ML20097J059
| ML20097J059 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/01/1982 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20097J052 | List: |
| References | |
| NUDOCS 8409210097 | |
| Download: ML20097J059 (271) | |
Text
. . .
INSERVICE INSPECTION AND TESTING PROGRAM QUAD CITIES NUCLEAR POWER STATION
, UNITS 1 AND 2 COMMONWEALTH EDISON COMPANY JULY 18, 1979 O
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TABLE OF CONTENTS Ji PAGE
1.0 INTRODUCTION
1.1 General Information......................... 1-1 1.2 System Classifications...................... 1-2 2.0 INSERVICE INSPECTION PROGRAM 2.1 Description of ISI Program.................. 2-1 2.2 Program Tables.............................. 2-8 A. Quad Cities Unit-1 B. Quad Cities Unit-2 2.3 Relief Requests............................. 2-9 30 INSERVICE TESTING PROGRAM FOR PUHFS Description of IST Program for Pumps....... 3-1 3.1 32 Program Tables.............................. 3-2 A. Quad Cities Unit-1 .
B. Quad Cities Unit-2 3.3 Relief Requests............................. 3-3
'4. 0 INSERVICE TESTING PROGRAM FOR VALVES Description of IST Program for Valves....... 4-1 4.1 4-10 4.2 Program Tables............................
A. Quad Cities Unit-1 B. Quad Cities Unit-2 4-11 43 Relief Requests............................
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1.0 INTRODUCTION
1.1 GENERAL INFORMATION The Inservice Inspection (ISI) and Inservice Testing (IST)
Programs for Qua.d 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, 1974 Edition including the Addenda through Summer 1975. Where these rules are determined to be impractical, specific relief is requested in writing.
The Inservice Inspection Programs for Class 1, 2 and 3 Components are applicable for the forty month period beginning October 18, 1979 and November 10, 1979 for Quad Cities Units 1 and 2, respectively. The Inservice Testing Programs for Class 1, 2-and 3 Pumps and Valves are appli-cable for the twenty month period beginning on the same above mentioned dates. It should be noted that the pro-posed rule change to 10CFR50.55a dated January 18, 1979, if adopted, would extend the applicable period for the Inservice Testing Programs from 20 months to 40 months.
The upcoming 40 month period is the third and final period of the first inspection interval for both Quad Cities Units 1 and 2.
'q Revision 1 1 -1 06-01-82'
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1.2 SYSTEM CLASSIFICATION
- 1. .
r 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 vessels. Piping, pumps, and valves were built primarily jn 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 inservice in.pection s requirements.
>* Components within the primary coolant pressure boundary, as defined in 10CFR50.2(v), are designated as ISI-Class I while c,cher 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 5 (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&) Ds) docu-menting the system classifications were developed to aid in the review and implementation of the subject programs. A legend explaining the cplor-coding scheme is included on the first page of the P& ids.
Revision 1 1-2 06-01-82
'r'$ '2.0 INSERVICE INSPECTION PROGRAM
- 2. I - 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, 1974 Edition through the Sununer 1975 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. 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 Num-l[ '
bers. The following information is included in the tables:
A . . Code Category lists the Section XI examination categories as defined in Table IWB-2500 for Class-1 components and Table IWC-2520 for Class 2 components. Since there are no such categories for pressure testing requirements or for Class 3 component examinations, the applicable paragraphs of the Code are referenced. Only those categories applicable to Boiling Water Reactors are includt .
1 Revision 1 2-1 06-01-82 w_________________________-_____-______._____
n rm.. Item Number and Description lists the Item Number and
,' B.
its description as 12.sted in Table IWB-2600 and IWC-2600. All applicable item numbers are listed for each Code Category.
C. Exam Method lists the examination method or methods that .will be performed for each component. Where no relief'has been requested, this reflects the Section !
XI requirements. Where relief has been requested, the exam methods that will be performed in lieu of the required Section XI methods are listed. The abbreviations used are as follows:
VOL.- Volumetric SUR - Surface V-A - Visual examination per IWA-2210 for components such as bolcing and vessel internal parts.
examination for evidence of leakage V-B - Visual conducted in conjunction with pressure testing requirements of IWB, IWC and IWD-5000.
V-C - Visual exa$iination of component supports to determine general condition as related to operability.
D. System _ lists the applicable Class 1, 2 or 3 system as indicated below. If NONE is listed in this column, i
there are no components applicable to that Item Num-ber.
Revision 1 '
2-2 06-01-82 i
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SYSTEM NAME 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 1 Reactor Core Isolation Cooling (RCIC) 1 Core Spray 1&2 High Pressure Coolant Injection (HPCI) 1&2 Main Steam 1 Feedwater 1&2
( ,
Diesel Generator Cooling Water 3 E. Line ' or Component Numbers lists all line numbers or component numbers applicable to each Item Number.
The first digit of the number indicates the appropriate unit number. The letter designation at the end of each line number indicates the piping material (A - stainless steel', B, C, DX, L and LX -
carbon steel).
F. P&ID and Coordinates references the applicable color-coded P&ID and Coordinates for the line or component.
G. No. of Items indicates the total. number of components s_. (i.e. welds, su'ppo r ts , valves, ete.) that apply to ,
23 Revision 1 06-01-82
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- - the particular Item Number. Where this number it appears in parentheses, it refers to the number of components exempted by the referenced relief request. For example, if an entry reads - 16 (1) -
then sixteen of seventeen total components will be examined per the Code and one component is exempted from a Code required examination by the referenced relief request.
Since this number indicates the total number of components for a particular Item Number, the number to be inspected each interval is some percentage of this total, based on the requirements stated in Section XI for each Category.
[
H. Relief Request 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 or surface examination by the applicable Code paragraph.
EX-1 - IWB-1220(b)(3): lines 1-inch nominal pipe size (n.p.s.) and less EX IVB-1220 (b) (1 ) : ILquid carrying lines 2-inch n.p.s. and less (see 2.1.3)
EX-3 -
IWB-1200 (b) (1 ) : steam carrying lines 3-inch n.p.s. and less (see 2.1.3)
. 2-4 Revision 1 06-01-82
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,1 EX IWC-1220 (a) : design pressure and temperature less than 200* F and 275 psig EX Deleted EX IWC-1220(d): 4-inch n.p.s. and less EX IWC-5222 (c) : open ended piping - hydro exempt EX IWD-5223(c): open ended piping - hydro exempt It should be noted that Section 2.3 - contains some generic relief requests that are not specifically referenced in the tables but apply to the ISI Program in general.
I. Remarks - lists general clarification remarks.
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to paragraph IWB-1220 (b) (1 ) , the max 2. mum size 2.1.3 Pursuant line break that can be made up by the reactor coolant makeup system has been calculated to be 2.08 inches inside diameter for liquid carrying lines and 4.16 inches for 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 3.0 inches n.p.s. were exempted.
2.1.4 Quad Cities Station will be implementing Cahss 2 and 3 inspection requirements for the first time with the acceptance of this program. For the remainder of the l- .
2-5 Revision 1 06-01-82
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A current ten year. interval the percentage of the required examinations completed will be only that which would have been' scheduled had Class 2 and 3 requirements been imple-mented at the beginning of the interval and the required examinations divided evenly among each of the three periods.
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2-6 Revision 1 06-01-82
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SECTION 2.2 TABLES FOR INSERVICE INSPECTION PROGRAM A. QUAD CITIES UNIT-1
.B. QUAD CITIES UNIT-2 ,
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. .l INSERVICE INSPECTION PROGRAM UNIT - 1 0 Commonwealth Edison ISI- CLASS 1. 2 & 3 COMPONENTS QUAD CITIES NUCLEAR POWER STATION 1 CODE CATEGORY CLASS REVISION - DATE PAGE B-C-1 PRESSURE RETAINING BOLTING, 2 INCllES DIAMETER AND LARGER 1 j .- 1-82 Page 7 of 40 LINE OR NUMBER ITE"i EXAM P & 10 AND REllEF 1 10N METHOD
" REQUESTS NUMBER NUM8ER COORDINATES IT MS RX VESSEL:
51.7 UU0SURE STUDS IN PLACE VOL RD CTOR VESSEL 1-201 NONE 92 Bl.8 CLOSURE STUDS AND NUTS, VOL&SUR 184 WilEN REMOVED B1.9 LIGAMENTS BETWEEN 111READED VOL 92 STUD 110LES Bl.10 CIDSURE WASilERS, BUSllINGS V-A 184 PIPING PRESSURE BOUNDARY:
B4.2 PRESSURE-RETAINING BOLTS, B4.3 STUDS AND BOLTING NA NONE NA NA 0 B4.4 .
PUMPS:
B5.1 PRESSURE-RETAINING BOLTS VOL RECIRCULATION 1A&B-202 35-2 B-6,3 32 AND STUDS, IN PIACE l
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1- 6-l-82 Page 13 of 40 LINE OR NUMBER EXAM P & ID AND RELIEF E
ITEM DESCRIPit0N "
l , METHOD NUMBER COORDINATES lifMS REQUESTS B4.5 CIRCUMFERENTIAL AND LONGITU- VOL RECIRCULATION 1-0201A-28"A 35-2 B-6 10 DINAL PIPE WELDS 1-0202A-28"A E-5 8
" 1-0201B-28"A B-3 10
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INSERVICE INSPECTION PROGRAM ,
UNIT - 1 O CommonwealthQUAD CITIES NUCLEAR POWER STATION Edison ISI- CLASS 1. 2 & 3 COMPONENTS CODE CATEGORY CLASS REVISION - DATE PAGE B-J PRESSURE RETAINING WELDS IN PIPING (Cont)- 1 j g_j,g7 Page 16 of 40 l LINE OR NUMBER ITEa EXAM P 610 AND REllEF
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l l B4.9 INTEGRALLY WELDED SUPPORTS VOL RECIRCULATION 1-0201A-28"A 35-2 B-6 4 IN PIPING 1-0202A-28"A E-5 1 1 -02018-28"A B-3 4 VOL 1-0202B-28"A E-4 1 (SUR) 1-0201A-22"A D-3 (1) CR-8 (SUR) 1-0201B-22"A D-3 (1) CR-8 VOL 1-0201-22"A B-4 1 CRD RETURN 1-0308-3&4"A 41 D-3 2 RHRS 1-1011-4"A/B 39 A-5 2 1 -1012A-16"A B-4 4
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- 1. 2 & 3 COMPONENTS O Commonwealth Edison ISI- CLASS QUAD CITIES NUCLEAR POWER STATION CLASS REVISICN - DATE PAGE CODE CATEGORY ~ ~
B-K-1 SUPPORT MEMBERS FOR PIPING, VALVES, AND PUMPS (Cont) 1 Pane 18 of 40 TE ITEM DESCRIPil0N "" COORD NATES RE UESTS
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INSERVICE INSPECTION PROGRAM UNIT - 1
- 1. 2 & 3 COMPONENTS O Commonwealth Edison ISI- CLASS QUAD CITIES NUCLEAR POWER STATION REVISION - DATE PAGE CLASS CODE CATEGORY B-K-2 SUPPORT COMPONENTS FOR PIPING, VALVES AND PUMPE I 1- 6-1-82 Page 19 of 40 LINE OR NUMMR RIllEF EXAM P & ID AND E
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CODE CATEGORY 2 1- 6-1-82 Page 30 of 40 ;
C-D PRESSURE RETAINING BOLTING EXCEEDING 1 INCll IN DIAMETER LINE OR NUMBER REllEF PfrID AND ITEM 1 ON EXAM " C00R0lNATES gyfyg REQUESTS EMARKS NUMBER METHOD NUMBER 2A-1003 79 A-2 (2) CR-3 BOLTING FOR 2 NONE RilRS l C1.4 PRESSURE RETAINING BOLTING FLANCES l IN PRESSURE VESSELS
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ITEM DESCRIPTION REMARKS
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INSERVICE INSPECTION PROGRAM >
1, 2 & 3 COMPONENTS UNIT - 2 ISI- CL ASS O CommonwealthOUAD CITIES NU(LEAR POWER STATION Edison CLASS REVISION - DATE PAGE l
CODE CATEGDRY C-G PRESSURE RETAINING WELDS IN PIPING, PUMPS, AND VALVES WHICil 2 1- 6-1-82 Page 36 of 40 CIRCULATE OTilER TilAN RX COOLANT
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SECTION 2.3 RELIEF REQUESTS FOR I!! SERVICE INSPECTION PROGRN1 l
9 2, g ' Revis ion 1
's 06-01-82
I RELIEF REQUEST NO. CR-1 IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE ,
I.
REQUIREMENTS The reactor . vessel is designed with one circumferential and six longitudinal welds in the core beltline region as shown on Figure 1.
The ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition through the Summer 1975 Addenda requires a volumetric examination of ten percent of the length of each longitudinal weld and five percent of the length of each circumferential weld
' each ten year interval (Code Category B-A) .
Relief is requested from the above mentioned Code requirements on the basis of inaccessibility.
II. BASIS FOR RELIEF Accessibility for examination of these welds was not ~ provided for in the original plant design which occurred prior to the issuance of Section XI inservice inspeccion requirements.
As indicated on Figure 1, examination from the reactor vessel outer surface is precluded due to the close proximity to the bio-U logical shield wall and obstruction by the vessel insulation.
29 Revision 1 06-01-82
' [~' . The mirror type insulation consists of interlocking panels which were not designed - to be easily removed at the weld locations. .
Furthermore, the annular dimensions between the shield wall and the insulation is not sufficient to allow direct access for per-sonnel. Access through the biological shield wall is only pro-vided at reactor vessel nozzle locations, however, there are no nozzle penetrations in the beltline region.
Examination of the beltline region welds from inside the vessel is impeded by vessel internal design features. The core shroud, jet pumps, and various brackets welded to the vessel wall are not designed to be removed.
III. ALTERNATE PROVISIONS f
Currently, it is not feasible to perform the required volumetric Commonwealth Edison will, however, examinations on these welds.
keep abreast of improvements in state-of-the-art NDE techniques that could provide a viable means of examination. P i
, i i %d 2 - 10 Revision 1 06-01-82 r
b T RELIEF REQUEST NO. CR-2 y
I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE
' REQUIREMENTS _
The-reactor vessel contains thirteen longitudinal welds and five
+
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, 1974 Edition through the Summer 1975 Addenda requires a volumetric examination of ten percent of the length of each longitudinal
- h shell weld and five percent of each circumferential shell and head weld each inspection interval (Code Category B-B) .
-As shown on Figure 1, all of the reactor vessel closure head
. welds are fully accessible for examination as are the vessel and l
head-to-flange welds and the three longitudinal welds in the upper shell course (No. 4) .
l I -The remaining shell and bottom head welds, however, are inacces-
'~
~s ible for examination.
i s.
. s 2-11 Revision 1 06-01-82 i
I II. BASIS FOR RELIEF As discussed ~ in Relief Request CR-1, accessibility for examina-tion of these welds was not considered in the plant design.
There is no access through the biological shield wall or between the wall and the vessel to permit examination of the shell welds from the vessel outer surface. Similarly, the bottom head welds cannot be examined because of the limited physical access, the inab'ility to remove vessel insulation . panels, and the interfer-ence from the forest of control rod drives and instrumentation penetrations.
III ALTERNATE PROVISIONS i
Currently, it is not feasible to perform the required volumetric i
exaninations on these welds. Commonwealth Edison will, however, keep abreast of improvements in state-of-the-art NDE techniques that could provide a viable means of exami. nation.
L l-f A._
L 2-12 -
Revision 1
' 06-01-82 l
RELIEF REQUEST NO. CR-3
- 1. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS This relief request addresses the Section XI Examination Cate-gories B-G-1 and B-G-2 for Class-1 bolting, and C-D for Class-2 bolting. Category B-G-1 in the 1974 Edition of the ASME Code, Summer '1975 Addenda covers bolting two-inches and greater in diameter and B-G-2 covers bolting less than two-inches in diam-eter. Category C-D covers bolting that exceeds one-inch in diam-eter.
./, However, in later editions- of the Code, Class-1 bolting exactly 4
two-inches in diameter'is~ shifted from Category B-G-1 to B-G-2 by revision of'~the category definition.
Similarly, Class-2 bolting between one and two-inch diameter is eliminated from Category C-D of the later editions of the Code.
Quad Cities Station concurs with the Category definitions of
^
later Editions of Section XI for Examination Categories B-G-1, B-G-2, and C-D and accordingly request permission to adopt these definitions.
(t
\ 2-13 Revision 1 06-01-82
- ~.
II. BASIS FOR- RELIEF I
This request for relief involves substitution of requirements from later Editions of the ASME Code. Adopting the more prac-tical requirements from these later editions will provide conti-nuity between the inspection program for this period and the .
program for subsequent intervals while reducing overall radiation exposure to inspection personnel. Plant safety margins will be unaffected by this change since modifications in the Code requirements are technically justified.
III. ALTERNATE PROVISIONS examinations will be performed as specified in the Visual appropriate Code Category for the bolting shif ted from Category B-G-1 to B-G-2. No alternate or augmented examinations are
- required for the bolting affected in Category C-D.
..4 .
l C Revision 1 2-14
'" 06-01-82 i
f' - RELIEF REQUEST NO. CR-4 u
OF COMPONENTS AND IMPRACTICAL CODE I. IDENTIFICATION REQUIREMENTS E
1The reactor vessel and associated closure head are stainless steel-cladded on the interior surfaces. Six patches, each having a 36 square inch area, are selected for examination in accessible locations of the reactor-vessel shell and closure head.
The' ASME. Boiler- and Pressure Vessel . Code ,Section XI, 1974 Edition through the . Summer 1975 Addenda requires that the clad reactor vessel be visually examined each patches in the interval. In addition, for ' closure head patches a visual and a surface or volumetric examination is required each interval.
'For the reasons described below, these examination requirements are unrealistic since they offer no meaningful check of reactor y
vessel integrity.
II. BASIS MR RELIEF Analysis has shown that flaws which initiate in the reactor ves-
'sel cladding at locations other than nozzles do not propagate through the clad-base metal interf ace. Therefore, their exist-ence poses no threat to reactor vessel integrity. The nozzle-
.g radii
. areas are covered by the requirement to examine the inner 2-15 Revision 1 1
06-01-82
I' volumetrically to detect the presence of flaws which may have propagated into base metal. Accordingly, the ASME has completely eliminated the B-I-1 and B-I-2 Examination Categories from later Editions of Section XI.
Performing these examinations only constitutes a needless expo-sure of personnel to radiation with no compensatory increase in safety. Quad Cities Station, therefore, will not perform the above mentioned examinations for the remainder of the present inspection interval. The examinations will not be required for subsequent intervals since the requirements have been deleted from the Code.
g( III. ALTERNATE PROVISIONS No alternate or augmented examinations are necessary in this Case.
L.
2-16 Revision 1 06-01-82 :
NN RELIEF REQUEST NO. CR-5 COMPONENTS AND IMPRACTICAL CODE I. IDENTIFICATION OF REQUIREMENTS
..Two Class-1 piping welds.are physically inaccessible for examina-
-tion.-
The -~ weld is in the Control Rod Drive System is on line number 0308-4, .
the other weld is in the RHR System on line number 1011-4". The weld in the CRD System cannot be examined because of interference from a . structural support as shown on Figure 4.- The weld in the RHR System is located just above the the floor separating the point at - which the line penetrates reactor cavity and the .drywell. The inaccessibility is due to as shown the . presence of a- water barrier and sleeve arrangement
[
on Figure 4.
Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition including the Summer 1975 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 will b'e inspectable at any time during the plant life. Relief is, therefore, requested from per-forming the volums cic examination requirements of Section XI.
g 2-17 Revision 1 06-01-82.
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/I II. BASIS FOR RELIEF The implications of this exemption are minimal due to the fact that safety margins inherent. in the desi(,n of the subject welds are typical of those in all other welds in the Class-1 systems.
Exempting these two welds from the total inspection sampling l program will"have negligible. statistical significance.
- III.- -ALTERNATE PROVISIONS No alternate or augmented examinations are feasible or necessary
' in this case. The examinations required by IWB-5000 will, how-ever, be conducted in accordance with the Code.
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2-18 Revision 1 06-01-82
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RELIEF REQUEST NO. CR-6 l I
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 .1011 -4", 1012A&B-16", 1025-20" Rx WATER CLEANUr - 1202-6" .
CORE SPRAY - 1403-10", 1404-10" HPCI - 2305-10" MAIN STEAM 3001 A , B , C , D-20" FEEDWATER 3204A&B-18"Section XI, 1974 The ASME Boiler and Pressure Vessel Code, Edition through the Stimmer 1975 Addenda requires that these Class-1 welds be volumetrically examined (Code Category B-J).
