ML20083Q350

Inservice Insp & Testing Program,Second 10-Yr Program,Quad Cities Nuclear Power Station Units 1 & 2

ML20083Q350
Person / Time
Site:	Quad Cities
Issue date:	08/18/1982
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20083Q335	List: ML20083Q332 ML20083Q350
References
PROC-820818, NUDOCS 8302250308
Download: ML20083Q350 (233)

Text

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Ao i-O' INSERVICE INSPECTION AND TESTING PROGRAM QUAD CITIES-NUCLEAR POWER STATION UNITS 1 AND 2 4

COMMONWEALTH EDISON COMPANY O

AUGUST 18, 1982 4

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8302250308 830217 PDR ADOCK 05000254 PDR g

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TABLE OF' CONTENTS PAGE l

1

1.0 INTRODUCTION

l.1 GENERAL ~INFORMATION....................... 1-1 1.2 SYSTEM CLASSIFICATIONS.................... 1-1 2.0 INSERVICE INSPECTION PROGRAM ,

2.1 DESCRIPTION

OF ISI PROGRAM................ 2-1 2.2 PROGRAM TABLES............................ 2-7 A. QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 .

2.3 RELIEF REQUESTS........................... 2-8 3.0 INSERVICE TESTING PROGRAM FOR SNUBBERS O

3.1 DESCRIPTION

OF SNUBBER IST PROGRAM........ 3-1 3.2 PROGRAM TABLES............................ 3-3

'A. QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 3 .' 3 RELIEF REQUESTS........................... 3-6 .

4.0 INSERVICE TESTING PROGRAM FOR PUMPS 4

4.1 DESCRIPTION

OF IST PROGRAM FOR PUMPS...... 4-1 4.2- PROGRAM TABLES............................ 4-2

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< A. QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 4.3 . RELIEF REQUESTS........................... 4-5 O

TABLE OF CONTENTS - continued

~T PAGE 5.0 INSERVICE TESTING PROGRAM FOR VALVES

5.1 DESCRIPTION

OF IST PROGRAM FOR VALVES..... 5-1 5.2 PROGRAM TABLES............................ 5-11 A.- QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 5.3 RELIEF REQUESTS........................... 5-13 5.4 COLD SHUTDOWN JUSTIFICATION FOR IST VALVE PROGRAM................................... 5-41 O

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1.0 INTRODUCTION

O 1.1 GENERAL INFORMATION ,

The Inservice Inspection (ISI) and Inservice Testing (IST) Programs for Quad Cities Nuclear Power Station, Units 1 and 2 are developed in compliance with the rules and regulations of 10CFR50.55a and Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition including the Addenda through Winter 1980. Where these rules are determined to be impractical, specific relief

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is requested in writing.

The Inservice Inspection and Testing Programs for Class

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1, 2 and 3 Components, Component Supports, Pumps, &

Valves are applicable for the ten year interval beginning February 18, 1983 and March 10, 1983 for Quad Cities Units 1 and 2, respectively. The upcoming ten year interval is the second inspection interval for both Quad Cities Units 1 and 2.

i 1.2 SYSTEM CLASSIFICATION The construction permits for Quad Cities Units 1 and 2 were issued on February 15, 1967. At that time the ASME Boiler and Pressure Vessel Code covered only nuclear O

COM-06-003 Revision 1 1-1

vessels. Piping, pumps, and valves were built primarily to the rules of USAS B31.1.0, therefore, the station has essentially no ASME Code Class 1, 2 or 3 designed systems. The system classifications used as a basis for the Inservice Inspection and Testing Programs are based on the requirements set forth in 10CFR50 and Regulatory Guide 1.26 and were developed for the sole purpose of assigning the appropriate .2nservice inspection requirements. Components within the primary coolant pressure baundary, as defined in 10CFR50.2(v), are

designated as ISI-Class 1 while other safety related components are designated as ISI-Class 2 and 3 in accordance with the guidelines of Regulatory Guide

{} 1.26. Pursuant to 10CFR50 paragraph (g)(1), inservice inspection requirements of Section XI of the ASME Code are then assigned to these components, within the constraints of existing plant design.

Color-coded Piping and Instrument Diagrams (P& ids) documenting the system classifications were developed to aid in the review and implementation of the subject programs. A legend explaining the color-coding scheme is included on the first page of the P& ids. l I

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COM-06-003 Revision 1 1-2

2.0 INSERVICE INSPECTION PROGRAM FOR COMPONENTS O

2.1 PROGRAM DESCRIPTION 2.1.1 The Inservice Inspection Program for ISI Class 1, 2 and 3 components meets the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition through the Winter 1980 Addenda. -Where these requirements are determined to be impractical, specific requests for relief have been written and included in Section 2.3.

2 .1. 2 The ISI Component Program is presented in Section 2.2 in

( ) a tabular format. The components and. associated requirements are listed according to ascending Code Category and Item Numbers. The following information is included in the taales

A. Code Category - The Section XI Examination Categories as defined in Table IWB-2500-1, IWC-2500-1, IWD-2500-1, and IWF-2500-1 for class 1,2, and 3 components.

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B. -Item Number and Item Description - The Item Number and its description as listed in Tables IWB, IWC, IWD-2500-1, and IWF-2500-2. Appilcable Iten l

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COM-06-003 Revision 1 2-1 1

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numbers and all item Descriptions are listed for

~g each Code . Category.

C. Section,XI Exam Required lists the examination method or methods. This reflects the Section XI requirements. The abbreviations used are as follows:

VOL -

Volumetric per IWA-2230 SUR -

Surface per IWA-2220 VT Visual per IWA-2211

.VT Visual per IWA-2212, VT Visual per IWA-2213 J

{} VT Visual per IWA-2214 D. Relief Requests references either a specific relief request contained in Section 2.3 or references one of the Code allowed exemptions listed below. If the latter is referenced, the particular line or component has been exempted from volumetric and/or surface examination by the applicable Code Components exempted from examination by

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paragraph.

Code allowed exemptions will not appear in the component tables of this program in most cases. It should be noted that section 2.3 contains some generic relief requests that are not specifically O-COM-06-003 Revision 1 2-2 i

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referenced in the tables but apply to the ISI

' Program in general. ,

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EX IWB-1220(b), lines 1-inch nominal pipe size (n.p.s.) and less.

EX IWB-1220(a), liquid carrying lines 2-inch n.p.s. and less (see 2.1.3)

EX IWB-1220(a), steam carrying lines 4-inch n.p.s. and less (see 2.1.3)

EX IWC-1220(c), head connections, 2-inches n.p.s. and less, made inaccessible by CRD penetrations EX IWC-1220(b), components not required to operate above a temperature of 200*F or above a pressure of 275 psig.

EX-6 'IWC-1220(c), component connections, piping and associated valves, and vessels and their attachments that are 4 in. n.p.s.

EX IWC-1220(a), lines not required during normal operating conditions but remain flooded under static conditions at l

a minimum of 80% of the pressure they would be subjected to when required to operate.

EX IWC-1230, piping support members and piping support components encased in concrete.

O COM-06-003 Revision 1 2-3

EX IWD-1220.1, integral aivachments of supports and. restraints to components that are 4 in.

n.p.s. and smaller.

EX IWD-1220.2(a), integral attachments of supports and restraints in systems Whose function is not required in support of reactor residual heat removal and emergency core cooling.

EX-ll - IWD-1220.2(b), integral attachments of' supports and restraints Where operating 4

pressure is 275 psig or less and operating temperature is 200*F or less.

EX IWD-5223(e), open ended vent and drain

{) lines from components extending beyond the last shut-off valve and open ended safety or' relief valve discharge lines.-

E. Alternate Exam lists the examination method or-methods that will be performed in lieu of the required Section XI methods When relief has been

. requested.

F. Remarks - lists general clarification remarks.

2.1.3 Pursuant to paragraph IWB-1220(a), the maximum size line break that can be reade up by the reactor coolant makeup

!O COM-06-003 Revision 1 2-4

system has been calculated to be 2.08 inches inside

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\~ diameter for liquid carrying lines and 4.16 inches for 1

steam carrying lines. In applying this exemption to the program, liquid carrying lines less than or equal to 2 inch nominal pipe size and steam carrying lines less than or equal to 4.0 inches n.p.s. were exempted.

2.1.4 Table 2.1-1 lists the applicable Class 1, 2 & 3 systems which are covered in the Inservice Inspection Program.

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TABLE 2.1-1 LIST OF SYSTEMS INDLUDED IN THE ISI PROGRAM SYSTEM CLASS Control Rod Drive 1& 2 Residual Heat Removal-(RHR) 1& 2 RHR Service Water 3 Standby Liquid Control (SBLC) 1& 2 Reactor Water Cleanup 3 Core Spray 1& 2 High Pressure Coolant Injection (HPCI) 1& 2 Main Steam 1 Feedwater 1& 2 Diesel Generator Cooling Water 3 4

COM-06-003 Revision 1 2-6

!O SECTION 2.2 1

TABLES'FOR INSERVICE INSPECTION PROGRAM i 3

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- A. QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 i

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CLASS 1.2 & 3 COMPONENTS OUAD ClllES NUCLEAR POWER STATION UNIT - 1 1 15 CLASS 1 , ,,3_18-82 I C00E ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER REQUIRED REQUESi$ EXAM B.A PRESSURE RETAINING WELDS IN REACTOR VESSEL l D1.10 Shell Welds Bl.11 Circumferential VOL CR-1 81.12 Longitudinal VOL CR-1 I

Bl.20 Head Welds .

B1.21 Circumferential VOL CR-2 D1.22 Heridional VOL CR-2 D1.30 Shell-to-Flange Weld VOL 01.40 Head-to-Flange Weld VOL AND SURF I

, D1.50 Repair Welds N/A 6.8 PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS N/A B.D FULL PENETRATION WELDS OF NO33LES IN VESSELS - INSPECTION PROGRAN B Reactor Vessel 03.90 Nozzle-to-Vessel Welds VOL D3.100 Nozzle Inside Radius Section VOL

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Pressurizer B3.110 Nozzle-to-Vessel Welds N/A B3.120 Nozzle Inside Radius Section N/A Steam Generators (Primary Side)

B3.130 Nozzle-to-Vessel Welds N/A B3.140 Nozzle Inside Radius Section N/A .

Deat Exchangers (Primary Side)

B3.150 Nozzle-to-Vessel Welds N/A B3.160 Nozzle Inside Radius Section N/A B-E PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS B4.10 Partial Penetration Welds EXTERNAL SURPACES B4.11 Vessel Nozzles VT-2 B4.12 Control Rod Drive Nozzles VT-2 B4.13 Instrumentation Nozzles VT-2 Pressurizer B4.20 lleater Penetration Welds N/A B-F PRESSURE RETAINING DISSINILAR NETAL WELDS Reactor Vessel B5.10 Nominal Pipe Size > 4 in.

Nozzle-to-Safe End Butt Welds VOL AND SURF

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CLASS 1.2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 1 ' " **

CLASS 1 ReWlon 1 l Dele 8-18-82 CM ITEM CATEGORY NUMBER ITEM DESCRIPT.'CN REMARedi REGINRED REQUESTS EXAM B5.11 Nominal Pipe Size < 4 in.

Nozzle-to-Safe End Butt Welds SURF F5.12 Nozzle-to-Safo End Socket Welds N/A Pressurizer B5.20 Nominal Pipe Size > 4 in.

Nozzle-to-Sate End Butt Welds N/A B5.21 Nominal Pipe Size < 4 in, Nozzle-to-Safe End Butt Welds N/A, B5.22 Nozzle-to-Safe Er.d Socket Welds N/A Steam Generator B5.30 Nominal Pipe Sise 1 4 in.

Nozzle-to-Safe End Butt Welds N/A B5.31 Nominal Pipe Size < *. in.

Nozzle-to-Safe End Butt Welds N/A B5.32 Nozzle-to-Safe End Socket Welds s N/A Neat Exchangers B5.40 Nominal Pipe Size > 4 in.

Nozzle-to-Safe End Butt Welds N/A B5.41 Nominal Pipe Size < 4 in.

Nozzle-to-Safe End Butt Welds N/A B5.42 Nozzle-to-Safe End Socket Welds N/A Piping B5.50 Nominal Pipe Size 1 4 in.

Dissimilar Netal Butt Welds VOL AND SURF SEE FIG. IWB-2500-8

a Commonwealth Ed; son ISI -

CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASSi "*" * " I' Revision 1 lpeie-18-82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMSER ITEM DESCRIPTM REQUIRED REQUESTS EXAM B5.51 Nominal Pipe Size < 4 in.

Dissimilar Netal Butt Welds SURP B5.52 Dissimilar Metal Socket Welds N/A D-G-1 PRESSURE RETAINING BOLTING, GREATER TNAN 2 in. IN DIAMETER Reactor Vessel

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B6.10 Closure llead Nuts SURF B6.20 Closure Studs, in place VOL B6.30 Closure Studs, when removed SURF AND VOL B6.40 Threads in Flange VOL B6.50 Closure Washers, Bushings VT-1 l Pressurizer B6.60 Bolts and Studs "

N/A I

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B6.70 Flange Surface, when connection disassembled N/A B6.80 Nuts, Bushings, and Washers N/A Steam Generators B6.90 Bolts and Studs N/A B6.100 Flange Surface, when connection disassembled N/A B6.110 Nuts, Bushings, and Washers N/A Neat Exchangers B6.120 Bolts and Studs N/A O

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CLASS 1 R sekm i l Dene 8-18-82 CODE ITEM l SECT XI EXAM RELIEF CATEGORY NUMBER ITEM DESCRINN RE01NRED REQUESTS l ALTERNATE EXAM REMS..

B6.130 Flange Surface, when connection disassembled N/A B6.140 Nuts, Bushings, and Washers N/A Piping D6.150 Bolts and Studs N/A D6.160 Flange Surface, when connection -

disassembled N/A B6.170 Nuts, Bushings, and Washers N/A Pumps B6.180 Bolts and Studs VOL B6.190 Flange Surface, when connection disassembled VT-1 .

B6.200 Nuts, Bushings, and Washers VT-1 1

Valves B6.210 Bolts and Studs N/A D6.220 Flange Surface, when connection disassembled N/A B6.230 Nuts, Bushings, and Washers N/A B-G-2 PRESSURE RETAINING BOLTING, 2 in. AND LESS IN DIAMETER Reactor Vessel D7.10 Bolts, Studs, and Nuts N/A

O Commonwealth Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER ' STATION UNIT - 1 CLASS 1

' " 15 wason i l pet. 8-18-82 C00E ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER REQUIRED REQUESTS EXAM R MARKS Pressurizer B7.20 Bolts, Studs, and Nuts Ff/A Steam Generators B7.30 Bolts, Studs, and Nuts N/A Heat Exchangers

• B7.40 Bolts, Studs, and Nuts N/A Piping B7.50 Bolts, Studs, and Nuts VT-1 ,

1 Pumps B7.60 Bolts, Studs, and Nuts N/A Valves .

B7.70 Bolts, Studs, and Nuts VT-1 CRD Housings B7.80 Bolts, Studs, and Nuts VT-1 WHEN DISASSEMBLED t

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B-H INTEGRAL ATTACHMENTS FOR VESSELS Reactor Vessel D0.10 Integrally Welded Attachments SURF VESSEL SUPPORT SKIRTS SEE FIG. IWB-2500-13 LIFTING LUGS, STABILI2ER LUGS SEE FIG. IWB-2500-15 l

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@ Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION l UNIT - 1 CLASS 1

. 1 s.8-18-82 CODE ITEM ITEM DESCRIPTION

' A j CATEGORY NUMBER REQUIRED REQUESTS EXAM REMARKS l

Pressurizer D8.20 Integrally Welded Attachments N/A Steam Generator B8.30 Integrally Welded Attachments N/A Heat Exchangers 88.40 Integrally Welded Attachments N/A, B-J PRESSURE RETAINING WELDS IN PIPING B9.10 Nominal Pipe Size > 4 in.

89.11 Circumferential Welds SURF AND VOL CR-3,CR-4 B9.12 Iongitudinal Welds SURF AND VOL CR-4 B9.20 Nominal Pipe Size < 4 in.

B9.21 Circumferential Welds SURP 09.22 tongitudinal Welds SURF 89.30 Branch Pipe Connection Welds 89.31 Nominal Pipe Size > 4 in. SURF AND VOL CR-5 VT-2 B9.32 Nominal Pipe Size < 4 in. SURF D9.40 Socket Welds N/A i

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@ Commonwealth Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASS 1 "* * # I' Rwason 1 lpete 8-18-82 CODE ITEM #

CATEGORY NUMBER ITEM DESCRIPTION REMARKS REOutRED REQUESTS EXAM D-K-1 INTEGRAI. ATTACHMENTS FOR PIPING, PUMPS AND VALVES Piping B10.10 Integrally Welded Attachments SURF SEE PIG. IWB-2500-13,15 Pumps -

B10.20 Integrally Welded Attachments SURF SEE FIG. IWB-2500-13,15 Valves B10.30 Integrally Welded Attachments SURF SEE FIG. IWB-2500-13,15 0)-L-1, PRESSURE RETAINING WELDS IN D-N-1 PUMP CASINGS AND VALVE BODIES N/A D-L-2, PUMP CASINGS AND VALVE BODIES B-M-2 D12.20 Pump Casing VT-3 CR-6 WHEN DISASSEMBLED D12.40 Valve Body, Exceeding 4 in.

Nominal Pipe Size VT-3 CR-7 WHEN DISASSEMBLED B-N-1 INTERIOR OF REAC'!OR VESSEL Reactor Vessel ,

B13.10 Vessel Interior VT-3 PER TABLE IWB-2500-1, CATEGORY B-N-1

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@ Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASS 1 "* * " *5 Revision 1 l Dele 8-18-82 CODE ITEM SECT XI EXAM REUEF ALTERNATE CATEGORY NUMBER M DESCRIPMN RE N REQUIRED REQUESTS EXAM Reactor Vessel (BNR)

B13.20 Interior Attachments VT-1 ACCESSIBLE NELDS B13.21 Core Support Structure VT-1 ACCESSIBLE SURFACES Reactor vessel (IWR)

B13.30 Core Support Structure N/A B-O PRESSURE RETAINING WELDS IN CONTROL ROD HOUSINGS Reactor Vessel B14.10 Nelds in CRD Housing VOL EX-2 EX-2 APPLIED TO THESE HOUSINGS DUE M CONFIGURATIOh B-P ALL PRESSURE RETAINING COMPOllENTS Reactor Vessel

! B15.10 Pressure Retaining Boundary VT-2 LEAKAGE TEST B15.ll Pressure Retaining Boundary VT-2 HYDROSTATIC TEST l

l Pressurizer D15.20 Pressure Retaining Boundary N/A B15.21 Pressure Retaining Boundary N/A.

l Steam Generators B15.30 Pressure Retaining Boundary 'N/A B15.31 Pressure Retaining Doundary N/A .

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@ Edison ISI -

CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASS 1

"* ** " 15 Rohion j lDete 8 18 82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY I NUMBER REOUIRED REQUESTS EXAM Heat Exchangers B15.40 Pressure Retaining Boundary N/A B15.41 Pressure Retaining Boundary N/A Piping B15.50 Pressure Retaining Boundary VT-2' LEAKAGE TEST B15.51 Pressure Retaining Boundary VT-2 HYDROSTATIC TEST ,

Pumps

B15.60 Pressure Retaining Boundary VT-2 LEAKAGE TEST B15.61 Pressure Retaining Boundary VT-2 HYDROSTATIC TEST Valves B15.70 Pressure Retaining Boundary VT,-2 LEAKAGE TEST B15.71 Pressure Retaining Boundary VT-2 HYDROSTATIC TEST B-O STEAM GENERATOR TUBING N/A l

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• ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION 11 d "

UNIT - 1 CLASS 2 , lo,, 8-18-82 CODE ITEM SECT XI EXAM REUEF ALTERNATE l CATEGORY NUMBER I DEMW REMARKS REQUIRED REQUESTS EXAM C-A PRESSURE RETAINING WELDS IN PRESSURE VESSELS C1.10 Shell Circumferential Welds VOL RHR HEAT EXCHANGER C1.20 Head Circumferential Welds VOL RHR HEAT EXCHANGER C1.30 Tubesheet-to-Shell Weld N/A C-B PRESSURE RELIMING NOZZLE WELDS IN VESSELS l

C2.10 Nozzles in Vessel i h in.

Nominal Thickness SURP CR-8 VT-2 RHR HEAT EXCHANGER C2.20 Nozzles in Vessel > h IN. .

Nominal Thickness N/A C-C INTEGRAL ATTACHMENTS FOR VESSELS, PIPING, PLMPS, AND VALVES Pressure Vessels C3.10 Integrally Welded Attachments SURP RHR HEAT EXCHANGER Piping C3.40 Integrally Welded Attachments SURP

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"* l' " '*5 Msion j l Dele 8-18-82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGGRY NUMBER DESCR N N EMS REOutRED REQUESTS EXAM Pumps l C3.70 Integrally Welded Attachments N/A

Valves C3.100 Integrally Welded Attachments '

N/A C-D PRESSURE RETAINING BOLTING GREATER THAN 2 in. IN DIANETER Pressure Vessels C4.10 Bolts and Studs

• N/A Piping C4.20 Bolts and Studs N/A Pumps C4.30 Bolts and Studs N/A Valves C4.40 Bolts and Studs N/A C-F PRESSURE RETAINING WELDS IN PIPING C5.10 Piping Welds < 14 in.

Nominal Wall Thickness C5.ll Circumferential Weld SURF C5.12 tongitudinal Weld SURP

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@ Edison ISI - CLASG 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION Page 13 et 15 UNIT - 1 CLASS 2

• sont l0 I 8-18-82 CODE ITEM '

CATEGORY NUMBER ITEM DESCRIPTION REMARKS REOUIRED REQUESTS EXAM C5.20 Piping Welds >l/2 in.

Nominal Wall Thickness C5.21 circumferential Wold SURF AND VOL C5.22 tongitudinal Weld N/A C5.30 Pipe Branch Connections C5.31 Circumferential Weld hDRP CR-10 VT-2 -

C5.32 tangit** *i.nal Weld N/A C-G PRESE'LE RETAINING WELDS IN PUNPS AND VALVES N/A C-Il ALL PRESSURE RETAINING CONPONENTS Pressure Vessels C7.10 Pressure Retaining Components VT-2 PRESSURE TEST-C7.11 Pressure Retaining Components VT-2 HYDROSTATIC TEST Piping C7.20 Pressure Retaining Components VT-2 PRESSURE TEST C7.21 Pressure Retaining Components VT-2 HYDROSTATIC TEST Pumps C7.30 Pressure Retaining Components VT-2 PRESSURE TEST C7.31 Pressure Retaining Components VT-2 HYDROSTATIC TEST

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@ Edison-ISI , CLASS .1.2 & 3 COMPOifENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASS 2

"* " 15 Revision 1 l o l. 6-16-62 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER TEM DESCRIPM REMARKS REQUIRED REQUESTS EXAM Valves C7.40 Pressure Retaining Components -2 PRCSSURE TEST C7.41 Pressure Retaining Components VT-2 I:YDROSTATIC TEST 9

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@ Edison ' ISl -- CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 1 Page 15 of 15 CLASS 3 Revisian lDet4-18-82 CODE ITEM '

CATEGORY NUMBER n DESCRINN SECT X! EXAM REUEF ALTERNATE RE M G REQUIRED REQUESTS EY.AM D-A SYSTEMS IN SUPPORT OF REACTOR Sil0TDOWN FUNCTION N/A D-B SYSTCMS IN SUPPORT OF ENERGENCY CORE COOLING, CONTAIHMENT BEAT REMOVAL, ATMOSPHERE CLEANUP, AND REACTM

• RESIDUAL HEAT REMOVAL ,

D2.10 Pressure Retaining Components VT-2 FUNCTIONAL TEST VT-2 HYDROSTATIC TEST D2.20 Integral Attachment

- Component Supports and Restraints VT-3 D2.30 Integral Attachment

- Mechanical and Hydraulic dtubbers N/A D2.40 Integral Attachment

- Spring Type Supports N/A D2.50 Integral Attachment

- Constant Load Type Supports N/A D2.60 Integral Attachment *

- Shock Absorbers '

N/A D-C SYSTEMS IN SUPPORT OF RESIDUAL HEAT REMOVAL FROM SPENT FUEL STORAGE POOL N/A

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[ Commonwealth Edison l ISI - CLASS 1,2 & 3 COMPONENTS

. QUAD CITIES NUCLEAR POWER STATION Ur41T - 2

• l CLASS 1 ,, j 82 CODE ITEW SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER "

REQUIRED REQUESTS EKAM B.A PRESSURE RETAINING WP".DS IN REACTOR VESSEL i Bl.10 Shell Welds B1.11 Circumferential VOL CR-1 Bl.12 Longitudinal VOL CR-1 Bl.20 Head Welds

Bl.21 Circumferential VOL CR-2,

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Bl.22 Meridional VOL CR-2  ;

Bl.30 Shell-to-t'lange Weld VOL Bl.40 I!ead-to-Flange Weld VOL AND SURF Bl.50 Repair Welds N/A B.B PRESSURE RETAINING WELnS IN VESSELS OTHER THAN REACTOR VESSELS N/A B.D l'OLL PENETRATION WELDS OF NO22LES IN VESSELS - INSPECT 7004 PROGRAM B Reactor Vessel B3.90 Nozzle-to-Vessel Welds y VOL ,

B3.100 Nozzlo Ins!Je Radius Section VOL 1

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ISI - CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR PO7/ER STATION UNIT - 2 CLASS 1

"* * " I' Revimon ? l osse 8-18-82 CODE ITEM SECT XI ET A.A RELIEF I ALTERNATE CATEGORY NUMBER REQUIRED REQUESTS #

EXAM Pressurizer 03.110 tiozzle-to-Vessel Welds N/A B3.120 %zzle Inside Radius Section N/A Steam Generators (Primary Side) a3.130 Nozzle-to-Vessel Welds N/A tl3.140 Nozzle Inside Radius Section N/A Heat Exchangers (Primary Side) 83.150 Nozzle-to-Vessel Welds , N/A 03.160 Nozzle Inside Radius Section N/A B-E PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS 84.10 Partial Penetration Welds EXTERNAL SURFACES 04.11 Vessel Nozzles VT-2 B4.12 Control Rod Drive Nozzles VT-2 04.13 Instrumentation Nozzles VT-2 4

Pressurizer B4.20 fleater Penetration Welds N/A B-P PRESSURE RETAINING DISSIMILAR METAL WELDS Reactor Vessel D$.10 Nominal Pipe Size > 4 in.

O O O Commonwee'lh l INSERVICE INSPECTION PROGRAM

@ Edison ISl - CL ASE 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASS 3' Revision 1 l Dalc 8-18-82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER HEM DESCRIPMN REMARKS REGINRED REQUESTS EXAM Nozzle-to-Safe End Butt Welds VOL AND SURP B5.11 Nominal Pipe Size < 4 in.

Nozzle *.o-Safe End Butt Welds SURF B5.12 Nozzle-to-Sate End Socket Welds N/A Pressurizer

• B5.20 Nominal Pipe Size > 4 in.

Nozzle-to-Safe End Butt Welds N/A B5.21 Nominal Pipe Size < 4 in. .

Nozzle-to-Safe End Butt Welds N/A B5.22 Nozzle-tQ-Safe End Socket Welds N/A Steam Generator B5.3C Nominal Pipe Size > 4 in.

hozzle-to-Safe End Butt Welds N/A B5.31 Nominal Pipe S!ze < 4 in.

Nozzle-to-Safe End Butt Welds N/A B5.32 Nozzle-to-Safe End Sucket Welds N/A Heat Ezchangers B5.40 Nominal Pipe Size 2 4 in.

Nozzle-to-Safe End Butt Welds N/A e B5.41 Nominal Pipe Size < 4 in.

Nozzle-to-Safe End Butt Welds N/A B5.42 Nozzle-to-Safe End Socket Welds N/A Piping B5.50 Nominal Pipe Size > 4 in.

Dissimilar Metal Butt Welds VOL AND SURP SEE FIG. IWB-2500-8 j

1

! O O O i

! co Edison co...,,, WMWM WWEUM MMMM ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATIOf4 l

Page 4 of 15 l

UNIT - 2 CLASS 1 l _

neviskm 1 lose 8-18-82

CODE ITEM I

CATEGORY NUMBER ITEM DESCRIPTION REMARKS i

REOSRED REQUESTS - EXAM B5.51 Nominal Pipe Size < 4 in.

