ML19350D100
| ML19350D100 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/27/1981 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML112990691 | List: |
| References | |
| PROC-810327-01, NUDOCS 8104130319 | |
| Download: ML19350D100 (200) | |
Text
4 -
NORTHERN STATES POWER COMPANY i
PONTICELLO NUCIESR GENERATING PIANT DOCKET NO. 50-263 LICENSE NO. DPR-22
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l INSERVICE INSPECTION AND TESTING PROGRAM l
O SECOND TEN YEAR INSPECTION INTERVAL JUNE 30, 1981 - JUNE 29, 1991 i
l SUBMITTED: March 27, 1981 s
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TABLE OF CONTENTS SECTION PAGE a
1 Nondestructive Examination Program t
Class 1 1-1 Class 2 1-47 Class 3 1-71 2
Pressure Testing Program 2-1 F
3 Inservice Testing of Pumps and Valves 3-1 4
Requests for Relief from ASME Code Section 4-1 XI Requirements Determined to be Impractical 5
Proposed Technical Specification Changes 5-1 6
Quality Group Classification Drawings 6-1 i
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t RECORD OF REVISIONS Pane No.
Revision No.
Pane No.
Revision No.
Cover Original ii Original iii Original iv Original 1-1 through 1-71 Original i
2-1 Original 3-1 tnrough 3-23 Original 4-1 through 4-32 Original i
5-1 through 5-31 Original' 6-1 through 6-21 Original j
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This report contains a description of our proposed program of inservice
("'/)
inspection and testing of components of the Monticello Nuclear Generating
\\_-
Plant. This program conforms to the requirements of 10 CFR 50, Section 50.55a(g).
The information presented in this section follows the recommendations contained in a letter dated November 24,~1976 from Mr D L Ziemann,. Chief, Operating Reactors Branch #2, Division of Operating Reactors, USNRCg and in a letter dated January '1',1978 from Mr D K Davis, Acting Chief, 6
Operating Reactors Branch #2.
The program is updated as r4 quired by changes to Section 50.55a(g) published in the Federal Register on October 7, 1979.
Inservice inspection and testing requirements are updated at 120 month intervals to conform to the latest edition and addenda of Section XI of the ASME Code referenced in paragraph (b) of 10 CFR' 50, Section 50.55a.
Thie acnual will be updated each time changes are made to the inservice inspection and testing program.
Deviat ions from Code requirements are also documented for NRC Staf f review in this manual.
The program description is arranged in the following 'tianner:
Nondestructive Examinstion Class 1
- Section 1.1 Class 2
- Section 1.2 Class 3
- Section 1.3 Pressure Testing Program
- Section 2 Inservice Tests of Pumps and Valves - Section 3 Deviations from Section XI
- Sect ion 4 Requirement s N
g/
Proposed changes to the Technical Specifications which implement this program were submitted to the Cmamission on August 30, 1977. A summary of these proposed changes is contained in Section 5 of this report.
System drawings rhowing ASME Code classification boundaries are included in Section 6 of this report. These drawings are used to define pressure test boundaries sed identify those Class 3 components subject to visual inspect ion as part of the nondestructive examination program.
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ASME SECTION XI NONDESTRUCTIVE EXAMINATION PROGRAM - CLASS 1
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PROGRAM PERIOD:
2nd Ten Year Interval l
June 30, 1981 through June 29, 1991 ASME SECTION XI:
1977 Edition through and including the Summer 1978 Addenda Exception:
(Note 3) 1975 Edition through and including the Summer 1975 Addenda i
NOTES:
1.
The following tables identify the specific Class I components and i
their supports to be examined.
These tables can be directly correlated with lable IWB-2500-1 of ASME Section XI.
The tables show the amount of items required to be examined during inspection period one, two,. and three and the corresponding percentage that will have been completed by the end of that period.
i 2.
Requests for reliraf from some specific ASME Section XI examination p/
requirements that have been determined to be impractical are included i
\\_,
-in Section 4 of this report. Specific Request for Relief numbers are referenced ir. tue tables, l
- 3. -The 1974 Edition through and including the Summer 1975, Addenda of ASME Section XI was utilized to determine the extent of examination f
for Class 1 pipe welds (Program Table 9.1).
t 4.
LEGEND:
- visual exanination S
- surface examination VOL - volumetric examination L
- length l
5.
INSPECTION PERIODS:
ONE
- June 30, 1981 to October 29, 1984 l
TWO
- October 30, 1984 to February 28, 1988 THREE - March 1,1988 to June 29, 1991 s
6.
Repairs will be performed in accordance with the applicable requirements of the latest Edition and Addenda of the ASME Code,Section XI.
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MONTICELLO NUCLEAR GENERATING PLANT TABLE l1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 2
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-A; PRESSURE RETAINING VELDS IN REACTOR VESSEL Bl.10 SHELL WELDS Bl.11 CIRCUMFERENTIAL FIGURE 2 L=57' RELIEF NO.16 VCBB-1 HD - SHELL VOL 1
5' 100%
ONE 9
5' 100%
THREE 18 VCBA-2 BELTLINE 1
NOT ACCESSIBLE VCBB-3 COURSE 2-3 1
NOT ACCESSIBLE Y
VCBB-4 COURSE 3-4 VOL 1
5' 100%
ONE 9
7' 100%
TWO 21 4'
100%
THREE 28' Bl.12 LONGITUDINAL FIGURE 2 L-11' RELIEF NO.16 VLAA-1 27" BELTLINE VOL 1
4' 100%
THREE 36 VLAA-2 27" BELTLINE VOL 1
4' 100%
TWO 36 VLBA-1 117"BELTLNE 1
NOT ACCESSIBLE VLBA-2 117"BELTLNE 1
NOT ACCESSIBLE VLCB-1 COURSE 3 VOL 1
4' 100%
THREE 36 VLCB-2 COURSE 3 VOL 1
4' 100%
ONE 36 VLDB-1 COURSE 4 VOL 1
57" 100%
THREE 43 VLDB-2 COURSE 4 VOL 1
57" 100%
ONE 43 Bl.20 HEAD WELDS Bl.21 CIRCUMFERENTIAL HCCB-2 CLOSURE HD VOL 1
8.5' MIN 100%
ONE 34 FIGURE 5 8' MIN 100%
TWO 66 L=25' 8.5' MIN 100%
THREE 100 FORM 17-457 9
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MONTICELLO NUCLEAR GENERATING PLANT TABLE l1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2
OF_2 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION eNSPECTIONRUNN:NG giuAngs NO.
METHOD ITEMS AMOUNT & E)rTENT PERIOO Bl.21 (CONTINUED)
HCAB-1 BOT HD VOL 1
3 100%
ONE 7
FIGUF'. 5 3
100%
TWO 14 L=44' 3
100%
THREE 20 Bl.22 MERIDONAL WELD N0'S FIGURE 5 VOL 16 HMCB-1 CLOSURE HD HMCB-2 L-7' VOL 1
7' 100%
ONE 100 HMCB-3 HMCB-4 VOL 1
7' 100%
THREE 100 HMCB-5 HMCB-6 7
HMAB-1 BOT HD VOL 1
2.5' 100%
ONE 100 HMAB-2 L=6'2" VOL 1
2.5' 100%
TWO 100 HMAB-3 VOL 1
2.5' 100%
TWO 100 HMAB-4 VOL 1
2.5' 100%
THREE 100 HMAB-5 VOL 1
2.5' 100%
THREE 100
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HMAB-6 VOL 1
2.5' 100%
ONE 100 HMAB-7 VOL 1
2.5' 100%
ONE 100 HMAB-8 VOL 1
2.5' 100%
ONE 100 HMAB-9 NOT ACCESSIBLE HMAB-10 NOT ACCESSIBLE 31.30 SHELL-TO-FLANGE WELD VCBC-5 FIGURE 6 VOL 1
19' MIN 100%
ONE 33 L=57' 19' MIN 100%
IWO 67 19' MIN 100%
THREE 100 31.40 HEAD-TO-FLANGE WELD HCCC-1 FIGURE S&6 VOL 1
19' MIN 100%
ONE 33 L=57' 19' MIN 100%
IWO 67 19' MIN 100%
IRREE 100 Bl.50 REPAIR WELDS NONE FORM 17-4571
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B B B B B B F
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TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 2
ITEM COMPONENT OR SYSTEM IDE."'clCATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERICO EXAlfINATION CATEGORY B-D, FULL PENETRATION WELDS IN VESSELS-INSPECTION PROGRAM A REACTOR VESSEL B3.10 N0ZZLE-TO-VESSEL WELDS &
B3.20 +N0ZZLE INSIDE RADIUS WELD N0'S FIGURE 4 VOL 29 7
SECTION vi HEAD VENT N7 HVAD-1 ISI-15 1
100%
ONE 33 HEAD SPRAY N6A RHDD-1 ISI-11D 1
100%
WO 67 HEAD SPARE N6B HSBD-1 ISI-14 1
100%
HIREE 100 STANDBY LIQUID CPAE-1 ISI-17
- TO THE EXTENT WO
- RELIEF NO. 19 CONTROL N10 POSSIBLE MAIN STEAM N3A MSAD-1 ISI-1 1
100%
ONE 25 MAIN STEAM N3B MSBD-1 ISI-2 1
100%
THREE 75 MAIN STEAM N3C MSCD-1 ISI-3 1
100%
WO 50 MAIN STEAM N3D MSDD-1 ISI-4 1
100%
THREE 100 FEEWATER N4 A WAD-1 ISI-5A 1
100%
ONE 25 FEEDWATER N4B WBD-1 ISI-5A 1
100%
WO 50 FEEDWATER N4C WCD-1 ISI-5B 1
100%
WO 75 FEEDRATER N4D WDD-1 ISI-5B 1
100%
HIREE 100 CORE SPRAY NSA CS AD-1 ISI-6A 1
100%
THREE 100 CORE SPRAY NSB CSBD-1 ISI-6B 1
100%
ONE 50 CONTROL ROD DRIVE CRAD-1 ISI-10 1
100%
ONE 100 RETURN N9 RECIRC OUTLET N1A RCAD-1 ISI-13A 1
100%
ONE 50 RECIRC OUTLET N1B RCBD-1 ISI-13B 1
100%
THREE 100
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FORM 37-4571
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%J MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2 OF 2
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPflON NDE 2OTAL EXAMINATION
'NSPECTION RUNNING REMARKS NO.
METriOD ITEMS AMOUNT & EXTENT PERIOD B3.106 (CONTINUED)
B3.20 RECIRC INLET N2A RRAD-1 ISI-13D 1
100%
ONE 10 N2D RRDD-1 ISI-13D 1
100%
ONE 20 N2J RRJD-1 ISI-13C 1
100%
ONE 30 N2H RRHD-1 ISI-13C 1
100%
WO 40 N2E RRED-1 ISI-13D 1
100%
WO 50 N2G RRGD-1 ISI-13C 1
100%
WO 60 N2B RRBD-1 ISI-13D 1
100%
THREE 70 N2F RRFD-1 ISI-13C 1
100%
THREE 80 N2C RRCD-1 ISI-13D 1
100%
THREE 90 N2K RRKD-1 ISI-13C i
1 100%
THREE 109 g
JET PUMP INSTR N8A JPAD-1 ISI-16 1
100%
ONE 50 N8B JPBD-1 ISI-16 1
100%
WREE 100 B3.306 PRESSURIZER VESSEL N/A B3.40 33.506 STEAM GENERATORS N/A B3.60 B3.70& HEAT EXCHANGERS N/A B3.80 l
FORM 9 7-4 5 71
MONTICELLO NUCLEAR GENERATING PLANT TABLE 4.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 1 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-E; PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS B4.10 REACTOR VESSEL AREAS SUBJECT TO PLANTS OPTION B4.ll PARTIAL PENET WELDS
1 100%
THREE 100 N13 HD SENSOR VT-2 1
THREE N14 HD SENSOR VT-2 1
THREE B4.13 CRD PENETRATIONS CRD N0ZZLES FIGURE 1 VT-2 121 10 100%
ONE 8
10 100%
TWO 17 11 100%
THREE 26 B4.14 INSTR PENETRATIONS NllA VT-2 1
1 100%
THREE 25 N11B VT-2 1
1 N12A VT-2 1
1 N12B VT-2 1
1 B4.20 PRESSURIZER N/A i
FORM 17-4 571
5.1 MONTICELLO NUCLEAR GENERATING PLANT TABLE TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 3
I l
ITEM COMPONENT OR SYSTEM IDENT!FICATION DESCRIPTION TAL XAMINATIO IN E lON NG REMARK'-
EXAMINATION CATEGORY B-F; PRESSURE RETAINING DISSIMILAR METAL WELDS REACTOR VESSEL B5.10 N0ZZLE-TO-SAFE END WELDS WELD N0'S FIGURE 4 S,VOL 25 HEAD VENT N7 HVAF-2 ISI-15 1
100%
ONE 33 HEAD SPRAY N6A RHDJ-2 ISI-11D 1
100%
WO 67 HEAD SPARE N6B HSBF-2 ISI-14 1
100%
THREE 100 STANDBY LIQUID CPAF-2 ISI-17
- 10 THE EXTENT WO
- RELIEF NO.19 CONTROL N10 POSSIBLE CORE SPRAY NSA CSAF-2 ISI-6A 1
100%
WREE 100 CORE SPRAY NSB CSBF-2 ISI-6B 1
100%
ONE 50 CRD RETURN N9 CRAF-2 ISI-10 1
100%
ONE 100 RECIRC OUTLET N1A RCAF-2 ISI-13A 1
100%
ONE 50 RECIRC OUTLET N1B RCBF-2 ISI-13B 1
100%
THREE 100 RECIRC INLET N2A RRAF-2 ISI-13D 1
100%
ONE 10 N2D RRDF-2 ISI-13D 1
100%
ONE 20 N2J RRJF-2 ISI-13C 1
100%
ONE 30 N2H RRHF-2 ISI-1.3C 1
100%
WO 40 N2E RREF-2 ISI-13D 1
100%
WO 50 N2G RRGF-2 ISI-13C 1
100%
WO 60 N2B RRBF-2 ISI-13D 1
100%
THREE 70 N2F RRFF-2 ISI-13C 1
100%
THREE 80 N2C RRCF-2 ISI-13D 1
100%
THREE 90 N2K RRKF-2 ISI-13C 1
100%
THREE 100 FORM 9 7-4571
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MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2
OF 3 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PEktOD B5.10 (CONTINUED)
JET PUMP INSTR N8A JPAF-2 ISI-16 1
100%
ONE 50 N8B JNF-2 ISI-16 1
100%
THREE 100 INSTRUMENT LINES N11A VIAF-2 ISI-18 1
100%
ONE 25 N11B VIBF-2 ISI-18A 1
100%
TWO 50 N12A VICF-2 ISI-19 1
100%
THREE 75 N12B VIDF-2 ISI-19 1
100%
THREE 100 B5.20 PRESSURIZER N/A B5.30 STEAM GENERATORS N/A B5.40 HEAT EXCHANGERS N/A PIPING B5.50 SAFE END WELDS CORE SPRAY A CSAJ-2A W7-8"EF S,VOL 4
3 100%
ONE 75 CSAF-10 ISI-6A 1
100%
THREE 100 CSAF-14 CSAF-18 CORE SPRAY B CSBJ-2A W11-8"EF S,VOL 4
2 100%
DNE 50 CSBF-9 ISI-6B 2
100%
HIREE 100 CSBF-12 CSBF-16 HPCI-STEAM PSAF-2B PS18-8"EF S,VOL 2
2 100%
IWO 100 PSAF-2C ISI-7 FORM 17-4 57 9
MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 3
OF 3
iEM COMPONENT OR SYSTEM IDENTillCATION DESCRIPTION N
TAL XAMINATI INSPE lONRUNt4NG REMARKS B5.50 (CONTINUED)
RHR REW10 RHAF-4 REW10-18"EF S,VOL 1
1 100%
WO 100 ISI-11A RHR W20 RHBF-4 W20-16"DC S,VOL 3
1 100%
WO 33 RHBF-20 ISI-11B 2
100%
DIREE 100 RHBF-24 RHR W30 RH CF-4 W30-16"DC S,VOL 3
2 100%
WO 67 RHCF-20 ISI-11C 1
100%
EREE 100 7
RHCF-23 5
RWCU CWAF-2 REW3-4"EF S,VOL 1
1 100%
ONE 100 ISI-9 i
FO R M 9 7-4 571
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G MONTICELLO NUCLEAR GENERATING PLANT TABLE 6.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 4
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION ND TOTAL XAMINATI INTE ION RU NG REMARKS EXAMINATION CATEGORY B-G-1; PRESSURE RETAINING BOLTING, LARGER THAN 2 IN. IN DIAMETER REACTOR VESSEL B6.10 CLOSURE HEAD NUTS PRB 1-64 FIGURE 3 S
64 22 100%
ONE 34 21 100%
WO 67 21 100%
DIREE 100 B6.20 CLOSURE STUDS, IN PLACE NONE (SEE B6.30)
B6.30 CLOSURE STUDS, WHEN REMOVED PRA 1-64 FIGURE 3 S,VOL 64 22 100%
ONE 34 21 100%
WO 67 21 100%
nlREE 100 B6.40 LIGAMENTS BENEEN PRE 1-64 FIGURE 3 VOL 64 22 100%
ONE 34 STUD HOLES 21 100%
WO 67 21 100%
111REE 100 B6.50 CLOSURE WASHERS & BUSHINGS WASHERS PRD 1-64 FIGURE 3 VT-1 64 22 PAIRS 100%
ONE 34 (PAIFS)
(PAIRS) 21 PAIRS 100%
WO 67 21 PAIRS 100%
HIREE 100 BUSHINGS PRC 1-64 FIGURE 3 VT-1 64 22 100%
ONE 34 l
21 100%
WO 67 21 100%
HIREE 100 FORM 17 4571
MONTICELLO NUCLEAR GENERATING PLANT TABLE _ 6.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2
op 4
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION NSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B6.60 PRESSURIZER N/A B6.90 SIEAM GENERATORS N/A B6.120 HEAT EXCHANGERS N/A PIPING NONE PUMPS g
i U
B6.180 BOLTS AND STUDS, IN PLACE RELIEF NO. 24 RECIRC PUMP A P-200A ISI-13A VOL 16 5
100%
ONE 31 FLANGE BOLTS 1-16 5
100%
IWO 63 6
100%
THREE 100 RECIRC PUMP B P-200B ISI-13B VOL 16 5
100%
ONE 31 FLANGE BOLTS 1-16 5
100%
IWO 63 6
100%
THREE 100 36.190 BOLTS AND STUDS, WHEN REMOVED RECIRC PUMP A & B P-200A &
ISI-13A S,VOL 37
- 100% WHEN FLANGE BOLTS P-200B ISI-13B REMOVED B6.200 BOLTING IN CONJUNCTION WIIH B6.180 AND RECIRC PUMP A P-200A ISI-13A VT-1 16 5
100%
ONE 31 B6.190 FLANGE BOLTS 5
100%
IWO 63 6
100%
IHREE 100 FORM 3 7-4 571
(
1 MONTICELLO NUCLEAR GENERATING PLANT TABLE 6.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF A ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION N
OTAL XAM NATION INSPE ION RU NG REMARKS O
1 B6.200 (CONTINUED)
RECIRC PUMP B P-200B ISI-13B VT-1 16 5
100%
ONE 31 FLANGE BOLTS 5
100%
WO 63 6
100%
DIREE 100 VALVES B6.210 BOLTS AND STUDS, IN PLACE RELIEF NO. 24 RECIRC A M02-53A ISI-13A VOL 24 8
100%
ONE 33 8
100%
WO 67 8
100%
THREE 100 RECIRC A M02-43A ISI-13A VOL 24 8
100%
ONE 33 8
100%
WO 67 8
100%
DIREE 100 RECIRC B M02-53B ISI-13B VOL 24 8
100%
DNE 33 8
100%
WO 67 8
100%
HIREE 100 RECIRC B M02-43B ISI-1.3B VOL 24 8
100%
JNE 33 8
100%
WO 67 8
100%
DIREE 100 B6.220 BOLTS AND STUDS, WHEN REMOVED RECIRC A & B M02-53A ISI-13A S,VOL 96
- 100% WHEN M02-43A REMOVED M02-53B ISI-13B M02-43B FORM 9 7-4S71
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MONTICELLO NUCLEAR GENERATING PLANT TABLE 6.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 4 OF.
4 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOO ITEMS AMOUNT & EXTENT PERIOD B6.230 BOLTING IN CONJUNCTION WITH B6.210 AND RECIRC A M02-53A ISI-13A VT-1 24 8
100%
ONE 33 B6.220 8
100%
WO 67 8
100%
THREE 100 RECIRC A M02-43A ISI-13A VT-1 24 8
100%
ONE 33
'7 8
100%
WO 6
8 100%
THREE 100 RECIRC B M02-53B ISI-13B VT-1 24 8
100%
ONE 33 8
100%
WO 67 7
8 100%
WREE 100 5
RECIRC B M02-43A ISI-13B VT-1 24 8
100%
ONE 33 8
100%
WO 67 8
100%
WREE 100 I
FORM 17 4571
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MONTICELLO NUCLEAR GENERATING PLANT TABLE 71 TEN YEAR INTERVAL-EXAMINATION
SUMMARY
PAGE 1
OF 6
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRWTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-G-2; PRESSURE RETAINING BOLT-ING 2 IN. AND SMALLER IN. DIAMETER REACTOR VESSEL B7.10 BOLTS, STUDS, AND NUTS HEAD VENT N7 1-8 ISI-15 VT-1 8
8 100%
ONE 33 HEAD SPRAY N6A 1-8 ISI-11D VT-1 8
8 100%
WO 67 HEAD SPARE N6B 1-8 ISI-14 VT-1 8
8 100%
THREE 100 CONTROL ROD HOUSINGS FLANGE FIGURE 1 VT-1 121*
41 100%
ONE 34
- 121 FLANGES 4
BOLTS 40 100%
WO 67 RITH 8 BOLTS EACH u
40 100%
THREE 100 B7.20 PRESSURIZER N/A B7.30 STEAM GENERATORS N/A 37.40 HEAT EXCHANGERS N/A PIPING B7.50 BOLTS, STUDS, AND NUTS EXAMINATIONS SCHEDULED / FLANGE MAIN STEAM A 4-FL.tNGES PSI-1-8"ED VT-1 4
1 100%
ONE 25 ISI-1 3
100%
THREE 100 MAIN STEAM B 1-FLANGE PS2-18"ED VT-1 1
1 100%
WO 100 ISI-2 MAIN STEAM C 1-FLANCE PS3-18"ED VT-1 1
1 100%
ONE 100 ISI-3 FORM 17-4571
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<AONTICELLO NUCLEAR GENERATING PLANT TABLE 71 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2
op 6
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B7.50 CONTINUED MAIN STEAM D 4-FLANGES PS4-18"ED VT-1 4
1 100%
ONE 25 ISI-4 1
100%
TWO 50 2
100%
THREE 100 RHR TW36 2-FLANGES TW36-4"ED VT-1 2
2 100%
WO 100 ISI-11D RECIRC A 1-FLANGE REW26-4" VT-1 1
1 100%
ONE 100 ISI-13A 7
5 RECIRC B 1-FLANGE REW27-4" VT-1 1
1 100%
THREE 100 ISI-13B RECIRC BYPASS A 1-FLANCE REW24-4" VT-1 1
1 100%
WO 100 ISI-13AA RECIRC BYPASS B l-FLANGE REW25-4" VT-1 1
1 100%
WO 100 ISI-13BB HEAD VENT LINE l-FLANGE V15-2"EP VT-1 1
1 100%
ONE 100 ISI-15 PUMPS B7.60 BOLTS, STUDS, AND NUTS RECIRC PUMP A P-200A ISI-13A VT-1 10 3
100%
ONE 30 GLAND BOLIS 1-10 3
100%
TWO 60 4
100%
THREE 100 FORM 17-4571
O O
O
~
MONTICELLO NUCLEAR GENERATING PLANT TABLE 7.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 6 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO-METHOD ITEMS AMOUNT & EXTENT PERIOD B7.60 CONTINUED RECIRC PUMP B P-200B ISI-138 VT-1 10 3
100%
ONE 30 GLAND BOLTS 1-10 3
100%
TWO 60 4
100%
THREE 100 VALVES B7.70 BOLTS, STUDS, AND NUTS EXAMINATIONS SCHEDULED / VALVE MAIN STEAM A A02-80A PSI-18"ED VT-1 4
2 100%
ONE 50
'i' A02-86A ISI-1 NONE WO 50 0
RV2-71E 2
100%
THREE 100 RV2-71A MAIN STEAM B A02-80B PS2-18"ED VT-1 4
NONE ONE A02-86B ISI-2 2
100%
WO 50 RV2-71B 2
100%
THREE 100 RV2-71G MAIN STEAM C A02-80C PS3-18"ED VT-1 4
2 100%
ONE 50 A02-86C ISI-3 NONE TWO 50 RV2-71C 2
100%
THREE 100 RV2-71H MAIN STEAM D A02-80D PS4-18"ED VT-1 4
NONE ONE A02-86D ISI-4 2
100%
TWO 50 RV2-71D 2
100%
THREE 100 RV2-71F FEEDWATER A FW-98-2 FW2B-14"ED VT-1 3
1 100%
ONE 33 FW-97-2 ISI-5A 1
100%
TWO 67 FW-94-2 1
100%
THREE 100 FO R M 17-4 S 71
l l
MONTICELLO NUCLEAR GENERATING PLANT TABLE 7.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 4
OF 6
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION rut 4 WO REMARKS NO.
