Text

Boston Edison Co Response to Generic Ltr 84-11: Insp of BWR Stainless Steel Piping, Summary Overview
	ML20213F263
Person / Time
Site:	Pilgrim
Issue date:	04/30/1987
From:	Gordon G, Ramp K; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20213F230	List: BECO-87-071, Forwards Rept Describing Generic Ltr 84-11/IGSCC Related Weld Insps,Per Verbal Request from NRC Project Manager. Valves 12-I-16 & RPV-N9B-1 Added to RFO 7 Insp List.Igscc Augmented Insps to Be Completed by 870701; ML20213F263;
References
	GL-84-11, NUDOCS 8705150275
	Download: ML20213F263 (79)
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{{#Wiki_filter:.. BOSTON EDISON COMPANY'S RESPONSE TO GENERIC LETTER 84-11: " INSPECTION OF BWR SS PIPING"

SUMMARY

OVERVIEW APRIL 30, 1987 r Prepared By: / K. S. Ramp Senior Engineer Reviewed By: / lN C. M. Gordon, Manager Fuel and Plant Materials Technology GENERAL ELECTRIC COMPANY San Jose, California G705150275 070506 PDR ADOCK 05000293 O PDR

TABLE OF CONTENTS PAGE .1 1.0 EXECUTIVE

SUMMARY

.1

2.0 INTRODUCTION

.5 3.0 CHRONOLOGICAL REVIEW OF NRC/BECo COMMUNICATIONS REGARDING IGSCC AT PILGRIM NUCLEAR POWER STATION . 10 4.0

SUMMARY

OF MITIGATING ACTIONS IMPLEMENTED.. AND PLANNED AT PILGRIM NUCLEAR POWER STATION 10 INSPECTION OF GENERIC LETTER d4-11.....'........ 5.0 IGSCC SUSCEPTIBLE WELDS AT PILGRIM NUCLEAR POWER STATION . 19

SUMMARY

OVERVIEW OF PILGRIM NUCLEAR POWER..... 6.0 STATION RESPONSE TO GENERIC LETTER 84-11 4 23 7.0 REFFRENCES 4 ,.m.- y .y,,

a LIST OF TABLES PAGE Status of Pilgrim Nuclear Power Station .2 Table 1: Class I Piping Welds. Inspection Statistics and Plans for Pilgrim Table 2: Nuclear Power Station GL84-11 IGSCC Susceptible.................. 3 Piping Welds. Status of Pilgrim Nuclear Power Station , 11 Table 3: Class I Piping Welds. Pilgrim Nuclear Power StationGL84-11 IGSCC Susceptible Welds.. 12 Table 4: Pilgrim Nuclear Power Station GL84-11 IGSCC Table 5: Susceptible Welds Scheduled for ExaminationDuring Re 16 Inspection Statistics and Plans for Pilgrim Table 6: Nuclear Power Station GL84-11 IGSCC Susceptible . 17 Piping Welds...................... =_.

~ F LIST OF ATTACHMENTS ATTACHMENT 1 MATRIX OF IGSCC INSPECTION ORDER EXAMINATION (REF. 6) ATTACHMENT 2 PNPS ISI IS0 METRICS FOR IGSCC SUSCEPTIBLE WELDS (REF. 6,7) ATTACHMENT 3 WELDS NOT SCHEDULED FOR EXAMINATION (REF. 6) ATTACHMENT 4 HYDROGEN WATER CHEMISTRY PROGRAM FOR PILGRIM NUCLEAR POWER STATION (REF. 6) ATTACHMENT 5 PRELIMINARY LIST OF PNPS WELDS SUSCEPTIBLE TO IGSCC AND SCHEDULED FOR EXAMINATION DURING REFUEL OUTAGE NUMBER 7 (REF. 12) ATTACHMENT 6 CRACK EVALUATION ANd REPAIR CRITERIA (REF. 2) ATTACHMENT 7 UPDATED PNPS ISI IS0 METRICS FOR IGSCC SUSCEPTIBLE WELDS I iii 4 -e-, r~. ,or- ,,y -,r-7w--- r-y, -.,,-<y--,, e---------,,,-f - -. v

4 1,0 Executive Summary Boston Edison Company's (BECo) efforts to mitigate intergranular stress corrosion cracking (IGSCC) at Pilgrim Nuclear Power Station (PNPS) include extensive piping replacement, heat sink welding, corrosion resistant cladding and S implementation of Hydrogen Water Chemistry (HWC). Furthermore, BECo. is installing a Crack Arrest Verification System (CAVS) in order to monitor crack 4 growth and demonstrate the mitigating effects of HWC. With the exception of one weld overlay repair applied to the jet pump instrument seal and some cracks in the thermal sleeve safe-ends reported during RFO 6, there are no other cracked 1 i welds remaining in-service at PNPS. A summary of the current mitigation status of PNPS Class I piping welds is presented in Table 1. 1 Inspection statistics and plans for PNPS GL 84-11 IGSCC susceptible piping welds 4 is presented in Table 2. For each pipe size, the percentage of welds previously inspected during RFO 6 and planned for reinspection during RFO 7 meets or exceeds the 20% sampling size specified in GL 84-11. With the exception of two Reactor Water Cleanup System piping welds that are inaccessible due to pipe l whip restraints, all IGSCC susceptible piping welds at PNPS that were not previously inspected in RFO 6 will be inspected during RFO 7. Hence, BECo's RFO 7 ,1 inspection program for PNPS complies with Generic Letter 84-11 requirements. 'h y In summary, BECo's plans and actions to address the IGSCC concern for Pilgrim Nuclear Power Station piping welds have met or exceeded NRC Generic Letter 84-11 requirements. 2.0 Introduction 4 Boston Edison Company (BECo) has implemented a comprehensive piping i inspection and replacement program at Pilgrim Nuclear Power Station (PNPS) in response to the Nuclear Regulatory Commission (NRC) August 26, 1983 Order and Generic Letter 84-11 (Ref. I and 2). BECo's plans and actions to address 4 f r y 1 i ~ ~.

~ r Tcbla la STATUS OF PILGRIM NUCLEAR POWER STATION CLASS I PIPING WELDS 1 s / _ ~ 3 1 ~ 2 GL 84-11 IGSCC SUSCEPTIBLE WELDS . TOTAL NO.

fW

<~ s CIASS I ICSCC RESISTANT CLAb.? I WELDS SUSC. WELD - SYSTEM SIZE ' NELDS REPIACED l CARBON STEEL l HSW l CRC MATERIAL OVERLAY OTHER TOTAL Recire. _28" 36 36 0 0 0 14 0 0 14 7 7 0 0 0 0 0, 0 0 22" 30 30 ' 0 0 0 10 0 0 10 12" RIE 10" 25 2 ~ _y

14 _

11 0 0 -3 0 0 3 18" 31 ~ '28 0 0 1 -14 0 'O 14 (IIS)4 6" 1 - 0 0 0 0-1 0 0 1 '. (lis) 4" 31 0 10 0 0 22 0 0 22 CORE SPRAY 12" 2 0 0 0 0 2' 0 0 2 10" 45 - _ 15 17 2 0 21 0 0 21 RWCU 6" 51 ^26-15 1 ~~~'O 15 0 0 15 4" 22 '0 0 0' O 22 0 0 22 JET PUMP 4" 2 0 0 0 0 1 1 0 .' 2 ~ INSTR TOTALS: 283 156 53 3 1 125 1 0 126 Notes: ' ~ ~ ~

1. These totals represent all Class I piping welds at PNPS which are equal to or greater than four inches in diameter and connected to the cactor coolant

.c pressure boundary, llowever. some of these welds are not considered susceptible to r IGSCC rince they a.e outside the Generic Letter 84-11 criteria (eg., operating temperatuie 2 200 F).

2. These totals represent Class I piping welds which fall within the boundary?

establish in Generic Letter 84-11 (i.e.,24" in diameter, operating temperature 2 200 F and connected to the reactor coolant pressure boundary out to the second isolation valve) but are ccasidered resistant to IGSCC due to the materials (i.e., 316NG) and/or welding processes.

3. These totals represent those piping welds which are still considered susceptable to IGSCC. For instance, those welds which are between the "new" pipe (i c.,316NG) and the "old" pipe (i.e.,304) are included becauseuf the potenital susceptibility to IGSCC.

These welds are included in the Generic letter 84-11 Augmented Inspection Program.

4. "HS" denotes lleat Spray Piping..~

INSPECTION STATISTICS AND PLANS FOR h iJIM NUCLEAR POWER STATION TABLE 2: GENERIC LETTER 84-11 IGSCC SUSCEPT UT1. PIPING WELDS (PAGE 1 0F 2) PREVIOUSLi INSPECTED IGSCC SUSC. WELDS IGSSS SUSC. WELDS NOT PREVIOUSLY INSPECTED TOTAL _ NUMBER NUMBER PLANNED % PLANNED FOR NUMBER NOT NO. PLANNED % PLANNED GL84-11 IGSCC INSPECTED FOR REINSPEC-REINSPECTION INSPECTED FOR INSPECTION FOR INSPECTION SIZE SYSTEM SUSLEPTIBLE RF0 6 TION RF07 RF07 RF06 RF07 RF07 28" RECIRC 14 14 3 21 0 20" RHR 3 3 2 67 0 18" RHR 14 14 3 21 0 12" RECIRC 10 10 2 20 0 ) f CORE SPRAY 2 2 1 50 0 Avg. 25% 10" CORE SPRAY 21 21 5 24 0 ., 3 -