Dissimilar Metal Butt Welds SURF B5.52 Dissimilar Metal Socket Welds ,

N/A I

B-G-1 PRESSURF. RETAINING BOLTING, GREATER Tsi'W 2 ist. IN DIAMETER Reactor Vessel B6.10 Closure llaad Nuts SURF D6.20 Closure Studs, in place VOL D6.30 Closure Studs, when removed SURF AND VOL B6.40 Threads in Flange VOL B6.50 Closure Washers, Bushings VT-1 0

Pressurizer B6.60 Bolts and Studs N/A B6.70 Flange Surface, when connection dissesemble<* N/A D6.80 Nuts, Bushings, and Washers '

N/A Steam Generators -

B6.90 Dolts and Studs N/A D6.100 Flange Surface, when concection disassembled N/A B6.110 Nuts, Dushings, and Washers N/A /

Ileat Exchangers D6.120 Bolts and Stude N/A 1

O O O Commonwealth

@ Edison ISI - CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 2 C!. ASS 1

"* 5 # 15 Revision i lcate 8-18-82 CODE ITEM TE CATEGORY NUMBER ITEM DESCRIPTION REMARKS REQUIRfD RECUESTS EXAM B6.130 Flange Surface, when connection disassembled N/A D6.140 Nuts, Bushings, and Washers N/A Piping B6.150 Bolts and Studs N/A D6.160 Flange Surface, when connection disassembled N/A D6.173 Nuts, Bushings, and Washers N/E Pumps

D6.180 Bolts and Studs VOL B6.190 Flange Surface, when connection disassembled VT-1 B6.200 Nuts, Bushings, and Washers VT-1 valves B6.210 Bolts and Studs N/A 96.220 Flange Surface, when connection disassembled '

N/A D6.230 Nuts, Bushings, and Washers N/A B-G-2 PRESSURE RETAINING BOLTING, 2 in. AND LESS IN DIAMETER s s Reactor Vessel B7.10 Bolts, Studs, and Nuts N/A

! O O O INSERVICE INSPECTION PROGRAM

@ Commonwealth ISI - CLASS 1,2& 3 COMPONENTS Edison OUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASS 1

"* ' " 15 Revision 1 lnate 8-18-82 CODE ITEM SECT XI EXAM RELIEF CATEGORY NUMBER ITEM MSCRINN ALTERNATE i REOLNRED REQUESTS EXAM Pressurizer n7.20 Bolts, Studs, and Nuts N/A l Steam Generators 87.30 Bolts, Studs, and Nuts i N/A

Heat Exchangers 87.40 Bolts, Studs, and Huts N/A

Piping i B7.50 Bolts, Studs, and Nuts VT-1 i

Pumps B7.60 Bolts, Studs, and Nuts N/A Valves 87.70 Bolts, Studs, and Nuts VT-1 CRD Housings B7.80 Bolts, Studs, and Nuts VT-1 WHEN DISASSEMBLED B-II INTEGRAL ATTACHMENTS n'OR VESSELS Reactor Vessel B8.10 Integrally Welded Attachments , SURF VESSEL SUPPORT SKIRTS SEE FIG. IWB-2500-13 LIFTING LUGS, STABILI2ER LUGS SEE FIG. IWB-2500-15

O O O Commonwealth INSERVICE INSPECTION PROGRAM

@ Edison ISI - CLASS 1,213 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASS 1 "" 7 # 15 Ikwision 1 l Dele 8-18-82 CODE ITEM SECT XI EXAM RELIEF ITEM DESCRIPTM ALTERNATE CATEGORY NUMBER REQUIRED REQUESTS EXAM MA M Pressurizer 08.20 Integrally Welded Attachments N/A Steam Generator +

08.30 Integrally Welded Attachments N/A Heat Exchangers 88.40 Integrally Welded Attachments N/A' i

D-J PRESSURE RETAINING WELDS IN PIPING B9.10 Nominal Pipe Size > 4 in.

09.11 Circumferential Welds SURP AND VOL CR-3,CR-4 B9.12 tongitudinal Welds SURF AND VOL CR-4 B9.20 Nominal Pipe Size < 4 in.

B9.21 Circumferential Welds GURP B9.22 tongitudinal Welds SURP B9.30 Branch Pipe Connection Welds B9.31 Nominal Pipe Size > 4 in. SURP AND VOL CR-5 VT-2 89.32 Nominal Pipe Size < 4 in. SURP B9.40 Socket Welds N/A i i

e- O Co .non...,,, WENM WWEWM MMMM ~

@ Edison ISI - CLASS 1,2 &3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION Page 8 of 15 I ~2 Revision 1 l Date 8-1S-82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE ITEM DESCRIPTION REMARKS CATEGORY NUMBER REQUIRED REQUESTS EXAM B-K-1 INTEGRAL ATTACHNENTS FOR PIPING, PUNPS AND VALVES I

f Piping B10.10 Integrally Welded Attachments SURF SEE FIG. TWB-2500-13,15

hgs Blo.20 Integrally Welded Attachments SURP SEE, FIG. IWB-2500-13,15 Valves B10.30 Integrally Welded Attachments SURF SEE FIG. IWB-2500-13,15 B-L-1, PRI23URE RETAINING WELDS IN B-N-1 PUNP CASINGS AND VALVE BODIES N/A B-L-2, PtBlP CASINGS AND VALVE BODIES B-N-2 B12.20 Pump Casing VT-3 CR-6 WHEN DISASSENBLED B12.40- Valve Body, Exceeding 4 in.

Nominal Pape Size VT-3 CR-7 Wi!EN DISASSENBLED B-N-1 INTERIOR OF REACTOR VESSEL Reactor Vessel B13.10 Vessel Interior VT-3 PER TABLE IWB-2500-1, CATEGORY B-N-1, J

l O O O C,,,,,,.,,,, INSERVICE- INSPECTION PROGRAM Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION

< UNIT - 2 CLASS 1 Revision 1 lo 6-16-6F CODE ITEM SECT XI EXAM RELIEF DEERMION ALTEhNATE CATEGORY NUMBER REQUIRED REQUESTS EXAM AN"I Reactor Vessel (BWR) 813.20 Interior Attachments VT-1 ACCESSIBLE WELDS 813.21- Core Support Structure VT-1 ACCESSIBLE SURFACES l

Reactor Vessel {PWR) ,

B13.30 Core Support Structure N/A

~

B-O PRESSURE RETAININ't WELDS l IN CONTROL ROD HOUSINGS Reactor Vessel B14.10 Welds in CRD Housing VOL EX-2 EX-2 APPLIED TO THESE HOUSINGS DUE TO CONFIGURATIOI I 3-P ALL PRESSURE RETAINING CONPONENTS Reactor Vessel B15.10 Pressure Retaining Boundary VT-2 LEAKAGE TEST B15.11 Pressure Retaining Boundary VT-2 HYDROSTATIC TEST Pressurlser B15.20 Pressure Retaining Boundary N/A B15.21 Pressure Retaining Boundary N/A r

Steam Generators u15.30 Pressure Retaining Boundary N/A B15.31 Pressure Retaining Doundary N/A

O O O .

INSERVICE INSPECTION PROGRAM O Commonwealth Edison ISI - CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASS 1 "* I" " 15 Revision i joale8-18-82 CODE ITEM SECT XI EXAM RELIEF CATEGORY NUMBER DEM DESCRNN ALTERNATE MARKS REQUIRED REQUESTS EXAM Heat.1;schangers B15.40 Pressure Retaining Boundary N/A B15.41 Preasure Retaining Boundary N/A Piping B15.50 Pressure Retaining Boundary VT-2 LEAKAGE TEST B15.51 Pressure Retaining Boundary VT-2 HYDROSTATIC TEST Pumps B15.60 Pressuro Retaining Boundary VT-2 LEAKAGE TEST B15.61 Pressure Retaining Boundary VT-2 , HYDROSTATIC TEST Valves B15.70 Pressure Retaining Boundary VT-2 LEAKAGE TEST B15.71 Pressure Retaining Boundary VT-2 HYDROSTATIC TEST B-O STEAM GENERATOR TUBING N/A i

I i i

1 O O O Commonwealth O Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASC 2

"* ** " 18 Revision 1 lDate 8-18-82 CODE ITEM SECT XI EXAM RELIEF CATEGORY NUMBER ITEM MSCNM ALTERNATE REOiNRED REOUESTS EXAM C-A PRESSURE RETAINING WELDS IN PRESSURE VESSELS C1.10 Shell Circumferential Welds VOL RHR HEAT EXCHANGER C1.20 IIead Circumferential Welds VOL RHR HEAT EXCHANGER C1.30 Tubesheet-to-Shell Weld N/A C-B PRESSURE RETAINING NOZILE WELDS IN VESSELS

• C2.10 NozzlesinVessel<hin.

Nominal Thickness SURP CR-8 VT-2 l

C2.20 NozzlesinVessel>hIN.

Nominal Thickness N/A C-C INTEGRAL ATTACHNENTS FOR VESSELS, PIPING, PUNPS, AND VALVES Pressure Vessels j C3.10 Integrally Welded Attachments SURP RHR HEAT EXCHANGER v

Piping C3.40 Integrally Welded Attachments SURP I

O O O "

c --

,,,, INSERVICE INSPECTION ' PROGRAM

@ Edson ISI - CLASS 1,2 & 3 ' COMPONENTS OUAD ' CITIES NUCLEAR POWER STATION :

UNIT - 2 CLASS 2 ** " 15 Revision i lpete 8-18-82 CODE ITEM SECT " EXAM ACLIEF ALTERNATE CATEGORY NUMBER I EM MSCRIPTM E U*""

REOL: '" REQUESTS EXAM Pumps 03.70 Integrally Welded Attachments  %/A Valves C3.100 Integrally Welded Attachments N/A I

l l C-D PRESSURE RETAINING BOLTING GREATER TRAM 2 in. IN DIANMTER s

Pressure vessels C4.10 Bolts and Studs N/A j Piping C4.20 Bolto and Studs N/A

• Pumps

, C4.30 Bolts and Studs i

Valves C4.40 Bolts and Studs N/A 4

C-F PRESSURE RETAINING WLLDS IN PIPING

! C5.10 Piping Welds i h in.

. Nominal Wall Thickness C5.11 Circumferential Weld SURF j C5.12 Iongitudinal Weld SURF l

1 l i

4

O O O Com onwealth 6 Ed35** ISI - CLASS 1,2 &3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASS 2 13 # **

Revision 1 lDate 8-18-82 CODE ITEM SECT XI EXAM RELIEF I ALTERNATE CATEGORY NUMBER REQUIRED REQUESTS EXAM ^

C5.20 Piping Welds > h in.

Nominal Wall Thickness C5.21 Circumferential Weld SURF AND VOL C5.22 tongitudinal Weld N/A j C5.30 Pipe Branch Connections

C5.31 Circumferential Weld SURF CR-lO VT-2 C5.32 tongitudinal Weld N/A C-G PRESSURE RETAINING WELDS

{ IN PUNPS AND VALVES N/A C-It ALL PRESSURE RETAINING CONIONENTS Pressure Vessels C7.10 Pressure Retaining Components VT-2 PRESSURE TEST C7.ll Pressure Retaining Components VT-2 HYDROSTATIC TEST Piping C7.20 Pressure Retaining Components VT-2 PRESSURE TEST C7.21 Pressure Retaining Components VT-2 HYDROSTATIC TEST Pumps C7.30 Pressure. Retaining Components ,

VT-2 PRESSURE TEST e C7.31 Pressure Retaining Components VT-2 HYDROSTATIC TEST  !

l

- m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

O -

O O O Commonwestth Edison ISI - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 2 " 14 15 CLASS 2

,, i ,, 6-16-62 CODE ITEM #

CATEGORY NUMBER ITEM DESCRIPTION REQUIRED REQUESTS EXAM REMARKS valves C7.40 Pressure Retaining Components VT-2 PRESSURE TEST C7.41 Pressure Retaining Components VT-2 HYDROSTATIC TEST n

4 i

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O O O Commonwealth INSERVICE INSPECTION PROGRAM

@ Edison ISI - CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION ,

UNIT - 2 CLASS 3 "* 15 # 15 Revision 1 lDate 8-18-82 CODE ITEM S CT M EXAM RN ITEM DESCRIPTION A DERNATE CATEGORY NUMBER REQUIRED REMARKS REQUESTS EXAM D-A SYSTENS IN SUPPORT OF REACTOR SHUTDONN FUNCTION N/A D-B SYSTENS IN SUPPORT OF ENERGENCY CORE COOLING, CONTAINNENT BEAT REHUVAL, ATNOSPHERE CLEANUP, AND REACTOR RESIDUAL HEAT RENOVAL D2.10 Pressure Retaining Components VT-2 FUNCTIONAL TEST VT-2 HYDROSTATIC TEST i

D2.20 Integral Attachment

- Component Supports and Restraints VT-3 D2.30 Integral Attachment

- Mechanical and Hydraulic Snubbers N/A D2.40 Integral Attachment

- Epring Tyre Supports N/A D2.50 Integral Attachment

- Constant toad Type Supports N/A D2.60 Integral Attachment

- Shock Absorbers

} N/A D-C SYSTENS IN SUPPORT OF RESIDUAL HEAT RENOVAL FRON SPENT FUEL STORAGE POOL N/A

O s

4

- TABLES FOR INSERVICE INSPECTION PROGRAM OF COMPONENT SUPPORTS QUAD CITIES UNITS I & 2 l

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COM-06-003 Revision 1 t

I

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d_ . _ _ O Commonwealth

@ Edison ISI - CLASS 1,2 & 3 COMPONENTS .

QUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASS 1 Pm i We Revision 1 lpete 8-18-82 COCE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER ITEM MSNMN REOiNRED EMARKS REQUESTS EXAM F-A PIATE AND SHELL TYPE SUPPORTS F-1 Mechanical Attachments, including bolting VT-3 i

F-2 Welded Attachments N/A j F-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3 Assembly of support items VT-3 i

F-B LINEAR TYPE SUPPORTS F-1 Mechanical Attachments including bolting VT-3 F-2 Welded Attachments 'VT-3 P-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3 Assembly of support items VT-3 P-C COMPONENT STANDARD SUPPORTS

! F-1 Mechanical Attachments, including l bolting VT-3

- ] O Ccmmonwealth

@Emon ISl - CLASS 1,2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION UNIT - 1 CLASS 1 Revision 1 lDete8-18-82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER ITEM MSCRIPTM MARKS REQUIRED REQUESTS EXAM F-2 Welded Attschments VT-3 P-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3

] Assembly of support items VT-3 i

F-4 Spring type supports VT-4 Constant load type supports VT-4 I

Shock absorbers VT-4 Hydraulic and mechanical type anubbers VT-4 ,

4 l

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. Commonwealth Edison ISI - CLASS 1.2 8 3 COMPONENTS 1 QUAD CITIES NUCLEAR POWER STATION l

UNIT - 1 CLASS 2 y 3 ma Rausion 1 lDate 8-18-82 CODE ITEM SECT XI EXAM RELIEF AITERNI.TE l

CATEGORY ' REMS NUMBER REOutRED REQUESTS EXAM F-A PLATF AND S!!CLL TYPE SUPPORTS N/A F-8 LINEAR TYPE SUPPORTS P-1 Mechanical Attachments, including bolting VT-3 F-2 Welded Attachments VT-3 3 F-3 CM ponent displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3, Asceably of support items VT-3 l

F-C COMPONENT STANDARD SUPPORTS F-1 Mechanical Attachments, including

• bolting VT-3 l F-2 Welded Attachments VT-3 F-3 Component displacement VT-3 Settings of guides and stops VT-3 ft+salignment of supports VT-3 Assembly of support items VT-3 F-4 Spring type supports VT-4 Constant load type supports VT-4 Shock absorbers VT-4 Hydraalic and mechanical typo snubbers VT-4 9

O O O C = .n ..u, INSERVICE INSPECTION PROGRAM

@ Edison ISI - CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION

. m UNIT - 2 CLASS 1 5 8 '

j , 8-18-82 i

' CODE ITEM SECT XI EXAM REUEF I DESCRIPMN ALTERNATE CATEGORY NUMBER REMARG REOUIRED REQUESTS EXAM F-A PLATE AND SHELI- TYPE SUPPORTS F-1 Mechanical Attachments, including bolting VT-3 F-2 Welded Attachments N/A i

F-3 Component displacement VT-3 Settings of guides and stops VT-3

• i Misalignment of supports VT-3 ,

. Assembly of support items .VT-3 1

i F-B I,1NEAR TYPE SUPPORTS F-1 Mechanical Attachments, including

bolting VT-3 j F-2 Welded Attachments VT-3 F-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignsent of supports VT-3 Assembly of support items VT-3 ,

I F-C COMPONENT STANDARD SUPPORTS i

• F-1 Mechanical Attachments, including bo1 ting VT-3

3. . G iU#i . .J; i -. . = Ef M - W S. *% E. . W. '.Y: . N . .

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Commonwealth I @ Edison ISI - CLASS 1,2 &3 COMPONENTS OUAD CITIES NUCLEAR POWER STATIO*1 UNIT - 1 CLASS 3 "* * "

• Revision 1 lDet*8-18-82 _

CODE ITEM REMARKS ITEM DESCRIPTICM REQUIRED REQUESTS EXAM CATEGOftY NUMBER ,

F-A PLATE AND SHELL TYPE SUPPORTS N/A F-B LINEAR TYPE SUPPORTS F-1 Mechanical Attachments, including bolting VT-3 F-2 Welded Attachments VT-3 P-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3 Assembly of support items VT-3 Y-C COMPONENT STANDARD SUPPORTS P-1 Hechanical Attachments, including bolting VT-3 P-2 Welded Attachments VT-3 ,_

F-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3 -

Assembly of suppGrt items VT-3 P-4 Spring type supports Vt-4 Constant load type supports VT-4 Shock absorbers VT-4 Ilydraulic and mechanical type snubbers VT-4

l .

O O O Commonwealth INSERVICE INSPECTION PROGRAM Edison ISI - CLASS 1,2&3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION

~

Revelon 1 l Dale d-15-5Z i CODE ITEM SECT XI EXAM REUEF ITEM DESCRI N N ALTERNATE CATEGORY NUMBER REQUIRED REMARKS REQUESTS EXAM l P-2 Welded Attachments VT-3 l

l F-3 Component displacement VT-2 Sattings of guidos and stops VT-3 Misaligt.sent of supports VT-3 Assembly of support items VT-3 P-4 Spring type supports VT-4 Constant load type sopports VT-4 Shock absorbers VT-4 Ilydraulic and piechanical type snubbers VT-4 e

h e

)

O O O Commonwealth -

@ Edison ISI - CLASS 1,2 1 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION Page 7 ci 8 tJNIT -- 2 CMSS 2 Revision 1 l Dele 8-18-82 CODE ITEM SECT XI EXAM RELIEF ALTERNATE CATEGORY NUMBER IT M DESCRW M E""U REQUIRED REQUESTS EXAM P-A PLATE AND SitELL TYPE SUPPORTS N/A I F-B LINEAR TYPE SUPPORTS

'l Mechanical Attachments, including bolting VT-3 t

F-2 Welded Attachments VT-3 P-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3 Assembly of support items VT-3 l

F-C COMPONENT STANDARD SUPPORTS P-1 Mechanical %ttachments, including

• boIting VT-3 P-2 Welded Attachments VT-3 P-3 Component displacement VT-3 Settings of guides and stops VT-3 Misalignment of suppcrts VT-3 Assembly of support items VT-3 F-4 Spring type supports VT-4 -

Constant load type supports VT-4 Shock absorbers VT-4 Ilydraulic and Mechanical Type Snubbers VT-4

O O O

,.,_,_,,, WMMM, MNMO,N MMMM 4

@ Edison (St - CLASS 1,2 & 3 COMPONENTS OUAD CITIES NUCLEAR POWER STATION UNIT - 2 CLASS 3 "* * #

• Revision 1 l Dene 8-18-82 CODE ITEM ITEM DESCRIPTION REMARKS CATEGORY NUMBER REQUIRED REQUESTS EXAM F-A PLATE AND SHELL TYPE SUPPORTS N/A F-B LINEAR TYPE SUPPORTS P-1 Mechanical Attachments, including .

i bolting VT-3 P-2 vielded Attachments VT-3 j P-3 Component displacement VT-3 Settings of guides and stops VT-3 i Misalignment of supports VT-3

! Assembly of support items VT-3 F-C COMPONENT STANDARD SUPPORTS P-1 Mechanical Attachments, including bolting VT-3 P-2 Welded Attachments VT-3 1

P-3 Component displa ::ement VT-3 Settings of guides and stops VT-3 Misalignment of supports VT-3 Assembly of support items VT-3 I

P-4 Spring type supports ,VT-4 -

Cor.:.itant load type supports VT-4 Shock absorbers VT-4 Ilydraulic and mechanical type snubbers VT-4

~.,

l O'

SECTION 2.3

\

i RELIEF REQUESTS FOR INSERVICE INSPECTION PROGRAM l

i O

O COM-06-003 Revision 1 2-8

RELIEF REQUEST NOe C R-1 O

I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS The reactor vessel is designed with one circumferential and six longitudinal welds in the core beltline region as shown on Figure la.

The ASME Eoiler and Pressure Vessel Code,Section XI, 1980 Edition through the Winter 1980 Addenda requires a volumetric examination of 100. percent of the length of one beltline longitudinal weld and one beltline circumferent'ial weld each ten year interval (Code Category B-A).

c( }

Relief is requested from'the above mentioned Code requirements on the basis of inaccessibility.

II. BASIS FOR RELIEF Accessibility for inspection of these welds was not provided for i

I in the' original plant design which occurred prior to the issuance l

l of Section XI inservice inspection requirements.

As indicated on Figure la, examination from the reactor vessel l-

~

outer surface is pre'cluded due the close proximity to the 4

i COM-06-003 Revision 1 2-9

-biological' shield wall and. obstruction by the vessel

! O insulation. The mirror type insulation consists of interlocking

panels which were not designed to be easily removable at the weld-locations. Furthermore, the annular dimensions between the shield. wall and the insulation is not sufficient to allow direct l access to personnel. Access through the biological shield wall' is only provided at reactor vessel nozzle locations, however, there are'no nozzle penetrations in:the belt line region.

. Examination of the beltline region welds from inside the vessel is impeded by vessel internal design features. The core shroud, i jet pumps,-and.various brackets welded to the vessel wall are not designed to be removable.

O III. ALTERNATE PROVISIONS i

Currently, it is not feasible to perform the-required volumetric examinations on these welds. Commonwealth Edison will, howeve'r, 4

keep abreast of improvements in state-of-the-art NDE techniques that could provide a viable means of examination.

!O COM-06-003 Revision 1 2-10

RELIEF REQUEST NO. CR-2 O

I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS The reactor vessel contains thirteen longitudinal welds and six circumferential welds in the shell sections and bottom head which are inaccessible for examination, in addition to the beltline region welds addressed in Relief Request CR-1.

Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition through the Winter 1980 Addenda requires a volumetric examination of 100 percent of the length of one meridional head weld and one circumferential head weld each inspection interval

)

(Code Category B-A).

As shown on Figure la and lb, all of the reactor vessel closure head welds are fully accessible for examination. The bottom head l

welds, however, are inaccessible for examination.

l l

II. BASIS FOR RELIEF As discussed in Relief Request CR-1, accessibility for examination of these welds was not considered in the plant design. The bottom head welds cannot be examined because of the limited physical access, the inability to remove vessel

()

COM-06-003 Revision 1 2-11

l insulation panels, and also because of interference from the O forest of control rod drive and instrumentation penetrations.

III. ALTERNATE PROVISIONS l

Currently, it is not feasible to perform the required volumetric examinations on the bottom head welds. -Commonwealth Edison will, however, keep abreast of improvements in state-of-the-art NDE i

techniques that could provide a viable.means of examination.

l l

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I COM-06-003

, Revision 1 2-12

RELIEF REQUEST NO. CR-3 I.. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL. CODE REQUIREMENTS

Two Class-1 piping welds in Unit 2 and one. weld in Unit 1 are physically inaccessible for examination. These-welds are in the 1

l Control Rod Drive System on line number 0308-4". These welds i cannot be examined because of interference from a structural support as shown on Figure 4.

i Section XI of the ASME Boiler and Pressure Vessel Code,1980 I '

! Edition including the Winter 1980 Addenda requires that twenty-five percent of the total number of circumferential pipe.

[]}

welds be volumetrically ~ examined each ten year interval (Code Category B-J)..

It is unlikely that these welds be-inspectable at anytime during the-plant life. Relief is, therefore, requested from performing the' volumetric examination requirements of Section XI.

II. BASIS FOR RELIEF The implications of this exemption are minimal due to the fact that safety margins inherent in the design of the subject welds are typical of those in all other welds in the Class-1 systems, i

Exempting these three welds from the total inspection sampling l

() -program will have negligible statistical significance.

COM-06-003 Revision 1 2-13 l

III'. ALTERNATE PROVISIONS O

• No alternate or augmented examinations are feasible or necessary in this case. The examinations required by IWB-5000 will, however, be conducted in accordance with the Code.

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i l COM-06-003 2-14 Revision 1 i .

RELIEF REQUEST NO. CR-4 o .

I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS

. -Each of the-lines listed below penetrates the primary containment by means of a penetration assembly similar in design to that shown in Figure-2. These Class-1 lines, due to the design of the penetration assembly, have one circumferential pressure retaining weld that is inaccessible for volumetric examination.

CRD RETURN - 0308-4" RHR.- 1012A&B-16", 1025-20" i Rx WATER CLEANUP - 1202-6" CORE SPRAY - 1403-10", 1404-10" HPCI - 2305-10" MAIN STEAM 3001A,B,C,D-20" FEEDWATER 3204A&B-18"

.The ASME Boiler and Pressure Vessel Code,Section XI,.1980 Edition through the Winter 1980 Addenda requires a volumetric and surface examination'on Class-1 welds (Code Category B-J).

Since this requirement is impractical due to plant design, relief is requested from the above stated examination requirements.

O COM-06-003 Revision 1 2-15

II. ' BASIS FOR RELIEF i-As stated in 10CFR50.55a (g)(1) for plants whose construction permits were issued prior to January 1,-1971, components shall meet Section XI requirements to the extent practical. Since examination requirements for these' welds did not exist at the time-Quad Cities Station was designed, accessibility for their examination was not a prime consideration. Figure-2 clearly illustrates the design constraints which make it extremely impractical to the examine the subject welds by volumetric or surface techniques. Commonwealth Edison feels that this consitutues a basis for relief from the volumetric examination requirements of Section XI.

The safety implications of this exemption are minimal due to the fact that the safety margins in the subject welds are typical of those in all welds in the applicable systems. Since the exempted welds represent only a small fraction of the total number of j circumferential, Category B-J welds in these systems (14 out of 291, and 14 out of 280, Unit 1 and 2 respectively), the statistical significance to the inspection sampling program due to exempting these welds is expected to be negligible.

O COM-06-003 Revision 1 2-16

s

( III. ALTERNATE PROVISIONS O

At the present time no alternate examinations are feasible because of the inaccessibility. The examinations required by IWB-5000 will be conducted in accordance with the Code.

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1 COM-06-003 Revision 1 2-17

~ RELIEF' REQUEST NO. CR-5 I. IDENTIFICATION'OF COMPONENTS AND IMPRACTICAL CODE l l REQUIREMENTS The design of certain Class-1 branch pipe connection welds calls for the use of reinforcement saddles. These saddles are fillet.

welded over the actual pressure retaining branch pipe to main.

pipe weld, completely encasing it as illustrated on Figure 3.

There are four such welds that are greater than 4 inches in

, diameter.

f i

Section XI of the ASME Boiler-and Pressure Vessel Code, 1980 Edition through the Winter 1980 Addenda requires that branch pipe

)

connection welds exceeding four inches diameter be surface and volumetrically examined. Twenty-five percent of these welds are required to be examined each inspection interval (Code Category B-J).

i Relief is requested from this requir ment due to the physical inaccessibility of the design.

i II. BASIS FOR RELIEF The fabrication of these joints precludes any type of surface examination.or meaningful volumetric examination. Additional O

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-CCM-06-003 Revision 1 2-18 i

t assurance of the continued integrity of joints fabricated in this

O fashion is afforded by'the fact that the reinforcement saddle strengthens the joint and reduces the stresses on the internal weld.

III.- ALTERNATE PROVISIONS i

A visual examination of these joints for evidence of leakage will be conducted during the pressure tests required by IWB-5000.

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O COM-06-003 Revision 1 ,

2-19

. . RELIEF' REQUEST NO. CR-6 O

I '. IDENTIFICATION OF COMPONENTS AND' IMPRACTICAL CODE '

REQUIREMENTS Each Quad Cities Unit has an ISI Class-1 recirculation pump in each of the two 28-inch diameter recirculation loops. These pumps function during normal reactor operation'to provide-forced recirculation through the core.. -

1 i

The ASME Boiler and Pressure Vessel Code,Section XI, 1980 Edition through the Winter 1980 Addenda requires that one of l

these recirculation pumps be examined visually during each inspection interval.- Specifically, the area of examination

)

includes all pump internal pressure boundary surfaces.

As discussed, in detail-below, Commonwealth Edison requests relief from the Section XI examination requirement to visually examine the recirculation pump internal surfaces on the basis of impracticality.

l II. BASIS FOR RELIEF The basis for this relief request is predicated on the following two points:

O COM-06-003 Revision 1 2-20

1) to complete the subject examination, large expenditures of manhours and man-rem are required with-essentially no compensating increase in plant safety, and

2) the structural integrity afforded by the pump casing material utilized will not significantly degrade over the lifetime of the pump.

Based on data' compiled from an actual recirculation pump disassembly, it is expected that approximately 1000 man-hours and 50 man-rem exposure would be required to disassemble, inspect,

-and reassemble one pump. Performing this visual examination under adverse conditions such as high dose rate (30-40 R/hr) and poor as-cast surface condition, realistically, provides little additional information as to the pump casing integrity.

The recirculation pump casing material, cast stainless steel (ASTM A351-CF-8), is widely used in the nuclear industry and has performed extremely well. The presence of some delta ferrite (typically 5% or more) imparts substantially increased resistance to intergranular stress corrosion cracking. The delta ferrite also results in improved pitting corrosion resistance in chloride containing environments.

O COM-06-003 Revision 1 2-21

r Commonwealth Edison feels that adequate. safety margins are

. inherent in the basic pump design and'that the health and safety.

of.the public'will not be adversely effected by-performing the visual examination of the pump internal pressure boundary surfaces only when the. pumps are required to be disassembled for maintenance.

i l III. ALTERNATE PROVISIONS As stated above, it is not felt that the visual examination  :

required by Code each ten year interval is warranted. However, as standard maintenance practice dictates, when a pump of this i type is disassembled for maintenance examination of the pump r

l internals and internal pressure boundary surfaces will be performed, to the extent practical.

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O COM-06-003 Revision 1 2-22 i

4 RELIEF REQUEST NO. CR-7'

,L-I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL' CODE REQUIREMENTS i-In the Class-1 system there are 51 valves which are greater than four inches nominal pipe size. These-. valves vary in size,

. design,_and manufacturer but are all manufactured from either

- cast stainless steel or carbon steel. None of the valve body casings are. welded.

Section XI of the ASME Code, 1980 Edition through the Winter 1980 l Addenda requires that a visual examination of the internal pressure boundary surfaces of one valve in each. group of valves

)

of the same constructional design and manufacturing method that perform similar functions in the system. These examinations:are required to be completed each inspection interval. (Code Category B-M-2)

Since these examinations must be met whether or not the valves have to be disassembled for maintenance, this requirement is considered impractical.