Since this requirement is impractical due to plant design, relief is. requested from the above stated examination requirements.
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2-19 Revision 1 06-01-82
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II. BASIS MR RELIEF l
l As stated ~ in 10CFR50.55a (g) (1 ) for plance 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 S t a t i o n w'a s designed, accessibility for their
-examination was not a prime consideration. Figure-2 clearly illustrates the design ' constraints which make it extremely
. impractical to. examine the subject welds by volumetric or surface techniques. Commonwealth Edison feels that this constitutes 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 welds in these systems (15 out of 445), the statistical signifi-cance to the inspection sampling program due to exempting these welds is expected to be negligible.
(' 2-20 Revision 1 06-01-82
l i ;-
III. ALTERNATE PROVISIONS
, 'l At the present time: no alternate examinations are feasible because of' the inaccessibility. 11owever , the examinations required by IWB-5000 will ' be conducted in accordance with the Code. --
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- i 2-21 Revision 1 06-01-82
f4 RELIEF REQUEST NO. CR-7
- 1. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE 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 incasing it as illustrated on Figure 3. As listed in the program, there is one such weld that is six inches in diameter and three welds that are greater than 6 inches.
Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addenda requires that branch pipe connection welds exceeding six inches diameter be examined vol-umetr'ically and those six inch diameter and smaller be surface examined. Twenty-five percent of these welds are required to be examined each inspection interval (Code Category B-J) .
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 or meaningful volumetric examination. Additional g assurance of the continued integrity of joints fabricated in this 2-22 Revis ion 1 06-01-82
['- fashion is afforded by the fact that the reinforcement saddle strengthens the joint and reduces the stresses on the internal
- . weld.
I 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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2-23 Revision 1 06-01-82
RELIEF REQUEST NO. CR-8
- 1. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS In the Class-1 system there are t2n integrally welded supports whose support lugs are welded to cast stainless steel
- components. Specifically, six are welded to the recirculation pump casingr., .two are welded to the bodies of' recirculation valves 0202-5A and 5B, and two are welded to a stainless steel crosses in the recirculation ring header piping.
Section XI of the ASME Boiler and Pressure Vessel Code, 1974
/. Edition through the ' Summer 1975 Addenda requires a volumetric V
inspection of each integrally welded support attachment each ten
/, year interval (Code Category B-K-1).
This examination requirement is impractical for these support attachments because of the material structure and weld geometry.
II. BASIS FOR RELIEF
~The high ultrasonic beam attenuation of the cast stainless steel l
I. ~
base material and the weld geometry inhibit meaningful examina-tion of the ten subject support attachments by either ultrasonic or radiographic methods. The substitution of a surface examina-A > tion, however, would be sufficient todetermh.netheintegrityof 2 24 Revis ion 1 06-01-82
i h
lr:. these attachment welds and the surrounding base metal since flaws which would be expected to occur in these areas would originate from the outer surface.
III. ALTERNATE PROVISIONS A surface examination will be substituted for the required vol-umetric examination for these components.
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2-25 Revision 1 06-01-82
y, RELIEF REQUEST NO. CR-9 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.
The ASME Boiler and Pressure Vessel Code, Section XI, 1974 Edition through the Sununer 1975 Addenda requires that one of 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, Quad Cities Station requests relief from the Section XI examination requirement to visually
' examine the ' recirculation pump internal surfaces on the basis of I
impracticality.
II. BASIS FOR RELIEF p
b The basis for this relief request is predicated on the following
?
two points:
q, 2-26 Revision 1 06-01-82
- y. .
- 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 disas-sembly, 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 add-itional 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.
Commonwealth Edison feels that adequate safety margins are inher-ent in the basic pump design and that the health and safety of k the public will not be adversely effected by performing the 2-27 Revision 1 06-01-82
visual examination of the pump internal pressure boundary sur-faces only when the pumps are required to be disassembled for maintenance.
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 type is disassembled for maintenance, examination of the pump internals and internal pressure boundary surfaces will be per-formed, to the extent practical.
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2-28 Revision 1 06-01-82
_ _ _ _ _ _ _ - _ _ _ _ _ _ . )
RELIEF REQUEST NO. CR-10
- 1. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE .
REQUIREMENTS 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 manuf acturer but are all manufactured from either cast stain-le.ss steel or carbon steel. None of the valve body casings are welded.
Section XI of the ASME Code, 1974 Edition through the Summer 1975 Addenda requires a visual examination of the internal pressure
( boundary surfaces of one valve in each group of valves of the same constructional design, manufacturing method and minufacturer 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 to implement.
II. BASIS FOR RELIEF The requirement to disassemble primary system valves for the sole
%. purpose of performing a visual examination of the internal pres 2-29 Revision 1 06-01-82
.~ . - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ .__ __ _._ __
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sure boundary surfaces has only a very small pocential 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 condi-
~
tions as high dose rates (10 R/hr) and poor as-cast surface con-dition, realistically, provides little additional information as to the valve casing integrity.
i For approximately 20 percent of these valves, the reactor vessel core must be completely unloaded and the vessel drained to permit disassembly for examination.
l
\ The performance of both carbon and stainless cast valve bodies has been excellent in all BWR applications. Based on this exper-ience 'and both industry and regulatory acceptance of these alloys, continued excellent service performance is anticipated.
VA more . practical approach that would essentially provide a::
equivalent sampling program and significantly reduce radiation exposure to plant personnel is to examine the internal pressure boundary of only those valves that require disassembly for main-tenance purposes. This would still' provide a reasonable sampling of primary system valves and give adequate assurance that the integrity of these components is being maintained.
+
. %. c 2-30 Revision 1 06-01-82
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f III. ALTERNATE PROVISIONS An examination of the internal pressure boundary surfaces will be Performed, to the extent practical, each time a valve is disas-
,- sembled for. maintenance purposes.
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2-31 Revision 1 06-01-82 b
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- O- + RELIEF REQUEST NO. CR-11 I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS There are two 18" diameter nozzles in the Class-2 portion of each of the two RHR System heat exchangers that are fabricated with ireinforcement saddles. These saddles are fillet welded over the l N' -
actual pressure-retaining nozzle-to-shell weld. The configura-tion is shown on Figure 5.
.A Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition .through the Summer 1975 A'ddenda requires a volumetric examination of two of these four nozzle-to-shell welds in the service lifetime of the unit. This requirement is impractical due to inaccessibility.
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! II. BASIS FOR RELIEF 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 j
stresses at the internal weld. Integrity of these joints will be g monitored by periodic system pressure and hydrostatic tests.
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2 - 32 Revision 1 06-01-82
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III. ALTERNATE PROVISIONS n >
'A visual examination for evidence of leakage will be conducted
" in accordance with the Subsection IWC-5000 requirements.
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2-33 Revision 1
' 06-01-82 t-i
j' RELIEF REQUEST NO. CR-12 t
& I. IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS L
l The pressure retaining components within each system boundary are t.. These test requirements are subject _ to system pressure tests.
not only an important part of inservice inspection but also clarity in their application. The hydrostatic test demand requirements in the 1974 Edition of Section XI are not as defini-tive as in later editions and addenda of the Code and for this reason, misinterpretation or misapplication could occur.
( The later editions <of the Code have revised various requirements regarding these system pressure tests which Quad Cities Station 7
feels are more practical to implement. Therefore, the following pressure testing requirements will be adopted:
- 1) The requirements of IWC&D 5200(a) in the 1974 Edition of.
the ASME Code, Section XI will be replaced with the r
x following: The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with design temperature of 200*F or less, and at least.
1.25 times the system pressure Psv for systems with Design Temperature above 200 *F. The system pressure Pav 4
is defined as the lowest pressure setting among the provided s,
i.
number of safety or relief valves for overpres 2-34 Revision 1 06-01-82
-4
- sure protection within the boundary of. the system to be L
tested.
- 2) The -following requirements regarding the holding ' time
.after pressurization (before visual examination) will be adopted for clarity.
a) System . Leakage Tests - no holding time required after attaining test pressure and temperature con-ditions.
b) System Functional Tests -
10 minutes after attaining the system operating pressure.
f
_Y c)_ System Inservice Tests - no holding time required ,
provided the system has been in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
g d) System Hydrostatic Tests - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> af ter attaining the test pressure and temperature conditions for
+
insulated systems, and 10 minutes for noninsulated
' systems or components, e) System Pneumatic Tests - 10 minutes after attatning the' test pressure.
2-35 Revision 1 06-01-82 g
II. BASIS FOR RELIEF s.
ic This request for relief involves the substitution of requirements from later Editions of the ASME Code. Substituting these more definitive and practical requirements will not only provide con-
- tinuity between the inspection program for this period and the program for subsequent intervals, but will also help reduce radi-ation ' exposure to inspection personnel. Plant safety margins will be unaffected by these substitutions since modification in the Code requirements are technically justified.
III. ALTERNATE PROVISIONS No alternate or augmented examinations are necessary in this
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2 - 36 Revision 1 06-01-82
1
,s-RELIEF REQUEST NO. CR-13 a ,
I; IDENTIFICATION OF COMPONENTS AND IMPRACTICAL CODE REQUIREMENTS l
)
Quad Cities Station currently utilizes a calibration block which . lacks documentation consistent with the requirements of current editions of the Code. The. documentation requirements .
~
existing- at. the time of their fabrication did not require tr'aceability to the material's chemical or physical certi-fications. As a result, the only documentation available for the
- i. existing block it verification of the appropriate P-number grouping.
!{
.y The Section XI requirements of the 1974 Edition of the ASME Code l ., including the 1975 Addenda specifies that the block will be fabricated as provided by Article I-3000 paragraph I-3121 re-
> quirements.
Relief is requested from this documentation requirement to
+ allow the continued use of the existing calibration block.
II. BASIS FOR RELIEF ,
y 1
' Previous inservice Lnspections have been performed utilizing the above mentioned block and its use would provide continuit/ in the ISI Program. It would bc impractical to fabricate a nca
- t. .
2-37 Revision 1 06-01-82 y
i .. . . . -
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calibration block in _ order to satisfy the documentation f-(
the current Code. Existing records which requirements of indicate the appropriate material P-grouping . prcvide adequate assurance that the block will establish the proper ultrasonic 1
calibration and sensitivity. Additionally, since both reactors vessels are 100% clad on the I.D. surface, there is no way to meet the requirement of verifying the acoustic properties of .
the block against the clad component.
III. ALTERNATE PROVISIONS The present reactor vessel calibration block will be demonstrated to have acoustic attenuation and velocity properties which fall 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 bloqk.
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2-38 Revision 1 06-01-82 1
. y_'
RELIEF REQUEST . NO. CR-14 c IDENTIFICATION OF COMPONENTS AND -IMPRACTICAL CODL I.
REQUIREMENT-1
. The rules of = Article 5 of Section V recommend that III indications which produce a ~ response greater than 207. of the reference level be investigated to the extent that the operator can evaluate the shape, identity and location of all such reflectors in terms of the ' acceptance-rejection standards of Section XI.
Section XI, 1974 The ASME' Boiler and Pressure Vessel Code,
- Edition through Summer 1975 Addenda states that the provisions of Article 5 of Section V shall apply where Appendix I is not appli-
- cable. However, in later editions of Section XI, the rules of
- Article 5, Section .V were amended such that only reflectors pro-ducing a response greater than 507. of the reference level are to be recorded, and.that all reflectors producing a response greater than 100% of the reference level shall be investigated to the extent that 'the operator can determine the shape, identity and location of all such reflectors in terms of the acceptance-rejec-tion standards of Section XI.
Commonwealth Edison concurs with the requirements of the later Code and addenda and therefore, a request for relief from the earlier requirement is sought.
\ -
2- 39 Revision 1 06-01-82 o
' II. BASIS FOR RELIEF As a result of the " noise" level in the typical UT response and the weld geometries present, no meaningful information is
- obtained from indications producing responses less than 50% of the reference level. Therefore, adopting the more current and practical requirements of Section XI is justified. In fact , the requirement to record these non-relevant indications results in i excessive examination times and personnel radiation exposures.
It is felt that the levels for recording and evaluating indica-tions specified in the later Codes are adequate and sufficiently reliable in detecting flaws.
( III. ALTERNATE PROVISIONS For_ examinations conducted to the requirements of Article 5 of Section V, the recording level shall be 50% of the reference level and all indications exceeding 100% of the reference level shall be investigated to the extent that the operator can deter-mine the shape, identity and location of all such reflectors in terms of the acceptance-rejection standards of Section XI.
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2-40 Revision 1 06-01-82
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2-41 REVIS!0ll1 06-01-82 i
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+
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18" N0ZZLES (TUBE SIDE-
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4 0 O 4
- A SECTION 3.2 TABLES FOR INSERVICE PUMP TESTING PROGRAM A. QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 y
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Revision 1 9/24/80
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- - 3 INSERVICE TESTING PROGRAM UNIT - 1 iSI- CL A SS 1. 2. & 3 PUMPS
@ Commonweahh Edison QUAD CITIES NUCLEAR POWER STATION REVISION - DATE 1 09/24/80 PAGE 1 of 1 TEST PARAMETERS E P b ID AND TEST PUMP NUM8tR PUMP NAME INTERVAL
$ COGADINATES SPEED INLET DIFF FLOW RATE vigAAil0N BEARING TEMP PRES PRES CORE SPRAY 2 36 E-9 NO YES YEE YES PR-1 PR-i PR-2 1A-1401 2 36 E-6 NO YES YES YES PR-1 PR-1 PR-2 13-1401 CORE SPRAY 2 37 B-4 NO YES YES YES PR-1 PR-1 PR-2 1A-1002 RESIDUAL HEAT REMOVAL 2 37 :'-4 NO YES YES YES PR-1 PR-1 PR-2 1B-1002 RESIDUAL HEAT REMOVAL 2 37 B-8 NO YES YES YES PR-1 PR-1 PR-2 1C-1002 RESIDUAL HEAT REMOVAL 2 37 E-8 NO YES YES YES PR-1 PR-1 PR-2 ID-1002 RESIDUAL HEAT REMOVAL 3 39 F-4 NO YES YES YES PR-1 PR-1 PR-2 1-1001-65A RHR SERVICE WATER 3 39 F-4 NO YES YES YES PR-1 PR-1 PR-2 1-1001-65B RilR SERVICE WATER 3 '39 F-7 NO YES YES YES PR-1 PR-1 PR-2 1-1001-65C RilR SERVICE WATER 3 39 F-7 NO YES YES YES PR-1 PR-1 PR-2 1-1001-65D RHR SERVICE WATER 2 40 D-7 NO PR-4 PR-4 YES PR-1 PR-1 PR-2 1A-1102 STANDBY LIQUID CONTROL 2 40 E-7 NO PR-4 PR-4 YES PR-1 PR-1 PR-2 1B-1102 STANDBY LIQUID CONTROL 1-2302 HIGli PRES COOLANT INJ 2 46 A-4 YES YES YES YES PR-1 PR-1 PR-2 1-3903 3 22 A-10 NO YES YES YES PR-1 PR-1 PR-2 D/G COOLING WATER 1/2-3903 D/G COOLING WATER 3 22 A-10 NO YES YES YES PR-1 PR-1 PR-2 1-5203 D/G FUEL DIL TRANSFER NC 29 F-3 NO PR-5 PR-5 YES PR-1 PR-1 PR .2 NC 29 F-3 NO PR-5 PR-5 YES PR-1 PR-1 PR-2 1/2-5203 D/G/ FUEL OIL TRANSFER 5
NOTE:
for are designed such that levels can be verified. The Luhrication core spraylevels will he (1401), observed RitR sluring (1002), and each inservice testEvel oil transfer (5203) pumps thatpumps are lubricated by pump flowage and, thus, lubricent the D/G 1svel or pressure measurements are no- re'esant.
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INSERVICE TESTING PRZGRAM UNIT - 2 gj, ISI- CL A S S 1. 2. & 3 PUMPS
@ Commonweahh QUAD CITIES NUCLEAR POWER STATION REviSm . DATE 1 09/24/00 PAGE 1 of 1 I II I A I I TEST E P6ID AND PUMP NUTiiBER PUMP NAME MM
$ COORDINATES SPEED INLET DIFF FLOW RATE VIBRATION BEARING TIMP PRES PRES 2A-1401 CORE SPRAY 2 78 E-9 NO YES YES YES PR-1 PR-1 PR-2 25-1401 CORE SPRAY 2 78 E-6 NO YES YES YES PR-1 PR-1 PR-l' 2A-1002 RESIDUAL HEAT REMOVAL 2 79 B-4 NO YES YES YES PR-1 PR-1 PR-2 25-1002 RESIDUAL HEAT REMOVAL 2 79 E-4 NO YES YES YES PR-1 PR-1 PR-2 2C-1002 RESIDUAL HEAT REMOVAL 2 79 B-8 HO YES YES YES PR-1 PR-1 PR-2 2D-1002 RESIDUAL HEAT REMOVAL 2 79 E-8 NO YES YES YES PR-1 PR-1 PR-2 2-1001-65A RRR SERVICE WATER 3 81 F-4 NO YES YES YES PR-1 PR-1 PR-2 2-1001-65B RHR SERVICE WATER 3 81 F-4 NO YES YES YES PR-1 PR-1 PR-2 2-1001-65C RRR SERVICE WATER 3 81 F-7 NO YES YES YES PR-1 PR-1 PR-2 2-1001-65D RHR SERVICE WATER 3 81 F-7 NO YES YE3 YES PR-1 PR-1 PR-2 2A-1102 2 82 D-7 NO PR-4 PR-4 YES PR-1 PR-1 PR-2 STANDBY LIQUID CONTROL 28-1102 STANDBY LIQUID CONTROL 2 82 E-7 NO PR-4 PR-4 YES PR-1 PR-1 PR-2 2-2302 HIGH PRES COOLANT INJ 2 87 A-4 YES YES YES YES PR-1 PR-1 PR-2 2-3903 - D/G COOLING WATER 3 69 A-10 NO YES YES YES PR-1 PR-1 PR-2 4
2-5203. D/G FUEL OIL TRANSFER NC 29 F-3 NO PR-5 PR-5 YES PR-1 PR-1 PR-2 i
4 NOTE:
Lubrication levels will be observed during each inservice test for are designed such that levels can be verified. The core spray (1401), RilR (1002), and the D/G fuel oil transfer (5203) pumps thatpumps are lubricated by pump flowage and, thus, lubricant level or pressure measurements are not relevant. -
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SECTI0f1 3.3 RELIEF REQUESTS FOR INSERVICE PUMP TESTlilG PROGRAM s
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Revision'l 9/24/80
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3.0 INSERVICE TESTING PROGRAM FOR PUMPS 3.1 GENERAL INFORMATION The Inservice Testing Program for ISI Class 1, 2 and 3 Pumps meets the requirements of Subsection IWP of Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addenda. Where these requirements are determined to be impractical, specific requests 'for relief have been written.
The tables in Section 3.2 list all Class 1, 2 and 3 pumps y to be tested along with the parameters to be measured-for each pump unless reference is made to a - relief request.
Section 3.3 includes all relief requests referenced in the tables plus any additional relief requests that are generic to 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.
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3-1 Revision 1 9/24/80
RELIEF REQUEST NO. PR-1 PUMP NUMBER: -All pumps in-program.
SECTION X1 REQUIREMENT: Detection of mechanical change per IWP-1700.
BASIS FOR RELIEF: ,
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 pr'ior 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.
Quad Cities' Station proposes an alternate program which is believed to be more comprehensive than that required L
by Section XI. This program consists of performing the required vibration readings in velocity rather than mils l- displacement. .This technique is an industry-accepted l
method which is much more meaningful and sensitive to
~ small changes that are indicative of developing mechanical problems. These velocity measurements detect not only high amplitude vibrations that indicate a major
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RELIEF REQUEST NO. PR-1 (CONTINUED)
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'l mechanical broblem but also the equally harmful low amplitude - high frequency vibrations due to misalign-ment, imbalance, or bearing wear that usually go undetected by simple displacement measurements.
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
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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 n'o t vary more than 3%),
causes problems with pumps having no recirculation / test loop. The temperature of the pumped fluid is also meaningful when attempting to trend any developing problems from year to year. It is easy to see that a program of bearing temperature trends and the evaluation of the results would in some cases be difficult to analyse. Improper interpretation of results could result in unnecessary pump maintenance.' In addition, it
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3-3 Revision 1 9/24/80
E I~ RELIEF REQUEST.NO. PR-1 (CONTINUED) l is impractical to 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-LOO 2 - same as above.