TABLE 2: INSPECTION STATISTICS AND FLANS FOR PILGRIM NUCLEAR POWER STATION GENERIC LETTER 84-11 IGSCC SUSCEPTIBLE PIPING WELDS (PAGE 2 0F 2) PREVIOUSLY INSPECTED ICSCC SUSC. WELDS IGSSS SUSC. WELDS NOT PREVIOUSLY INSPECTED TOTAL NUMBER NUMBER PLANNED % PLANNED FOR NUMBER NOT NO. PLANNED % PLANNED GL84-11 IGSCC INSPECTED FOR REINSPEC-REINSPECTION INSPECTED FOR INSPECTION FOR INSPECTION SIZE SYSTEM SUSCEPTIBLE RF0 6 TION RF07 RF07 RF06 RF07 RF07 6" RHR(HS) 1 1 0 0 0 RWCU 15 14 3 21 1 1 100% Avg. 20% 4" RHR(HS) 22 22 5 23 I) d1) 0 RWCU 22 20 5 25 ) 1 100% JP INSTR. 1 0 Avg. 23% i l d 4" JP INSTR. I 1 1 100% (OVERLAY) Notes:

1. Not accessible due to pipe whip restraints.

2. Jet pump Instrument Nozzel N9B-1 was PT during RFO#6. A UT inspection will be conducted during RFO#7.

3. Jet pump Instrument Nozzel N9A-1 was PT and UT during RFO#6 and 7. -- - _ _____- _ -_

the intergranular stress corrosion cracking (IGSCC) concern are documented in recent transmittsis to the NRC. However, BECo felt that it would be beneficial to create a single document that contains a comprehensive summary of the PNPS response to Generic Letter 84-11, 3.0 Chronological Review of NRC/BECo Communications Regarding IGSCC at Pilgrim Nuclear Power Station In response to increasing incidence of BWR pipe cracking, the NRC issued an Order on August 26, 1983 requesting BECo to conduct extensive ultrasonic l (UT) inspections of PNPS piping welds during the sixth refueling outage (RF0 6). In response to the Order, BECo submitted a letter to the NRC dated December 8, 1983 (Ref. 3) stating plans to commence UT inspections of 1 IGSCC susceptible piping in accordance with the December 13, 1983 Order. This letter informed the NRC that BECo was considering piping replacement in the event that inspections identified cracked welds. Following discov-ery of crack indications, BECo decided to implement Recirculation System piping replacement during RF0 6. In April, 1984, the NRC issued Generic Letter 84-11; " Inspections of BWR Stainless Steel Piping", which specified actions that would be considered an acceptable response to the current IGSCC concerns. This document called for implementation of a reinspection program that should commence during the first scheduled outage following piping inspections performed in response to the August 1983 Order. Generic Letter 84-11 also specified categories of piping welds that should be inspected, defined an acceptable sampling size, and discussed other issues such as UT examiner competence, leak detection, crack evaluation and repair criteria. BWR operating reactor licensees were requested to submit plans relative to inspections for IGSCC and interim leakage detection. -- -

BECo responded co Generic Letter 84-11 by letter dated June 4, 1984 during BECo informed the NRC that inspection and piping replace-RF0 6 (Ref. 4). 03, ment operations were being conducted in accordance with IE Bulletins 82-26, 1983 Order, and stated the Order's interim meas-83-02 and the August l ures would be used as leakage guidelines pending submittal of a Technica BECo also communicated plans to use the crack Specification Amendment. Attach-evaluation and repair criteria contained in Generic Letter 84-11: SCC ment 2 until a guideline for flaw evaluation and remedial actions for IG approved by the Boiling Water Reactors Owners Group (BWROG) becomes available. the NRC staff and representatives of BECo met to discuss On June 15, 1984, l between the resolution of crack indications found in Inconel weld materia S (Ref. 5). the safe end and the nozzles of the recirculation system at PNP Crack indications were identified by dye penetrant tests during resurfacing of the nozzle butters for recirculation system safe end replacement. General Electric Company proposed rewelding, providing that sufficient Otherwise, Inconel 182 butter remained on the nozzles following machining. local post weld heat treatment or halfbead repair could be applied to In addition, BECo announced their facilitate restoration of the buttering. plan to implement Hydrogen Water Chemistry (HWC) in order to mitigate to The Staff stated that they had no objections cracking in the future. The NRC requested that BECo submit a report proceeding with repair plans. d during regarding the PNPS piping repair and replacement program implemente RF0 6. supplemental letter dated In a letter dated September 11, 1984 and a 1984 (Ref. 6,7), BECo submitted reports to the NRC that re-November 9, 3 Order. viewed actions undertaken at PNPS in response to the August 26, 198 identified welds examined in accordance with the Order, This response d carbon presented UT inspection results, and provided stress r"% index an The matrix of IGSCC Inspection Order content information where available. examination data is presented in Attachment 1. Isometric sketches that i

show locations of inspected welds and portions of the piping which were Attachment replaced by Type 316NG SS piping are presented in Attachment 2. 3 contains a list of welds for which BECo requested an exemption from inspection along with technical justification for not conducting these inspections during RF0 6. The September 11, 1984 BECo report (Ref. 6) included discussions about BECo informed the NRC of the decision te replace piping replacement plans. from the a portion of the Residual Heat Removal (RHR) System piping (i.e., Reciret.lation System piping connections out to the drywell penetrations) as well as the entire Recirculation System piping in order to eliminate interference and boundary design specification problems and thus achieve a BECo discussed plans to replace a portion more complete remedy for IGSCC. of RER piping outside the containment that included a weld with several UT indications (10-IB-14A), and corrosion resistant clad the existing weld within the penetration and the penetration to new piping sub-asseitbly In addition, BECo material using IGSCC resistant Type 308L weld material. portion of the Core Spray System piping communicated plans to replace a from the manual isolation valves to the penetratipn, and the suction piping in the ASME Class 1 portion of the Reactor Water Cleanup (RWCU) System from It was stated that the RHR piping connection to the drywell penetration. last pass heat sink welding would be implemented on the Core Spray and RHR systems where new Type 316NG stainless steel piping connects with existing Type 304 SS within the drywell at the reactor containment penetrations. (Ref. 6) also included a discuwsion of The September 11, 1984 BEco report corrective actions planned at PNPS for IGSCC affected piping components. BECo reported that a safety evaluation for operation with cracked recircu-lation inlet thermal sleeves had been conducted. The results indicated no safety concern associated with full-power operation exists since that the thermal sleeves are not part of the pressure boundary, and any poten-tial damage to jet pumps caused by thermal sleeve failure would be detected immediately with in plant instrumentation. For the remaining Core Spray _- - ~

System piping weld with crack indications less than 25% through wall, BECo presented plans to perform an evaluation using the GC acceptance criteria for the purpose of demonstrating that operation during the next fuel cycle was acceptable without repair. (BECo completed the evaluation and subse-quently replaced the portion of Core Spray System piping that contained crack indications.) Furthermore, additional information describing the recircul.ation nozzle repair program was provided to the NRC during a June 15, 1984 meeting. This information outlined the use of HWC to modify the reactor coolant as a means of mitigating IGSCC in BWR's (see Attachment 4). On November 8, 1984 (Ref. 8), BECo forwarded a letter to the NRC regarding PNPS jet pump instrumentation nozzle repairs. UT examinations had revealed crack indications in a band of the original furnace sensitized Type 304 SS jet pump instrumentation safe end that remained following safe-end replace-ment operations performed in 1969. BECo repaired this cracked weld with a full structural reinforcement weld overlay. Details of the weld overlay repair and the analytical basis for the design were included in Reference S. The NPC forwarded a letter to BECo dated December 4, 1984 (Ref. 9) in response to BECo's efforts to comply with the August, 1983 Order. This letter stated that the NRC accepted BECo's actions, and granted permissien to restart PNPS. However, BECo was requested to submit a Technical Speci-l fication Amendment concerning reactor coolant system leak detection and leakage limits. In addition, the NRC requested that BECo continue to study the cracking mechanism in the recirculation inlet thermal sleeves and submit a plan for mitigation or repairs prior to RF0 7. BECo submitted a Technical Specificatfor Amendment concerning leak detec-tion and leakage limits to the NRC on February 4, 1985 (Ref. 10). The Amendment details the new liaiting condition for operation (i.e., the rate of increas, of reactor coolant leakage into the primary containment from unidentified sources). Tha Amendment establishes a tro gallon per minute.

(2 gpm) limit increase averaged over any 24-hour period. This limiting condition for operation would apply only when the reactor has been in the RUN mode for greater that 24-hours. 1 1 f In a letter dated January 2, 1986 (Ref. 11), the NRC requested that BECo submit ' an augmented inspection program for RF0 7 and suggested that UT In personnel be requalified by an upgraded program at the EPRI NDE center. addition, BECo was informed of interim guidelines developed by the staff regarding derivation of allowable flaw sizes for flux welds which could require overlay repair of unrepaired welds that were previously justified for continued service. BECo's augmented inspection program was submitted to the NRC in a letter dated September 22, 1986 (Ref. 12). This letter included a preliminary list of welds identified as susceptible to IGSCC and scheduled for examin-ation during RPO 7, along with isometric drawings showing the location of identified welds. This information is prescated in Attachment 5. BECo stated that the scope of planned inspections is subject to change due to tentative plans to reduce the number of IGSCC susceptible welds through modifications of the Head Spray Line in the RHR sy' stem. BECo submitted a report on January 2, 1987 (Ref. 13) in response to the December 4, 1984 NRC request for continued study of recirculation inlet thermal sleeve cracking. This report summarized results of a etudy con-ducted by General Electric Ccapany to determine the cause of the isolated cracks found on the outer diameter of the N-2 recirculatica nozzle thermal sleeves. The focus of the study was to determine whether it was possible to sensitize the caterial in the area of concern. Results showed that sensitization could occur in th1s area, and therefore IGSCC was identified as the most likely cracking mechanism. EECo presented its plan to utilize j' HWC to mitigate IGSCC in recirculation nozzle thermal sleeves. ) f 9_