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O COM-06-003 Revision 1 2-23 l

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II. BASIS FOR~ RELIEF  !

() l The requirement'to disassemble primary system valves.for the sole purpose. of performing a visual examination ~of' the - internal pressure _ boundary surfaces has.only a very small potential of increasing' plant safety ~ margins and a very disproportionate impact.on expenditures of plant manpower and radiation exposure.

- Performing these visual examinations, under such adverse conditions as high dose rates (10 R/hr) and poor as-cast surface

' condition, realistically, provides little additional information as to the valve casing integrity.

For approximately 20 percent of these valves, the reactor vessel core must be completely unloaded and the vessel drained to permit  :

disassembly for examination.

The performance of both carbon and stainless cast valve-bodies has been excellent in all BWR applications. Based on this experience and both industry and regulatory acceptance of these alloys, continued excellent service performance is anticipated.

A more practical approach that would essentially provide an equivalent sampling program and significantly reduced radiation exposure to plant personnel is to inspect the internal pressure boundary of only those valves that require disassembly for

()

COM-06-003

Revision 1 2-24

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

maintenance purposes. This would still provide a reasonable .

O sampling of primary system values and give adequate assurance that the integrity of these components is being maintained.

. III. ALTERNATE PROVISIONS

\ 0 An examination of the internal pressure boundary surfaces will be f performed, to the extent practical, each time a value is )

disassembled for naintenance purposes.

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COM-06-003 Revision 1 2-25

I RELIEF REQUEST NO. CR-8 i _ ().

I. . IDENTIFICATION'OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS 1

There are-two 18" diameter nozzles in'the Class-2 portion of each l of the two RHR System heat exchangers that are fabricated with reinforcement saddles. .These saddles are fillet welded over the actual pressure retaining' nozzle to shell weld. The.

configuration is shown on Figure-5.

Section XI of the ASME Boiler and Pressure Vessel Code,1980 Edition through the Winter 1980 Addenda requires a= surface

() ' examination of two of these four nozzle-to-shell welds in the inspection inteval. This requirement is impractical due_to inaccessibility.

II. BASIS FOR RELIEF f

The fabrication of.these nozzle-to-shell welds precludes any type of volumetric or surface examination. The design does, however, provide additional strength at the joint and results in lower stresses at the internal weld. Integrity of these joints will be ,

) monitored by periodic system pressure and hydrostatic tests.

4

()

COM-06-003 Revision 1 2-26

-- . III. ALTERNATE PROVISIONS A. visual examination for evidence'of leakage will be conducted in accordance with the Subsection IWC-5000 requirements.

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O COM-06-003 Revision 1 2-27

RELIEF REQUEST NO. CR V I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS ..

Quad ' Cities Station currently utilizes 'a calibratton block which lacks documentation consistent with the requirements of current editi'ons of the Code. The documentation requirements' existing at the time of their fabrication did not require traceability to the material's chemical or physical certifications. As a result, the only documentation available for the existing blocks.is verification of the appropriate P-number grouping.

-O The Section XI requirements of the 1980 Edition of the ASME Code

's V 4

including the Winter 1980 Addenda specify that the block will be fabricated as provided by Article III-3400, paragraph III-3411 .

requirements.

Relief is requested from this documentation requirements to allow the ~ continued use of the existirig calibration blocks.

i l

[ II. BASIS FOR RELIEF Previous inservice inspections-have been performed utilizing the above mentioned block and its use would provide continuity in the ISI Program. It would be impractical to fabricate a new COM-06-003 Revision 1 2-28

9

' calibration block- in order to satisfy the documentation

,e k_ requirements of the-current Code. Existing records which l

indicatefthe appropriate material P-grouping provide-adequate ,

assurance that the' block' will establish the proper ultrasonic calibration and sensitivity. Additionally,:since both reactors

. vessels are 100% clad on the I.D. surface, there is no way to meet tdus requirement of . verifying the acoustic properties of the block against the clad. component.

4 III.- ALTERNATE PROVISIONS The present reactor vessel calibration block will be demonstrated to have acoustic attenuation and velocity properties which fall l ,( J within the range of' straight beam longitudinal wave velocity and attenuation as found in the reactor vessel. However, since Quad-Cities Station reactor vessels are 100% clad on the I.D. Surface,-

this check will be completed on the clad component and appropriate reviews made by the C.E.C.O. Level III Examiner to verify the acceptability of the block. ,

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COM-06-003

' Revision 1 2-29

RELIEF REQUEST NO. CR-10 o

1. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENT The design of certain Class-2 branch pipe connection welds calls for.the use of reinforcement saddles. These saddles are fillet welded over the-actual pressure retaining' branch pipe to main

~ pipe weld, completely encasing-it as illustrated on Figure 3. As listed in the. program, there are 40 cuch wolds that are greater than 4 inches in diameter.

f Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition through the Winter-1980 Addenda requires that branch pipe connection welds exceeding 4 inches diameter be. surface examined. Twenty-five percent of these welds are required to be examined each inspection interval (Code Category C-F).

Relief from this requirement is requested due to the physical ,

inaccessibility of the design.

II. BASIS FOR RELIEF The fabrication of these joints precludes any type of surface

. examination. Additional assurance of the continued integrity of joints fabricated in this fashion is afforded by the fact that O

COM-06-003 Revision 1 2-30

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

the reinforcement saddle strengthens the joint and reduces the stresses on the internal weld.

I III. ALTERNATE PROVISIONS A visual examination of these joints for evidence of leakage will be conducted during the pressure tests required by IWC-500.

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O COM-06-003 Revision 1 2-31 l

(*

()) RX VESSEL FLANGE l N a.c SHELL COURSE #4 l J REACTOR VESSEL v e e c 7

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• COM-06-003 Revision 1 2-32

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• Iw Head Nozzles sii ,,

Typical Lifting

• Closure Head Lugs All Welds B-B

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FIGURE 1.b Reactor Closure Head Weld Identification O COM-06-003

~

Revision 1 2-33

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. INSilLATION -

- ~~rs N N N N N N N_ N N N~N N N N NT / TNNNNA INACCESSIBLE WEl,D i

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< INACCESSIBLE WELD

, (PRESSURE-RETAINING)

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BRANCH PIPE RUN s

p . . . . /. . - -- .

- 4 . . .

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REINFORCEMENT PLATE i -

\ MAIN PIPE RUN

<p INACCESSIBLE WELD (PRESSURE-RETAINING) id i l .

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' BRANCH PIPE RUN l l /

/ , REINFORCEMENT l / SADDLE

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FIGURE 3 TYPES OF REINFORCED BRANCH PIPE CONNECTIONS. ,

COM-06-003 i

Revision 1 2-35 1

\. 7 -

Guide O^

~ >

I I

Line 0308-4' Weld # 03- Fil (U-2)

C .

' STRUCTURAL STEEL GUIDE

PENETRATION

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/ LINE 0308-4"

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_Q FIGURE 4 COM-06-003 WELD OBSTRUCTION DETAILS Revision 1 2-36

f 18" N0ZZLES (TUBE SIDE-

• eq RHRS SERVICE WATER) 4" RELIEF VALVE CONN.

7 .

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RHR HEAT EXCHANGER SHELL l .q REINFORCEMENT lg PLATE 18" N0ZZLE 2" RELIEF VALVE CONN .s (SHELL SIDE -

1 RHRS WATER)

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CAT C-A L,_ -

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FILLET WELDS 6Li-13/8"j',"I CAT C-D INACCESSIBLE CAT C-A ILLET WELD (PRESSURE RE-DETAIL "A" TAINING) CAT C-B l

l 1" DRAIN CONN l

FIGURE 5 l

' e COM-06-003 RHRS HEAT EXCHANGERS Revision 1 2-37

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

3.0 INSERVICE TESTING PROGRAM.FOR SNUBBERS'

.O:

3.1 GENERAL INFORMATION The -Inservice Testing Program for snubbers meets the requirements of subsection IWF, of Section XI of the ASME Boiler and Pressure Vessel Code,1980 Edition through the Winter 1980 Addenda. Where the requirements are determined to be impractical, specific requests for relief have been written.

The IST Program for snubbers is presented in Section 3.2 in a tabular format. The snubbers are listed without Code category as there is no category assigned by Section XI of the ASME Code. The information given in the tables is explained below.

A. ISI"SupporE ~No. - lists the number assigned to the support assembly which belongs to ISI class 1, 2 and-3 (i.e. structural attachments and snubbers).

Tech ~ Spec' Snubber No. - lists the unique number

~ ~

B..

assigned to the snubber as shown in the plant's Technical Specifications.

O COM-06-003 4

Revision 1 3-1

C.~ Snubber Type - identifies the snubber type (i.e. M

= mechanical, H = hydraulic D. System - gives the system abbreviation (see tables 3.2-1) of this program for explanation of

\_ abbreviations.

E. Test Parameters - lists the appropriate test to be performed cn1 anubbers. These tests are referenced to the appropriate paragraph within IWF-5400.

F. Remarks - gives general clarification remarks.

Tables 3.2-1 list the systems and their respective P&ID numbers which are covered in the Snubber Testing Program.

Section 3.3 includes a relief request that is generic to the Snubber Testing Program.

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O COM-06-003 -

! Revision 1 3-2

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SECTION 3.2 4

TABLES FOR INSERVICE TESTING PROGRAM FOR SNUBBERS A. QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 l

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COM-06-003 Revision 1 3-3

- TABLE 3.2-la LIST OF SYSTEMS INCLUDED IN THE SNUBBER PROGRAM FOR UNIT-1 SYSTEM' REFERENCE SYSTEM NUMBER P&ID Recirculation (Recirc.) 0200 35-2 Control Rod Drive (C.R.D.) 0300 41 Residual Heat Removal (RHRS) 1000 37 & 39 Core Spray 1400 36 Main Steam 3000 13-1 & 2 O

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COM-06-003 Revision 1 3-4

O O O INSERVICE TESTING PROGRAM

@ Commonwealth SNUBBER TESTNG (fWF-5000)

OUAD CITES NUCLEAR POWER STATION UNIT - 1 Peet 1 of 3 Revision 'bv lDate 8/18/82 TE G $PEC $NUtBER TEST PARAMETER $

151 $UPPORT NUMSER SYSTEM REMAngs

$NUSSER NOL TYPE

{,)

1403-W-103 1-1 M Core Spray X X 1403-W-103 1-2 M Core Spray + X X 1404-W-103 1-3 M Core Spray X X 1404-W-103 1-4 M Core Spray X X 1012A-W-101 1-5 M RIIRS X X 1012A-W-102 1-6 M RIIRS 'X X 1012B-W-102 1-7 M RilRS X X 1012D-W-101 1-8 M RiiRS X X

~

1025-W-103 1-9 M RilRS X X 1025-W-101 1-10 M , RilRS X X 0200-W-109 1-11 M Recire (1A-202 PMP) X X Note (4) 0200-W-127 1-12 M Recire (18-202 PMP) X X Note (4) 0200-W-107 1-13 M Recirc (IA-202 PMP) .X X Note (4) 0200-W-129 1-14 H Recire (1B-202 PMP) X X Note (4) 0200-W-120 1-15 M .Recire (1B-202 PHP) X X Note (4) 0200-W-154 1-16 M Recirc Ring lleader X X Note (4) 0200-W-143 1-17 M Recirc Ring Ileader X X Note (4) 0200-M-110 1-18 M Recirc (1A-202 Motor) X X Note (4)

! 0200-M-130 1-19 M Recirc (1B-202 Motor) X X Note (4) 0200-W-169 1-20 M Recirc X X Note (4) 0200-W-108 1-21 M Recirc (1A-202 PMP) X X Note (4) 0200-M-111 1-22 H Recire (1A-202 Motor) X X Note (4) 0200-M-112 1-23 M Recirc (IA-202 Motor) X X Note (4) 0200-M-131 1-24 H Recire (1B-202 Motor) X X Note (4) 0200-M-132 1-25 H Recirc (18-202 Motor) X X Note (4)

N/A 1-26 M Main Steam (SRV disch) X X Note (4)

NOTES: (1) Test as required by IWF 5400(b)(1) (4) Snubber 50 kips or Greater - No test specification per (2) Test as required by IWF 5400(b)(2) IWF-5300. Will be tested for full stroke, freedom of (3) Test as required by IWF 5400(b)(3) movement in compression & tension.

. D 0 D

%,,,,, INSERVICE TESTING PROGRAM

@ Edson SNUBBER TESTNG (lWF-5000)

DUAD OTES NUCLEAR . POWER - STATION UNIT - 1 #

Revision 1: l Dale 8/18/82 TECH $PEC SNUSSER TEST PARAMETERS 454 M M T NUMBER SNutBER NOL SYSTEM TYPE REMARKS (4) (2) (3)

N/A 1-27 M Main Steam (SRV disch) X X Note (4)

N/A 1-28 M Main Steam (SRV disch) X X N/A 1-29 M Main Steam (SRV disch) X X Note (4)

N/A 1-30 M CRD X X N/A 1-31 M CRD X X N/A 1-32 M CRD X X +

N/A 1-33 M CRD X X N/A 1-36 11 RilRS/SDC X X X N/A 1-37 11 RIIRS/SDC X

~

X X N/A 1-38 il RIIRS/SDC X X

' X N/A 1-39 Il RIIRS/SDC X X X N/A 1-40 11 Ri!RS/SDC X X X 1015A-W-202 1-41 Il RilRS X X X 1015A-W-203 1-42 11 R11RS X X X 0200-M-171 1-43 M Recirc X X 0200-H-172 1-44 M Recirc X X 0200-M-175 1-45 M Recirc X X 0200-H-174 1-46 M Recirc X X 0200-M-173 1-47 M Recirc X X 1 0200-M-170 1-48 M Recirc X X 1011-M-111 1-49 M RIIRS/IIead Spray X X N/A 1-50 M Main Steam X X N/A 1-51 M Note (4)

Main Steam X X N/A 1-52 Note (4)

M Main Steam X X N/A 1-53 Note (4)

M Main Steam X X i

N/A 1-54 Note (4)

M Main Steam X X Note (4) '

NOTES,* (1) Test as required by IWP 5400(b)(1) (4) Snubber 50 kips or Greater - No test specification per (2) Test as required by IWP 5400(b)(2) IWF-5300. Will be tested for full stroke, freedom of (3) Test as required by IWP 5400(b)(3) movement in compression & tension.

. O O O

%,,,,, INSERVICE TESTING PROGRAM

@ N SNUBBER TESTNG (IWF-5000)

OUAD OTES NUCLEAR POWER STATION UNIT - 1 * ' " 3 Revision b lDete8/18/82 TECH SPEC SNUSSER TEST PARAMETERS 151 NT MR SNUBBER NCE SYSTEM REMARKS TYPE (1) (2) (3)

N/A 1-55 H Main Steam X X Note (4)

N/A 1-56 M Main Steam i X X Note (4)

N/A 1-57 H Main Steara X X Note (4) l l

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NOTES: (1) Test as required by IWF 5400(b)(1) (4) Snubber 50 kips or Greater - No test specification per (2) Test as required by IWF 5400(b)(2) IWP-5300. Will be tested for full stroke, freedom of (3) Test as required by IWF 5400(b)(3) movement in compression & tension.

TABLE'3.2.lb O

LIST OF SYSTEMS INCLUDED IN THE SNUBBER PROGRAM FOR UNIT-2 SYSTEM REFERENCE i SYSTEM NUMBER P&ID l l

' Recirculation-(Recirc.) 0200 77-2 Control Rod Drive (C.R.D.) 0300 83 Residual Heat Removal (RHRS) 1000 79 & 81 l Standby Gas Treatment (SBGTS) 7500 44 i Reactor. Cleanup (RWCU) 1200 88 l Reactor Core Isolation Cooling (RCIC) 1300 89 -

l Core Spray 1400 78 Pressure. Suppression (Press. Supp.) 1600 76 High Pressure Coolant-Injection (HPCI) 2300 87 ,

Main Steam 3000 60-1 & 2 I

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COM-06-003 '

Revision 1 3-5

O O O .

@o m _ _ -.- alth Edison SNUBBER TESTNG IIWF-5000)

QUAD OTES NUCLEAR POWER STATION UNIT - 2 Page 1 of 4 TEGI SPEC 3 4 s/is/s2 SNUBSER TEST PARAMETERS ISI SUPPORT NUMBER SYSTEM

$NUSSER NOL TYPE (t) (2) (3) 1403-W-103 2-1 M Core Spray X X 1403-W-103 2-2 M Core Spray . X X 1404-W-103 2-3 M Core Spray X X

, 1404-W-103 2-4 M Core Spray X X 1012A-W-101 2-5 M RIIRS X X 1012A-W-102 2-6 M RIIRS X X 1012D-W-102 2-7 M Rl!RS X X 1012D-W-101 2-8 M RilRS X X 1025-W-101 2-9 M RilRS X X

1025-W-101 2-10 M RI!RS X X 0200-W-107 2-11 M Recire (2A-202 PMP) X X Note (4) 0200-W-127 2-12 M '

Recire (2B-202 PMP) X X Note (4) 0200-W-109 2-13 H Recirc (2A-202 PMP) X X Note (4) 0200-W-129 2-14 H Recirc (2B-202 PMP) X X. Note (4) 0200-W-128 2-15 M Recirc (2B-202 PHP) X X Note (4) 0200-W-154 2-16 M Recirc Ring IIeader X X Note (4) 0200-W-143 2-17 M Recirc Ring Ileader X

~

X 12ote (4) 0200-M-110 2-18 H Recirc X X Note (4) 0200-M-130 2-19 M Racirc X X Note (4) 0200-W-169 2-20 M Recirc Ring Ileader X, X Note (4) 0200-W-108 2-21 M Recire (2A-202 PMP) X X Note (4) 0200-M-111 2-22 M Recire (2A-202 Motor) X X Note (4) 0200-M-112 2-23 M Recirc (2A-202 Motor), X X Note (4) 0200-M-131 2-24 M Recirc (2B-202 Motor) X X Note (4) 0200-M-132 2-25 M Recirc (2B-202 Motor) X X Note (4)

N/A 2-26 H Main Steam X 'X Note (4) #

NOTES: (1) Test as required by IWP 5400(b)(1) (4) Snubber 50 kips or Greater - No test specification per (2) Test as required by IWP 5400(b)(2) IWP-5300. Will be tested for full stroke, freedom of (3) Test as required by IWP 5400(b)(3) movement in compression & tension.

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,, INSERVICE TESTING PROGRAM i E**oa SNUBBER TESTNG (IWF-5000)

QUAD OTES NUCLEAR POWER STATION i

Page 2 og 4

) UNIT 2 7 lo.ees/is/s2 TEOt SPEC ENUGAER TEST PAR M j 158 SUPPORT NUAASER SNUSSER NO, SYSTEAR TYPE REAAARKS -

i {,) gg) (3) 4

-N/A 2-27 M Main Steam X X Note (4) 3001A-N-105 2 M Main Steam i X X Note (4)

3001A-M-106 2-29 M Main Steam X X i

0308-M-101 2-30 Note (4)

M CRD X X 0308-N-102 2-31 M CRD X X 0308-M-103 2-32 M CRD X X -

} 0308-M-104 2-33 M CRD X X 1015B-W-202 2-34 II RHRS X X X l 10158-W-202 2-35 II Rt!RS X X X N/A 2-36 H Ri!RS/SDC X X X l 2305-M-207 2-37 H HPCI X X X N/A 2-38 H Press Supp X X X N/A 2-39 H SBGTS X X X N/A 2-40 H SBCTS X X X N/A 2-41 H RCIC X- X X l

N/A 2-42 II RCIC X X X N/A 2-43 M Main Steam X X N/A 2-44 Note (4)

M Main Steam X .X N/A 2-45 Note (4)

M Main Steam X X l N/A 2-46 M Note (4)

. Main Steam X X N/A 2-47 Note (4)

M Main Steam X X N/A 2-48

, Note (4)

M Main Steam X X N/A 2-49 Note (4)

M Main Steam X X N/A 2-50 Note (4)

M Main Steam X X N/A 2-51 Note (4)

M RIIRS/ Head Spray X X 1403-M-212 2-52 M Core Spray ,

X X '

NOTES: (1) Test as required by IWP 5400(b)(1)

(2) Test as required by IWP 5400(b)(2) (4) Snubber 50 kips or Greater - No test specification per (3) Test as required by IWP 5400(b)(3) IWF-5300. Will be tested for full stroke, freedom of movement in compression & tension.

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@c Edson SNUBBER TESTNG (lWF-5000)

DUAD CITES NUCLEAR POWER STATION UNIT 2 g 3 gg 4 i ..s/1s/s2 TECH SPEC SNUBSER TEST PARAMETERS ISI SUPPORT NUMBER SNUSSER M TYPE SYSTEM REMARKS (2) Q) 1403-M-213 2-53 M Core Spray X X 3001A-N-107 2-54 H Main Steam I X X 3001A-H-100 2-55 H Main Steam X X 3001A-M-109 2-56 M Main Steam X X N/A 2-57 M Main Steam (SRV disch) X X N/A 2-50 M Main Steam (SRV disch) X X -

N/A 2-59 M Main Steam (EMRV disch) X X N/A 2-60 M Main Steam (EMRV disch) X X 30018-M-106 2-61 M Main Steam X X 3001D-M-107 2-62 H , Main Steam X' X N/A 2-63 H Main Steam (EMRV disch) X X

N/A 2-64 M Main Steam (EMRV disch) X X N/A 2-65 M Main Steam (EMRV disch) X X 3001C-N-106 2-66 M Main Steam X X-

, 3001C-M-107 2-67 M Main Steam X X 3001C-M-100 2-68 M Main Steam X X N/A 2-69 M Main Steam (EMRV disch) X X l

N/A 2-70 M Main Steam (EMRV disch) X X 3001C-M-105 2-71 M Main Steam X X 3001C-M-106 2-72 M Main Steam X X -

3001C-M-107 2-73 M Main Steam X X N/A 2-74 M Main Steam (EMRV disch) X X N/A 2-75 M Main Steam (EMRV disch) X X 2305-M-104 2-76 M IIPCI X X 2305-M-105 2-77 M IIPCI X X 2305-M-106 2-78 M IIPCI #

I X X NOTES: (1) Test as required by IWF S400(b)(1)

(2) Test as required by IWF 5400(b)(2) (4) Snubber 50 kips or Greater - No test specification per (3) Test as required by IWF 5400(b)(3) IWF-5300.

movementincompressYon&Will be t sted for full stroke, freedom of tension.

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O O O m,,,,,, INSERVICE- TESTING PROGRAM

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SNUBBER TESTN3 (IWF-5000)

DUAD OTES NUCLEAR POWER STATION UNIT - 2 Pg 4 W 4 Revision ? 1 ' lDete s/1a/82 TFOt SPEC SNUSSER TEST PARAMETERS ISI SWPORT NUMSER $NUBGER M SYSTEA4 TYPE g g REMARKS 2300-M-107 2-79 M Recirc X X -

2-80 M Recirc X X 0200-M-170 2-81 M Recirc X X 2-82 M Recirc X X 1025-M-105 2-83 H RilRS/SDC 5C X 1025-M-105 2-84 M RIRS/SDC X X 1025-M-106 2-85 M RIIRS/SDC X X 1025-M-106 2-86 M R11RS/SDC X X 1202-M-115 2-87 M RWCU X X

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1202-M-115 2-88 M ,

RWCU X X

3001B-M-108 2-89 M, Main Steam X X i

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NOTES: (1) Test as required by IWP 5400(b)(1) (4) Snubber 50 kips or Greate r - No test specification per (2) Test as required by IWP 5400(b)(2) IWP-5300. Will be tested for full stroke, freedom of (3) Test as required by IWF 5400(b)(3) movement in compression & tension.

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SECTION 3.3 RELIEF REQUESTS FOR SNUBBER TESTING PROGRAM

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COM-0 6-003 Revision 1 3-6

RELIEF REQUEST NO. CSR-1 i() '

SUPPORT NUMBERi ~All Mechanical and Hydraulic Snubbers in the Program

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SECTION XI"REQUIREMENTi Snubbers shall be tested to ensure that during low velocity displacements, the specified maximum drag will initiate movement in both tension and compression.and that activation is achieved within the specified range of velocity in both tension and compression.

BASIS'FOR'RELIEFi Until snubber testing equipment is ,

{} commercially available on a competitive basis it is not possible to test Cnubbers as per IWF-5400(b)(l&2). It is also not practical to remove the snubbers and send them to an off-site testing facility because of time involved with decontamination and transportation to and from the site. Alternative testing will be used on.an interim basis until snubber testing equipment is available, at which time, testing will be to the IWF Section of the ASME Code.

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, COM-06-003

! Revision 1 3-7

ALTERNATIVE TESTING:' Snubbers which are to be tested will be fully cycled in both tension and compression to ensure freedom of movement and inspected for mechanical soundess. The;e tests will be done manually.

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O COM-05-003 Revision 1 3-8

l 4.0 INSERVICE TESTING PROGRAM FOR PUMPS

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4.1 GENERAL INFORMATION i

.The Inservice Testing'Progrmn for ISI Class 1, 2 and 3 pumps meets .tdte requirements of Subsection IWP of Section XI of the ASME Boiler and Pressure Vessel. code, 1980 Edition through the i

Winter 1980 Addenda. Where these requirements are determined to be impractical, specific. requests for relief have been written.

The tables in Section 4.2 list all Class 1, 2 and 3 pumps which are within the scope of IWP-llOO to be tested along with the parameters to be measured for each pump unless reference is made to a relief request. Section'4.3 includes all relief requests

.({}

. referenced in the tables plus any additional relief requests that

(

( are generic to the pump testing program.- Table 4.2-1 lists the systems and their respective P&ID numbers which are covered in the pump testing program.

It should be noted that pump speed' is not measured for synchronous type pumps per IWP-4400. Where pump suction is from

! a tank or the river, inlet pressure will be calculated from the measured tank or river level.

COM-06-003 Revision 1 4-1

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l SECTION 4.2 TABLES FOR INSERVICE PUMP TESTING PROGRAM A. QUAD CITIES UNIT-1 i

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[' B.- QUAD CITIES UNIT-2 l

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COM-06-003 Revision 1 4-2 i

TABLE 4.2.la i

LIST OF SYSTEMS INCLUDED IN THE PUMP PROGRAM FOR UNIT-1 SYSTEM REFERENCE SYSTEM NUMBER P&'ID Core Spray 1400 36 Residual Heat Removal 1000 37 & 39 Residual Heat Removal Service Water 1000 37 & 39 Standby Liquid Control 1100 40 High Pressure Coolant Injection 2300 46 s Diesel Generator Cooling Water 3900 22

' ('~/

Diesel Generator Fuel Oil Transfer 5200 29 COM-06-003 Revision 1 4-3

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INSERVICE TESTING PR2 GRAM UNIT - 1

@ Com;onwealth Edison ISI- C L A S S 1. 2. & 3 PUMPS QUAD CITIES NUCLEAR POWER STATION REW hjh82 1o 1 N PfrID AND TEST PARAMETERS PUMP NUMBER PUMP NAME = TEST Eg COORDINATES Diff SPEED INLET fl0W VIRRAT10N REARING INTERVAL PRES PRES RATE TEMP 1A-1401 CORE SPRAY 2 36 E-9 NO YES YES YES PR-1 PR-1 QUARTERLY 1B-1401 CORE SPRAY 2 36 E-6 No YES YES YES PR-1 PR-1 QUARTERLY 1A-1002 RESIDUAL IIEAT REMOVAL 2 37 B-4 No YES YES YES PR-1 PR-1 QUARTERLY IB-1002 RESIDUAL HEAT REMOVAL 2 37 E-4 NO YES YES YES PR-1 PR-1 QUARTERLY IC-1002 RESIDUAL IIEAT REMOVAL 2 37 B-8 NO- YES YES YES PR-1 PR-1 QUARTERLY ID-1002 RESIDUAL IIEAT REMOVAL 2 37 E-8 NO YES YES YES PR-1 PR-1 QUARTERLY 1-1001-65A RIIR SERVICE WATER 3 39 F-4 NO YES YES YES PR-1 PR-1 QUARTERLY 1-1001-65B RIIR SERVICE WATER 3 39 F-4 NO YES YES YES PR-1 PR-1 QUARTERLY 1-1001-65C RilR SERVICE WATER 3 39 P-7 NO YES YES YES PR-1 PR-1 QUARTERLY 1-1001-65D RitR SERVICE WATER 3 39 F-7 NO YES YES YES PR-1 PR-1 QUARTERLY 1A-1102 STANDBY LIQUID CONTROL 2 40 D-7 NO PR-3 PR-3 YES PR-1 PR-1 QUARTERLY 1B-1102 STANDBY LIQUID CONTROL 2 40 E-7 NO PR-3 PR-3 YES PR-1 PR-1 QUARTERLY 1-2302 IIIGli PRES COOLANT INJ 2 46 A-4 YES YES YES YES PR-1 PR-1 QUARTERLY 1-3903 D/G COOLING WATER 3 22 A-10 NO YES YES YES PR-1 PR-1 QUARTERLY 1/2-3903 D/G COOLING WATER 3 22 A-10 NO YES YES YES PR-1 PR-1 QUARTERLY 1-5203 D/G FUEL OIL TRANSPER NC 29 F-3 NO PR-4 PR-4 YES PR-1 PR-1 QUARTERLY 1/2-5203 D/G FUEL OIL TRANSFER NC 29 F-3 NO PR-4 PR-4 YES PR-1 PR-1 QUARTERLY

___= _

NOTE: Lubrication levels will be observed diiring each inservi ce test f or pumps that are designe<. such th<it levels can be verified. Tl e core spray (1401), RIIR (1002), and 1.he D/G fuel all trans fer (520 l} pumps are lubri cated by pump flosage and, thus, lubrict nt level or pressure measuremen'.s arit not relevarit.

TABLE 4.2.lb LIST OF SYSTEMS INCLUDED IN THE PUMP PROGRAM FOR UNIT-2 .