METHOD ITEMS AMOUNT & JGENT PERIOD B7.70 CONTINUED FEEDWATER B FW-98-1 FW2A-14"ED VT-1 3
1 100%
ONE 33 FW-97-1 ISI-5B 1
100%
WO 67 FW-94-1 1
100%
THREE
- 100 CORE SPRAY A MO-1754 W7-8"EF VT-1 3
2 100%
ONE 67 A014-13B ISI-6A 1
100%
TWO
-100 POS-1758 NONE THREE 100 CORE SPRAY B MO-1753 W11-8"EF VT-1 3
1 100%
ONE 33 A014-13A ISI-6B NONE WO 33 POS-1757 2
100%
THREE 100 m
HPCI STEAM MO-2034 PS18-8"ED VT-1 2
1 100%
ONE 50 MO-2035 ISI-7 1
100%
WO 100 RWCU MO-2398 REW3-4"EF VT-1 3
1 100%
ONE 33 RC-1 ISI-9 1
100%
WO 67 M0-2397 1
100%
THREE 100 RHR REWIO POS-2028 REW10-18"ED VT-1 3
1 100%
ONE 33 M0-2030 ISI-11A 2
100%
WO 100 MO-2029 NONE THREE 100 RHR W20 POS-2019 W20-16"DB VT-1 3
1 100%
ONE 33 A010-46B ISI-11B 2
100%
TWO 100 MO-2015 NONE THREE 100 RHR TW30 MO-2014 W30-16"DB VT-1 3
1 100%
ONE 33 POS-2018 ISI-11C 1
100%
WO 67 A010-46A 1
100%
THREE 100 FORM 17-4 571
O p) c O
MONTICELLO NUCLEAR GENERATING PLANT TABLE 7.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 5
OF 6
iEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION N
OTAL XAN INATI INSPE ION NG REMARKS B7.70 CONTINUED RHR TW36 RHR-21 TW36-4"F" VT-1 3
1 100%
ONE 33 M0-2027 ISI-11D 2
100%
WO 100 MO-2026 NONE THREE 100 RCIC STEAM MO-2076 PS17-3"ED VT-1 2
1 100%
TWO
.50 MO-2075 ISI-12 1
100%
THREE 100 RECI'tC BYPASS A M02-54A REW24-4" VT-1 1
1 100%
THREE 100 ISI-13AA 7
G RECIRC BYPASS B M02-54B REW25-4" VT-1 1
1 100%
THRJE 100 ISI-13BB RECIltC 'iANIFOLD M02-65A REW32-22" VT-1 4
2 100%
ONE 50 M02-66A ISI-13C 2
100%
THREE 100 M02-65B REW32-22" M02-66B ISI-13D HEAD VENT LINE XDV-1 V15-2"ED VT-1 3
1 100%
WO 33 XDV-2 ISI-15 2
100%
THREE 100 i
XDV-3 BOTTOM HEAD DRAIN XDV-4 REW31--2"ED VT-1 1
1 100%
THREE 100 ISI-21 STANDBY LIQUID XP-7 CH2-13"DC VT-1 3
1 100%
ONE 33 5
CONTROL XP-8 ISI-22 1
100%
TWO 67 XP-6 1
100%
THREE 100 MAIN STEAM DRAIN MO-2373 PS15-3"ED VT-1 2
1 100%
ONE 50 MO-2374 ISI-23 1
100%
WO 100 1
FORM 9 7-4571
6 SKR p
A 1 o M
7 ER 6
EE G
L BG N
0 0
0 AA N%
0 0
0 TP A
1 1
1 RL N
O ID TO E
E O
CI ER N
N W
PE O
O T
SP N
I 0
0 0
T 0
0 0
N 1
1 1
NE OT I X TEA&
N I T MN U
AXO EMA 1
2 2
LS ATM E
OT 1
2 2
TI D
EO 1
1 1
DH NT T
T T
EM V
V V
YRA N
D E
O M
I "2
T 2C 2
P M
I 4
6
- 6 R
82 8 2 2
9 U
C 1 -
2 -
2 -
S S
E DI WI WI D
RS ES ES TN CI RI RI N O AI L T N
P A O
G N A
3 IT NI C
3 1 1 22 TM I
I F
A A I
3 67 67 TN REX E
V RR RR D
NE I
M XX XX EG-RL AA E
R LV E
CR M
D N
N E
U E T
A I
I T
S E
A A
N Y
N S
H R
R E
D D
OI R
R O
D MN L
E AI A
B L
T EA N
U RL CE I
N C
C C
E TY N
I SN R
R N
O T
I I
I P
N DA C
C O N M
O RR E
E E
MT O
C CD R
R C
1 7
5 4
7 0
9 7
M M.O R
ETN 7
I B
O F
78
o o
O MONTICELLO NUCLEAR GENERATING PLANT TABLE 8.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 1
ITEM COMPONENT OR SYSTFM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION
- NSPI CTION RUNNt?O REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT F_,710D EXAMINATION CATEGORY B-H; VESSEL SUPPORTS REACTOR VESSEL B8.10 INTEGRALLY WELDED ATTACHMENTS SUPPORT SKIRT HCAH-2 FIGURE 5 S
1 17 100%
ONE 32 L=53' 18 100%
TWO.
66 18 100%
THREE 100 STABILIZER LUGS LUGS 1-4 FIGURE 6 S
4 1
100%
ONE 25 y
1 100%
TWO 50 N
2 100%
THREE 100 B8.20 PRESSURIZER N/A l
B8.30 STEAM GENERATORS N/A B8.40 HEAT EXCHANGEL N/A e
ronu n-4 sri
)
O m
p U
V O
MONTICELLO NUCLEAR GENERATING PLANT TABLE 9.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAG 5 1
OF 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION
&PECTIONRUNNIMG EEMARKS NO.
METHOD, ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-J;
- ASME SECTION XI PRESSURE RETAINING 1974 EDITION THRU WELDS IN PIPING SUMMER 1975 ADDENDA USED FOR B9.10 NOMINAL PIPE SIZE DETERMING THE 4 IN. AND GREATER EXTENT OF EXAM-INATIONS.
39.116 CIRCUMFERENTIAL AND B9.12
- LONGITUDINAL WELDS
- THE GREATER OF 12 IN. OR 1 PIPE MAIN STEAM A BUTWELDS PS1-18"ED S,VOL 21 3
100%
ONE 14 DIAMETER LENGTH 7
ISI-1 NONE WO 14 FROM SCHEDULED U
3 100%
THREE 29 CIRC WELD INTER-SECTION WILL BE BUTTWELDS PS1-6"ED S,VOL 6
1 100%
ONE 17 EXAMINED.
ISI-1 NONE WO 17 1
100%
THREE 33 MAIN STEAM B BUTWELDS PS2-18"ED S,VOL 26 NONE ONE ISI-2 4
100%
WO 15 3
100%
TiiREE 27 1
BUTWELDS PS2-6"ED S,VOL 3
NONE ONE O
i ISI-2 1
100%
TWO 33 NONE THREE 33 l
MAIN STEAM C BUTTWELDS PS3-18"ED S,VOL 27 2
100%
ONE 7
ISI-3 2
100%
WO 15 3
100%
THREE 26 l
BUTWELDS PS3-6"ED S,VOL 3
1 100%
ONE 33 ISI-3 NONE WO 33 NONE THREE 33 FORM 9 7-457 9
MONTICELLO NUCLEAR GENERATING PLANT TABLE _._9 1
TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2_ OF 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION ND TOTAL X MINAT) tJ
- NSPE ION NG REMARKS B9.116 (CONTINUED)
B9.12 MAIN STEAM D BUTWELDS PS4-18"ED S,VOL 24 2
100%
ONE 8
ISI-4 1
100%
WO 13 3
100%
THREE 25 BUTWELDS PS4-6"ED S,VOL 6
1 100%
ONE 17 ISI-4 NONE WO 17 1
100%
W REE 33 FEEDWATER A BUTWELDS W2B-10"ED S,VOL 12 2
100%
ONE 17 ISI-5A NONE WO 17 1
100%
THREE
.25 BUTWELDS W2B-14"ED S,VOL 12 NONE ONE ISI-SA 2
100%
WO 17 1
100%
WREE 25 FEEDWATER B BUTWELDS W2B-10"ED S,VOL 10 NONE ONE ISI-5A 2
100%
WO 20 1
100%
THREE 30 FEEDWATER C BUTWELDS W2A-10"ED S,VOL 10 1
100%
ONE 10 ISI-5B NONE WO 10 1
100%
THREE 20 FEEDWATER D BUTWELDS W2A-10"ED S,VOL 12 2
100%
ONE 17 ISI-5B NONE WO 17 1
100%
THF.EE 25 BUTWELDS N2A-14"ED S,VOL 12 2
100%
ONE 17 ISI-5B 1
100%
WO 17 NONE THREE 25 FORM 3 7-4 5 71
MONTICELLO NUCLEAR GENERATING PLANT TA3LE 91 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE
'l OF 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION fNSPECTIONRUNNING REMARKS NO-METHOD ITEMS AMOUNT & EXTENT PERIOD B9.116 (CONTINUED)
B9.12 CORE SPRAY A BUTTWELDS W 7-8"EF S,VOL 17 1
100%
ONE 6
ISI-6A 2
100%
WO 18 1
100%
THREE 24 CORE SPRAY B BUTWELDS W11-8"EF S,VOL 15 2
100%
ONE 13 ISI-6B NONE WO 13 2
100%
THREE 27 0
HPCI-STEAM BUTWELDS PS18-8"ED S,VOL 16 NONE ONE ISI-7 2
100%
WO 13 2
100%
THREE 25 RWCU LINE BUTWELDS REW3-4"EF S,VOL 17 NONE ONE ISI-9 2
100%
WO 12 2
100%
THREE 24 RHR REW10 BUTWELDS REW10-18"ED S,VOL 20 3
100%
ONE 20 ISI-11A NONE WO 20 2
100%
DIREE 25 RHR W20 BUTWELDS W20-16"DB S,VOL 19 2
100%
ONE 11 ISI-11B 2
100%
WO 21 1
100%
THREE 26 BUTWELDS W20-18"DB S,VOL 2
NONE ONE ISI-11B NONE WO 1
100%
THREE 50 FORM 17-4571
MONTICELLO NUCLEAR GENERATING PLANT TABLE 9*1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE OF 11 1EM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION ND TOTAL X MINATIO IN C N NG REMARKS B9.116 (CONTINUED)
B9.12 RHR W30 BUTWELDS W30-16"DB S,VOL 20 2
100%
ONE 10 ISI-11C 2
100%
WO 20 1
100%
THREE 25 BUTWELDS W30-18"DB S,VOL 2
NONE ONE O
ISI-11C 1
100%
WO 50 NONE IBREE 50 RHR W36 BUTWELD3 W36-4"ED S,VOL 22 NONE ONE u
ISI-11D 3
100%
TWO 14 3
100%
THREE 27 RECIRC A BUTWELDS REW13A-28" S,VOL 17 1
100%
ONE 6
ISI-13A 2
100%
TWO 18 2
100%
THREE 29 RECIRC B BUTWELDS REN13B-28" E,VOL 16 2
100%
ONE 13 ISI-13B NONE WO 13 2
100%
THREE 25 RECIRC BYPASS A BUTWELDS REW25-4" S,VOL 12 2
100%
ONE 16 ISI-13AA NONE WO 16 1
100%
'IEREE 25 RECIRC BYPASS B BUTWELDS REW26-4" S,VOL 13 2
100%
ONE 15 ISI-13BB 2
100%
WO 31 NONE THREE 31 RECIRC MANIFOLD BUTTWELDS REW32-22" S,VOL 17 2
100%
ONE 12 ISI-13C &
2 100%
WO 18 ISI-13D 1
100%
THREE 29
{
FORM 9 7 4571 l
l
o-o o
MONTICELLO NUCLEAR GENERATING PLANT TABLE _.
91 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 5
OF 11 ITEM l-COMPONENT OR SYSTEM IDENTIFICAtlON DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD I1 EMS AMOUNT & EXTENT PERIOD B9.116 (CONTINUED)
B9.12 RECIRC RISERS RISER F BUTWELDS REW14-12" S,VOL 4
NONE ONE ISI-13C NONE WO 2
100%
THREE 50 RISER G BUTWELDS REW15-12" S,VOL 4
NONE ONE ISI-13C NONE WO NONE THREE w
E RISER H BITrWELDS REW16-12" S,VOL 4
1 100%
ONE 25 ISI-13C 2
100%
Wo 75 NOWE THREE 75 RISER J BUTWELDS REW17-12" S,VOL 4
NONE ONE ISI-13C NONE WO NONE THREE RISER K BUTWELDS REW18-12" S,VOL 4
NONE ONE ISI-13C NONE WO 2
100%
THREE 50 RISER A BUTWELDS REW23-12" S,VOL 4
NONE ONE ISI-l'3D NONE WO NONE THREE RISER B BUTrWELDS REW22-12" S,VOL 4
NONE ONE ISI-13D NONE WO 2
100%
WREE 50 F O R M 17-4 5 71
1
{N J
LJ MONTICELLO NUCLEAR GENERATING PLANT TABLE 91 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 6
op 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B9.ll6 (CONTINUED)
B9.12 RISER C BUTWELDS REW21-12" S,VOL 4
NONE ONE ISI-13D NONE WO NONE THREE RISER D BUTWELDS REW20-12" S,VOL 4
NONE ONE ISI-13D NONE WO NONE THREE 7
RISER E BUTIV!'.LDS REW19-12" S,VOL 4
NONE ONE y
ISI-13D NONE WO NONE THREE HEAD SPARE BUTWELDS CLOSURE HD S,VOL 1
NONE ISI-14 HEAD VENT BUTWELD CLOSURE HD S,VOL 2
1 100%
ONE 50 ISI-15 JET PUMP INSTR BUTWELDS N8A & N8B S,VOL 2
NONE ONE ISI-16 NONE WO 1
100%
THREE 50 li"DC S,VOL 8
1 100%
ONE 13 CONDENSII;G AND INSTRUMENT LINES BUTWELDS s
FROM N11A&N11B ISI-18&l8A NONE WO 13 CONSTANT HEAD 1
100%
THREE 25 CHAMBERS CRD SCRAM HDR 8" BUTWELDS CRD26A-8" S,VOL 6
1 100%
ONE 17 CRD26B-8" 1
100%
Wo 33 ISI-24A&B NONE THREE 33 FO R BA 17-4 S 71
i MONTICELLO NUCLEAR GENERATING PLANT TABLE 9.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 7
OF 11 l
ITEM ~
COMPONENT OR SVSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION HSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOO B9.11&
(CONTINUED)
B9.12 CRD SCRAM HDR 6" BUTWELDS CRD14A-6" S,VOL 17 NONE ONE CRD14B-6" 2
100%
WO 12 ISI-24A&B 3
100%
THREE 29 CRD SCRAM HDR 4" BUTWELDS CRD13A-4" S,VOL 28 2
100%
ONE 7
CRD13B-4" 2
100%
WO 14 CRD15A-4" 3
100%
THREE 25 CRD15B-4" y
ISI-24A&B M
SCRAM DISCHARGE BUTWELDS CRD18-12" S,VOL 2
NONE ONE VOLUME TANK ISI-24C NONE WO 1
100%
THREE 50 39.20 NOMINAL PIPE SIZE LESS THAN 4 IN.
39.216 CIRCUMFERENTIAL AND
- THE GREATER OF 39.22
15 2
100%
ONE 13 FROM SCHEDULED ISI-l'2 NONE WO 13 CIRC WELD INTER-2 100%
VIREE 27 SECTION WILL BE EXAMINED MAIN STEAM BUTTWELDS PS15-3"ED S
10 2
100%
ONE 20 CONDENSATE LEAK 0FF ISI-23 NONE WO 20 1
100%
DIREE 30 FORM 17-4 57 9
o-o c,
MONTICELLO NUCLEAR GENERATING PLANT TABLE 9.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 8 OF 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITcMS AMOUNT & EXTENT PERIOD B9.30 BRANCH CONNECTION WELDS 39.31 NOMINAL PIPE SIZE GREATER THAN 2 IN.
MAIN STEAM A 6" BRANCH PS1-18"ED S,VOL 6
1 100%
ONE 17 WELDS ISI-1 NONE WO 17 1
100%
THREE 33 MAIN STEAM B 6" BRANCH PS2-18"ED S,VOL 3
NONE ONE Y
WELDS IDI-2 1
100%
WO 33 NONE THREE 33 8" BRANCH MSBJ-22 S,VOL 1
NONE WELD ISI-2 MAIN STEAM C 6" BRANCH PS3-18"ED S,VOL 3
1 100%
ONE 33 WELDS ISI-3 NONE WO 33 NONE TEREE 33 3" BRANCH MS CI-22 S,VOL 1
NONE WELD ISI-3 MAIN STEAM D 6" BRANCH PS4-18"ED S,VOL 6
1 100%
ONE 17 WELDS ISI-4 NONE WO 17 1
100%
THREE 33 RWCU 4" BRANCH REW3-4"EF S,VOL 1
1 100%
ONE 100 WELD ISI-9 RECIRC A 4" BRANCH REW13A-28" S,VOL 1
NONE WELD ISI-13A FORM 17-4571
MONTICELLO NUCLEAR GENERATING PLANT TABLE 91 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 9 OF 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION
- NSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B9.31 (CONTINUED)
RECIRC B 4" BRANQi REW13B-28" S,VOL 1
1 100%
THREE 100 WELD ISI-13B RECIRC BYPASS A 4" BRANCH REW24-4" S,VOL 2
2 100%
WO 100 WELDS ISI-13AA RECIRC BYPASS B 4" BRANCH REW25-4" S,VOL 2
1 100%
100%
THREE 100
[
RECIRC MANIFOLD 12" BRANOI REW32-22" S,VOL 8
1 100%
ONE 13 o
WELDS ISI-13C&D 1
100%
WO 25 B9.32 NOMINAL PIPE SIZE 2 IN.
AND LESS_
MAIN STEAM B 2" BRANCH PS2-18"ED 5,VOL 1
1 100%
THREE 100 WELD ISI-2 RWCU 2" BRANQi REW3-4"ED S
1 NONE WELD ISI-9 MAIN STEAM 2" BRANCH PS15A-2"ED S
8 1
100%
ONE 13 CONDENSATE LEAKOFF WETDS PS15B-2"ED PS15C-2"ED 1
100%
THREE 25 PS15D-2"ED ISI-23 CRD SCRAM HDR 2" BRANQi CRD18-2"DB S
1 NONE WELD ISI-24C RECIRC DRAIN A&B 2" BRANCll REW28-2" S
2 1
100%
ISI-26 FORM 17-4571 i
MONTICELLO NUCLEAR GENERATING PLANT TABLE 91 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 10 _ op 11 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B9.40 SOCKET WELDS HEAT VENT
- * '*~:IUKET'**
N15-2"ED S
54 4
100%
ONE 7
C WELDS 151-15 5
100%
WO 17 5
100%
THREE 26 i
w INSTRUMENT LINES SOCKET 1 "DC S
36 3
100%
ONE 8
WELDS ISI-18,18A 3
100%
WO 17 ISI-19 3
100%
THREE 25 BOTTOM HEAD DRAIN SOCKET REW31-2"ED S
40 3
100%
ONE 8
7 WELDS ISI-21 3
100%
WO 15 g
4 100%
THREE 25 STANDSY LIQUID SOCKET CH2-1 "EF S
15 1
100%
ONE 7
CONTROL WELDS ISI-22 1
100%
WO 13 2
100%
THREE 27 WIN STEM 4
- SOCKET '
PS15A-2" S
34 3
100%
ONE 9
"* '50NDENSATE LEAKOFF WELDS D S15B-2" 3
100%
...v.*
PS15C-2" 3
100%
THREE 26 PS15D-2" ISI-23 CRD SCRAM HDR SOCKET CRD16A-2" S
36 3
100%
ONE 8
DISCHARGES WELDS CRD16B-2" 3
100%
WO 17 ISI-24Ca24D 3
100%
THREE 25 CRD SCRAM SOCKET CRD18-2" S
8 NONE ONE HEADER DRAIN WELDS ISI-24C 1
100%
WO 13 1
100%
p) o o
QJ U
u r
MONTICELLO NUCLEAR GENERATING PLANT TABLE ___9 1
l TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 11 OF 11 REM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTON RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B9.40 (CONTINUED)
RECIRC MANIFOLD SOCKET VB5-2"DC S
28 4
100%
ONE 19 BYPASS OF M02-65A WELDS VB6-2"DC NONE WO 19 AND M02-65B ISI-25 3
100%
THREE 25 RECIRC A & B DRAIN SOCKET REW28-2" S
26 2
100%
ONE 8
WELDS REW29-2" 2
100%
WO 15 ISI-26 3
100%
THREE 27 7
0 FORM 17 4571
_.,__r.__
p f~s
/m Q)
\\,b (J)
MONTICELLO NUCLEAR GENERATING PLANT TABLE 10.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 3
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY INCLUDES THE B-K-1; SUPPORT MEMBERS CORRESPONDING FOR PIPING, PUMPS, AND Bil.10 (VT-3 &
VALVES VT-4) EXAMINA-TIONS WHERE PIPING APPLICABLE.