(3) RHR Service Water 1(211001-65A,B,C,D -
Bearings are contained in an oil-filled reservoir. The 4
ambient temperature of the pump space is changeable thereby varying the start temperature of the data. Results would be difficult if not impossible to trend from test to test.
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( (4) High Pressure Coolant Injection -
this pump is driven by a steam turbine which exhausts steam into the pressure suppression chamber. Extended r run times to stabilize bearing temperatures would
, create problems in keeping suppression pool temperatures below the Technical Specification l
l limit of 95'F.
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_ _ _ _ _ _ _ _ _ _ _ _ .._ _ _- _ __.- _ _ . _ . . _ . _ . . . . ~ . _
g- RELIEF REQUEST NO. PR-1 (CONTINUED) i (5) Diesel Generator Cooling Water 1(2)(1/2)-3903 --
Same as RHR Service Water (6) Diesel Generator Fuel Oil Transfer 1(2)(1/2)-5203
- this transfer pump pumps fuel oil from the 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
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program of vibration measurements is a more practical method of testing which exceeds the requirements of the ASME Code.
Pump vibra. tion measurements will be j ALTERNATE TESTING:
taken in vibration velocity (in/sec). The evaluation of the readings will be per the attached table.
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1 ALLOWABLE RANGES OF TEST QUANTITIES ALERT RANGE REQUIRED ACTION RANGE LOW HiGil LOW HIGH.
QUANTITY ACCEPTABLE RANGE VALUES VALUES VALUES . VALUES v When 0 < vr < .15 in/sec 0 to .3 in/sec None .3 in/see to None v > .45 in/sec
.45 in/sec v When .15 in/see < v r 0 to .45 in/sec None .45 in/see to None v> .75 in/sec
< .3 in/sec .75 in/sec v When .3 in/sec < v r 0 to 0.9 in/sec None 0.9 to 1.5 None, y > 1.5 in/sec
< .6 in/sec in/sec v When .6 in/sec < v r 0 to .1.1 in/see None 1.1 to 1.5 None v > 1.5 in/see 4
< 1.0 in/sec in/sec w
I e ____________________
i Where:
v = velocity measured in inches /second , peak.
v r = reference velocity measurement (initial measurement after installation or rework.
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l' RELIEF REQUEST NO. PR-2 e
PUMP NUMBER: All pumps in program.
'SECTION XI REQUIREMENT: Measure pump parameters monthly.
BASIS FOR RELIEF: Relief is requested from the require-ments of Subsection IWP-3400 to measure the basic pump parameters identified in Table IWP-3100-1 on a monthly basis. Changes in these hydraulic and mechanical parameters will not significantly change over the period of one month because the pumps are primarily run only for operability and remain in a standby mode of operation. Quarterly measurement of these parameters is more than adequate in determining pump degradation.
The original intent to require monthly testing was based on the premise that damage can occur to bearings if a pump remains stagnant for long periods of time. This concern can be mitigated by running pumps on a monthly basis to lubricate the main bearings.
A change to the Code of a similar nature, recently passed the Section XI Main Committee and will be published in a forthcoming Addends to Section XI. It is 4
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g RELIEF REQUEST NO. PR-2 (CONTINUED)
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not- felt that this relief request represents a relaxa-tion in safety requirements, only that it allows more practical implementation of Section XI requirements.
ALTERNATE TESTING: All pumps will be exercised on a monthly basis to lubricate the bearings. Pump parameters will be measured quarterly.
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e- RELIEF REQUEST NO. PR-3
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PUMP NUMBER: All pumps in the program.
SECTION XI REQUIREMENT: The requirements of IWP-3230(c), Correc-tive Action.
BASIS FOR RELIEF: Relief is requested from the require-ments of IWP3230(c) regarding corrective action when pump parameters are found to be within the " Required Action Range" of Table IWP-3100-2. Some means should be allowed for conducting an analysis to demonstrate that the condition of a pump does not impair pump operability and that the pump can still perform its intended
' function. Later editions of the Code do address this concern by allowing such an analysis to serve as the corrective action.
ALTERNATE TESTING: When measured pump parameters fall into the " Required Action Range", pump operability and corrective action will be based on the limits specified in the Limiting Conditions for _ Operation of the plant Technical Specifications. A pump may remain operable if it meets all Technical Specification requirements and an analysis indicates that, even though a pump parameter is in the " Required Action Range", the pump can still A fulfill its intended functions.
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r RELIEF REQUEST No. PR-4 4 !
I PUMP NUMBER: 1(2)-1102 (SBLC)
SECTION XI REQUIREMENT: Measure pump inlet pressure BASIS FOR RELIEF: It is impractical to measure standby liquid control pump inlet pressure in accordance with Section XI requirements. During pump testing,the pump suction is from a test tank rather than the main standby liquid control tank. No instrumentation is provided for measuring inlet pressure, and therefore , the only means 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 discharge pressure and the flow rate is adequate to detect changes in the hydraulic characteristics of the pumps.
ALTERNATE TESTING: Pump discharge pressure will be monitored at each inservice test.
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I-3-10 Revision 1
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9/24/80
RELIEF REQUEST NO. PR-5 7
PUMP NUMBER: 1-5203, 2-5203, 1/2-5203 SECTION XI REQUIREMENT: Measure Pump Inlet Pressure BASIS FOR RELIEF: Relief is requested from the requirement of. measuring pump inlet pressure during pump tests.
This pump is uilized in transfering fuel oil from the diesel generator fuel oil storage tank to the diesel fuel oil day tank. The configuration of the piping is such that the pump is located above the storage tank.
The pump is a positive displacement gear type pump not requiring a positive auction head for proper I operation. Since this pump is a positive displacement type, the discharge pressure is. independent of the suction pressure and, therefore, inlet pressure data is not important in evaluating pump performance.
ALTERNATE TESTING:- Pump discharge pressure will be monitored at each inservice test.
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/. 4.0' INSERVICE TESTING PROGRAM FOR VALVES 1
4.1 -GENERAL INFORMATION l l
The Inservice Test'ing 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, 1974
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ECition through the Summer 1975 Addenda. Where these requirements are determined to be impractical, specific requests for relief have been written and included in Section 4.3.
The tables in Section,4.2 list all ISI Class 1, 2 and 3 i .
valves . that have been assigned valve categories; valves exempt per IWV-1300 are not listed. The tables are organized by system in order of the assigned system numb'er . A list of these systems and their respective P&ID numbers is given in Table 4.1-1. The following infor-L mation is included in the tables:
l-l l A. Valve Number lists the valve identification number as shown.on the color-coded P& ids. The first digit of the valve number indicates the appropriate unit.
B. P&ID and Coordinates references the color-coded P&ID on which the valve appears and its coordinates.
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4-1 Revision 1 9/24/80
f-C. Class it, the ISI Classification of the valve. Valves in 'the diesel fuel oil and air start systems as well as some primary containment isolation valves are included in the program, even though they do not have an ISI Classification. These valves are designated as Class NC (Not Classified).
D. Valve Category indicates the category assigned to the valve based on the definitions of IWV-2110.
E. Valve Size lists the nominal pipe size of the valve in inches.
F.. Valve Type lists the valve design as indicated by the following abbreviations.,
GATE GA GLOBE GL CHECK CK SAFETY - SV RELIEF RV ELECTROMATIC RELIEF ERV l-l BUTTERFLY BTF i
l STOP CHECK SCK
! BALL BALL 7,
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/' RUPTURE DIAPHRAM RPD EXCESS FLOW CHECK XFC G. Actuator Type lists the type of valve actuator as j indicated by the following abbreviations.
MOTOR OPERATOR MO AIR OPERA, TOR AO SOLENOID QPERATOR SO PILOT SOLENOID ACTUATOR PS EXPLOSIVE ACTUATOR EXP SELF ACTUATED SA MANUAL M i
H. Valve Position indicates the normal position of the valve during plant. operation. This is specified as open (O), - closed (C), locked open - (LO), and locked closed (LC).
I. Stroke Direction indicates the direction which an active valve must stroke to perform its safety 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).
4-3 Revision 1 9/24/80
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J.. T e's t lists the test or tests that will.be performed for each- valve to ful' fill the requirements of-Subsection IW. Th'e - following tests and
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abbreviations are used:
Seat Leak Test (IFV- 3420 ) AT Valve will be seat leak tested at the
- appropriate functional differential pressure. (See VR-16)
-Full Stroke Exercise Test ( IW-3410 a , b , c ) BT Valve will be full stroke exercised for operability in the direction necessary to
. fulfill its safety function.
Partial Exercise Test (IW-3410b ,1 ) BTP Valve will be part-stroke exercised when
- full stroke exerc'ising .-
is impractical. -
t Check Valve Exercise Test (IW-35 20 ) 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 valvt shall be adequate. demonstration of valve operability.
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. - - _ - - - - - _ . . , _ - . . ~ . - . . , , , - - - , , _ _ _ . , _ . _ , - . . _ , . . _ - . _ . _ , . , _ _ , - - _ - _ - . , _ . , . _ . , ~ . - , . _ . . .
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- Relief Valve Set Point Check (IW-3510) CT-2
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Relief and safety valve set points will be verified in accordance with IW-3510.
Explosive Valve Tests (IW-3 6'10 ) DT Explosive valves will be tested in accordance with IW-3610.
5 Fail-Safe Test (IW-3410e) FST All valves with fail-safe actuators will be tested to verify proper fail-safe operation upon loss.of actuator power.
Position Indication Check (IW-3300 ) PIT All valves with remote position indicators that are inaccessible for direct observation during normal plant operation i will be checked to verify that remote valve indications accurately reflect valve i
operation.
K. Test Mode indicates the frequency at which the above mentioned tests will be performed. The following abbreviations are used:
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t Normal Operation OP p
Tests which are conducted at least once every 3 months during normal plant operation.
Cold Shutdown CS Inservice valve testing at cold shutdown is valve testing which commences within two hours after the plant reaches a cold shut-down condition but in no case later than 48 j hours 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 case of frequent s
cold ' shutdowns , valve testing need not be performed more often than once every three months for Category A, B, and C valves.
In the case of longer planned cold shut-downs, the testing. need not be started t
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- r 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
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these instances, all valve ~ testing must be completed prior to start-up.
Note: It is expected that the required cesting will normally be completed in 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 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 I
specific valves.
In the event that a valve must be declared i inoperable as a result of cold shutdown testing, the applicable unit start-up i.
limitations will be as stated in the Tech-nical Specification, Limiting Conditions for operation.
Reactor Refueling RR Tests which are conducted during plant refueling outages but not less than once every two years.
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4-7 Revision 1 9/24/80
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L. Max Stroke Time lists the maximum allowable full stroke time in seconds for power operated valves in Category A or B.
.M. Relief Request references the relief request contained in Section 4.3 that applies to the particu-lar valve. Also included in Section 4.3 are generic relief requests that are not 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 4.1-2 for the explanation of isolation valve groupings.
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,..;., TABLE 4.1-1 LIST OF SYSTEMS INCLUDED IN THE VALVE PROGRAM UNIT-1 UNIT-2 SYSTEM REFERENCE REFERENCE SYSTEM NUMBER 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 Stsndby Liquid Control 1100 40. 82 Reactor Water Cleanup 1200 47 x 88 Reactor Core _ Isolation Cooling 1300 50 89
. Core Spray 1400 36 78 Pressure Suppression 1600 34 76 High Pressure Coolant Injection 2300 46 87 Main Steam 3000 13-l&2 60-1&2 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 I
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4-9 Revision 1 9/24/80
/' TABLE 4.1-2 3
AUTOMATIC ISOLATION VALVE GROUPINGS Group 1: The valves in. Group 1 are closed upon any one of the following conditions:
- 1. Reactor low-low water level
- 2. Main steamline high radiation
- 3. ' Main steamline high flow
- 4. Main steamline tunnel high temperature
- 5. Main steamline low pressure j l
Group 2: The. actions in Group 2 are initiated by any one of the following conditions:
- 1. Reactor low water level
.tL 2. High.drywell pressure Group 3: ' Reactor low water level alone initiates the following:
- 1. Cleanup demineralizer system isolation Group 4: Isolation valves in the high pressure coolant injec-tion system (HPCI) are closed upon any one of the following signals:
- 1. HPCI steamline high flow
- 2. High temperature in the vicinity of the HPCI steamline
- 3. Low reactor pressure Group 5: Same as Group 4 except applies to RCIC u-4-10 Revision 1 9/24/80
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SECTION 4.2 TABLES.FOR INSERVICE VALVE TESTING PROGRAM A, QUAD CITIES UNIT-1 B. QUAD CITIES UNIT-2 t
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INSERVICE TEDTING PREGRAM UNIT - 1
@ Commonwealth Edison ISI- CLASS 1.2.& 3 VALVES QUAD CITIES NUCLEAR POWER '3 T A T IO N SYSTEM P 6 ID REVISION - DATE PAGE NUCLEAR BOILER INSTRUMENTATION ISI-35 Sh.1 1 - 09/24/80 1 of 36 VALVE NUMBER [ REMARKS AT RR l-263-2-15A D-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1-263-2-13A D-5 1 AC 0.5 MC SA O C CT-1 RR VR-14
. AT RR 1-263-2-19A C-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR
- l-263-2-17A D-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-11 E-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1-220-54 E- 5 ,( l AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-15B D-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263 2-13B D-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1-263-2-178 D-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14
.... ______. ........ ___..... _ . _ . . . ...____...__... .______ ____.... ...__.. .......~.___.......___. .......... ..__..................__.
AT RR l-263-2-198 C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14
......... __ .__..... ...___.. . . . . . . . .___......__.._ ______ ..__.... ..___.. ...__._~...__...._____. .._______. .....___...........___...
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth ISI- CL ASS 1. 2. & 3 VALVES
- Edison QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-35 Sh.1 1 - 09/24/80 2 of 36 VAtVE NUMBER [ [ REMARKS AT RR l-263-2-20A B-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-23A C-5 1 AC 0.5 XFC SA O C CT.1 RR VR-14 AT RR l-763 2-31B C-5 1 AC 0.5 XFC SA O C CT.1 RR VR.14 AT RR 1 263 2-31G C-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1 263-2-31C C-5 1 AC 0.5 XFC JA O '
C CT-1 RR VR-14 AT RR l-263-2-3111 C-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-31D C-5 1 AC 0.5 XFC SA O C CT-1 RR VR.14 AT RR 1 263-2 27 A-5' 1 AC O.5 XFC SA O C CT.1 RR VR-14 AT RR l-263-2-25 B-5 1 AC 0.5 XFC SA O C CT-1 RR VR.14 AT RR l-263-2-31J C-5 1 AC O.5 XFC SA O C CT-1 RR VR.14 F
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\ . i INSERVICE TESTING PROGRAM UNIT - 1
@ CommonwealthQUAD CITIES NUCLEAR POWER ST ATION Edison ISI - CL ASS 1, 2. & 3 VALVES SYSitM . P 6 ID REVISION - DATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-35 Sh.1 1 - 09/24/80 3 of 36 VALVE NUMBER [ REMARKS AT RR l-263-2-31E C-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR l-263-2-311t C-5 1 AC 0.5 XFC SA 0 C CT.1 RR VR-14
.....--.....~--..--.
AT RR 1-263-2-23B C-5 1 AC 0.5 XFC SA 0 C Cr.1 RR VR-14 AT RR l-263-42A C-5 1 AC 0.5 XFC SA 0 C CT.1 RR VR-14 AT RR l-263-2-208 B-5 1 AC 0.5 XFC SA 0 C CT.1 RR VR-14 AT RR 1 263-2-20C B-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-263-2-23C C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR.14 AT RR 1-263-2-31M C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR
- 1-263-2-31T C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR l-263-2-31N C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 O
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INSERVICE TESTING PROGRAM UNIT - 1
@ CommonmalthQUAD CITIES NUCLEAR POWER STATION Edison ISI - CL ASS 1.2.& 3 VALVES SYSTEM P 610 REVISION - DATE , FAGE NUCl. EAR BOILE.1 INSTRUHENTATION (CONTINUED) ISI-35 Sh.1 1 - 09/24/80 4 of 36 VAtVE NUMBER [ [ REMARKS AT KK 1-261-2-31U C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1-263-2-31P C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14
_ _ _ _ _ _ _ . . . . . ......... .....__ .____... .__ __. ........ .___.__..... __.. _ _ . . . . ......... ......L ................__ .__......................
AT RR l-263-2-33 E-3 1 AC 0.5 XFC SA O C Cr-1 RR VR-14 AT RR l-263-2-31V C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-31R C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-31W C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 .
AT RR 1-263-2-23D C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-42B C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-263-2-20D B-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 l
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI - CLA SS 1.2.& 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM P b 10 REVISION - DATE PAGE RECIRCULATION ISI-35 Sh. 2 1 - 0')/24/80 5 of 36 VALVE NUMBER [ REMARKS BT CS 45 VR-6 1-202 5A D-6 1 B 28 GA MO O C PIT RR BT CS 45 VR-6 1-202-55 D-3 1 B 28 GA HO O C PIT RR AT RR BT OP 5 1-220 44 E-2 1 A 0.75 GL A0 0 C PIT RR GROUP 1 ISOLATION FST OP AT RR l-220-45 E-1 1 A 0.75 GL A0 0 C BT OP 5 GROUP 1 ISOLATION FST OP i AT RR 1-220-67A F-5 1 AC 0.5 XFC SA, O C CT-1 RR VR-14 AT RR
! l-220-67B F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR l-220-67C E.F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR j 1-220-67D F-5 1 AC O.5 XFC SA 0 C CT-1 RR VR-14 AT RR l-320-89A E-1 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR
, 1-220-89B E-1 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 4 ...._____... . . . . . . . . ...__.. ........ ..._. ......... .....__ ....___. ____ .. .....__.,__.____ ........ __........ ____...__._______........
AT RR 1
1-220-67E E-5 1 AC 0.5 KFC SA 0 C CT-1 RR VR-14
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INEERVICE TESTING PREGRAM UNIT - 1
@ Commonwealth Edison ISI - C L A S S ' 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE RECIRCULATION (CONTINUED) ISI-35 Sh. 2 1 - 09/24/80 6 of 36
/
VAtVE NUMBER [ REMARKS I
AT RR 1-220-67F F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14
.........__... .__...__....... ......... . . . . . . ....__.__......_~.__.... . _ _ _ _ _ ...........__...... ...............__........
o AT RR 1-220-67C E-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1 220 67H F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-263-2-6A B-7 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-263-2-5A B-7 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-220-20A B-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR l-220 19A B-6 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-220-22A D-8 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-220-21A D-8 1 AC 0.5 XFC SA 0 C CT-1 PR VR-14
..___...__..~.__..__..____. .__..... .______ ........ ....... ..________..... ..___... ...____ .........__.____...__...__.......__.__.....
AT RR l-220-20B A-3 1 AC 0.5 .XFC SA 0 C CT-1 RR VR-14
......._....~...._......... .__. ... ...._.. . ...... ....... ............... ........ ....... .............................-_.- .........
a AT RR 1-220-198 A-3 1- AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR l-262-2-6B B-2 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 I
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C INSERVICE TESTING PRCGRAM UNIT - 1
@ CommonwealthQUAD CITIES NUCLEAR POWER STATION Edison ISI - CL ASS 1.2.& 3 VALVES
/ i l
SYSTIM P 6 ID REVISION - DATE PAGE l RECI:tCULATION (CONTINUED) ISI-35 Sh. 2 1 - 09/24/80 7 of 36 l
/ l
/
VALVE NUMBER [
- RE M ARKS
/
AT RR 1-262-2-58 B-2 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1-220-22B D-1 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR l-220-218 D-1 1 AC 0.5 XFC SA O C CT-1 RR VR-14
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- P INSERVICE TESTING PROGRAM UNIT - 1
@ CommonwealthQUAD CITIES NUCLEAR POWER STATION Edison ISI - CL ASS 1. 2. & 3 VALVES P is ID REVISION - DATE PAGE .
SYSTEM CONTROL ROD DRIVE ISI-41 1 - 09/24/80 8 of 36 VALVE NUMBER [ REMARKS
/
(177)
- VR-11
- SCRAM TESTING 1-0305-127 D-9 1 B 0.75 GA A0 C O BT (177)
- VR-ll 1-0305-126 D-10 1 B 1.0 CA AO C O BT (177) CT-1 * " "
1-0305-114 E-9 2 C 0.75 CK SA C O VR-ll BT CS 1-0302-21A F.2 2 B 1.0 GL A0 0 C FST CS VR-19 BT CS 1-0302-21B F-7 2 B 1.0 GL A0 0 C FST CS VR-19 BT CS 1-0302-22 F-3 2 B 2.0 GL A0 O C FST CS VR-19 a
s .
INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI - CL A SS QUAD CITIES NUCLEAR POWER STATION
- 1. 2. & 3 VALVES SYSTEM P 6 ID REVISION -
DAT E PAGE RESIDUAL HEAT REMOVAL ISI-37 1 - 09/24/80 9 of 36
/
VAtyt NUMBER [ D REMARKS I
1-1001-7A B-b Z B 14 GA MO O O BT OP 90 1-1001-75 E-6 2 B 14 CA MO O O BT OP 90 1-1001-7C B-6 2 8 14 GA HQ O O BT OP 90 1-1001-7D E-6 2 8 14 CA HQ O O BT OP 90 1-1001-67A B-3 2 C 12 CK SA C O CT-1 OP 1-1001-67B E-3 2 C 12 CK SA C O CT-1 OP 1-1001-67C B-9 2 C 12 CK SA C 0 CT-1 OP 1-1001-67D E-9 2 C 12 CK SA C 0 CT-1 OP 1 1001-125A B-5 2 C 1 RV SA C 0 CT-2 RR 1 1001-125B E-5 2 C 1 RV SA C 0 CT-2 RR
........__..~... __. ....___ ....____ ____... ..____ . ....... ........ ....... ........ .. ____ ...__... ..._...........___....______..__...
1-1001-125C B-7 2 C 1 RV SA C 0 CT-2 RR 1-1001-125D E-7 2 C 1 RV SA C 0 Cr-2 RR 1-1001-43A B-4 2 B 14 GA MO C C BT OP 105 1-1001-43B E4 2 B 14 CA MO C C BT OP 105 1 1001-43C B-5 2 B 14 GA H0 C C BT OP 105 1
__ .. ___.._~.___ .. ....... ........ .__.... ........ .. _... ...___ . ....... . _ _ _ _ _ _ . .._____ ..__________.___.........._________...__...
1-1001-43D E-8 2 B 14 GA H0 C C BT OP 105
.....__. ...~..__ .. ....... ...... . a._.... .._._ .. ___.... ._______ ....... ....___. ...__.. .._..___ ...______....__________..........__
l-1001-6A F-5 2 E 24 PTF M LO NA t
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INSERVICE TEDTING PRZ' ORAM UNIT - 1
@ Commonwealth E& son ISI- CLASS QUAD CITIES NUCLEAR POWER STATION
- 1. 2. & 3 VALVES SYSTEM P b 10 REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-37 1 - 09/24/80 10 of 36 k
VALVE NUMBER
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REMARKS 1-1001-65 n-6 2 E 24 BTF M LO NA .
1-1001-42A C-5 2 E 14 CA H LC NA 1 1001-428 E-5 2 E 14 GA H LC NA
._____....._~.............. ............... ........ . . . . . . . ..__...___ _ . . . . . ..___ __. . . . . . . . .......__....._____ .____..___.______........
1-1001-42C C-6 2 E 14 CA H LC NA 1-1001-42D E-6 2 E 14 CA H LC NA 1-1001-66A C.2 2 E 12 CA H LO NA 1-1001-66B E-2 2 E 12 CA H LO NA 1-1001-66C C_8 2 E 12 CA H ID NA 1-1001-66D E-9 2 E 12 CA H IA NA 1-1001-15A B-2 2 E 18 CA H LO NA 1-1001-158 B-9 2 E 18 CA H LO NA 1-1001-17A B-2 2 E 18 CA H LO NA 4
1-1001-178 s-9 2 E 18 CA H LO NA 1 1001 141A B.3 2 E 2 CA H LO NA 1-1001-1418 E-3 2 E 2 CA H LO NA 1 1001-141C B-9 2 E 2 GA H LO NA i iodi~ii.ib'""[T' "~j"~ ""E "' ~" "~
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............~.............. ........ .....__ ........ ____... ........ ._..... .. .___. ____... ....___...__............................__.
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.hh INSERVICE TESTING PROGRAM UNIT - 1
@ c. Edison ISI - C L. A S S QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P 610 REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-37 1 - 09/24/80 11 of 36 valve NUMBER [ [ REMARKS 1-1001-142A B.3 2 C 2 CK SA C 0 CT-1
- VR-20
- SEE VR-20 1-1001-1428 E-3 2 C 2 CK SA C 0 CT-1
- VR-20
- BEE VR-20 1-1001-142C B-9, 2 C 2 CK SA C 0 CT-1
- VR-20
- SEE VR-20 1-1001-142D E-9 2 C 2 CK SA C 0 CT-1
- VR-20
- SEE VR-20
INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth [d son ISI - CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSIEM
- P 6 ID REVISION - DATE PAGE RESIDUAL HEAT REHOVAL (CONTINUED) ISI-39 1 - 09/24/80 12 of 36 VAlvt NUMBER [ REMARKS i
AT RR l-1001-29A A-5 1 A 16 GA HO C O BT CS 25 VR-22
___.........m.............. ........ ....... ........ ... ._. ........__.... ........ _ _ _ . . . . ...___ ............ ................. _____..
AT RR l-1001-298 A-7 1 A 16 CA HO C O BT CS 25 VR-22
............~_____.. ....... ........ ....... ..... .. ....... ......._. . . . . _ . ......... . . . . . . . ....._............. .............. ..........
AT RR 1-1001-47 C-5 1 A 20 GA HQ OLC C BT CS 40 VR-9 GROUP 2 ISOLATION AT RR l-1001-50 B-5 1 A 20 CA HD OLC C BT CS 40 VR-9 GROUP 2 ISOLATION AT RR l-1001-60 A-7 1 A 4 CA HO O&C C BT CS 25 VR-9 GROUP 2 ISOLATION AT RR l-1001-63 A-6 1 A 4 CA HQ OLC C BT CS 25 VR-9 GROUP 2 ISCLATION PIT RR 1-1001-68A A-5 1 C 16 CK SA C O CT-1 CS VR-7 PIT RR 1-1001-68B A-6 1 C 16 CK SA C O QT-1 CS VR-7 1-1001-16A D-2 2 B 18 GL HQ OkC O BT OP 125 1 1001-168 D-10 2 B 18 GL HQ OkC O BT OP 125 l-1001-18A B-4 2 B 3 CA HO C OkC BT OP VR-8
....m_. .... ....... ..._.... ....... ... ... ....... ........ ....___ .._ .... ....... ........... _.._..._..._________. __......
1-1001-19A D-2 2 B 18 CA HQ O O BT CS 125 VR-21 1-1001-198 D-9 2 B 18 CA HO O O BT CS 125 VR-21 i
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INSERVICE TESTING PROGRAM g' _ ,
@ CommonwealthQUAD CITIES NUCLEAR POWER STATION E& son ISI- CL A SS 1. 2. & 3 VALVES SYSTIM P 6 ID REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-39 1 - 09/24/80 13 of 3G VALVE NUMBER [ REMARES AT KR 1-1001-20 C-8 2 A 3 CA MO O&C C BT OP 25 GROUP 2 ISOLATION i
AT RR 1-1001-21 C-8 2 A 3 CA MO O&C C BT OP 25 GROUP 2 ISOLATION i
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 O CT-2 RR AT RR 1-1001-23A A-5 2 A 10 GA M0 C C BT OP 15 AT RR 1-1001-23B A-6 2 A 10 GA MO C C BT OP 15 AT RR 1-1001-26A A-5 2 A 10 GA M0 C C BT OP 15 AT RR 1-1001-26B A-6 2 A 10 CA M0 C C BT OP 15 1-1001-2EA A 2 8 16 GL MO O O BT CS 90 VR-22
. . . . . . ... -4......... . _ ....
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AT RR '
1-1001-36A B-2 2 A 14 GL MO C OkC BT OP 60
...... _____~_ ............ ........ ....... ........ ..._.._ ........ ....__. .....__. ... ... ............_________ .....................
AT RR 1-1001-368 B-8 2 A 14 GL MO C OGC BT OP 60 AT RR 1_1001-37A B-3 2 A 6 GL M0 C O&C BT OP 60 AT RR 1-1001-37B B-7 2 A 6 GL MO C OEC BT OP 60 I
4 d INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI- CLASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P fr ID REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-39 1 - 09/24/80 14 of 36 D
VAtVE NUMBER
[ REMARKS i
AT RK 1-1001-34A A-2 2 A 16 GA MO C OEC BT OP 125 1-1001 34B B-7 2 A 16 GA MO C OEC BT OP 125 1-1001-2A F-3 3 C 12 CK SA C 0 CT-1 OP 1-1001-25 -
F-3 3 C 12 CK SA C O CT-1 OP 1-1001-2C F-7 3 C 12 CK SA C 0 CT-1 OP 1-1001-2D F-7 3 C 12 CK SA C O CT-1 OP .
4 1-1001-5A E3 3 B 12 GL M0 C OEC BT OP 90 ,
1-1001-5B E-7 3 5 12 CL M0 C OEC BT OP 90 1-1001-1A G-4 3 . E 14 GA H LO NA 1 1001-1B G-4 3 E 14 GA N LO NA 1-1001-1C G-6 3 E 14 GA M LO NA i . . . . . . . . . . . . ......... ....... ........ ....... ........ ....... ........ ....... . . . . . . . . ....... ........ .........................__........
1 1001 1D G.6 3 E 14 GA H LO NA l 1-1001-3A G-3 3 E 12 GA N LO NA 4
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI- CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 val.VES SYSTEM PIr10 REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) 131-39 1 - 09/24/80 15 of 36 VALVE NUMBER [ [ REMARKS 1-1001-35 G-3 3 E 12 GA H LO NA 1-1001-3C G-7 3 E 12 CA M LO NA 1-1001-3D G-7 3 E 12 CA M LO NA 1-1001-201A F-3 NC E 14 BTF M 14 NA 1-1001-2015 F-7 NC E 14 BTF M LO NA 1/2-1099-1 G-2 3 E 16 GA N LC NA 1-1001-33A B-5 3 E 16 CA H LO NA 1 1001-338 B-6 3 E 16 GA H LO HA 4
(- v) _)
INSERVICE TESTING PROGRAM ,,,,7 _ ,
ISI - CLASS 1.2.& 3 VALVES
@ Commonwealth %
QUAD CITIES NUCLEAR POWER STATION P 610 REVISION - DATE PAGE STSTE,M STANDBY LIQUID CONTROL ISI-40 1 - 09/24/80 16 of 36 VAlvt NUMBER [ REMARKS C.3 C 1.5 CK SA C 0 CT-1 CS/RR VR-10 1 1101-15 1 1-1106A C-4 2 D 1.5 EXP C 0 IYr RR 1-1106B D-4 2 D 1.5 EXP C ' O DT RR 1-1101-43A D-6 2 C 1.5 CK SA C 0 CT-1 0F 1-1101-43B E-5 2 C . 1.5 CK SA C 0 CT-1 0F 1 1105A C-6 2 C 1.5 RV SA C 0 CT-2 RR 1 1105B D-5 2 C 1.5 RV SA C 0 CT.2 RR 1-1101-4 E-8 2 E 2.5 CA M ID NA 1-1101-8 D-8 2 E 2.5 GA M LC NA 1-1101-3A D-7 2 E 2.5 CA M LO NA 1-1101-38 E-7 2 E 2.5 GA M LO HA
.......... ._~.......
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 1101-28 E-5 2 E 1.5 GL M LO NA
.............~....... ....... ........ ....... ........ ....... ...___.. ....... .__..... ....... ....__.........................__.........
1-1101-22 C-4 2 E 1.5 CL 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 CL M Lo NA t
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/. . s' l2 INSERVICE TESTING PROGRAM , ,
@ Commonwealth Edison ISI- CL ASS QUAD CITIES NUCLEAR POWER STATION
- 1. 2 & 3 VALVES SYSTfu Pfr10 nEVIS10N - DATE PAGE REACTOR CORE ISOLATION COOLING ISI-50 1 - 09/24/80 18 of 36 valve nousin
(( f [ p neuAnKs AT RR 1-1301-16 B-2 1 A 3 GA HQ 0 C BT OP 25 GROUP 5 ISOLATION PIT RR AT RR 1-1301 17 B-3 1 A 3 CA HQ O C BT OP 25 GROUP 5 ISOLATION AT RR 1-1301-40 D-2 NC AC 2 CK SA C C CT-1 RR VR-13 AT RR 1 1301-41 D-2 NC AC 8 CK SA C C CT-1 RR VR-13 j AT RR 1-1301-15A B-2 1 AC .5 ,XFC SA 0 C CT-1 RR VR-14 1 AT RR i 1-1301-155 B-2 1 AC .5 XiC SA 0 C CT-1 RR VR-14 4
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INSERVICE TEDTING PROGRAM UNIT - 1 ISI - CL ASS
@ Commonwealth Edison
- 1. 2. & 3 VALVES.
QUAD CITIES NUCLEAR POWER ST ATIO N P fr ID REVISION - DAff PAGE SYSTEM CORE SPRAY ISI-36 1 - 09/24/80 19 of 36
/
REMARKS valve NUM8ER u
CT-1 CS C-3 C 10 CK SA C 0 PIT RR VR-7 1-1402-9A 1
............. ....... .......~....... . . _ . . . . .................
CT-1 CS C-4 10 CK SA C O PIT RR VR-7 1-1402-95 1 C
............. ....... . . . . . - . ........ .......~...............
1 1402-25A C-2 1 B 10 CA MO C 0 BT OP 15 i
1-1402-25B C-5 1 8 10 GA HQ C 0 BT OP '15 B-2 10 CA HQ O O BT OP 15 1-1402-24A 2 B B-5 8 10 CA HQ O O BT OP 15 1-1402-24B 2 1 1402-28A C-9 2 C 2 RV SA C 0 CT-2 RR 4
1-1402-288 D-6 2 C 2 RV SA C 0 CT-2 RR 1-1402-38A C-8 2 8 1.5 CA MO O C BT OP VR-8 1-1402-388 D-7 2 B 1.5 GA MO O C BT OP VR-8 j . . . - - . . - _ _ . . ..-___.. . . . . . . ...__....
l 1 1402-8A E-9 2 CE 12 SCK SA C/LO O CT.1 OP 1-1402-8B E-6 2 CE 12 SCK SA C/LO O CT-1 OP AT RR 1-1402-31A E-3 1 AC .5 XFC SA O C CT-1 RR VR.14 AT RR 1-1402-31B E-3 1 AC .5 XFC SA O C CT-1 RR VR-14 1-1402 6A D.3 1 E 10 GA H LO HA 7
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI - CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P Ir ID REVISI0tt - DATE PAGE CORE SPRAY (CONTINUED) ISI-36 1 - 09/24/80 20 of 36 VALVE NUM8ER [ REMARKS
/
1-1402-2A G-7 2 E 12 GA M LC NA 1-1402-25 G-4 2 E 12 CA M LC NA 1-1402-34A G-4 2 E 18 BTF M IA NA 1-1402-345 F-3 2 E 18 BTF M LO NA 1-1402-4A A-8 2 8 8 GL @ C C BT OP 60 1-1402-4B C-7 2 8 8 GL M0 C' C BT OP 60 1-1402-13A E-9 2 CE 1.5 SCK SA/M C/14 O CT-1
- VR-20
- SEE VR-20 1-1402-13B E-6 2 CE 1.5 SCK SA/M C/I4 0 CT-1
- VR-20
- SEE CR-20 I
6 1
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9 INSERVICE TESTING PROGRAM , .,
@commith gig ISI - CLASS 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER - STATION SYSTIM P 610 REVIM006 - DATE PAGE PRESSURS SUPPRESSION N-34 ,
1 - 09/24/80 21 of 36 REMARKS VALVE NUMSER 4
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1-1601-21 C.6 NC A 18 BTF A0 C C BT OP 10 GROUP 2 ISOLATION AT RR 1-1601-22 C-6 NC A 18 BTF AD C C BT OF 10 GROUP 2 ISOLATION AT RR 1-1601-55 A-6 NC A 4 CA A0 0 C BT OP 10 GROUP 2 ISOLATION AT RR 1-1601-56 D-6 NC A 18 BTF A0 O C BT OP 10 CROUP 2 ISOLATION FST OP AT RR 1-1601-57 C-9 NC A 1 CL MO O C BT OP 15 GROUP 2 ISOLATION AT RR 1-1601-58 D-7 NC A , 1 CL A0 C C BT OP 15 GRMP 2 ISOLATION AT RR 1 1601-59 D-7 NC A 1 GL A0 0 C BT OP 15 GROUP 2 ISOLATION FST OP AT RR 1 1601-20A D-9 NC A 20 BTF A0 C OkC BT CS 10 VR-23 FST CS AT RR 1-1601-31A D-9 NC AC 20 CK SA C OLC CT-1 OP AT RR 1-1601-208 E-9 NC A 20 BTF A0 C OkC BT CS 10 VR-23 FST CS
- AT RR 1-1601-31B E-9 NC AC 20 CK SA C O&C CT-1 OP 1-1601-23 B-3 NC A 18 BTF AO C C h 10 GROUP 2 ISOLATION i I
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth E& son ISI - CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P Ir 10 REVISION - DATE PAGE PRESSURE SUPPRESSION (CONTINUED) N.34 1 - 09/24/80 22 of 36 valve mumsER [ 8 [ p REMARKS
/ /
ar un i 1-1601-24 B-2 NC A 18 BTF AD C C BT GP 10 GROUP 2 ISOLATION
............. ....... ................ . . . . . . . . ...._.. . . . . . . . . ....... ...__... . . . . . . . . . . . . . . . 4....._ . . . . . . . . . . ..........................
AT RR l-1601-60 B-3 NC A 18 BTF A0 C C BT OP 10 GROUP 2 ISOLATION AT RR 1-1601-61 B-2 NC A 2 GL A0 C C BT OP 15 GROUP 2 ISOLATION AT RR 1-1601-62 E-2 NC A 2 GL A0 C C BT OP 15 GROUP 2 ISOLATION AT RR 1-1601-63 E-2 NC A 6 BTF AD C C BT CP 10 GROUP 2 ISOLATION
............. . . . .: . . ....__.m....... . . . . . . . ......... ........ ....... ...___ . ....... . . . . . . . . ....... ....................................
AT RR 1-8803 C-6 NC A 2 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 1-8804 D-6 NC A 2 CL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 1-8801A C-3 NC A 0.5 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 1-8801B D-3 NC A 0.5 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 1-8801C D-3 NC A 0.5 GL A0 O C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 1-8801D E-3 NC A 0.5 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FS
...T.... . . .O. .P-------------------- - - - - - - - - - - - - - - - - - - - - - - - - -
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wh INSERVICE TESTING PROGRAM UNIT - 1
@c % ISi- CLASS QUAD CITIES NUCLEAR POWER STATION
- 1. 2. & 3 VALVES SYSTEM P 610 REVISION - DATE PAGE PRESSURE SUPPRESSION (CONTINUED) M-34 1 - 09/24/80 23 of 36 VALVE NUM8ER [ REMARKS r
AT RR
- 1-8802A C-3 NC 'A 0.5, GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 1-88028 D-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP
............~....... ....... ........ ...... ........ ....... ........ ....... ........ ....... .....___ ....._.............................
AT RR 1-8802C D-3. NC A O.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OF AT RR 1-8802D E-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION F"JT OP 1-1601-32A E-2 NC C 18 CK SA C OkC CT.1 OP l 1-1601-325 E-2 NC C '18 CK SA C OkC CT-1 OP i
i 1-1601-32C E-2 NC C 18 CK SA C OkC CT-1 OF 1-1601-32D E-2 NC C 18 CK SA C OkC CT-1 OP 1-1601-32E E-2 NC C 18 CK SA C OkC CT-1 OF 1-1601-32F E-2 NC C 18 CK SA C OkC CT-1 OP 1-1601-33A E-7 NC C 18 CK SA C OkC CT-1 OF 1-1601-33B E-7 NC C 18 CK SA C OEC CT-1 OP 1-1601-33C E-7 NC C 18 CK SA C OEC CT-1 OP 1-1601-33D E-7 NC C 18 CK SA C OkC CT-1 OP
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI - CLASS 1.2.& 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM . P 610 REVISION - DATE PAGE ,
PRESSURE SUPPRESSION (CONTINUED) M-34 1 - 09/24/80 24 of 36 VAtVE NUM4(R [ [ REMARKS 1-1601-33E E-7 NC C 18 m SA C OkC CT-1 OP .