Summary of Mitigating Actions Implemented and Planned 4.0 at Pilgrim Nuclear Power Station ~ include implementation of HWC, extensive piping BECo's mitigation efforts Sink Welding (HSW) and limited Corresion Reef s-f replacement, limited Heat Other remedies commonly applied in BWR piping such as tant Cladding (CRC). solution heat treatment and Induction Heating Stress Iaprovement (IHSI) have not been utilized for 84-11 IGSCC st.sceptible Clast I piping at PNPS. There are no unrepaired cracked welds remaining in-service at the PNPS. However, a weld overlay repair was performed on a cracked jet pump instru-A summary of the current mitigation status of PNPS Class I ment seal weld. Welds identified as susceptible to piping welds,is presented in Table 3. IGSCC in accordance with Generic Letter 84-11 cre listed in Table 4. HWC is expeted to provide full protection against IGSCC in RHR, RWCU and It is expected that crack growth Recirculation System piping weldments. will be inhibited in presently cracked recirculation inlet thermal sleeves, and additional IGSCC initiation in this cocponent will be effectively implementation of HWC is expected to provide some mitigsted. In additicn, Spray System piping and the overlay repaired Jet protection to the Core Pump Instrument Seal. Furthermore, BECo has installed a Crack Arrest during RF0 7 in order to monitor crack growth Verification System (CAVS) and demonstrate the mitigating effects of HWC, Inspection of Generic Letter 34-11 IGSCC Susceptible Welds 5.0 at Pilgrim Nuclear Power Stat 3cn Pre-service excminations in accordance with ASME Section XI were condu;ted Furthermore, near:1y all of on all new piping welds installed during RF0 6. the rema.ining accessible IGSCC susceptible velds were UT examined or A to the August 26, 1983 Order. A radiographed during RF0 6 in response comprehensive inspection plan for the current refueling outage (RF0 7) was Welds identified for inspection developed to address Generic Letter 64-11. during RF0 7 are listed in Table 5. A compilation of the number of welds F exaAined in each system during P.F0 6 and planned for exanination during R 0 7 is presented in Table 6..

Tnbire 3: STATUS OF PIISRIM NUCLEAR POWER STATION CLASS I PIPING HELDS 1 3 TOTAL NO. 2 GL 84-11 IGSCC SUSCEPTIBLE WELDS CLASS I IGSCC RESISTANT CLASS I WELDS SUSC. WELD SYSTEM SIZE WELDS REPLACED l CARBON STEEL l HSW -l CRC MATERIA _L OVERLAY OTHER TOTAL Racire. 28" 36 36 0 0 0 14 0 0 14 22" 7 7 0 0 0 0 0 0 0 12" 30 30 0 0 0 10 0 0 10 RHR 20" 25 14 11 0 0 3 0 0 3' 18" 31 28 0 0 1 14 0 0 14 (IIS)4 6" 1 0 0 0 0 1 0 0 1 (HS) 4" 31 0 10 0 0 22 0 0 22 CORE SPRAY 12" 2 0 0 0 0 2 0 0 2 10" 45 15 17 2 0 21 0 0 21 RWCU 6" 51 26 15 1 0 15 0 0 15 4" 22 0 0 0 0 22' O O 22 JET PUMP 4" 2 0 0 0 0 1 1 0 2 i. INSTR TOTALS: 283 156 53 3 .'1 125 1 0 126 i Notes:

1. These totals represent all Class I piping welds at PNPS which are equal to or greater than four inches in diameter and connected to the reactor coolant pressure boundary, flowever, some of these welds are not considered susceptibic to IGSCC since they are ouaide the Generic letter 84-11 criteria (eg., operating temperature 2 20(PF).

2. These totals represent Class I piping welds which fall within the boundary establish in Generic Letter 84-11 (i.e.,24" in diameter, operating temperature 2 200 F and connected to the reactor coolant pressure boundary out to the second isolation valve) but are considered resistant to IGSCC due to the materials (i.e.,

316NG) and/or welding processes.

3. These totals represent those piping welds which are still considered susceptable to IGSCC. For instance those welds which are between the "new" pipe (i.e.,316NG) and s

the "old" pipe (i.e,304) are included because of the potenital susceptibility to IGSCC. These welds are included in the Generie Letter 84-11 Augmented Inspection Program.

4. "IIS" denotes IIcat Spray Piping.

, 11_

TABLE 4. PILGRIM NUCLEAR POWER STATION CL-84-11 ICSCC SUSCEPTIBLE WELDS. (Page 1 of 4) WELD INSPECT INSPECTION SIZE SYSTEM IDENT ISOMETRIC RFO# TECHNIQUE 28" kECIRC (14) 2R-N1A-1 ISI-I-2R-B 67 UT 2R-NIA-10 ISI-I-2R-B 6 UT 2R-N1A-11 ISI-I-2R-B 6 UT 2R-NIA-6 ISI-I-2R-B 6 UT 2R-NIA-7 ISI-I-2R-B 6 UT 2R-NIA-8 ISI-I-2R-B 6 UT 2R-N1A-9 ISI-I-2R-B 6 UT 2R-N1B-1 ISI-I-2R-A 6 UT 2R-N1B-10 ISI-I-2R-A 6 UT 2R-N1B-11 ISI-I-2R-A 6 UT 2R-N1B-12 ISI-I-2R A 6 UT 2R-N1B-13 ISI-I-2R-A 6 UT 2R-N1B-8 ISI-I-2R-A 67 UT 2R-N1B-9 ISI-I-2R A 67 UT 20" RHE (3) 10R-0-12 ISI-I-10-1A 6 UT 10R-0-7 ISI-I-10-1A 6 7 UT 10R-0-8 ISI-I-10-1A 6 7 UT 18" RHR (14) 10R-IA-12 IS1-I-10-1 6 UY 10R-IA-5 ISI-I-10-1 6 UT 10R-IA-6 ISI-I-10-1 6 'UT 10R-IA-7 ISI-I-10-1 6 UT 10R-IA-8 ISI-I-10-1 6 UT 10R-IB.12 ISI-I-10-1 67 UT 10R-IB-14 ISI-I-10-1 67 UT 1 10R-IB-16 ISI-I-10-1 6 UT 10R-IB-5 ISI-I-10-1 67 UT 10R-IB-6 ISI-1-10-1 6 UT 10R-IB-7 ISI-I-10-1 6 UT 10R-IB-8 ISI-I-10-1 6 UT 10-IA-14 ISI-I-10-1 6 UT 10-IA-15 ISI-I-10-1 6 UT 12" RECIRC (10) 2R-N2A-1 ISI-I-2R-A 67 UT i 2R N2B-1 ISI-I-2R-A 67 UT 2R-N2C-1 ISI-I-2R-A 6 UT 2R-N2D-1 ISI-I-2R-A 6 UT 2R-N2E-1 ISI-I-2R-A 6 UT 2R-N2P-1 ISI-I-2R-B 6 UT 2R-N2G-1 ISI-I-2R-B 6 UT 2R-N2H-1 ISI-I-2R B 6 UT 2R-N23-1 ISI-I-2R-B 6 UT 2R-N2K-1 ISI-I-2R-B 6 UT W PIIERIM NUCLEAR POWER STATION GL-84-11 IGSCC SUSCEPTIBLE TABLE 4. (cont. ) WELDS. (Page 2 of 4) INSPECT INSPECTION WELD SIZE, SYSTEM IDENT IS'JMETRIC RF0# TECHNIQUE 12" CORE SPRAY (2) 14 A-1 ISI-I-14-1 67 UT 14-B-1 ISI-I-14-1 6 UT 10" CORE SPRAY (21) 14-A-3 ISI-1-14-1 6 UT 14-A-19 ISI-I-14-1 6 UT 14-A-18 ISI-I-14-1 6 UT 14-A-17 ISI-I-14-1 6 UT 14-A-10A ISI-I-14-1 6 UT 14R-A-16 ISI-I-14-1 6 UT 14R-A-14 ISI-I-14-1 6 UT 14R-A-13 ISI-I-14-1 6 UT 14R A-11 ISI-I-14-1 6 UT 14-B-3 1S1 1-14-1 67 UT 14-B-20 ISI-I-14-1 67 UT 14-B-19 ISI-I-14-1 67 UT 14-B-18 ISI-I-14-1 67 UT 14-B-17 ISI-I-14-1 67 UT 14-B-10A ISI-I-14-1 6 UT 14R-B-21A ISI-I-14-1 6 UT 14R-B-21 ISI-I-14-1 6 ' UT 14R-B-16 ISI-I-14-1 6 UT 14R-B-14 ISI-I-14-1 6 UT 14R-B-13 ISI-I-14-1 6 UT 14R-B-11 ISI-I-14-1 6 UT 6" RHR (1) 10-HS-1 ISI-I-10-5A 6 UT 6" RWCU (15) 12R-0-10 ISI-1-12-1 6 UT 12R-0-11 ISI-I-12-1 6 UT 12R-0-23 ISI-I-12-1 6 UT 12R-0-5 ISI-I-12-1 6 UT 12R-0-6 ISI-I-12-1 6 UT 12-I-16 151-1-12-2 7 RADIOGRAPH 12-0-24 ISI-I-12-1 6 UT 12-0-25 ISI-I-12-1 6 UT 12-0-26 ISI-I-12-1 6 UT 12-0-27 ISI-1-12-1 6 UT 12-0-28 ISI-I-12-1 6 UT + 12-0-29 ISI-I-12-1 67 UT 12-0-30 ISI-I-12-1 67 UT 12-0-31 ISI-I-12-1 67 UT 12-0-32 ISI-I-12-1 6 UT PILGRIM NUCLEAR POWER STATION GL-84-11 ICSCC SUSCEPTIBLE TABLE 4. (cont.) WELDS. (Page 3 of 4) INSPECT INSPECTION VELD SILE , SYSTEM IDENT ISOMETRIG RF3# TECHN!QUE 4" RHR (22) 10-HS-10 ISI-I-10-5A 6 7 UT 10-HS-11 ISI-I-10-5A 6 7 UT 10-HS-12 ISI-I-10-5A 6 7 UT l 10-HS-13 ISI-I-10-5A 6 UT f 10-HS-14 ISI-I-10-5A 6 UT j 10-HS-15 ISI-I-10-5A 6 UT l 10-HS-16 ISI-I-10-5A 6 UI l 10-HS-17 ISI-I-10-5A 6 UT 10-HS-18 ISI-I-10-5A 6 UT 10-HS-19 ISI-1-10-5A 6 UT l 10-HS-2 ISI-I-10-5A 6 UT 10-HS-20 ISI-1-10-5A 6 UT 10-HS-21 ISI-I-10-5A 6 UT 10-HS-22 ISI-I-10-5A 6 UT 10-HS-23 ISI-I-10-5A 6 UT and RADIOGRAPH 10-HS-3 ISI-I-10-5A 6 UT 10-HS-4 ISI-I-10-5A 6 UT 10-HS-5 ISI-I-10-5A 6 UT 10-HS-6 ISI-1-10-5A 6 UT 10-HS-7 ISI-I-10-5A 6 , UT l 10-HS-8 ISI-I-10-5A 6 7 UT 10-HS-9 ISI-I-10-5A 6 7 UT 4" RUCU (22) 12-I-17 ISI-I-12-2 6 UT 12-I-18 ISI-I-12-2 6 UT 12-I-19 ISI-I-12-2 6 UT 12-I-20 ISI-I-12-2 (NOT ACCESSIBLE) 12-1-21 ISI-I-12-2 6 UT 12-I-22 ISI-I-12-2 (NOT ACCESSIBLE) 12-I-23 ISI-I-12-2 6 UT 12-I-23A ISI-I-12-2 6 UT 12-I-24 ISI-I-12-2 6 UT 12-I-25 ISI-I-12-2 6 UT 12-I-25A ISI-I-12-2 6 UT l 12-I-26 ISI-I-12-2 6 UT 12-I-27 I;' -I 2 6 UT 12-I-28 ISI-I-12-2 6 UT 12 1-29 ISI-I-12-2 6 UT 12-I-30 ISI-I-12-2 67 UT 12-I 31 ISI-I-12-2 67 UT 12-I-32 ISI-I-12-2 6 UT 12-I-33 ISI-I-12-2 6 UT -