B10.10 INTEGRALLY WELDED ATTACH-MENTS AND B11.10 COMPONENT SUPPORTS Y
MAIN STEAM A WELDED PSl-18"ED S
2 1
100%
ONE 50 d
SUPPORT ISI-l VT-3 NONE WO 50 VT-4 1
100%
THREE 100 MAIN STEAM E WELDED PS2-18"ED S
2 NONE ONE SUPPORT ISI-2 VT-3 2
100%
WO 100 VT-4 NONE THREE 100 MAIN STEAM C WELDED PS3-18"ED S
2 NONE ONE SUPPORT ISI-3 VT-3 1
100%
WO 50 VT-4 1
100%
THREE 100 MAIN STEAM D WELDED PS4-18"ED S
2 NONE ONE SUPPORT ISI-4 VT-3 1
100%
WO 50 VT-4 1
100%
THREE 100 FEEDWATER A&B WELDED W2B-10"ED S
3 NONE ONE SUPPORTS ISI-5A VT-3 2
100%
WO 67 VT-4 1
100%
WREE 100 FEEDWATER A WELDED N2-14"ED S
1 1
100%
ONE 100 SUPPORTS ISI-5A VT-3 NONE WO 100 VT-4 NONE WREE 100 d=
FO R M 9 7-4 5 7 9
O O
O l
MONTICELLO NUCLEAR GENERATING PLANT TABLE 10.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2 OF 3
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION D
OTAL MINATI INSPE lON NG REMARKS B10.10 (CONTINUED)
FEEDWATER C&D WELDED W2A-10"ED S
3 1
100%
ONE 33 SUPPORTS ISI-5B VT-3 2
100%
WO 100 VT-4 NONE DIREE 100 FEEDWATER D WELDED W2A-14"ED S
1 1
100%
ONE 100 SUPPORTS ISI-5B VT-3 NONE WO 100 VT-4 NONE THREE 100 RWCU WELDED REW3-4"EF S
1 NONE ONE 7
SUPPORT ISI-9 VT-3 NONE WO VT-4 1
100%
THREE 100 RHR W36 WELDED W36-4"ED S
1 NONE ONE SUPPORT ISI-11D VT-3 1
100%
WO 100 VT-4 NONE THREE 100 l
RECIRC A WELDED REW13A-28" S
8 3
100%
ONE 37 SUPPORTS ISI-13A VT-3 2
100%
WO 63 VT-4 3
100%
THREE 100 RECIRC B WELDED REW13B-28" S
8 2
100%
ONE 25 SUPPORTS ISI-13B VT-3 2
100%
WO 50 VT-4 4
100%
HIREE 100 RECIRC MANIFOLD WELDED REW32-22" S
10 4
100%
ONE 40 SUPPORTS ISI-13C&D VT-3 3
100%
WO 70 VT-4 3
100%
THEEE 100 SCRAM DISCHARGE WELDED CRD18-12"DB S
1 1
100%
ONE 100 SUPPORT ISI-24C VT-3 NONE WO 100 VT-4 NONE THREE 100 FORM 17-4571
R R
E E
3 D
D S
N N
KR U
U F
A O
M D
D E
E E
1 R
D0 D0 0
U1 U1 1
L L
3 C0 C0 N1 N1 IB IB l
EE G
L BG IN AA N%
TP N
U R
l I
TN NE OT I X TEA&
N I T MN U
AX O EMA LSAM TOE T
TI D
EO DH NTEM YRA N
O M
I TP M
i R
U C
S S
ED TN N O AI L T N
P A O
G N A
I T
NI C
TM I
E E
I I
F A A N
N TN R
E O
O EX NE I
N N
D EG-RL AA ELV CR M
E U E T
N T SY N
S OI R
LR O
L T
EA N
CE E
TY N
I N
O P
O N M
E MT OC S
S E
P V
w M
L U
A 1
7 P
V 5
'q
-4 0
0 7
2 3
9 M O M
ITN 0
0 R
E 1
1 O
B B
F 7&
l MONTICELLO NUCLEAR GENERATING PLANT TABLE 11 1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE _1 OF 5
ITEM COMPONENT OR SYSTFM IDENTIFICATION DESCRiPTN N
TOTAL XAMINATIO IN E ION NG REMARKS EXAMINATION CATEGORY B-K-2; COMPONENT SUPPORTS RELIEF NO. 23 FOR PIPING, PUMPS, AND VALVES PIPING B11.10 COMPONENT SUPPORTS MAIN STEAM A SUPPORTS PS1-18"ED VT-3 4
2 100%
ONE 50 ISI-1 VT-4 NONE WO 50
[
2 100%
THREE 100 m
MAIN STEAM B SUPPORTS PS2-18"ED VT-3 2
NONE ONE ISI-2 VT-4 2
100%
WO 100 NONE THREE 100 MAIN STEAM C SUPPORTS PS3-18"ED VT-3 2
2 100%
ONE 100 ISI-3 VT-4 NONE WO 100 NONE THREE 100 MAIN STEAM D SUPPORTS PS4-18"ED VT-3 4
1 100%
ONE 25 ISI-4 VT-4 2
100%
WO 75 1
100%
THREE 100 FEEDWATER A SUPPORTS W2B-10"ED VT-3 3
2 100%
ONE 67 ISI-5A VT-4 NONE WO 67 1
100%
THREE 100 FEEDWATER A SUPPORTS W2B-14"ED VT-3 3
1 100%
ONE 33 ISI-5A VT-4 NONE Wo 33 2
100%
THREE 100 FORM 17-4571
o-O O
MONTICELLO NUCLEAR GENERATING PLANT TABLE 11 1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2
OF 5 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION TOTAL X MINATIO N C RUNP41NG REMARK 5 311.10 (CONTINUED)
FEEDWATER D SUPPORTS W2A-10"ED VT-3 3
1 100%
ONE 33 ISI-5B VT-4 NONE WO 33 2
100%
IHREE 100 FEEDWATER D SUPPORTS W2A-14"ED VT-3 3
1 100%
ONE 33 ISI-5B VT-4 1
100%
WO 67 1
100%
THREE 100 CORE SPRAY A SUPPORTS W7-8"EF VT-3 2
1 100%
ONE 50 ISI-6A VT-4 1
100%
WO 100 h
NONE IRREE 100 CORE SPRAY B SUPPORTS Wll-8"EF VT-3 2
1 100%
ONE 50 ISI-6B VT-4 NONE WO 50 1
100%
THREE 100 HPCI-STEAM SUPPORTS PS18-8"ED VT-3 1
NONE ONE VT-4 NONE WO 1
NONE 100%
THREE 100 RWCU SUPPORTS REW3-4"EF VT-3 2
NONE GNE ISI-9 VT-4 NONE WO 2
100%
THREE 100 RHR REW10 SUPPORTS REW10-18"ED VT-3 6
NONE ONE ISI-11A VT-4 4
100%
WO 67 2
100%
THREE 100 RHR W20 SUPPORTS W20-16"DB VT-3 6
NONE ONE ISI-llB VT-4 2
100%
WO 33 4
100%
IRREE 100 FORM 9 7-4 571
n g-ga v
MONTICELLO NUCLEAR GENERATING PLANT TABLE 11 1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 3
op 5
ITEM COMPONENT OR SYSIEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNi & EXTENT PERIOD B11.10 (CONTINUED)
RHR W30 SUPPORTS W30-16"DB VT-3 5
2 100%
ONE 40 ISI-11C VT-4 3
100%
WO 100 NONE THREE 100 RHR W36 SUPPORTS W36-4"ED VT-3 1
NONE ONE ISI-11D VT-4 1
100%
WO 100 NONE THREE 100 RCIC-STEAM SUPPORTS PS17-3"ED VT-3 3
1 100%
ONE 33 ISI-12 VT-4 1
100%
WO 67 0
1 100%
THREE 100 m
RECIRC A SUPPORTS REW13A-28" VT-3 12 4
100%
ONE 33 ISI-13A VT-4 4
100%
WO 67 4
100%
THREE 100 RECIRC B SUPPORTS REW13B-28" VT-3 12 4
100%
ONE 33 ISI-13B VT-4 4
100%
WO 67 4
100%
THREE 100 RECIRC BYPASS A&B SUPPORTS REW24-4" VT-3 2
1 100%
ONE 50 REW25-4" VT-4 NONE WO 50 ISI-13AA&BB 1
100%
THREE 100 RECIRC MANIFOLD A&B SUPPORTS REW32-22" VT-3 10 4
100%
ONE 40 ISI-13C&D VT-4 3
100%
WO 70 3
100%
THREE 100 RECIRC RISERS SUPPORTS ISI-13C &
VT-3 10 3
100%
ONE 30 MANIFOLD A & B ISI-13D VT-4 3
100%
WO 60 4
100%
THREE 100 FORM 17-4579
{3 O
()
V s
MONTICELLO NUCLEAR GENERATING PLANT TABLE 11 1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 4 OF 5
IEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION D
TAL XAMINATION INSPE l@ RU NG REMARKS B11.10 (CONTINUED)
HEAD VENT LINE SUPPORTS V15-2"ED VT-3 2
1 100%
ONE 50 ISI-15 VT-4 NONE WO 50 1
100%
THREE 100 BOTTOM HEAD DRAIN SUPPORTS REW31-2" VT-3 5
3 100%
ONE 60 ISI-21 VT-4 2
100%
WO 100 NONE THREE 100 STANDBY LIQUID SUPPORTS CH2-11" VT-3 2
1 100%
ONE 50 5
d, CONTROL ISI-22 VT-4 1
100%
WO 100 NONE THREE 100 CRD SCRAM SUPPORTS CRD26A-8"DB VT-3 14 4
100%
ONE 29 HEADER A CRD13A-4"DB VT-4 5
100%
100%
THREE 100 CRD15A-4"DB ISI-24A CRD SCRAM SUPPORTS CRD26B-8"DB VT-3 15 5
100%
ONE 33 HEADER B CRD13B-4"DB VT-4 5
100%
100%
THREE 100 CRD15B-4"DB ISI-24B CRD SCRAM HEADER SUPPORTS CRD16A-2"DB VT-3 18 4
100%
ONE 33 DISCHARGES A & B CRD16B-2"DB VT-4 7
100%
WO 67 ISI-24 C&D 7
100%
THREE 100 CRD SCRAM "EADER SUPPORTS CRD18-2"DB VT-3 1
1 100%
ONE 100 DRAIN ISI-24C VT-4 NONE WO 100 NONE THREE 100 FORM 9 7-4 571
o o
o MONTICELLO NUCLEAR GENERATING PLANT TABLE 11 1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 5 OF 5
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION
!NSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD B11.10 (CONTINUED)
SCRAM DISCHARGE SUPPORT CRD18-12"DB VT-3 1
1 100%
ONE 100 VOLUME TANK ISI-24C VT-4 NONE WO 100 NONE THREE 100 RECIRC VALVE SUPPORTS VBS-2"DC VT-3 2
1 100%
ONE 100 BYPASS A & B VB6-2"DC VT-4 1
100%
WO 100 ISI-25 NONE THREE 7
B11.20 PUMPS C_0MPONENT SUPPORTS NONE INCLUDED UNDER B11.10 B11.30 VALVES COMPONENT SUPPORTS NONE INCLUDED UNDER B11.10 l
L FORM 17-4 571
A g^g
% j' 0
MONTICELLO NUCLEAR GENERATING PLANT TABLE 12.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 3
I ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-L-1, B-M-17 PRESSURE RETAINING WELDS IN PUMP CASING AND VALVE BODIES B-L-2, B-M-2; PUMP CASINGS AND VALVE BODIES PUMPS B12.10 PUMP CASING WELDS NONE B12.20 PUMP CASING RECIRC PUMPS P-200A/
REW13A-28" VT-1 2
- RELIEF NO. 41 AND B P-200B REW13B-28" ISI-13A6B
- VALVES, B12.10 VALVE BODY WELDS NONE 312.20 VALVE BODY, EXCEEDING 4 IN. NOMINAL PIPE SIZE ATWOOD MORRILL A02-80A PSI-18"ED VT-1 8
EXAMINE THE THREE 100 GLOBE VALVES A02-86A INTERNALS OF A02-80B PS2-18"ED ONE VALVE A02-86B A02-80C PS3-18"ED A02-86C A02-80D PS4-18"ED A02-86D FORM 17-4 571
O p
p V
U C#
MONTICELLO NUCLEAR GENERATING PLANT TABLE 12.1 TEN YEAR INTERVAL-EXAMINATION
SUMMARY
PAGE 2
OF 3
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION (NSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERtOD 312.20 CONTINUED TARGET ROCK RV2-71A PSI-18"ED VT-1 8
EXAMINE THE THREE 100 RELIEF VALVES RV2-71E INTERNALS OF RV2-71B PS 2-18"ED ONE VALVE RV2-71G RV2-71C PS3-18"ED RV2-71H RV2-71D PS4-18"ED RV2-71F 7
ANCHOR W-9 7 -2 W2B-14"ED VT-1 4
EXAMINE THE THREE 100 CHECK VALVES FW-94-2 INTERNALS OF W-9 7-1 W2A-14"ED ONE VALVE FW-94-1 ATWOOD MORRILL A010-46B FW2A-14"ED VT-1 2
EXAMINE THE THREE 100 CHECK VALVE A010-46A W23-14"ED INTERNALS OF ONE VALVE ROCKWELL A014-13B TW7-8"EF VT-1 2
EXAMINE THE THREE 100 CHECK VALVE A014-13A TW11-8"EF INTERNALS OF ONE VALVE ANCHOR POS-1753 TV-7-8"EF VT-1 15 EXAMINE THE THREE 100 GATE VALVE MO-1754 INTERNALS OF POS-1757 TW11-8"EF ONE VALVE MD-1753 MO-2034 PS18-8"ED MO-2035 M0-2029 REW10-18"ED MO-2030 POS-2028 FORM 9 7-4 57 9
MONTICELLO NUCLEAR GENERATING PLANT TABLE 12.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 3
op 3
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD 312.20 CONTINUED W-9 8-2 W2B-14"ED W-98-1 W2A-14"ED POS-2019 W20-16"DB MO-2015 POS-2018 W30-16"DB M0-2014 CRANE CHAPMAN M02-65A REW32-22" VT-1 6
- RELIEF NO. 42 GATE VALVE M02-65B M02-53A REW13A-28" M02-43A M02-53B REW13B-28" M02-43B FORM 17-4 57 9
l l
13.1 MONTICELLO NUCLEAR GENERATING PLANT TABLE TEN YEAR INTERVAL-EXAMINATION
SUMMARY
PAGE 1
OF 1
NSPECTON RUNNING REMARKS l
ITEM COMPONENT OR SYSTEM IDENTiftCATION DESCRIPTION NDE TOTAL EXAMINATION NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-N-1, INTERIOR OF REACTOR VESSEL; E-N-2, INTEGRALLY WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACHMENTS TO REACTOR VESSELS; B-N-3, REMOVABLE CORE SUPPORT STRUCTURES REACTOR VESSEL Tg B13.10 VESSEL INTERIOR SPACE ABOVE AND BELOW VT-3 VISUALLY ONE 100 DIE REACTOR CORE DIAT ACCESSIBLE AREAS WO 100 IS MADE ACCESSIBLE FOR THREE 100 EXAMINATION BY THE RE!'^ VAL OF COMPONENTS DURINJ NORMAL REFUELING OUTAGES.
B13.20 INTERIOR ATTACHMENTS &
ALL ATTACHMENTS AND VT-1 VISUALLY ONE 100 B13.30 CORE SUPPORT STRUCTURE CORE SUPPORT STRUCTURES ACCESSIBLE WELDS TWO 100 AND SURFACES THREE 100 REACTOR VESSEL (PWR)
B13.30 CORE SUPPORT STRUCTURE N/A l
FORM 9 7-4571
1 1
ll LA 1
RS EG SK HN R
PI O
1 F
A I S O
M RU 4
E EO R
1 PH 4
1 2
EE G
LBG N
0 I
AA N%
370 TP N
361 U
R NOD E
I TO E
I TN 000 NE 000 OT I
X 1 1 1 TE A &
N I
T MN U
A O XEMA 1 1 1 LS A M 1
TOT 2
E TI 1*
D L
EO O
O V, O DH L
NTEM SV YRA N
1 O
M I
E TP M
I R
R U
U C
G S
S E
I D
F TN N O AI LT N
P A O
G N A
G I
T NI I
C N
TM I
I S F
A A I
SD TN UL R
OE EX E
NE I
HW D
EG-RL 0
S AA D
S E
B L
G LV E
N CR M
Y W I
G E
R S
N UE I
O G U L
I T
S N
Y G N O E
S N
S E
I H
S U
OI R
T N
S O
LR O
L A
I D
E H
T EA N
C A O V
ICE E
T R
D TY N
N E
R R N
O O R L O C O N P
I O
T M
T E
R C N E
MT O
A R T A
I C
N U N E
I S
O R
S M S C
D 0
A E L
X R N E
E P
I W
0 1
M.O ETN 4
I 1
3 1l L
l!lll
O b)
~
l MONTICELLO NUCLEAR GENERATING PLANT TABLE 15.1 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1 OF 1 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION MNG REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY B-P; ALL PRESSURE RETAINING
_ COMPONENTS B15.10 REACTOR VESS2L PRESSURE VT-2 PRESSURE RETAIN-100%
- SYSTEM LEAKAGE B15.50 PIPING RETAINING ING BOUNDARY TEST PERFORMED B15.60 PUMPS BOUNDARY BY PLANT EAGI B15.70 VALVES REFUELING OUTAGE B15.11 REACTOR VESSEL PRESSURE VT-2 PRESSURE RETAIN-100%
- SYSTEM HYDRO-315.51 PIPING RETAINING ING BOUNDARY STATIC TEST B15.61 PUMPS BOUNDARY PERFORMED BY B15.71 VALVES PLANT EACH INTERVAL B15.20 PRESSURIZER N/A B15.30 STEAM GENERATORS
,"/A B15.40 HEAT EXCHANGERS N/A FORM 17-457 9
- - ~. - - - - - - -
(n]
ASME SECTION XI NONDESTRUCTIVE EXAMINATION PROGRAM - CIASS 2 PFOGRAM PERIOD:
2nd Ten Year Interval June 30, 1981 to June 29, 1991 ACME SECTION XI:
1977 Edit' ion through and including the Summer 1978 Addenda Exception:
(Note 3) 1975 Edition through and including the Summer 1975 Addenda NOTES:
1.
The following tables identify the specific Class 2 components and their supports to be examined.
These tables can be directly r
correlated with Table IWCH2500-1 of ASME Section XI.
The tables show the amount of items required to be examined during inspection period one, two, and three and the corresponding percentage that i
will have been completed by the end of that period.
2.
Requests for relief from some specific ASME Section XI examination
[
O' requirements that have been determined to be impractical are included i
in Section 4 of this report. Specific Request for Relief numbers are referenced in the tables.
3.
The 1974 Edition through and including the Summer 1975 Addenda of ASME Section XI was utilized to determine the extent of examiaation for class 2 pipe welds (Program Table 5.10).
4.
LEGEND:
- visual examination S
- surface examination VOL - volumetric examination i
L
- length 5.
INSPECTION PERIODS:
ONE
- June 30, 1981 to October 29, 1984 TWO
- October 30, 1984 to February 28, 1988 THREE - February 28,10S8 to June 29, 1991 i
6.
Repairs will be perforned in accordance with the applicable requirements of the latest Edition and Addenda of the ASNE Code,Section XI.
l 1-47 l
1 e
O O
O MONTICELLO NUCLEAR GENERATING PLANT TABLE 1.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 1
ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRtPTION MET D T S A
NT & E NT PER EXAMINATION CATEGORY C-A, PRESSURE RETAINING WELDS IN PRESSURE VESSELS C1.10 SHELL CIRCUMFERENTIAL WELDS RHR HEAT EXCHANGERS (6)
(3)
(100%)
(100)
MULTIPLE VESSELS E-200A SHELL TO WELD 1 VOL 3
1 100%
ONE 33 7
100%
TWO 67 E-200B ISI-50 WELD 2 70L 3
1 100%
THREE 100 C1.20 HEAD CIRCUMFERENTIAL WELDS RHR HEAT EXCHANGERS (2)
(1)
(100%)
(100)
E-200A flEAD TO WELD 4 VOL 1
E-200B SHELL WELD 1
1 100%
TWO 100 C1.30 IUBE SHEET TO SHELL WELDS NONE FOR M 9 7-4 5 71
lI 1
Y -
BA N
DI EM 1
TA SK NX R
EE o
F A
M O
M EE E
LC 0
R PAS 1
PFN URO 2
1 SUI
- ST EE G
LBG N
5050 AA
'N %
2570 TP N
1 U
R N
OD E
TO E
I CI EO0R R
PE N W '4 H E
SP OTTT N
i 0000 0000 TN 1 1 1 1 NE OT I X TE A &
N I
T MN U
AXO EM A
E 2
2 OT TI L
L D
O O
EO DH V,
V, o
T NE S
S M
YR 8
8 A
N 6
O M
I 7
7 TP M
i D
D S
S R
U C
L L
S S
E E
E D
W W
TN N O AI L T N
P A O
G N A
4 4
iT NI I
C E
N N
TM I
N F
A A I
O TN N
REX E
3 3
D NE I
N N
EG-
, E n
RL B L i
K AA Z
R C
C Z 2
E I
EV 0
/
V H LCR M
Y N
1 O T E
R U E T
T S
O G S
L S
L N
Y G N L A L
A N
S E
I S
E N
E N
OI R
T N L S
I S
I LR O
L A
I E
S M
S M
EA N
C A S
E O
E O
T T
S V N V
N SA B
ICE E
N N E E
R0 0
TY N
O O R V N N N N TE0 0
P I
I I
S I
I AG2 2
O N M
T E N S
EN -
E MT O
A R I
S S
E S
HAE E
C N U E
S N
E n
H I
S S
L E
K L
i RC M S D
Z L
C Z
S HX o
A E
L Z
I Z
2 S
RE 1
X R E O R H 0
/
E 7
E PW N O T N
1 N
5 4
7 0
0 1
2 M
1 M O TN 2
2 R
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OF
[e 1llIl
6
)
v v
i MONTICELLO NUCLEAR GENERATING P: ANT TABLE 3.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 10 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECilON RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD b
EXAMINATION CATE T RY C C AND C-E, SUPPORT MEMBERS C3.10 INTEGRALLY WELDED SUPPORT ATTACHMENTS RHR HEAT EXCHANGERS (6)
(3)
(100%)
(100)
MULTIPLE VESSELS E-200A WELDED E-200A S
3 1
100%
ONE 33 SUPPORTS 1
100%
TWO 67 E-200B E-200B S
3 1
1(.0%
THREE 100 C3.20 COMPONENT SUPPORTS RHR HEAT EXCHANGERS E-200A SUPPORTS E-200A VT-3 3
2 100%
ONE 33 1
100%
TWO 50 E-200B SUPPORTS E-200B VT-3 3
1 100%
TWO 67 2
100%
THREE 100 C3.30 3UPPORTS-MECHANICAL AND NONE IYDRAULIC FORM 37-4571
7 Ns G
.-I MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 2
op 10 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTION RUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD p7p g.g
- INCLUDES THE CORRESPONDING C3.40
- INTEGRALLY WELDED C3.50(VT-3) &
SUPPORT ATTACIDfENTS T0S MAIN STEAM A WELDED FSI-18"ED S
1 1
100%
ONE 100 SUPPORT ISI-26 VT-3 NONE WO 100 VT-4 NONE THREE 100 MAIN STEAM B WELDED PS2-18"ED S
1 NONE ONE SUPPORT ISI-27 VT-3 1
100%
WO 100 y
0, VT-4 NONE THREE 100 MAIN STEAM C WELDED PS3-18"ED S
1 NONE ONE SUPPORT ISI-28 VT-3 1
100%
WO 100 VT-4 NONE THREE 100 MAIN STEAM D WELDED PS4-18"ED S
1 NONE ONE SUPPORT ISI-29 VT-3 NONE WO VT-4 1
100%
'HIREE 100 SUPPLY TO STEAM WELDED PS14-6"ED S
1 NONE ONE SEAL SYSTEM SUPPORT ISI-30 VT-3 NONE WO VT-4 1
100%
THREE 100 HPCI WATER WELDED W3-12"ED S
2 1
100%
ONE 50 DISCHARGE SUPPORTS ISI-31 VT-3 1
100%
WO 100 VT-4 NONE THREE 100 HPCI STEAM WELDED PS18-8"ED S
2 NONE ONE SUPPORTS ISI-32 VT-3 1
100%
WO 50 VT-4 1
100%
THREE 100 FORM 17-4571
O O
O MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 3
OF 10 ITEM CO*APONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD C3.40 CONTINUED HPCI STEAM WELDED RS2-16"HE S
2 NONE ONE DIS CHA. 'CE SUPPORTS ISI-33 VT-3 2
100%
WO 100 VT-4 NONE THREE 100 CORE SPRAY A WELDED W7-10"GE S
2 1
100%
ONE
.50 DISCHARGE SUPPORTS ISI-34A VT-3 1
100%
WO 100 VT-4 NONE fHREE 100 CORE SPRAY B WELDED W11-10"GE S
2 NONE ONE
[
DISCHARGE SUPPORTS ISI-35 &
VT-3 NONE IUO N
ISI-35A VT-4 2
100%
THREE 100 REACIOR WATER FROM WELDED REWil-8 'HE S
3 1
100%
ONE 33 SKIMMER SYSTEM SUPPORTS ISI-36 VT-3 1
100%
WO 67 VT-4 1
100%
THREE 100 RHR SERVICE WELDED SW9-8"G';
S 1
1 100%
ONE 100 WATER SUPPORT ISI-39 VT-3 NONE WO 100 VT-4 NONE THREE 100 RHR SUCTION A WELDED W16-14"HE S
2 NONE ONE SUPPORTS W18-14"HE VT-3 2
100%
WO 100 ISI-40 7T-4 NONE THREE 100 RHR DISCHARGE A WELDED W30-16"GE S
1 NONE ONE SUPPORTS ISI-41 7T-3 1
100%
WO 100 VT-4 NONE THREE 100 RHR SUCTION B WELDED W15-14"HE S
2 NONE ONE SUPPORTS W17-14"HE VT-3 NONE WO ISI-42 VT-4 2
100%
THREE 100 FORM 37-4571
,r i-V) k[%
MONTICELLO NUCLEAR GENERATING PLANT TABLE 3,10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 4 OF 10 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNN!NG REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD C3.40 CONTINUED RHR DISCHARGE B WELDED W19-10"GE S
1 NONE ONE SUPPORT ISI-43 VT-3 1
100%
WO l100 VT-4 NONE THREE 100 WELDED W20-16"GE S
1 NONE ONE SUPPORT ISI-43 VT-3 NONE WO VT-4 1
100%
THREE 100 CONTAINMENT WELDED TW32-12"GE S
4 1
100%
ONE 25 SPRAY A & B SUPPORTS W23-10"GE VT-3 1
100%
WO 50 h
W33-12"GE VT-4 2
100%
THREE 100 ISI-44 C3.50
- COMPONENT SUPPORTS
- INCLUDES THE CORRESPONDING MAIN STEAM A SUPPORTS PS1-18"ED VT-3 6
2 100%
ONE 33 C3.60 (VT-4)
VT-4 2
100%
WO 67 EXAMINATIONS 2
100%
THREE 100 WHERE APPLICABLE.
MAIN STEAM B SUPPORTS PS2-18"ED VT-3 6
2 100%
ONE 33 VT-4 2
100%
WO 67 RELIEF NO. 23 2
100%
THREE 100 MAIN STEAM C SUPPORTS PS3-18"ED VT-3 6
2 100%
ONE 33 VT-4 2
100%
WO 67 2
100%
THREE 100 MAIN STEMI D SUPPORTS PS4-18"ED VT-3 6
2 100%
ONE 33 VT-4 2
100%
WO 67 2
100%
THREE 100 FORM 17-457 9
p p
,D G
n l
G l
\\
MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 5
op 10 iEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION ND OTAL XAMINA ION INSPEC N NG REMARKS C3.50 CONTINUED SUPrLY TO STEAM SUPPORTS PS11-6"ED VT-3 3
2 100%
ONE 29 SEAL SYSTEM PS12-6"ED VT-4 1
2 100%
3 100%
THREE 100 PS14-6"ED 1
ISI-30 7
3 100%
ONE 27 PS7-8"ED VT-4 2
3 100%
WO 55 ISI-30 11 5
100%
THREE 100
~
Oi*
MAIN STEAM SUPPORTS PS30-18"EDB VT-3 3
NONE ONE EQUALIZER HDR ISI-30A VT-4 2
100%
WO 67 1
100%
mREE 100 HPCI WATER SUPPORTS W3-12"ED VT-3 17 6
100%
ONE 35 DISQlARGE ISI-31 VT-4 6
100%
WO 71 5
100%
m REE 100 HPCI WATER SUPPORTS W1-14"HE VT-3 4
NONE ONE SUCTION ISI-31A VT-4 2
100%
WO 50 2
100%
THREF 100 HPCI STEAM SUPPORTS PS18-8"ED VT-3 13 3
100%
ONE 23 ISI-32 VT-4 5
100%
WO 62 5
100%
THREE 100 HPCI STEAM SUPPORTS RS2-16"HE VT-3 6
2 100%
ONE 33 DISCHARGE VT-4 4
100%
WO 100 NONE THREE 100 FORM 97-4571
f I
MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 6
OF in ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION N
TAL XAMINATI iNSPE lON NG REMARKS C3.50 CONTINUED CORE SPRAY A SUPPORTS W6-12"HE VT-3 5
2 100%
ONE 40 SUCTION ISI-34 VT-4 2
100%
TWO 80 1
100%
THREE 100 CORE SPRAY A SUPPORTS W7-10"GE VT-3 16 5
100%
ONE 31 DIS CHARGE W7-8"GE VT-4 6
100%
WO 69 ISI-34 &
5 100%
THREE 100 ISI-34A CORE SPRAY B SUPPORTS TW10-12"HE VT-3 4
1 100%
ONE 25 SUCTION ISI-35 VT-4 NONE WO 25 3
100%
THREE 100 CORE SPRAY B SUPPORTS W11-10"GE VT-3 12 3
100%
ONE 25 DISCHARGE W11-8"GE VT-4 4
100%
WO 58 ISI-35 &
5 100%
THREE 100 ISI-35A REACTOR WATER FROM SUPPORTS REW11-8"HE VT-3 6
2 100%
ONE 40 SKIM >ER SYSTEM ISI-36 VT-4 NONE WO 40 4
100%
THREE 100 RCIC WATER SUPPORTS W5-6"HE VT-3 2
1 100%
ONE 50 SUCTION ISI-38 VT-4 1
100%
WO 100 NONE THREE 100 RCIC STEAM SUPPORTS RS3-8"HE VT-3 6
2 100%
ONE 33 DISCHARGE ISI-38 VT-4 2
100%
WO 67 2
100%
THREE 100 FORM 17-4 571
A
/~N N
m)
MONTICELLO NUCLEAR GENERATING PLANT TABLE
'i 10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 7 OF 10 iEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION OTAL XAMINATIO INSPEC N RU NG REMARKS O
C3.50 CONTINUED RHR SERVICE WATER SUPPORTS SW9-8"GE VT-3 16 5
100%
ONE 31 ISI-39 VT-4 5
100%
WO 62 6
100%
THREE 100 RHR SUCTION A SUPPORTS REW10-1 f3"HE VT-3 4
2 100%
ONE 50 ISI-40 VT-4 2
100%
WO 100 NONE THREE 100 SUPPORTS W14B-20"HE VT-3 3
2 100%
ONE 67 Oi ISI-40 VT-4 1
100%
WO 100 NONE DIREE 100 SUPPORTS N28-20"HE VT-3 3
NONE ONE ISI-40 VT-4 NONE WO 3
100%
THREE 100 RHR DISCHARGE A SUPPORTS N29-10"GE VT-3 4
1 100%
ONE 25 ISI-41 VT-4 1
100%
WO 50 2
100%
THREE 100 SUPPORTS W30-14"GE VT-3 8
3 100%
ONE 38 ISI-41 VT-4 3
100%
WO 75 2
100%
THREE 100 SUPPORTS W30-16",DB VT-3 1
NONE ONE ISI-41 VT-4 1
100%
WO 100 NONE THREE 100 RHR SUCTION B SUPPORTS REW10-18"HE VT-3 4
NONE ONE ISI-42 VT-4 NONE WO 4
100%
THREE 100 FORM 17 4571
O f~%
O b
d
()
MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.10 TEN YEAR INTERVAL.- EXAMINATION
SUMMARY
PAGE 8
op 10 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION
- NSPECTIONRUNNtNG REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD C3.50 CONTINUED SUPPORTS W14A-20"HE VT-3 4
2 100%
ONE 50 ISI-42 VT-4 NONE WO 50 2
100%
THREE 100 SUPPORTS W27-20"HE VT-3 3
NONE ONE ISI-42 VT-4 2
100%
WO 67 1
100%
THREE 100 RHR DISCHARGE B SUPPORTS W19-10"GE VT-3 2
NONE ONE
[
ISI-43 VT-4 NONE WO u
2 100%
THREE 100 SUPPORTS W19-14"GE VT-3 1
NONE ONE ISI-43 VT-4 1
100%
WO 100 NONE THREE 100 SUPPORTS W20-14"GE VT-3 9
3 100%
ONE 33 ISI-43 VT-4 2
100%
WO 56 4
100%
THREE 100 SUPPORTS N22-14"GE VT-3 1
1 100%
ONE 100 ISI-43 VT-4 NONE WO 100 NONE THREE 100 CONTAINMENT SUPPORTS N23-12"GE VT-3 6
2 100%
ONE 33 SPRAY A & B W23-10"GE VT-4 1
100%
WO 50 ISI-44 3
100%
WREE 100 SUPPORTS WS3-12"GE VT-3 7
2 100%
ONE 29 W33-10"GE VT-4 1
100%
WO 43 ISI-44 4
100%
THREE 100 FORM 17 4571
p) p g
cl v
MONTICELLO NUCLEAR GENERATING PLANT TABLE 3.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 9 OF 10 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCR PTION NDE TAL XAMINAil INS E lON RUNNING REMARKS C3.60
- SUPPORTS - FECHANICAL
- INCLUDED UNDER AND HYDRAULIC C3.40 & C3.50 PUMPS C3.70
- INTEGRALLY WELDED
- INCLUDES THE SUPPORT ATTACHMENTS CORRESPONDING C3.80 (VT-3)
RHR PUMPS WELDED P-200A S
4 1
100%
IWO 50 EXAMINATIONS SUPPORTS P-200B VT-3 1
100%
THREE 100 P-200C 1
100%
TWO 75 P-200D 1
100%
ONE 25 ISI-48 CORE SPRAY PUMPS WELDED 14-1A S
2 1
100%
THREE 100 SUPPORTS 14-1B VT-3 1
100%
ONE 50 C3.80 COMPONENT SUPPORTS HPCI TURBINE & PUMPS SUPPORTS TURBINE VT-3 11 3
100%
ONE 27 DVS PUMP 3
100%
TWO 55 DVMX PUMP S
100%
THREE 100 ISI-45 &
ISI-46 RCIC TURBINE & PUMP SUPPORTS TURBINE VT-3 4
1 100%
ONE 25 PUMP 1
100%
IWO 50 2
100%
THREE 100 FORM q 7 4 5 71
MONTICELLO NUCLEAR GENERATING PLANT TABLE L 10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 10 OF 10 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNN'NG REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD C3.90 SUPPORTS - FECHANICAL NONE AND HYDRAULIC VALVES 33.100 INTEGRALLY WELDED SUPPORT ATTACHMENTS
- INCLUDED UNDER C3.40, C3.50,
& C3.60 23.110 COMPONENT SUPPORTS 23.120 SUPPORTS - MECHANICAL AND HYDRAULIC i
i FORM 17-4 571
MONTICELLO NUCLEAR GENERATING PLANT TABLE 4.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 1
! TEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD ITEMS AMOUNT & EXTENT PERIOD EXAMINATION CATEGORY C-D, PRESSURE RETAINING BOLTING EXCEEDING 2 IN. IN DIAMETER PRESSURE VESSELS C4.10 BOLTS AND STUDS NONE PIPING y
Ino C4.20 BOLTS AND STUDS NONE PUMPS C4.30 BOLTS AND STUDS NONE VALVES C4.40 BOLTS AND STUDS NONE 4
FORM 17 4 57)
G O
O l
L.)
O MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL.- EXAMINATION
SUMMARY
PAGE 1 OF 8 _
EXAMINATION AMOUNT ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NOE TOTAL
- NSPECTION REMARKS NO.
- cIHOD ITEMS 40 YR 10 YR PERIOD
'4 (40YR)
REQUIRED %
10YR RUNNING %
EXAMINATION CATEGORY C-F*
PRESSURE RETAINING WELDS
- EXTENT OF EXAMS IN PIPING ARE DETERMINED USING 1974 ED THRU SUMMER 1975 C5.10 PIPING WELDS is IN. OR ADDENDA 0F ASME LESS NOMINAL WALL SECTION XI THICKNESS RELIEF No. 15 e-- C5.116 CIRCUMFERENTIAL A!a' m C5.12
- LONGITUDINAL WELDS
(' 75 CATEGORY C-F)
SUPPLY TO STEAM SEAL SYSTEM (19)
(5)
(100)
SINGLE STREAM PS10-5" CIRC WELDS ISI-30 S
19 19 3
ONE 16 5" X.375" 2
THREE 26 (22)
(6)
(2)
(100)
MULTIPLE STREAMS PS11-6"ED CIRC WELDS ISI-30 S
7 2
1 ONE 17 PS12-6"ED 6" X.432" S
5 1
PS13-6"ED S
5 1
PS14-6"ED S
5 2
1 THREE 33 RHR SUCTION A&B (24)
(6)
(100)
SINGLE STREAM REW10-18"HE CIRC WELDS ISI-40 S
24 24 2
ONE 8
18" X.375" ISI-42 2
IWo 17 2
THREE 25 FORM 17-457 9
l p
,A p)
V
(
O MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE
'2 OF 8 EXAMINATION AMOUNT ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TAL
'N E ION REMARKS 3.11&
(40YR)
REQUIRED %
3.12 (CONTINUED) 10YR RUNNING %
(10)
(5)
(1)
(100)
MULTIPLE STREAMS W14B-20"HE CIRC WELDS ISI-40 S
S 3
1 ONE 20 W14A-20"HE 20" X.375" ISI-42 S
5 2
(28)
(7)
(2)
(100)
MULTIPLE STREAMS W16-14"HE CIRC WELDS ISI-40 S
7 2
1 WO 14 W18-14"HE 14" X.375" S
7 1
W15-14"HE ISI-42 S
7 2
W17-18"HE S
7 2
1 THREE 29 7
RHR DISCHARGE A&B (37)
(19)
(5)
(100) MULTIPLE STPIAMS W29-10"GE CIRC WELDS iSI-41 S
18 9
2 WO 11 W19-10"GE 10" X.365" ISI-43 S
19 10 1
WO 16 2
THREE 26 (19)
(10)
(3)
(100) MULTIPLE STREAMS W29-14"GE CIRC WELDS ISI-41 S
7 4
1 ONE 10 W19-14"GE 14" X.375" ISI-43 S
12 6
2 WO 30 (67)
(34)
(9)
(100)
MULTIPLE STREAMS W30-14"GE CIRC WELDS ISI-41 S
38 19 2
ONE 6
W20-14"GE 14" X.375" ISI-43 S
29 15 3
WO 15 2
WO 24 2
THREE 26 (8)
(4)
(1)
(100)
MULTIPLE STREAMS W30-16"GE CIRC WELDS ISI-41 S
4 2
1 WO 50 W20-18"GE 16" X.375" ISI-43 S
4 2
1 ONE 25 (5)
(1)
(100)
SINGLE STREAM W22-14"GE CIRC WELDS ISI-43 S
5 5
1 ONE 20 14" X.375" FORM 17-4 5 71
b,o V,A p
MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL-EXAMINATION
SUMMARY
PAGE 3 OF__8 EXAM l NATION AMOUNT N
TA E
REMARKS ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION O
R D
'4 3.11&
(CONTINUED)
(40YR)
REQUIRED %
25.12 10YR RUNNING %
(' 75 CATEGORY C-G)
HPCI WATER SUCTION (26)
(13)
(3)
(50)
SINGLE STREAM W1-14"HE CIRC WELDS ISI-31A S
24 12 2
WO 15 C16-14"HE 14" X.375" ISI-31A S
2 1
1 THREE 23 HPCI STEAM (15)
(4)
(52)
SINGLE STREAM PS18-8"ED CIRC WELDS ISI-32 S
29 15 2
WO 13 8" X.500" 2
THREE 27 O
HPCI STEAM DISCH CIRC WELDS (27)
(14)
(4)
(52)
SINGLE STREAM RS2-16"HE 16" X.375" ISI-33 S
19 10 2
WO 14 1
THREE 21 RS2-18"HE 18" X.375" S
4 2
1 THREE 29 RS2-20"HE 20" X.375" S
4 2
CORE SPRAY A&B SUCTION (52)
(13)
(4)
(50)
MULTIPLE STREAMS W6-12"HE CIRC WELDS ISI-34 S
27 7
2 ONE 15 W10-12"HE 12" X.375" ISI-35 S
25 6
2 THREE 31 CORE SPRAY A&B DISCHARGE CIRC WELDS (76)
(19)
(5)
(50)
MULTIPLE STREAMS W7-10"GE 10" X.365" ISI-34 S
21 5
2 ONE 11 ISI-34A S
16 5
28 6
2 THREE 26 ISI-35A S
11 3
(12)
(3)
(1)
(50)
MULTIPLE STREAMS W7-8"ED CIRC WELDS ISI-34A S
6 2
1 ONE 33 W11-8"ED 8" X.500" ISI-35A S
6 1
FORM 17-4 57 9
f')
V U
[J
/'
'\\
(
l l
MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 4
op 8
EXAMINATION AMOUNT ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION N
OTAL
- N E lON REMARKS O
(40YR)
REQUIRED %
C5.116 (CONTINUED) 10YR RUNNING %
C5.12 1
(4)
(2) 1 (50)
MULTIPLE STREAMS W8-8"GE CIRC WELDS ISI-34A S
2 W12-8"GE 8" X.322" ISI-35A S
2 1
1 THREE 50 REACTOR WATER FROM SKIMMER SYSTEM (14)
(4)
(50)
SINGLE STREAM REW11-8"HE CIRC WELDS ISI-36 S
28 14 1
ONE 7
8" X.322 2
WO 21 y
1 THREE 29 S
RCIC WATER SUCTION (23)
(12)
(3)
(52)
SINGLE STREAM W5-6"HE CIRC WELDS ISI-38 S
21 11 1
ONE 8
6" X.280" 1
WO 17 C17-6"HE S
2 1
1 THREE 25 RCIC STEAM DISCHARGE (14)
(4)
(52)
SINGLE STREAM RS3-8"HE CIRC WELDS ISI-33 S
27 14 1
ONE 7
8" X.322" 2
WO 21 1
THREE 29 RHR SERVICE WATER (24)
(6)
(51)
SINGLE STREAM SW9-8"GE CIRC WELDS ISI-39 S
47 24 2
ONE 8
8" X.322" 2
WO 17 2
THREE 25 RHR SUCTION A&B (18)
(9)
(2)
(50)
MULTIPLE STREAMS W28-20"HE CIRC WELDS ISI-40 S
9 5
1 ONE 11 W27-20"HE 20" X.375" ISI-42 S
9 4
1 WO 22 FORM 17-4 5 7 9
h l
MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL-EXAMINATION
SUMMARY
PAGE 5 OF 8
l EXAMINATION AMOUNT ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION N
TAL
'N E lON REMARKS C5.11&
(CONTINUED)
(40YR; REQUIRED %
C5.12 10YR RUNNING %
CONTAINMENT SPRAY A&B (33)
(9)
(3)
(55)
15 4
1 ONE 11 W33-12"CE 12" X.375 S
18 5
1 WO 22 1
IHREE 33 (27)
(7)
(2)
(52)
17 4
1 ONE 14 W33-10"GE 10" X.365" S
10 3
1 THREE 29 25.20 PIPING WELDS OVER 15 IN.
NOMINAL WALL THICKNESS 25.216 CIRCUMFERENTIAL AND
- 2.5T MIN FROM 15. 2 2
(' 75 CATEGORY C-F)
MAIN STEAM A,B,C,&D (63)
(16)
(4)
(100) MULTIPLE STREAMS PS1-18"ED CIRC WELDS ISI-26 S,VOL 16 4
1 ONE 6
PS2-18"ED 18" X.937" ISI-27 S,VOL 15 4
1 WO 12 PS3-18"ED ISI-28 S,VOL 16 4
1 WO 19 PS4-18"ED ISI-29 S,VOL 16 4
1 IHREE 25 FORM 17-4$71
MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL-EXAMINATION
SUMMARY
PAGE 6
op 8
EXAMINATION AMOUNT,
"IM^h5 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL R D NO.
METHOD ITEMS 40 YR 10 YR C5.21&
(CONTINUED)
(40YR)
REQUIRED %
25.22 10YR RUNNING %
SUPPLY TO STEAM SEAL SYSTEM (7)
(2)
(100)
SINGLE STREAM PS7-8"ED CIRC WELDS ISI-30 S,VOL 7
7 2
ONE 29 8" X.593" (18)
(5)
(100)
SINGLE STREAM PS7-10"ED CIRC WELDS ISI-30 S,VOL 18 18 3
WO 17 10" X.593" 2
THREE 28 7
MAIN STEAM EQUALIZER HDR (21)
(5)
(100)
SINGLE STREAM PS30-18"EDB CIRC WELDS ISI-40 3,VOL 21 21 2
ONE 10 18" X.937" 1
WO 14 2
WREE 24 10" DRIPLEG CIRC WELDS ISI-40 3,VOL 2
2 SINGLE STREAM 10" X.594" FEEDWATER A&B (8)
(4)
(1)
(100) MULTIPLE STREAMS W2A-14"ED CIRC WELDS ISI-37 S,VOL 4
2 1
2 RHR DISCHARGE A&B (6)
(3)
(1)
(100)
MULTIPLE STREAMS W30-16"DB CIRC WELDS ISI-41 S,VOL 3
1 W20-16"DB 16" X.843" ISI-43 S,VOL 3
2 1
(' 75 CATEGORY C-G)
~
HPCI WATER DISCHARGE CIRC WELDS (47)
(24)
(6)
(51)
SINGLE STREAM W3-12" ED 12" X.687" ISI-31 S,VOL 7
3 1
ONE 4
12" X.843" S,VOL 38 20 2
ONE 12 3
WO 25 8" X.594" S,VOL 2
1 1
WREE 29 FORM 97-4575
N (d
V
( )s G
^
x.
MONTICELLO NUCLEAR GENERATING PLANT TABLE 5.10 TEN YEAR INTERVAL. - EXAMINATION
SUMMARY
VAGE.
7 OF 8 EXAMINATION AMOUNT ITEM COMPONENT OR SYSTEM IDENilflCATION DESCRIPTION NDE TOTAL
'NSPECTION REMARxs NO.
METHOD ITEMS 40 YR 10 YR PERtOO 25.30 PIPE BRANCH CONNECTIONS (401 0 REQUIRED %
10YA RUNNING %
25.31&
CIRCUMFERENTIAL AND 25.32
- LONGITUDINAL WELDS
(' 75 CATEGORY C-F)
EXAMINED SUPPLY 10 STEAM SEAL SYSTEM (4)
(1)
(2 )
MULTIPIE STREAMS 7
PSil-6"ED WELD 0 LEIS ISI-30 S
1 PS12-6"ED 18" X 6" S
1 PS13-6"ED S
1 1
1 THREE (100)
PS14-6"ED S
1 RHR SUCTION A&B (4)
(2)
(1)
(100) MULTIPLE STREAMS TW16-14"HE WELDOLETS ISI-40 S
2 1
1 IWO 50 IW18-14"HE 20" X 14" IW15-14"HE ISI-42 S
2 1
1W17-14"HE RHR DISCHAEGE B (1)
(100)
MULTIPLE STREAMS IW22-14"GE WELDOLET ISI-43 S
1 1
14" X 8" (CATEGORY C-G)
REACTOR WATER FROM SKIMMER SYSTEM (1) 1 (100)
SINGLE STREAM REWil-8"HE WELD 0LET ISI-36 S
1 1
1 IWO 100 18" X 10" FORM 17-4 57 8
SMA E
R 8
S T
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EE
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TP 41
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SN TN U
O
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M
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M 11 -
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T 211
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YRA N
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44 P
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SS S
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TN N O AL TI N
P A O
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IR U
C S
S ED TN N O AI L T N
P A O
G N A
IT INI C
TM I
F A A I
E E
TN N
N REX E
O O
NE D
I N
N EG-G RL AA C
ELV Y
CR M
R E
UE T
O G D N T S
G N N S
Y N
S E
I A
D S
OI R
T N L
D R
O A
I S
E L
L C A P W
E L
T EA N
T M W
CE N E U G
I E
TY N
O R P N
Y N
O I
I S
D O N P
T E N S
S E
O M
A R I
P A
V B
E MT O
N U S
M C
L C
I S
S E
U A
E M S D V P
P V
V A
E L L
M L
9 X R E A U
A E
P W V P
V 1
7 5
4 0
0 7
M.
1 E. O 1
2 M
. N 6
6 R
C C
O F
[-
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MONTICELLO NUCLEAR GENERATING PLANT TABLE 7.10 TEN YEAR INTERVAL - EXAMINATION
SUMMARY
PAGE 1
OF 1 ITEM COMPONENT OR SYSTEM IDENTIFICATION DESCRIPTION NDE TOTAL EXAMINATION INSPECTIONRUNNING REMARKS NO.
METHOD tTEMS AMOUNT & EXTENT PERIOD l
EXAMINATION CATEGORY C-H, ALL PRESSURE RETAINING COMPONENTS C7.10 PRESSURE VESSELS PRESSURE IWC-5221 VT-2 PRESSURE RETAIN-100%
- SYSTDI 2RESSURE C7.20 PIPING RETAINING ING BOUNDARY TEST PERFORMED C7.30 PUMPS BOUNDARY BY PLANT EACH C7.40 VALVES INSPECTION PERIOD 7
a C7.11 PRESSURE VESSELS PRESSURE IWC-5222 VT-2 PRESSURE RETAIN-100%
- SYSTEM HYDRO-C7.21 PIPING RETAINING ING BOUNDARY STATIC TEST C7.31 PUMPS BOUNDRAY PERFORMED BY C7.41 VALVES PLANT EACH INSPECTION INTERVAL FORM 17-4571
i l
9 ASME SECTION XI NONDESTRUCTIVE EXAMINATION PROGRAM - CIASS 3
/ 'l PROGRAM PERIOD:
2nd Ten Year Interval
\\s J June 30, 1981 to June 29, 1991 ASNE SECTION XI:
1977 Edition through and including the Summer 1978 Addenda NOTES:
- 1. The classification diagrams in Section 6 of this report identify the systems that are required to be examined in accordance with IWD-2000 (Quality Group C).
- 2. The scope of the inspection program for Class' 3 components is' based on the classification of the plant's inspection boundaries and exemptions as allowed for in IWD-2600 and IWD-5200. The inspection program will conform to IWD-2400.
- 3. Visual examination will be conducted for evidence of component leakage, structural distress, or corrosion when the system is undergoing either a system inservice test, component functional test, or a system pressure test.
- 4. Supports and hangers for components will be visually examined to detect any loss of support capability or evidence of inadequate restraint.
s_,)
- 5. Repairs will be performed in accordance with the applicable requirements of the latest Edition and Addenda of the ASNE Code,Section XI.
- 6. INSPECTION PERIODS:
ONE
- June 30, 1981 to October 29, 1984 TWO
- October 30, 1984 to February 28, 1988 THREE - February 28, 1988 to June 29, 1991 o
O 1-71
SECTION 2 PRESSURE TESTING PROGRAM
'\\s ASME Section XI Pressure Testing Program ASME Code Edition and Addenda:
1977 Edition through and including Sununer 1978 Addenda Program Period:
June 30, 1981 through June 29, 1991 APPLICABLE ASME TEST TEST CODE CIASS TYPE FREQUENCY R$UEST FOR RELIEF 1
Leakage Refueling 30 (Quality Group A)
Hydrostatic 10 years 30 2
Pressure 10 years 30, 31 (Quality Group B) 3 Pressure 10 years 30 (Quality Group C)
Except as noted in the Requests for Relief, pressure tests will confonn to IWA-5000, IWB-5000, IWC-5000, and IWD-5000.
ASME Code Class boundaries are shown on the figures in Section 6.
These figures do not include small instrument, leak test, vent, and drain lines.
t e
O 2-1
i SECTION 3 ' INSERVICE TESTING OF PUMPS AND VALVES ASME Section XI Pump and Valve Testing Program ASME Code Edition and Addenda:
1977 Edition through and including Sununer 1978 Addenda Program Period:
June 30, 1981 through June 29, 1991 Pump tests are summarized on the table on page 3-2.
Valve tests are summarized on the table beginning on page 3-3.
Requests for Relief are included in Section 4.
Key for Pump Testing Table l
M Monthly Not Required (constant speed drive or fixed resistance system)
NR Not Applicable (sealed bearings)
NA RR - See Request for Relief r
Key for Valve Testing Table Q
Quarterly NR
- Not Required RR
- See Request for Relief Cold Shutdown, Not More Often Than Quarterly CSIQ O'
IWV-3510 - In Accordance With the Requirements of Paragraph IRV-3510 IWV-3610 - In Accordance With the Requirements of Paragraph IWV-3610 i
9
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O Table 3-1 ASME Code Section XI Pump Testing Applicable ASIE Code Pump Description Class Test Parameter
- Requests for Relief b
N Pi AP Q
V
/p Tb 11 Emergency Service Water 3
NA RR 2, 3, 27, 45 12 Emergency Service Water 3
NA RR 2, 3, 27, 45 11 Standby Liquid Control 2
hR M
M M
M M
RR 2, 3, 44, 45 12 Standby Liquid Control 2
hR M
M M
M M
RR 2, 3, 44, 45 11 Core Spray 2
hR M
M M
M NA RR 2, 3, 45 12 Core Spray 2
hR M
M M
M NA RR 2, 3, 45 11 Residual Heat Removal 2
NR M
M M
M NA RR 2, 3, 45 12 Residual Heat Removal 2
NR M
M M
M NA RR 2, 3, 45 13 Residual Heat Removal 2
NR M
M M
M NA RR 2, 3, 45 14 Residual Heat Removal
~2 NR M
M M
M NA RR 2, 3, 45 y
11 RHR Service Water 3
hR RR RR M
M NA RR 1, 2, 3, 45 N
12 RIR Service Water 3
hR RR RR M
M NA RR 1, 2, 3, 45 13 RIR Service Water 3
hR RR RR M
M NA RR 1, 2, 3, 45 14 RIE Service Water 3
NR RR RR M
M NA RR 1, 2, 3, 45 High Pressure Coolant Injection 2
M M
M M
M M
RR 2, 3, 45 Reactor Core Isolation Cooling 2
M M
M M
M M
RR 2, 3. 45
- Refer to Table IWP-3100-1, Inservice Test Quantities L/P refers to lubricant level or pressure
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'R.J Table 3-2 ASME Code Section XI Valve Testing FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Descriptgn Code Class Category Frequency Test For Relief hbin Steam A0 2-80A 80-A hbin Steam Isolation 1
A Q
Full Stroke-Time 28 hbin stean AD 2-80B 80-B bbin Steam Isolation 1
A Q
Full Stroke-Time 28 hbin Steam A0 2-80C 80-C bhin Steam Isolation 1
A Q
Full Stroke-Time 28 bbin stean A0 2-50D 80-D hhin Steam Isolation 1
A Q
Full Stroke-Time 28 bbin Steam A0 2-86A 86-A bbin Steam Isolation 1
A Q
Full Stroke-Time 28 hbin Steam A0 2-86B 86-B bbin Steam Isolation 1
A Q
Full Stroke-Time 28 hhin Y
Steam A0 2-86C 86-C h!ain Steam Isolation 1
A Q
Full Stroke-Time 28 hbin Steam A0 2-86D 86-D Main Steam Isolation 1
A Q
Full Stroke-Time 28 hbin Steam M0-2373 74 Steamline Drain Isolation 1
A Q
Full Stroke-Time 28 bhin Steam M0-2374 77 Steamline Drain Ir.olatior, 1
A Q
Full Stroke-Time 28 Main Steam RV-2-71A RV-71-A hbin Steam Safety Relief 1
C IhV-3510 Setpoint hbin Steam RV-2-71B RV-71-B bbin Steam Safety Relief 1
C IhV-3510 Setpoint hbin Steam RV-2-71C RV-71-C Main Steam Safety Relief 1
C IhV-3510 Setpoint Bhin Steam RV-2-71D RV-71-D bhin Steam Safety Relief 1
C IhV-3510 Setpoint bbin Steam RV-2-71E None Main Steam Safety Relief 1
C IhV-3510 Setpoint Lin Steam '
RV-2-71F None Main Steam Safety Relief 1
C IhV-3510 Setpoint Main Steam RV-2-71G None Main Steam Safety Relief 1
C IhV-3510 Setpoint hbin Steam W-2 iH None min Steam Safety Relief 1
C IhV-3510 Setpoint
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FSAR Valve Valve Applicable A9E Valve Test Request.
System Number Number Description Code Class Category Frequency Test For Relief Main Full Stroke Steam FCV-7682 FCV 27 Recombiner Steam Supply 2
B CS1Q Time 28 Main Reactor IIead Seal Full Stroke Steam CV-2369 17 Leak-Off Valve 2
B Q
Time 28 Main Reactor IIcad Seal Full Stroke Steam CV-2370 18 Leak-Off Valve 2
B Q
Time 28 Main Reactor llead Full Stroke Steam CV-2371 20 Vent Valve 2
B Q
Time 28 Main Reactor IIcad Full Stroke Steam CV-2372 21 Vent Valve 2
B Q
Time 28 w
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FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief FW FW 91-1 27B FW Inlet Check Valve 2
C RR RR 7
FW FW 91-2 27A FW Inlet Check Valve 2
C RR RR 7
FW FW 94-1 96B Outboard Isolation 1
A, C RR RR '
8 FW FW 94-2 96A Outboard Isolation 1
A, C RR RR 8
y FW FW 97-1 28B Inboard Isolation 1
A, C RR RR 8
FW FW 97-2 28A Inboard Isolation 1
A, C RR RR 8
Recirc CV-2790 39 Rx Water Sample Isolaticn 2
A Q
Full Stroke-Tite 10, 28 Recirc CV-2791 40 Rx Water Sample Isolaticn 2
A Q
Full Stroke-Tire 10, 28 Recirc 50-2-43/. 43A Recirc Suction 1
B CSIQ Full Stroke-Tire 28 P.ecirc h0-2-431 43B Recirc Suction 1
B CSIQ Full Stroke-Tire 28 28 Recirc 30-2-53/
53A Recirc Discharge 1
B CSIQ Full Stroke-Tire 28 Recirc hD-2-53I 53B Recirc. Discharge 1
B CSIQ Full Stroke-Tire Recirc 30-2-541. 54A Recirc Disch. Bypass 1
B Q
Full Stroke-Tire 28 28 Recirc bO-2,54I 54B Recirc Disch. Bypass 1
B Q
Full Stroke-Tire Recirc h0-2-651, 65B Recirc Loop Crosstic 1
B CSIQ Full Stroke-Tire 28 Recirc FO-2-65I 65A Recirc Loop Crosstie 1
B CSIQ Full Stroke-Tire 28 g
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FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Recirc h0-2-66A 66B Recirc Crosstie Bypass 1
B CSIQ Full Strcke-Time 28 Recirc h0-2-66E 66A Recirc Crosstie Bypass 1
B CSIQ Full Stroke-Tire 28 Block Valve on Recirc None Uoner Seal Leakoff 2
B RR Full Stroke-Tine 5,2B Block Valve on Recirc None Uoper Seal Leakoff 2
B RR Full Stroke-Tine 5,28 mm RV-1990 72A Pump Suction Relief 2
C IhV-3510 Setpoint mm RV-1991 72B Pump Suction Relief 2
C IhV-3510 Setpoint pJa RV-1992 72C Pump Suction Relief 2
C IhV-3510 Setpoint
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mm RV-1993 72D Pump Suction Relief 2
C IhV-3510 Setpoint RIR RV-2004 35A Pump Disch Relief 2
C GV-3510 Setpoint mm RV-2005 35B Pump Disch Relief 2
C IhV-3510 Setpoint mm None ik Shell Side Relief 2
C IhV-3510 Setpoint pJa None ik Shell Side Relief 2
C IhV-3510 Setpoint mm A0-10-464 A0-46A LPCI Loop Check 1
A, C CSIQ Exercise Rim A0-10-46B A0-46B LPCI Loop Check 1
A, C CSIQ Exercise Rim RIR-2-1 48A RIR Pump Discharge Check 2
C Q
Exercise pj a RIR 2' 48B RIR Pump Discharge Check 2
C Q
Exercise mR RIR-2-3 48C RER Pump Discharge Check 2
C Q
Exercise pJg RER-2-4 48D RIR Pump Discharge Check 2
C Q
Exercise l
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FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief RIR RIR-21 29 Rx Head C1g Check 1
C RR RR 7
RIR RIR-8-1 19A RIR Min Flow Check 2
C RR RR 7
RIR 50-1988 FO-15A Shutdown C1g Suction 2
B CSIQ Full Stroke-Time 28 RIR 50-1989 FO-15B Shutdown C1g Suction 2
B CSIQ Full Stroke-Tima 28
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RIR h0-2006 50-39A Disch to Torus 2
A Q
Full Stroke-Time 28 RIR 50-2007 h0-39B Disch to Tonis 2
A Q
Full Stroke-Time 28 RIR h0-2008 hD-34A Torus C1g Inlet 2
A Q
Full Stroke-Time 28 RIR 50-2009 BD-34B Torus C1g Inlet 2
A Q
Full Stroke-Time 28 RIR 50-2010 50-38A Torus Spray 2
A Q
Full Stroke-Time 28 RIR hD-2011 hD-38B Torus Spray 2
A Q
Full Stroke-Tima 28 RIR 50-2012 hD-27A LPCI Injection 2
B Q
Full Stroke-Tim:
28 RIR 50-2013 hD-27B LPCI Injection 2
B Q
Full Stroke-Tim:
28 RIR 50-2014 hD-25A LPCI Injection 1
A Q
Full Stroke-Time 28 RIR 50-2015 BD-25B LPCI Injection 1
A Q
Full Stroke-rim:
28
LJ LJ R.)
FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Cont Spray RIR 50-2020 50-26A Outboard Isolation 2
A 0
Full Stroke-Tine 28 Cont Spray RIR 50-2021 50-26B Outboard Isolation 2
A 0
Full Stroke 'ine 28 Cont Sprav RIR 50-2022 h0-31A Inner Iso'1ation 2
A 0
Full Stroke-Tine 28 Cont Spray RIR 50-2023 50-31B Inner Isolation 2
A 0
Full Stroke-Tir e 28 l
RIR 50-2026 50-33 Ilead Sprav Isolation 1
A CSIQ Full Stroke-iire 28
[
RIR h0-2027 50-32 IIead Spray Isolation 1
A CSIQ Full Stroke-Tire 28 RIR 50-2029 50-18 Shutdown C1n Isolation 1
A CSIQ Full Stroke-Tine 28 28 RIR 50-2030 50-17 Shutdown Clc Isolation 1
A CSIQ Full Stroke-Tire RIR 50-2032 h0-57 Disch to Waste Surne 2
B Q
Full Stroke-Tine 28 RIR CV-1994 CV-153A RIR Pump !!in Flow 2
B Q
Full Stroke-Tire 28 RIR
,CV-1995 CV-153B RIR Pump hiin Flow 2
B Q
Full Stroke-Tire 28 RIR CV-1996 CV-153C RIR Pump hiin Flow 2
B Q
Full Stroke-Tire 23 RIR CV-1997 CV-153D RIR Pump Slin Flow 2
B Q
Full Stroke-Tire 28 RIR RV-2023 44 Ilead Spray Line Relief 2
C IhV-351f Setpoint RIR RIR 2 19B RIR Sfin Flow Check 2
C RR RR 7
RIR RIR-SN-1 7 182 SW Emere Supply to RIR 2
C RR RR 11 RIE RV-2031 40 RIR Shutdown Cle Relief 2
C IhV-3510 Setpoint
m
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J FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Core A0-14-13A A0-13A Loop Inj. Check 1
A,C CSIQ Exercise Sprav A0-14-13E A0-13B Loop Inj. Check 1
AC CSIQ Exercise E$$y 10-1753 FD-12A Core Spray Injection 1
A Q
Full Stroke-Tine 28 Core Spray 10-1754 MO-12B Core Spray Injection 1
A Q
Full Stroke-Time 28 Core 10-1751 10-11A Core Spray Injection 2
n n
Full Stroke-T b 28 Spra,,
w a
[3rv Full Stroke-Time 28 10-1752 10-11B Core Spray Injection 2
B n
h#8 RV-1745 20A Disch Line Relief 2
C IhV-3510 Setpoint h ".
RV-1746 20B Disch Line Relief 2
C IhV-3510 Setpoint
^
Core CS-9-1 10A Pump Disch Check 2
C Q
Exercise cprny h,".
CS-9-2 10B Pump Disch Check 2
C Q
Exercise h ",
FD-1741 7A Core Spray Suction 2
B Q
Full Stroke-Tim 3 28 30-1742 7B Core spray Suction 2
B Q
Full Stroke-Tim:
28 y
C're'Spr o
CS-10-1 18A Min Flow Block 2
E NR Valve Lineup Sprnv Core Fu'1 Stroke-Tima 28 1
50-1749 FD-26A Test Line to Torus 2
B Q
sprnv h "y 3D-175d m-26B Test Line to Toms 2
B Q
Full Stroke-Timy 28 m
RV-2056 66 Relief Valve 3
C IhV-3510 Setpoint HPCI HPCI-18 130 C1g Water Return Check 3
C Q
Exercise y-
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,.J FSAR Valve Valve Applicable A9tE Valve Test Request System Ntsnber No.
Description Code Class Category Frequency Test For Relief IPCI 50-2068 30-19 Purnp Discharge Isol 2
B Q
Full Stroke-Tim 28 HFCI AO-23-1E AO-18 C1g Wtr Disch Check 2
C CSIQ Exercise IEI 50-2071 FO-21 Test return to CST 2
B Q
Full Stroke-Timu 28 IPCI m-2067 h0-20 Coolant Ptznp Disch.
2 B
Q Full Stroke-Timu 28 IPCI
'CV-2065 41 Min Flow Bypass 2
B Q
Full Stroke-Timu 28 IPCI IPCI-42 62 Min Flow Bypass Check 2
C RR RR 7
u IFCI RV-2064 34 Relief Valve 2
C IhV-3510 Setpoint IFCI HPCI-32 32 CST Suction Check 2
C Q
Exercise IPCI h0-2063 h0-17 CST Suction 2
B Q
Full Stroke-Tise 26 ffCI 50-2062 50-57 Torus Suction 2
B Q
Full Stroke-Tiie 26 IPCI HPCI-31 61 Torus Suction Check 2
C RR RR 7
IPCI BO-2061 FO-58 Torus Suction 2
B Q
Full Stroke-Tire 28 IPCI 50-2034 50-15 Steam Supply Isolation 1
A Q
Full Stroke-Tire 28 IPCI 50-2035 B0-16 Steam Outboard Isolatior 1
A Q
Full Stroke-Tide 28 IPCI 50-2036 50-14 Turbine Steam Supply 2
B Q
Full Stroke-Tire 28 ffPCI 10-7 10 Turbine Stop Valve 2
B Q
Full Stroke 10 IIPCI 10-8 10 Turbine Control Valve 2
B Q
Full Stroke-Tim 28 Turbine Exhaust HPCI HPCI-60 None 2
C RR RR 7
Vacuum Breaker
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p FSPR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Cooling Water IPCI PCV-3492 PCV-50 Supply Cont.
2 B
RR RR Cooling Water ilPCI HPCI-20 131 Return Check 2
C Q
Exercise llPCI IPCI-14 56 Ex. Line Drain Pot Check 2
A, C RR RR 7
IIPCI IPCI-15 45 Ex. Line Drain Pot Check 2
C RR RR 7
IPCI IPCI-9 65 Turbine Ex. Line Check 2
A, C Q
Exercise IIPCI IIICI-10 12 Ex. Line Stop Check 2
C Q
Exercise Y
IPCI HPCI-65 None Vac. Bkr Check 2
C RR RR 7
0 IPCI IIPCI-7I None Vac. Ekr Check 2
C RR RR 7
IPCI PSD-2038 None Ex. Line Rupture Disc 2
D NR RCIC 50-2096 50-2096 Cooling Water to Cond.
2 B
Q Full Stroke-Time 28 RCIC RV-2097 RV-2097 Relief Valve 3
C IW-351C Setpoint RCIC RCIC-14 None Condenser Cond Pump Disc 1 2
C Q
Exercise
_RCIC RCIC-17 None Vac Pump Disch Check 2
C RR RR 7
RCIC RCIC-9 None Turbine Exhaust Check 2
A,C Q
Exercise RCIC RCIC-10 None Steam Exh Stop Check 2
C Q
Exercise RCIC RCIC-57 None Vac Brkr Check 2
C RR RR 7
RCIC RCIC-59 None Vac Brkr Check 2
C RR RR 7
RCIC PSD-2089 PSD-2089 Rupture Disc 2
D NR
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v FSAR Valve Valve Applicable ASME Valve Test Request System M =her No.
Description Code Class Category Frequency Test For Relief RCIC 50-2075 MD-2075 Steam Supply Isolation 1
A Q
Full Stroke-Tin 28 RCIC h0-2076 h0-2076 Steam Supply Isolation 1
A Q
Full Stroke-Tim 28 RCIC h0-2078 FO-2078 Steam Supply to Turbine 2
B Q
Full Stroke-Tim 28 RCIC RCIC-7 None Throttle Trip Valve 2
B Q
Full Stroke 10 RCIC 10 None RCIC Goveming 2
B RR RR RCIC PCV-2092 PCV-2092 Condenser Press Cont 2
B RR RR RCIC RCIC-16 None Vac Pump Disch Check 2
A, C RR RR 7
RCIC h0-2100 10-2100 Inboard Torus Suction 2
B Q
Full Stroke-Tim 28 RCIC RCIC-31 None Check Valve to Torus 2
C RR RR 7
RCIC 50-2101 FO-2101 Outboard Torus Suction 2
B Q
Full Stroke-Tim 28 RCIC RCIC-41 None Check Valve to CST 2
C Q
Exercise RCIC h0-2102 h0-2102 CSF Suction 2
B Q
Full Stroke-Tim 28 RCIC RV-2103 RV-2103 Q Suction Line Relief 2
C IhV-351f.Setpoint RCIC CV-2104 CV-2104 Min Flow Bypass 2
B Q
Full Stroke-Time 28 RCIC RCIC-37 None Min Flow Bypass Check 2
C RR Ta 7
PCIC h0-2106 50-2106 Pump Discharge 2
B Q
Full Stroke-Tin 23 RCIC 50-2107 50-2107 Pump Discharge 2
B.
Q Full Stroke-Timc 28 RCIC A0-13-22 AO-13-22 Punp Disch Check 2
C CSIQ Exercise
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Applicable A9E Valve Test Request FSAR System Number No.
Description Code Class Category Frequency Test For Relief Valve Valve RCIC FO-2110 50-2110 Test Return to Cond Stor 2
B Q
Full Stroke-Time 28 2
C Q
Exercise SBLC XP-3-1 43A Pump Disch Check 2
C Q
Exercise SBIf XP-3-2 43B Pump Disch Check 13 SBLC XP-6 16 Outboard Isolation Check 1
A, C RR R
13 SBLC XP-7 17 Inboard Isolation Check 1
A, C RR 1R 2
'C IhV-35105etpoint U
SBlf RV-11-39A 39A Relief Valve 2
C IhV-35105ctpoint SBlf RV-11-393 39B Relief Valve 2
D IhV-3610 ctuation SBLC 11-14A 14A Explosive Actuated Valve 2
D IhV-3610 Actuation SBLC 11-14B 14B Explosive Actuated Valve 2
A Q
Full Stroke-Tilne 10,28
2 A
Q Full Stroke-Ti:ne 10,28 TIP TIP 1-2 None Bal a,
- 3 TIP Isolation 2
A Q
Full Stroke-Ti me 10,28 TIP TIP 1-3 None Ball Valve s ation 2
D IWV-3610 Actuation TIP TIP 2-1 None y,
2 D
IWV-3610 Actuation hefVaf
,TIP TIP 2-2 None olation 2
D IWV-3610 Actu ation hTIP TIP TIP 2-3 None 2
A, C RR RR 26 l
TIP TIP 3 None TIP System Purge Check
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FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief CRD CV-3-32A CV-32A Scram Disch Volume Vent 1
B Q
Full Stroke-Tim 28 CRD CV-3-32B CV-32B Scram Disch Volume Vent 1
B Q
Full Struke-Tim 28 CRD CV-3-33 CV-33 Scram Disch Volume Drain 1
B Q
Full Stroke-Tim 28 bCram DisCh Volume CRD RV-3-34 34 Relief Valve 2
C IhV-3510 Setpoint CRD CRD-114 114 Scram Riser Check 2
C RR RR '
9 Accumulator Charging CRD CRD-115 115 Water Check 2
C RR RR 9
CRD CRD-138 138 Cooling Water Check 2
C RR RR 9
w CRD CV-126 CV-126 Inlet Scram Valve 1
B RR RR 9
CRD CV-127 CV-127 Outlet Scram Valve 2
B RR RR 9
RHR SW CV-1728 CV-1728 RHR SW Control Valve 3 3
B Q
Full Stroke-Tim 28 RIR SW' CV-1729 CV-1729 RHR SW Control Valve 3
B Q
Full Stroke-Tim 28 RHR-SW RHR-SW RIR SW Pump RIR SW 1-1 1-1 Disch Check 3
C Q
Exercise mm-SW mR-SW m m SW Pump RHR SW 1-2 1-2 Disch Check 3
C Q
Exercise
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G' FSAR Valve Valve Applicable ASME Valve Test Request Systen Number No.
Description Code Class Category Frequency Test For Relief RIR-SW-RIR-SW RIR Si Pump RIR SW 1-3 1-3 Disch ch
(
c n
Exercise RIR-SW RIR-SW RIR SW Pump RIR SW 1-4 1-4 Disch Ch 3
C 0
Frerrico I
Pump Motor Cooling RIR SW RV-3038 RV-3038 Line RV 3
C Thv-T;10Rotnnint Pump Motor Cooling RIR SW RV-3039 RV-3039 Line RV 3
c Thv-T;1 nqntnnint Pump Motor Cooling RIR SW FCV-3004 PCV-3004 Line PCV 3
R RR RR Pump lbtor Cooling RIR SW PCV-3005 PCV-3005 Line PCV 3
R RR RR RIR SW AV-3147 AV-3147 Punn Disch Air Vent 3
C 0
Pynrcico RIR SW AV-3148 AV-3148 Pumn Disch Air Vent 3
C 0
Prercien RIR SW AV-3149 AV-3149 Pumn Disch Air Vent 3
C 0
Prercise Rim SW AV-3150 AV-3150 Punn Disch Air Vent 3
C 0
Prorci se RIR SW SW-21-1 None Motor Cooline Line Check 3
C RR 1R 7
RIR SW SW-21-2 None Motor Cooline Line Check 3
C RR 1R 7
RIR SW RV-3202 None RIR IN RV 3
C IhV-35103etnnint RIR SW RV-3203 None RIR IN RV 3
C IkV-35105etnoint Emerg Serv Wtr ESW-4-1 None S.W. Check Valve 3
C RR M
7 Emerg Serv Ntr ESW-4'2 None S.W. Check Valve 3
C RR m
7 Emerg Serv Ntr SW-101 None E.S.W. Check Valve 3
C RR M
7 Emerg Serv Ntr SW-102 None E.S.W. Check Valve 3
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V FSAR l
Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Emerg Serv Wtr SW-103 None E.S.W. Check Valve 3
C RR RR 7
Emerg Serv Wtr SW-104 None E.S.W. Check Valve 3
C RR RR 7
Emerg Serv Wtr ESW-1-1 ESW-1-1 Pumo Check Valve 3
C 0
Frercise Emerg Serv Wtr ESW-1-2 ESW-1-2 Pumn Check Valve 3
C 0
Exercise Emerg Serv Wtr SW-16 SW-16 E.S.W. Check Valve 3
C RR RR 7
Emerg Serv Wtr SW-18 SW-18 E.S.W. Check Valve 3
C RR RR 7
Y Emerg E
Serv Wtr AV-3155 AV-3155 Pump Disch Air Vent 3
C O
Exercise Emerg Serv Wtr AV-3156 AV-3156 Pump Disch Air Vent 3
C 0
Exercise Primag Containm.
A0-2377 None Cont. Purge Isolation 2
A O
Full Stroke-Timy 28 Primary 2ntainm.
A0-2378 None Torus Purge Isolation 2
A 0
Full Stroke-Tim?
28 Primary
- ontainm.
A0-2379 None Tonis Vac Bkr Isolation 2
A 0
Full Stroke-Tim 28
=
Primary 28
- ontainm.
A0-2380 None Torus Vac Bkr Isolation 2
A O
Full Stroke-Tim?
Primary
- ontainm.
A0-2381 None Drywell Purge Isolation 2
.A 0
Full Stroke-Tins 28 Primary 2ntainm.
A0-2383 None Torus Vent Isolation 2
A 0
Full Stroke-Tim s 28 Primary intainm.
A0-2386 None Drvwell Vent Isolation 2
A 0
Full Stroke-Tim 28 Primary intainm.
A0-2387 None Drywell Vent Isolation 2
A 0
Full Stroke-Tins 28
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FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Primary
- Containm, A0-2896 None Torus Vent Isolation 2
A O
Full Stroke-Timo 28 Primary Sec Cont to
- Containm, DhV-8-1 None Torus Vac Bkr 2
A. C Q
hercise Primary Sec Cont to
- Containm, DhV-8-2 None Toms Vac Bkr 2
A. C 0
hercise Primary
- Containm, CV-7436 None N9 Ptmpback Isolation 2
A RR 1R 28 Primary Containm.
CV-7437 None N, Pumpback Isolation 2
A RR 1R 28 Primary Containm.
CV-2384 None Torus Vent Isolation 2
A 0
Full Stroke-Time 28 Primary w
,L Containm.
CV-2385 None Drvwell Vent Isolation 2
A 0
Full Stroke-Time 28 Primary Containm.
CV-3267 None Torus N, Makeup Iso.
2 A
0 Full Stroke-Time 28 Primary
- Containm, CV-3268 None Drywell N2 Makeug Iso.
2 A
Q Full Streke-Time 28 Primary Containm.
CV-3269 None Cont N, Makeup Iso 2
A 0
Full Stroke-Time 28
~
Primary Containm.
CV-3305 None Drywell 02 Analy Iso 2
A Q
Full Stroke-Time 28 Primary Containm.
CV-3306 None Drywell 07 Analy Iso 2
A Q
Full Stroke-Time 28 Primary Containm.
CV-3307 None Drywell 02 Analy Iso 2
A Q
Full Stroke-Timc 28 Primary 2
A Q
Full Stroke-Time 28 Containm.
CV-3308 None Drywell 02 Analy Iso Primary Containm.
CV-3309 None Drywell 02 Analy Iso 2
A Q
Full Stroke-Time 28 Primary Containm.
CV-3310' None Drywell 02 Analy Iso 2
A Q
Full Stroke-Time 28 Primary Containm.
CV-3311 None Drywell 02 Analy Iso 2
A Q
Full Stroke-Timo 28 Primary Containm.
CV-3312 None Drywell 02 Analy Iso 2
A Q
Full Stroke-Time 28
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l FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief Primary "nntninm.
CV-3313 None Dnvell 0, Analy Iso 2
A O
Full Stroke-Time 28 Primary
-nntninn_
CV-3314 None Drvwell 0, Analy Iso 2
A O
Full Stroke-Time 28 Primary
'nntninm.
CV-7440 None Torus to Drvwell Ny Iso 2
A O
Full Stroke-Time 28 2ond Serv System IN-58 None Drwell Demin Etr Iso 2
A RR RR 1PCCW 50-1426 50-1426 Drwell RPCCW Isolation 2
A CSIO Full Stroke-Tin:
28 1BCCW RPCC-15 Ncne Dnvell RPCCW Isolation 2
A. C RR RR 6
Y 1h m 50-2397 5D-2397 Pump Suction Isolation 1
A O
Full Stroke-Time 28 5
th m 50-2398 50-2398 Pump Suction Isolation 1
A Q
Full Stroke-Tim 28 Liquid ladwaste AD-2541A None Dnvell Floor Dm Sep Iso 2
A O
Full Stroke-Tim:
28 Liquid bdwaste A0-2541B None Drwell Floor Drn Snp Isc 2
A Q
Full Stroke-Time 28 Liquid Radwaste A0-2561A None Drvwell Equip Sump Iso 2
A Q
Full Stroke-Time 28
.iquid bdwaste A0-2561B None Drwell Equip Sump Iso 2
A Q
Full Stroke-Tim 28 uel Pool 21g 5 C1p PC-20-1 None Fuel Storage Pool Check a
C RR RR 7
- uel Pool 3
C RR RR 7
- Ig 5 C1p PC-20-2 Nc.ne Fuel Storage Pool Check
- omp (ir CV-1478 CV-1478 Drwell Comp Air Iso 2
A CSIQ Full Stroke-Time 28
- omp
.Gr CV-7956 None Torus Inst Air Iso 2
A Q
Full Stroke-Tine 28
- omp
.kir AS-39 None Service Air Iso 2
A RR RR
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FSAR Valve Valve Applicable ASME Valve Test Request System Ntrber No.
Description Code Class Category Frequency Test For Relief RX INST.
X-27A None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-27B None Excess Flow Check Valve 1
A, C RR RR 39 l
RR 39 RX INST., X-27C None Excess Flow Check Valve 1
A,C RR RX INST.
X-28A None Excess Flow Check Valve 1
A, C l
RR RR 39 RX INST.
X-28B None Excess Flow Check Valve 1
A, C RR RR 39 RX. INST.
X-28C None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-28D None Excess Flow Check Valve 1
A, C RR RR 39 G
RX INST.
X-28E None Excess Flow Check Valve 1
A,C RR RR -
39 RX. INST.
X-29A None Excess Flow Check Valve 1
A, C RR RR 39 l Excess Flow Check Valve 1
A, C RR RR 39 -
RX INST.
X-29B None RX INST.
X-29C None Excess Flow Oleck Valve 1
A, C RR RR 39 RX. INST.
X-29D None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-31A Mone Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-31B None Excess Flow' Check Valve 1
A, C -
RR RR 39 RX INST.
X-31D None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-31E None Excess Flow Check Valve 1
A, C RR RR 39
' RX INST.
X-31F None Excess Flow Check Valve-1 A,- C RR RR 39
FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief RX INST.
X-32A None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-32B None
A, C RR RR 39 RX INST.
X-32D None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-32E None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-32F None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-33A None Excess Flow Check Valve 1
A, C RR-RR 39 RX INST.
X-33B None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-33C None Excess Flow Check Valve 1
A, C -
RR RR-39 lRXINST.
X-33D None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-33E None Excess Flow Check Valve 1
A, C RR RR 39 -
RX INST.
X-33F None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40A-A None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40A-I.
~ 1 A, C RR RR 39 RX INST.
X-40A-C None Excess Flow-Check Valre 1
A-C-RP.
ER 39 RX INST.
X-40A-1)
None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40A-1!
None Excess Flow Check Valve 1
A, C RR FR 39 RX INST.
X-40 A-;:
None Excess Flow Check Valve-1 A, C RR RR 39 RX INST.
X-40B-A None Excess Flow Check Valve 1
A, C RR RR 39
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FSAR Valve Valve Applicable ASME Valve Test Request System Number No.
Description Code Class Category Frequency Test For Relief 39 RX INST.
X-40B-B None Excess Flow Check Valve 1
A, C RR RR RX INSI'.
X-40B-C None Excess Flow Check Valve 1
A, C RR RR 39 i
RX INST.l X-40B-D None i Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40B-E None Excess Flow Check Valve 1
A, C RR RR 39 i
RX INST.
X-40B-F None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40C-A None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40C-B None
A, C RR RR 39 w
RX INST.
X-40C-C None Ewess Flow Check Valve 1
A, C RR RR-39 RX INST.
X-40C-D None Excess Flow Check Valve 1
A, C RR RR 39 1
A, C RR RR 39 -
RX INST.
X-40C-E None l Excess Flow Check Valve RX INST.
X-40C-F None Excess Flow Check Valve 1
A, C RP.
RR 39 RX INST.
X-40D-A None Excess Flow Check Valve 1
A, C RR RR 39 RX INST.
X-40D-B None Excess Flow Check Valve 1
A, C RR RR 39
~
RX INST.
X-40D-C None Excess Flow Check Valve 1
A, C-RR RR 39 RX INST.
X-40D-D None Excess Flow Check Valve 1
A,-C RR RR 39 RX INST.
X-40D-E None Excess Flow Check Valve 1
A, C RR RR 39 f
i RX INST.
X-40D-F None Excess Flow Check Valve-1 A, C RR RR 39
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O FSAR l
Valve Valve Applicable ASME Valre Test Request Systen Nu-ber No.
Description Code Class Category! Frequency Test For Relief i
RX INST.
X-49A None Excess Flow Geck Valve 1
A, C RR-l RR 39 I
RXINST.lX-49B None Excess Flow Geck Valve 1
! A, C RR RR 39 RX INST. X-49C None Excess Flow Check Valve
'l A, C l
RR RR 39 RX INST.
X-49D None Excess Flow Geck Valve i
X-49E None Excess Flow Geck Valve 1
A, C RR RR 39 RX INST.
X-49F None Excess Flow Geck Valsv 1
A, C RR RR 39 RX INST.
X-50A None Excess Flow Check Va' te 1
A, C RR RR 39 Yy RX INST.
X-503 None Excess Flow Geck Valve 1
A, C -
RR RR-39 lRXINST.X-50C None Excess Flow Geck Valve 1
- I A, C RR RR 39 i
RX INST.
X-50D None Excess Flow Check Valve 1
A, C RR FR 39 -
RX INST.
X-51A None Excess Flow Geck Valm 1 -
A, C RR RR 39 RX INST.
'X_51B None Excess Flow Check Valve 1
A, C RF RR 39 RX INST.
X-51C None Excess Flow Check Valve 1
, A, C RR RR 39 RX INST.
X-51D None Excess Flow Check Valve 1
A, C-RR RR 39 RX INST.