1-1601-33F E-7 NC C 18 m SA C OEC CT-1 OP
.______.... ~....__. .._____ .. __... .... __ ...__... .....__ ........___.... . _____. ___.... .....___ ___________....__..................
i 1-220-81A E NC C 1. & SA C O CT-1 CS VR
. _ _ . ... -4.. ._ .. ........ ... ... ... ... ... .. .____... ..._... ... .. ..__ .. ..__.... .___ -24 ...______ ______..........
l-220-81C E-4 NC C 1 CK SA C O CT-1 CS VR-24 1-220-81D E-5 NC C 1 CK SA C 0 CT-1 CS VR-24 1-220-81E E-5 NC C 1 m SA C 0 CT-1 CS VR-24 I
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- INSERVICE TEDTING PROGRAM UNIT - 1
@ commen.nlih ISI- CL ASS 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER STATION 3YSTEM P fr 10 REVISION - DATE PAGE HIGH PRESSURE COOLANT INJECTION ISI-46 1 - 09/24/80 25 of 36 vatvE muusta [ [ [
- REMARKS AT RR 1-2301-4 c-9 1 A 10 GA MO O OEC BT CS 50 VR-15 GROUP 4 ISOLATION FIT RR AT RR 1-2301-5 B-10 1 A 10 CA MO O OliC BT CS 50 VR-15 GROUP 4 ISOLATION
.......___..m___........... .. ..... ....... ......_. ....__. ....____................ ....... ...........................................
1-2301-3 A-6 2 8 10 CA MO C 0 BT OP 25 1-2301-68 A-6 2 D 16 RPD SA C 0 *
- RPD NOT TESTABLE 1-2301 69 A-6 2 D 16 RPD SA C O
- OPEN STROKE VERI.
1-2301-34 D-7 2 AC 2 CK SA C OEC CT-1 VR-13 FIED BY PUMP TEST
............~.........__... ...__... ....... ........ ....... ......... . . . . . . ......... .
.O.P./.R.R.. ............................................
AT RA
- OPEN STROKE VERI.
1-2301-45 B-8 2 AC 24 CK SA C OEC CT-1 VR-13 FIED BY PUMP TEST
. . . . . . . . _ _ . ..___........... ............. . ........ ....... ......... . . . . . . ......... .O.P./.R.R ................................_.__......__
1-2301-35 E-7 2 8 , 16 GA M0 C OEC BT OP 120
............~....... ....... ........ ....... ........ ....... ........ ....... ........ ....___ ....................._____.................
1-2301-36 E-9 2 B 16 CA HQ C O&C BT OP 120 1-2301-6 F-2 2 8 16 CA MO O OkC BT OP 120
.......___. ~.............. ........ ....... ........ ....... ......... . . . . . . ......... ....... ...............___.........................
1-2301-20 E-2 2 C 16 CK SA O O CT-1 OP *
............~...........__. ........ ....... ........ ....... .__..... ................ __..... ......................................... .
1-2301-14 C-6 2 8 4 GL MO C Obc BT OP VR-8
............m.... .. ....... ........ ....... ........ ....... ........ ....... ........ ....... ._.........................................
1-2301-39 E-8 2 C 16 CK SA C O CT-1
- VR-12
- SEE VR-12
............m.___.......... ........ ....... ........ ....... ........ ....... .....__. ....... ...........................................
1-2301-40 D-7 NC C 4 CK SA C O CT-1 ** VR-20 ** SEE VR-20 t
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l- J INSERVICE TESTING PROGRAM UNIT - 1 ISI- CLASS
@ can nmaith Edrson QUAD CITIES NUCLEAR POWER ST ATION
- 1. 2. & 3 VALVES P 610 REVISION - DATE PAGE SYSTIM HICH PRESSURE COOLANT INJECTION (CONTINUED) ISI-46 1 - 09/24/80 26 of 36 VAtVE NUMBER [ REMARKS 1-2301-8 D-6 2 B 14 CA MO C O BT OP 45 PIT RR D-6 14 & SA C O CT-1 CS VR-7 1 2301-7 2 C SCK SA C/li 0 CT-1 OP* *OPEN STROKE VERI-1-2301-74 B-8 2 CE 12 FIED BY PUMP TEST AT RR 1-2301-26 D-9 1 AC .5 XFC SA O C CT-1 RR VR-14 AT RR D-9 .5 XFC SA O C CT-1 RR VR-14 1-2301 27 1 AC
. 1-2301-22 B-1 2 E 16 CA N LO
. . . . . . . . . . . . ................ ....._ . ....... ........ ....... ........ . . . . . . . ......... . . .N.A. . . .............................. ............
1-2301-56 F-8 2 E 16 BTF M LO NA
............~.............. ............... ......_. ....... .....4................ ....... .._..........._............................
SCK, SA/M C/II O CT-1 OP2
- OPEN STROKE VERI-1-2301-71 D-7 2 CE 2 FIED BY PUMP TEST
............~....... ....... ........ ....... ........ ....... ........ ....... ........ ....... ...........................................
1-2301-9 D-5 2 B 14 CA HG O O BT OP 45
............~....... ....... ........ ....... ........ ....... ........ ....... ........ . . . . . . . ........
7.......
1-2301 10 E-5 2 5 12 CL MO C C BT OP 60 I
............~....... ....... ........ ....... ........ ....... ........ ....... ........ ....... ........ ..........l.........................
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@cm==.alth Edeson OUAD CITIES NUCLEAR POWER STATION P 6 ID REVISION - DATE PAGE SYSTIM MAIN STEAM ISI-13 Sh. I 1 - 09/24/80 27 of 36
/
VALVE huMSER [ [ REMARKS
/
AT RR BTP OF BT CS 5 VR-1 GROUP 1 ISOLATION F-4 A 20 CL AO O C FST CS VR-1 I 1-203-1A 1 l PIT RR
............~............ .
AT RR BTP OP BT CS 5 VR.1 GROUP 1 ISOLATION D-4 A 20 CL A0 0 C FST CS VR-1 1-203-1B 1 RR FIT AT RR BTP OF BT CS 5 VR-1 GROUP 1 ISOLATION 20 CL A0 0 C FST CS VR-1 1-203-lC C-4 1 A PET RR AT RR
- BTP OP BT CS 5 VR-1 CROUP 1 ISOLATION 1-203-1D B-4 1 A 20 CL A0 0 C FST CS VR-1 PIT RR AT RR l-220-1 E-4 1 A 3 GA MO C C BT OP 35 CROUP 1 ISOLATION BT *
- VR-2
- SEE VR-2 1-203-3A F-4 1 BC 6 ERV/SV PS/SA C 0 CT-2 RR VR-3 BT VR-2 1-203-3B D-6 1 BC 6 ERV PS C 0 CT-2 RR VR-3 BT
I J 1 INSERVICE TESTING PROGRAM UNIT - 1
@ CommeewealthQUAD CITIES NUCLEAR POWER STATION Edison ISI- CLASS 1.2.& 3 VALVES SYSTEM P 410 REVISION -
DATE PAGE MAIN SIEAM (CONTINUED) ISI-13 Sh. I 1 - 09/24/80 28 of 36 i
VALVE NUMBER [ [ R! MARKS ST *
- SEE VR-2 BT *
- VR-2 "
1-203-3E D-7 1 BC 6 ERV PS C 0 CT-2 RR VR-3 1-203-4A F-8 1 C 6 SV SA C 0 CT-2 RR VR-26 1-203-4B D-5 1 C 6 SV SA C 0 CT-2 RR VR-26 1-203-4C C-5 1 C 6 SV SA C 0 CT-2 RR VR-26 1-203-4D B-5 1 C 6 SV SA C 0 CT-2 RR VR-26
............~.............. ........ ....... ........ ....... ........ ....... ........ ....... ...........................................
1-203-4E F-8 1 C 6 SV SA C 0 CT-2 RR VR-26 1-203-4F D-5 1 C 6 SV SA C 0 CT-2 RR VR-26 1-203-4G C-5 1 C 6 SV SA C 0 CT-2 RR VR-26 1-203-4H B-5 1 C 6 SV SA C 0 CT-2 RR VR-26 f I
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edrson ISl- CL ASS 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM P fr 10 REVISIO88 - DATE PAGE MAIN STEAM (CONTINUED) ,
ISI-13 Sh. 2 1 - 09/24/80 29 of 36 i
REMARKS VA.VE NUMBER ['
. BT CS 5 VR-1 1-203-2A E-7 1 A 20 GL A0 0 C FST CS VR 1 CROUP 1 ISOLATION PIT RR AT RR BTP OP BT CS 5 VR-1 20 A0 0 FST CS VR-1 GROUP 1 it,0LATION 1-203 28 E-7 1 A GL C PIT RR AT RR BTP OP BT CS 5 VR-1 1-203 2C D-7 1 A 20 GL A0 0 C FST CS VR.1 GROUP 1 ISOLATION PIT RR AT RR
- BTP OP BT CS 5 VR-1 1-203-2D B-7 1 A 20 GL A0 0 C FST CS VR-1 GROUP 1 ISOLATION PET RR AT RR l-220-2 E-7 1 A 3 GA HQ C C BT OP 35 CROUP 1 ISOLATION AT RR 1-220-17A E-B 1 AC 0.5 "FC SA O C CT-1 RR VR-14 AT RR l-220-17B D-8 1 AC 0.5 XFC SA O C CT-1 RR VR-14 l AT RR l-220-17C C-8 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 1:320:17D ... . . . D:S_. ...l... ....oC.. .. 0.5. ... 3fC.. . _S&_. ... 0... .__C._. . .CT:1.. ...BB._ ........ .._YB:36-- -------------------------
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INSERVICE TEDTING PRICRAM UNIT - 1
@ CommonwealthQUAD CITIES NUCLEAR POWER STATION Edison ISI- CL ASS 1.2.& 3 VALVES SYSTEM P 610 REVISION - DATE PAGE MAIN STEAM (CONTINUED) ISI-13 Sh. 2 1 - 09/24/80 30 of 36
--l VALVE NUMBER [ REMARKS AT RR 1-220-18A h,-8 1 AC .5 XFC SA 0 C CT-1 RR VR-14 AT RR 1-220-18B D-8 1 AC .5 XFC SA 0 C CT-1 RR VR-14 1-220-18C C-8 1 AC .5 XFC SA 0 C Cr-1 RR VR-14 AT RR 1-220-18D B-8 1 AC .5 XFC SA 0 C CT-1 RR VR-14 t
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INZERVICE TEDTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI - CL AS S 1. 2 & 3 VALVES QUAD Cl, TIES NUCLEAR POWER STATION SYSTEM P 6 ID REVISNHe - DATE PAGE FEEDWATER ISI-15 1 - 09/24/80 31 of 36 VALVE NUM8ER [ REMARKS AT RR 1-220-58A E-3 1 AC 18 m SA O C CT-1 RR VR-4 AT RR 1-220-58B F-3 1 AC 18 m SA O OEC CT-1 RR VR-4 1-220-59B F-3 2 C 18 3 SA O C CT-1 RR VR-4 1-220-62A E-3 1 AC 18 M SA O C CT-1 RR VR-4 AT RR 1-220-62B F-3 1 AC 18 M SA O OEC CT-1 RR VR-4 i
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@ Commonweshh Edison ISI- CLASS QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE DIESEL GENERATOR STARTING AIR (SERVICE AIR SYSTEM) H-25 1 - 09/24/80 i 33 of 36 VALVE NUM8ER [ -[ REMARKS
/
1-4699-121 E-9 ;!C E 1.5 CA H LO HA .
1-4699-122 E-9 FC E 1.5 CA H LO NA 1-4699-225 D-8 NC E 1.5 BALL H LO HA 1-4699-123 E-9 NC C 1.5 CK SA C 0 CT-1 OP 1-4699-196 E-9 NC C 1.5 CK SA C 0 CT-1 OP 1-4699-226 D-8 NC B 1.5 GL A0 C 0 BT OP 5 1
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI- CLASS 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER STATION A
SYSTEM P fr 10 REVISION - DATE PAGE INSTRUMENT AIR PIPING H-24 Sh. 2 1 - 09/24/80 34 of 36 VALVE NUMBER [ REMARKS AT RR 1-4720 D-3 NC A 1 CA A0 0 C BT OP 10 CROUP 2 ISOLATION FST OP AT RR 1-4721 D-3 NC. A 1 CA A0 0 C MT OP 10 CROUP 2 ISOLATION FST OP AT RR 1-733-1 P-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION AT RR 1-733-2 P-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION AT RR 1 P-7 NC A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION
_-733-3
~
AT RR 1-733-4 F-7 NC A 0.375 BALL SO C C BT OP 5 CROUP 2 ISOLATION AT RR 1-733-5 F-7 NC A 0.375 BALL SO C C BT' OP 5 GROUP 2 ISOLATION j _ _ _ . . . . . . . . . ............____ ........ ....... ........ ....... ........................ ....... ......._......___........__..........__....
AT RR 1-743 B-7 NC AC 0.375 CK SA O C CT-1 RR VR-5 1-736-1 P-7 NC D 0.375 EXP O C DT Ra 1-736-3 P7 NC D 0.375 EXP O C DT RR l ,
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INSERVICE TESTING PROGRAM UNIT - 1
@ Commonwealth Edison ISI - CL ASS 1.2.& 3 VALVES QUAD CITIES NUCLEAR POWER STATION I
1 1
SYSitM P 6 (D REVISION - DATE PAGE f RX BUILDING EQUIPMENT DRAINS (Es SilARED UNIT 1/2 DIESEL CEN. AIR START PIPING) H-43 1 - 09/24/80 35 of 36 l l
i l
N N REMARKS VALVE NUMBER
/ e[ e vs A
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/
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l AT RK 1-2001-15 E-3 NC A 3 CA AD C C BT OP 20 GROUP 2 ISOLATION FST OP l AT RR l 1 2001-16 E-3 NC A 3 CA A0 C C BT OP 20 GROUP 2 ISOLATION FST OP AT RR j 1-2001-3 F-7 NC A 3 GA A0 C C BT OP 20 CROUP 2 ISOLATION I 1
FST OP AT RR 1 2001-4 F-7 NC A 3 CA A0 C C BT OP 20 GROUP 2 ISOLATION FST OP 1/2-4699-46 D-9 NC E 1.5 GA H , In NA 1/2-4699-47 D-9 NC E 1.5 CA H LO NA 1/2-4699 225 D-8 NC E 1.5 BALL M LO NA 1/2 4699-48 D-9 NC C 1.5 CK SA C 0 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 A0 C 0 BT OP 5 9
i i
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INSERVICE TESTING PROGRAM
@ common =aith ISl- CL ASS 1.2.& 3 VALVES UNIT - 1 Edison QUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID REVISION - DATE PAGE DIESEL GENERATOR FUEL UIL
{,UgrT 1 ann 1/2) H-29 1 09/24/80 36 of 36 VALVE NUMBER [ REMARKS l
1-5299-5 E-4 NC C 1.5 CK SA C 0 CT-1 OP 1-5201 E-3 NC B 1 CA SO C 0 BT OP 5 1-5199-155 B-5 NC E 1 CL H 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 0 CT-1 OP 1/2-5201 E-3 NC B 1 CA SO C 0 BT OP 5 1/2-5199-157 C-5 NC C .5 CK SA C C CT-1 OP E
3 4
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INSERVICE TEOTING PRE 2 RAM
@ Commonwealth Edison ISI- CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES UNIT - 2 SYSTEM P & 10 REVISION - DATE PAGE MIP1 5' AR Hof f ER YN9TRIIMFNTATf 0N ISI-77 Sh. I 1 - 09/24/80 1 of 36 VALVE NUMBER e [ [ REMARKS
/
AT RR 2-263-2-15A D-5 1 AC 0.5 XFC SA 0 C CT-1 RR . VR-14 AT RR i
2-263-2-13A D-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-19A C-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-17A D-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-11 E-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2 120-54 E-5,6 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-15B D-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-13B D-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 d
AT RR i
2-263-2-17B D-3 1 AC 0.5 XFC CA 0 C CT-1 RR *VR-14 AT RR
, 2-263-2-198 C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-20A B5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-23A C-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 I
i .~a T.
. .a INSERVICE TECTING PREGRAM UNIT -
2
- 1. 2, & 3 VALVES
@ commonwealthQUAD CITIES NUCLEAR POWER STATION Edisoi; ISI - CL ASS P 610 REVISION - DATE PAGE SYSTEM NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-77 Sh. I 1 - 09/24/80 2 of 36 vetvs muusER
[ 8 g nEMARKS 1
AT RR 0.5 XFC SA 0 C CT-1 RR VR-14 2-263-2-315 C-5 1 AC AT RR 2-263-2-31G C-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-31C C-5 1 AC 0.5 XFC SA 0 C CT 1 RR VR-14 AT RR 2-263-2-31H C-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2 31D C ~> 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-27 A-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2 25 B-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14
- AT R g VR-14 2-263-2-31J C-5 1 AC 0.5 XFC SA 0 C CT-1 RR AT RR 2-263-2-31E C-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 j . . . . . . . . . . . . .....___.
AT RR 2-263-2-31K C-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 3
I i
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@ commonwaith Edison ISI - CL A SS QUAD CITIES NUCLEAR POWER STATION 1,2.& 3 VALVES SYSTEM P Ir 10 REVISION ~ DATE PAGE NUCLEAR BOILER INSTRUMENTATION (CONTINUED) ISI-77 Sh. I 1 - 09/24/80 3 of 36 VAIVE NUMBER REMARKS AT RR 2 263-2-23B C-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR '
1 2-263-42A C-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-208 B-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-20C 5-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263 2-23C C 'i 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT ' RR 2-263-2-31M C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-31T C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-31N C-3 It AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2 31U C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263 2-31P C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-33 B-3 1 AC O.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-31V C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 1
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INSERVICE TESTING PROGRAM UNIT - 2
@ CommonmalihQUAD CITIES NUCLEAR POWER STATION Edison ISI - CL ASS 1.2.& 3 VALVES SYSTEM P lir ID REvlsl0N -
DATE PAGE NI'PT E AR ROYIJR Y NRTilflMFMTATTON (enNTTMllFn) TRY_77 Rh_ 1 1 . OQ/24/AO 4 n f 'M
,/ . -
VALVE FOMBER [ [ h REMARKS AT RR .
2-263-2-31R C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-31W C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2-23D C-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-428 C-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-263-2 20D B-3 1 AC 0.5 XFC SA 0 C CT-1 Rk VR-14 l
4
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. .,4 INSERVICE TESTING PREGRAM UNIT -
2
@ CommonwealthQUAD CITIES NUCLEAR POWER STATION Edison iS1 - CL ASS 1.2.& 3 VALVES SYSTEM P 6 ID REVISION - DATE PAGE i RECIRCULATION ISI-77 Sh. 2 1 - 09/24/80 5 of 36 VALVE NUMBER I [ [ REMARKS
/
BT CS 45 VR-6 2-202-5A D-6 1 B 28 GA MO O C PIT lLR BT CS 45 VR-6 2 202-5B D.3 1 B 28 GA MO O C PIT RR AT RR BT OP 5 2-220-44 E-2 1 A 0.75 GL A0 0 C PIT RR GROUP 1 ISOLATION FST OP AT RR 2-220-45 E-1 1 A 0.75 GL A0 0 C BT OP 5 GROUP 1 ISOLATION FST OP AT RR 2-220-67A F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2 220-67B F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-67C- E . F- 5 1 AC 0.5 XFC SA O C CT.1 RR VR-14 AT RR 2-220-67D F-5 1 AC 0.5 XFC SA O C CT-1 RR VR-14 AT RR 2-220-89A E-1 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-89B E-1 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 i AT RR 2 220-67E E-5 1 AC 0.5 XFC SA O C CE.1 RR VR-14 l t
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INSERVICE TESTING PROGRAM UNIT - 2
@ Commonwealth Edison ISI - CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSHM P fr ID REVisl0N - DATE PAGE RECIRCULATION (CONTINUED) ISI-77 Sh. 2 1 - 09/24/80 6 of 36 vatvE uuMsER REMARKS AT RR 2-220-67F F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-67G E-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR i 2-220-67H F-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-263-2-6A B-7 1 AC 0.5 XFC SA 0 C CF-1 RR VR-14 ,
t . . . _ _ . . . . . . . ......... ....... ........ ....... ........ ....... ........ ....... .__..... ....... ..... .....___ ............................