PILGRIM NUCLEAR POWER STATION GL-84-ll IGSCC SUSCEP TABLE 4. (cont. ) WELDS. (Page 4 of 4) INSPECT INSPECTION WELD SIZE SYSTEM IDENT ISOMETRIC, RFO# TECHNIQUE 3 4" RWCU (cont.) 12-I-34 ISI-I-12-2 67 UT 12-I-35 ISI-I-12-2 67 UT 12-I-36 1S1-1-12-2 67 UT 4" JET PUMP (2) RPV-N9A-1 'ISI-I-54-4 67 UT (OVERLAY) RPV-N9B-1 ISI-I-54-4 7 UT TOTAL: 126 WELDS 3 4 m. I 1 I l i I 4 .5-t m

PILGRIM NUCLEAR POWER STATION CL-84-11 ICSCC SUSCEPTIBLE . TABLE 5. VELDS SCHEDULED FOR EXAMINATION DURING REFUELING OUTAGE NUMBER 7. INSPEC1 ION WELD SIZE SYSTEM IDENT-ISOMETRIC TECHNIQUE 28" RECIRC (3) 2R-NIA-1 ISI-I-2R-B UT 2R-N1B-8 ISI-I-2R-A UT 2R-N1B-9 ISI-I-iR-A UT 20" RHR (2) 10R-0-7 ISI-I-10-1A UT 10R-0-8 ISI-I-10-1A UT 18" BHR (3) 10R-IB-12 ISI-I-10-1 UT 10R-IB-14 1S1-1-10-1 UT 10R-IB-5 ISI-I-10-1 UT 12" RECIRC. (2) 2R-N2A-1 ISI-I-2R-A UT 2R-N2B-1 ISI-I-2R-A UT 12" CORE SPRAY (1) 14-A-1 ISI I-14-1 UT 10" CORE SPRAY (5) 14-B-17 ISI-I-14-1 UT 14-B-18 101-I-14-1 UT 14-B-19 ISI-I-14-1 UT 14-B-20 ISI-I-14-1 UT 14-B-3 ISI-I-14-1 UT 6" RUCU (4) 12-I-16 ISI-I-12-2 RADIOGRAPH 12-0-29 ISI-I-12-1 UT 12-0-30 151-1-12-1 UT 12-0-31 ISI-I-12-1 UT 4" RHR (5) 10-HS-10 ISI-I-10-5A UT 10-HS-11 ISI I 10-5A UT 10-HS-12 ISI-I-10-5A UT 10 HS-8 ISI-I-10-5A UT 10-HS-9 ISI-I-10-5A UT 4" RWCU (5) 12-I-30 ISI-I-12-2 UT 12-I-31 151-1-12-2 UT 12-1-34 1S1 1-12-2 UT 12-I-35 ISI-I-12-2 UT 12-I-36 ISI-I-12-2 UT 4" JET PUMP (2) RPV-N9A-1 ISI I-54 4 UT RPV N9B-1 ISI-I-54-4 UT TOTAL: 32 WELDS 9

INSPECTION SIATISTICS AND PLANS FOR PIIERIM NUCLEAR POWER STATION TABLE 6: GENERIC LETTER 84-11 IGSCC SUSCEPTIBLE PIPING WELDS (PAGE 1 0F 2) PREVIOUSLY INSPECTED IGSCC SUSC. WELDS IGSSS SUSC. WELDS NOT PREVIOUSLY INSPECTED TOTAL NUMBER NUMBER PLANNED % PLANNED FOR NUMBER NOT NO. PLANNED % PLANNED GL84-11 IGSCC INSPECTED FOR REINSPEC-REINSPECTION INSPECTED FOR INSPECTION FOR INSPECTION SIZE SYSTEM SUSCEPTIBLE RF0 6 TION RF07 RF07 RF06 RF07 RF07 28" RECIRC 14 14 3 21 0 20" RHR 3 3 2 67 0 18" RHR 14 14 3 21 0 12" RECIRC 10 10 2 20 0 CORE SPRAY 2 2 1 50 0 Avg. 25% 10" CORE SPRAY 21 21 5 24 0 INSPECTION STATISTICS AND PLANS FOR PILGRIM NUCLEAR POWER STATION TABLE 6: CENERIC LETTER 84-11 IGSCC SUSCEPTIBLE PIPING WELDS (PAGE 2 0F 2) PREVIOUSLY INSPECTED IGSCC SUSC. WELDS IGSSS SUSC. WELDS NOT PREVIOUSLY INSPECTED TOTAL NUMBER NUMBER PLANNED % PLANNED FOR NUMBER NOT NO. PLANNED % PLANNED ( GL84-11 IGSCC INSPECTED FOR REINSPEC-REINSPECTION INSPECTED FOR INSPECTION FOR INSPECTION SIZE SYSTEM SUSCEPTIBLE RF0 6 TION RF07 RF07 RF06 RF07 RF07 6" RHR(HS) 1 1 0 0 0 RWCU 15 14 3 21 1 1 100% Avg. 20% I, 4 4" RHR(HS) 22 22 5 23 I) dI) 0 RWCU 22 20 5 25 ) I 100% JP INSTR. 1 0 Avg. 23% 4 h 1 4" JP INSTR. I 1 1 100% (OVERLAY) Notes:

1. Not accessible due to pipe whip restraints.

2. Jet pump Instmment Nozzel N9B-1 was Fr during RFO#6. A UT inspection i

willbe conducted during RFO#7.

3. Jet pump Instrument Nozzel N9A-1 was PT and UT during RFO#6 and 7.

1, I l j 1 I 1

+-A.- Summary Cverview of Pilgrim Nuclear Power Station 6.0 Response to Generic Letter 84-11 Generic Lette.T 84-11; Item 1. A reinspection program of piping susceptible to IGSCC should be undertaken. The reinspection should commence within about two calendar years, adjusted from the previous inspection to coincide with the next scheduled outage, 82-03, 83-02, or our August 26, 1983 Order. performed under IE Bulletins BECo Response: A reinspection program covering ASME Code Class I piping weldments iden-l t tified as IGSCC susceptible in accordance with GL 84-11 has been imp emen - This complies with ed during the current refueling outage at PNPS (RF0 7). refueling outage following the GL 84-11; Item i since RF0 7 is the first h outage during which initial inspections were performed in response to t e August 26, 1983 Order. Generic Letter 84-11; Item 2: These reinspections should include the following stainless steel welds, susceptible to IGSCC, in piping equal to or greater than 4" in diameter, in systems operating over 200*F, that are part of or connected to the reactor coolant pressure boundary, out to the second isolation valve as follows. Inspection of 20% of the welds in each pipe size of (a) IGSCC sensitive welds not inspected previously (but no less than 4 welds) and reinspection of 20% of the welds in each pipe size inspected previously (but not lous This sample than 2 welds) and found not to be cracked. should be selected primarily from weld locations shown by experience to have the highest propensity for cracking. i 1 -~---"-----.m.

(b) All unrepaired cracked welds. Inspection of all weld overlays on welds where (c) circumferential cracks longer than 10% of circumference were measured., Disposition of any findings will be Criteria for reviewed on a case-by-case basis. operation beyond one cycle with overlaid joints are under development. Inspection of any weld treated by induction heating (d) stress improvement which has not been post treatment UT acceptance tested. In the event new cracks or significant growth of old (e) cracks are found, the inspection scope should be expanded in accordance with IEB 83-02. BECo Response: I piping welds were reviewed for susceptibility to PNPS ASME Code Class A compilation of all-welds IGSCC in accordance with GL 84-11 requirements. An updated set of identified as IGSCC susceptible is presented in Table 4. IS1 Isometrics for PNPS IGSCC susceptible welds is presented in Attac 7. 7. BECo plans to inspect all PNPS piping welds listed in Table 3 during is A summary of completed RF0 6 and planned RF0 7 inspection statictics For each pipe size and GL 84-13 inspection catcgory, presented in Table 6. 0 7 is equal to the percentage of welds planned for reinspection during RF With the exception of or greater than the specified 20% sampling size. (i.e., RWCU welds welds that are inaccessible due to pipe whip restraintsPNPS that were not 12-1-20 and 12-I-22), all IGSCC susceptible welds at,

Hence, inspected during RF0 6 are planned for inspection during RF0 7.