X-51E-None Excess Flow Check Valve 1
A, 'C l
RR RR 39 l
39 RX INST.
X-51F None Excess Flow deck Valve 1
A, C
- RR RR RX INST.
X-52A None Excess Flow Check Valve-1 A, C RR RR 39
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9 9
9 9
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3 3
3 3
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R R
R R
R s
R R
R R
R R
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ycneu R
R R
R R
R tq R
R R
R R
R se er TF yro eg C
C C
C C
C ve lt aa A
A A
A A
A VC E
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'0 C 1
1 1
1 1
2 ace
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p A
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e v
v v
v v
v l
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l a
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a V
V V
V V
V k
k k
k k
k c
c c
c c
c e
e e
e e
e h
h h
h h
h n
C C
C C
C C
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w w
w w
w w
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l l
l l
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F F
F F
F rc s
s s
s s
s s
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c c
c c
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x x
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E E
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e e
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n n
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n Al Sao o
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N r
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C D
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2 2
2 2
8 au 5
5 5
5 5
2 VN X
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SECTION 4 REQUESTS FOR RELIEF FROM ASME CODE SECTION XI REQUIREMENTS DETERMINED TO BE IMPRACTICAL v
This Section contains Requests for Relief from those ASME Code Section XI requirements which are impractical to implement on the Monticello Nuclear Generating Plant.
Requests for Relief are numbered consecutively. Requests submitted in earlier ASME Code Section XI Inservice Inspection and Testing Program descriptions have, where possible, been carried over to this program with their original identification numbers.
ASME Code changes or NRC review may result in the deletion of a particular Request for Relief. The following Monticello Requests have been deleted:
4 29 12 32 14 33 17 34 20 35 21 36 22 37 25 38 43 O
The following Requests for Relief are new or substantially revised from those contained in the ASME Code Section XI Inservice Inspection and Testing Program used for the first ten year interval:
15 24 16 41 19 42 23 44 45 e
O 4-1
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1.
REOUEST FOR RELIFS APPLICABLE C3fP(MNT FUNCTION A9fE CODE CIASS 11, 12, 13, 14 RHR Provide cooling water to the RER 3
Service F1ter Pumps heat exchangers.
Code Requirement Inlet pressure and differential pressure will not be measured directly as required by IWP-3100 and IWP-4240.
Basis There is no installed instnanentation for directly measuring the inlet pressure and differential pressure of these punps. These punps are submerged and take suction several feet below the river level.
Alternate Testing The river level elevation will be measured to detemine the inlet pressure for these pumps.
Differential pressure will be detemined by taking the difference between the discharge pressure and calculated inlet pressure.
Scheduled for Implementation February 28, 1978
i 2.
Reauest for Relief i
(
j O NPONENT FUNL7 ION APPLICABLE A9E CODE CLASS l
11, 12 Emergency Provide cooling water to the emergency 3
Service W,ter diesel generators and critical reactor building equipment.
11, 12 Standby Provide a redundant means of reactor 2
Liquid Control shutdown as a backup to the Control Rod Drive System.
11, 12 Core Spray Provide cooling water to the reactor 2
l under emergency conditions.
?
11, 12, 13, 14 Provide cooling water to the reactor 2
Residual lieat and to containment under accident Removal conditions.
11, 12, 13, 14 Provide cooling water to the RIR heat 3
RER Service Water exchangers, i
High Pressure Provide cooling water to the reactor 2
Coolant Injection under emergency conditions.
Reactor Core Provide cooling water to the reactor 2
Isolation Cooling under anergency conditions.
Code Requirement Pump bearing temperature will notebe measured as required by IWP-3100 and INP-4310.
Basis There is no instrtmentation installed to measure lube oil or bearing temperature. The use of external temperature measuring devices is not considered meaningful because of the environmental influence on these parameters.
j
I 2.
REQUEST IUR RELIEF (Cont'd.)
i Altemate Testing i
The mechanical coalition of the ptanp will be assessed by using vibration data.
Schedule for Implementation j
February 28, 1978 l
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l l
1 l
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i a
1
, - - - - - - - -.. - -,. - - - - -. - -... -.,.. - ~.. -..
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3.
REQUEST FOR RELIEF l
C04PONENT FUNCTION APPLICABLE CODE CIASS 11, 12 Emergency Provide cooling water to the emergency 3
Service Water diesel generators and critical reactor Ptanps building equipment.
11, 12 Standby Provide a redundant means of reactor 2
Liquid Control shutdown as a backup to the Control Ptsnps Rod Drive System.
11, 12 Core Provide cooling water to the reactor 2
Spray Ptznps under emergency conditions.
11, 12, 13, 14 Provide cooling water to the reactor 2
Residual Heat and to containment under accident Removal Pteps conditions.
?
11, 12, 13, 14 Provide cooling water to the RIR heat 3
vi RIR Service exchangers.
Water Planps High Pressure Provide cooling water to the reactor 2
Cooling Injection under emergency conditions.
Reactor Core Provide cooling water to the reactor 2
Isolation Cooling under emergency conditions.
Code Requirement Displacement vibration anplitude will not be used to evaluate the condition of the pianp as required by IWP-3110, 3210, and 4500.
Basis We prefer to measure vibration velocity due to its superiority in detecting wear and interior machine failure. Existing instrumentation reads out in velocity units.
Altemate Testing Vibration velocity measurements will be used to evaluate the condition of the pump.
Schedule for Implementation February 28, 1978 i
O O
O 5.
REQUEST FOR RELIEF Applicable Valve Component Function A9fE Code Class Category Shutoff Recirc Punp #11 Upper Seal 2
B Flow hhen Pump is Shutdom Shutoff Recirc Pump #12 Upper Seal Flow 2
B hhen Ptr:p is Shutdom Code Requirement
... l I
These valves cannot be tested at the frequency required by IhV-3410.
Basis
?
These valves are located inside primary centainment which has an inerted atmosphere. The only way to verify valve stroke and measure stroke time is by direct observation of the valve stem.
Altemate Testing These valves will be full stroked and timed during each refueling outage when the Containment is de-inerted and cpen for general access.
Schedule for Implementation February 28, 1978
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6.
Request for Relief Applicable Valve Component Function A9E Code Class Category RBCC-15 To provide containment isolation for the 2
A, C Reactor Building Closed Cooling water drywell inlet line.
Code Requirement This valve will not be exercised as required by IW-3520.
Basis
'Ihere are no means provided for detemining that the disc travels to the seat pronptly on cessation 4
or reversal of flow.
Alternate Testing This line will be modified upon concurrence of the NRC to allow leak testing of this valve (see letter from L. O. Mayer to Victor Stello, subject, " Planned Modifications to Pemit Testing to be Conducted in Accordance with 10CFR50, Appendix J", dated May 5, 1976). Proper seating of the valve disc will be verified during the leak rate test.
Schedule for Implementation See Alternate Testing.
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7.
REQUEST FOR RELIEF Applicable A9fE Valve Corponent Function Code Class Category SW-101 Prevent flow of mergency service water 3
C into the nomal service water systm 3
C SW-102 when the emergency service water system SW-103 is operating.
3 C
3 C
SW-104 ESW-4-1 Prevent reversal of flow into redundant 3
C emergency service water line, 3
C m'-4-2 FW-91-1 Prevent reversal of flow into the 2
C feedwater system.
2 C
FW-91-2 c-RIR 8-1 Prevent reversal of flow into RIR 2
C Ptrip Discharge Line.
2 C
gg RHR-21 Prevent reversal of flow into Head 1
C Oooling line HPCI-14 Prevent reversal of flow from 2
C m into HPCI System.
HPCI-15 HPCI-42 2
C HPCI-31 Prevent reversal of flow from IFCI Systen 2
C into Torus.
HPCI-65 Prevent IPCI Exhaust Steam flow to 2
C Pace.
2 C
HPCI-71 RCIC-57 Prevent RCIC sxhaust steam flow to 2
C m Air Space.
2 C
RCIC-59 RCIC-37 Prevent Reversai of Flow from the 2
C Torus Into the Rt.IC System.
2 C
RCIC-16 RCIC-17 2
C I
RCIC-31 Prevent Reversal Flow of PflC Flow 2
C to Torus.
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7.
RBQUEST POR RELIEF (Cont'd.)
Applicable ASE Valve Couponent Function Code Class Category SW-21-1 Prevent Reversal of Normal Cooling 3
C Flow Into the Service Water System.
3 C
SW-21-2 SW-16 Prevent Reversal of Flow From 3
C rgency Se M ce h ter Syst m 3
C SW-18 Into Service Water System.
PC-20-1 Prevent siphoning of Water From 3
C 1 Stomge Pool Into Fuel Pool 3
C PC-20-2 Cleanup System.
HPCI-60 Turbine Steam Exhaust Vac Brkr 2
C e
Code Requirement These valves will not be tested as required by IhV-3520.
Basis There is no means available to verify that the disc travels pronptly to the seat on cessation or reversal of flow for normally open valves or that the disc moves praaptly away from the seat when the closing differential is removed and flow through the valve is initiated for normally closed valves.
Alternate Testing The systems in which these valves are located will be functionally tested on a periodic basis to demonstrate proper operation.
~
, Schedule For Implementation NA
8.
RBQUEST FOR RELIEF Applicable ASMi Valve Component Function Code Class Category FW 94-1 1
A, C FW 94-2 To provide containment isolation for 1
A, C pg 97 7 the feedwater injection lines.
1 A, C FW 97-2 1
A, C Code Requirement These valves will not be exercised as required by IW-3520.
e Basis
.L There are three check valves in series in each of the feedwater injection lines. Verification that each valve disc travels to the seat promptly on cessation of flow cannot be completed by observing pressure indications. The valves cannot be directly observed and there is no instrtmentation installed to monitor disc position.
Alternate Testing Proper seating of the valve disc will be verified during the valve leak rate testing.
Schedule for Implementation Febmary 28, 1978 l
1 I
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(U)
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9.
RBQUEST IOR RELIEF Applicable ASfE Valve C aponent FuncCon Code Class Category CRD-114 Prevent scram discharge flow fr a 2
C flowing back into the CRD during a scram.
CRD-115 Prevent scram accmulator pressure 2
C from discharging into CRD accmulator charging water circuit during a scram.
CRD-138 Prevent scram accunnlator pressure fra 2
C discharging into CRD cooling water circuit during a scram.
CV-126 Provide scram accm ulator pressure 1
B to the bottom of the control rod drive piston during a scran.
CV-127 Exhaust scram discharge water from 2
B 7
the top of the control rod drive p
piston during a scram.
Code Requirement These valves will not be tested as required by IhV-3410 and IhV-3520.
Basis The above listed valves are located on each of the 121 hydraulic control units. 'Ihere is no practical method of testing these valves in accordance with Section XI requirements. There is no instrmentation installed to verify proper seating of the check valves and the ccntrol valves operate too rapidly to measure stroke time. Technical Specifications require all control rods to be scram tested once per operating cycle. These valves are all exercised one full cycle during a scram. Proper operation of these valves and the safety function of the control rod drive system are verified by the scram testing.
Alternate Testing See Basis Schedule For Implementation February 28, 1978
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- 10. REQUEST FOR RELIEF Applicable A9E Valve Component Function Code Class Category 10-7 Imediately stop the steam flow to the 2
B IPCI Turbine.
RCIC-7 1 mediately stop the steam flow to the 2
B RCIC Turbine.
CV-2790 Drywell Isolation for reactor water sap]o 2
A line from recirculation Loop B.
2 A
CV-2791 TIP 1-1 2
A TIP 1-2 Drywell isolation for TIP System.
2 A
b TIP 1-3 2
A Code Requirement These valves will not be stroke timed as required by IW-3410.
Basis These valves operate too fast to obtain meaningful stroke time.
Alternate Testing These valves will be full stroked as required by IW-3410 and proper operation will be verified.
Schedule for Implementation February 28, 1978
- 11. REQJEST FOR RELIEF Applicable ASME Valve Conponent Function Code Class Category RER SW-17 Prevent reversal of flow of RfR water 2
C into RIR Service Water System.
Code Requirement This valve cannot be exercised as required by IW-3520.
Basis Exercising of this valve would require ptmping river water into the RIR Systen.
Il f
Alternate Testing U
None Schedule for Implanentation NA P
i
.. -.. - ~,... -, _. - - - _ -.. - - - - _. - -.
m
- 13. RBQUEST IOR RELIEF Applicable A9fE Valve Component Function CM e Class Category XP-6 Prevent reversal of flow of reactor 1
A, C XP-7 water into SBLC System.
1 A, C Code Requirement These valves will not be exercised at the frequency required by IW-3520.
Basis Exercising of these valves can only be accmplished by initiation of the SBIf Systs, including actuation of an explosive valve, and pumping to the reactor vessel.
Altemate Testing These valves will be exercised by initiation of the SBIf System, actuating an explosive valve and ptmping deineralized water to the reactor vessel during each refueling outage.
Schedule For Inplmentation Febmary 28, 1978
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1 15.
REQUEST FOR RELIEF CODE PROGRAM CODE EXAM COMPONENT OR ITEM CLASS TABLE ITEM CATEGORY PIPING WELDS 2
5.10 C5.ll C-F 2
5.10 C5.12 C-F 2
5.10 C5.21 C-F 2
5.10 C5.22 C-F 2
5.10 C5.31 C-F 2
5.10 C5.32 C-F CODE REQUIREMENT By reference in 10CFR50.55a (b)(2)(IV), paragraph IWC-1220 of the 1974 Edition through and including the Summer 1975 Addenda shall be used for the exemption criteria for determining the extent of examination for piping welds.
I B ASIS 0;
This exemption criteria will not be used to d;/elop the ISI program. NSP does not technically concur with the basis for many of the exemptions and especially the control of system chemistry. This type of control l
eliminates one mode of possible failure, but it does not totally eliminate the need for examinations.
ALTERNATE EXAMINATION The Class 2 NDE exemption criteria established f u paragraph IWC-1220 of the 1977 Edition through and including the Summer 1978 Addenda of ASME Section XI wil'. be utilized to develop the Monticello ISI program. This exemption criteria is considered more conservative and the use of these exemptions is consistant with recent revisions to 10CFR50.55a which references the Summer 1978 Addenda.
SCHEDULE FOR IMPLEMENTATION June 30, 1981
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f 16.
REQUEST FOR RELIEF CODE PROGRAM CODE EXAM CLa3S TABLE ITEM CATEGORY COMPONENT OR ITEM REACTOR VESSEL 1
1.1 Bl.ll B-A CIRCUMFERENTIAL WELDS VCBB-1, VCBA-2, VCBB-3, & VCBB-4 1
1.1 Bl.12 B-A LONGITUDINAL WELDS VLAA-1, VLAA-2, VLBA-1, VLBA-2, VLCB-1, VLCB-2, VLDB-1, & VLDB-2 CODE REQUIREMENT Perform a volumetric examination of one circumferential and one longitudinal beltline region weld.
d BASIS A
The examination of the circumferentis' weld and portions of the longitudinal welds will not be performed. The Fbnticello RPV was constructed with a 2'3h" thick biological shield wall surrounding it, with the exception of i
the top eight feet. Between this wall and the reactor vessel shell is a space of appro::imately 1 foot that l
houses the thermal insulation. The only access areas to the reactor vessel are:
- 1. at the top eight feet above the biological shield wall, 2. through openings in the wall at each nozzle location and two inspection ports below the skirt weld, 3. by the control rod drives under the reactor head, and 4. from the vessel inside diameter.
The area above the biological shield wall and at the nozzle openings is further obstructed by non-removable A good portion of the vessel insulation was not designed to be removed and therefore it was installed insulation.
prior to the installation of the piping, electrical conduits, vessel stabilizers, duct work, etc.
A very thorough review was performed, using drawings, sketches, and previous examination reports, to try and locate weld areas that possibly could be inspected.
It was concluded that some of the vessel welds appear to be close enough to nozzle openings for performing the examinations provided the insulation can be removed.
An attempt will be made to remove or modify the insulation and to examine many of the reactor vessel welds during the 1981 refueling outage. Each of the welds that can be examined will be sketched to show the examination amount, extent, and location.
The examination areas and amount shown in Table 1.1 were scheduled from the drawing review. As these areas are examined, the specific amount and extent in Table 1.1 will be changed to reflect the actual measurements.
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REQUEST FOR RELIEF (continued)
ALTERNATE Oue to the inaccessibility of the circumferential beltline region veld (VCBA-2), all of the accessible areas on the remaining circumferential welds will be examined. A portion of the beltline region longitudinal welds may be accessible. These two areas, if accessible, and all of the accessible areas on the remaining longitudinal welds will be examined.
SCHEDULE FOR IMPLEMENTATION June 30, 1981
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1 18.
REQUEST FOR RELIEF CODE PROGRAM CODE EXAM COMPONENT OR ITEM CLASS TABLE ITEM CATEGORY REACTOR VESSEL FLANGE LEAKAGE SENSORS 1
4.1 B4.12 B-E (N0ZZLES N-13 AND N-14)
CODE REQUIREMENT Perform a visual (VT-2) examination of the external surfaces of the flange leakage sensor nozzles.
BASIS 7
The area surrounding these two penetrations will not be visually examined for evidence of leakage during the 5;
vessel pressure test as required by Exam Category B-E.
These penetrations never see pressure during either operation or vessel pressure test, unless the vessel flange o-rings leak.
Inspection during pressure testing therefore serves no purpose.
In addition, the nozzle area is not accessible without damaging insulation.
j ALTERNATE The nozzles will be hydrostatically tested to insure seal integrity at or near the end of the inspection interval.
In addition, the areas surrounding these two penetrations will be visually examined if insulation is removed for maintenance or other inspection activities.
SCHEDULE FOR IMPLEMENTATION l
February 28, 1978 l
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19.
REQUEST FOR RLLIEF
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CODE PROGRAM CODE EXAM COMPONENT OR ITEM CLASS TABLE ITEM CATECORY i
STANDBY LIQUID CONTROL N0ZZLE TO VESSEL WELD 1
3.1 B3.10 B-D N0ZZLE INSIDE RADIUS SECTION 1
3.1 B3.20 B-D N0ZZLE TO' SAFE END WELDS 1
5.1 B5.10 B-F CODE REQUIREMENT Perform volumetric examinations on these welds and a surface examination on the nozzle to safe end weld.
BASIS e
These examinations will be performed to the extent possible. The design of the biological shield wall, the h
vessel and nozzle insulation, and the physical surroundings presents access to portions of these welds. The vessel insulation around the standby liquid control nozzle was not designed with removable panels. The nozzle protrudes through a hole in the insulation with minimal clearance.
ALTERNATE l
l The volumetric and surface examinations required on these welds will be performed to the extent possible. An attempt will be made during the 1981 refueling outage to modify the insulation to allow accessibility and to l
sketch.those portions of the welds that remain inaccessible.
SCHEDULE FOR IMPLEMENTATION l
1, June 30, 1981 l
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23.
REQUEST FOR RELIEF b
i CODE PROGRAM CODE EXAM COMPONENT OR ITEM CLASS TABLE ITEM CATEGORY COMPONENT SUPPORTS FOR PIPING, PUMPS, AND VALVES 1
11.1 Bil.10 B-K-2 SUPPORT MEMBERS 2
3.10 C3.50 C-E PIPING COMPONENT SUPPORTS COMPONENT SUPPORTS AND RESTRAINTS 3
Dl.2 D-A D2.2 D-B D3.2 D-C CODE REQUIREMENT Examination Category B-K-2 and C-E of ASME Section XI requires all areas of the support component from the piping, valve, and pump attachment to and including the attachment to the supporting structure be examined.
i BASIS E
Insulation will not be removed for the visual examination provided that all mechanical connections and welds can be examined.
It has been our experience that any loss of support capability or inadequate restraint can usually be detected through the inspection of the uninsulated portion of the support and the surrounding insulation.
The governing Codes and Regulations used in the design and construction of those systems that are now classified as Class 2 and 3 did not require provisions for inspection access for these systems. Thus, it would be an undue burden without compensating incresce in safety to require insulation removal for support inspection.
ALTERN5TIVE The insulation will be removed from a supported component for further inspections whenever the connections and welds can not be examined or an abnormality is detected that may have been a result of a loss of support capa-l bility or inadequate restraint.
I SCHEDULE FOR IMPLEMENTATION _
i June 30, 1981 s
- - - - -, - - - - - ~ - - - - -
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i 24.
REQUEST FOR RELIEF ICODE PROGRAM CODE EXAM CLASS TABLE ITEM CATEGORY COMPONENT OR ITEM BOLTS AND STUDS RECIRCULATION PUMP FLANGE BOLTS 1
6.1 B6.180 B-C-1
& B6.190 P-200A & P '.00B RECIRCULATION VALVE BONNET BOLTING 1
6.1 B6.210 B-G-1
& B6.220 M02-53A, M02-43A, M02-53B, & M02-43B CODE REQUIREMENTS l
Ultrasonic examinations shall be performed in accordance with Article 5 of Section V when the provisions of j
Article 4 of Section V or Appendix III of Section XI do not apply.Section V requires that calibration be established on a test bar that has certain physical and chemical parameters.
s~
h BASIS The Sect. ion V technique utilizing the calibration test bar was not used for the baseline examinations and it is not as sensitive to detect discontinuties as the presently applied back reflection method.
In addition, when using the back reflection method, the poorer the end reflecting surfaces (painted, corroded, etc.) the more conservative the examinations are.
ALTERNATE The items will be examined using the back reflection method correlated with an as built sketch of the particular bolt or stud being examined. ASME Section XI will be used for evaluation criteria.
SCHEDULE FOR IMPLEMENTATION June 30, 1981
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- 26. RMUEST FOR RELIEF l
APPLICABLE AfME VALVE COFONENT FUNCTION CODE CLASS CATEGORY TIP 3 Prevent reversal of flor 2
A, C in TIP purge line.
4 Code Requirement This valve cannot be exercised as required by IW-3520.
Basis This is a nomally open check valve that is in service during all i
modes of operation.
In addition, there is no means available to U
verify that the disc travels promptly to the seat on cessation or reversal of flow.
Altemate Testing Proper seating of the valve disc will be verified during the valve leak rate testing.
Schedule for Implementation February 28, 1978
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27.
REQUEST FOR RELIEF APPLICABLE ASME COIPONENT FUNCTION CODE CLASS 11,12 Emergency Service Provide cooling water to the emergency 3
Water Pumps diesel generators and critical reactor building equipment.
Code Requirement Pump flowrate will not be measured to determine pump performance as required by Ih?-3100.
Basis e4 There is no installed instrumentation for measuring the flowrate of these pumps. Flowrate varies due to the seasonal variations in cooling requirements making it impractical to w
establish a reference value and acceptance criteria for this parameter.
Alternate Testing The Emergency Service Water pumps will be tested to shutoff pressure. Pump differential pressure will be measured under these conditions.
Schedule for Implementation January 1, 1979
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- 28. REQUESF FOR RF1IEF APPLICABLE AS4h VALVE CGIPONEhT FUNCTION CODE CLASS CATEGORY All power operated valves.
1, 2, 3 A, B Code Requirement The acceptance criteria for valve-stroke time as stated in IWV-3413(c).
will not be used.
Basis
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Z Stroke time acceptance criteria outlined in IhV-3410 is general and is not based on system functional requirements and normal valve variability.
Alternate Acceptance criteria for valve stroke times will be based on nomal valve variability and on system functional requirements.
Schedule for Implementation February 28, 1978 i
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- 30. REQUEST FOR RELIEF APPLICABLE ASME CGIPOND(F FUNLTION CODE CLASS All Class 1, 2 and Pressure Retaining 1,2,3 3 Components Code Requirements The test pressure requirements of INA-5000, IWB-5000, INC-5000 and IWD-5000 will not be met on certain components.
Basis The code does not recognize that non-isolable junctions of components A
with different design pressures or different ASME Classes exist (i.e.,
pump suction and discharge lines, piping upstream and downstream of restricting orifices, etc.).
Pressurizing components to the require-ments of the code may result in overpressurizing the non-isolable cc.ponents.
Alternate Testing Where these junctions exist, test pressure will be based on the component with the lowest test pressure requirement.
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Schedule for Implementation February 28, 1978 i
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- 31. REQUEST R)R RELIEF APPLICABLE KNE CCMPONENT FUNCTION CODE CIASS Head vent and leak Provide connection for 2
test connections on leak testing and for Class 1 piping.
venting the reactor head.
Code Requirement s~
4 These lines will not be pressure tested in accordance with IWC-5200.
cn Basis These line.c are connected to Class 1 piping and are classified as Quality Group B lines (applicable ASME Code Class 2) due to line size only. These lines will be tested in accordance with Class 1 requirements (IWB-5000).
Alternate Testing See Basis Schedule for Implenentation February 28, 1978
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- 39. REQTST FOR RELIEF APPLICABLE AShE VALVE C0bFONENT FUNCTION mm CLASS CATEGORY Excess Flow Check Minimize the blowdown to the Secondary Valves, Typical of Containment in the event of an instrument 1
A, C X-27A through X-52F sensing line break.
Except X-28F.
00T REQUIREhENT These valves will not be tested as required by IWV-3410 or IWV-3520.
' BASIS s~
4 The Excess Flow Check Valves are located in sensing lines for the Plant Protection System and ECCS Instruments. Testing the check valves during operation is not practical since it would make the vital instrumentation inoperable.
Cold shutdown testing is also impractical since it would require pressurizing the reactor vessel to operating pressure.
' ALTERNATE TESTING The Excess Flow Check Valves are tested each refueling outage when the vessel is pressurized to 1000 psig during the Reactor Vessel Hydro Test.
SOEIXEI FOR IhPLEFENTATION This testing was implemented since initial plant startup as required by the original Technical Specifications.
- 40. REQUEST FOR RELIEF g g,7 g,gyg APPLICABLE AShE VALVE 00DE CLASS
. CATEGORY Excess Flow Check Minimize the Blowdown to the secondary Valve X-28F containment in the event of an instru-2 A, C ment line break and vessel head seal failure.
CODE REQUIREhENT This valve will not be tested as required by IWV-3410 or INV-3520 BASIS
?>
There is no way to test this valve except to remove it from the line since there is nonna11y no pressure between the vessel head seals. Removal of the valve is extremely difficult because the piping is rigid and thread damage has resulted from previous attempts to remove the valve. A sensing line break would result in blowdown to the secondary containment only if the vessel head seal was leaking. The blowdown would be limited by the seal leak which would, in all probability, be less than the leakage allowed by the check valve. The probability of a line break and seal leak occuring to cause a significant blowdown is extremely small, therefore it is felt this valve should not be tested to avoid further thread damage to the piping.
ALTERNATE TESTING None
'SOIEIULE FOR IFPLEhENTATIN l
N/A
41.