SYSTEM REFERENCE

. SYSTEM NUMBER P&ID Core Spray 1400- 78 Residual Heat Removal 1000 79 & 81 Residual Heat Removal Service Water 1000 79 & 81 Standby Liquid Control 1100 82 i

High Pressure Coolant Injection 2300 87

() Diesel Generator Cooling Water 3900 5200 69 29

[

Diesel Generator Fuel Oil Transfer l

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(]h l COM-06-003 Revision 1 4-4 l

g Commonwealth INSERVICE TESTING PREGRAM UNIT - 2 Edison ISI- CL A S S 1, 2. & 3 PUMPS QUAD CITIES NUCLEAR POWER STATION RE -

8 8/82 1o 1

=> TEST PARAMETERS PUMP NUMBER PUMP NAME g P&lD AND TEST g COORDINATES INtET DIFF FLOW BEARING INTERVAL SPEED VitRATION PRES PRES 2 ATE TEMP 2A-1401 CORE SPRAY 2 78 E-9 NO YES YES YES PR-1 PR-1 QUARTERLY .

2B-1401 CORE SPRAY 2 78 E-6 NO YES YES YES PR-1 PR-1 QUARTERLY 2A-1002 RESIDUAL llEAT REMOVAL 2 79 B-4 NO YES YES YES PR-1 PR-1 QUARTERLY 2D-1002 RESIDUAL HEAT REMOVAL 2 79 E-4 NO YES YES YES PR-1 PR-1 QUARTERLY 2C-Ir 12 RESIDUAL ilEAT REMOVAL 2 79 B-8 NO YES YES YES PR-1 PR-1 QUARTERLY 2D-1002 RESIDUAL IIEAT REMOVAL 2 79 E-8 NO YES YES YES PR-1 PR-1 CUARTERLY 2-1001-65A RilR SERVICE WATER 3 81 F-4 NO YES YES YES PR-1 PR-1 QUARTERLY 2-1001-65D R!lR SERVICE WATER 3 81 F-4 NO YES YES YES PR-1 PR-1 QUARTERI.Y 2-1001-65C R!lR SERVICE WATER 3 81 F-7 NO YES YES YES PR-1 PR-1 QUARTERLY 2-1001-65D Ri!R SERVICE WATER 3 81 P-7 NO YES YES YES PR-1 PR-1 QUARTERLY 2A-1102 STANDBY LIQUID CONTROL 2 82 D-7 NO PR-3 PR-3 YES PR-1 PR-1 QUARTERLY 2D-1102 STANDBY LIQUID CONTROL 2 82 E-7 NO PR-3 PR-3 YES PR-1 PR-1 QUARTERLY 2-2302 IIIGli PRES COOLANT INJ 2 87 A-4 YES YES YES YES PR-1 PR-1 QUARTERLY 2-3903 D/G COOLING WATER 3 69 A-10 NO YES YES YES PR-1 PR-1 QUARTERLY 2-5203 D/G FUEL OIL TRANSFER NC 29 P-3 NO PR-4 PR-4 YES PR-1 PR-1 QUARTERLY

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.==. .___ .---- . -------- ......

NOTE: Lubrii:ation levels will be observed during each inservsce test for pumps that ar e designell such tlat levels can be verified. T16e core spray (1401), RilR (1002), and the D/G fuel oil transfer (52( 3) pumps are lubricated by pump fic, wage and, thus, lubric< int level or pressure measuremer to ara not releva it.

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SECTION 4.3 RELIEF REQUESTS FOR INSERVICE PUMP TESTING PROGRAM l

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l COM-06-003 Revision 1 4-5

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RELIEF REQUEST NO. PR-1 O

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PUMP NUMBER: All pumps in program.

SECTION'X1" REQUIREMENT':~ IWP-1500, Detection of Change  ;

BASIS'FORRELIEFi Pump vibration and bearing temperature are required to be measured to detect any changes in the  ;

mechanical characteristics of a pump. This is to detect developing problems so repairs'can.be initiated prior to a pump becoming inoperable (i.e. unable to perform its function). The ASME Code minimum standards require measurement of the vibration amplitude displacement in mils (thousands of an inch) every three months and bearing temperatures once per year.

t Quad Cities Station proposes an alternate program which is believed to be more comprehensive than that required ,

! by Section XI. This program consists of performing the required vibration readings in velocity rather than mils displacement. This technique is an industry-accepted method which is much more meaningful and sensitive to small changes that are indicative of developing mechanical problems. These velocity measurements detect .

1 not only high amplitude vibrations that indicate a major mechanical problem but also the equally harmful low

()

COM-06-003 Revision 1 4-6

-, .,-----n , , - - - - - . . . , - - . -. . , ..- ., .

RELIEF REQUEST NO. PR-1 (CONTINUED)

-( )

amplitude - high frequency vibrations due to misalign-ment, imbalance, or bearing wear that usually go undetected by simple displacement measurements.

1 In addition, these readings go far beyond the capabilities of a bearing temperature monitoring program, which requires a bearing to be seriously degraded prior to the detection of increased heat at the bearing housing. The vibration velocity readings on a schedule of once every three months achieves a much higher probability of detecting developing problems than the once per year reading of bearing temperatures. Data gathering on bearing temperatures also is not without its own problems. The enforced thirty minute run time, (i.e. IWP-3500 (b) - three successive readings taken at ten minute intervals that do not vary more than 3%),

causes problems with pumps having no recirculation / test loop. It is easy to see that a program of bearing l temperature trends and the evaluation of the results would in some cases be difficult to analyze. Improper

. interpretation of results could result in unnecessary pump maintenance. In addition, it is impractical to lO l COM-06-003 l Revision 1 4-7 1'

RELIEF REQUEST NO. PR-1 (CONTINUED) measure bearing temperatures on many of the pumps in the-program. 'Some specific examples are as follows:

(1) Core' Spray 1(2)A,B-1401 - pump bearings are lubricated by pump flowage. Temperature of the pumped liquid would seriously affect the accuracy of trends.

(2) RHR 1(2)A,B,C,D-1002 - same as above.

(3) RHR SERVICE WATER 1(2)1001-65A,B,C,D -

Bearings are contained in an oil-filled reservoir. The

-ambient temperature of the pump space is changeable thereby varying the start temperature of the data. Results would be difficult if not i impossible to trend.from test to test.

(4) High Pressure Coolant Injection -

this pump is driven by a steam turbine which exhausts steam into the pressure suppression chamber. Extended run times to stabilize bearing temperatures would create problems in keeping suppression pool l-l temperatures below the Technical Specification l

limit of 95*F.

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COM-06-003 i Revision 1 4-8 1

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6

RELIEF REOUEST NO. PR-1 (CONTINUED)

() -

, (5) Didsel'Gsn5rator~ Cooling'Wsh6r'l(2)(1/2)-39'03 -

Same as RHR Service Water (6) Diesel' ' ensrdtor G ' Fuel' ' Oil ~ ~Transf 6r' 1( ~2 )Yl/2 i-5203

- this transfer pump pumps fuel oil from tne fuel oil storage tank to the D/G fuel oil day tank. '

There is no recirculation test loop for these pumps, thereby, limiting the run necessary to gather bearing temperature data.

The foregoing reasons demonstrate that the proposed program of vibration measurements is a more practical j}

method of testing which exceeds the requirements of the l ASME Code.

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ALTERNATE TESTING: Pump vibration measurements will be taken in vibration velocity (in/sec). The evaluation of the I readings will be per the attached table.

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COM-06-003 Revision 1 4-9

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

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o o ALIDOBM RANGES OF VIBRATION V 2DCITY1 to o

<3 yi AEIC' RAW 2 REQUIRED' ACTIN RANGE

! rm IIM HIG IIM HIGH 0$o GIANPITY ACCEFfABW RANGE VAWES VAWES VAWES VAWES H Lab v When 0 < vr < .15 in/sec 0 to .3 in/sec. None .3 in/sec to None v > .45 in/sec v When .15 in/sec < vr 0 to .45 in/sec None .45 irt/sec to None v > .75 in/sec

< .3 in/sec .75 in/sec v Wen .3 in/sec < vr 0 to 0.9 in/sec None 0.9 to 1.5 lbne v > 1.5 in/sec

< .6 in/sec in/sec v When .6 ir./sec < vr 0 to 1.1 in/sec None 1.1 to'1.5 None v > 1.5 in/sec

< l.0 in/sec in/sec

& Where:

o v = velocity measured in inches /second, peak.

! vr. = reference velocity measurement (initial measurement after installaticn or rework.

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i 1 See ASME Technical Paper 78-WA/NE-5, 'hble 2.

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RELIEF REOUEST NO. PR-2 PUMP NUMBERi -All pumps in program.

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SECr10N'XI'RnOUIREM2NTi The requirements of IWP-3230(c).  ;

1, corrective ~ action.

h BASIS FOR' RELIEF: Relief is requested from the requirements of l.

L sstablishing a new set of reference values following an

+  :

analynis-of a deviation which falls into the " Required

Action Range." It is felt that osing the original i

values as a reference will alert the testers to future I

degradation of the pump more quickly than using a new cet-of reference values. Pumps which fall into the l

l " Required Action Range" will be analyzed and will'reraain 1

operative as long as pump operability is not impaired and pump performance meets Technical Specification limits. Maintenance will.be performed at the first opportunity to rectify the deviation.

ALTERNATE TESTINGi Pump testing will be doubled (i.e., every 45

' days) during the interim period and a new set of reference values will be established following the correction of the deviation.

~ Testing will then return to the original frequency (i.e., every 92 days).

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COM-06-003 Revision l' 4-11

RELIEF REQUEST No..PR-3 10 PUMP NUMBER: 1(2)-llD2 (SBLC) i L _

SECTION XIREQUIREMENTi licasare pump inlet pressure  ;

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BASIS FOR'RELIEFE It it impractical to measure standby liquid control pump inler pressure in accordance with Section ,

XI requirements. .During pump testing,the pump suction i

! is from a test tank rsther than the rhain standby liquid L control rank. Ho instrumentation is provided for ,

r measuring inlet pressure, and therefore, the only means

( .I available is to correlate tank-level to inlet ,

pressure. Since these pumps are. positive displacement- ,

designs, the measurement of inlet pressure is not critical in judging pump performance. .. Measuring the j discharge pressure and the flow rate is adequate to detect changes:in the hydraulic characteristics of the pumps.

ALTERNATE' TESTING'i Pump discharge pressure will be monitored at each inservice test.

O COM-06-003 Revision 1 4-12 1 -- . - - .. - - ._. ._ - . _- ~ _. - . . . . . -, . . . . - . .-

RELIEF REQUEST NO. PR-4 PUMP NUMBER: 1-5203, 2-5203, 1/2-5203

,SECTION XI REOUIREMENT: Measure Pump Inlet Pressure BASIS FOR RELIEF: Relief is requested from the requirement of I measuring pump inlet pressure during pump tests. This 1

pump is utilized in transfering fuel oil from the diesel generator fuel oil storage tank to the diesel fuel oil The configuration of the piping is such that H day tank.

the pump is 1ccated above the storaga tank. The pump is l'

a positive displacement gear type pump uct requiring a I positive suction head for proper operation. Since this pump is a positive displacement type, the discharge l pressure is independent of the suction pressure and, l

-therefore, inlet pressure data is not important in evaluating pump performance.

l ALTERNATE TESTING: Pump discharge pressure will be monitored at each inservice test.

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l COM-06-003 Revision 1 4-13 l

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1 5.0- INSERVICE TESTING PROGRAM FOR VALVES O

5.1 ~ GENERAL INFORMATION The Inservice Testing Program for ISI Class 1, 2 and 3 Valves meets the requirements of Subsection IWV of Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition through the Winter-1980 Addenda. Where these, requirements are determined to be impractical, specific requests for relief have been written and included in Section 5.3.

The tableir in Section 5.2 list nJ1 ISI Class 1, 2 and 3 valves that have been assigned valve categories; valves exempt per IWV-1200 are not listed. The tables are organized by system in order of the assigned system number. A list of these systems and their respective P&ID numbers is given in Table 5.2-1. The following information is included in the tables:

l A. Valve Number lists the valve' identification number I as shown on the color-coded P& ids. The first l

I digit of the valve number indicates the i

appropriate unit.

1o COM-06-003 Revision 1 5-1

B. P&ID and Coordinates references the color-coded

() P&ID on which the valve appears and its coordinates.

C. Class is the ISI Classification of the valve.

Valves in the diesel fuel oil and air start systens as well as some primary containment ,

i isolation valves are included in the program, even though they do not have an I.ST Classification.

I These valves are designated 23 Class-NC (Not Classified).

D. Valve"Categdry indicates the category assigned to the valve based on the definitions of IWV-2200.

l

() Note that the Category E valves, valves which are normally locked (or sealed) open or locked (or-sealed) closed to fulfill their function, are not within the scope of Subsection IWV. They are listed in the Tables for information only.

E. Valve'S'ize lists the nominal pipe size of the valve in inches.

F. Valve Type lists the valve design as indicated by the following abbreviations.

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O COM-06-003 l Revision 1 5-2

GATE GA GLOBE GL CHECK CK SAFETY SV RELIEF RV ELECTROMATIC RELIEF ERV 4

BUTTERFLY BTF 1

STOP CHECK SCK BALL BALL d

RUPTURE DIAPHRAM RPD EXCESS FLOW CHECK XFC i

G. Actdat6r'Tvns lists the type of valve actuator as indicated by the following abbreviations.

MOTOR OPERATOR MO AIR OPERATOR AO SOLENOID OPERATOR SO PILOT SOLENOID ACTUATOR PS

• EXPLOSIVE ACTUATOR EXP

l. SELF ACTUATED SA MANUAL M H. Normal'Pos'itib'n indicates the normal position of the valve during plant operation. This is O

COM-06-003 Revision 1 5-3

i specified as open (O),' closed (C),-locked open l

I) (LO), and locked closed (LC).

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! I. Stroke Direction indicates the direction which an l l

, active valve must stro'ke to perform its safety '

L function. Also,.the direction in which the valve <

will be stroked to satisfy the exercising

[ requirements of IWV-3410 or IWV-3520. This may be specified as open (O), closed (C),-or both (O&C).

i' J. Test lists the test or tests that will be

, performed for each valve to fulfill the requirements of Subsection IWV. The following tests and abbreviations are'used

Seat Leak Test (IWV-3420)

a. Type C Air Leak Test - Primary contain- AT-1 ,

ment isolation valves. The acceptance criteria for this test is defined in Relief Request VR-11. Note that 10CFR50 Appendix J, Type A & B tests are not included in this program even though thest '.sts will be conducted.

!O COM-06-003 Revision 1 5-4

b. Excess Flow Check Valve Test - Excess I.T-2 flow check valves will-be tested in accordance with the Technical Specifi-t cations requirements. For further ,

.information see Relief Request VR.-9.

i-Full Stroke Exercise Test (IWV-3411, BT

• I 3412, 3413)

Valve will be full stroke exercised for operability in the direction necessary to i fulfill its safety function.  ;

Partial Exercise Test (IWV-3412(a)) BTP l.

)

Valve will.be part-stroke exercised when full-stroke exercising is impractical.

Check Valve Exercise Test (IWV-3520) CT-1 Check valve will be exercised fully open, closed or both depending on the safety function of the. valve. Verification of acceptable system flow through a valve l

shall be adequate demonstration of valve operability.

O COM-06-003 Revision 1 5-5

l Relief Valve Set Point Check (IWV-3510) CT-2

(_/ Relief and safety valve set points will be verified in:accordance with IWV-3510.

Explosive Valve Tests (IWV-3610) DT Explosive valves will. be tested in accordance with IWV-3610.

l l

Fail-Safe Test (IWV-3415) FST All valves with fail-safe actuators will be tested to verify proper fail-safe operation upon loss of actuator power.

Position Indication Check (IWV-3300) PIT O All valves with remote position indicators will be checked to verify that remote valve indications accurately reflect valve operation.

l K. Test Mode indicates the frequency at which the above mentioned tests will be performed. The following abbreviations are used:

Normal Operation OP I

Tests which are conducted at least once every 3 months during normal plant operation.

l

()

COM-06-003 Revision 1 5-6

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

l-l-

Cold Shutdown CS l

~

• I Inservice valve testing at cold shutdown is valve-testing which commences within two i

. hours after the plant reaches-a cold shutdown-condition but in no' case later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after cold-shutdown is_ reached.'- This testing continues'until all' valves are tested or the unit'is ready for start-up. Completion of all testing is not a prerequisite to plant start-up. Valve testing which is not completed during a cold shutdown shall be completed during subsequent cold shutdowns-that may occur before refueling to meet the code specified testing frequency. In the O case of frequent cold shutdowns, valve testing need not be performed more often than once every three months for category A, B, and C valves.

l t

In the case of longer planned cold shutdowns, the testing.need not be started within the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> limitation. However, in these I instances, all valve testing must be i

completed prior to start-up.

l Note: It is expected that the required testing will normally be completed in 96

.( J COM-06-003 Revision 1 5-7

i hours following cold shutdown. However,.

). completion of all valve testing during. cold  !

shutdown is not required if plant operating  !

. conditions will not permit the testing of specific valves.

! In the event that a valve must be declared inoperable as a result of cold shutdown b

testing,-the applicable unit start-up limitations will be as stated in the Tech-nical Specification, Limiting Conditions for

i. Operation.

. RiactoW ' Rsf uel'ing - RR

() Tests which are conducted during plant i

refueling outages but not less than once every two years.

Every'S'Y ars SY Tests which are conducted during plant refueling outages.but not less than once every five years (See Article IWV-3511).

L. Msk"Stioks"T'ime lists the maximum allowable full stroke time in seconds for power operated valves in Category A or B.

O COM-06-003 Revision 1 5-8

M. Relief Request references ^either'a specific request

() contained in Section 5.3 or references a cold Shutdown Justification contained in Section 5.4.

Also, included in Section 5.3 are generic relief requests that are pot specifically referenced in this column'of the tables, but apply to the valve program in general.

N. Remarks lists' clarification remarks or indicates that a valve receives an automatic isolation -

signal. See Table 5.1-1 for the explanation of isolation valve groupings. ,

i i-

-( ) I .

I l

l l

1

()

COM-06-003 Revision 1 5-9 l

8

< TABLE 5.1 i

-(3 sj' AUTOMAT.IC ISOLATION VALVE GROUPINGS  ;

i Groap 1: The valves in. Group 1 are closed upon any one of the j following' conditions:

l L 1. Reactor icw-low water level l

2. Main ateamline high radiation I l 3. Main staamline high flow

4. Main steamline tunnel high temperature

5. Main stsamline low prosaure-I j' Group 2: The actions'in Group 2 are initiated by any one of the l

I following conditions:

I

!- 1. Reactor lov water level l

2. High dryvell pressure Group 3: Reactor lov vater level alone initiates.the following:

-( ) 1. Cleanup derineraliser system isolation Group 4: Isolation valves in the high pressure coolant injection system (HPCI) are closed upon any-one of the l

)

following signals:

1

1. HPCI_steamline high flow

2. High temperature in the vicinity of the HPCI steamilne l

3. Low reactor pressure I

Group 5: Same as Group 4 except applies to RCIC.

1 l '

l i

, i l-O COM-06-003 Revision 1 5-10  :

l J

,5 s

f .'-

l '.

l l . SECTION 5.2 1

!- ' TABLES FORl INSERVICE VALVE TESTING PROGRAW l.

A. QUAD CITIES UUIT-1 ,

I l

B. GUAD CITIES-UNIT-2 l l

l l

1 r;

O- l i

i t

f i

O COM-06-003 i Revision 1 5-11 i

P-Wg'.99-*qrW=VMgWm u 7+7-tymw gy

D a- .

TABLE 5.2-1

. ' LIST OF SYSTEMS INCLUDED IN THE VALVE PROGRAM .

UNIT-1 UNIT-2 SYSTEM' REFERENCE REFERENCE

. GYSTEM NUMECR P&ID P&ID Nuclear Boiler 0200 35-1 77-1 Recirculation 0200 35-2 77-2 Control Rod Drive 0300 41 83 Residual Heat Removal 1000 37&39 79&81- l 4

1100 40 Standby Liquid Control ,

2 1200 47 88

-( ) React r Water Cleanup Reactor Core Isolation Cooling 1300 50 89 Core Spray 1400 36 78 Pressure Suppression 1600 34 76 High Pressure Coolant Injection 2300 46 87 i

Main Steam 3000 13-l&2 60-1&2 i

Feedwater 3200 15 62 Service Water 3900 22 69 Instrument Air 4700 24-2 71-2 Diesel Air Start 4600 25 72 Rx Building Equipment Drains 4800 43 85 Diesel Fuel Oil 5200 29 29 l

O COM-06-003 Revision 1 5-12 l

= x

. _ g- _

O 7 g PROGRAM UNIT - 1 TESTING VALVES

- - INSERVICE 1. 2 & "OWER 3 ST ATION q ISI-CLASS PAGE Commonwealth CIT IE S NUCLE /R ,m m_= ____ =m .

P 6 60 I REVISION - GATE 1 of 36 QUAD _____ 1 - 8/18/82 Edison ISI- 35 Sh. I jp_/

L,--__.-y ION ,

/ /

SySygg NUCLEAR BOILEH INSTRUMENTAT , , "'"^""5 l

g

1. / (( (-

eAmt muusta

[/ - q RR

/

I- ,

f VR-9 i

r l AT-2 __ ._ ________-_- ___

CT-1 RR ._____ __. __

C ___._______. ____ g O . _ -

SA ______. _ _ _ _ _

O.5 XFC __ __ _ .AT-2 RR  ! V R-9 g -_ __ - ____-________

1 AC 1R .______.f I -263-2-15A D-5

__,. __ _______._____ ._____- .SA_____.

• O C

CT-1.__._..___.sL_____b

_____________ O.5 XFC _______, ___ .__ ___ AT-2 )

LE VR 9

-263-2-13A _ __..

D-5 1 AC

. _ _.______.,. _____. ___ . O C CT-1 < 1G L__ _ _.______.__ __ ____ _ ____________

SA 1

AC O.5 XFC

_AT-_ _ _____ j Rh_2_ __I _Jf V R-9 C-5 1 ._______ CT-1 l RR . _ _ _

C y __ __

1-263-2-19A ______. ____ _ ______-.__ __

O SA __ __ ..__-__._.-_ AT-? _

_____ XFC RR VR-9 -___ __ __

1 AC 0.5 _ __ _ -- __

CT_1 KR __ ...

D-5 ______.___ _ _____ _

I-263-2-17A ___ -_. SA O ~~C .___.. __., _____ Rn I _ _-

______~.__ _ - 0.5 XFC ._____. AT-2 V K-9 _____ __ ______

AC RE [

l-263-2-11 E-5 ___

1 O

l C CT_1

____ .____. ~ _ _ _ _ _ . _ . .

SA /,__ ___ ____ . . -,

__ _____ ___..____. - SPC VR-9 E-5,6 1 AC O . '2

__ __ ._- __._.______.- ___ . RP l

__.______. O C lCT? ___ _ _____

SA _ -__

1-220-54_ _ _ - _ _ _ _ _ . _ O.5 XFC l VR-9 ________________ ____.

D-3 1 AC

_.. ___ .__-. __ _ ____._ O

___ C C T-: JpH _- ___

-_ __t____._ ...

SA _ _ _ .

f 1-263-2-ISu_________ __. AC __- _ __ O.5___.___ XFC _ _-_-__ _ ___ _______ AT -2 #

KR ' V9-9 b -_ -___ ________--.

D-3 1

__ ___________._____ j O c eT-1

__-__p._...____.....__

RR I I

SA 1-263-2-13B_______-_-_____ ;XFC _ _.._.. ____ _

. . . - _ RL l l Vn-9 D-3 1 AC O.5 ..- -

C cT-1 lAT-2 ._.-__ L _____9 W l .. .____.__ ______ __ __________

O I-263-2-17D_ - _ _ - _____ . . _ _ _ . 0.5 FC SA

______. _ _ _ _ _ _ _ . . . f _ ER _ _ _ fwl R R Vet 9. -._ -__-___.____ _ ______

AC ._____

C-3 1  ; FAT-4 CT ...__y._. . __.

l-263-2-19H .______ _______ .__~_ SA O j r

._ _ j _ - 1 _

~

____-____________ _ O.5 XFC AT-2 -- l RR VR-9 ____...__. .. _______

1 AC l ( P-1 RP l i u-5 I-263-2-20A SA O

C

_...._ _____..-m-L______.____..  ! -

AC O.5 XFC

_______{.... AT-2 i RR VM-9 ___________- _____.'

C-5 1

_______.. .._.. _ ___ CT-1 1 - 2t> 3 2 3 A_________

___--____ XFC SA O

C q'RD ,

' \

AC O.5 1

C-5 1-263-2-31D

l. . , T

~

a -

^

INSERVICE TESTING PROGRAM

_@ Commonwealth Edison ISI- CL ASS QUAD CITIES NUCLEAR POWER ST ATION

1. 2. & 3 VALVES UNIT - 1 SYSilM P 6 ID REVISON - DATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-35 Sh. I fI- 8/18/82 2 of 36

/

~c ~~

l / // p# p /* // 4;/ #

p /l p/,fh//

1-263-2-31G C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9 1-263-2-31C C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 VR-9 g RR 1-263-2-3111 C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9 -

1-263-2-31D C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1

- - - - - - - = -

RR VR-9 1-263-2-27 A-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 PR VR-9 1-263-2-25 D-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR

' ' ~ ~ ~ ~ ~ ^ - -

VR-9 AT-2 263-2-31J C-5 ER 1 AC O.5 XFC SA O C CT-1 RR VR 9

....... = _ _ . . . . . . . __

l-263-2-31r. C-5 AT.2 RR 1 AC O.5 XFC SA O

~ -

~ --

C CT-1 RR VR-9 1-263-2-31K C-5 AT-2 En 1 AC O.5 XFC SA O C CT-1 RR VR-9 l

...a...- -

I 1-263-2-23D C.5 AC AT-2 RR 1 O.5 XFC SA O C CT-1 RR

..... =_- _

VR-9 1-263-42A C-5 1 AC AT-2 RR O.5 XFC SA O C CT-1 RR VP-9

......y....4...... ..-

1-263-2-20B B-5 AT-2 RR I 1 AC O.5 XFC SA O C CT-) RR Vh-9 1-263-2-20C B-3 AT-2 RR 1 AC XFC SA O l

_- -- .-_ I. 0.5 C

.. ... .C.T..1.. . ..R.k. VR-9

. . ..l ... -- -__.

t

h Commonwedth INSERVICE TESTING PROGRAM Edison ISI - CL ASS 1. 2. & 3 VALVCS UNIT - 1 QUAD CITIES NUCLEAf1 P O W Eft S T ATIGli SYSTEM P b 10 REYtSiON -

DATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-35 Sh. 1 .),- 8/18/82 3 of 36

/ /

v ~ ~~ ~

l / // l p# /If;/fll4'lh// ./,s  !

p

/

""*~

1-263.2-23C C-3 AT-2 F.2 AC O.5 1 XFC

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

SA O C CT.* i RR VR-9

, _ , , , _, t.,_,_, ,,, _

1-263-2-31M C AT-2 PR

.._-3 1 AC O.5 XFC SA O C CT-1 RR

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

VR-9 1-263-2-31T C.3 t AT-2 RR 1 AC O.5 XFC SA O l C CT-1 RR

-. J.. . -

VR-9 1-263-2-31N C-3 AT-2 R9 1 AC O.5 XFC SA O C CT-1 RR VR-9 1-263-2-31U C-3 AT RR 1 AC O.5 XFC SA O C CT-1

= .

RR VR-lO 1-263-2-31P C-3 AT ltR 1 AC O.5 XFC SA O C CT-1 RR VR-10

. .......a.. . - _ . . ......- -

1-263-2-33 D-3 AT RR 1 AC O.5 XFC SA O C CT-1 RR VR-lO 1-263-2-31V C-3 AT RR 1 AC O.5 XFC SA O C CT-1 RR

_. __ . . _ - . _ , , _ , _ _ , , _ , .,____,_, __ VR_ -10 ,

1-263-2-31R C-3 AT RR 1 AC O.5 XFC SA O C CT-1 RR VR-10 1-263-2-31W C-3 AT RR 1 AC O.5 XFC SA O C CT-1 RR

---

==- - -

VR-10 1-263-2-23D C.3 AT RR 1 AC O.5 XFC SA O C

-_-- =-__. _.

CT-1 RR g VR 10 1-263-2-42n C-3 AT RR 1 AC O.5 XFC SA O C CT-1 d

'R

_-_ -_ VR-10

• 1 263-2-20D D-3 AT RR 1 AC O.5 XFC SA

_ , , ., _O C__ CT-1_ ,R R _ ,, _ _ _ , , , , , _VG 1,0,,_ _ ,_, ,

_ _ _ . _ _ _ _ _ . _ _ _ _ _ - _ - - - - - - - - - - - - - - ^ - - - - - - - - - - - - - - - - - - - - ^ - - - - - - - -~ --

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CL ASS 1. 2. a 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSi!M P le 13 MVIslud -

DATE PAGE RECIRCULATION ISI-35 Sh. 2 r.1 - 8/18/82 4 of 36

/ /

~ '""""

l$fjj$ # f if j l E, jl c '""*"

,/ / <

/

l -

l 1-202-5A D-6 B? C3 45 '

1 B 28 CA MO O C PTT

--- = --

RR J-N 1-202-5B B1- CS 45 D-3 1 B 28 MO O l pit RP

_'G A

-_ -_h~C

J-3 A7 RA

_-_.g__......

g 1-220-44 BT be 5 E-2 1 A 0.75 GL AO O C GROUP 1 ISOLATION l

l Plf RR l OP

__-.l ._FST - . . . _ ..._ __. . _ _ _ . . . . _._-.

' D%

l-220-45 E-1 1 A 0.75 GL AO O C l. Ys?

A? Ob ,

5 ,

GROUP 1 ISOLATION g FST CT l

1-220-67A F-5 AT-2 RM 1 AC 0.5 XFC SA O l

C t?t-1 ER l VR-9

._ __ _ l._ _ . - _[ __ .-_-- ---

l-220-67B F-5 *T-2

.  : RR }

1 AC 0.5 XFC SA u C

=- _- _ _ _ _ .. CT- 8 Rit VR

_ - _ ~.___

- ___ ___ [-___.._ l ..___-9_ _ _ _.__-=.