BECo's RF0 7 inspection plans for PNPF satisfy the sample size inspection criteria specified in GL 84-11 Item 2a. There are no unrepaired cracked welds or induction heating stress improved welds at PNPS. Hence, sections (b) and (d) of GL 84-11; Item 2 do not apply. In response to GL 84-11; Item 2c, BECo plans to inspect the overlay re-Inspection results for this paired Jet Pump Instrument Seal during RF0 7. overlay will be submitted to the HEC for review. The inspection scope will be expanded in accordance with IE3 83-02 in the event that new cracks are found or significant growth occurs in the overlay repaired crack in the Jet Pump Instrument Scal. Generic Letter 84-11; Item 3: should 5 emonstrate competence in 1 All level 2 and level 3 UT examiners field accordance with IEB 83-02 and level 1 examiners should demonstrate performance capability. BECo Response: Generic Letter 84-11 refers to IEB 83-02 for competence demonstrations of However, in September, 1985, all all le tel 2 and level 3 UT examiners. requalified under the new "NRC/EPRI/Bk'ROG examiners were required to be forth Coordination Plan", which references the competence requirements set BEco selected General in the EPRI NDE Center's training program manuals. All GE level 2 Electric Company (GE) to conduct the required examinations. and 3 examiners performing inservice inspection have either been requal-This initial qualification under the new plan. ified or obtained their includes examiners performing manual UT and those performing automatic UT

with the GE Smart UT System (which was also qualified along with the and the examiner). In addition, all icvel I examiners have procedure received appropriate training and have demonstrated their field performance Hence, all PNPS UT examiners have capability in accordance with GL 84-11. in accordance with requirements specified in IE Bulletin been qualified 83-02 and level I examiners have demonstrated field performance capability. Generic Letter 84-11; Item 4: Leak detection and leakage limits should be sufficiently restrictive to ensure timely investigation of unidentified leakage. BECo Response: On February 4, 1985 (Ref. 10), BEco submitted a proposed Technical Speci-fication AmerJment which addresses reactor coolant leak de If this proposed change is approved by the NRC, then a new leakage limits. Limiting Condition for Operation (LCO) for the rate of increase of reactor coolant leakage into the primary containment from unidentified sources wi The proposed Technical Specification Amendment establir.hes be established. increase averaged over ony 24-hour (2 gpm) limit two gallon per minute In addition, new operability requirements were proposed for the a period. Leakage Detection System and the Drywell Continuous Atmo-Reactor Coolant Radioactivity Monitoring System. Greater specificity for the sphere f the operational requirements was proposed to account for the redundancy o Foliowing NRC l systems and the redundancy of components within subsystems. l approval of the amendment, leak detection and leakage limits for Ph i should be sufficiently restrictive to ensure timely investigation of unidentified leakage. Generic Letter 84-11; Item 5: For crack evaluation and repair criteria see GL 84-11: Attachment 2. (Presented as Attachment 6 in this report.) '

BECo Response: In the event that new cracks are found or the overlay repaired crack in the Jet Pump Instrument Seal exhibits significant growth, GL 84-11 recommended p crack evaluation and repair criteria described in Attachment 6 will be j applied. In summary, BEco's plans and actions to address the IGSCC concern for Pilgrim Nuclear Power Station piping welds have met or exceeded NRC Generic Letter 84-11 requirements. 7.0

References:

1. NRC Docket No. 50-293, "IGSCC Inspection Order Confirming Shutdown", August 26, 1983. 2. NRC Generic Letter 84-11, " Inspections of BWR Stainless Steel Piping", April 19, 1984. 3. BECo letter to NRC, #83-292, December 8, 1983. 4. BECo letter to NRC, " Generic Letter 84-11: Inspections of BWR Stainless Steel Piping", #84-074, June 4, 1984. 5. NRC letter to BECo, " Meeting on June 15, 1984 with Boston Edison Company Regarding Cracks in Inconel Weld Material at the Pilgrim..uclear Power Station", Docket

50-293, July 13, 1984.

6. BECo letter to NRC, " Response to IGSCC Inspection Order Confirming Shutdown", #84-146, September 11, 1984. 7. BECo letter to NRC, " Supplemental Response to IGSCC Inspection Order Confirming Shutdown", #84-191, November 9, 1984. 8. BECo letter to NRC, " Pilgrim Jet Pump Instrumentation Nozzle Repair Program", #84-189, November 8, 1984. l t -

7.0 References (Continued) 9. NRC letter to BECo, "IGSCC Inspection per August 26, 1983 Order", Docket No. 50-293, December 4, 1984.

10. BECo letter to NRC, " Pilgrim Station Propesed Technical Specification Change", #85-023, February 4, 1985.

11. NRC letter to BECo, " Generic Letter 84-11, Inspection of Stainless Steel Piping", Docket No. 50-293, January 2, 1986. 12. BECo letter to NRC, " Generic Letter 84-11 Augmented Inspection Program", #86-145, September 22, 1986.

13. BECo letter to NRC, #87-003, January 2, 1987.

T: ATTACHENT 1 MATRIX OF IGSCC INSPECTION ORDER EXAMINATION From Ref. 6 BEco Letter to NRC, " Response to ICSCC Inspection 11, 1984, Order Confirming Shutdown," #84-146, Septemb'er h.

MATRIX OF IGSCC INSPECTION ORDER EXAMINATIONS Page 1 of 7 EXAMINATION CORRECTIVE ORDER SYSI_EH REQUIREMENTS BECO'S ACTION __,_ STRESS. RULE _ _ CA_R, BON. CON. TENT, RESULTS __ ACTION-REMARKS RECIRC 1001 UT exam-Examined Ination of 2-N2A-5 1.963 .046 Indications Replace the Reported. entire Recirc-all welds. (approx. 125) 2-N2C-3 1.188 .079 No Indications. ulation System with 316 NG 2-N20-5 1.808 .065 Indications material. Reported. i 2-N2E-4 1.215 .053 No Indications. i 1 2-N2K-5 1.907 .046 No Indications 2-NIA-3 1.163 .070 Indications i Reported. 2-NIA-9 1.153 .070 Indications Reported. 1 RHR 1001 UT of Examined (Suction the ASME Code 10-0-1 1.849 .050 No Indications. Replace the RHR System and Ols-Class I charge) Portion of the 10-0-2 1.213 .050 No Indications, within contaln-ment along with system. i 10-0-3 1.228 .050 No Indications Recirculation System with 10-IA-2 2.178 .050 No Indications. 316 NG material. ] 10-IA-14 Not Available. Not Avallable. No Indications. Outside Containment 10-IA-15 Not Available. Not Available. No Indications. Outside Containment 10-18-14 Not Available. Not Available. No Indications. Outside l Containment i 10-IB-14A Not Available. Not Available. Indications Hill be Outside Reported. replaced Containment i with 316 NG material. I

MATRIX OF IGSCC INSPECTION ORDER EXAMINATIONS Page 2 cf 7 EXAMINATION CORRECTIVE ORDER SYSTEM REQUIREMENTS BECO'S ACTION STRESS RULE CARBON CONIENI RESULTS ACTION-REMARKS 10-IB-15 Not Avallable. Not Available. Indications Evaluated Reported. as counter-bore and con-firmed by construction x-ray. 10-1B-16 Not Available. Not Available. No Indications. No Indications. Inside and RHR (Head 1001 Ultra-Ultrasonically outside con-Spray sonic Exam-examined twenty-tainment. Ination of two of the the ASME Code twenty-three { Class 1 Portion welds. of the system. Exemption f l Held 10-HS-23* from Order is scheduled requested. 3I for radiography Scheduled k (trl-metallic for RI. Held). Exemption Core Spray 1001 Ultra-

1) The following from Order sonic Exam-welds will be -

requested. ination of the radiographed: Scheduled ASME Code 14-A-l.14-A-3, for RI. Class 1 14-A-10A. 14-B-1, Portion of the 14-B-3,14-B-10A s system. Excluded

2) 14-A-2 and from Order 14-B-2 not because examined as it welds are is 316L material.

not 304SS. j l { )

s MATRIX OF IGSCC INSPECTION ORDER EXAMINATIONS Page 3 of 7 EXAMINAil0N CORRECTIVE ORDER SYSTEM REQUIREMENTS BECO'S ACTION STRESS RULE CARBON CONTENT RESUliS ACil0N-REMARKS

3) All 30455 Excluded i

from 14-A-3 and from Order 14-B-3 to the because manual isolation welds are valves was replaced not 30455. I with carbon steel during the 1977 refueling outage.