REQUEST FOR RELIEF mm CODE PROGRAM CODE EXAM COMPONENT OR ITEM CLASS TABLE ITEM CATECORY PUMP CASINGS RECIRCULATION PUMPS P-200A & P-200B 1
12.1 B12.20 B-L-2 CODE REQUIREMENT Perform a visual examination (VT-1) of all internal surfaces in at least one piimp.
BASIS Disassembly of the recirculation pumps for the sole purpose of visual examination of the casing internal pressure surfaces requires many manhours from skilled maintenance personnel.
Increased radiation exposures i
O result from this activity. The probability of pump failure is increased by unnecessarily disassembling the units. Deferring the examination has no affect on integrity of the pumps.
ALTERNATE i
Recirculation Pump internal pressure surfaces will be visually examined when the pumps are disassembled for maintenance.
I SCHEDULE FOR IMPLEMENTATION l
June 30, 1981
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9 42.
REQUEST FOR RELIEF CODE PROGRAM CODE EXAM COMPONENT CLASS TABLE ITEM CATEGORY VALVE BODIES 1
12.1 B12.20 B-M-2 CRANE CHAPMAN GATE VALVES RECIRCULATION VALVES MO 2-65A, MO 2-65B M0 2-53A, MO 2-53B M0 2-43A, MO 2-43B CODE REQUIREMENT Perform a visual examination (VT-1) of all internal surfaces in at least one valve in this group.
l BASIS
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g Disassembly of the recirculation valves for the sole purpose of visual examination of the internal pressure surfaces requires many manhours from skilled maintenance personnel.
Increased radiation exposures result from this activity. The probability of valve failure is increased by unnecessarily disassembling the units.
Deferring the examination has no affect on the integrity of the valves.
ALTERNATE I
Recirculation Valve internal pressure surface will be visually examined when the pumps are disassembled for mainten'ance.
f SCHEDULE FOR IMPLEMENTATION June 30, 1981
- 44. REQUEST FOR RELIEF COMPONENT FUNCTION APPLICABLE ASME CODE CIASS 11 & 12 Standby Liquid Provide a redundant means of 2
Control System Pumps reactor shutdown as a backup to the control Rod Drive System CODE REQUIREMENT The flow rate will not be measured by using a rate or quantity meter installed in
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the Standby Liquid Control System pump circuit as required by IWP-4600.
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BASIS There is no rate or quantity meter installed in the system.
ALTERNATE TESTING The flow rate will be determined by measuring the change in water level in the Standby Liquid Control System test tank over a fixed period of time.
t SCHEDULE FOR IMPLEMENTATION
' June 30, 1981
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- 45. REQUEST FOR RELIEF I
COMPONENT FUNCTION APPLICABLE ASME I
CODE CIASS All pump instrumentation Used to measure pump reference 2&3 values and measure pump test quantities CODE REQUIREMENT The full-scale range of each instrument will not be three times the reference value or i
less as required by IWP-4120.
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BASIS Some of the instruments presently installed have a range greater than three times i
the reference value. All the instrumentation presently installed have. a. full-scale range four times the reference value or less which was the requirement of 1
IWP-4111 of the 1974 Edition of the ASME Code,Section XI, through and including the Sununer 1975 Addenda (previous approved Edition and Addenda).
Replacement of instrumentation is not practical.
i ALTERNATE TESTING a
The full-scale range of each instrument will not be greater than four times the reference value as required by the Edition and Addenda of the Code previously approved by the NRC.
SCHEDULE FOR IMPLEMENTATION June 30, 1981 i
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SECTION 5 PROPOSED TECHNICAL SPECIFICATION CHANGES
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Reproduced in this section are the proposed Technical Specification changes included in Northern States Power Company's License Amendment Request dated August 30, 1977. These changes revise the surveillance requirements in the Technical Specifications to conform to 10 CFR Part 50, Section 50.55a(g).
Changes proposed in Northern States Power Company's License Amendment Request dated January 18, 1978 also appear here.
These changes were submitted to incorporate a program of augmented inservice inspection for piping susceptible to stress corrosion cracking.
Due to the extended period of time from initial filing of these proposed Technical Specification changes and completion of NRC Staff review and issuance of the changes, other Technical Specification changes have been made which also affect these pages.
This section will be kept current by showing these changes.
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRENDITS 3.4 STANDBY LIQUID CONTROL SYSTD4 4.4 STANDBY LIQUID CONTROL SYSTEM Applicability:
Applicability:
Applies to the operating status of the Applies to the periodic testing require-standby liquid control system.
ments for the standby liquid control system.
Objective:
Objective:
To assure the availability of an To verify the operability of the standby independent reactivity control mechanism.
liquid control system.
SPECIFICATION:
SPECIFICATION:
A.
Nomal Operation A.
The operability of the standby liquid control system shall be verified by I
- 1. 'lhe standby liquid control system perfomance of the following tests:
shall be operable at all times when fuel is in the reactor and the reactor is not shutdown by control 1.
At least once each ope mting cycle rods, except as specified in 3.4.B.
manually initiate one of the two standby liquid control systems and
- 2. Ihch standby liquid control system pump Pump deminemlir.ed water into the I
shall be capable of delivering 24 gpm reactor vessel.
Both systems shall against a reactor pressure of 1275 psig.
be tested and inspected in the course of two operating cycles.
3 The system pmssure relief valves shall be opemble with a setpoint between 1350 and 1450 psig.
3.4/4.4 A
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30 N E N ITIW S FW OPERATIM k.O SLEVIIIJJICE REQUIR9HNf3
- 2. Inaervice inspection and testing of components shall be conducted in accordance with Specification 4.XX.
i Tw B.
Operation with Inoperable Components B.
Surveillance with Inoperable Courponents From and after the date that a redundant When a component becomes inopemble, its component is made or found to be inoperable, redundant component shall be demonstrated Specification 3 4.A shall be considered to be operable immediately and daily fulfilled, provided that the component is thereafter.
returned to an operable condition within seven days.
1 94 3.4/4.4 REV l
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30 LIMITING conditions FOR OPERATION 4.0 SURVEIILANCE REQUIREMENES C.
Vo1 M oncentration Requiremments C.
%e availability of the proper boron he liquid poison tank shall contain bearing solution shall be verified by a boron bearing solution that satisfies performance of the fo14owing tests; the voltune-concentration requirements 1.
At least once per month -
i of Figure 3.4.1 and at all times when the standby liquid poison system is r*
Boron concentration shall be quired to be operable the tesaperature determined.
In addition, the boron shall not be less than the solution tW ersture presented in Figure 3.h.2.
In concentration shall be determined u
j addition, the heat tracing on the pump any time water or boron are added suction lines shall be operable whenever or if the solution temperature drops the room temperature is less than the below the limits specified by Figure 3.4.2.
solution temperature presented in Figure 3.4.2.
1 l
3.4/4.4 95 REV
gases 3.4 and 4.4:
A.
The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn con-trol rods can be inserted.
To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of 900 ppm of boron in the reactor core in less than 125 minutes.
900 ppm boron concentration in the reactor core is required to bring the reactor from full power to a 3% A k suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin for possible imperfect mixing of the ch===f cal solution in the reactor water and dilution from the water in the cooldown circuit. A minimum not quantity of 1400 gallons of solution having a 21.4% sodium pentaborate concentration is required to meet this shut-down requirement.
he time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.
he maximum net storage volume of the baron solution is 2095 gallous. (256 gallons are contained below the pump suction and, therefore, have not been used in the net quantities above. )
Boron concentration, solution temperature, and volume (including checx of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Experience with pump operability demonstrates that testing at a three-sonth interval is adequate to detect if failures have occurred.
Standby liquid contml system components are inspected and tested in acconiance with the requimments of 10 CFR 50, Section 50 55a(g). Rese mquirements are delineated in Specification 4.XX.
Bis inspection and testing pmgram, combined with the additional surveillance aquirements contained in this section, pmvide a high degne of assurance that the standby liquid contml system will perform as requimd when needed.
he relief valves in the standby liquid control system protect the system piping and positive dis-placement pusps which are nominally designed for 1500 psi from overpressure, he pressure relief
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valves discharge back to the standby liquid control solution tank.
I 3,h/4.k. BASES 99 REV
a 30 U MITING CONDITIONS FOR OPERATION 4.0 SURVEIIIANCE REQUIRDG2fTS 35 CORE AND CONTAINMENT COOLING SYSTDE 4.5 CORE AND CONTAINMDff COOLING SYSTB4S Applicability:
Applicability:
Applies to the operational status of the Applies to periodic testing of the essergency esmergency cooling systems.
cooling systens.
Objective:
Objective:
y To insure adequate cooling capability for heat To verify the operability of the emergency resnoval in the event of a lose of coolant cooling systems, as accident or isolation from the nonnal reactor heat sink.
Specification:
Specification:
Loet Pressure Core Cooling Capability Iow Pressure Core Caali v 8% - bility A.
Core Spray System A.
Surveillance of the core spray system shall be perforised as follows:
1.
Except as specified in 3 5.A.2.,
3.5.A.3., and 3 5.A.S. below, both core 1.
Routine Testing spray subsystems shall be operable when--
ever irradiated fuel is in the reactor
- a. A simulated automatic actuation test vessel and reactor coolant water tempera-shall be conducted each refueling outage.
ture is greater than 212 F.
- b. Core spray headerap instrumentation shall be checked once each day, tested once each month, and calibrated once each 3-month period.
101 3 5/4.5 REV
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMDrFS Inservice inspection and testing of c.
components shall be conducted in accordance with Specification 4.XX.
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I 2.
From and after the date that one of the 2.
When it is determined that one core core spray systans is made or found to be inoperable for any reason, reactor opera-spray system is inopemble, the oper-tion is permissible only during the suc-able core spray system and the IECI ceeding fifteen days unless such system mode of the RER systen and the diesel generator 6 required for operation of is sooner made operable, provided that such components (if no external source during such fifteen days all active compo-of power were available) shm11 be nents of the other core spray syetem and the LPCI mode of the RHR system and the demonstrated to be operable immedia-tely. 'Ibe operable core spray system diesel generators required for operation of such components (if no external source shall be demonstrated to be operable daily thereafter.
of power were available) shall be operable.
3 From and after the date that both core 3
When it is det' ermined that both core spray systems are made or found to be i
inoperable for any reason, reactor spray systems are inoperable, the LPCI mode of the RHR system and the 35/4.5 102 REV
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s 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIRD4ENTS operation is permissible only during the succeeding seven days unless at diesel generators required for least one of such systems is sooner operation of such components (if no external ~ source of power were made operable, provided that during available) shall be demonstrated such seven days all active components of the LPCI mode of RHR system and the to be operable inanediately and daily thereafter.
diesel generators required for operation of such components (if no external source of power were available) shall be opera-ble.
4.
Each core spray system shall be capable of delivering 3,020 gpm against a reactor pressure of 130 psig.
If this rate of delivery requirement cannot be met, the systema shall be considered inoperable.
5 If the requirements of 3 5.A.1 - 3 cannot be met, an orderly shutdown of the reactor will be initiated and the reactor water temperature shall be reduced to less than 212 0F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.5/4.5
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEIIZANCE REQUIREM WPS B. Iow Pressure Coolant Injection (LPCI) Subsystem B. Surveillance of the Iow Pressure Coolant (LPCI mode of RHR system)
Injection (LPCI) Subsystem (LPCI mode of RHR system) shall be perfomed as follows:
1.
Except as specified in 3.5.B.2 and 3 5.B.3 below, the LPCI shall be operable
- 1. Routine Testing whenever irradiated fuel is in the reactor vessel and reactor coolant temperature is
- a. A simulated automatic actuation test greater than 212 F.
shall be conducted each refueling outage.
- b. Inservice inspection and testing of components shall be conducted in accordance with Specification 4.XX.
Y
- c. During each five year period, an air test aball be perfomed on the dryvell spray headers and nozzles.
2.
From and after the date that one of the 2.
When it is detemined that one of the LPCI pumps or admission valves is nede LPCI pumps is inoperable, the remaining or found to be inoperable for any reason, active components of the LPCI and con-reactor operation is permissible only tainment coolin6 subsystem, both core during the succeeding thirty days unless such pump or admission valve is sooner spray systems and the diesel generstors required for operation of such components made operable, provided that during such (if no external source of power were thirty days the remaining active components available) shall be demonstzsted to be of the LPCI and containment cooling sub-operable immediately and the operable system and all active components of both LPCI pumps daily thereafter, core spray systems and the diesel genera-tors required for operation of such com-ponents (if no external source of power were available) shall be operable.
35/h.5 10' REV
9 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.
From and after tha date that two of the LPCI 3.
When it is determined that the LPCI pumps or admission valves are made or found subsystem is inoperable, both core to be inoperable for any reason, reactor spray systems, the containment operation is permissible only during the cooling subsystem, and the diesel succeeding seven days unless such pumps or generators required for operation admission valves are made operable sooner, of such components (if no external provided that during such seven days all source of power were available) active components of both core spray sys-shall be demonstrated to be operable tems, the containment cooling subsystem immediately and daily thereafter.
(including 2 LPCI pumps) and the diesel y
generators required for operation of such g
components (if no external source of power were available) shall be demonstrated to be operable at least once each day.
4.
A maximum of one drywell spray loop (contain-ment cooling mode of RHR) may be inoperable for 30 days when the reactor water tempera-ture is greater than 212 F.
If the loop is not returned to service within 30 days, the orderly shutdown of the reactor will be initiated and the reactor water temperature shall be reduced to less than 212 F.
5.
Each LPCI subsystem (RHR) pump shall be capable of delivering 4,000 gpa +10%
against a system head corresponding to three pumps delivering 12,000 gpa at a reactor pressure of 20 psi above the suppression chamber pressure. If this 3.5/4.5 105 REV a
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l 3.0 LIMITING CONDITIONS FOR OPERATION k.O SURVEILIRCE REQUIRD4ENTS rate of delivery requirement cannot be met, the ptunp anall be considered inoper-able.
6.
If the requirements of 3.5.B.1-4 cannot be met. an orderly shutdown of the reactor will be initiated and the reactor water t
rature shall be mduced to less than
'within 24 houis.
Y Containment Cooling Capability Containment Cooling Capability C.
Residual Heat Removal (RHR) Service Water C.
Surveillance of the RHR service water System system shall be performed as follows:
1.
Except as specified in 3 5.c.2 and 3 5.C.3 1.
Inservice inspection and teettog of below, both RHR service water system loops components shall be conducted in shall be operable whenever irradiated fuel accordance with Specification 4.XX.
is in the reactor vessel and reactor coolant temperature is greater than 2120F.
2.
From and after the date that one of the 2.
When it is determined that one RHR RHR service water system pt=ps is made or service water pump is inoperable, found to be inoperable for any reason, the redundant components of the 106 3 5/4.5 REV
30 LIMITING CONDITIONS M R OPERATICE 4.0 SURVEILLUK2 RIENJIRBENTS reactor operation is pemissible only remaining subsystem shall be during the succeeding thirty days unless demonstrated to be operable inumedi-such pump is sooner made operable, pro-ately and daily thereafter.
vided that during such thirty days all other active components of the RHR service water system are opemble.
3 From and after the date that one of the 3
When one RER service water systeen RHR service water systems is made or found becomes inoperable, the operable to be inoperable for any reason, reactor system shall be demonstrated to be operstion is pemissible only during the operable imunediately and daily T
succeeding seven days unless such system thereafter.
G is sooner made operable, provided that during such seven days all active compo-nents of the operable BHR service water system shall be denonstrated to be opers-ble at least once each day.
4.
Ib be considered operable, a RER service water ptaip shall be capable of delivering 3500 gym against a head or 500 feet.
5 If the requirements of 3 5.C.1-3 cannot be met, an orderly shutdown of the reactor will be initiated and the reactor water temperature shall be reduced to less than 212oF vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 3 5/4.5 107 REV
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLAICE REQUIREMEITIS High Pressure Core Cooling Capability High Pressure Core Cooling Capability D.
High Pressure Coolant Injection (HPCI) System D.
Surveillance of HPCI System shall be performed as follows:
1.
Except as specified in 3 5.D.2 below, 1.
Ibutine Testing the HPCI system shall be operable when-ever the reactor pressure is greater than 150 psig and irradiated fuel is in the
- a. A sinalated automatic actuation test y
reactor vessel.
shall be conducted each refueling Un outage.
- b. Inservice inspection and testing of components shall be conducted in accordance with Specification 4.XX.
I 2.
From and after the date that the HPCI 2.
When it is determined that HPCI system is made or found to be inoperable system is inoperable, the RCIC system, for any reason, reactor operation is per-the LPCI subsystem, and both of the core missible only during the succeeding seven spray systems shall be demonstrated days unless such system is sooner made to be operable immediately, operable, provided that during such seven days all of the Automatic Pressure Relief system, the RCIC system, both of the core spray systems, and the LPCI subsystem and containment cooling mode of the RHR system are operable.
108 3.5/4.5 REV
30 LIMITING CONDITIONS FOR OPERATION
!.. O SURVEILLANCE REQUIRD'.DITS 3
To be considered opernble, the HPCI system shall meet the following conditions:
he HPCI shall be capable of delivering a.
3,000 spa into the reactor vessel for the reactor pressure range of 1120 psig to 150 psig.
b.
he condensate storage tanks shall y
contain at least 75,000 gallons of g
condensate water.
c.
he controls for automatic transfer of the HPCI pump suction from the condensate storage tank to the suppression chamber shall be operable.
4.
If the requirements of 3 5.U.l 2 cannot be met, an or11erly reactor shutdown shall be initiated immediately and the zwactor g.Masure shall be reduced to 150 Psig
. within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
35/4.5 109 REV
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i 3.G LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIN:CE REQUIRDtENTS E.
Automatic Pressure Relief System E.
Surveillance of the Automatic Pressure Relief System shall be performed as follows:
1.
Except as specified in 3 5.E.2 and
- 1. Routine Testing 3.5.E.3 below, the entire autoaatic pressure relief syctem shall be operable
- a. A simulated automatic actuation at any time the reactor pressure is above test shall be conducted each oper-150 psig and irradiated fuel is in the ating cycle.
reactor vessel.-
- b. Once each operating cycle, valve oper-u 2.
From and after the date that one of the ability shall be verified by cycling the valves and observing a compensating l-.
automatic pressure relief system valves is change in turbine bypass valve position.
made or found to be inoperable for any reason, reactor operation is permissible
- c. Inservice inspection and testing of only during the succeeding saven days components shall be conducted in unless such valve is sooner made operable, accordance with Specification 4.XX.
provided that during such seven days both remaining automatic relief system valves and the HPCI system are operable.
3 From and after the date that more than one of the automatic pressure relief valves are made or found to be inoperable for any reason, reactor operation is 2.
Mien it is determined thet one or permissible only during the succeeding more automatic pressure relief valves 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless repairs are made and of the Automatic Pressure Relief provided that during such time the HPCI system is inoperable, the HPCI systest system is gerable.
shall be demonstrated to be operable inmediately and weekly thereafter.
4.
If the requirements of 3 5.E.1-3 cannot i
be met, an orderly reactor shutdown shall be initiated immediately and the reactor shall be reduced to 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
110 35/4.5 REV
-o o
,o 3.0 LDERING CMDITICMB FGt OPERATION 4.0 SURVEILIANCE REGIN F.
Reactor Core Isolation Cooling System (RCIC)
F.
Surveillance of Reactor Core Isolation Cooling System (RCIC)
Surveillance of the ECIC System shall be performed as follows:
1.
Except as specified in 3 5.F.2 below, the
- a. A si -lated automatic actuation test psig and irradiated fuel is in the reactor shall be conducted each refueling outage.
vessel.
T
- b. Inservice inspection and testing E
a.
To be considered operable, the RCIC of components shall be conducted system shall be capable of delivering in accordance with Specification 4.XX.
1.00 gym into the mactor vessel.
2.
From and after the date that the RCIC sys-2.
tem is made or found to be inoperable for When it is determined that the RCIC sys-any reason, reactor operation is permissible tem is inoperable, the HPCI systes shall only during the succeedir.g 15 days unless be demonstrated to be operable icmediately and daily thereafter.
such system is sooner made operable, provided that during such 15 days all active compo-nents of the HPCI system am operable.
3 If the requirements of 3 5.F.1 - 2 cannot be met, an orderly shutdown of,the reactor shall be initiated inanediately,and the mactoi pressure shall be mduced to 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> themefter.
111 35/4.5
.I 3.0 LIMITING CONDITIONS FOR OPEAATION 4.0 SURVEILIANCE REQUIREDENTS I.
R: circulation System I.
Recirculation System 1.
Except as specified in 3.5.1.2 below, whenever 1
Once per month, when irriated fuel is in the irradiated fuel is in the reactor, with reactor reactor with reactor coolant temperature greater coolant temperature greater than 212 F and both 0
than 212 F and both reactor recirculation reactor recirculation pumps operating, the pumps operating, the recirculation system cross recirculation system cross tie valve interlocks tie valve interlocks shall be demonstrated to shall be operable.
be operable by verifying that the cros tie valves cannot be opened using the normal control u 2.
The recirculation system cross tie valve inter-switch.
E locks may be inoperable if at least one cross tie valve is maintained fully closed.
2.
When a recirculation system cross tie valve interlock is inoperable, the position of at least one fully closed cross tie valve shall 3.
Valves in the equalizer piping between the be recorded daily, recirculation loops shall be closed. Reactor operation with one loop shall be limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3 Inservice inspection and testing of components shall be conducted in accordance with Specification 4.XX.
1 114 l
3.5/4.5 REV 1
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Bases 4.5:
The testing interval for the core and containment cooling systems is based on a quantitative reliability analysis, judgeent, and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, the core spray final admission valves do not open until reactor pressure has fallen to 450 psig; thus, during operation even if high drywell pressure were simulated, the final Mves would not open. In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel, which is not desirable.
The systems can be automatically actuated during a refueling outage and this will be done. To increase the availability of the individual components of the core and containment cooling systems, the components etich make up the system, i.e., instrtamentation, pungs, valve operators, etc., are tested more frequently.
The instrumentation will initially be functionally tested once per month until a trend is established and thereafter according to Figure 4.1 (see Section 3.1/4.1) with an interval not greater then three Y
months.
Core and containment cooling system components are inspected and tested in accordance with E
the requirements of 10 CFR So, section 50 55a(g). These requirements are delineated in specification 4.XX.
This inspection and testing program, combined with the additional surveillance requirements contained in this section, provide a high degree of assumnce that the core and containment cooling systems will perfom as requimd when needed.
With components or subsystems out-of-service, overall core and containment cooling reliability it, main-tained by dam-matrating the operability of the remaining cooling equipment. The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components. However, if a failure, design deficiency, etc., caused the out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problem does not exist on the remaining components. For exasyle, if an out-of-service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the operability checks.
4.5 BASES 120 REV
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e 3.0 LIMITING CONDITIONS POR OPERATION 4.0 SURVEILLANCE REQUIREMENTS E.
Safety / Relief Valves E.
Safety / Relief Valves 1.
During power operating conditions 1.
a.
A minimum of seven safety / relief and whenever reactor coolant pressure valves shall be bench checked or is greater than 110 psig and replaced with a bench checked temperature is greater than 345 F.
valve each refueling outage.
The nominal setpoint of all opera-a.
The safety valve function (self-tional safety / relief valves shall actuation) of seven safety /
): 1108 psig.
4 relief valves shall be operable.
b.
At leas t two of the safety / relief b.
The solenoid activated relief valves shall be disassembled and function (Automatic Pressure inspected each refueling outage.
Relief) shall be operable as required by Specification 3.5.E.
c.
The integrity of the safety / relief valve bellows shall be continuously monitored.
d.
The operability of the bellows monitoring systes shall be demon-strated at least once every three months.
i 2.
Inservice inspection and testing of components shall be conducted in accordance with Specification 4.XX.
3.6/4.6 127 REV
O O
O 30 LIMITING CO.'fDITIONS FOR OPERATION I4. 0 SURVEILIANCE Rl!RUIRDENTS F.
deleted F.
deleted vi G.
Jet Pumps G.
Jet Ptaaps Whenever the reactor is in the Startup Whenever there is recirculation flow with the or Run modes, all Jet pumps shall be oper-reactor.in the Startup or Run modes, jet pump able. If it is determined that a jet Pump is operability shall be checked daily by verify-inoperable, the plant shall be placed in a ing that all the following conditions do not cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
occur simuftaneously:
1.
The two recirculation loop flows are unbalanced by 15% or more when the recirculation pumps are operating at the same speed.
2.
The indicated value of core flow rate is 10% or more less than the value de-rived from loop flow measurements.
36/4.6 128 REV
Bases Continued 3.6 and 4.6:
he safety / relief valves have two functions; i.e. power relief or self-actuated by high pressure.
%e solenoid actuated function (Automatic Pressure Relief) in <tich external instrumentation signals of coincident high dryvell pressure and low-low veter level initiate opening of the valves. This function is discussed in Specification 3 5.E.
In addition, the valves can be operated manually.
he safety function is perfomed by the same safety / relief valve with self-actuated integral bellows and pilot valve causing main valve operation. Article 9 of the AINE Pressure Vessel Code Section III Nuclear Vessels requires that these bellows be monitored for failure since this vould defeat the safety function of the safety / relief valve.
It is realized that there is no way to repair or replace the bellows during operation and the plant must be shut down to do this. The thirty-day peried to do this allows the operator flexibility to choose his time for shutdown; meanwhile, because of the redundancy present in the design and the continuing monitoring of the integrity of the other valves, the overpressure pressure protection has not been vi 4
ccepromised. De auto-relief function vould not be impaired by a failure of the bellows. However, the self-setuated overpressure safety function would be impaired by such a failure.
Provision also has been made to detect failure of the bellows monitoring system.
Testing of this system quarterly provides assurance of bellows integrity.
When the setpoint is being bench checked, it is prudent to disassemble one of the safety / relief valves to ernmine for crud buildup, bending of certain actuator members or other signs of possible deterioration.
Se program of safety / relief valve testing confoms to the requirements of 10 CFR 50, Section 50 55a(s).
Rese requirements are delineated in Specification 4.XX.
Bis inspection and testing progrsm, combined with the additional surveillance requirements contained i this section, provide a high degree of assurance that the safety / relief valves vill perfor:n as required il en needed, l
151 3.6/4.6 BASES REV
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G.
Jet pumps Failure of a jet pu=p nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow aree for blowdown following the design basis double-ended line break.
herefore, if a failure occurred, repairs must be made.
Se detection technique is as follows. With the tvc recirculation punps balanced in speed.to within
+ 5%, the flow rates in both recirculation loops will be verified by Control Room monitoring instnaments.
If the two flow rate values do not differ by more thnn 10%, riser and nozzle assembly integrity has been verified. If they do differ by 10% or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of v, loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10%
4 or more (with the derived value higher) diffuser measurements vill be taken to define the location
" within the vessel of failed Jet pump nozzle (or riser) and the plant shut down for repairs. If the potential blevdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pu=p vill "run out" to a substantially higher flow rate (approximately 115% to 120% for a single nozzle failure). If the two loops are balanced in flov at the same pump speed, the resistance characteristics cannot have chanEed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate.