1 AT RR 2-263-2-5A B-7 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-20A B-5 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-19A B-6 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-22A' D-8 1 AC 0.5 XFC ' SA 0 C CT-1 RR VR-14 AT RR 2-220-21A D-8 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-208 A-3 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-19B A-3 1 AC 0.5 XFC SA 0 C C1-1 RR VR-14 i ............ ........ __..... . . _ _ _ . . . ....... . . . . . . . . ....... ...__... ..__... ..__'.... ....... .. ..__.. ...........___.......__..........
AT RR 2-262-2-6B B-2 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 e
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INSERVICE TESTING PROGRAM UNIT - 2
@ Commonwealth Edison ISI- CLASS 1.2.& 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM P b 10 REVISION - DATE PAGE CONTROL ROD DRIVE ISI-83 1 - 09/24/80 8 of 36 vatvE NuMe!R
[ f [ p REMARKS (177)
- 2-0305-127 D-9 1 B 0.75 CA AD C O BT VR-11
- SCRAM TESTING (177)
- 1.0 VR-11 2-0305-126 D-10 1
- 0.75 0 CT-1 VR-11 2-0305-114 E-9 2
C CK SA C
< BT CS 2-0302-21A .F-2 2 B 1.0 GL A0 0 C FST CS VR-19 BT CS 2-0302-21B F-7 2 B 1.0 GL A0 0 C FST CS VR-19 BT CS 2-0302-22 F-3 2 B 2.0 GL AD 0 C FST CS VR-19 i
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INSERVICE TESTING PROGRAM UNIT - 2
@ Commonwealth Edison 131 - C L A S S QUAD CITIES NUCLEAR POWE FI STATION 1.2.& 3 VALVES SYSTEM P ir ID REVISION - DATE PAGE RESIDUAL HEAT IMHOVAL ISI-79 1 - 09/24/80 9 of 36 VALVE NUMMR
[ REMARKS
/
2-1001-7A B-6 2 2 14 GA HQ O O BT OP 90 2-1001-75 E-6 2 5 14 CA H0 0 0 BT OP 90
............_.............. ........ .6..... ........ . . . . . . . ................-...._.. . . . . . . . ................_ . .........................
2-1001-7C B-6 2 5 14 CA HQ 0 0 BT OP 90
............-.............. ............... ._...... . . . . . . .....___........~....... . . . . . . . ..........____..... .... _.......... ......_.
, 2-1001 7D E-6 2 B 14 CA PO O O BT OP 90
............m.............. ........ ....... ........ . . . . . . . .......... . . . . . . .. . . . . . . . . ....... .................. .....____................
2-1001-67A B-3 2 C 12 CK SA C 0 CT-1 OP 2 1001-67B E-3 2' C 12 CK SA C 0 CT-1 OP 2-1001-67C i ) 2 C 12 CK SA C 0 CT-1 0F 2 1001-67D -9 2 C 12 CK SA C 0 CT-1 OP'
,. 2-1001-125A B-5 2 C 1 RV SA C 0 CT-2 RR 2-1001-125B E-5 2 C 1 RV SA C 0 Cr.2 RR 2 1001 125C B-7 2 C 1 RV SA C 0 CT-2 RR
...... .....~..... . . __... ........ ....... ........ ....... ......... . . . . . . .. ...... ....... ...........................................
2-1001-125D E-7 2 C 1 RV SA C 0 CT-2 RR
- 2-1001-43A B-4 2 B 14 CA HO C C BT OP 105 2-1001-43B E-4 2 8 14 GA HQ C C BT OP 105 i ........... .........__..... . __... ........
2 1001-43C B-8 2 B 14 CA HQ C C BT OP 105 2-1001 43D E-8 2 B 14 CA Ho C C BT OP 105 i
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INCERVICE TEDTING PREGRAM UNIT - 2 i
@ Commonwealth Edison ISI - CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTIM P 6 ID REVISION - DATE PAGE n Rni ntfAT. HRAT REMOVA1. (fMMTf MIf RD) 147 74 1 _ OQ/7/a/AO
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VALVE NUMBER k REMARKS
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2 1901-6A F-5 2 E 24 BTF M LO NA .
2-1001-6B B-6 2 E 24 BTF M LO NA 2-1001-42A C-5 2 E 14 GA M LC NA 2-1001 42B E-5 2 E 14 CA M LC NA 2-1001-42C C-6 2 E 14 GA M ~ LC NA 2-1001-42D E.6 2 E 14 CA M LC NA 2-1001-66A C-2 2 E 12 CA H LO NA
- 2-1001-66B E-2 2 E 12 CA M LO HA i 2-1061 66C C-8 2 E 12 GA H LO NA 2-1001-66D E-9 2 E 12 GA N LO NA 2-1001 15A B-2 2 E 18 GA M LO H;.
2 1001-15B B-9 2 E 18 GA N Lo NA 2-1001 17A B-2 2 E 18 GA H ID NA 2-1001-141A B-3 2 E 2 CA H 14 NA 2-1001-141B E-3 2 E 2 CA H ID HA
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INC'ERVICE TEE; TING PREGRAM UNIT - 2
@ Commonwealth i
Edison ISI- CLASS OUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P 6 ID REVISION - DATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-79 1 - 09/24/80 11 of 36 VALVE NUMB (R REMARKS
[ /
2-1001-141C B-9 2 E 2 GA M LO NA 2-1001-141D E-9 2 E 2 CA M ID NA 2-1001-142A B-3 2 C 2 CK SA C 0 CT-1
- VR-20
- SEE VR-20 2-1001 1428 E-3 2 C 2 CK SA C 0 CT-1
- VR-20
- SEE VR-20 2-1001-142C B-9 2 C 2 CK SA C O CT-1
- VR-20
- SEE VR-20 2-1001-142D E-9 2 C 2 CK SA C O CT-1
- VR-20
- SEE VR-20 1
4
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INCERVICE TESTING PREGRAM UNIT - 2
@ Commonwealth Edison ISI- CL ASS 1. 2. & 3 VALVES QUAD CITIES HUCLEAR POWER STATION SYSTEM P fr ID REVISION - DATE PAGE RRRf hlfAL HFAT GElff) VAT. (fMNTf NIIFD) IST-81 1 09 /2/a /R0 17 nf %
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t
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i AT RR 2 1001-29A A-5 1 A 16 GA HO C 0 BT CS 25 VR-22 2 1001 29B A-7 1 A 16 GA HO C 0 BT CS 25 VR-22 AT RR i
2 1001-47 C-5 1 A 20 CA HQ O&C C BT CS 40 VR-9 GROUP 1 ISOLATION AT RR
- 2-1001-50 B-5 1 A 20 GA HQ OEC C BT CS 40 VR-9 GROUP 2 ISOLATION AT RR 2-1001-60 A-7 1 A 4 GA HQ OEC C BT CS 25 VR-9 GROUP 2 ISOLATION AT RR 2 1001-63 A-6 1 A 4 CA HQ OEC C BT CS 25 VR-9 GROUP 2 ISOLATION PIT RR 2-1001-68A A-5 1 C 16 CK SA C O CT-1 CS VR-7 PIT RR 2-1001-688 A-6 1 C 16 CK SA C O CT-1 CS VR-7
-----"- - - - - - - - ~ ~ ~ ~ ~
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............. ....... . . . . . . . ......... .......~__..._.. . . . . . . . ....... .._..... ........ . . . . . . ___.... .._......_ ...__...._...............
2 1001-18A B-2 2 B 3 CA H0 C OEC BT OP VR-8 2-1001-18B 78D- 3 2 a 3 GA HQ C OkC BT OP VR-8 2-1001-19A D-2 2 8 18 CA H0 0 0 BT CS 125 VR-41 2-1001 19B D-9 2 B 18 CA HO O O BT CS 125 VR-21
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INEERVICE TECTING PREGRAM UNIT - 2
@ Commonwealth Edison ISI- CL ASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P 610 REVISION - DATE PAGE RESIDUAL HEAT REHOVAL (CONTINUED) ISI-81 1 - 09/24/80 13 of 36 VALVE NUMBER REMARKS AT RR 2-1001-20 C-8 2 A 3 GA HQ Obc C BT OP 25 GROUP 2 ISOLATION AT RR 2-1001-21 C-8 2 A 3 GA HQ OEC C BT OP 25 GROUP 2 ISOLATION 2-1001-22A A-2 2 C 1 RV SA C 0 CT-2 RR 2-1001-22B A-9 2 C 1 RV SA C 0 CT-2 RR AT RR 2 1001-23A A-5 2 4 10 GA H0 C C BT OP 15 AT RR 2-1001-23B A_6 ; 2 A 10 CA HQ C C BT OP 15
............. . . . . . . . . . . . . . . .. .............. .....s.. ..__.... ....... ........ ........ . . . . . . ......... .......... .........................
AT RR i 2-1001-26A A-5 2 A 10 GA H0 C C BT OP 15 AT RR 2-1001-265 A-6 2 A 10 CA H0 C C BT OP 15 2-1001-28A A-4 2 5 16 GL HQ O O BT CS 90 VR-22 l 2-1001-28B A-7 2 8 16 GL HQ O O BT CS 90 VR-22 AT RR B2 2 1001-36A 2 A 14 GL HQ C OLC BT OP 60 AT RR 2-1001-36B B-8 2 A 14 GL HO C O&C BT OP 60 AT RR 2-1001 37A B-3 2 A 6 GI . HO C OLC BT OP 60 AT RR 2-1001 37B B-7 2 A 6 GL H0 C O&C BT OP 60
( = + - .W) ~. ,,3 pe INCERVICE TESTING PREGRAM UNIT - 2
@ CommonwealthQUAD CITIES NUCLEAR POWER STATION Mson ISI - CL ASE 1.2.& 3 VAI.VES SYSitM P 6 ID nfVISION - DATE PAGE azSIDUAL HEAT REMOVAL (CONTINUED) ISI-81 1 09/24/80 14 of 36 8 8 U4 vAtvE noustn
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AT RR 2 1001 34A A-2 2 A 16 CA M0 C OEC BT OP 125 AT RR 2 1001-34B B-7 2 A 16 CA MO C OEC 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 0F 2-1001-2C F-7 3 C 12 CK SA C 0 CT-1 OP 2-1001-2D F-7 3 C 12 CK SA C 0 CT-1 OP 2-1001-5A E-3 3 3 12 GL M0 C 0&C BT OP 90 2-1001-5B E-7 3 8 12 GL M0 C O&C BT OP 90 2-1001-1A G-4 3 E 14 GA N LO NA 2-1001-1B G-4 3 E 14 GA N LO NA 2-1001-1C G-6 3 E 14 CA M LO NA 2 1001-1D G-6 3 E 14 GA M LO NA 2-1001-3A G-3 3 E 12 GA M LO NA 2
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INCERVICE TEDTING PR'2 GRAM UNIT -2
@ CommonwealthQUAD CITIES NUCLEAR POWER ST ATION Edison ISl- CL ASS 1. 2. & 3 VALVES SYSTEM r lr ID REVISION - CATE PAGE RESIDUAL HEAT REMOVAL (CONTINUED) ISI-81 1 - 09/24/80 15 of 36
/
vatvE nuusin
[ f [ p 4'9 ntuanns 2-1001-3B G-3 3 E 12 GA H LO NA 2-1001-3C G-7 3 E 12 CA M LO NA 2-1001-3D G-7 3 E 12 GA H 14 NA 2-1001-201A F-3 NC E 14 BTF M- ID HA 2-1001-?918 F-7 NC E 14 BTF M lh NA 2-1001-33A B-5 3 E 16 GA M IA NA 2-1001-338 B-6 3 E 16 GA M LO HA I
I e
INCERVICE TEDTING PREGRAM UNIT - 2
@ Commonwealth Edison ISI- CLASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P 610 REVISION DATE PAGE STANDBY LIQUID CONTROL ISI-82 1 - 09/24/80 16 of 36 VALVE NUMBER
[ REMARKS 2 1101-15 C-3 1 C 1.5 CK SA C 0 CT.1 CS/RR VR-10 2 1101 16 C.3 1 C 1.5 CK SA C O CT-1 CS/RR VR-10
......... ... .....__ _ _ _ _ _ . . ......... .......~...._.. ........ ....... ........ ....... ........ ....... . . . . . . . . . . ........................._
2-1106A C-4 2 D 1.5 EKP C O [Yr RR 2-11065 D-4 2 D 1.5 EXP C O IYr RR 2-1101-43A D-6 2 C 1.5 CK SA C O CT-1 OP 2-1105A C-6 2 C 1.5 RV SA C O CT-2 RR 2-1105B D-5 2 C -
1.5 RV SA C O CT-2 RR 2-1101-4 E-8 2 E 2.5 GA M LO HA 2-1101-8 D-8 2 E 2.5 CA M LC HA 2-1101-3A D-7 2 E 2.5 CA M la NA 2-1101-35 E-7 2 E 2.5 CA M ID NA I
i 2 1101 10 D.7 2 E 1 CL M LC NA 2 1101-2A D-5 2 E 1.5 GL M LO HA E-5 2 E 1.5 GL M LO HA .
2__1101-28 2-1101 22 C-4 2 E 1.5 GL M LC HA 2-1101-98 D-4 2 E 1 GL M LC NA 2 1101-23 D-3 2 E 1.5 GL M LO HA 2-1101 1 E 1.5 GL M LO NA
........ 1'..... ...D-2 .... ....... ............... .._..... . . . . . . _ .,_....... . . . . . . ......... .....__ ...........................................
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Commonwealth INEERVICE TEOTING PREGRAM Edison ISI - CL ASS 1, 2. & 3 VALVES UNIT - 2 QUAD CITIES NUCLEAR POWER STATION SYSTEM P 6 ID nEVIS10N - DATE PAGE RX WATER CLEAN-UP ISI-88 1 - 09/24/80 17 of 36 vAtvE nousta
(( 8 (( g stunnus AT RR 2-1201-2 B-6 1 A 6 GA H0 0 C BT OF
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INSERVICE TESTING PROGRAM UNIT - 2 ISI - CL ASS 1. 2. & 3 VALVES
@ Commonwealth Edison QUAD CITIES NUCLEAR POWER STATION P & 10 REVISION - DATE PAGE SYSTIM ISI-89 1 - 09/24/80 18 of 36 REACTOR CORE ISOLATION COOLING vAtyt noustR
(( # [ p PEMARKS AT RR H0 0 C BT CS 25 GROUP 5 ISOLATION 2-1301-16 B-2 1 A 3 GA PIT RR .........................
AT RR GA HQ 0 C BT CS 25 GROUP 5 ISOLATION 3-1301-17 B-3 1 A 3 AT RR 2 CK SA C C CT-1 ' RR VR-13 2-1301-40 D-2 NC AC AT RR 8 CK SA C C CT-1 RR VR-13 2-1301-41 D-2 NC AC AT RR 0.5 XFC SA 0 C CT.1 RR VR-14 2-1301-15A B-2 1 AC AT RR 0.5 XFC SA 0 C CT-1 RR VR-14 2-1301-155 B-2 1 AC i
6 Q v) d INSERVICE TESTING PROGRAM UNIT - 2
@ CommonwealthQUAD CITIES NieCLEAR POWER STATION EJison ISI - CL ASS
- 1. 2. & 3 VALVES SYSTEM P fr ID REVIS108d - DATE PAGE CORE SPRAY l ISI-78 1 - 09/24/80 19 of 36
/ /
- REMARKS VALVE NUMBER CT-1 CS 2-1402-9A C-3 1 C 10 CK SA C O PIT RR VR-7 CT-1 CS 2-1402-95 C-4 1 C 10 CK SA C 0 PIT- RR VR-7 2-1402-25A C-2 1 B 10 CA MO C 0 BT OP 15 2-1402-25B C-5 1 B 10 CA HO C 0 BT OP 15 2-1402-24A 2 10 CA HQ O O BT OP 15
.... ....-2........
B B 2-1402-28A C-9 2 C 2 RV SA C O CT-2 RR 2-1402-28B D-6 2 C 2 RV SA C 0 CT-2 RR 2-1402-38A C-8 2 B 1.5 CA HQ O C BT OP VR-8 2-1402-388 D-7 2 8 1.5 CA HQ O C BT OP VR
.... ... .. _ _ _ _ _ _ ......___... ...... _____ ..... _... .......... ... . . . . ....._ .. ._____-8 _ .. .______________..........
2-1402-8B E-6 2 CE 12 SCK SA C/LO O CT-1 0F AT RR 2-1402-31A E-3 1 AC 0.5 XFC SA O C C'" 1 RR VR-14 AT RR 2-1402-31B E-3 1 AC 0.5 XFC SA O C CT-1 RR VR-14 2-1402-6A U-3 1 E 10 CA H LO NA 2-1402-6B D-3 1 E 10 CA H LO NA I I i s
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INCERVICE TEUTING PR2G' RAM UNIT - 2
@ Commonwealth Edison
-3.
ISI- CLASS 12.&3 YAl.VES QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 REVISION - DATE PAGE rnte spaAY (CONTINilED) 19f.78 1 09 /7/,a /An ?n nr M
/
VALVE NUM8ER 8' [ REMARKS 2 1402-2A G-7 2 E 12 CA N LC NA 2-1402-25 G-4 ,2 E 12 GA M LC NA 2 1402-34A G-I j 2 E 18 BTF M ID NA
............m...__. .s....... ........ ....... ....__.. ....... ............... .._..... ....... .................. .... ....................
2-1402-145 F-3' 2 E 18 BTF M 14 NA 2 1402-4A A-8 2 B D GL M0 C C BT OP 60
............m....... ....... ........ ....... ........ ....... ......... . . . . . . ......... ....... .................. .........................
2 1402-4B C-7 2 5 8 GL M0 C C BT OP 60 4 2-1402-13A E-9 2 CE 1.5 SCK SA/M C/li 0 CT-1
- VR-20
- SEE VR-20 2 1402-138 E-6 2 CE 1.5 SCK SA/M C/11 0 CI-1
- VR-20
- SEE VR-20 T
9
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INSERVICE TESTING PROGRAM UNIT - 2
@ Commonwealth Edison ISI- CL ASS QUAD CITIES NUCLEAR POWER STATION
- 1. 2. & 3 VALVES SYSTEM P le ID REVISION - DATE PAGE PRESSURE SUPPRESSION M-76 1 - 09/24/80 11 of 36
/
VALVE NUMBER [ [ REMLRKS AT RR
, 2-1601-21 C-6 NC A 18 BTF AO C C BT OP 10 CROUP 2 ISOLATION .
AT RR o 2 1601-22 C-6 NC A 18 BTF AO C C BT OF 10 CROUP 2 ISOLATION AT RR 2 1601-55 A-6 NC A 4 GA AO O C BT OF 10 GROUP 2 ISOLATION AT RR 2 1601-56 D6 NC A 18 BTF AO O C BT OP 10 GROUP 2 ISOLATION FST OF AT RR 2-1601-57 C-9 NC A 1 GL NO O C 3T OP 15 GROUP 2 ISOLATION AT RR
- 2-1601-58 D-7 NC A 1 GL AO C C BT OP 15 GROUP 2 ISOLATION AT RR 2-1601-59 D-7 NC A 1 GL AO O C BT OP 15 GROUP 2 ISOLATION FST OP AT - RR 2-1601 20A D-9 NC A 20 BTF AO C OEC BT CS 10 VR-23 FST CS
............~.............. ........ ...__.. ........ ....___ ........ ....... .......- ....... ........ ...................................
AT RR 2-1601-31A D-9 NC AC 20 CK SA C O&C CT-1 OP AT RR
, 2-1601 20B E-9 NC A 20 BTF AO C OLC BT CS 10 VR-23 s FST CS AT RR
! 2-1601-318 E-9 NC AC 20 CK SA C OkC CT-1 OP 4
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INSERVICE TESTING PR2 GRAM UNIT - 2
@ CommonwealthEdison ISl- CL ASS QUAD CITIES NUCLhn POWER STATION
- 1. '. &3 VALVES c.