1-220-67C A? *2 RR E.F-5 1 AC O.5 XFC SA O i C CT-1 PR VR-9 ,

_ a ._-____ ___ -_

• _. =-

I

. ..=- _.__.y __-__-__ .

1-220-67D F-5 1 AC AT-2 RR I 0.5 XFC SA O C

_____ - C f-l S R*t V2-9

._-___p_RR AT-2 1-220-89A E-1 1 AC O.5 XFC SA O e

C CT-1  !.R VR 9 -

1-220-09D E-1 AT-1 l klr 1 AC 0.5 XFC SA O C CT

• _ - - ___. i RR VR-9

_ _ _ . _ ....____ - _ _ . . _ _ . _ _ _ . - _ _ _ - _ . ._--_,-_-.,l.__---- = - -

l-220-67E E-5 AT-2 RR 1 AC 0.5 XFC SA o C CT-1 &

____- -=- .__ -- --

VR-9 g

l-220-67F F AT-2 Ek #

_ -5 1 AC O.5 XFC SA O C

[' CT 1 stR VM l

___-9

_ . _(_ _ -_ _ . --

_/; ~') Commonwealth INSERVICE TESTING PROGRAM

\

Edison ISt - CLASS 1.2.& 3 val.VES UNIT - 1

\.C_/ H QUAD CITIES NUCLEAR POWER S T ATION

$1$11M P610 e DATE PAGE l* RiVISiDN RECIRCULATION (CONTINUED) :ISI-35 Sh* 2

[

'l - 8/'18/82

' 5 of 36

/ / i VALVE NUMBIR

[ , j REMARKS

/ -- / _/ L/ I 1-220-67G E-5 AT-2 RE 1 AC 0.5 XFC SA O C CTal ME

_ - - - _ _ _ .. VR-9 1-220-6711 F-5 4 P"-2 RR 1 AC O.5 XFC SA O C CT 1 RR VK-9 1-263-2-6A B-7 AT-2 RM 1 AC O.5 XFC SA O C CT-1 RR VM-9 .

.-___. .___a._...___ __..- ..___.- .-

AT-2 1-263-2-SA D-7 PR 1 AC O.5 XFC SA O C CT-1

._ RR VF-9 1-220-20A B-5 AT-2 RR 1 AC O.5 XFC SA O i C

= -

CT-1 RI. VR-9

=

1-220-19A B-6 AT-2 RR 1 AC 0.5 XFC SA O C CT-1 RR VR 9

[.

- - _ _ _ _ _ . =

1-220-22A D-8 AT-2 RR 1 AC O.5 XFC SA O C CT_1 g RR VR-9 4 ..,._ _ _ ____ _- _____

1-220-21A D-8 AT-2 Rf.

1 AC 0.5 XFC SA O C CT-1

_ _ . _ - kR

...___.f.VR-9

-_ -_.__.... __+ _ _ . .- . _ . ~ . e.._.=

1-220-20B A-3 1 AC i* N_2 RR 3 O.5 XFC SA O C (CT_1

__._4 ____.

RR

___...i lVR-9 g

1-220-19B A-3 9.T_2 Rx 1 AC . 0.5 XFC SA O C RR

.r*.1 VR-9 1-262-2-6B B-2 1 AT-2 RR AC O.5 XFC SA O CT-1

t. RR V R-9

=- -=-

1-262-2-5B B-2 AT.2 RR 1 AC 0.5 XFC SA G C CT 1 94 VR 9

__._.__+._..___..__.___.-

l-220-22B D-1 AT-2 RR 1 AC O.5 XFC SA O C CT 1 RR VR 9 t

INSERVICE ii! STING PROGRAM ,

O Commonwealth Edison ISI- CL ASS 12 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION -

SYSTIM P &10 REVISIGN - DATE PAGE RECIRCULATION (CONTINUED) ISI-35 Sh.2 1* - 8/18/82 6 of 36

. 4 d' y 49 REMARKS

! l / . _ /

1-220-21B p-1 RR ,

1 AC O.5 XFC SA o e l --- .-

l CT 1 na VR-9 i

I i

i 1

I 4

l l

4 Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CL A SS 1. 2. & 0 VALVES UNIT - 1

-QUAD CITIES NUCLEAri POWER STATION

_ w.m SYSIEM P Ir ID l REVISION - DATE PAGE CONTROL ROD DRIVE ISI-41 1 - 8/18/82 7 of 36 l

r / /

VALVE NUMBER

@ 4 REMARKS

/ /

(177) 1-0305-127 D-9 1 B O.75 GA AO C O *

= -- .... .

BT VR-6

• SCRAM TESTING (177) ...__. .,

1-0305-126 D-10 1 B 1.0 CA AO C 0

• __= --

BP VR-6

• SCRAM TESTING (177) - __ .

1-0305-114 E-9 2 C 0.75 CK SA C

_ = - - -

O CT-1

• VR-6

• SCRAM TESTING DT CS 1-0302-21A F-2 2 B 1.0 GL AO O C FST CS VR 12 BT CS 1-0302-21D F-7 2 B 1.0 GL An O O FST C?, VR-12

_9..___.. . . , _ _ . _ _ _... __ . . . . . . . _ _ _ .-__ __..

1-0302-22 11 7 CS F-3 2 B 2.0 GL AO O C FST CS VR-12 i

?

m .___-m _ . __- m ,

Commonwealth INSERVICE TESTING PROGRAM Edsn ISI- CLA SS 1. 2. & 3 VALVES UNIT - 1 .

QUAD CITIES NUCLEAR POWER STATION

-=::= :u.-

SYSifM P 6 ID '

RIVlil0N - DATE PALE RESIDUAL IIEAT REMOVAL ISI-37 1 - 8/18/62 8 of 36

/ /

VALVE NUMBER

[ REMARKS

/

1-1001-7A B-6 2 B 14 GA MO O O BT OP 90

.. -_ __... ... _. ......___._ .. =

1 1001-7B E-6 2 B 14 GA MO O O BT OP 1-1001-7C B-6 4._90 ____ . .. ... -- -

2 B 14 CA MO O O BT OP 90

..___ =

1-1001-7D E-6 2 B 14 GA MO O O DT OP 90 1-1001-67A B-3 2 C 12 CK SA C 0 CT-1 OP

_..._ . - = . . . . . . ___. . _ _ . . . .... _... ...... ..-__ . . . . , . _ . . . . ....___ ....___ ... . ~ . ...-

1 -1001-6's B E-3 2 C 12 CK SA C O CT-1 GP

- CP- - 1 . ----- r-- ------

1 1001-67C D-9 2 C 12 CK SA C O CT-1 1

. . . . . . . . .==

! 1 1001-67D E-9 2 C 12 CK SA C O CT_1 OP 1 1001-125A B-5 2 C 1 RV SA C C CT-2 SY

. __.- .._____... - - _ _ .._____ _ . , _ .. _ _ . . . . . - _ _ _ _ . . . .__ . _.. ._ ._.. _ L____ .___ ..___-

1_1001-125B E-5 2 C 1 RV SA C

• O CT-2 SY

.===- - -

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

1 1001-126C D-7 2 C 1 RV SA C 0 CT-2 SY

l. . . . . . _....... ..._ -_... ._-_-. .....

1 1001-125D E-7 2 C 1 RV SA C O OT-2 SY 1_1001-43A D-4 2 B 14 GA MO C C BT Ge 105

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

1-1001-43B E-4 2 B 14 GA MO C C BT CP 105

= . . . - --.a.. .... - ..._......_.._. .__

1 1001-43C B-8 2 B 14 GA MO C C BT OP 105 8 1-1001_43D E-8 2

. . . . ....... ...- _.-- _ . . . _ . .. .._.. ...._-_t___.... .....__ L _..... -_

B 14 GA MO C C BT OP f 105 1 1001-6A F-5 2 E 24 BTP M LJ > NA 1-1001-6B B-6 2 E 24 BTP M LO NA 1 1001-42A C-5 2 E 14 GA M LC NA

. . ._ .._.... ._. ..__ _ ... .=....... . . . .......

5

O, . --

Commonwealth INSERVICE TESTING PROGRAPA Edison ist - CLASS % ;e, a 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER S1 A flON

._ wa ,- m -- - -

$1S W j r L ID AWIS!Dft - DATE PAGE RESIDUAL llEAT REMOVAL (CONTINUED) l 8-

- = 8/10/82 9 of 36 lIS1-rf g

/ / _ ! '