4) 304 pipe material from the manual Isolation val-ves to'the dry-well penetration is being replaced i

with 316 NG. This eliminates the following i' welds: A-Loop B-Loop 14-A-Il 14-B-11 14-A-IIA 14-B,12 14-A-12 14-B-13 l 14-A-13 14-B-14 14-A-14 '14-B-15 14-A-15 15-B-16 14-A-16 i

MATRIX OF IGSCC INSPECTION ORDER,EXAMINAT.10NS Page 4 of 7 EXAMINATION CORRECTIVE i ORDER SY, STEM REQUIREMENTS BECO'S ACTION STRESS RULE CARBON CONTENT RESULTS ACTION-REMARKS __

5) The following welds outside containment to the Class I boundary were examined:

A-L99P 14-A-17 Not Available. Not Available. No Indications. 14-A-18 Not Available. Not Available. Indications Evaluated as Reported. counterbore and 14-A-19 Not Available. Not Available. No Indications. confirmed by construction l X-ray. B-Logg' 14-B-17 Not Available. Not Available. No Indications. i 14-B-18 Not Available. Not Available. No Indications. 14-B-19 Not Available. Not Available. No Indications. 14-B-20 Not Available. Not Avallable. No Indications. 3: 14-B-21 Not Available. Not Available. Indications Within 25% of Reported. wall thickness, a therefore, evaluation in accordance with l NRC criterla shall be r performed. RNCU 1001 Ultra-

1) RNCU Suction sonic Exam-pipe from the Ination of the RHR supply line ASME Code to the drywell Class i System.

penetration will be replaced l with 316NG material. =e w

MATRIX OF IGSCC INSPECTION ORDER EXAMINATIONS Page 5 cf 7 EXAMINATION CORRECTIVE SYSTEM REQUIREMENTS tECO'S ACTION STRESS RULE CARBON CONTENT RESULTS ACTION

REMARKS, ORDER

+ Welds 12-0-24 Not Available. Not Available. No Indications, through 12-0-31 (Outside contain-ment.) RWCU Discharge Not Available. Not Available. No Indications. pipe from the transition weld 12-1-17 through i weld 12-1-36 were l examined with the following excep-tions: a) 12-I-20,12-I-22 i pipe whip restraints. t l P;netrations 1001 Ultra-BECo requested J RNR sonic exam-exemption No 2> ation. X-SIA Not Available. Not Available. examination conducted; J, penetration welds are inaccessible for ultra-sonic exam-Ination. ,~ j

MATRIX OF IGSCC INSPECTION ORDER EXAMINAi!ONS Page 6 of 7 EXAMINATION CORRECilVE ORDER SYSTEM REQulREMENTS BECO'S ACTION STRESS RULE CARBON _ CON T EN T._, _,,,RE.50t TS _, ACTION

REMA_RKS, X-51B Not Available.

Not Available. ID of dye was To be dye penetrant corrosion inspected. resistant N0 IGSCC. clad when piping subsequently associated with welds Nos. 10-IB-14A and 10lb-15 is replaced. Not Available. Not Available. No Core X-16A Spray conducted; penetration welds are inaccessible for ultra- (( sonic exam-Ination. X-16B . Not Available. Not Available. No examination conducted; penetration welds are inaccessible for ultra-sonic enam-Ination. e

a MATRIX Of IGSCC INSPECTION ORDER _EXAMINAllONS Pag? 7 cf 7 ~ EXAMINAllON CORREC11VE ORDER SYSitM REQUIREMENTS BECO'S ACTION STRESS RULE CARBON CONTENT RESULTS ACTION REMARKS X-14 Not Available. Not Available. No RNCU examination conducted; penetration welds are inaccessible for ultra-sonlC exam-Inatlon. L

e wesa e ese

ATTACHMENT 2 PNPS ISI IS0 METRICS FOR-IGSCC SUSCEPTIBLE WELDS From Ref. 6 BECo Letter to NRC, " Response to IGSCC Inspection 11, 1984; and Ref. 7, Order Confirming Shutdown," #84-146, Septemb,erBECO L Confirming Shutdown," #84-191, November 9, 1984. l I 4 I i - - -, ~, - -,. - _.

1 SYMBOLS j

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Nec REQUIRES OT EXAMINATION REPLACE k/ITH 316NG MATERI AL g NOT EXAMINED DUE TO OBSTR0tfloN,14AT'L ETC. arnasonic TestCometeTeo X g WELD SCHEDOLED FOR RADIOGRAPHY (TRI-METALLIC) INDICATIONS REPORTED h NOTES 1. Section will be removed rather than replaced. ~ 2. ID of weld was dye penetrant inspected. No IGSCC. To be corrosion resistant clad when piping subsequently associated with welds Nos.10-IB-14A and 10lb-15 is replaced. 3. Indications reported. 4. Evaluated as counterbore artti confinned by construction x-ray. 5. Within 25% of wall thickness, therefore, evaluation in accordance with NRC criteria shall be perfonned. = a _ s @ = - - i ~ Ei h Ts"a~o'*'a~ ~~ FOR INFORMATION (,6 ryg<M -= u _ _. r_ 2. x :.x;.- g t 1s1-I o o ~- M.%" ;';'f =_J = 2 .- L m = 3_: l = = -

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1 ATTACHMENT 3 WELDS NOT SCHEDULED FOR EXAMINATION i From Ref. 6 BEco Letter to NRC, " Response to ICSCC Inspection Order Confirming Shutdown," #84-146, September 11, 1984. e t 4 1 I

WELDS NOT SCHEDULE 0 FOR EXAMINATION SYSTEM WELO IDENTIFICATION

DIAMETER DRAWING NO.

EXEMPT CODE' Head Spray 10-HS-23 4" ISI-I-10-5A 4 Core Spray 14-A-1* 10" IS!-I-14-1 2 (loop A) 14-A-3* 10" ISI-I-14-1 2 14-A-10A* 10" 151-!-14-1 2 Penetration Weld (X-16A) 10" 15!-!-14-1 1,3 Core Spray 14-8-l* 10" ISI-I-14-1 2 (Loop B) 14-B-3* 10" ISI-I-14-1 2 14-8-10A* 10" 151-1-14-1 2 Penetration Weld (X-168) 10" ISI-I-14-1 1,3 RWCU 12-I-20 4" ISI-I-12-2 3 12-I-22 4" 151-!-12-2 3 12-I-16 6" ISI-I-12-2 4 NOTES: A. Carbon steel welds shown on drawings are not listed. B. Exemption Codes are: 1. Normal system operating temperature less than 200*F. 2. Weld is in 316 stainless steel piping material having low carbon content (below 0.02%) or is a dissimilar metal weld consisting of low carbon 316 stainless steel and carbon steel base materials, clad with Inconel and welded out with Inconel weld metal. 3. Weld is"not accessible for ultrasonic examination because of component configuration. 4. Tri-met &1lle weld. Weld 10-HS-23 will require radiography due to its configuration. Welds 14-A-1*, 14-B-1*, 14-A-3*, 14-B-3*, 14-A-10A*, and 14-8-10A' In the ASME Code Class I portion of the Core Spr,ay System external to the Reactor Vessel will be radiographed in lieu of UT examination because these welds are trl-metallic. Welds 14-A-2 and 14-B-2 were fabricated with 316L base metal and 308L weld metal and consequently were excluded from the Order. A3-1

ATTACllMENT 4 IlYDROCEN WATEA CilEMISTRY PROGRAM FOR PILCRIM NUCLEAR POWER STATION From Ref. 6 BEco Letter to NRC, " Response to ICSCC Inspection Order Confirming Shutdown," #84-146, September 11, 1984.

HYDROCEN/ WATER CHEMIS?RY PROGRAM FOR l P!LCRIM NUCLEAR POWER STATION i* The Hydrogen / Water Chemistry Program for Pilgrim Nuclear Power Station will i i contain three major objectives. The first objective alll be to develop and i permanently implement Hydrogen Chemistry Control for the reactor Coolant sys tem. The purpose of this objective is to suppress !GSCC in the reactor i coolant system piping by reducing the electrochemical driving force for 4 IGSCC. A comercially feasible hydrogen chemistry control program / system does not esist at this time. As such, commencing with startup from the current refueling outage, feasibility studies will be conducted at Pilgrim Station to develop tne necessary design criteria to develop such a system. Injection and Instrument taas will be installed and other modifications made to the reactor I coolant system to facilitate testing. Various methodologies will be evaluated and a final conceptual design is targeted for implementation during RF0 C. The implementation of Hydrogen chemistry control at Pilgrim Station will significantly impact station operations. Radiation levels are espected to rise which in turn will cregte snielding problems in several areas of the plant. Estensive analyses and evaluations will have to be conducted to ) address this impact. From a licensing perspective relief from the Technical Spectftcations will be needed for the Main Steam Line Radiation 1.imits. The second objective of the Hydrogen / Water Chemistry 3rcgram Drovides interim measures, during the ucccming fuel cycle, to minimite tne potentlei for IGSCC l propa ;4 tion. This ocjective creates a general water :'1em!stry control progrsm for tne reactor coolant. It is our intention to make every effort to comoiy i with the Electric Power Research Institute guideline 1 for SWR water chemistry. Achieving this second objective will also require several major I changes to system operations. Lessons learned during this Interim period will be Incorporated in the ongoing program to modify the reactor coolant environment to mitigate IGSCC. f The final ojective is to effect a reorganization sucn that water chemistry i j control at Pilgrim Station is given sufficient attention and importance to i ma6e it an integral part of station operations. Dedicated manpower and resources 4l11 De allocated to facilitate a permanent and comprehensive l Hydrogen < Water Chemistry Program. 4 l I i j A4-1 i

1 t l l ATTACHMENT 5 f PRELIMINARY LIST OF PNPS WELDS SUSCEPTIBLE TO IGSCC AND SCHEDULED FOR EXAMINATION DURING REFUEL OUTAGE NUMBER 7 l I i F l l i From Ref. 12 BEco Letter to NRC, " Generic Letter 84-11 f Augmented Inspection Progras," #86-145, September 22, 1986. l l t I i l l

PRELIMINARY LIST OF WELDS _ SUSCEPTIBLE TO IGSCC AND SCHEDULED FOR EXAMINATION 00RlNG REFUEL OUTAGE.NUM8ER 7 SIZE SYSTEM WELD # ISOMETRIC # 28" Rectre (3) 2R-NIA-1 ISI-I-2R-8 2R-N18-8 ISI-I-2R-A 2R-N18-9 ISI-I-2R-A 12" Rectre (2) 2R-N2A-1 ISI-I-2R-A 2R-N28-1 ISI-I-2R-A 12" Core Spray (1) 14-A-1 ISI-I-14-1 20" RHR (2) 10R-0-7 ISI-!-10-IA 10R-0-8 ISI-I-10-IA i 18" RHR (3) 10R-18-5 151-1-10-1 10R-18-12 151-1-10-1 10R-18-14 151-1-10-1 10" Core Spray (5) 14-8-17 15!-I-14-1 14-8-18 15!-I-14-1 14-8-19 151-1-14-1 14-8-20 151-1-14-1 l l 14-8-3 151-1-14-1 l 6" RNCU (3) 12-0-29 151-1-12-1 12s0-30 151-1-12-1 12-0-31 151-1-12-1 4" RNCU (5) 12-I-34 151-1-12-2 12-I-35 !$1-1-12-2 n 12-I-38 151-I-12-2 12-I-30 151-1-12-2 12-I-31 151-I-12-2 4" RNR (5) 10-NS-8 !$1-1-10-5A 10-MS-9 151-I-10-5A { 10-H5-10 151-1-10-5A 10-MS-11 ISI-I-10-5A 10-MS-12 151-1-10-5A t 4" Jet Pump Instru. .RPV-N9A-1 ISI-t-54-4 (weld overlay) TOTAL: 30 welds A$ 1