SYSitM P 610 AEVISION - DATE PAGE PRESSURE SUPPRESSION (CONTINUED) H-76 1 - 09/24/80 22 of 36
/
VALVE NUMBER REMARKS 4
AT RR 2-1601-23 B-3 NC A 18 BTF A0 C C BT OP 10 GROUP 2 ISOLATION AT RR 2-1601-24 B-2 NC A 18 BTF A0 C C BT OF 10 GROUP 2 ISOLATION 4 AT RR 2-1601-60 5-3 NC A 18 BTF A0 C C BT OP 10 GROUP 2 ISOLATION AT RR 2 1601-61 B-2 NC A 2 GL A0 C C BT OP 15 GROUP 2 ISOLATION AT RR 2 1601-62 E-2 NC A 2 GL A0 C C BT OP 15 GROUP 2 ISOLATION AT RR
- 2-1601-63 E-2 NC A 6- BTF A0 C C BT OP 10 GROUP 2 ISOLATION AT RR '
2-8803' C-6 NC A 2 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 2-8804 D-6 NC A 2 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP
............. ....... . . . . . . ~....... ......................_. ....... _..... ....................... .......... .........................
AT RR 2-8801A C-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION i FST OP AT RR 2-8801B D-3 NC A 0.5 GL A0 0 C BT OP 10 GROUP 2 ISOLATION FST OP I
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INSERVICE TESTING PROGRAM UNIT - 2
@ Commonwealth Edison ISI- CLASS QUAD CITIES NUCLEAR PC NER STATION
- 1. 2. A 3 VALVES SYSTEM PfrID REVISION -
DATE PAGE PRESSURE SUPPRESSION (CONTINUED) M.76 1 - 09/24/80 23 of 36
/
VALVE NUMBER [ REMARKS AT .RR
- 2-8801C D-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 2-8801D E.3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP '
AT RR 2-8802A C-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OF AT RR 2-88028 D-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION
.l FST OP i ..... ......- .............. ....._ ......__ ......... . . . . . . .......... . . . . . . ......... . . . . . . . ..............._ .. .........................
AT RR
= 2-8802C D-3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION FST OP AT RR 2-8802D E.3 NC A 0.5 GL AO O C BT OP 10 GROUP 2 ISOLATION 1 FST OP 2-1601-32A E-2 NC C 18 CK SA C OLC CT.1 OP 2 1601-32B E2 NC C 18 CK SA C OLC CT.1 OP E.2 NC C 18 CK SA C OkC CT-1 OP j 2-1601_32C i 2-1601-32D E-2 NC C 18 CK SA C OkC CT.1 OP 2-1601-32E E-2 NC C 18 CK SA C OLC CT-1 OP 2-1601-32F E-2 NC C 18 CK SA C OkC CT-1 OP
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INCERVICE TESTING PREGRAM UNIT - 2
@' Commonwealth Edison ISI - CL A SS QUAD ClilE S NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P 6 ID REVISION - DATE PAGE PRESSURE SUPPRESSION (CONTINUED) N-76 1 - 09/24/80 24 of 36 VALVE NUMBER [ REMARKS 2-1601-33A E-7 NC C 18 CK SA C O&C CT-1 OP 2-1601-338 E-7 NC C 18 CK SA C OGC CT-1 OF 2-1601-33C E-7 NC C 18 CK SA C OEC CT-1 OP 2 1601-33D E-7 NC C 18 CK SA C OEC CT-1 OP 2-1601-33E E-7 NC C 18 CK SA C OLC CT-1 OP
............. ....... . . . _ . . . ......... .__....~............... ....... . . . . . . . ..........__. . ....... .................__.................
i 2-1601-33F E-7 NC C 18 CK SA C OkC CT-1 OF 2-220-81A E-4 NC C 1 CK SA C 0 CT-1 CS VR-24 2-220-815 E-4 NC C 1 CK SA C 0 CT-1 CS VR-24 2-220-81C E-4 NC C 1 CK SA C O CT-1 CS VR-24 2-220-81D E-5 NC C 1 CK SA C O CT-1 CS VR-24
, 2-220-81E E-5 NC C 1 CK SA C 0 CT-1 CS VR-24
............. ........ . . . . . . .,..............n........ . . . . . . . ....... . . . . . . . . ........ . . . . . . ......... .......... .........................
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\, 3 I
UNIT - 2
@ Commonwealth Edison ISI - CL ASS 1.2.& 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM P 610 A! VISION - DATE PAGE umn percenter mni Aut T u t ri.rf nN YRY_A7 1 09/24/80 2% of 36 valve NUh8ER REMARKS A? RR 2 2301-4 C-9 1 A 10 CA M0 O OLC BT CS 50 VR-15 GROUP 4 ISOLATION PIT RR AT RR 2 2301-5 B-10 1 A 10 GA HQ O OEC BT CS 50 VR-15 GROUP 4 ISOLATION 2-2301-3 A-6 2 5 10 GA HQ C O BT OP 25 2-2301-68 A-6 2 D 16 RPD SA C O *
- RPD NOT TESTABLE 2-2301-69 A-6 2 D 16 RPD SA C 0
- VERIFIED OPEN 2-2301-45 B-8 2 AC 24 CK SA C OEC CT-1 OP/RR* VR.13 DURING PUMP TEST 2-2301-35 E-7 2 B 16 CA HQ C OkC BT OF 120 2-2301-36 E-9 2 B 16 GA H0 C OEC BT OP 120
............~....... ....... ........ ....... ........ ....... ........ ...___. .._..... ....... ........ ...... ............................
I 2-2301-6 F-2 2 B 16 CA HO 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 8 4 GL HO C OLC BT OP VR-B 2-2301-39 E-8 2 C 16 CK SA C O CT-1
- VR-12
- SEE VR-12 i
J
~ -
.J INSERVICE TESTING PRECRAM
@ Commonwealth Edison ISI- CLASS QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES UNIT - 2 SYSitM P 610 REVISION - DATE PAGE HICH PHMMURE COOLANT INJECTION (CONTINUED) 191-87 1 - 09/24/80 26 of 36 VALVI NUM8tR [ [ REMARKS 2-2301-8 D-6 2 B 14 CA M0 C 0 BT OP 45 PIT RR 2-2301-7 D-6 2 C 14 CK SA C 0 CT-1 CS VR-7 2-2301-74 B-8 2 CE 12 SCK SA C/14 0 CT-1 OP*
- VERIFIED OPEN DURING PUMP TEST AT RR 2-2301 26 D-9 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-2301-27 D-9 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14
. 2-2301-22 3-1 2 E 16 CA H 14 NA 2-2301-56 F-8 2 E 16 BTF M LO NA 2-2301-71 D-? 2 Cli 2 SCK SA/M C/I4 0 CT-1 OP*
- VERIFIED OPEN DURING PUMP TEST 2-2301-9 D-5 2 B 14 GA MO O O BT OP 45
............. .............. ............... ....e... . . . . . . . .................,_...... ...__.. ...__. .... ...... ._. ...........___.......
2-2301-10 E-5 2 B 12 GL HO C C BT OP 60 2-2301-40 D-7 NC C 4 CK SA C- O CT-1
- VR-20
- SEE VR-20 T
e
(. . . \ 9 ,
INZERVICE TESTING PR'sGRAM UNIT - 2
@ CommonwealthQUAD CITIES NUCLEAR POWER STATluN Edison ISI- CLASS 1. 2. & 3 VALVES i
SYSTEM P 610 REVISION - DATE PAGE NAIN STEAM ISI-60 Sh. I 1 - 09/24/80 27 of 36 l
VALVE NUMBER [ [ REMARKS I AT RR BTP OP l BT CS 5 VR-1 GROUP 1 ISOLATION l 2 203-1A F-4 1 A 20 CL A0 0 C FST CS VR-1 l PIT RR 1
............ .......- ....... ....-_.. ....... ........ ....... ........ ....... . . . . . . . . ....... ........ ...__............................. l AT RR l BTP OP l BT CS 5 Vn-1 GROUP 1 ISOLATION I 2-203-1B D-4 1 A 20 GL A0 0 C FST CS VR-1 l PIT RR AT RR BTP OP
- BT CS 5 VR-1 GROUP 1 ISOLATION 2-203-1C C.4 1 A 20 CL A0 0 C FST CS VR-1 PIT RR AT RR BTP OP BT CS 5 VR-1 GROUP 1 ISOLATION 2-203-1D B-4 1 A 20 CL A0 0 C FST CS VR-1 PIT RR
............... ........ . . . . . . . ....... ........ ....... .......~....... ... .... ....__.... .__......................
AT RR 2-220-1 E-4 1 A 3 CA HQ C C BT OP 35 CROUP 1 ISOLATION BT *
- VR-2
- SEE VR-2 2-203-3A F-4 1 BC 6 ERV/LV PS/S4 C 0 CT-2 RR VR-3 BT VR-2 2-203-35 D-6 1 BC 6 ERV PS C 0 CT-2 RR VR-3 BT VR-2 2-203-3C C-7 1 BC 6 ELV PS C 0 CT-2 RR VR-3 9
f 3 ) '
INDERVICE TEDTING PREGRAM UNIT - 2
@ Commonwealth Edison ISI - CL A SS QUAD CITIES NUCLEAR POWER STATION 1,2.& 3 VALVES SYSTIM f (r 10 REVISION - DATE PAGE j HAIN STEAM (CONTINUED) ISI-60 Sh. 1 p 1 - 09/24/80 28 of 36 )
l VALVE NUMBER [ [ REMARKS BT *
- SEE VR-2 BT *
- VR-2 2-203-3E D-7 1 BC 6 ERV PS C 0 CT-2 RR VR-3 2-203-4A F-8 1 C 6 SV SA C 0 CT.2 RA VR-26 2-203-4B D5 1 C 6 SV SA C 0 CT.2 RR VR-26 2-203-4D B-5 1 C 6 SV SA C 0 CT-2 RR VR-26 2-203-4E F-8 1 C 6 SV SA C 0 CT-2 RR VR-26 ,
.......... ......................... 1 I
2-203-4F D-5 1 C 6 SV SA C O CT-2 RR VR-26
. . . . . . . ................. ....... ........ ............... ....... .......... ......................... j 2-203-4G C-5 1 C 6 SV SA C 0 CT-2 RR VR-26 2-203-4H B-5 1 C 6 SV SA C 0 CT-2 RR VR-26 1
i l
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/. ,. i l INSERVICE TE^; TING PROGRAM UNIT - 2
@ Commonwealth Edison ISI - CL ASS 1. 2. & 3 VALVES OUAD CITIES NUCLEAR POWER STATION SYsitM P 6 ID REVISION - DATE PAGE MAIN STEAM (CONTINUED) ISI-60 Sh. 2 , 1 - 09/24/80 29 of 36 WALVI NUMBER REMARKS AT RR
- BTP OP BT CS 5 VR-1 2-203-2A E-7 1 A 20 GL A0 0 C FST CS VR-1 GROUP 1 ISOLATION FIT RR AT RR BTP OP BT CS 5 VR-1 2-203-28 E-7 1 A 20 GL A0 0 C FST CS VR-1 GR0(JP 1 ISOLATION PIT RR AT RR BTP OP BT CS 5 VR-1 2-203-2C D-7 1 A 20 GL A0 0 C FST CS VR-1 CROUP 1 ISOLATION PIT RR AT RR BTP OP BT CS 5 VR-1 2-203-2D B-7 1 A 20 GL A0 0 C FST CS VR-1 GROUP 1 ISOLATION PIT RR AT RR 2-220 2 E-7 1 A 3 CA NO C C BT OP 35 GROUP 1 ISOLATION AT RR 2-220-17A E-8 1 AC 0.5 KFC SA 0 C CT-1 RR VR-14 AT RR 2-220-178 D-8 1 AC 0.5 KFC SA 0 C CT-1 RR VR-14 AT RR 2-220-17C C-8 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14 AT RR 2-220-17D B ,8 1 AC 0.5 XFC SA 0 C CT-1 RR VR-14
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INSERVICE TEDTING PROGnAM UNIT - 2
@ CommenmalthQUAD C1YlES NUCLEAR POWER STATION Edson 'ISI - CL ASS 1.2.& 3 VALVES SYSTEM . P fr 10 REVISION - DATE PAGE FEEDWATER ISI-62 1 - 09/24/80 31 of 36 VALVE NUMEIR [ REMARKS l AT RR l m
! 2-220-58A E-3 1 AC 18 SA 0 C Cr-1 RR VR-4 AT RR 2 220-585 F-3 1 AC 18 m SA 0 OEC CT-1 RR VR-4
............. ....... . . . . . . . ......... .......~........ . . . . . . . ....... . . . . . . . . ............... ....... .......... .........................
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INSERVICE TEDTING PWi2 RAM UNIT - 2
@ Commonwealth E& son ISI- CLASS 1. - 2. & 3 VALVES QUAD CITIES NIJC LE A R POWER STATION SY3i!M P 610 REVISION - DATE PAGE DEESEL GENERATOR STARTING AIR (SERVICE, AIR SYSTEM) M. 72 1 - 9/24/80 33 of 36
/
I VALVE NUMBER [' [ REMARKS ,
2-4699-121 E-9 NC E 1.5 GA N LD NA 2-4699 122 E-9 NC E 1.5 CA N ID NA 2.t699 225 D-8 NC E 1.5 l BAI.L
.......s.......
M 14 NA 2-4599-123 E-9 NC C 1.5 CK SA C 0 CT-1 OP .
2-4699-196 E-9 NC C 1.5 CK SA C 0 CT-1 OP 2-4699 226 D-8 NC B 1.5 GL AO C 0 BT OP 5 s
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INSERVICE TESTING PROGRAM UNIT - 2 ISI- CLASS 1. 2. & 3 VALVES
@ commenmann E& son OUAD CITIES NUCLEAR POWER STATION Pfr10 REVISION - DATE PAGE SYSTEM INSTRUMENT AIR PIPING N-71 Sh. 2 i 1 - 09/24/80 34 of 36 VMVE NUM4(R [ REMARKS I .
AT RR i A GA AO O C BT OP 10 GROUP 2 ISOLATION l 2-4720 D-3 NC 1 FST OP AT RR A GA AO O C BT OP 10 GROUP 1 ISOLATION i 2-4721 D-3 NC 1 FST OP I
....... . . . . . . . ..........___.......... ..... .... ......................... I AT RR A 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION 2-733-1 F7 NC AT RR 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION 2-733-2 F-7 NC A AT RR 0.375 BALL C C BT OP 5 GROUP 2 ISOLATION 2-733-3 P-7 NC A_ SO AT RR ,
0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION 2-733-4 F-7 NC A AT RR 0.375 BALL SO C C BT OP 5 GROUP 2 ISOLATION 2-733-5 F-7 NC A AT RR B-7 0.375 CK SA O C CT-1 RR VR-5 2-743 NC AC 2-736-2 F-7 NC D 0.375 EXP O C DT RR 2 736-3 P-7 NC D 0.375 EXP O C [rf RR 2-736-4 P-7 NC D 0.375 EXP O C frf RR 2-736-5 F.7 NC D 0.375 EXP O C DT RR
I V .)
INSERVICE TEOTING PROGRAM UNIT - 2
@ CommonwealthEdison ISI- CL ASS 1. 2. & 3 VALVES QUAD CITIES NUCLEAR POWER STATION SYSTEM P610 AEVISION - DATE PAGE RX BUILDING EQUIPNENT DRAINS M-85 1 09/24/80 35 of 36 1
VALVE IHiMSER [ REMARKS 1 /
l AT RR l 2-2001-15 E-3 NC A 3 CA AD C C BT OP 20 GROUP 2 ISOLATION FST OF AT RR 2 2001 16 E-3 NC A 3 CA AD C C BT OF 20 GROUP 2 ISOL.iTION l FST OP AT RR !
2-2001 3 F-7 NC A 3 GA AD C C BT OP 20 GROUP 2 ISOLATION l
- FST OP AT RR 2 2001-4 F-7 NC A 3 CA A0 C C BT OP 20 GROci 2 ISOLATION !
FST OP I G
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INSERVICE TESTING PROGRAM UNIT - 2
@ Commonwealth Edison ISI - CL A S S QUAD CITIES NUCLEAR POWER STATION 1.2.& 3 VALVES SYSTEM P(RID REVISION - DATE PAGE DIESEL GENERATOR FUEL OIL M-29 1 - 09/24/80 36 of 36 VALVE bum 8ER d* REMARKS 2-5299-5 E-4 NC C 1.5 CK SA C 0 CT-1 OP 2-5201 E-3 NC B 1 GA SO C 0 BT OP 5 2-5199-155 5-5 NC E 1 CL M LO NA 2-5199-157 C-5 NC C 0.5 CK SA C C CT-1 OP D
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SECTI0i! L1.3 RELIEF REQlJESTS FOR INSERVICE VALVE TESTING PROGRAF 1
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Revision 1 9/24/30
g RELIEF REQUEST NO. VR-1 t.
l SYSTEM: Main Steam COMPONENT: 1(2)-203-1A, B, C, D 1(2)-203-2A, B, C, D
~
CATEGORY: A
_ FUNCTION: Primary containment isolation valves for the main steam lines.
TEST REQUIREMENT: BT -
Exercisa and time valves every three
] months.
FST -
Check the fail-safe operation of the valves upon loss of actuator power every three months.
BASES FOR RELIEF: Full stroke testing these valves during normal reactor operation requires isolating one of the four main steam lines. Isolation of these lines results in primary system pressure spikes, reactor power fluctuations, and increased 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 4-11 Revision 1 9/24/80
A g- RELIEF REOJEST NO. VR-1 (CONTINUED)
-testing MSIVs increase the chances of actuating primary system relief valves.
It.is proposed that only partial stroke testing be performed during power operation and that full stroke testing be performed at cold shutdowns.
These valves are provided with the circuitry to
- . permit partial stroking to a 10% closed position.
This partial stroke exercising provides an accept-able means of verifying vaJ va performance during 4 plant operation without affecting safety margins.
This request also contributes to the reduction of
}
the relief valves cha11ange rate as recommended in NUREG-0626.
ALTERNATE TESTING: These valves will be part stroke exercised' every three months and full stroke exercised during cold shutdown. The fail-safe operation of these valves will also be checked during cold shutdown since this is done coincident with full stroke exercising. The fail-safe testing of valves 1(2)-
203-1A,B,C and D however, will be completed only at cold shutdowns when that primary containment is de-inerted since access to the valves to perform this u testing requires entry into the drywell.
. 4-12 Revision 1 9/24/80
RELIEF REQUEST NO. VR-2 -
. SYSTEM: Main Steam COMPONENT: 1(2)-203-3A, B, C, D, E 3A-Target Rock Safety Relief Valve 3B-E-Electromatic Relief Valves.
CATEGORY: B/C .
' FUNCTION: 1) Open upon receipt of an auto depressurization signal- to blowdown reactor: 2) Act as a primary system relief valve which actuates .on high system pressure.
.t -
TEST REQUIREMENT: BT - Exercise and time valves every three months.
BASIS FOR RELIEF: 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
, primary system resulting in undesirable thermal F gradients in the reactor vessel. Excessive testing 4-13 Revision 1 9/24/80
. RELIEF REQUEST NC. VR-2 (CONTINUED)
- c 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 thatLsubmitted by Quad Cities Station in a proposed Technical Specification . change required by the August 3, 1977 letter from Don K. Davis (liRC-DOR) to Commonwealth Edison Company. In this Technical Specification change, a program was committed to which specified a variable testing frequency related to demonstrated reliability and operability. The testing interval is based on the number of valve failures during the required test interval. The 4-14 Revision 1 9/24/80
7 ;
i RELIEF REQUEST NO. VR-2 (CONTINUED) 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 0 18 months 1 25%
1 184 days i 25%
2 92 days 1 25%
> 3 31 days 1 25%
Additionally, stroke times for these valves will not be measured since there is no position indication circuitry to show disc movement.
4-15 Revision 1 9/24/80 L -__
l-RELIEF REQUEST NO. VR-3 SYSTEM: Main Steam COMPONENT: 1(2)-203-3A (Target Rock Safety / Relief) i 1(2)-203-3B,C,D,E (Electromatic Relief)
CATEGORY: BC FUNCTION: 1) Open upon receipt 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 n .
Valve functions the same as above except, it also acts as a safety valve.
TEST REQUIREMENT: CT-2 - Verify pressure set point in accor-dance with IWV-3510.
BASIS FOR RELIEF: The electromatic relief valves and the relief function of the Target Rock valve are operated by actuation of a pilot solenoid valve which opens th e main valve by applying system pressure to a diaphragm. The pilot valve is actuated from an electric signal from either the control switch, the auto-depressurization logic, cr a pressure switch q
that senses system pressure.
4-16 Revision 1 9/24/80
\
RELIEF REQUEST NO. VR-3 (CONTINUED)
The requirement of IW-3 510 (b) to check relief and safety valve set points in accordance with PTC-25.2-1966 is not applicable in this case. Therefore, relief is requested from compliance with this require' ment.
