)(2)" Valves Not Meeting Acceptance Criteria" 4 "Any valve exceeding its stamped set pressure by 3% or greater shall be repaired or replaced, the cause of failure shall be determined ar.d corrected, and the vah e shall successfully pass a retest before it is retumed to senice.

REOUEST FOR RELIEF:

Reliefis requested on the basis that the proposed altematives would provide an acceptable level of quality and safety.

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Vcrm:nt Ycnken Nucirr Pcwer Stiti:n Ins:rvice Testing Progrcm l

o RELIEF REOUEST Number

RR-V08, Revision 0 (Sheet 2 of 2)

REOUEST FOR RELIEF (CONT.):

Vermont Yankee Technical Specifications 4.4.A requires testing of valves SR-11-39A & B at least once every operating cycle. This testing frequency represents a six-fold increase over the testing frequency required by Part 1 of the Code and, as such, yields more accurate valve degradation trending data. The testing is currently performed at the Vermont Yankee plant site. I The acceptable range ofreliefvalve actuation given in Technical Specifications 4.4.A is 1400 psig to 1490 psig. The 1400 psig value ensures that a sufficient injection pressure can be established prior to lifting of the relief valve. The 1490 psig value ensures relief valve actuation prior to reaching the system design pressure of 1500 psig. 1490 psig translates into a maximum relief valve setpoint of 99% of system design pressure as opposed to 110% allowable by the piping code. Applying the Part I tolerance of 3% to the current setpoint would unnecessarily reduce the acceptable range to 1400 psig to 1442 psig (1442 psig =

96% ofsystem design pressure).

The increased testing frequency and present setpoint requirements provide adequate assurance of the operational readiness of valves SR-11-39A & B and, as such, no significant increase in the level of safety or quality can be expected if the subject Code requirements are imposed.

/~T U ALTERNATE METHOD: i In accordance with Vermont Yankee Technical Specifications 4.4.A, valves SR-1139A & B shall be tested at least once every operating cycle. The setting of the valves shall be between 1400 and 1490 psig.

USNRC EVALUATION STATUS:

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Reliefwas granted in the September 1993 SER [ Reference (s)] for ReliefRequest RR-V08, Revision 0.

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Verm:nt Ycnkee Nucle:r Pcwer Stati:n Inservice Testing Prcgram RELIEF REOUEST 1

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Number: RR-V09, Revision 1 (Sheet 1 of1)

SYSTEM: ResidualHeatRemoval COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V10-17 A 2 G-191172 D-06 V10-27A A 2 G-191172 D-06 V10-27B A 2 G-191172 D-12 )

Valve V10-17 is the Residual Heat Removal (RHR) shutdown cooling supply outboard primary containment isolation valve. The valve has a safety function in the closed position to provide primary containment and pressure isolation, and in the open position to provide RHR pump suction during j

shutdown cooling operation.

Valves V10-27A & B are the RHR system low pressure coolant inje<aon primary containment isolation valves. These valves have a safety function in the closed position to provide primary containment and pressure isolation, and in the open position to pass low pressure coolant injection water to the reactor.

( USNRC EVALUATION STATUS:

' TIUS RELIEF REQUEST WAS WITHDRAWN IN REVISION 16 OF THE IST PROGRAM PLAN.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Vermrtt Ycnkee Nucl:cr P wer Stati:n Its:rvice Testing Pr: gram s RELIEF REOUEST s

Number: RR-V10, Revision 0 (Sheet 1 of1)

SYSTEM: Residual Heat Removal Service Water COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V10-89A B 3 G-191179 Sh 2 G-01 V10-89B B 3 G-191179 Sh 2 C-01 I

These valves are the Residual Heat Removal Service Water (RHRSW) flow control valves for the RHR heat exchangers. They have a safety function to provide and regulate cooling water to the RHR heat exchangers.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 15 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal SER's.

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Verm: t Ycnkee Nuclear Pcw:r Stcti n Inservice Testing Pr: gram RELIEF REOUEST Number: RR-Vil, Revision 0 (Sheet 1 of1) l SYSTEM: Reactor Core Isolation Cooling COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

RCIC-CONTROL B 2 G-191174 Sh 2 D-12 This valve is the Reactor Core Isolation Cooling (RCIC) turbine governor control valve. This valve has a l

safety function to operate to provide and regulate steam to the RCIC turbine.

EXAM OR TEST CATEGORY-  !

I Category B.

THIS RELIEF REQUEST WAS WITHDRAWN IN REVISION 18 OF THE IST PROGRAM.

This RR number is being maintained for traceability to the Third Interval IST Program Submittal A SER's.

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Vcrm:nt Ycnkee Nuclear Pcwer Statio: Eservice Testing Program RELIEF REOUEST O-Number: RR-V12, Revision 0 (Sheet 1 of3)

SYSTEM: ServiceWaterSystem COMPONENTS:

Valve Number OM Cat. Safety Class Drawing Number Dwg. Coord.

V70-43A C 3 G-191159 Sh 1 J-12 V70-43B C 3 G-191159 Sh 1 B-12 These valves are the Service Water System header discharge check valves. These valves have a safety function to close to prevent the backflow of RHRSW pump discharge to the suction of the pumps when operating in the Alternate Cooling mode ofoperation.

EXAM OR TEST CATEGOR11 Category C CODE REOUIREMENT: Part 10 O para. 4.3.2.1 " Exercising Tests for Check Valves"

" Check valves shall be exercised nominally every 3 months, except as provided by paras.4.3.2.2, 4.3.2.3, l 4.3.2.4 and 4.3.2.5.

i REOUEST FOR RELIEF: <

Reliefis requested on the basis that compliance with the Code requirements is impracticable and that the proposed altematives would provide an acceptable level of quality and safety.

It is impractical to full or part-stroke exercise these valves in the' closed direction on a quarterly, cold shutdown or refueling outage basis. Closure of these valves would require shutdown of each of the Senice Water Pumps in their individual trains when the Senice Water System is required to supply cooling water to core standby cooling equipment and the emergency diesel generators.

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Vermut Yc kee N clear Pcuer Statim Inservica Testing Program RELIEF REOUEST Number: RR-V12, Revision 0 (Sheet 2 of3)

ALTERNATE TEST METHOD:

These valves are 8 inch swing check valves p?+he same design, manufacturer, size, model and materials of construction. Additionally, these valvs are bd oriented in the horizontal position and see similar service conditions. These valves will be partially disassembled, inspected and nuunally exercised on a sampling basis (one valve per refueling outage) when the Service Water System flow to the required loads can be isolated in accordance with NRC StaffPosition 2 identiSed in NRC Generic Letter 89-04. During the valve disassembly, the internals of the valve will be verified to be stmeturally sound (no loose or corroded parts).

If the disassembled valve is not capable of being full-stroke exercised or there is binding or failure of the valve intemals, the other valve will also be disassembled, inspected and manually exercised during the same outage.

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Vcrm::t Ycnkee Nuclear P:w r Statina Inservics Testing Program f- RELIEF REOUEST

'%)s Number: RR-V12, Revision 0 (Sheet 3 of 3)

USNRC EVALUATION STATUS j

l This reliefrequest was approved in for use in USNRC Generic 89-04, NRC StaffPosition 2 i 1

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