VALVE NUMSER i /

/

1-1001-42D E-5 2 E 14 GA M LC

_ HA 1-1001-42C C-6 2 E 14 CA M LC NA 1-1001-42D E-6 2 E 14 GA M LC NA

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

1-1001-66A C-2 2 E 12 GA M LO

-==- --

HA 1-1001-66B E-2 2 E 12 CA M LO

- =-_- _= -

NA 1-1001-66C C-8 2 E 12 CA M LO NA 1-1001 66D E-9 2 E 12 CA M LO NA 1-1001-15A B-2 2 E 18 GA M LO

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

NA 1-1001-1SB D-9 2 E 18 GA M LO NA

. = - -

1-1001-17A D-2 2 E 18 GA M LO

.- - - - = = - -- --

NA 1-1001-17B B-9 2 E 18 GA

~~~~

M LO HA

... ~ - - -

1-1001-141A B-3 2 E 2 GA M LO

--~~--= _ _ .

NA

= - _ . . ....

1-1001-141B

. _ _ ...... a ...___-

E-3 2 E 2 GA M LO NA

_ _ .._ ._=, __

1-1001-141C B-9 2 E 2 GA M I4

~ - - -

NA 1-1001-141D E-9 2 E 2 GA M 14

~ - NA 1-1001-142A B-3 2 C 2 CK SA C O CT 1

• VR-13

-===- - -

• SEE VR-13 1-1001-142B E-3 2 C 2 CK SA C

• O CT-1 VR-13

---

_ _= - _

s *SEE VR-13 1-1001-142C

= . . . . . . ..

B-9 2 C 2 CK SA C

-.---__ ...--- - ...-.... ==-

O CT-1

• VA-13 *SEE VR-13 1-1001-142D ___. _

E-9 2 C 2 CK SA C O CT-1

• VR-13 *SEE VR-13 .

^

Commonwealth INSERVICE TESTIPjG PROGRAM Edison ISI - CL AS6  ?. 2. E 3 VALVES , ,

QUAD CITIES NUCt. EAR POWER STATION SYSTEM P 6 ID REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-39 1.,- 8/18/82 10 of 36

/

VALVE NUMBER

[ ,

REMARKS

/ c 1-1001-29A AT-1 RR A-5 1 A 16 CA MO C O BT CS j 25 J-7 1-1001-29B A-7 AT-1 RR 1 A 16 GA MO C O BT CS 25 J-7

. _ _ _ . - . __. ___ . . . . 1.__

1-1001-47 C-5 AT-1 RR 1 A 20 GA MO OEC C B1 CS 40 J4 GEOUP 2 ISOLATION 1-1001-50 B-5 AT-1 RR A 20 osc 1

GA MO C

. ~ - .___....

trr CS 40 lJ GROUP 2 ISOLATION 1-1001-60 A-7 1 A 4 GA AT-1 RR _.7._-4 _ ....

MO O&C C BT CR 25 l Ja4 GROUP 2 ISOLATION AT-1 F.R

___....p_.______ ._-

1-1001-63 A-6 1 A 4 GA MO Osc C g

9 IIT CS 25 , J /. GROUP 2 ISOLATION 7._-_.m kk I

__-_.-- _. __=

1-1001-68A A-5 1 C 16 CK SA C C l PIT

CT 1 ch J-2 1-1001-600 A-6 l

M.T RR 1 C 16 CK SA C O C1*1 CS

_ - _ _ _ .. - -  : J-2

1. ~

- - - . . ._ _ _ - _ _~-.. _ _..___ .

1-LOO 1-16A D-2 2 B 18 GL HO Osc O RT OP 125 l 1-1001-16B D-10 2 B 18 GL MO Of C C BT OP 1'45 I

_.. .... . . , _ . _ _ , ..__. . . . _ - . . ..__. I

. . _ _..Ot.C('_.L_BT__ _.- ..-

1-1001-10A B-4 2 B 3 CA MO C OP I? I 1-1001-18B B

', . , _ _ _ . . . .. __..l.--_ __. .-

2 B 3 GA MO C l DT

____ .__-7_ _ _ _ .. -- 00 17 1-1001-19A D-2 2

....4__OaC _____. . _ . _ . . . _ _ . . __ _ _ _ _ _ . . . . . . . . ___.__.--

B 18 GA MO 0 p? C . J_6 1-1001-19B

- - - -_- _= _

D-9 2 B '18 GA

--MO . d _cU _ . . . .m___..m.. . . . . . . .

C

...'I. .. .h.12*' __. . _.,... . ...__.__. ..... _

BT C3 125 J

__ -6

.._....._}.__.._

1-1001-20 C RR 2 A 3 GA

_____________ .__-8 __

MO f

_ . 4C C IAT-1 IIT OP 25 y.- ----

GROUP 2 ISOLATION

. . . _ I----- -~I -. - i 6

m__ __._ ____ . . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

• 9 .

Commonwealth INSERVICE TESTING PROGRAM Hison 131 - CL A S S 1. 2. A 3 VALVES y_3 QUAD ClTIES NUCLEAR POWER STATION SYsitu ' P 610 l REVIStak - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) SI .19 { 4,- 8/18/82 11 of 36

/_ /

/

$ Af f f$ ll? /_

f -

$l "**"'

1-1001-21 C-8 7.T RR 2 A 3 GA M0 O&C

-- C BT OP 25 GROUP 2 ISOLATION 1-1001-22A A-2 2 C 1 RV SA C 0 CT-2 RR 1-1001-22B A-9 2 C 1 RV SA C

---

" 0 CT-2 ER

=__. _ _ _ . . . . ____ _ _ _ _

1-1001-23A A-5 AT RR 2 A 10 CA MO C C

- - BT OP 15 1 1001-23B A-6 2 AT RA A 10 GA Mo C C BT OP 15 1_1001-26A A-5 2 AT RR A 10 GA Mo C 0

.- hT OP 17

..____.q........ ...__. .....-.. . . _ ~ . . -

1-1001-26B A-6 2 A 10 j a na .

GA Mo C 3 C Br OP 4 15

._. ..__.____L,. . . _ _ .

1-1001-28A A-4 . . . ...__. .

2 B 16 GL MO O

- - - O BT CS 90 J-7 1-1001-28B A-7 2 B 16 GL MO O o fit CS 90 J-7 1-1001-36A B-2 2 AT-1 RR A 14 GL MO C OLC BT OP 60

- . _ _ _ . .. _ _ . . .._ _ ___. _ __._.i _____

1-1001-36B B-8 2 AT-1 RR A 14 GL MO J C&C BT OP 60 AT

_ _ _ i 1-1001-37A B-3 2 Rn A 6 GL MO C . OLC

= -

PT i 60 I,

____ _.E ..'P.---t------- --- -

_ = _ _ _

I AT-1 1-1001-37D D-7 2 A 6 GL MO C RR OLC BT OP 60

-. De6____e

.___..P._e.__.66'd'.-- p f, 1

1

_l

- ,oy m

,m

/

i, i'J ')7 common..aith INSERVICE TESTING PROGRAM Edison ISI - CI. A S S "

. 2. & 3 VALVES UNIT - 1 k/ QUAD CITIE S 740 CLEAR POWER STATION .

~

SYSTEM F 6 tD REVISION - DATE PAGE RESIDUAL llEAT REMOVAL (CONTINUED) IP.I- 39 T- 8/18/82 12 of 36

/

)

VALVE NUMBER 5 4 e4 4 D s REMARKS N

4'A C *s* 8

@d J 8 5

-r

<P - c y A y#v

/

/_

AT 1 RR 1 1001 34A A2 2 A 16 GA C HO O&C RT OP

__.....g....__..

125 AT -1 RR 1 1001 34B B.7 2 A 16 GA MO C O&C BT OP 125 1 1001-2A F.3 3 C 12 CK SA C O CT.1 OP 1-1001 2B F.3 3 C 12 CK SA C O CT.1 OP 1_1001 2C F-7 3 C 12 CK SA C O CT.1 OP 1 1001 2D F-7 3 C 12 CK SA C 0 CT-1 OP g g .....___....__..........

1 1001-5B E_7 3 B 12 GL MO C OEC BT OP 90

........ _... ._____.. ... __.. . . . _ _ _ ..__..... ....... ...____. .....__ ..__..................t.. .___.................__.....___.__....

1 1001-1A G-4 3 E 14 GA M LO NA 1 1001-1B G-4 3 E 14 GA M LO NA 1_1001_1C G-6 3 E 14 CA N LO NA

._. .....____ . . . . _ . . . . ._.... _ . . . . ......... ....... . _....- .._..., .. n. .. .......- . . . . . . . . . . . . . . . .....................-- ..._..._.

1 1001-1D G6 3 E 14 CA M '

LO HA

.......___... ........ ........ . . . . . ......... ....... ........ .......e........

r ........ . . . . . . ...._. ............................. ......

1-1001-3A G-3 3 E 12 GA M l NA

___...____... ..__...--___.... . . . . . . .._____.. ....... . . . . . . . !. . .LG....L.......

f 1 1001_3D G-3 3 E 12 GA M 1.5 NA

............. . _ _ _ . . . . ........____._ ..____.. ___.... .____..- ..__...gg....... ......q........ ...... ,.......__. .__...........____......

1 1001 3C G-7 3 E 12 GA M LO

_..._____.... . . . . . _ _ ....____. .______ . . _ _ . . . _____.. . . _ _ . . . . .......')....... . . . . . . . ' . . . ._.........__...................___....___

L NA 1-1001-3D G-7 3 E 12 CA N r.O NA f

....... ....... ._____ ......}.......g .. ........._____.__. .._____.....

1 1001-201A F- 3 NC E 14 BTF M 30 i  ! NA j

, ....___...... ..____.~__..__. ..___.. ...__... ...........___. _...... ... .... ....... .........,__..),,........ .__...__...___....._____

1 1001-201B F-7 NC E 14 BTF M LO NA I

i 1

I

{

Commonwealth INSERVICE TESTING PROGRAM Edison Isi- CL ASS 0. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSi!M P & ID RIVISION - DATE PAGE RESIDUAL !! EAT REMOVAL (CONTINUED) ISI-39 .1 5 - 0/18/82 13 of 36 VAtVE NUMBER

[ 4 REMARKS T

1/2-1099-1 G-2 3 E 16 GA M LC NA 1-1001-33A B-5 3 E 16 CA M LO NA 1-1001-33D B-6 3 E 16 GA M LO NA e

4

N-Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CL AS S 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSTEM P l'r ID REVISION - DATE PAGE STANDBY LIQUID CONTROL ISI-40 1 - 8/18/82 14 of 36 VALVE NUMBER

[ [ REMARKS 1-1101 15 C.3 1 C 1.5 CK SA C 0 CT-1 CS/RS VR-5 1-1101 16 C-3 1 C 1.5 CK SA C 0 CT.1 CS/RR VR-5 1-1106A C-4 2 D 1.5 EXP C O DT RR 1 1106D D-4 2 D 1.5 EXP C 0 DT RR 1-1101 43A D-6 2 C 15 CK SA C 0 CT-1 OP 1 1101-43B E-5 2 C 1.5 CK SA C O CT.1 OP 1 1105A C-6 2 C 1.5 RV SA C O CT-2 SY 1-1105D D-5 2 C 1.5 RV SA C' O CT.2 SY 1-1101-4 E.8 2 E 2.5 GA M LO NA 1-1101-8 D-8 2 E 2.5 GA M LC NA 1 1101 3A D.7 2 E 2.5 GA M LO NA 1-!!01-3B E-7 2 E 2.5 GA M LO NA 1 1101 10 D.7 2 E 1 GL M LC NA 1 1101 2A D.5 2 E 1.5 GL M LO NA 1 1101 2D E.5 2 E 1.5 GL M LO NA j

1 1101-22 C.4 2 E 1.5 GL M LC NA 1 1101 9B D-4 2 E 1 GL M LC NA 1 1101 23 D.3 2 E 1.5 GL M LO NA 1 1101 1 D.2 1 E 1.5 GL M LO NA

(y)

Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CL A SS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID REVIsl0N -

DATE PAGE RX WATER CLEAN-UP ISI-47 ,1,' '- S /18 / 8 2 15 of 36 VALVE NUMBER p [ REMARKS AT-1 RR 1-1201-2 B.6 1 A 6 CA MO O C BT OP 30 GROUP 3 ISOLATION PIT RR AT-1 RR 1-1201 5 C.6 1 A 6 GA MO O C BT OP 30 CROUP 3 ISOLATION t i e

l t

l l

l

Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CL A SS 1.2 & 3 VALVES , ,

QUAD CITIES NUCLEAR POWER STATION SYSIEM P 6 10 REVis10N - DATE PAGE REACTOR CORE ISOLATION COOLING ISI-50 1 - 8/18/82 16 of 36 VALVE NUMBER

[ REMARKS

/

1-1301-16 B-2 AT-1 RR 1 A 3 GA MO O C BT OP 25 GROUP 5 ISOLATION PIT RR 1-1301-17 B-3 AT-1 RR 1 A 3 GA MO O C BT OP 25 GROUP 5 ISOLATION 1-1301-40 D-2 AT-1 RR NC AC 2 CK SA C C CT-1 RR VR-8 1-1301-41 D-2 AT-1 RR NC AC 8 CK SA C C CT-1 RR VR-8 1-1301-15A AT-2 RR B-2 1 AC .5 XFC SA O C CT-1 RR VR-9

• 1-1301-15B B-2 AT-2 RR 1 AC .5 XFC SA O C CT-1 RR VR-9 e

a Y

Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CLASS 1.2,& 3 VALVES UmT-1 QUAD CITIES NUCLEAR POWER STATION SYS1(M WSM P 6 LD -M PAM CORE SPRAY I ~

~

l VALVE NUMBER N [

1-1402-9A C-3 CT-1 CS 1 C 10 CK SA C O PIT RR J-2

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

1-1402-9D C-4 CT-1 CS -

1 C 10 CK SA C O PIT

=

RR J-2 1-1402-25A C-2 1 D 10 CA MO C O DT. OP 15

-1402-25D C-5 1 D 10 GA MO C O BT OP 15 1 1402-24A B-2 2 B 10 GA MO O

=- ------- --

-. .--.=

O BT OP 15 1-1402-24D D-5 2 D 10 GA HO O O BT CP 15 1-1402-28A . _ _ _ _ _ .._____-

=.

C-9 2 C 2 RV SA C O CT-2 SY 1-1402-20B D-6 2 C 2 RV SA C O CT-2 SY 1-1402-38A C-0 2 B 1.5 GA MO O C BT OP 12 1-1402 BOD D-7 2 B 1.5 CA MO O C BT OP 12 1-1402-DA E-9 2 CE 12 SCK SA C/14 O CT-1 OP 1-1402-0D E-6 2 CE 12 SCK SA C/LO O CT-1 OP

=.

-_____ .- . _ _ . =.___-

1-1402-31A E-3 AC AT-2 RR 1 .5 XFC SA O C CT-1 RR

. VR-9 1-1402-31D E-3 AT-2 RR 1 AC .5 XF2 SA O C CT-1 RR VR-9 1-1402-6A D-3 1 E 10 GA M LO

= .___. _ , _ _ _ _ _

. _ _ ___._ NA

. -6_. -

1-1402-6B D-3 E 10 -M_.-- -

1 CA H LO NA 1-1402-2A G-7 2 E 12 CA M LC

_-- - NA

.~

commonwniih INSERVICE TESTING. PROGRAM Edison ISI - CL ASS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYS11M P & ID REVISION - DATE PAGE CORE SPRAY (CONTINUED) ISI-36 y - 8/18/82 18 of 36

/

WAIVE NUMBER A [ [ REMARKS i

1-1402-28 G-4 2 E 12 GA M LC NA 1-1402-34A G-4 2 E 18 BTF M LO NA 1-1402-343 F-3 2 E 18 BTF M LO NA 1-1402-4A A-8 2 B 8 GL MO C C BT OP 60 1-1402-48 C-7 2 B 8 GL MO C C BT OP 60 1-1402-13A E-9 2 CE 1.5 SCK SA/M

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

C/LO O CT.1

• VR-13 *SEE VR-13 1-1402-13B ..____._.._..

E-6 2 .

CE 1.5 SCK SA/M C/LO O CT-1

• VR-13 *SEE VR-13 i

I

Commonwealth INSERVICE TESTING PR2 GRAM Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSifM P Ir ID REVISION -

DATE PAGE PRESSURE SUPPRESSION M-34 d - 8/18/82 19 of 36

/

VAtVE NUMBER [ REMARKS 1-1601-21 AT-1 RR C-6 NC A 18 BTF AO C C BT OP 10 GROUP 2 ISOLATION FST OP 1-1601-22 AT-1 RR C-6 NC A 18 BTF AO C C BT OP 10 GROUP 2 ISOLATION

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

FST OP 1-1601-55 AT-1 RR A-6 NC A 4 GA AO O C BT OP 10 GROUP 2 ISOLATION FST OP 1-1601-56 D-6 AT-1 RR NC A 18 BTF AO O C BT OP 10 GROUP 2 ISOLATION FST OP 1-1601-57 AT-1 RR C-9 NC A 1 GL MO O C BT OP 15 GROUP 2 ISOLATION FST OP AT-1 RR 1-1601-58 D-7 NC A 1 GL AO C C DT OP 15 GROUP 2 ISOLATION

_____=_ - - __ _

FST OP 1-1601-59 AT-1 RR D-7 NC A 1 GL AO O C BT OP 15 GROUP 2 ISOLATION

__- -== _-- _ -

FST OP AT-1 RR 1-1601-20A D-9 NC A 20 BTF AO C O&C BT CS 10 J-8 FST OP

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

AT-1 RR 1-1601-31A D-9 NC AC 20 CK SA C O&C CT-1 OP

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

AT-1 RR 1-1601-20B E-9 NC A 20 BTF AO C O&C BT CS 10 ,J-8

-z... -

PST OP

Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CL A SS 1.2.& 3 VALVES UNIT - 1 OUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID REVISION - DATE PACE PRESSURE SUPPRESSION (CONTINUED) M-34  % - 8/18/82 20 of 36

/

VAtyt NUMBER

[ REMAhKS AT-1 RR 1-1601-31B E-9 NC AC 20 CK SA C O&C CT-1 OP AT-1 RR 1-1601-23 D-3 NC A 18 BTF AO C C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 1-1601-24 D-2 NC A 18 BTF AO C C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 1-1601-60 D-3 NC A 18 BTF AO C C DT OP 10 GROUP 2 ISOLATION FST OP 1-1601-61 AT-1 RR B-2 NC A 2 GL AO C C BT OP 15 GROUP 2 ISOLATION FST OP

_.==

AT-1 RR

• 1-1601 62 E.2 NC A 2 GL AO C C DT OP 15 GROUP 2 ISOLATION

_ _ _ _ _ _ = = - - _-- _-

FST OP 1 1601-63 AT-1 RR E-2 NC A 6 BTF AO C BT C OP 10 GROUP 2 ISOLATION FST OP 1-8803 AT-1 RR C-6 NC A 2 GL AO O C DT OP 10 GROUP 2 ISOLATION

_.. . . = - -

PST OP AT-1 RR 1-8804 D-6 NC A 2 GL AO O C BT OP 10 GROUP 2 ISOL\ TION

-..__. =___ . _ _ _ ..._

OP AT-1 RR 1-8801A C-3 NC A O.S GL AO O C BT OP 10 GROUP 2 ISOLATION

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

PST OP

)

l

m -

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- GL A SS 1. 2 & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSTIM P 610 PAGE REV.ISION - DATE PRESSURE SUPPRESSION (CONTINUED) M-34 .1- 8/18/82 21 of 36 VALVE NUMBIR

[ REMARKS AT-1 RR 1 8001D D.3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT -1 RR 1 0801C D-3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT -1 RR 1 8001D E-3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT -1 RR l-8802A C.3 hc A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT -1 RR 1 0802D D.3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT -1 RR l-8802C D-3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT -1 RR 1 8802D E-3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP 1-1601-32A E-2 NC C 18 CK SA C O&C CT-1 OP 1-1601 32B E-2 NC C 18 CK SA C OEC CT-1 OP 1-1601-32C E-2 NC C 18 CK SA C O&C CT.1 OP  !

............. .............. ........ ....... ........ ....... ........ ....... ........ ....... . . . . . . . .. . . . . . . . . . ......................... e 1 1601 32D E.2 NC C 18 CK SA C O&C CT.1 OP a 1 1601 328 I

E.2 NC C 18 CK SA C OEC CT.1 OP

- -__ - --- - - - _ - --_. - A

c3 Commonwealih INSERVICE TESTING PROGRAM C/ Edison ISI - CL A SS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION sysTIM r g, ID PAGE REVJSiON - DATE PRESSUPE SUPPRESSION (CONTINUlin) M-34 31- 8/18/82 22 of 36 VAtVE NUMagg g REMARKS 1-1601-32P E-2 NC C 10 CK SA

=-- . . . . ..--

C O&C CT-1 OP l-1601-33A E-7 NC C 18 CK SA C O&C CT-1 OP 1-1601-33D E-7 NC C 18 CK SA C OEC CT-1 OP 1-1601-33C -......_-

E-7 NC C 18 CK SA

--- - - ==-

C O&C CT 1 OP 1-1601-33D E-7 NC C 18 CK SA C O&C CT-1 OP 1-1601-33E E-7 NC C 18 CK SA C O&C CT-1 OP

-1601-33P E-7 NC C 18 CK

=-

SA C O&C CT-1 OP 1-220-OlA E-4 NC C 1 CK SA C O CT-1 CS VR-14 1-220-01B E-4 NC C 1 CK SA C O CT-1 CS

=-- --_. -

.._ _-- VR-14 1-220-81C E-4 NC C 1 CK SA C O CT-1 CS VR-14 1-220-81D E-5 NC C 1 CK SA C O CT-1 CS VR-14

$b 220-81E E-5 NC C CK SA C O CT-1 CS VR-14

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

i

Commonwealth . INSERVICE TESTING PROGRAM Edison ISI - CL ASS 1. 2. & 3 vat.VES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE IIIGli PRESSURE COOLANT INJECTION ISI-46 1 - 8/18/82 23 of 36 VALVE NUMBER

[ ftEMARKS AT RR 1-2301-4 C-9 1 A 10 GA MO O OEC BT CS 50 J-5 GROUP 4 ISOLATION PIT RR -

AT RR 1-2301-5 B-10 1 A 10 GA MO O O&C BT-

. _

CS 50 J-5 GROUP 4 ISOLATION 1-2301-3 A.-6 2 B 10 GA MO C O

_= --

BT , OP 25 1-2301-68 A-6 2 D 16 RPD SA C O *

• PER MANUFACTURER RECOMMENDATION 1-2301-69 A-6 2 D 16 RPD SA C O *

• PER MANUFACTURER

..__.......- .= .- -.... .... -

RECOMMENDATION BT OP 1-2301-29 B-9 2 B 1 GI. AO O C FST OP 10 AT-1 RR *OPEN STROKE VERIFIED 1-2301-34 D-7 2 AC 2 CK SA C O&C CT-1

=

OP/RR* VR-8 BY PUMP TEST 1-2301-45 AT-1 RR *OPEN STROKE VERFIED B-8 2 AC 24 CK SA C O&C CT-1 OP/RR* VR-8 BY PUMP TEST

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

1-2301-35 E-7 2 8 16 GA MO C O&C BT

_- . _ _ - - _ = - -

OP 120 1-2301-36 E-9 2 8 16 GA MO C O&C BT OP 120

....____ __ =- . . _ _ .. _ _ . .__. ...___

1-2301-6 P-2 2 0 16 CA MO O O&C BT OP 120 2301-20 E-2 2 C 16 CK SA O O CT-1 1-2301-14 C-6 2 D 4 G1. HO C O&C BT OP 26 -

l-2301-39 E-8 2 C 16 CK SA C

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

O CT-1

• VR-7 *SEE VR-7 1-2301-40 D-7 NC C 4 CK SA C O CT-1 **

---

=

VR-13 **SEE VR-13

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CLASS 1. 2 & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSTIM P fr ID REVISION - DATE PAGE IIIGII PRESSURE COOLANT INJECTION (CONTINUED) ISI-46 $-8/18/82 24 of 36 VALVE NUMBER

[ [ 4 h REMARKS 1-2301-8 D-6 2 B 14 GA MO C O BT OP 45

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

PIT RR l-2301-7 D-6 2 C 14 CK SA C O CT-1 CS

---

=- -- , - - -

J-2

. _... .---.__ a.....

1-2301-74 B-8 2 CE 12

• OPEN STROKE VERIFIED SCK SA C/I4 O CT-1 OP* BY PUMP TEST AT-2 RR 1-2301-26 D-9 1 AC .5 XFC SA O C CT-1 RR VR-9 AT-2 RR 1-2301-27 D-9 1 AC .5 XFC SA O C CT-1 RR VR-9

---

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

1-2301-22 B-1 2 E 16 CA M LO NA

---

== ---. --.

1-2301-56 F-8 2 E 16 BTF M In NA 1-23bi-71 D-7 2 CE 2 SCK SA/M C/I4 CT-1 OP* *OPEN STROKE VERIPIED BY PUMP TEST 1-2301-9 D-5 2 B 14 CA MO O O BT OP 45 1 2301-10 E-5 2 B 12 GL MO C C BT OP 60 l

i i

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CL ASS 1.2.& 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSilM P 6 ID REVISION -

DATE PAGE MAIN STEAM ISI-13 Sh. 1 'l- 8/18/62 25 of 36 VALVE NUMBER

[ REMARKS AT-1 RR BTP OP BT Cs 5 J-1 GROUP 1 ISOLATION 1-203-1A P-4 1 A 20 GL AO O C FST CS PIT RR AT-1 RR i BTP OP '

1-203-1B DT CS 5 J-1 GROUP 1 ISOLATION D-4 1 A 20 GL AO O C FST CS

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

PIT RR AT-1 RR BTP OP BT CS 5 J-1 GROUP 1 ISOLATION 1-203-IC C-4 1 A 20 GL AO O C PST CS PIT RR AT-1 RR DTP OP 1-203-1D BT CS 5 J-1 GRdUP 1 ISOLATION B-4 1 A 20 GL AO O C FST CS PIT RR 1-220-1 E-4 AT-1 RR 1 A 3 GA MO C C BT OP 35 GROUP 1 ISOLATION 1-203-3A BT *

• VR-1 *SEE VR-1 F-4 1 BC 6 ERV/SV

..__--- _=_. __

PS/SA C 0 CT-2 RR VR-2 I BT *

• 1-203-3B D-6 1 BC 6 ERV VR-1 *SEE VR-1 PS C O CT-2

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

RR VR-2 BT *

• 1-203-3C C-7 1 BC 6 VR-1 *SEE VR-1 ERV PS C 0 CT-2 RR VR-2 BT *

• 1-203-3D D-7 1 BC 6 VR-1 ERV PS C O C RR

..T ... __ . . . . , _ _ , _ . ._V_R-. 2 _.-

• SEE VR-1 -

+.

INSERVICE TESTING PROGRAM O Commonweahh Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 1 OUAD CITIES NUCLEAR POWER STATION r

SYSi!M P 6 ID REVISION - DATE PAGE MAIN STEAM (CONTINUED) ISI-13 Sh. I 1 - 8/18/82 26 of 36 VALVE NUMBER

[ REMARKS BT *

• 1-203-3E D-7 VR-1 i BC 6 ERV PS C 0 CT-2 RR VR-2

• SEE VR-1 1-203-4A F-8 1 C 6 SV SA C O CT-2 RR VR-15

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

1-203-4B D-5 1 C 6 SV SA C 0 CT-2

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

RR VR-15 1-203-4C C-5 1 C 6 SV SA C O CT-2 RR VR-15 1-203-4D D-5 1 C 6 SV SA C 0 CT RR VR-15 1-203-4E F-8

.__ . _ _ _ _ . . . ..-2 _.___.. . _ _ _ _ . - - ______....._____

1 C 6 SV SA C O CT-2 RR VR-15

_-__ ._____... = - - - .

1-203-4F D-5 1 C 6 SV SA C O CT-2 RR VR-15

_- ._ __... .__.. _m , ,

1-203-40 C-5 1 C 6 SV SA C O

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

CT-2 RR VR-15 1-203-411 D-5 1 C 6 SV SA C O CT-2 RR VR-15 9

INSERVICE T'ESTING PROGRAM

@ Commonwealth Edison ISI - C L A S S QUAD CITIES NUCLEAR POWER STATION

1. 2. & 3 VALVES UNIT - 1 SYSTEM P fr ID REVISION - DATE PAGE MAIN STEAM (CONTINUED)

ISI-13 Sh. 2 4 - 8/18/82 27 of 36 VALVE NUMBER

[ [ REMARKS AT-1 RR BTP OP 1-203-2A BT CS 5 J-1 E-7 1 A 20 GL AO O C GROUP 1 ISOLATION FST CS

= --

PIT RR AT-1 RR BTP OP 1-203-20 E-7 BT CS 5 J-1 1 A 20 GL AO O C GROUP 1 ISOLATION FST CS

.=-- ==...____~,.._____ m.-

PIT RR AT-1 RR BTP OP 1-203-2C BT CS S J-1 D-7 1 A 20 GL AO O C GROUP 1 ISOLATION FST CS PIT RR

.__.= . ...

AT-1 RR BTP OP 1-203-2D B-7 BT CS 5 J-1 l 1 A 20 CL AO O C FST GROUP 1 ISOLATION

' CS

==- -- - --- -- -

PIT RR 1-22C-2 E-7 1 A AT-1 RR 3 GA MO C C BT

= -- -

OP 35 GROUP 1 ISOLATION 1-220-17A E-8 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9 l I

1-220-17D D-8 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR

.. -. - VR-9 1-220-17C C-8 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR 1 --

VR-9 1-220-17D B-8 AT-2 RR 1 AC O.5 XFC SA O C CT l' RR

---

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

VR-9 i

Commonwealth INSERVICE TESTING PR$ GRAM Edison ISI - CL A SS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYS11M PinID REVISION - DATE PAGE MAIN STEAM (CONTINUED) ISI-13 Sh. 2 1 - 8/18/82 28 of 36 VAtVE NUMBIA * -

REMARKS I

e 1-220-18A E-8 AT-2 RR 1 AC .5 XFC SA O l C CT-1 RR VR-9

... == . .__. ___... ..

1-220-18B D-8 1 AC AT-2 RR

.5 XFC SA O C CT-1 RR VR-9 1-220-18C C-8 AT-2 RR 1 AC .5 XPC SA O C CT-1 RR VR-9 1-220-18D D-8 AT-2 RR 1 AC .5 XFC SA O C

.. . = . -

CT-1 RR VR-9 a

e

l Commonwealth INSERVICE TESTING PROGRAM l

Edison ISI- CL ASS 1.2.& 3 VALVES UNIT 1 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 RIVISION - DATE PAGE FEEDWATER ISI-15 15 - 8/18/82 29 of 36 VALVE NUMBER

[ [ REMARKS AT-1 RR 1-220-58A E-3 1 AC 18 CK SA O C .

CT-1 RR VR-3

_-___. -=

AT-1 RR 1-220-58B F-3 1 AC 18 CK SA O OEC CT-1 RR VR-3 1-220-59B F-3 2 C 18 CK SA O C CT-1 RR VR-3 AT-1 RR 1-220-62A E-3 1 AC 18 CK SA O C CT-1 RR VR-3 AT-1 RR 1-220-62B F-3 1 AC 18 CK SA O O&C CT-1 RR VR-3 t

1 O

Commonwealth INSERVICE TESTING PROGRAM Edison ISI - C t. A S S 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYSTEM P & ID REVIS1011 - DATE PAGE SERVICE WATER ISI-22 1 - 8/18/82 30 of 36 vatyt muusta

[f f g stuAnts 1-3999-85 A 3 C 8 CK SA C O CT-1 OP

_... .__- 10 _ . . . . - - .__.. . _ . - - - -- - .._ - _ _ . . . .

1 3999-86 B-9 3 C 8 CK SA C O CT OP

. . . _ . .. .._-- . ... . . . ..__._ _ ..-1 .....___ . ... .

1-3999-88 D-10 3 C 8 CK SA C O CT-1 OP e

l l

l l

I

Commonwealth INSERVICE TESTING PREGRAM Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 1 OUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID REVE001 - DATE PAGE DIESEL GENERATOR STARTING AIR (SERVICE AIR SYSTEM) M-25 li - 8/18/82 31 of 36 VAWE EMBER  % REMARKS 1-4699-121 E-9 NC E 1.5 CA M LO NA

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

1-4699-122 E-9 NC E 1.5 GA M LO NA 1-4699-225 D-8 NC E 1.5 BALL M LO NA 1-4699-123 E-9 NC C 1.5 CK SA C O CT-1 OP

_= -- _ _ . . _ _ .

1-4699-196 E-9 NC C 1.5 CK SA C O CT-1 OP -

1-4699-226 D-8 NC B 1.5 GL AO C O BT OP 5 I

Commonwealth INSERVICE TESTING PREGRAM , ,

ISI- CLASS 1. 2 & 3 VALVES Edison QUAD CITIES NUCLEAR POWER STATION SYSTEM P fi ID REVISION -

DATE PAGE INSTRUMENT AIR PIPING M-24 Sh. 2 1 - 8/18/82 32 of 36 VALVE NUMBER 4 p f REMARKS AT-1 RR 1-4720 D-3 NC A 1 GA AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR ,

1-4721 D-3 NC A 1 GA AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 1-733-1 F-7 NC A O.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION

• AT-1 RR

• 1-733-2 P-7 NC A O.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION

_._.=- -

AT-1 RR 1-733-3 F-7 NC A O.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION AT -1 RR 1-733-4 F-7 NC A O.375 BALL SO C C DT OP 5 GROUP 2 ISOLATION AT-1 RR 1-733-5  ?-7 NC A O.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION

_ ... . _ . ~ . = .....-- -..... - . . ,

AT-1 RR f 1-743 B-7 NC AC O.375 CK SA O C CT-1 RR VR-4

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

1-736-1 F-7 NC D O.375 EXP O C DT RR

.. .- ... ==

1-736-2 P-7 NC D O.375 EXP O C DT RR 1-736-3 F-7 NC D O.375 EXP O C DT RR 1-736-4 F-7 NC D O.375 EXP O C DT RR 1-736-5 F-7 NC D O.375 EXP O C DT RR

- .___..._ ._ . ....... .. _ _ . . . ..._ _ . .. ._- .~. _.

,m g

) commonwealih INSERVICE TESTING PROGRAM

('9

(  %-

J Edison ISI- CLASS 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER STATION UNIT - 1 SYSTEM RX DUILDING EQUIPMENT DRAINS P 6 ID REVISION - DATE PAGE

(& SHARED UNIT 1/2 DIESEL GEN. AIR START PIPING) M-43 1 - 8/18/82 33 of 36 VALVE NUMBER

[ [ [ REMARKS AT-1 RR 1-2001-15 E-3 . NC A 3 GA AO C C BT OP 20 GROUP 2 ISOLATION FST OP ,

AT-1 RR 1-2001-16 E-3 NC A 3 GA AO C C BT OP 2G GROUP 2 ISOLATION FSf OP AT-1 RR 1-2001-3 P-7 NC A 3 GA AO C C BT OP 20 GROUP 2 ISOLATION I FST OP l AT-1 RR 1-2001-4 P-7 NC A 3 CA AO C C BT OP 20 GROUP 2 ISOLATION.

FST OP

=. ____. .__. __ - - . - _ _ _ ... _ - -

1/2-4699-46 D-9 NC E 1.5 GA H LO NA 1/2-4699/47

= -

D-9 NC E 1.5 GA M LO NA 1/2-4699-225 D-8 NC E 1.5 BALL M I4 NA 1/2-4699-48 D-9 NC C 1.5 CK SA C O CT-1 OP

-__ -- ._____- __ = -

1/2-4699-196 D-9 NC C 1.5 CK SA C 0 CT-1 OP 1/2-4699-226 D-8 NC B 1.5 GL AO C O BT OP S e

commenmalth INCERVICE TESTING PR$ GRAM Edison ISI- CLASS 1.2.& 3 vat.VES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION DIESEL CENERATOR FUEL OIL P & 10 REVISION - DATE  ? AGE (UNIT 1 AND 1/2) M-29 1.- 8/18/82 34 of 36

/

VALVE NUMBER

[ REMARKS ,

1-5299-5 E-4 NC C 1.5 CK SA C 0 CT-1 OP 1-5201 E-3 NC B 1 GA SO C O BT OP 5 1-5199-155 B-5 NC E 1 CL M LO NA 1-5199-157 C-5 NC C .5 CK SA C C CT-1 OP 1/2-5299-5 E-4 NC C 1.5 CK SA C O CT-1 OP ~

1/2-5201 E-3 NC B 1 GA SO C O BT OP 5 1/2-5199-157 C-5 NC C .5 CK SA C C CT-1 OP e

E

commonweatih INSERVICE TESTING PROGRAM Edison ISl- CL ASS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION x

SYSTEM

• P b ID REVISION - DATE FAGE CONTAINMENT ATMOSPl!ERE MONITOR M-641 Sh. 1 'l - 8/18/82 35 of 36 WAtVE NUMBER D p REMARKS

/

A S0-1-2499-1A D-7 NC A 0.5 GA SO C O AT-1 RR

- VR 10 S0-1-2499-1B D-2 NC A 0.5 CA SO C O AT-1 RR VR-10 S0-1-2499-2A D-6 NC A O.5 GA SO C* O AT-1 RR VR-10 S0-1-2499-2B D-3 NC A O.5 GA SO C 0 AT-1 RR VR-10 S0-1-2499-3A B-7 NC A 0.5 GA SC C O AT-1 RR VR-10 0-1-2499-3B -2 NC A O.5 GA SO C O AT-RR VR-10 80-1-2499-4A B-6 NC A 0.5 GA SO C O AT-1 RR VR-10 S0-1-2499-4B B-3 NC A 0.5 GA SO C O AT-1

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

RR VR-10

\

r i.

Commonwealth INSERVICE TESTING PREGRAM Edison ISI - CL ASS 1. 2. & 3 VALVES UNIT - 1 QUAD CITIES NUCLEAR POWER STATION SYsitM P 6 ID SEVisl0N - DATE PAGE ATMOSPHERIC CONTAINMENT ATMOSPHERE DILUTION M-642 Sh. 1 ,.1 - 6/18/82 36 of 36 VAtVE NUM8la

[ [ [ 4 REMARKS

/

FCV-1-2599-1A C-1 NC A 1.5 GL AO C O AT-1 RR VR-10 FCV-1-2599-1B C-8 NC A 1.5 GL AO C O AT-1

..=

RR VR-10 AO-1-2599-2A C-4 NC A 1.0 GL AO C O AT-1 RR VR-10 AO-1-2599-28 C-5 NC A 1.0 GL AD C O

.=

AT-1 RR VR-10 AO-1-2599-3A C-3 NC A 1.0 GL AO C 0 AT-1 RR VR-10 AO-1-2599-33 C-6 NC A 1.0 GL AO C O AT-1 RR VR-10 1-2599-23A C-3 NC A 1.0 CK SA C O AT-1 RR VR-10 1-2599-238 C-6 NC A 1.0 CK SA C O AT-1 RR VR-10 1-2599-24A C-3 NC A 1.0 CK SA C 0 AT-1 RR VR-10 1-2599-24B C-6 NC A 1.0 CK SA C O AT-1 RR VR-10 AO-1-2599-4A F-3 NC A 1.0 GL AO C O AT-1 RR VR-10 AO-1-2599-4B E-3 NC A 1.0 GL AO C O AT-1 RR VR-10 FCV-1-2599-5A F-4 NC A 1.0 GL AO C O AT-1 RR VR-10 FCV-1-2599-SD E-4 NC A 1.0 GL AO C O AT-1 RR

=- -

VR-10 h

e O Commonwealth INSERVICE TESTING PREGRAM Edison ISI- CLASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTIM P fr 10 REVIS1001 - DATE PAGE NUCLEAR BOILER INSTRUMENTATION ISI-77 Sh. 1 il- 8/18/82 1 of 36 VALVE NUM8ER p REMARKS AT-2 RR 2-263-2-15A D-5 1 AC O.5 XFC SA O C CT-1 RR VR-9

__ =. ... . . . .. . .... . ..

AT-2 RR 2-263-2-13A D-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-19A C-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 .

AT-2 RR 2-263-2-17A D-5 1 AC O.5 XPC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-11 E-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-54 E-5.6 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-15B D-3 1 AC O.5 XFC SA O C CT-1 RR VR-9

.: - =-- . ........... ..,, ......... .. == ..-

AT-2 RR 2-263-2-13B D-3 1 AC O.5 XFC SA O ~C CT-1 RR VR-9 AT-2 RR .

2-263-2-17D D-3 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-190 C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9

= -

AT-2 RR 2-263-2-20A D-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-23A C-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-310 C-5 1 AC O.5 XFC SA O C CT-1 RR .

- _. . VR-9.. =

O 9

i 1

1 1

Commonwealth INSERVICE TESTING PR2 GRAM Mson ISI- CLASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P fr ID REVISI0ld - SATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-77 Sh. I 1- 8/18/82 2 of 36 VALVE NUM8tR D REMARKS 2-263-2-31G AT-2 RR C-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 2-263-2-31C C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1

- ------ -- RR VR-9

_..

2-263-2-3111 C-5 AT-2 RR

.1 AC O.5 XFC SA O C CT-1 RR VR-9 .

2-263-2-31D AT-2 RR C-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 2-263-2-27 A-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9 2-263-2-25 D-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9

• 2-263-2-31J C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9 2-263-2-31E C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9 2-263-2-31K C-5 AT-2 RR .

1 AC O.5 XFC SA O C CT-1 RR VR-9 2-263-2-23D C-5 AT-2 RR 1 AC O.5 XFC SA O C CT-1 RR VR-9

~

2-263-42A C-5 AT-2 RR 1 AC O.5 XFC SA O j C CT-1 RR VR-9 ,

2-263-2-20B B-5 AT-2 RR 1 AC 0.5 XFC SA O C CT-1 RR VR-9 2-263-2-20C D-3 AT-2 RR 1 AC O.5 XFC SA O C CT-1

-....- .==--- RR, VR-9 '

Commonwealth INSERVICE TESTING . PROGRAM Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSTIM P 6 ID REVISION - DATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-77 Sh. 1 il- J/19/82 3 of 36 VALVE NUMBER

[ [ REMARKS AT-2 RR 2-263-2-23C C-3 1 AC O.5 XFC SA O C. CT-1 RR VR-9 AT RR 2-263-2-31H C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9

... _=.-- -

. - =

AT-2 RR 2-263-2-31T C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9 ,

AT-2 RR 2-263-2-31N C-3 1 AC O.5 XFC SA O C CT-1 RR VF-9

= __-- =-

AT-2 RR 2-263-2-31U C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-31P C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9

.. =.

AT-2 RR 2-263-2-33 D-3 1 AC O.5 XFC SA O C CT-1

__ ... .__ _ __. .=

RR VR-9 AT-2 RR 2-263-2-31V C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-31R C-3 1 AC O.5 XFC SA O C CT-1 RR

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

VR-9 AT-2 RR 2-263-2-31W C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-23D C-3 1 AC O.5 XFC SA O C CT-1