The pressure set point of these valves is set by 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 i paragraph IW-3510.
ALTERNATE TESTING: 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.
4-17 Revision 1 9/24/80
- - RELIEF REQUEST NO. VR SYSTEM
- FEEDWATER COMPONENT: 1(2)-220-58A, B 1(2)-220-59B 1(2)-220-62A, B CATEGORY: C & AC FUNCTION: The 58 and 62 valves close for containment isola-
~
tion. The.59B valve closes for HPCI injection.
TEST REQUIREMENT: CT-1 -
Exercise check valve every th ree months.
BASIS FOR RELIEF: These check valves cannot be tested for oper-ability during reactor operation because the feed-water 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 r eg' aired to be operable. In addition, to verify that ' these check valves stroke to the full closed positien, a leak rate test must be performed. Since
,.4 leak rate testing is performed only daring ref.seling 4-18 Re*fgog0
4/
9
- +$g&@/ IMAGE EVALUATION
////
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% TEST TARGET (MT-3) 4 k//77g, s<.
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1.0 'd m En 5E E I.I
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[" lE lJi-1.25 1.4 1.6 4 150mm >
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RELIEF REQUEST NO. VR-4 (CONTINUED) i outages, these valves will be demonstrated to be in the full closed position at each refueling outage.
ALTERNATE TESTING: These check valves will be exercised closed during each reactor refueling outage.
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4.
4-19 Revision 1 9/24/80
. - --- - .. . - ~ - . . . - . . . - - _ . - - - - . , . - . . _ . _ _ _ - . . . - - . - .
~
RELIEF REQUEST NO. VR-5
~
SYSTEM: ~ Neutron Monitoring System COMPONENT: 1(2)-743 CATEGORY: C FUNCTION: Primary containment isolation valve for the T.I.P.
System nitrogen purge line.
TEST REQUIREMENT: CT Exercise valve every three months.
BASIS FOR RELIEF: This check valve cannot be exercised for 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 ou -
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 s
each refueling outage.
ALTERNATE TESTING: The valves will be full stroke exercised each refueling outage.
t
- 4-20 Revision 1 9/24/80
- _ . _ . ~ . _ . _ - _ . _ . _ _ . . _ - _ _ _ _ _ _ . _ _ _ _ _ . _ _ . _
1 RELIEF REQUEST NO. VR-6 i
SYSTEM: Recirculation COMPONENT: 1(2)-202-5A, B CATEGORY: B FUNCTION: In a design basis loss of coolant accident, one of these valves will close depending on the location of the line break.
TEST REQUIREMENT: BT - Exercise and time valves for operability every three months.
t BASIS FOR RELIEF: 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.
ALTERNATE TESTING: These valves will be full stroke exercised during cold shutdown.
l'
~
4-21 Revision 1 9/24/80
RELIEF REQUEST NO. VR-7 -
~
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 CATEGORY: C FUNCTION: Open upon System Injection 4
TEST REQUIREMENT: CT-1-Exercise valve every three months.
BASIS FOR RELIEF: These valves have air-operators and remote position indicators for remote testing purposes.
However, during normal operation the high differ-ential pressure across the valve seats prohibits exercising. Additionally, the residual heat removal and_ core spray system valves (ie.,1(2)-1001-68A, 683
& 1(2)-1402-9A,98) are located inside the primary containment which is inerted with nitregen during normal operation. The high pressure cociant injection valve (l(2)-2301-7) is located inside the main steam isolation valve room 4-22 Revision 1 9/24/80
i
, RELIEF REQUEST NO. VR-7 (CONTINUED) 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 area (120* to 140*F) which further increases the hazards involved 'n i entering the area for this testing. The accumulated dose to conduct this test would be approximately 1.5 man-rem.
ALTERNATE TESTING: These valves will be full stroke exercised during cold shutdown.
t t
a 4-23 Re"91in:iaio:01 1
p RELIEF REQUEST NO. VR-8 -,
SYSTEM: Core Spray, Residual Heat Removal, and High Pressure Coolant Injection.
COMPONENT: 1(2)-1402-38 A&B 1(2)-1001-18 A&B 1(2)-2301-14 CATEGORY: B FUNCT~ON: The valves close when pump flow is adequate (i.e.,
minimum flow recirculation valves).
TEST REQUIREMENT: Exercise and time valves every three months.
BASIS FOR RELIEF: Relief is requested from measuring the stroke time of these valves. Since the valves close auto-matically when adequate pump flow is reached, it is difficult to accurately measure the stroke time. An equally meaningful test would be to just verify that the valves do close automatically as the pump flow increases.
ALTERNATE TESTING: Operator will verify that thes e valves auto-matically close as pump flow increases during quar-terly pump operability test.
4-24 Revision 1 9/24/80
I l
RELIEF REQUEST NO. VR-9 a l l
. SYSTEM: RHR-Head Spray / Shutdown Cooling Subsystems COMPONENT: 1(2)-1001-60, 63, 47, 50 CATEGORY: A FUNCTION: Primary' containment isolation valves for RHR Head Spray and Shutdown Cooling Subsystems.
TEST REQUIREMENT: BT - Exercise and time valves every 3 months BASIS FOR RELIEF: Relief is requested from partial or full stroke testing these valves during operation. These valves, which are normally closed during plant oper-ation, serve as isolation between the high and low
> pressure piping. Protective interlocks prevent opening these valves while the reactor is at oper-ating pressure.
ALTERNATE TESTING: The valves will be exercised during cold shutdown.
+
4-25 Revisign 1 9/ 24,80
RELIEF REQUEST NO. VR-10 .
SYSTEM: Standby Liquid Control COMPONENT: 1(2)-1101-15, 16 CATEGORY: C s
. FUNCTION: The safety function of these check valves is to ,open upon a system injection.
TEST REQUIREMENT: CT Exercise valve every three months.
BASIS FOR RELIEF: 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 described above, only 26 gpm of the required 39 can 4-26 Revision 1 9/24/80
RELIEF REQUEST NO. VR-10 (CONTINUED) be injected. The station is confident that during refueling outages a method can be devised to obtain full flow through' these valves. If necessary, although not desired, the system pumps could be used to provide this flow, but this would require exten-sive cleaning of the system to remove residual boron.
ALTERNATE TESTING: These valves will be part stroke exercised during cold shutdown and full stroke exercised at each refueling outage.
4-27 Revieion 1
- 9/24/80
RELIEF REQUEST NO. VR-ll
' SYSTEM:' Control Rod Drive i
COMPONENT: 1(2)-0305-126, 127, 114 CATEGORY: B&C
-FUNCTION: These valves operate on a scram signal to drive the control rods in.
TEST REQUIREMENT: BT - Exercise and time valves every 3 months.
CT Exercise valves every three' months.
4 BASIS FOR RELIEF: There are 177 of each 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 e'ach 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.
4-28 Revision 1 9/24/80
RELIEF REQUEST NO. VR-ll (CONTINUED)
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ALTERNATE TESTING: Individual scram insert.on i 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.
i 4-29 Revision 1 9/24/80
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RELIEF REQUEST NO. VR-12 -
SYSTEM: High Pressure Coolant Injection COMPONENT: 1(2) - 2301-39 CATEGORY: C FUNCTION: See Basis for Relief i
TEST REQUIREMENT: CT-1 Exercise check valve every three months.
BASIS FOR RELIEF: This valve is designed to prevent backflow 3
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.
4-30 Revision 1 9/24/80
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RELIEF REQUEST NO. VR-12 (CONTINUED) 1 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.
- g. It should also be mentioned that an evaluation is being conducted to determine if these valves can be removed from the system. If this proves to be feasible, the valve internals will be removed and this relief request would no longer be required.
ALTERNATE TESTING: Each valve will be disassembled every third refueling outage to verify that the disc swings freely to the full open position.
s 4-31 Revision 1 9/24/80
u RELIEF REQUEST NO. VR-13 _
L SYSTEM: High Pressure Coolant Injection, Reactor Core Isolation Cooling COMPONENT: 1(2)-2301-34, 45, 1(2)-1301-40, 41 CATEGORY: C FUNCTION: Primary containment isolation.
TEST REQUIREMENT: CT-1 - Exercise valve for operability every three months.
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. BASIS FOR R2 LIEF: It is impractical to demonstrate closure of these check valves during normal operation or cold shutdown. To verify closure upon reversal of flow a pressure test must be performed. This requires that the systems be taken out-of-service. The safety significance of these components is minimal since leakage past these valves would be contained within the HPCI and RCIC piping which returns to the containment.
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I 4-32 Revision 1 l- 9/24/80 l
f RELIEF REQUEST NO. VR-13 (CONTINUED)
ALTERNATE TESTING: These valves will be demonstrated to close upon reversal of flow during each refueling outage
- per Appendix J test.
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4-33 Revision 1 9/24/80
l RELIEF REQUEST NO. VR i SYSTEM: Nuclear Boiler Instrumentation, Recirculation, Reactor. Core Isolation Cooli ng, Core Spray, High Pressure Coolant Injection, Main Steam.
COMPONENT: Excess flow check valves as listed in program.
CATEGORY: 'AC FUNCTION: Limit flow (leakage) from instrument lines penetrating primary containment; perform containment isolation function.
TEST REQUIREMENT: AT - Seat leak rate test.-
CT-1 Exercise check valves to the closed position every three months.
BASIS FOR RELIEF: These valves are currently tested per
-Technical Specification requirements ' which consists of a leakage test conducted during , primary system 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.
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. 4-34 Revision 1 9/24/80
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RELIEF REQUEST NO. VR-14 (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.
ALTERNATE TESTING: These valves will be tested in' the manner described above prior to start-up from each refueling outage.
>b 4-35 Revision 1 9/24/80
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FE RELIEF REQUEST NO. VR-15 -
, i SYSTEM: High Pressure Coolant Injection.
COMPONENT: 1(2)-2301-4&5.
CATEGORY: A FUNCTION: Primary containment isolation.
TEST REQUIREMENTS: 'BT-Exercise valve for operability every three months.
BASIS FOR RELIEF: The above valves are normally open to supply steam to the turbina 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 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.
4-36 Revision 1 9/24/80
h RELIEF REQUEST NO. VR-15 (CONTINUED)
L ALTERNATIVE TEST: These valves will . be full stroke exercised during cold shutdowns.
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I 4-37 Revision 1 .
9/24/80 :
RELIEF REQUEST NO. VR-16 SYSTEM: All Systems COMPONENT: All primary containment isolation valves (listed in program as Category A).
CATEGORY: A FUNCTION: Primary containment isolation.
TECT REQUIREMENT: AT - seat leakage tests per IWV-3420.
BASIS FOR RELIEF: 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 requirements of 10CFR50.
At this functional differ-ential 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: Valves will be seat leak tested in accordance with 10CFR50 Appendix J.
4-38 Revision 1 9/24/80
x RELIEF REQUEST NO. VR-17 Specific relief is requested from requirements of paragraphs IW-3410(g) and IW-3520(c) of Section XI of the 1974 Edition of the ASME. Boiler and Pressure Vessel Code including the Addenda through Summer 1975. 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 within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
- i. These paragraphs do not take into account the plant Technical 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-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.
For these reasons, Quad Cities Station will evaluate the condi-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.
4-39 Revision 1 9/24/80
i-i RELIEF REQUEST NO. VR-18 i.
k SYSTEM: All Systems COMPONENT: All power operated valver requiring full stroke timing tests.
CATEGORY: A and B FUNCTION: Power operated valves.
TEST REQUIREMENT: Stroke time accuracy per IWV-3410(c)(2)
BASIS FOR RELIEF: The code requires that. stroke timing accuracy be either to the nearest second or 10 percent of the maximum stroke time, whichever is less. Quad cities Station feels that this requirement is impractical for valves with stroke times less than 10 seconds.
For these valves, the accuracy of the measured stroke time is less than one second which is incon-sistent with the method used for timing (i.e., hand i.
held stopwatch) and difficult to certify. Since the purpose for the test is to detect deterioration of
- valve function, establishing a lower limit of one second for valve timing accuracy will have no effect 4-40 Revision 1 9/24/80 L-.--_______.________________._____-_--. . - _ _ - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
RELIEF REQUEST NO. VR-18 (CONTINUED) on the ability to detect changes in valve perfor-mance. Recent changes in Section XI of the ASME Code have reflected this position.
ALTERNATE TESTING: As revised in later editions of the Code, the stroke time of all power operated valves will be measured to the nearest second, for stroke times 10 seconds or less, or 10 percent of the specified limiting stroke time for full-stroke times longer than 10 seconds.
s 4-41 Revision 1 9/24/80
.r RELIEF REQUEST NO. VR-19 -
SYSTEM: Control Rod Drive COMPONENT: 1(2) - 0302-21 A & B 1(2) - 0302-22 CATEGORY: B FUNCTION: Scram discharge volume vent and drain valves.
TEST REQUIREMENT: BT - Exercise and time every 3 months.
FST - verify fail-safe operation upon loss of actuator power every 3 months.
BASIS FOR RELIEF : 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 suf ficient volume is always available to accept 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 4-42 Revision 1 9/24/80
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RELIEF REQUEST NO. VR-19 (CONTINUED)
NRC Staff guidelines concerning the cycling of valves that could potentially place the plant in an unsafe mode of operation, it is felt that these valves should be tested at cold shutdown. This applies to the vent and drain Valves since the air supply to the valves and the test circuit is common for all three valves.
In addition, the system is designed such that the test circuit bleeds the air from these air-operated valves at a very slow rater much slower than during normal operation of the valve. Timing these valves during testing, therefore, has no relevance, and because of the slow bleed rate the test time repeatability is poor.
ALTERNATE TESTING: These valves will be full stroke exercised without timing during each cold shutdown.
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RELIEF REQUEST NO. VR-20 SYSTEM: Core . Spray, Residual Heat Removal, High Pressure Coolant Injection COMPONENT: 1(2)-1402-13A&B 1(2)-1001-142A,B,C,&D
.1(2)-2301-40 CATEGORY: C & CE FUNCTION: Pump minimum flow line check and stop-check valves required to open during pump low flow conditions for pump cooling.
TEST REQUIREMENT: CT-1 Exercise valve every three months.
BASIS FOR RELIEF: There are no provisions in the current system design for exercising or determining the position of these valves. Based on the record 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 the inability to demonstrate that the valves stroke open, Quad cities Station has initiated sys t e:n modifications to install flow instrumentation in the 4-44 Rev'.sion 1 9/24/80
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RELIEF REQUEST NO. VR-20 (CONTINUED) 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 will be out of service during the installation. The modifi-cations will be implemented at the earliest possible date which is contingent on the availability of-materials. Relief is therefore requested from the requirement to demonstrate that the subject valves i
stroke open until system modifications provide the necessary instrumentation.
f ALTERNATE TESTING: No specific alternate test is applicable during this interim period.
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l 4-45 Revision 1 9/24/80 i
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RELIEF REQUEST NO. VR-21 SYSTEM: Residual Heat Removal COMPONENT: 1(2)-1001-19A & B CATEGORY: B FUNCTION: RHR System cross-tie line isolation valves.
TEST REQUIREMENTS: BT-Exercise and time in the open direction every three months.
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- - SASIS FOR RELIEF
- These valves are normally in their safety position (open) and are only closed a very small percentage of plant operating time when the system is in the residual heat removal mode. Testing these valves during normal operation places the plant in an unsafe mode because a failure of either valve in the closed position renders the low pressure coolant 1njection (LPCI) function inoperable. The LPCI function of RH3 is designed such that three of the four pumps are required to provide makeup flow to either recirculation loop in the event of a design basis loss of coolant accident. Tnis requires the a
4-46 Revision 1 9/24/80
RELIEF REQUEST NO. VR-21 (CONTINUED) 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, these valves will be exer-cised during cold shutdown conditions.
ALTERNATE TESTING: These valves will be exercised and timed during cold shutdowns.
i 4-47 Revision 1 9/24/80
RELIEF REQUEST NO. VR-?. 2 SYSTEM: Residual Heat Removal COMPONENT: 1(2)-1001-28 A&B 1(2)-1001-29A&B CATEGORY: B,A FUNCTION: LPCI injection valves; primary containment isolation (29A&B); pressure isolation.
-TEST REQUIREMENT: BT -
Exercise and stroke time every three months.
BASIS FOR RELIEF: Relief is requested from exercising these valvas 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.
4-48 Revision 1 9/24/80
l-RELIEF REQUEST NO. VR-22 (CONTINUED)
To ensure that valve exercising procedures do not place the plant in an unsafe mode of operation, these valves will be full stroke exercised only at cold shutdowns.
ALTERNATE TESTING: These valves will be full stroke exercised at cold shutdowns.
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L 4-49 Revision 1 l 9/24/80 l
RELIEF REQUEST NO. VR-23 l
SYSTEM: Pressure Suppression COMPONENT: 1(2)-1601-20A&B CATEGORY: A FUt4CTION: Reactor building to torus vacuum breaker isolation valves and primary containment isolation.
TEST REQUIREMENT:- BT - ' Exercise and time valves in both the open and closed direction every three months.
t BASIS FOR RELIEF: Exercising these valves open during normal plant operation compromises primary containment integrity and reduces safety margins by leaving only a single check valve (1601 - 31 A or B) to maintain l-the primary containment boundary.
ALTERNATE TESTING: These valves will be full stroke exercised during. cold shutdown.
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4-50 Revision 1 9/24/80 L
RELIEF REQUEST NO. VR-24 --
SYSTEM: Main Steam COMPONENT: 1(2)-220-81A,B,C,D & E CATEGORY: C.
FUNCTION: Vacuum breakers for the main steam relief valve discharge lines.
TEST REQUIREMENT: CT-1 Excercise check valve in the open direction every three months.
, BASIS FOR RELIEF: These check valves have no external means of actuation for exercising. The only practical method for exercising these valves open is by manually pusl$ing 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.
71 TERNATE TESTING: These check valves will be verified to freely swing to their full open position at cold shutdowns when the drywell is de-inerted.
I 4-51 Revision 1 9/24/80 i.
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RELIEF REQUEST NO. VR-25
. SYSTEM:' All Systems
. COMPONENTS: All power operated valves requiring full , stroke testing.
CATEGORY: A and B FUNCTION: Power operated valves TEST REQUIREMENT: Stroke time evaluation per IWV-3410(c) (3).
BASIS- FOR RELIEF: Paragraph IWV-3410(c) (3) requires that valve stroke times be evaluated against the previous stroke time to determine if corrective action is required. To establish consistency in evaluating stroke times and make program implementation more practical, Quad Cities Station proposes to establish
^a reference stroke time for each valve which will be used for evaluating performance. This reference value will be determined by averaging stroke times. This actually results in a tighter band of acceptable. stroke times, but is much easier to administer. The limiting value of full _ stroke tice for ' each valve will remain as listed in the IST t
Program tables.
4-52 Revision 1 9/24/80 .
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.l RELIEF REQUEST NO. VR-25 (CONTINUED)
In addition, it-is impractical to apply the require-ments of IW-3410 (c ) (3) to valves with very short stroke times (i.e. < 5 seconds) particularly solenoid valves which typically have full stroke times under one second. For these short stroke time valves, variances of 50 percent or more can occur in the measured times for reasons that are in no way related to valve performance, for example, operator reaction times. In these specific cases, verifying that the valve stroke times do not exceed 5 seconds would be sufficient to evaluate valve performance.
ALTERNATE TESTING: Based on this relief request paragraph IW-3410 (c) (3) would, in effect, read as follows:
If an increase in stroke time of 25% or more from the established reference value for valves with stroke times greater than ten seconds or 50% or more for valves with stroke times between 5 and 10 seconds or 5 seconds or more for valves with stroke times less than or equal to 5 seconds is observed, test frequency .....
-1 4-53 Revision 1 9/24/80
' RELIEF REQUEST NO. VR-26 SYSTEk: Main Steam COMPONENT: .l(2)-203-4A through 4H CATEGORY: C FUNCTION: Safety relief 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-3510, 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.
Currently, these valves are being removed from the system, cleaned and rebuilt, and then shipped off-site for re-verification of valve set points.
Therefore, IWV-3510(c) 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 t
, 4-54 Revision 1 9/24/80 e
RELIEF REQUEST NO. VR-26 (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 1
proper set point.
This accelerated maintenance schedule provides adequate assurance that these valves will perform reliably.
4 ALTERNATE TESTING: One-half (4) of the total number of safety valves will be removed and replaced with valves that have been rebuilt and had their set points verified each refueling outage.
4 4-55 Revision 1 9/24/80
(