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? /-' A6[I @ Q-0-32 Nil p -tG 7 gj g EL 4i-4Will T [D E.,',; ' * - EL.64' w F- -33 3 , 3yg 8 -25 4',,, /. *S.k?- ?. 2*D&4 d 7' 6 REACTOR VESSEL ( d -t3 f) <3> { l DRAIN LINE 4 B FL 43' w 7 , t 's b 4 s5 ' - ' ..i t qy f _ _4.fd}9 -16 9 9 [ptPE DATA 5 k3*.c 6'DC A-12 ', / De -20 < 2k " q 5' ID Ee41084 OCAt21 OlEA 62 h - -S l 4JEREC- -10{g )*W DB '13 NPS 2 2 61 6 6 .-~ .i4 ~* AOy g a *I I - -4 a NwT .344 219 VIN.378 432

4 %"

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DISSMLAR METAL WELD

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] l 8 l 7 l 6 l 5 4 4 l 3 l 2 l 1 Im.i =-~ 1 l [ALLOySTEEL ALLOY STEEL b. TAiNLESS STEEL S l CARBON STEEL D D NOZZLE l El )ISng>SA) HEAD SPRAY TO JET PUMP NOZZLE l lI N98 NSTRUMENTATION g N7A N7A-1 T _N99-i STA <EL L [ALLOYSTEEL i CARBON STEEL STAINLESS STEEL i C TO VESSEL LEVEL C NO7ZLE , I N NSTRUMENTATOJ f N78 f N78-1 N6A 2 N6Al INACCESSIBLE WELD l + Y' + E l l INCOPEL [ALLOYSTEEL CARBON STEEL STAN_ESS STEEL l K CR E j TO VESSEL LEVEL 8 NOZZLE j I R INSTRUMENTATION NO N81 N168 3 O N68 2 NI681 ALLO APPROVEDjf h i g REVIEwE Dtre_/?' i / SS STEEL TO JET PUMP CHECKED +4* NOZZLE ~~ A N9A INSTRUMENTATION

;r-mz MIE

~ ~ ~' A 1 WELD OVERLAY I

===aa a w, 1 VESSEL NOZZLES t ( 6 8-11S11s4-4 o same- _ t

I ATTACHMENT 6 CRACK EVALUATION AND REPAIR CRITERIA i - i' i i From Ref. 2,'NRC Generic Letter 84-11 " Inspections of BWR Stainless Steel Piping: Attachment 2," April 19, 1984. l t 1 f + i f i 1 4 5 m.. . _... _. _.,,. _.. _. _. - _ _, _, ~

CRACK EVAltlATION AND REPATR CRITERIA

===1. Background=== (a) Code Reouirements The ASME Boiler and Pressure Vessel Code Section XI has rules for evaluating the acceptability of flaws for further operation. Table IWB 3514-3 provides rules for acceptability of flaws without further evaluation; although the specific dimensions of such acceptable flaws depends on both the length and depth of the flaws, the practical effect is that flaws less than about 10% of the wall thickness are acceptable for further operation without analysis or repair. A new section has recently been added to the Code, IWB 3600. This extends the Code flaw evaluation rules for piping to include specific rules whereby flaws deeper than those allowed by IWB 3514-3 can be accepted for further operation without repair. Section IWB 3600 also requires that these acceptable flaw sizes include considerations of crack growth by stress corrosion and fatigue. In other words, if a crack is to be considered acceptable for further operation without repair, it must be shown that it will not grow to be larger than the IWB 3640 limits during the time period for which the evaluation is parformed. AG-1 ~, m y r

= (b) Crack Growth Assessment IGSCC at welds ir BWRs is primarily initiated by very high tensile residual welding stresses on the inside surface at the heat affected (sensitized) zones of the base metal very near the welds. This tensile residual stress changes to a compressive stress toward the middle of the pipe wall; this reduction in stress reduces the crack growth rate through the center portion of its pipe wall. As the crack progresses further through the wall, the relative effect of the pressure and bending stresses increases, and the crack growth rate will increase. The residual stress patterns and calculational methods for crack growth rates are fairly well established by considerable research and correlations with service experience. The staff has selected parameters that should lead to overprediction of growth. This is intended to comoensate for uncertainties i discussed in more detail below. (c) Staff Treatment of Uncertainties One of the main uncertainties associated with the evaluation of pipe cracks is the uncertainty of crack sizing, both depth and length of IGSCC cracks. Although this technology is beino improved, the uncertaintiv in crack sizira will likely remain. A6-2

e The staff has used a relatively simple approach to cover sizing uncertainties. In practice, the staff approach permits operation of unrepaired cracks but only if calculations show that they would not exceed Code limits even if the crack at the start of operation were actually twice as large as reported. 2. Staff Acceptance Criteria (a) Criterion for Operation without Repair Plant operation is permitted with cracked welds only for the time period that the cracks are evaluated to not exceed 2/3* of the limits for depth and length provided in ASME Code Section XI, Paragraph IWB-3640. Crack growth analyses must include any additional stress imposed on the weld by other weld repair operations, and each analysis must be approved by the NRC. (b) Criteria for Cracked Regnirs i fi) If cracked welds are repaired by weld overlay, the-thickness of the overlay must be sufficient to provide full IW8-3640 margin during the proposed operating period, assuming that the cracks are or will grow completely through the original pipe wall and the first overlay layer to the low carbon and low ferrite portion of the overlay, unless it is demonstrated that the crack (s) are shallow enough to be arrested by the weld overlay.

This ciiterion allows for an uncertainity of up to 100% in crack depth sizing for raported cracks up to 25% of wall thickness.

AS-3

4 Effective overlay thickness is defined as the thickness of overlay deposited after the first weld layer that clears dye-penetrant testing (PT) inspection. (ii) The minimum effective overlay thickness permitted, even for very short cracks in either longitudinal or circumferential direction, is two weld layers after the first layer to clear PT inspection. (iii) Full structural strength weld overlays must be provided for lono cracks with total circumferential extent approaching the length that would cause limit load failure if they were actually through-wall. (iv) Multiple short circumferential cracks are to be treated as one crack with a length equal to the sum of-the circumferential lengths. 3. Discussion of Staff Acceptance Criteria Since the period of operation between inspections could vary from plant to plant and the applied stress level varies fro $ location to location, use of a fixed simplified repair criterion established on the bases of crack size prior to the period of operation would be difficult. In any case, however, flaws less than about 10% of the wall thickness are acceptable for further operation without repairs. For a typical 18 month operating cycle, the staff criteria vould cenerally require that AS-4

' d 5-cracks greater than 30% of the circumference and cracks with reported depth of 25% or greater of the thickness will likely need some fem of repair. For the same 18 month cycle, cracks nf smaller size down to 10% of wall thickness may be acceptable without repair but would require evaluation in accordance with the staff criteria. il

e 3

J AS-5

6 ATTACHMENT 7 UPDATED PNPS ISI IS0 METRICS FOR FOR IGSCC SUSCEPTIBLE WELDS Piping material is identified on isometrics as follows: Note: DC Type 304 SS -Type 316 NGSS - DCA 3 s -n n-- -m ee.m, ,n..~,-----.-,v,,e-a-.m.+ ..--.,.-----.-r -r

l 3 y 2 1 4 IN(EX DWG NO TITLE (ISI I- ) s c. O NOTATION 54-3 VESSEL BOITOd HEAD H M AIN STE AM & DRAIN M ANIFOLD 54-4 VESSEL NOZ'ZLE$h ~~ 2R-A REACTOR RECIRCUL ATION LOOP 'A' q.g p, 4 {,'. 2 R-B REACTOR RECIRCULATION LOOP'B' Kp 3-1 REACTOR CONTROL ROD drive 44 {Q # 6-1 FEE 0 WATER LOOPS ALB 10-1 RHR RETURN LINES 44. y 47 g' [t IOiA RHR SUPPLY s C C IO-IB RHR PUMP SUCTION CROSS CONNECTION "x N IO-IC FUEL PCOL COOLING TO RHR SUCTION CROSS CONNECT 10-2 A RHR PUMPS P203 A,C SUCTION t'T' 10-2 8 RHR PUMPS P2038.D SUCTION A ki 3 IO-3A RHR PUMPS P203 A,C DISCHARGE t L{'g 10-3B RHR PUMPS P2038.0 DISCHARGF.,, lO-4A RHR DISCHARGE TO DRYWELL i-4 x y ? 10-4 8 RHR CROSS CONNECT Dt r 4, 1 %4 r-1 10-5 A RHR SPRAY HEADER 4 ~ 'd il 10-58 RHR& CORE SPRAY DLS Il-l STANDBY LIQUID DONT 8 i ~+ 12-1 RWCU SUCTION A RPV AMll ht 's 3 I h D @, ', l 12-2 RWCU Hx E-2CB4DISCH ARdE 7 o%Q RCIC SYSTEM Y3,_ c9!$ Mi P ' l l 4. RCIC PUMPdbCTIONEY[j B 13-1 13 -2 13-3 RCIC. TURBINE EXHAUST