'Ihe reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3% to 6%)
in the total core flow measured. is decrease, together with the loop flow increase, vould result in a lack of correlation isetween measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal vould be reduced because the backflow would be less than the nomal forward flow.
A noeste-riser system failure could also generate the coincident failure of a jet pay body; however, the converse is not true.
He lack of any substantial stress in the jet pung body makes failure impossible without an initial nozzle-riser system failure.
t 3.6/4.6 BASES
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3.0 LIMITING CONDITIONS FDR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.
Pressure Suppression Chamber -
3 Pressure Suppression Chamber -
Reactor Building Vacuum Breakers Reactor Building Vacuum Breakers a.
Except as specified in 3.7.A.3.b a.
The pressure suppression chamber-reactor below, two pressure suppression building vacuum breakers and associated in-chamber-reactor building vacuum strumentation including set point shall be breakers shall be opet'bke at all checked for proper operation every three times when the primary m ntainment months.
integrity is required. The set 4
point of the differential pressure b.
Inservice inspection and testing of e
instrumentation which actuates the cosaponer.ts shall be conducted in pressure suppression chamber-reactor building vacuum breakers shall be accordance with Specification 4.XX.
0.5 psi.
b.
From and after the date that one of the pressure suppression chamber-reactor building vacuum breakers is made or found to be inoperable for any reason, reactor operation is permiscible only during the suceed-ing seven days unless such vacuum breaker is socner made operable, provided that the repair procedure does not violate primary containment integrity.
3.7/4.7 163 asy
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENfS d.
The fuel cask or irradiated fuel is not being moved within the reactor building.
D.
Primary Contairunent Isolation Valves D.
Primary Containment Isolation Valves 1.
During reactor power operating conditions, 1.
The primary containment isolation valves Y
all isolation valves listed in Table 3.7.1 surveillance shall be performed as follows:
U and all primary system instrument line flow check valves shall be operable except At least once per operating cycle the a.
as specified in 3.7.D.2.
operable isolation valves that are power operated and automatically initiated shall be tested for sinalated automatic initiation and closure times.
b.
Inservice inspection and testing of components shall be conducted in accordance with Specification 4.XX.
3 7/4.7 170 REV
O V
G 3.0 LIMITING CONDITIONS FOR OPERATION k.O SURVRTr4NCE REQUIRDOffS Y
2.
In the event any isolation valve specified 2.
Whenever an isolation valve listed in in Table 3.7.1 becomes inoperable, reactor operation in the run mode may continue Table 3 71 is inoperable, the position of provided at least one valve in each line at least one fully closed valve in each line having an inoperable valve is closed.
having an inoperable valve shall be recorded daily.
3.
If Specification 3.7.D.1 and 3.7.D.2 cannot be met, initiate normal orderly shutdown and have reactor in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3 7A.7 171 REV l
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3:0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVPTTJANCE REQUIRENENPS 3.XX INSERVICE INSPECTION AND TESTING 4.XX INSERVICE INSPECTION AND TESTING Applicability:
Applicability:
Ipplies to co=ponents which am part of Applies to the periodic inspection and the reactor coolant pressure boundary and testing of components which are part of their supports and other safety-related the reactor coolant pressure boundary pressure vessels, piping, pumps, and and their supports and other safety-valves.
related pressure vessels, piping, pumps, and valves.
Objective:
Objective:
w E
To assure the integrity of the reactor
'Ib verify the integrity of the reactor coolant pressure boundary and the coolant pressure boundary and the operational madiness of safety-related operational m adiness of safety-pressure vessels, piping, pumps, and related pressum vessels, piping, pumps, valves.
and valves.
Specification:
Specification:
A. Inservice Inspection A. Inservice Inspection 1.
To be considered operable, Quality
- 1. Inservice inspection of Quality Group A, B, and C components shall Group A, B, and C components shall be perfomed in accordance with satisfy the requirements contained the requirements for ASME Code Class in Section XI of the ASME Boiler and Pressure Vessel Code and appli-1, 2, and 3 components, respectively, contained ir.Section XI of the ASME cable Addenda for continued service Boiler and Pr?ssure Vessel Code and of ASME Code Class 1, 2, and 3 compo-applicable Addenda as requimd by nents, respectively, except where 10 CFR 50, Section 50 55a(g), except relief has been requested from the Commission pursuant to 10 CFR 50, where relief has been requested from the Commission Pursuant to 10 CFR 50, Section 50.55a (g)(6)(i).
Section 50 55a(gJ(6)(i),
3.XX/4.XX XXX REV
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n 30 LIMITING CONDITIO!G FOR OPERATION 4.0 SURVRTTJANCE REQUIRDENTS
- 2. For Non-Confoming Lines which are not Service Sensitive, inspections required by 4.XX.A.1 during the first lO-year inspection interval shall be completed by the end of the 1978 refueling outage.
If these examim tions reveal no incidence of stress corrosion cracking, the examination schedule may revert to that specified in 4.XX.A.l.
- 3. For Non-Conforming Lines which are Service u
4 Sensitive:
o>
- a. 'Ihe velds and adjoining areas of bypass piping of the discharge valves in the main recirculation loops, and of the austenitic stainless steel reactor core spray piping up to and including the second isolation valve, shall be examined at each reactor refueling outage or at other scheduled or unscheduled plant cold shutdowns.
Successive examinations need not be closer than six months apart.
In the event three successive examinations find the piping free of unacceptable indications, the examination may be extended to each 36 month interval, plus or minus 12 months, coinciding with a refueling outage, and may be limited to one bypass pipe run and one reactor core spray pipe run.
3.XX/4.XX
1
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30 LIMITING CONDITIONS FOR OPERATION h.0 SURVEILIANCE REQ'JIRDENTS
- b. If Service Sensitive Lines other than those listed in 4.XX.B.3.a above are identified, the velds and adjoining areas of this piping shall be subjected to examination at each reactor refueling outage or at other scheduled or unscheduled plant cold shutdowns on a sampling basis.
Successive examinations need not be closer than six months apart.
If unacceptable flaw indications are detected in any branch run, the mmaining branch runs with similar functions and configumtions y
shal1 be examined.
In the event three suc-cessive examinations find the piping free of unacceptable indications, the examination schedule may revert to that specified in 4.XX.A.1 with the exception that all examina-tions nomally completed over a ten-year interval shall be completed each 80-month period.
B, Inservice Testing of Pumps and Valves B. Inservice Testing of Pumps and Valves 1.
To be considered operable, Quality Group A, B, and C pumps and valves
- 1. Inservice testing of Quality Group shall satisfy the requirements con-A, B, and C pumps and valves shall be tained in Section XI of the ASME Boiler perfomed in acconiance with the and Pressure Vessel Code and appli-requirements for ASME (ble Class 1, cable Addenda for operability of 2, and 3 pumps and valves, respectively, ASFE Code Class 1, 2, and 3 pumps contained in Section Xi of the ASME and valves, respectively, except Boiler and Pressure Vessel Code and where relief has been requested from applicable Addenda as required by the Commission pursuant to 10 CFR 50, 10 CFR 50, Section 50 55a(g), except Section 50 55a(g)(6)(1).
where relief has been requested from the Commission pursuant to 10 CFR 50, Section 50 55a(g)(6)(1).
3.XX/4.XX XXX REV
fO Ch s
NY Bases 3.XX and 4.XX he inservice inspection and testing program confoms to the requirements of 10 CFR 50, Section 50 55a(g).
Where pmetical, the inspection and testing of components classified into NRC Quality Groups A, B, and C vill confom to the requirements for ASME Code Class 1, 2, and 3 components contained in Section XI of the ASME Boiler and Pressure Vessel Code.
Using Regulatory Guide 1.26, Revision 3, " Quality Gmup Classifications and Standards for Water, Steam, and Badioactive-Waste-Containing Components of Nuclear Power Plants," as a guide, all hticello components have been classified into Quality Groups. 'Ihis classification serves as the basis for determining which ASME Code Class inspection and testing requirements are applicable to a given component. 10 CFR 50, Section 50 55a(g) requires components which are part of the reactor coolant pressure boundary and their supports to meet the inservice inspection and testing requirements applicable to components classified as ASME Code Class 1.
Other safety-re3ated components must meet the inservice inspection and testing requirements applicable to components classified as ASME Code Class 2 or 3 u
h The inservice inspection pmgram must be updated at 40 month intervals. The program for testing pumps and valves for operational readiness must be updated every 20 months. A description of the updated programs should be submitted to the NRC for review at least 90 days before the start of each period. A suggested fomat for this description is contained in Appendix A to reference (1).
'Ibe inservice inspection and testing program must, to the extent practical, comply with the requirements in editions and addenda to the ASME Code that are "in effect" no more than six nonths before the start of the period covered by the updated progmm. The tem "in effect" means both having been published by the ASME, and having been referenced in paragraph (b) of 10 CFR 50, Section 50 55a.
If a code required inspection or test is impractical, requests for deviations are submitted to the Conmiission in accordance with 10 CFR 50, Section 50 55a(g)(6)(1).
The infomation specified in Appendix B to reference (1) should be submitted for each deviation requested. Deviation requests should, if possible, be submitted to the NRC for revi.a at least 90 days before the start of each period.
Deviations identified during an inspection reriod nay be grouped and requested at the end of each calendar q' tarter.
It is expected th:.; a small number of deviations vill be identified during the inspection period, particularily che first period when new inspection and testing techniques vill be utilized. A requested deviation request may be considered acceptable to the Cbmission until a romal disapproval has been received.
Q p
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Bases 3.XX and 4.XX (Continued)
Small, hairline cracks in austenitic stainless steel piping in BWR facilities has been observed on geveral occasions.
Data indicates that Type 304 austenitic stainless steel piping in the reactor coolant pressure boundary of the boiling water reactor is susceptible to st ess corrosion cracking. Such cracking is caused by a combination of significant amounts of oxygen in the coolant, high stresses, and some sensitization of metal adjacent to welds.
Cracks have occurred in the heat affected zones adjacent to welds, but are not expected to occur outside these amas, provided the pipe material is properly annealed.
Pipe runs containing stagnant or low velocity fluids have been observed to be more susceptible to stress corrosion cracking than pipes containing a continuously flowing fluid during plant operation. Historically, these cracks have been identified either by volumetric examination, by leak detection systems, or by visual inspection.
Because of the inherent high material toughness of austenitic stainless steel piping, stress corrosion cracking is unlikely to cause a rapidly propagating failure resulting in a loss of coolant accident.
Although the probability that stress corrosion cracks will propagate far enough to create a significant safety hazani is slight, the presence of such cracks is undesirable. 'Ihe following steps have been taken to minimize this problem:
- 1. Where practical, pipe runs constructed of material susceptable to stress corrosion cracking and which contained stagnant or low velocity fluid have been replaced with materials not susceptable to cracking or they have been eliminated.
- 2. 'ihe reactor coolant leakage detection technical specifications have been amended to enhance the ability to detect unidentified leakage that may include through-wall cracks.
- 3. 'ite program of inservice inspection has been augmented to increase the probability of crack detection in lines susceptable to stress corzosion cracking.
This program confoms to the Commission's guidelines for plants with operating licenses U cference 2).
References:
- 1. letter from D. L. Ziemann, Chief, Operating Beactors Branch #2, USNRC, to L. O. Mayer, NSP, dated November 24, 1976.
- 2. NURE-0313, " Technical Report on Nterial Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," July,1977
/^)
SECTION 6 QUALITY GROUP CIASSIFICATION DRAWINGS V
System Page Main Steam System 6-2 Feedvater System 6-3 Reactor Recirculation System 6-4 f
Core Spray System 6-5 Residual Heat Removal Syste loop A 6-6 Residual Heat Removal System Icop B 6-7 High Pressure Coolant Injection System (steam side 6-8 High Pressure Coolant Injection System (water side 6-9 RCIC (steam side 6-10 RCIC (Water side 6-11 Standby Liquid Control System 6-12 Primary Containment System 6-13 Emergency Service Water 6 14 RHR Service Water 6-15 CRD Hydmulic Control Unit 6-16 Control Rod Drive System 6-17 Fuel Pool Cooling & Clean-up System 6-18 Compressed Air System 6-18 Condensate Service System 6-18 Reactor Building Cooling Water System 6-19 O
Reactor Water Clean-up System 6-19 Liquid Radvaste System 6-19 Traversing In-core Probe System 6-20 Excess Flow Check Valves 6-21 Key Indicates Quality Indicates Quality 4-- Group B Group A Indicates Quality cmnp c --->
- O 6-1
'- ~
~
Raset::r Head Vent IDV-2,,
IDV-3 r,
MS-22-1 1
=
h CV-7371 T CV-2372 To CRW FCV-7682 Primary Steam to Air Ejectors 9
l
\\
MS-24-2
--@h2Ste8m SUPP 7 to
\\
MS-24-1 13 Recombiner System AO-2-80 B
From Main Steam
/
CV-1243 " 4 6 fg Line A & B Only 9
- x
6
/e l4 D.
MS-22-2 MS-13-L XDV-1
~[A al pl AO-2-86 From Other O
O Three Main f
RV-2-71 Steam Lines
~~
v From Other RCIC Steam Supply From Steam Line C Only.HPCI Steam From Steam Line
/
B only
'B CV-2369 3N
[
h 2564
[
g 11 f
Reactor Head CV-2370
]N LA Seal Leak-Off Q
4
(
Turbine Stop p
Bypass Contro L Control Valve To Main Condenser Valve To CRW From other MD-1046 MD-2374 MO-2373 h l617
~d N
gg,,,
\\
h k
[ /BJ 2
f MD-1045
/
W Deaerating To Condenser P
/B I
e Primary Steam to Steam Seal System
o--
o o
1 i
I 7
Reactor Clean up W
Return Reactor Vessel i
s*
From RCIC I
From HPCI s -
W 91-2 W-94-2 W-97-2
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g 97_1 g_
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N N
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l i
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s N
cs
/ AI REACTOR AN ca VESSEL RHR SYSTEM gTEM
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7 l
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) [4-MO 2-54A M0 2-43B I
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4l_ 2 ^ 1.. g N Recirculation
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N
/
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g p]
p o
LJ
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Service H/[HFromConde CST-96 Water MD-1751 A0 13A Q
CS-13-1 k
MD-1753 i
K'CSrd0s sate Service
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<h X 9
9 M - 75'
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/
CS-13-2 Vessel X
2 M
i j r'
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d RV-1745 W
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MD-1749 HO-1750 CS-8-2
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CS-9-2 7
y CS-9-1 1 r
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N 1 CS-10-2 MI CS CS-10-1 m
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z s
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g
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+ CS-3-1 ORW ORW From Condensate Storage CORE SPRAY SYSTEM
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RV-2025 MO-2027 MO-2026 h
h RHR-21 CV-2024 4 j, kT JQ g
fr' s CST-97 CS From CNDS Service Bx
<From CNDS Serv Water Water ;
MO-2020 Reactor M~
CST-99 Vessel To Contain-1
[ CST-ment Spray 7 r,
r, b
CST-92 w-From CNDS A0-10-46A MO-2014 CST-95
/ Serv Water
[
Fron
(
r, W hecirc RHR-6-1 4
y
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/
N MO-2012 MO-2006 From Core + 5 g
s*
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., Torus..__..
7 h\\
MO-2008-)
Waste Surge Tank -
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S
(
RHR-8-1 RHR 7 CV-1996 F-
/ 4 R0 2031
?From RHR Emergency Fuel 7
Pool Cooling
- J-Loop B
)
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~ ' h' rom f
CV-1994 " Gy FPCC
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a HX f/
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CST-90 RV-1992 RV-1990 From CNDS
/
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j h
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RHR-2-1 RER-4-1 RHR Service f
Condensatq7 Water Storage 9
RHR-3-1 RHR-47 RHR-10-1 RESIDUAL HEAT REMOVAL SYSTEM LOOP A
CST-119 From Condensate % ; g g
2023 CST-93
- CST-MO-2021 g A0-10-46B RV-2005 MO-2015 j To Containment N. \\
RHR-6-2 MO-2013 S
L N pray f' I
~-
2 m rc0 MO-2007
- )& ~
~
2 CST-91
\\
h M
CST-88 l [C+-MO-2011 To Recirculating Pump JI
\\From Condensate Service Water fg MD-2009
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To RHR
{l 3.
gc Loop A d
l
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5
"* '"" ** "" ?
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RHK-18-2 CV-1997 Prom Conden-g M
'do-2003 sate storage RHR-3-4 MD-1987 N
l/;
D.]
Y
/3
] [
RHR-2-4 RHR-54 Fred RHR n
RHR-30-4 M Loop A RV-1993 h
N CV-1995 I
g RV 991 RHR-4-2 MD-1988 RIIR-3-2 t
RHR-1-2 k
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id 1
k f,
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g RHR-30-2 U
RESIDUAL HEAT REMOVAL SYSTEM LOOP B
r p
p
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[
MO-2034 MO-2035 I
\\
/
l
'I Ah O
h O
N X
./
.B\\
Turbine Valves:
REACTOR I
Control Ston
- VESSE, HO-8 HO-7 HO-2036
\\
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j l
9-9.
1 i
S 7'
I
{
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=
T
^
,HPCI P c i
(
3 Drive Turb HPCI-3 ST-2045 r,
s.1 CV-204 y
/
l X HPCI-59 y
HPCI-58 HPCI-4 b
[ HPCI-7 aE7 JU4
- vH.
7
\\,
Condenser PSD-2038 CV-2046 &
HPCI-10 HPCI-9 nV-2056
- M
/
k PCV-3492 To Main N
K 7
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h
_lL I HPCI-15 HPCI-13 HPCI-12
/
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h M
HPCI-70 L
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HPCI-65
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f' f
f k
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I I
j HPCI-16 HPCI-14 HPCI-27 \\ HPCI-24 7
g i
HMfgm l
I Lube 011 Cooler 8
8 ST-2052
- j;-
7 HPCI-71 Torus f' Condenser
/
~
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Cooling Water ;
'/,
Refer to CI (Water side)
HIGH PRESSURE C001. ANT INJECTION SYSTEM (STEAM SIDE)
l f^~
()
,r '
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Y.)
/ B1 m
' *-To Condensate Storage Tank t
HPCI-32 + *
\\
T MD-2063 1
e
- -HPCI-33 l
Reactor l
RV-2064 To Condensate D
Vessel Storage MO-2071 + g F Turbine l
A0-23-18
(
l l
MO-2068 MO-2067 s
x9 x
- - $' rom FI RWCU
{je,} --l' j---l'-j- - y y -c Feedwater CV-2065
.B ng m
e$
y; Cooling Water Return To Core Spray Test Line
~
^
g HPCI-42
/
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/
TORUS MO-2061 HPCI-31 M0-2062 i
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N i
HIGH PRESSURE COOLANT INJECTION SYSTEM (WATER SIDE)
Q j'~~)
O e
NJ Y
L}
MO-2075 M0-2076
\\
o a y
N""*?"
i 2bs e
i H
HO MO-2078 From Condensate l
4 4
Storage Tank T
}
S
(
Main Steam es RCIC-3 Yg ST-2081 CV-2079]
/
- ' ump Drive Turb
_3- -
+ To 4
Reactor o
CV-2082A ee RCIC M o
PSD-2089 RCIC-23 lgp7
\\
RCIC-9 RCIC-24 To M p
g Condense RCIC-1
' To Barometric gg K~~
RCIC-10 PCV-2092 t
Condenser RCIC-61 o
FROM VACUUM ST-2088 B
PUMP 1
RCIC-59 + \\--
RCIC-18RCIC-17 RCIC-16 D<}-
RCIC-20 RV-2097 B
~
f f
RCIC-26 N
!/I lM2 i
-- )
MO-2096 I
/
4 1
RCIC-5 RCI - 5 Torus y
M
'U 1
z'B _j I/ h' N l
Lube Cooler RCIC-14 # From Condensate Pump RCIC (Steam Side)
'N O
-2 To Condensate
^
y N Storage Tank RCIC-41 MO-2102 To Condensate l
g l
Storage RCIC-32
! Reactor MO-2101--yp Vessel i
U
- O e RV-2103 7
i
% MO-2110 h.
f l
RCIC PUMP 4
(
i DRIVE I
s il A0-13-22 TURBI l
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{
MO-2107 MO-2106 d
f i
N.
,/
ri Q
O a
r-
- p, y
pt t
__ _- y _.. g e Feedvater 1
's v
Ai 0
Core Spray Test Line
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[ TORUS RCIC-37
\\ L *'
- l *-CV-2104
\\e o
{
[~-l. J l MO-2100 RCIC-31
\\
\\g 4
A N
RCIC (WATEh SIDE
(
To rest Tank u
XP-13 XP-5 y
g r
a 6
1
~
Standby Liquid Control Tank ll-MB
{
Drain From Test Tank 5N XP-18 XP-17 %
IP-6 XP-24-1 o
9
--uiat.r XP-1 IP-7 B
ll-14A ng RV-11-39A
\\
y XP-8
.XP-4-1 XP-3-1 XP-2-1
[
~~I Reactor d
Vessel XP-4-2 XP-3-2 k
lM XP-2-2
-L
'T h'
q' s
t U
L PxRv-11-3,o Accumulato
-~
STANDBY LIQUID CONTROL SYSTEM
pm
(.
d C)
C/
To Nitrogen Storage N
h CV-3269 f To Standby Gas Treatment cud Reactor Building Plenum V-326
- \\ AO-2387 CV-3267 A0-2386 M-23X X M-23
/
Air Purge i
A0-2377 CV-2385 r
iding I
l l
?
96 2
Back PRIMARY Vacuum relief N
?-
CONTAINMENT A0-237 8 -+ _M DWV-8-2 from Secondary m
4 CV-7436 X ntainment A0-2379 +
s
\\
)
u A0-2383 CV-7437
\\
A0-2380 +
M
?
i
(
)
i CV-2384 1
Torus I
DWV-8-1 J
Vacuum Relief from Secondary CV-7440
"**I""*"E
\\ToDrywell Pressurization System PRIMARY CONTAINMENI SYSTEM
p).
O
[
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L/
'w /
/
/
l
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2
./ SW-16 j
+-SW-18 r
r-----
g Standby Diesel Generators ESW-5-2 Reactor ESW-5-1---,g
}{h f
d Building i
ESW-3-q ESW-7 ESW-4-1 ESW-4-2 m
ESW-2-1 W-11d f
ESW-1-2 A
ESW SW-118I ESW-1-1 C
~
SW-105-1 SW-105-2 /
~
SW-110-2ESW-110-4+
SW-112-2X
"~' 2
~
SW-106-1+
SW-106-2 m
n J
gSW-107-1 s
SW-107-2 SW-109
. SW-11
__SW-111-4k _SW--113-22 1
1 ECCS Pump Motors and Vent Units SW-112-1 SW-110-3 SW-110-SW-114
/
l C
y EMERCENCY SERVICE WATER PUMPS V SW-111-3 SW-111-1 SW-115 SE-113-1 i
EMERGENCY SERVICE WATER
fm fg 7-~S i
_/
./
RHR-SW-4-2 RHR-SW-12 RHR-SW-14 h
&[
RHR-SW-3-2 X
AA RHR-SW-2-2
{+ -RHR-SW-4-1 RHR-SW-13s1 2
) [e-RHR-SW-3-1 RHR-SW-1-4 RHR-SW-2-4
\\ " [ " ~b ~ ~l
"~
~
DM-49-2 DM-49-1 DemineralizedWater\\
\\
-RHR-SW-2-3 h
/ e-RHR-SW-2-1 /( f lc
/
t
-1 RV-3203 5
RV-3202 +
-RHR-SW-1-3 l
C [r,
{
[
[,*-- RHR-SW-1-1 0
'l o
n Q
O BH-316-1 BH-316-2 RHR RHR
~
~
~
~
Demineralized Water HX HX CV-1728 CV-1729 g
\\
C RHR SERVICE WATER PUMPS
~
To Cire Water Discharge
O 1
HYDRAULIC CONTROL UNIT (typical of 121)
From Drive Water Header e
i M
M S0-121g pgS0-120
^
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N-50-122 S0-123
.M.I\\
iM1 138 102 I
I b
OI p' 101 Control w
g Drive l
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CV-127 CV-126
~
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115 2
i m
114 5
g g
A 112
[
i 107
~
l v
l To Scram Accumulator 1
Discharge Header O
t
~
6-16
O O
o i
l C
D
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- 5-Q Scram Risers
/ _,[
N f_
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i N,
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h i1PH B
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9 Pi
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h
/
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d e
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cl 4
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p 1
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RV-3-34 '
/
c 1
W-3-33 Piping Class B Valve CorrRot non nRivE SYSTEM (SCRAM DISCHARGE PIPING)
52-1 Fuel Storage P
[
PC-19-1 O
Q f
^
PC-20-L PC 9-2 52-2
\\;
PC-20-2
'N b
\\.
- (
'N b
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[
Instrument Air
/
f, H l
- o Containment 5
!"l T
(l CV-1478
/
AS-39 Instrument Air to Torus
(
2
]q g Service Air To Containment i
COMPRESSED AIR SYSTEM DM-58
/
f N
2 y
k
/5l CONDENSATE SERVICE SYSTEM 6-18
To Cooling From Reactor Building eat xchangers Water Pump l
f
,Ci c
V HD 1426 j
RBCC-15 o
J N
N To Drywell Coolers Co a
REACTOR BUILDING COOLING WATER SYSTEM rom Cleanup HPCI I MW Heat Exchangers
\\
MO-2397 MO-2398 From Reactor #
Recire Loop B RC-1
/
From Reactor To Cleanup Vessel Drain Recirc Pumps REACTOR WATER CLEAN-UP SYSTEM From Drywell Equipment Drain 1 rD A0-2541A A0-2541B A0-2561A A0-2561B
\\
g
/
/
u h
d' h
d i
To Floor Drain To Waste Collector Tank Collector Tank LIQUID RADRASTE 6-19
O O
O
[
's j
4 Reactor Vessel Shear Valves p
Ball Valves Index Mechanisms 4
TIP l
N To Drive 7
TIP 1-2 2-2 l
M N
?
TIP l-3 2-3 4
i
\\
.l 7
TRAVERSING IN-CORE PROBE SYSTEM
PENETRATION l~ 'l l
MANUAL VALVE 1,' LINE I
bl
,/TO INSTRUMENT 1" LINE 4'
$$RY f
!K M
SYSTEM EXCESS-FLOW l
l TYPICAL OF X-27A THROUGH X-52F EXCEPT X-28F O
VESSEL FLANGE PENETRATION HEAD SEALS
_ _I
~ I r
1 MANUAL VALVE h
[TOINSTRUMENT l
N RACK l
r r
EXCESS FLOW i
j, _{
CHECK VALVE X-28F ONLY o
"M 6-21
_ _ _ _ _.. _ _ _ _ _ _ _ _ _ _.. - - _ _