-=- __ =_ . - . - -

RR VR-9 AT-2 RR 2-263-2-42D C-3 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-263-2-200 D-3 1 AC O.5 XFC SA O C CT-1

_ - - RR .

_= -.V-.R- 9 6

e

e 9 -

Commonwealth INSERVICE TESTING PREGRAM Edison ISI - CL ASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P & ID REVISION - DATE PAGE RECIRCULATION ISI-77 Sh. 2 ^1- 8/18/82

4 of 36 VALVE NUMBER

[ [ REMARKS BT CS 45 2-202-5A D-6 1 B 20 GA MO O C PIT RR

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

J-3 BT CS 45 2-202-5B D-3 1 B 28 GA HO O C PIT RR J-3 AT-1 RR BT OP 5 GROUP 1 ISOLATION 2-220-44 E-2 1 A 0.75 GL AO O C PIT RR

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

PST OP AT-1 RR 2-220-45 E-1 1 A 0.75 GL AO O C BT OP S GROUP 1 ISOLATION FST OP

. AT-2 RR 2-220-67A F-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-678 F-5 i AC 0.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-67C E.F-5 1 AC 0.5 XPC SA O C CT-1 RR VR-9

---

- _ = _ .. - - .-

A't-2 RR 2-220-67D F-5 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-09A E-1 1 AC 0.5 XFC SA O C CT-1 RR VR-9

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

AT-2 RR 2-220-09B E-1 1 AC 0.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-67E E-5 1 AC 0.5 XFC SA O C CT-1 RR VR-9

---

=

AT-2 Ed 2-220-67F F-5 1 AC 0.5 XFC SA O C CT-1 RR VR - - - - - - - -

t

e 9 Lo.nmenwealth INCERVICE TECTING PREGRAM Edison ISI - CL ASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTIM P 610 REVISION - DATE PAGE

. RECIRCULATION (CONTINUED) ISI-77 Sh. 2 1 - 8/18/82 5 of 36 VALVE NUMata f [ REMARKS AT-2 RR 2-220-67G E-5 1 AC 0.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-67tl F5 1 AC 0.5 XFC SA O C CT-1 RR

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

VR-9 AT-2 RR 2-263-2-6A B-7 1 AC O.5 XFC SA O C CT-1 RR VR-9

• AT-2 RR 2-263-2-5A D-7 1 AC 0.5 XFC SA O C CT 1 RR VR-9 AT-2 RR 2-220-20A D-5 1 AC 0.5 XFC SA O C CT-1 RR VR-9

= ..... .__ .... ... ....... .. .--

AT-2 RR 2-220-19A B-6 1 AC 0.5 XFC SA C O CT-1 RR VR-9 AT-2 RR 2-220-22A D-8 1 AC 0.5 XFC SA O C CT-1 RR

___. .- =

VR-9 AT-2 RR 2-220-21A D-8 1 AC O.5 XFC SA O C CT-1 RR VR-9 2-220-20D AT-2 RR A-3 1 AC O.5 XFC SA O C CT-1 RR VR-9

- - _ _= .

AT-2 RR 2-220-19B A-3 1 AC 0.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-262-2-60 D-2 1 AC 0.5 XFC SA O C CT-1 RR VR-9

.. . . ............ ...== __ .__ . ___

AT-2 RR 2-262-2-5B B-2 1 AC 0.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-22B D-1 1 AC O.5 XFC SA -

O ...

. C.- CT-1

. _ - - RR ..-- - VR-9 -- ----

4

_ _ m Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CLASS 1.2 & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM PfrID REVIS40N - DATE PAGE RECIRCULATION (CONTINUED) ISI-77 Sh. 2 'l - 8/18/82 6 of 36 VALVE NUMBER 8 g [ g REMARKS AT-2 RR 2-220-21B D-1 1 AC 0.5 XFC SA O C . CT-1 RR VR-9 G

h 9

l.

w -

Commonwealth INSERVICE TESTING PREGRAM Edison ISI - CL A S S 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSifM P fr ID REVIS10N - DATE PAGE CONTROL ROD DRIVE ISI-83 .,1 - 8/19/82 7 of 36 VALVE NUMBER [ REMARKS (177)

• 2-0305-127 D-9 1 B O.75 GA AO C O BT

• VR-6

• SCRAM TESTING (177) 2-0305-126 D-10 1 B 1.0 CA AO C 9 BT

• VR-6

' *SCRAM TESTING (177) 2-0305-114 E-9 2 C 0.75 CK SA C 0 CT-1 *

=_- __- _. . _ . . .. . . . . . _____...

VR-6

• SCRAM TESTING BT CS 2-0302-21A F-2 2 B 1.0 GL AO O C FST CS VR-12 BT CS 2-0302-21D F-7 2 B 1.0 GL AO O C FST CS VR-12 BT CS 2-0302-22 P-3 2 B 2.0 GL AO O C FST CS VR-12 I

? O 9 INSERVICE TESTING PREGRAM

@ Commonwealth Edison ISI - CL A SS QUAD CITIES NUCLEAR POWER STATION

1. 2. & 3 VALVES UNIT - 2 SYSTEM P 1: ID REvtSION - DATE PAGE RESIDUAL llEAT REMOVAL ISI-79 T - 8/18/s2 a of 36 valve NUMBEA [ REMARKS 2-1001-7A D-6 2 B 14 GA HO O O BT OP 90

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

2-1001-7B E-6 2 B 14 GA MO O O DT OP 90 2-1001-7C B-6 2 B 14 GA MO O O BT OP 90 2-1001-7D E-6 2 B 14 GA MO O O BT OP 90 2-1001-67A D-3 2 C 12 CK SA C 0 CT-1 OP -

2-1001-670 E-3 2 C 12 CK SA C 0 CT-1 OP h-10bl-67C D-9 2 C 12 CK SA C O 1 OP

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

2-1001-67D E-9 2 C 12 CK SA C O CT-1 OP 2-1001-125A B-S 2 C 1 RV SA C O CT-2 SY 2-1001-125B E-5 2 C 1 RV SA C O CT-2 SY

...= -- --. -.. . - ..... ..--..

2-1001-126C D-7 2 C 1 RV SA C O CT-2 SY 2-1001-125D E-7 2 C 1 RV SA C O CT-2 SY 2-1001-43A D-4 2 B 14 GA MO C C BT OP 105 ,

2 kUU5-43D E-4 2 B 14 GA MO C C BT OP 105

~

2-iO0i-43C B-o 2 B 14 GA IE'~~ C C BT OP 10s 2-1001-43D E-0 2 B 14 GA MO C C BT OP 105

---

=_. __ =

2-1001-6A F-5 2 E 24 BTF H I4 NA

-= ---

2-1001-6B B-6 2 E 24 BTF M LO NA 2-1001-42A C-5 2 E 14 GA H LC NA

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

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n I , - - -

o n S , - -

_ E 5 6 6 2 2 8 9 2 9 2 m o R - - - - _- - - - -

9 3

3 9

9 3 3 99 _

mi s E C E - - - _ - - - -

od C E _C E B B D B D E B E B E B E _

CE

_ R _

E B

M A B C D A

=B C D U B C D A B _C D A B A B 1 1 1 1 2 2 _ 2 2 N 2 2 2 6 _ 6 -

6 5 5 7 7 4 4 4 4 4 6 . 6 _6

_6 6 4 4 4 4 4 4

_ M E

E V

L 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1 . 1 1

1 1

1 1

1 1

1 1

1 1

T A 0 0 _ 0 0 0 -

0 0 - 0 0,0 0 0 0 - 0 0 00 0 0 S V 0 0 _ 0 0 . 0 0 0 - 0 0 0 0 0 0 - 0 0 Y 1 1 1 1 - 1 1 0 0 0 _

S 1 1 1 1 1 1 - 1 1 - 1

- - - - -_1- - - - - - - - - - - - -

1 1 1 2 2 _ 2 222 2 2 2 2 2 2 2 2 2 2 2 - 2 2 2 _

l

Commonwealth INSERVICE TESTING PROGRAM Edison ISI - CLA SS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM PfrID REVISION -

DATE PAGE RESIDUAL IIEAT REMOVAL (CONTINUED) ISI-81 ,1 - 8/18/82 10 of 36 VALVE NUMBER D REMARKS W

AT-1 RR 2-1001-29A A-5 1 A 16 GA MO C O BT CS 25 J-7 2-1001-29B AT-1 RR A'-7 1 A 16 GA MO C O BT CS 25 J-7 3-1001-47 C-5 AT-1 RR 1 A 20 GA MO O&C C DT CS 40 J-4 GROUP 2 ISOLATION 2-1001-50 D-5 AT-1 RR 1 A 20 GA MO O&C C BT

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

= _ _

CS 40 J-4 GROUP 2 ISOLATION 2-1001-60 A-7 AT-1 RR 1 A 4 GA MO O&C C BT CS 25 J-4 GROUP 2 ISOLATION 2-1001-63 A-6 AT-1 RR 1 A 4 GA MO O&C C BT CS 25 J-4 GROUP 2 ISOLATION 2-1001-68A A-5 PIT RR 1 C 16 CK SA C O CT-1 CS J-2 2-1001-6BB A-6 PIT RR 1 C 16 CK SA C O CT-1 CS J-2 3-1001-16A D-2 2 B 18 GL MO O&C O

= - - - - - - -

BT OP 125 .

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

3-1001-16B D.10 2 B 18 GL MO O&C O BT OP 125

.==

2-1001-18A D-4 2 8 3 GA HO C OEC DT

\

OP 17 2-1001-100 D-7 2 B 3 GA MO C O&C DT

_ - - = _ _ . --

OP 17 3-1001-19A .-__. __.__..-

D-2 2 B 18 GA MO O O

_ - . . _ - - = -

BT CS 125 J-6 2-1001-19B D-9 2 B 18 GA MO O

_ _ _ _ _ - - - _= _===

O BT CS 125 J-6 '

_____.- ..=

2 AT-1 RR C-8 2 A 3 GA MO O&C C

__-1001-20 - _ ........

DT OP 25 '

GROUP 2 ISOLATION ._-

i s

INSERVICE TESTING PREGRAM

@ Commonwealth Edison ISI - CL A S S QUAD CITIES NUCLEAR POWER STATION

1. 2. & 3 VALVES UNIT - 2 SYSTEM PfrID REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED)

ISI-81 8/18/82 11 of 36 VALVE NUMBER [ [ REMARKS

/

• AT-1 RR 2-1001-21 C-8 2 A 3 GA MO O&C C BT OP 25 GROUP 2 ISOLATION 2-1001-22A A-2 2 C RV SA C 0 T-2 SY 2-1001-22B A-9 2 C 1 RV SA C 0 CT-2 SY AT-1 RR 2-1001-23A A-5 2 A 10 CA MO C C BT OP 15 AT-1 RR 2-1001-23B A-6 2 A 10 GA MO C C BT OP 15 AT-1 RR 2-1001-26A A-5 2 A 10 GA MO C C BT OP 15 AT-1 RR 2-1001-26B A-6 2 A 10 GA MO C C BT OP 15 2-1001-28A A-4 2 B 16 GL MO O J BT CS 90 J-7 2-1001-28B k-7 2 B 16 GL MO O O BT CS 90 J-7 AT-1 RR 2-1001-36A B-2 2 A 14 GL MO C O&C BT OP 60 ,

AT-1 RR 2-1001-36B B-8 2 A 14 GL MO C O&C BT OP 60 AT-1 RR 2-1001-37A B-3 2 A 6 GL MO C O&C BT OP 60 AT-1 RR 2-1001-37B B-7 2 A 6 GL MO C O&C BT OP 60

=. .-_. . --

O e

o

INCERVICE TESTING PREGRAM O commonwnlin OUAD CITIES NUCLEAR POWER STATION Edison ISI- CL A SS 1. 2. & 3 VALVES UNIT - 2 SYSTEM P lb 10 REVISION - DATE PAGE RESIDUAL llEAT REMOVAL (CONTINUED) ISI-81 1 - 8/18/82 12 of 36 VALVE NUMBER

[ REMARKS AT-1 RR 2 1001-34A A-2 2 A 16 GA HO C OEC BT OP 125 AT-1 ER 2-1001-34B B-7 2 A 16 GA MO C O&C BT OP 125 2-1001 2A F-3 3 C 12 CK SA C 0 CT-1 OP 2-1001 2B F-3 3 C 12 CK SA C 0 CT-1 OP

• ______....... .____ .. ....... ........ .__.... ........ ..__........... .............. ........ .___...~ . . . . . . . . . .........__..............

2 1001 2C F-7 3 C 12 CK SA C O CT-1 OP 2-1001-2D F.7 3 C 12 CK SA C 0 CT-1 OP 2-1001 5A E-3 3 B 12 GL MO C O&C BT OP 90 2-1001-5B E-7 3 B 12 GL MO C O&C BT OP 90 2-1001 1A G-4 3 E 14 GA M LO NA 2-1001-1B G-4 3 E 14 GA H LO NA l 2-1001-1C G-6 3 E 14 GA M LO NA i

l l

2-1001-ID G-6 3 E 14 GA M LO NA 2-1001 3A G-3 3 E 12 GA M LO NA l . . . _ _ _ . . . . . . . ......__. ...___..___.... .___.... _ _ . . . . . __......._____ .........__.... ........ ..... . ...___.... ....._______........__..

l 2 1001-3D G-3 3 E 12 GA M LO NA

( _ _ _ _ _ _ _ _ . . . . ......... ..___...__..... ........ _ _ _ . . . . .____ ._ . . _ _ _ _ ......... ..__... ....____ __..... ._.....___ .......______. .___.....

2-1001-3C G-7 3 E 12 GA M LO NA 2-t001-3D G-7 3 E 12 GA M LO NA g 2-1001-201A F-3 NC E 14 BTF M LO NA I 2-1001-2010 F-7 NC E 14 BTF M LO NA .

_ _ _ _ _ _ _ . . . . . . .._____.. ........ . . _ _ _ . ........ . . . . . . . .............. ..____ . ....... ____.... ._..... . . _ _ . ~ . . . . . .....................__.

4 4

commonwann INSERVICE TESTING PROGRAM Edison ISI - CLASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM PfrID REVISION -

DATE PAsit RESIDUAL !! EAT REMOVAL (CONTINUED) ISI-81 _1 - 8/18/82 13 of 36 VALVE NUMBER

[ REMARKS r

2-ROO1-33A B-5 3 E 16 GA M LO NA 2-100.-330 B-6 3 E 16 CA M LO NA e

9 9

f

• INSERVICE TESTING PROGRAM 0 Commonwealth Hison ISI - Cl. A S S QUAD CITIES NUCLEAR POWER STATION

1. 2. & 3 VALVES UNIT - 2 SYS11M P f 10 REVISION -

DATE PAGE STANDBY LIQUID CONTROL ISI-82 .k - 8/18/82

. 14 of 36 VMVE NUMBER [ REMARKS 2-1101-15 C-3 1 C 1.5 CK SA C 0 CT-1 CS/RR VR-5 2-1101-16 C-3 1 C 1. -. CK SA C O^ CT-1 CS/RR VR-5

_............ .............. ........ ........ _ . . . . . ....... ........ ....... ........ ....... .. 4_... .......... ...._.......__..--......

2-1106A C-4 2 D 1.5 EXP c 0 DT RR 2-1106B D-4 2 D 1.5 EXP C O DT RR 2 -1101-43 A D-6 2 C 1.5 CK SA C O CT-1 OP ,

2-1101-43B E-5 2 C 1.5 CK SA C O CT-1 OP

• g g .. .. .. ... ........ ................. ........__..____

2-1105D D-5 2 C 1.5 RV SA~ C O CT-2 SY 2-1101-4 E-8 2 E 2.5 GA M LO NA 2-1101-8 D-8 2 E 2.5 GA M LC NA 2-1101-3A D-7 2 E 2.5 GA M LO NA ,

2 1101-3D 3-7 2 E 2.5 GA M LO NA 2-1101-10 D-7 2 E 1 GL M LC NA 2-1101-2A D-5 2 E 1.5 GL M LO NA ,

2-1101-2B E-5 2 E 1.5 GL M LO NA 2-1101-22 C-4 2 E 1.5 GL M LC NA 2-1101-9B D-4 2 E 1 GL M LC NA 2-1101-23 D-3 2 E 1.5 GL M Lo NA 2-1101-1 D-2 1 E 1.5 GL M LO NA 9

g _

Commonwealth INSERVICE TESTING PR2 GRAM Edison ISI- CLA SS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P & 10 REVISION - DATE PAGE RX WATER CLEAN-UP ISI-88 4 - 8/18/82 15 of 36 VA1VE NUMSER

[ p REMARKS AT-1 RR 2-1201-2 B-6 1 A 6 GA MO O C BT OP 30 GROUP 3 ISOLATION PIT RR AT-1 RR 2-1201-5 C-6 1 A 6 GA MO O C BT OP 30 GROUP 3 ISOLATION 9

i l

t 9

1

INSERVICE TESTING PREGRAM O Commonwealth Hisen ISI- CL A SS 1. 2. 1 3 VALVES QUAD CITIES NUCLEAR POWER STATION UNIT - 2 SYSTEM P fr ID REVISION - DATE PAGE REACTOR CORE ISOLATION COOLING ISI-89 1 - 8/18/82 16 of 36 VALVE NUMBER

[ REMARKS AT-1 RR 2-1301 16 D-2 1 A 3 CA MO O C BT OP 25 GROUP 5 ISOLATION PIT RR AT-1 RR 2-1301-17 D-3 1 A 3 CA MO O C BT OP 25 GROUP 5 ISOLATION AT-1 RR 2-1301-40 D-2 NC AC 2 CK SA C C CT-1 RR VR-8 AT-1 RR 2-1301-41 D-2 NC AC 8 CK SA C C CT-1 RR VR-8 AT-2 RR 2-1301-15A D-2 1 AC .5 XFC SA O C CT-1 RR V R-9 AT-2 RR 2-1301-15D D-2 1 AC .5 XFC SA O C CT-1 RR VR-9 E

h

?

l

Commonwealth INSERVICE TESTING PREGRAM Edison ISI - CL A SS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION -

DATE PAGE CORE SPRAY ISI-78 l'- 8/18/82 17 of 36 VALVE NUMBER

[ [ REMARKS CT-1 CS 2-1402-9A C-3 1 C 10 CK SA C O PIT RR J-2

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

CT-1 CS 2-1402-9B C-4 1 C 10 CK SA C O PIT RR 'J-2 2-1402-25A C-2 1 B 10 GA MO C .O BT OP 15

==.--- ...... .-

2-1402-25D C-5 1 B 10 GA MO C 0 BT OP 15 2-1402-24A B-2 2 B 10 GA MO O O BT OP 15 2-1402-24D D-5 2 B 10 GA MO O O BT OP 15 2-1402-28A C-9 2 C 2 RV SA C O CT-2 SY 2-1402-28B D-6 2 C 2 RV SA C O CT-2 SY 2-1402-38A C-8 2 B 1.5 GA MO O C BT OP 12

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

2-1402-38B D-7 2 B 1.5 GA HO O C BT OP 12

• l 2-1402-8A E-9 2 CE 12 SCK SA C/LO O CT-1 OP

= - - .-

2-1402-8B E-6 2 CE 12 SCK SA ,

C/14 O CT-1 OP ,

AT-2 RR 2-1402 31A E-3 1 AC .5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-1402-318 E-3 1 AC .5 XFC SA O C CT-1 RR VR-9 2-1402-6A 1 E 10 GA M LO NA 2-1402-6B D-3 1 E 10 GA M ID NA

==

2-1402-2A G-7 2 E 12 GA M LC NA B

INSERVICE Commonwealth TE@ TING PROGRAM Edison . ISI- CLAS S 1.2.13 VALVES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSTEM P & ID -

REVISION DATE PAGE CORE SPRAY (CONTINUED) ISI-78 1 - 8/18/82 18 of 36

• i WALVE NUMBER

[ [ REMARKS 2-1402-2B G-4 2 E 12 CA M LC NA 2-14b2-34A G-4 2 E 8 BTF M LO NA 2-1402-34D F-3 2 E 18 BTF M LO NA

.== __

2-1402-4A A-8 2 B 8 GL MO C C BT OP 60 2-1402-4D C-7 2 B B GL MO C

• C' BT OP 60 2-1402 13A E-9 2 CE 1.5 SCK SA/M C/LO .O CT-1

• _=- __

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

VR-13 *SEE VR-13 2-1402-138 E-6 2 CE 1.5 SCK SA/M C/LO O CT-1

• VR-13 *SEE VR-13 4

5

e .

Commonwealth INSERVICE TEDTING PROGRAM Edison ISI- CL ASS 1.2 & 3 VALVES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSI!M P 6 ID REVISION - DATE PAGE PRESSURE SUPPRESSION M-76 h - 8/18/82 A9 of 36 VALVE NUMBER

[ REMARKS AT-1 RR 2-1601-21 C-6 NC A 18 BTF AO C C BT OP 10 GROUP 2 ISOLATION FST OP

. AT-1 RR 2-1601-22 C-6 NC A 18 BTF AO C C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 2-1601-55 A-6 NC A 4 GA AO O C BT OP 10 GROUP 2 ISOLATION FST OP

_.== .___.. ._

AT-1 RR 2-1601-56 D-6 NC A 18 BTF AO O C BT OP 10 GROUP 2 ISOLATION z__,--

. FST OP AT-1 RR 2-1601-57 C-9 NC A 1 GL HO O C BT OP 15 GROUP 2 ISOLATION FST OP AT-1 RR 2-1601-58 D-7 NC A 1 GL AO C C BT OP 15 GROUP 2 ISOLATION FST OP AT-1 RR -

2-1601-59 D-7 NC A 1 GL AO O C BT OP 15 GROUP 2 ISOLATION FST OP AT-1 RR 2-1601-20A D-9 NC A 20 BTF AO C O&C BT CS 10 J-8 FST OP AT-1 RR 2-1601-31A D-9 NC AC 20 CK SA C OEC CT-1 OP AT-1 RR 2-1601-20D E-9 NC A 20 BTF AO C OEC BT CS 10 J-8 FST OP O

e

O Commonwealth INSERVICE TESTING PROGRAM Edison ISl- CLASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P fr ID REVISION - DATE PAGE PRESSURE SUPPRESSION (CONTINUED) M-76 [ - 8/18/82 20 of 36 VALVE NUMBER

[ REMARKS AT-1 RR 2 1601 31B E.9 NC AC 20 CK SA C ,0&C CT.1 OP AT-1 RR 2-1601-23 B.3 NC A 18 BTF AO C C BT OP 10 ' GROUP 2 ISOLATION FST OP AT-1 RR 2 1601-24 B-2 NC A 18 BTF AO C .C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 2 1601 60 B.3 NC A 18 BTF AO C C BT OP 10 GROUP 2 I3OLATION FST OP AT-1 RR 2-1601 61 B-2 NC A 2 GL AO C C BT OP 15 GROUP 2 ISOLATION FST OP AT-1 RR 2-1601 62 E.2 NC A 2 GL AO C C BT OP 15 GROUP 2 ISOLATION FST OP i .................... . . . . . . . ................ ......... . . . . . . ................ ........ ....... ........ ..................................

l l AT-1 RR i 2-1601 63 E.2 NC A 6 BTF AO C C DT OP 10 . GROUP 2 ISOLATION PST OP l .................... ....................... ............... ............... .--..... ....... ........ .......... ........................

AT.1 RR 2 8803 C.6 NC A 2 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT.1 RR 2 8004 D.6 NC A 2 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT.1 RR 2 8801A C.3 NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP '

t

- - x

' Commonweahh INSERVICE TESTING PROGRAM Edison ISI- CLASS 1. 2. & 3 VALVES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID -

REVISION DATE PAGE PRESSURE SUPPRESSION (CONTINUED) M.76 tb - 8/18/82 21 of 36 VAtVE NUM8tR [ REMARKS I -

AT-1 RR 2 88010 D.3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 2 8801C D.3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT.1 RR 2 8801D E.3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 2 8802A C= 3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR 2-8802D D.3 NC A O.5 GL AO O C DT OP 10 GROUP 2 ISOLATION FST OP AT-1 RR

• 2 8802C D.3 NC A 0.5 GL AO O C DT OP 10 GROUP 2 ISOLATION FST OP AT.1 RR .

2-8802D E.3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP 3 1601 32A E.2 NC C 18 CK SA C O&C CT.1 OP 2 1601 32D E.2 NC C 18 CK SA C O&C CT.1 OP

..................... ........ ....... . . . . . . . ,......... . . . . . . ................ ........ ....... ........ .......... ....................... f 2 1601 32C E.2 NC C 18 CK SA C O&C CT.1 OP  !

2 1601 32D E.2 NC C 18 CK SA C OEC CT.1 OP t 2 1601 32E E.2 NC C 18 CK SA C OEC CT.1 OP l

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CLASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSifM P fe 10 REVIS10Il - DATE PAGE PRESSURE SUPPRESSION (CONTINUED) M-76 1 - 8/18/82 22 of 36 VALVE NUMBER

[ REMARKS 2-1601-32P E-2 NC C 18 CK SA C O&C CT-1 OP 2-1601-33A E-7 NC C 18 CK SA C O&C CT-1 OP 2-1601-33B E-/ NC C 18 CK SA C O&C CT-1 OP

. _ . __..=

2-1601-33C E-7 NC C 18 CK SA C

- =- O&C CT-1 OP 2-1601-33D E-7 ....----

NC C _ _ . . . . _ _ -

18 CK SA C O&C CT-1 OP 2-1601-33E E-7 ---.. ---.. ._.--

HC C 18 CK SA C O&C CT-1 OP 2-1601-33F E-7 NC C 18 CK SA C O&C CT-1 OP 2-220-81A E-4 NC C 1 CK SA C 0 CT-1 CS VR-14 2-220-81B E-4 NC C 1 CK SA C O

_ _ _ _ - CT CS VR-14 2-220-81C

=- -

.. - 1 . __ ----_. ...

E-4 NC C 1 CK SA C O CT-1 CS VR-14 2-220-81D E-S NC C 1 CK SA C i

=- - _ . - . . . _-

O CT-1 CS VR-14 2-220-81E .-

E-5 . _ - - - - .. . . _ _ - - - - - ..-___----.

NC C 1 CK SA C O CT-1 CS VR-14 l

I 1

1 i

Commonwealth INSERVICE TESTING PR$ GRAM Edison ist - CL ASS 1. 2. & 3 VALVES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE IIIGII PRESSURE COOLANT INJECTION ISI-87 Jf5 - 8/18/82 23 of 36 VALVE NUMBEg

[ [ REMARKS AT-1 RR 2-2301-4 C-9 1 A 10 GA MO O O&C BT CS 50 J-5 GROUP 4 ISOLATION PIT RR 2-2301-5 AT-1 RR B-10 1 A 10 GA MO O O&C BT CS 50 J-5 GROUP 4 ISOLATION 2-2301-3 A-6 2 B 10 GA MO C O BT OP 25 .

2-2301-68 A-6 2 D 16 RPD SA C O * *

• PER MANUFACTURER

-.---. ___= . _ . -- -__. _._ -

RECOMMENDATION 2-2301-69 A-6 2 D 16 RPD SA C O * *

• PER MANUFACTURER

= - - - - _

= - _ - . ..

RECOMMENDATION BT OP 2-2301-29 D-9 2 0 1 GL AO O C FST OP 10 2-2301-34 AT-1 RR *VERIPIED OPEN D-7 2 AC 2 CK SA C O&C CT-1 OP/RR* VR-8 DURING PUMP TEST 2-2301-45 D-8 AT-1 RR *VERFIED OPEN 2 AC 24 CK SA C O&C CT-1 OP/RR* VR-8 DURING PUMP TEST 2-2301-35 E-7 2 B 16 GA MO C OEC BT

---

- -= _

OP' 120 .

2-2301-36 E-9 2 B 16 GA MO C O&C BT OP 120 2-2301-6 P-2 2 D 16 GA MO O OEC BT OP 120 2-2301-20 E-2 2 C '16 CK SA O O CT-1 OP 2-2301-14 .__.

C-6 2 0 4 GL MO C O&C BT OP 26 2-2301-39 E-8 2 C 16 CK SA C O CT-1

• VR-7 *SEE VR-7 2-2301-40 D-7 NC C 4 CK SA C O CT-1 **

VR-13 * *SEE VR-13 4

0

A

_A_

_ w Commonwealth INSERVICE TESTING PROGRAM , ,

Edison ISI- CLASS 1,2.13 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTIM PfrID REVISION - DATE PAGE IIIGil PRESSURE COOLANT INJECTION (CONTINUED) ISI-87 1 - 8/18/82 24 of 36

/

VALVE NUMBER [ REMARILS 2-2301-8 D-6 2 B 14 GA MO C O BT OP 45 PIT RR 2-2301-7 D-6 2 C 14 CK SA C O CT-1 CS J-2

• VERIFIED OPEN 2-2301-74 B-8 2 CE 12 SCK SA C/ID 0 CT-1 OP* DURING PUMP TEST AT-2 RR 2-2301-2G D-9 1 AC .5 XFC SA O C CT-1 RR VR-9

=_

AT-2 RR 2-2301-27 D-9 1 AC .5 XFC SA O C CT-1 RR VR-9 2-2301-22 D-1 2 B 16 GA M LO NA 2-2301-56 P-8 2 E 16 BTF M LO NA 2-2301-71 D-7 2 CE 2 SCK SA/M C/LO O CT-1 OP*

• VERIFIED OPEN DURING PUMP TEST 2-2301-9 D-5 2 D 14 GA MO O O BT OP 45 2-2301-10 E-5 2 B 12 GL MO C C BT OP 60 t

O

e Commonwealth INSERVICE TESTING PROGRAM Hison ISI - CL ASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P in ID REVIS1001 -

DATE PAGE MAIN STEAM ISI-60 Sh. 1 1 - 8/12/82 25 of 36 VALVE NUM81R [ [ REMARKS AT RR BTP OP BT CS 5 J-1 2-203-1A P-4 1 A 20 GL AO O C FST CS GROUP 1 ISOLATION PIT RR

.= -

AT RR BTP OP .

BT CS 5 J-1 GROUP 1 ISOLATION 2-203-1B D-4 1 A 20 GL AO O O FST CS PIT RR AT RR BTP OP BT CS 5 J-1 GROUP 1 ISOLATION 2-203-1C C-4 1 A 20 GL AO O C FST CS

+

I PIT RR AT RR DTP OP BT CS 5 J-1 GROUP 1 ISOLATION 2-203-1D D-4 1 A 20 GL AO O C FST CS PIT RR AT-1 RR 2-220-1 E-4 1 A 3 GA MO C C BT OP 35 GROUP 1 ISOLATION

.==

. =

BT I I VR-1 *SEE VR-1 2-203-3A F-4 1 BC 6 ERV/SV PS/SA C 0 CT-2 '

= _ - - - --

RR VR-2 BT *

• 2-203-3B D-6 VR-1 *SEE VR-1 1 BC 6 ERV PS C O CT-2 RR VR-2

. - - - _ _ = ...___

BT *

• 2-203-3C C-7 VR-1 *SEE VR-1 1 BC 6 ERV PS C O CT-2 RR VR-2

_ _ . = . --- ___- __.. .-  ;

BT *

• 2-203-30 D7 VR-1 *SEE VR-1 l 1 BC 6 ERV PS C O CT-2 RR j

== --

VR .2 --= ---

J I

l g -

Commonwealth INSERVICE TESTING PREGRAM Edison ISI - C L A S S 1. 2. & 3 VALVES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID REVISION - DATE PAGE MAIN STEAM (CONTINNED) ISI Sh. 1 1. - 8/18/82 26 of 36 D

WALVE NUMBER

[ [ REMARKS BT *

• VR-1 2-203-3E D-7 1 BC 6 ERV PS C O CT-2 RR VR-2 2-203-4A F-8 1 C 6 SV SA C O CT-2 RR VR-15 b-2b3-4B D-5 1 C 6 SV SA C 0

b2 hk VR-15 2-203-4C C-5 1 C 6 SV SA C 0 CT-2 RR VR-15 .

2-203-4D B-5 1 C 6 SV SA C O CT-2 R3 VR-15 2-203-4E F-8 1 C 6 SV SA C 0 CT-2 RR VR-15.

2-203-4F D-5 1 C 6 SV SA C 0 CT-2 EE VR-15 2-203-4b C-5 ,1 C 6 SV SA C 0 b-2 RR VR-15 2-203-411 B-5 1 C 6 SV SA C

---

~ ~~

0 CT-2 RR VR-15 F

a s . l

INSERVICE TESTING PREGRAM

@ Commonwealth Edison ISl- CLASS 1. 2, & 3 VALVES QUAD CITIES NUCLEAR POWER STATION UNIT - 2 SYS11M.

P fr ID REVISION -

DATE PAGE MAIN STEAM (CONTINUED) ISI-60 Sh. 2 1 - 8/18/82 27 of 36 VALVE NUMBER

[ [ REMARKS AT-1 RR BTP OP BT CS 5 J-1 GROUP 1 ISOLATION 2-203-2A E-7 1 A 20 GL AO O C FST CS PIT RR AT-1 RR' BTP OP BT CS S J-1 GitOUP 1 ISOLATION 2-203-20 E-7 1 A 20 GL AO O C FST CS PIT RR

= - . .

AT-1 RR BTP OP BT CS 5 J-1 GROUP 1 ISOLATION 2-203-2C D-7 1 A 20 GL AO O C FST CS PIT RR AT RR BTP OP BT CS 5 J-1 GROUP 1 ISOLATION 2-203-2D D-7 1 A 20 GL AO O C FST CS PIT RR 2-220-2 AT-1 RR .

E-7 1 A 3 GA Ho C C BT OP 35

---

= -

GROUP 1 ISOLATION AT-2 RR 2-220-17A E-8 1 AC O.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-17B D-8 1 AC 0.5 XFC SA O C CT-1 RR VR-9 2-220-17C AT-2 RR C-8 1 AC 0.5 XFC SA O C CT-1 RR VR-9 AT-2 RR 2-220-17D D-8 1 AC 0.5 XFC SA O C CT-1 RR VR-9

g Commonwealth INSERVICE TE'3 TING PREGRAM Edison ISI - CLA S S 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTIM P fr ID REVISION - DATE PAGE MAIN STEAM (CONTINUED) ISI-60 Sh. 2 :al- 8/18/82 28 of 36 VALVE NUMBER

[ REMARKS 2-220-18A E-8 AT-2 RR 1 AC .5 XFC SA O C CT-1 RR VR-9 2-220-18B D-8 AT-2 RR 1 AC .5 XFC SA O C CT-1 RR VR-9 AT-2 2 220-18C C-8 1 AC .5 XFC SA O RR C CT-1 RR VR-9 2-220-18D B-8 AT-2 RR

, 1 AC .5 XFC SA O C M-1 RR

~~-- -- -

VR-9 f

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P fr 10 REVIsl0N - DATE PA6E FEEDWATER ISI-62. .1 - 8/18/82 29 of 36 VALVE NUMBER [ [ [ 4 REMARKS 2-220-5BA AT-1 RR E-3 1 AC 18 CK SA O C CT-1

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

RR VR-3 AT-1 RR 2-220-50B F-3 1 AC 18 CK SA O O&C CT-1

-= --

RR VR-3 2-220-59B F 2 C 18 CK SA O C CT-1

_.-3...

RR VR-3 3-220-62A AT-1 RR E-3 1 AC 18 CK SA O C CT-1

= -

RR VR-3 2-220-62B AT-1 RR P-3 1 AC 18 CK SA O O&C CT-1 RR VR-3

commonweatih INSERVICE TESTING PROGRAM Edison ISI- CL ASE 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION svSTEM P & ID REVISION -

DATE PAGE SERVICE WATER ISI-69 8/18/82 30 of 36 WALVE NUMBER

[' -

REMARES 2-3999-86 B-9 3 C 8 CK SA C O CT-1 OP 2-3999-88 B-10 3 C 8 CK SA C O CT.1 OP e

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Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CLASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE DIESEL GENERATOR STARTING AIR (SERVICE AIR SYSTEM) M-72 ~1;- 8/18/82 31 of 36 VAWE NMER A [ 4 REMARKS 2-4699-121 E-9 NC E 1.5 GA M LO NA 2-4699-122 E-9 NC E 1.5 CA M LO NA 2-4699-225 D-8 NC E 1.5 BALL M I4 NA 3-4699-123 E-9 NC C 1.5 CK SA .C O CT-1 OP

-. .= -- - -

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

2-4699-196 E-9 NC C 1.5 CK SA C O CT.1 OP

• 2-4699-226 D-8 NC B 1.5 GL AO C O BT OP 5 t

Commonwealth INSERVICE TESTING PR2 GRAM Edison ISI- CL A SS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P & ID REVISION - OATE PAGE INSTRUMENT AIR PIPING M-71 Sh. 2 15 - 8/18/82 32 of 36 VALVE NUMBER

[ REMARKS AT-1 RR 2-4720 D-3 NC A 1 GA AO O C BT OP 10 GROUP 2 ISOLATION FST OP

-=. -

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

AT-1 RR 2-4721 D-3 NC A 1 GA AO O C BT OP 10 GROUP 2 ISOLATION FST O?

AT-1 RR

• 2-733-1 F-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION AT-1 RR 2-733-2 P-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION AT-1 RR 2-733-3 F.7 NC A 0.375 BALL SO C C BT OP S GROUP 2 ISOLATION

.. = . ~ - . .----- ----. ==

AT-1 RR 2-733-4 F-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION ATal RR 2-733-5 F-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION AT-1 RR 2-743 0-7 NC AC 0.375 CK SA O C CT-1 RR ,VR-4 2-736-1 P-7 NC D 0.375 EXP O C DT RR 2-736-2 P-7 NC D 0.375 EXP O C DT RR

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

2-736-3 P-7 NC D 0.375 EXP O C DT RR 2-736-4 P-7 NC D 0.375 EXP O C DT RR

.. _=

2-736-5 F-7 NC D 0.375 EXP O C DT RR

-.- .-_. . .--. _-- . t

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INSERVICE TESTING PROGRAM

@ commonwealthQUAD CITIES NUCLEAR POWER STATION Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 2 SYSTEM P fr ID REVISION -

DATE PAGE RX BUILDING EQUIPMENT DRAINS M-85 1 - 8/18/82 33 of 36 VALVE NUMBER 4 y [ [ REMARKS AT-1 RR 2-2001-15 E-3 NC A 3 GA AO C C BT OP 20 GROUP 2 ISOLATION FST OP AT-1 RR 2-2001-16 E-3 NC A 3 CA AO C C BT OP 20 GROUP 2 ISOLATION FST OP AT-1 RR l 2-2001-3 P-7 NC A 3 GA AO C C BT OP 20 GROUP 2 ISOLATION l

FST OP A? 1 RR 2-2001-4 P-7 NC A 3 GA AO C C BT OP 20 GROUP 2 ISOLATION FST OP

.... . . .. . - - - ~~..--- - - -

l l

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CL A SS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID REVISION - DATE PAGE DIESEL GENERATOR FUEL OIL M-29 1.1 - 8/18/82 34 of 36 VALVE NUMFf7 4 5 [ REMARKS 2-5299 E-4 NC C 1.5 CK SA C O CT-1 OP 2-5201 E-3 NC B 1 GA SO C O DT OP 5 2-5199-155 0-5 NC E 1 GL M LO NA

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

2-5199-157 C-5 NC C .5 CK SA C C CT-1 OP e

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Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CL ASS 1. 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTIM P610 REVISI001 - DATE PAGE CONTAINMENT ATHOSPIIERE MONITOR M=641 Sh. 2 ',1 - 8/18/82 35 of 36 WALVE NUMBER [ REMARKS 30-2-2499-1A D-7 NC A O.5 GA So C O AT-1 RR VR-10 30-2-2499-1D D-2 NC A O.5 GA SO C O AT-1 RR VR-10 SO-2-2499-2A D-6 NC A O.5 GA SO C O AT-1 RR VR-10 bO2-2499-2D D-3 NC A O.5 GA SO C O AT-1 RR VR-lO SO 2-2499-3A D-7 NC A O.5 GA SO C 0 AT-1 RR VR-10 S0-2-2499-3D D-2 NC A O.5 GA SO C O AT-1 RR VR-lO So-2-2499-4A D-6 NC A O.5 GA SO C O AT-1 RR VR-lO G0-2-2499-4D D-3 NC A O.5 GA SO C O AT-1

-- - RR VR-10 e

e

Commonwealth INSERVICE TESTING PROGRAM Edison ISI- CLASS 1. 2. & 3 val.VES UNIT - 2 OUAD CITIES NUCLEAR POWER STATION SYSTIM PfrID REVISION - DATE PAGE ATHOSPl!ERIC CONTAINMENT ATMOSPl!ERE DILUTION -

M-642 Sh. 2 .1 ~ 8/18/82 36 of 36 VAtVE NUMBER

[ REMARKS FCV-2-2599-1A C-1 NC A 1.5 GL AO C O AT-1 RR VR-10 FCV-2-2599-1B C-8 NC A 1.5 GL Ao C 0 AT-1 RR'

= =---

VR-10

_=__=

AO-2-2599-2A C-4 NC A 1.0 GL AO C O AT-1 RR VR-10 AO-2-2599-2B C-5 NC A 1.0 GL AO C. O AT-1 RR VR-10 AO-2-2599-3A C-3 NC A 1.0 GL AO C O AT-1

~

RR VR-10 AO-2-2599-3D C-6 NC A 1.0 GL AO C O AT-1 RR VR-10 2-2599-23A C-3 NC A 1.0 bK SA C O AT-1 RR VR-10

-2599-23B C-6 NC A 1.0 CK SA O AT-1 RR VR-10 2-2599-24A C-3 NC A 1.0 CK SA C O AT-1 RR VR-10 2-2599-24B C-6 NC A 1.0 CK SA C O AT-1 RR VR-10 AO-2-2599-4A F-3 NC A 1.0 GL AO C O AT-1 RR VR-10 AO-2-2599-4D E-3 NC A 1.0 GL AO C O AT-1

- - - = = - - - -

RR VR-10 FCV-2-2599-5A F-4 NC A 1.0 GL AO C O AT-1 RR

---

.VR-10 FCV-2-2599-5B E-4 NC A 1.0 GL AO C O AT-1 RR VR-10 s

. _ _ . _ . . _ . . _ _ . _ . . . _ . _ _ - . - . _ _ . _ - - - -_- .. _m . ._. . _ _ _ _ . _ _ _ - . . . _ - . _ - <- - - .

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j- SECTION 5.3 j RELIEF REQUESTS FOR INSERVICE-VALVE TESTING PROGRAM i' i i

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10 COM-06-003 Revision 1 5-13 J

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RELIEF REQUEST NO. VR-1 SYSTEMi Main Steam COMPONENTi - 1(2)-203-3A, B, C, D, E 3A-Target Rock Safety Relief Valve 3B-E-Electromatic Relief Valves.

CATEGORYi B/C FUNCTION: 1) Open upon receipt of an auto depressurization '

signal to blowdown reactory 2) Act as a primary system relief valve which actuates on high system

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pressure.

TEST'REOUIREMENT: BT -

Exercise and time vslves every three months.

i BASIS FOR'RELIEFi Relief is requested from the Section XI required testing frequency of once every three months. These electromatic relief valves are not tested routinely during reactor operation because of the resultant primary system pressure transients.

In addition, a failure of any valve to close would cause an uncontrolled, rapid depressurization of the

()

COM-06-003 l Revision 1 5-14 )

l

RELIEF REQUEST NO. VR-1 (CONTINUED) primary system resulting in undesirable thermal gradients in the reactor vessel. Excessive testi'ng of those valves is inadvisable because each relief valve discharge'to the suppression pool detracts from the limited fatigue life of the containment.

These valves cannot be tested at cold shutdown or refueling since a system pressure of greater than 150 psig is needed to actuate the valves. Surveil-lance testing of these valves is, therefore, completed.at very, low reactor power levels. Verifi-

.(]) cation of relief valve actuation is accomplished by first opening a turbine' bypass valve, actuating the-relief valve, and then observing a corresponding closure response of the turbine bypass valve.

The-frequency of such testing requested herein is that submitted by Quad Cities Station in a proposed Technical Specification change required by the  !

August 3, 1977 letter from Don K. Davis (NRC-DOR) to Commonwealth Edison Company. In this Technical Specification change, a program was committed to which specified a variable testing frequency related i l

to demonstrated reliability and operability. The )

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COM-06-003 Revision.1 5-15

w

- RELIEF REQUEST NO. VR-1 (CONTINUED) testing interval is based on the number of valve failures during the required test interval. _The frequency ranges from a maximum of 18 months to a minimum of 31 days. This testing frequency is provided to ensure operability and demonstrate reliability of the valves. .Since the frequency varies with observed valve failures, this proposed testing scheme should result in a uniform level of reliability.

ALTERNATE TESTING: The following schedule will be used to

() determine the required test interval.

Number of Relief Valves Found Inoperable Next Required During Testing or Test Interval Test-Interval D 18 months j- 25%

1 84 days j- 25%

2 92 days j 25%

>3 31 days j- 25%

Additionally, stroke times for these valves will not be measured since there is no position indication circuitry to shou disc movement.

o COM-06-003 Revision 1 5-16

RELIEF REQUEST NO. VR-2 SYSTEMi- Main-Steam COMPONENTi 1(2)-203-3A (Target Rock Safety / Relief) 1(2)-203-3B,C,D,E (Electromatic Relief)

CATEGORY: BC FUNCTIONi 1) Open upon re-'ipt of an auto depressurization signal to blow down the reactor, and 2) act as a primary system relief valve actuating on a high pressure condition. The Target Rock Safety / Relief Valve functions the same as above except, it also

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acts as a safety valve.

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TESTREOUIREMENTi CT-2 -

Verify pressure set point in

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accordance with IWV-3510.

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BASIS 1FOR RELIEFi The electromat'ic relief valves and the relief functior cf the Target Rock valve are operated by actuation of a pilot solenoid valve which opens the main valve by applying a pressure differential to the main valve piston. The pilot valve is actuated from an electric signal from either the control switch, the aute-depressurization logic, or a

() pressure switch that senses system pressure.

COM-06-003 -

Revision 1 5-17

RELIEF REQUEST NO. VR-2 (CONTINUED)

, The requirement of IWV-3512 to check relief and safety valve set points in accordance with PTC-25.2-1976 is not applicable in this case. Therefore, re' lief is requested from compliance with this requirement.

The pressure set point of these valves is set by i

calibrating the pressure switch rather than testing

, the complete valve assembly. The combination of the pressure switch calibration and the exercising test

,. for operability (BT) satisfies the intent of Q paragraph IWV-3510.

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ALTERNATE TESTINGi The pressure switch for each of these valves will be calibrated to verify the correct set point and the exercise test (BT)-will verify operability of the valve.

O COM-06-003 Revision 1 5-18

RELIEF REQUEST NO. VR-3 i SYSTEMi FEEDWATER COMPONENTi 1(2)-220-58A, B 1(2)-220-59B 1(2)-220-62A, B CATEGORYi C & AC FUNCTIONi The 58 and 62 valves close for containment isolation. The 59B valve closes for HPCI injection.

() TEST' ' REQUIRE'M ENT'i CT-1 -

Exercise check valve every three months.

BASIS EFO'R' RELIEF'i These check valves cannot be tested for operability during reactor operation because the feedwater system is needed to maintain primary coolant inventory. It is impractical to test these valves during cold shutdown because the reactor water clean-up and feedwater systems are generally I required to be operable. In addition, to verify that these check valves stroke to the full closed position, a leak rate test must be performed. Since leak rate testing is performed only during refueling COM-06-003 Revision'l 5-19 0

RELIEF REQUEST NO. VR-3 (CONTINUED) outages, these valves will be demonstrated to be in the full closed position at each refueling outage.

ALTERNATE'TESTINei These check valves'will be exercised closed during each reactor refueling outage.

O O

COM-06-003 Revision 1 5-20

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l RELIEF REQUEST NO. VR-4 SYSTEMi Neutron Monitoring System 1

COM'PONENTi 1(2)-743 i

CATEGORYi C l

FUNCTIONi Primary containment isolation valve for the T.I.P.

System nitrogen purge line.

~~

TEST REQUIREMENTi CT Exercise valve every three months, i

B'A SIS FbR'RELIEFi. This check valve cannot be exercised for

~~

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operability every three months because the T.I.P.

system is required to be purged constantly during operation. Since there is no external means of position indication, the system must be taken out-of-service and a leak rate test performed to verify operability. Since leak rate testing is performed only during refueling outages, these valves will be demonstrated to be in the full closed position at each refueling outage.

~

ALTERNATE' TESTING'i The valves will be full stroke exercised each refueling outage.

C)

COM-06-003 Revision 1. 5-21

RELIEF' REQUEST NO. VR-5 O-SYSTEMi Standby Liquid Control COM'PONENT'i 1(2)-1101-15, 16 CnTEGdRYi C FUNCTION'i The safety function of these check valves is to open upon a system injection.

TEST'R'EQUIREMENTi CT Exercise valve every three months.

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BASIS FOR' 'REL'IE'F i Exercising these valves by system initiation is not feasible during operation due to the require-ments to maintain (a) boron to reactor water separa-tion, and (b) requirements to maintain system operability per Technical Specifications.

Since the valve operability test, in this case, must be performed with the system out of service by injecting clean demineralized water from some exter-nal source, it is more practical in terms of system availability to perform this test during cold shutdown. Currently it is not possible to achieve full flow through the valves using the method

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COM-06-003 Re'. ision 1 5-22 i

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RELIEF REQUEST NO. VR-5.(CONTINUED) described above, only 26 gpm of_the required 39 can be injected. However, the station has developed a methoduto achieve full flow through these valves during refueling outages using the system pumps.

ALTERNATE' TESTING'i These valves will be part stroke exerciseo during cold shutdown and full stroke exercised at each refueling outage.

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O COM-06-003

' Revision 1 5-23

RELIEF REQUEST NO. VR-6 SYSTEMi Control Rod Drive

~

COMPONENTi 1(2)-0305-126, 127, 114 CATEGDRYi B &.C FUNCTIDN'i These valves operate on a scram signal to drive the control. rods in.

TEST'~REQUIREMENTi BT - Exercise and time valves every 3 months.

CT Exercise valves every three months.

O BASIS ' FOR' ' RELIEF'i There are 177 of eeth of the valves listed, i.e.,one for each of the 177 control rod drives.

The proper operation.of each of these valves is-

demonstrated during scram testing. During scram testing each drive's scram insertion time is measured. The Technical Specifications limit indi-vidual scram insertion times to specific values.

This insures that the above mentioned valves are functioning properly.

i

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i COM-06-003 Revision 1 5-24

RELIEF.REOUEST NO. VR-6 (CONTINUED)

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~~

ALTERNATE TESTING'i Individual scram insertion tests will be performed per the Technical Specifications frequency. The frequency is: 1) 100% of control rod drives after each refueling with reactor power equal to.or less than 30%, and 2) 50% of the CRD's every 16 to 32 weeks with 100% completed every year.

4 VO i 1

O COM-06-003 Revision 1 5-25 l

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4 RELIEF REQUEST NO. VR-7 h

SYSTEM: High Pressure Coolant Injection COMPONENTi 1(2) - 2301-39 CATEGJRYi C FUNCTIbNi See Basis for Relief

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TEST REOUIREMENfi CT-1 Exercise check valve every three months.

BASIS'FORRELIEFi This valve is designed to prevent backflow

(} into the suppression pool in the event of a pump suction shift from the contaminated condensate storage tank (CCST) to the suppression pool. The safety related stroke direction of this valve is in the open direction to provide suction flow to the-HPCI pump. There is no acceptable method for verifying this valve's ability to swing to its full ,

open position. The system test circuit utilizes the CCST as the pump suction rather than the suppression pool. The suppression pool is not used as the pump suction for testing because of the desire to keep the system free of the dirt and contamination typically found in torus water.

O COM-06-003 Revision 1 5-26

RELIEF REQUEST NO. VR-7 (CONTINUED)

O In lieu of the Code required full stroke test, Commonwealth Edison proposes to demonstrate valve operability by disassembling the valve and verifying ,

that.the disc swings freely to the full open position. Since this valve is not normally used, there will be no expected wear-induced degradation of the valve internals. Therefore disassembly and inspection of these valves once every third refueling outage is felt adequate to insure valve operational readiness.

~

ALTERNATETESTINGi Each valve will be disassembled every third

)

refueling outage to verify that the disc swings freely to the full open position.

COM-06-003 Revision 1 5-27 )

1:

RELIEF REQUEST NO. VR-8 O <

SYSTEMS High Pressure Coolant Injection, Reactor Core Isolation Cooling b

~

COMPONENTi 1(2)-2301-34, 45, 1(2)-1301-40, 41 CnTEGORYi C FUNCTIONi Primary containment isolation.

TESTREQUIREMENTi CT-1 -

Exercise valve for operability every three months.

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BASIS ' 'FOR ' ' RELIEF i It is impractical to demonstrate closure of 4

these check valves during normal operation or cold

, shutdown'. To verify closure upon reversal of flow a l

pressure test must be performed. This requires that l

the systems be taken out-of-service. The safety significance of these components is minimal since leakage past these va.1ves would be contained within the HPCI and RCIC piping which returns to the containment.

O

-COM-06-003 Revision 1 5-28  ;

I 1

RELIEF REQUEST NO. VR-8 (CONTINUED)

O

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ALTERNATE TESTINGi These valves will be demonstrated to close upon reversal of flow during each refueling outage per Appendix J test.

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COM-06-003 Revision 1 5-29

RELIEF REQUEST NO. VR-9 0

SYSTEMi' Nuclear Boiler Instrumentation, Recirculation, Reactor Core Isolation Cooling, Core Spray, High Pressure Coolant Injection, Main Steam. .

COMPONENT: Excess flow check valves as listed in program.

CATEGORYi AC

~

FUNCTIONi Limit flow (leakage) from instrument lines penetrating primary containment; perform containment isolation function.

TEST'REQUIREMENTE AT - Seat leak rate test.

CT-1 Exercise check valves to the closed position every three months.

BASIS'FOR'RELIEFi

~

These valves are currently tested per Technical Spec'ification requirements which consists of a leakage test conducted during primary system i ,

pressure tests at the completion of each refueling outage. The testing involves uncoupling the instru-ment lines and verifying that each valve strokes to the closed position. The operator also observes that the valve limits flow to an acceptable level.

C:t COM-06-003 Revision 1 5-30

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RELIEF REQUEST NO. VR-9 (CONTINUED) )

This method and frequency of testing has been justified in the plant FSAR and has proven to be an adequate verification of valve performance.

1 ,

ALTERNATE TEST AGi These valves will be tested in the manner-described above prior to start-up from each refueling outage.

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COM-06-003 Revision 1 5-31

._ . __ . _ _ . _ _ _ _ . _ . _ . _z

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RELIEF REQUEST NO. VR-10 SYSTEMi 'All Systems

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COMPONENTi All primary containment isolation valves (listed in program as Category.A).

CATEGORYi A FUNCTION'i Primary containment isolation.

TEST'RE'QUIREMENTi AT - seat leakage tests per IWV-3420.

l

{) B ASI$ 1 FOR '. RELI'EF':' Primary containment isolation valves whose functional differential pressure does not exceed the primary containment accident pressure will be seat leak tested in accordance with the Appendix J i requirements of 10CFR50, type C tests. At this functional differential pressure Section XI testing requirements are essentially equivalent to those of Appendix J. No additional information concerning valve leakage would be gained by performing separate tests to both Section XI and Appendix J.

ALTERNATE TESTING'i Valves will be seat leak tested in accordance with 10CFR50 Appendix J, type C tests.

O

! COM-06-003 l Revision 1 5-32

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

i-P I

RELIEF REQUEST NO. VR-ll O

i Specific relief is requested from requirements of paragraphs

, IWV-3417(b) and IWV-3523 of Section XI of the 1980 Edition of the

' ASME Boiler and Pressure' Vessel Code including the Addenda

. through Winter 1980. These paragraphs state the corrective actions to be.taken when. valves fail to exhibit a required change

.of disk position. These actions include requirements to take corrective action prior to plant startup should a failure occur during cold shutdown testing. .Also stated are requirements to declare valves. inoperable if corrective action is unsuccessful I

within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

These paragraphs do not take into account the plant Technical

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Specification requirements for limiting conditions for operation.

, which state the minimum conditions necessary for safe operation

~

of the plant. The failure of a particular valve may not neces-

sarily require a plant shutdown or prevent a startup. In addi-tion, valves not capable of performing'their safety-related func-i tion are declared inoperable as soon as that condition has been

, verified, not'after a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period has elapsed.

i For these reasons, Quad Cities Station will evaluate the condi-i tion of each valve with respect to its safety related function and take the appropriate corrective action as stated in the Tech-nical Specification-Limiting Condition for Operation.

lC1 l

COM-06-003 Revision 1 5-33 f

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RELIEF REQUEST NO. VR-12 O

SYSTEM: Control Rod Drive COMPONENT'i 1(2) - 030'2-21 A & B f

1(2) - 0302-22 CATEGORYi B

FUNCTION

Scram discharge volume vent and drain valves.

)

i TESTI.REOUIREMENTE Full-stroke timing as required by IWV-3413(b). .

O BASIS FOR' RELIEF'i These valves are normally in the open position to allow water which enters the scram discharge volume.from normal CRD leakage to drain into the reactor building equipment drain tank. This assures that a sufficient volume is always available to accept scram discharge water following a scram. The system is designed such that the test circuit bleeds the air from these air-operated valves at a very slow rate; much slower than during normal operation 1

of the valve. Timing these valves during testing, therefore, has no relevance, and because of the slow i

bleed rate the test time repeatability is poor.  :

I).

COM-06-003 Revision 1 5-34

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RELIEF REQUEST NO. VR-12 (CONTINUED)

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ALTERNATE'TESTINGi These valves will be full stroke exercised and fail-safe operability will be observed without timing during each cold shutdown. (See Justification No. J-9 for further information on testing frequency).

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O COM-06-003 Revision 1 5-35

RELIEF REQUEST NO. VR-13 SYSTEMi. Core Spray, Residual Heat Removal, High Pressure Coolant Injection COMPONENTi 1(2)-1402-13A&B  ;

1(2)-1001-142A,B,C,&D 1(2)-2301-40 CATEGORYi C& CE FUNCTIONi Pump minimum flow line chech and stop-check valves required to open during pump low flow conditions for pump cooling.

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TEST REOUIREMENTi CT-1 Exercise valve every three months.

BASISJFOR' REL'IEFi There are no provisions in the current system design for exercising or determining the position of these valves. Based on the recora of satisfactory pump performance and lack of pump overheating

~

problems, it is evident that these valves have performed in an acceptable manner. However, due to i the inability to demonstrate that the valves stroke l

j open, Quad Cities Station has initiated system modifications to install flow instrumentation in the O

COM-06-003 Revision 1 5-36

RELIEF REQUEST NO. VR-13 (CONTINUED) j O .

minimum flow lines to indicate that the valves are, in fact, opening and passing adequate flow for pump cooling purposes. The modifications must be made during a refueling outage because the system sill be  !

out of service during the installation. The modifi-cations have been initiated and will be completed at the earliest possible date which is contingent on.

the availability of materials. Relief is therefore requested'from the requ'irement to demonstrate that ,

the subject valves stroke open until system modifications provide the necessary instrumentation.

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ALTERNATETESTINGi No specific alternate test is applicable during this interim period.

(

O COM-06-003 Revision 1 5-37

RELIEF.REOUEST NO. VR-14 SYSTEMi Main Steam i

COMPONENT: 1(2)-220-81A,B,C,D & E

\

CATEGORYi C Vacuum breakers for the main steam relief valve

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FUNCTIONi discharge lines.

TESTREOUIREMENT: CT-1 Excercise check valve in the open direction every three months.

O BASIS FOR RELIEFi These check valves have no external means of actuation for exercising. The only practical method for exercising these valves open is by manually pushing the disc from its seat using a small diameter rod. Since this requires access to the valves which are located within primary containment, the test must be deferred to cold shutdowns when the primary containment is de-inerted.

ALTERNATETESTINGi These check valves will be verified to freely swing to their full open position at cold shutdowns when the drywell is de-inerted.

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RELIEF REQUEST NO. VR-15 l

' SYSTEM: Main Steam COMPONENT . 1(2)-203-4A through 4H

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CATEGORY: C FUNCTION: Gafety relici valves for the primary coolant pressure boundary.

TEST REQUIREMENT: CT-2, Verify safety valve set point BASIS FOR RELIEF:'

(]) It is impractical for Quad Cities Station to meet the requirements of IWV-3512, in that "as-found" set points for these safety relief valves cannot be determined. The station has no on-site facility for testing safety valve set points.

Carrently, these valves are being removed from the system, cleaned and rebuilt, and then shipped off-site for re-verification of valve set points.

l Therefore, IWV-3512 cannot be applied because "as

< found" set points are not verified.

The frequency of removal and maintenance of these

[ valves, however, is on a greatly accelerated basis

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COM-06-003 Revision 1 5-39

RELIEF REQUEST NO. VR-15 (CONTINUED) compared to the Section XI requirements. The Technical Specification frequency for these valves has been to remove one-half (4) of the eight safety valves each refueling outage and replace them with valves that have been rebuilt and verified for proper set point.

This accelerated maintenance schedule provides adequate assurance that these valves will perform reliably.

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ALTERNATE'TESTINGi One-half (4) of the total number of safety

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valves will be removed and replaced with valves that have been rebuilt and had their set points verified each refueling outage.

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SECTION 5.4

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COLD SHUTDOWN JUSTIFICATIONS FOR INSERVICE VALVE TESTING PROGRAM 1

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COLD SHUTDOWN JUSTIFICATIONS Justifications 1 through 8 inclusive are applicable when plant operating conditions are such that specific valves cannot be full-stroke tested. Those valves which have part-stro'ke capabilities will be tested in that manner every three months and full-stroke tested during cold shut-downs. Those valves which cannot be full-stroke or part-stroke tested during plant operation will be full-stroke tested during cold shut-downs. The aforementioned testing meets the requirements of IWV 3412(a).

Justification 9 is applicable when fail-safe valves cannot be tested during plant operation by observing their operation upon a loss of actuator power. These valves will be tested during cold

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shutdowns. This testing meets the requirements of IWV-3415.

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COM-06-003 Revision 1 5-42

_ ' JUSTIFICATION NO. J-1 l SYSTEM: Main Steam FUL'CTION : Primary containment isolation. valves for the main steam lines.

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COMPONENT: 1(2)-203-1A, B, C, D 1(2)-203-2A, B, C, D

'EASONS:

R Full stroke testing these valves during normal reactor operation requires isolating one of the four main steam lines. Isolation of these lines results 4 () in primary system pressure spikes, reactor power fluctuations, and incradsed flow in the unisolated steam lines. This unstable operation can lead to a reactor scram, and as discussed in NUREG-0626 pressure transients resulting from full stroke .

testing MSIVs increase the chances of actuating primary system relief valves.

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i JUSTIFICATION NO. J l O l SYSTEM: Residual Heat Removal, Core Spray, High Pressure Coolant Injection.

COMPONENT: 1(2)-1001-68A & B 1(2)-1402-9A & B 1(2)-2301-7 '

FUNCTION: Open upon System Injection REASONS: These valves have air-operators and remote position indicators for remote testing purposes. However,

(} during normal operation the high differential pressure across the valve seats prohibits exercising. Additionally, the residual heat removal and core spray system valves (i.e., 1(2)-1001-68A, 68B & 1(2)-1402-9A,9B) are located inside the primary containment which is inerted with nitrogen during normal operation. The high pressure coolant.

injection valve (l(2)-2301-7) is located inside the main steam isolation valve room which is a designated high radiation area where normal power operation radiation dose rates are one to two rem / hour. Also high temperatures exist in this are J

( (120' to 140*F) which further increases the hazards COM-06-003 Revision 1 5-44

c JUSTIFICATION NO. J-2 (CONTINUED) l involved in entering the area for this testing. The accumulated dose to conduct this test would be approximately 1.5 man-rem..

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COM-06-003 Revision 1 5-45

JUSTIFICATION NO. J-3 O

SYSTEM: Recirculation COMPONENT:- 1(2)-202-5A, B FUNCTION: In a design basis loss of coolant accident, one of .

these valves will close depending on the location of the line break.

REASONS: These' valves cannot be fully stroke tested or partial stroke tested during normal operation since isolation of a recirculation loop would cause a recirculation pump trip. One loop operation is restricted by the Technical Specifications.

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JUSTIFICATION NO. -J-4 O

SYSTEM: RHR-Head Spray /Shu,tdown Cooling Subsystems COMPONENT: 1(2)-1001-60, 63, 47, 50 FUNCTION: Primary containment isolation valves.for RHR Head

' Spray and Shutdown Cooling Subsystems.

REASONS: Relief is requested from partial or full stroke testing these valves during operation. These valves, which are normally closed during plant I

operation, serve as isolation between the high and

() low pressur'e piping. Protective interlocks prevent opening these valves while the reactor is at operating pressure.

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JUSTIFICATION NO. J-5 4

<O SYSTEM: High_ Pressure Coolant Injection.

COMPONENT: .l(2)-2301-4&5.

FUNCTION: Primary containment isolation.

REASONS:' The above valves are normally open to supply steam to the turbine driven HPCI injection pumps.

Conservatively these valves are left in the open position to insure that-driving steam can be '

supplied to these turbines at all times during

() operation. Also, these valves serve a primary containment isolation function (Group _4).

Quad Cities Station feels that to close these valves f during operation would place the operation of the system in an untenable condition. Further, if either were to fail closed it would render the HPCI system inoperable.

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JUSTIFICATION NO. J -6 D

SYSTEM:- Residual Heat Removal COMPONENT: 1(2)-LOOL-19A.& B

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FUNCTION: RHP System cross-tie line isolation valves.

REASONS: These' valves are normally in their safety pos'ition (open) and are only closed a very small percentage of plant operating time when the system.is in the residu'al heat removal mode. Testing these valves during normal operation places the plant in an

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unsafe mode because a failure of either valve in .te cloaed position renders the low pressure coolant injection (LPCI) function inoperable. The LPCI function of RHR is designed such that three of the four pumps are required to provide makeup flow to

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either recirculation loop in the event.of.a design basis loss of coolant accident. This requires the crosstie.line to be open and, hence, both the 1001-19A and B valves. In accordance with NRC Staff guidelines on excluding the cycling of valves whose failure in a non-conservative position would cause a loss of system function.

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F JUSTIFICATION NO. J-7 SYSTEM: Residual Heat Removal C0 4PONENT: 1(2)-1001-28 A&B 1(2)-1001-291A&B FUNCTION: LPCI injection valves; primary containment isolation (29A&B); pressure isolation. ,

REASONS: Relief is requested from exercising these valves during normal reactor operation. Both sets of valves are included because they are interlocked

{} such that one of the two valves must be closed at all times to provide the pressure isolation func-tion. A failure of any one of these valves in the closed position would render the' entire LPCI function technically inoperable since both injection-loops must be available in the design basis accident to provide coolant to the unbroken recirculation loop and this loop could be either one of the two.

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JUSTIFICATION NO, J-8

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SYSTEM: Pressure Suppression COMPONENT: 1(2)-1601-2 A&B 4

FUNCTION: Reactor building to' torus vacuum breaker isolation valves and tirimary containment isolation.

1 REASONS: Exercising these valves open during normal plant operation compromises primary containment integrity and reduces safety margins by leaving only a single 1

check valve (1601 -~31 A or B) to maintain the primary containment boundary.

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r JUSTIFICATION NO. J-9 O

SYSTEM: Control' Rod Drive COMPONENT: 1(2)-302-21 A&B 1(2)-302-22 FUNCTION: Scram discharge volume vent and drain valves.

REASONS: These valves are normally in the open position to allow water which enters the cram discharge volume from normal CRD leakage to drain into the reactor building equipment drain tank. This assures that a sufficient volume is always available to accept.

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scram discharge water following a scram.

The testing of these valves;during plant operation has the potential of isolating the scram discharge volume and the increasing _ water level in the volume would then cause a reactor scram. Consistent with i NRC Staff guidelines concerning the cycling of i

l valves that could potentially place the plant in an

. unsafe mode of operation, it is feld that these valves should be tested at cold shutdown. This l applies to the vent and drain valves since the air

( supply to the valves and the test circuit is common for all three valves.

COM-06-003 Revision 1 5-52 l

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