iy M%

i i ' 8 CORE -SPRAY PUMP P215A,8 DISCHARGE K(f Tk 14 -1

CORE { SPRAY PUMP P2tSA Q

l,j 14-2 A I4-2B . CORE-SPRAY PUMP P2158

Eb. @

23-1 HPCI TORBINE STEAM SUPPLY 6 Put# DISCHARGE B 23 2 s HPCI TURBINE STEAM SUPPLY 23-3 HPCI TURBINE EXHAUST STEAM 23-4 HPCI PUMP SUCTION 23-5 HPCI PUMP TO RE ACTOR i-ic 29-1 SERVICE WATER y_m e3"/ 30-1 RBCCW PUMPS DISCHARGE 30-2 RBCCW PUMPS SUCTION sosioN EDtSON CORAPANY A A 50-1 CTMT ATMOSPHERE CONTROL INDEX 54-1 VESSEL WELDS & NOZZLES 54-2 VESSEL HEAD WELDS & NOZZLES g g ISI-O I a= o =-

l 3 / 2 1 4 SYMBOLS NOTATION IDENTIFICATION ID #5 CONSIST OF; D o VALVES g VALVE # GATE / GLOBE VALVES <C SUPPORT # BUTTERFLY I I SPOOL # -- - I ~ - -[= ( ) PENETRATION # gp qQ CHECK (FLOW -) - - F _ _ _. FIELD 4dILD .UfPORTS A (SINGLE LETTEP) gCg{ELDYb kb. H-33 SUPPORTS b"( digit) ,p A 'Ts $p9, l E RESTRAINT (SR) h CLASS BOdkDARY~ki, $$hI g Fo-E SNUBBER (55) A INF,OR ATI ONLY fJ.S. F++-E S//AY BPACE (liE) A ChiNUATION ON i I GUIDE (SG) t$y% INDCATED.DFAWING I > SPRING HANGER (SH) i T0!OR 'FROM' DENOT ES (LINE#)(SPOOL #)(WE LD #) D W TTJTERVENING LINE 4 f(gy 'I /4 NOT SHOWN)._ d FIELD ANCHOP (SA) L b f5 \\Y - --(LINE #)(WEL D*) l - I- - i9 l l RIGID HANGER (RH) SUPPORT LUG (HL)... 'Q.jfQ $ r @Q.) l ' g b;.j g STlFFENER COLL 4R (Sc) Ai T' %g,.gEXAMPLE l l ( ) ' I rp -p v w e+ tpp:p, 'u-8 y PIPSDATA"33 NT_3 '~ 9 8, 18 HL-lO F2_ o-i i o p~ l ~' . /; C# OTHER i l1 NOZZLE'. IR Pl&lITNTIFICATION ' I g.L MAgigt CS. CARBON STEEL , IN. '8 1 4 gg STAINLESS FLANGE-(CONNECTION NE@AL P:PE SIZE n REDUCER *.- NWT'" " ~ WALL T HICKNESS H lO-t-lO H-lO-l-12S R 10 01-9 H L 8 B HL-lO F26 WELD' MWT MINIMUM (EQUALS NC NCMINAL CIRCUMFERENCE CONST ISO # ) WELDOLET W10-26) D""'"^ ""~"~' - ~ .._ _ ~__ ~ {} SOCKET WELD O. [ EXPANSION BELLOWS A A INTEGRAL WELD A JSI 150 NOTATION f .mo.o o X mow ~' w 'M-1SI-I- O l no wm a sin C L"ss "

( l 3 y 2 1 4 N2E N2D / W2C /j N2B N2A / j 2R N2C-L 2R N2 9-l. 2R A2 A-is 2R+42 E-I 2RN20-1 _+ + 'N 't 4 D D - 2. - 2' -2 -2 /4e 2R-NIB-l 12'DCA-2 ~ / C# 12'DC A-2 12 DCA-2 .,:p / / s 'DCA-2 / -2 12 V-28 hg,92(2 DCA.2 - Q - 3-_ 3_ 22,DCA-2 s g4 - HA-4 T .s g f .h; k, 2R-Hk l /, .A &h (e nh f II-LOK r%nyt1A-l i -NIB I', 4_ 4 C C <l@l A i h. Sh g Q 4HL-b4) ' _$_ _7 $ ' -5 HL-l (4) 5l4 k + 4, M MT> O dps egt..$G - NC -14R-l(3) r7 y Tep ONN __ i. M i b ~ 'I4 CC 2 R 6PA .(-, br+ PIPE DATA j - t'3 ,e _ R- ?. e g. v't?:. h 'hitq%,Td. a re . y 202 lD DC A-2 l l s g y 5A M AT L 55 l ! @$% f;f* 7 t \\ NPS 2 12 22 28 l ' ~.[D. f MWT .218.% 3 f.S O NOTEI y s 'O E,. 3 f > -6HL f(3) Whl, S.4 NC 6.3 l 41.2 78.5 94.8 LI - 6 RANCH,, _ ~ 9 ONN,T 8 28' DCA-2 2R-N!B96C-1 f DISSMLAR PETAL,pELD A N vEC [/M4f //.3 BRANCH) 2 R-NIB-9 BC-2 NOTE l-ALL 28 DCA IS MWT 1.084 RiviEwf 0 kg(( tg.g 202 -. -2A CONN 4A EXCEPT WHERE NOTED ON DWG MIE-Al SH 1. (%E(nto gt c,3/e. ,,j,,,, @l l - w w mauw s4 r. n i~.. ic 2'DCA -3 " 2 ?" s< t. s, _ e. 2kI U" OR sb 63A A h %'ifSsW P REACTOR RECIRC. A ms,o LOOP 'A' SIET14 nio ce. sx na m TO CRW ISI-I-2R-A 0 ,o =:.-

l 3 y 2 1 4 / ICK N2J NH 2 R-N2 K-1 2R-N2J-Is 2R-N2H-I 2R-N 2 G-1 2R N2F-t 's t + y g D - 2s - 2' s -2 -2 -2 s s 2R-NIA-I NtA 2 .h 28"DC A 2 3 , 3_ DCA-2 ,3,_,' h -2' 12 0 2 2 I2 DCA2 - --3 , 2 2~ D C A. -H b4 -Hd-3 -H pg.g - 3 a IO R-I84 '^ C C ~ g s -13 - k,; q., , 4 "-Q h Ii2HL-2(4 hC d^ t $ \\ g4% kFW - 28 DCA 2 f. + g g hp i 1 - 4HL-l(4) TOR e (- g. .,[#"}j hd-- :D D C A-2 >p QHL-l (3) PIPE D ATA Cp g .? ^I-fl4 6 _d. . _ /i -:y MATL SS c 3$ NPS 2 12 22 28 l 69 6 MwT .2iE 563 a.50 NOTEI I s NC 6.3 41.2 /8.5 94.8 l ,ye k,' g I ' -4 H L-2(3) L 'i% .? 9 i 8

h.

' BRANCH Ce d g - ., -l ;}$ / CCNN __.k kI 5 8'DC A-2 2R NlA 7BC-1 B 4 DISS.M.LAR METAL WELD -- 2R-NI A-78C-2 (BR ANCH} -6 -7 CONN NOTE I ALL 28 DCA IS MPT=1.084 EXCEPT ppg WHERE NCTED ON DWG MIE-Al SHl. uvoto n,y _, g,,,, 48 mg g oe tc, y,w .v. s., 2DC A-2,_ _,3 z m., m u o,o w s<- i-.- ic Nm h t$fEh asm s A ^ OdYE E T 796E9tO w&.8 iMiaVdiro cm st n:47 ~ TO CRW 0 ISI-I-2R-8 0 ,,,,, o met

l 3 2 1 4 DC k DCA 15 -16

B" RHR 1

VALVE RM D FROM RHR h DCA D PUMP CISCH 14 vc x XDI RECLRC DC y ,'2 LOOP'B' DRYWELL DCg _ _'I2 RECRC o /A LOOP A' A,. 's - - A OR 18-1 9 DCA 4 [9 c ~ A 4 DCA O Of M lOR-I A-l k vs b, oo -] l - k, d' W g' ' DC ay$to , 'f-2 -i2 p _..o '@l slA 'h C C -9 -9PS g .g a ~0 '6tf 2(4) 6 -ehLK41 IO-1A -14 4 IC IA-15 -4 g {ll.j g:tNN Q9A/ p g T .g i O #. 7, 7[ % Q \\j -ag; f -7 if ?;j$ TA

FROM RHR

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M SYS T E M -241 ED 30$T000 EDISON COtsPANY T5 CCNST g.JiTION isa 2i L"stMRdPJ^" ui js RHR LOOP A(8 a-i o 151-1-10-1 8 = r..-

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i. 4

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e 8 I 7 ) 6 ) 5 7 4 l 3 j 2 i I LLg7 4 mf;'? -w-22g DCA DR}WELL

r. - N

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l I e p C + g g 7 m.~ %S n i O e, h _ l c w_ f,$ t yx.Y = a 8-2 d u2 S 0 1 2 T g N t. / G I AL n, i r

q.. _ s T I

A t .hn T D R e t E g s f T s 0 CA S V u y I P t ,ds E E R ~ 1 H R A f N 2R s M 5 4 KwnS uEC $A A 4 5 1 y o S A L rHD 9 M: I; 5 l"- A E aX E r 2 E L S d l -f L A 4 W3 I 5 T - s 6 P f A A 4 D 1C i. ,F.v 37 f 4 0 6 TR s. i i, U f l,. A 0 3 D AE F L 2 (C P T s f = E. _ 7 :- 1 fN Pt D A P C . - ~ 1 \\ g I V N N 2> U CH S ,G g ~ __. _1 C L AX E p=Dp.- I GE u.2 3 7 l 3 r 2 ~ '6 O ~ +7! T )H> 2 Tfj !sd 2 uZ ) 8 3 a(t-q 2 43 7 v ( s- - H3 =g fm$ 9 l p N<A LT t e 3 H3 2G4,j ( gw a 2 t[Y 3 6i y- .y ) ,3, p 3 T 5 4 y ip a-p#7 sD,A2. %wd 2 M aRg j >a;, ECp A - g2 l s % gf'r;.n-3 1 4 M>',g')< l A ag\\ Jm3a+- .f. F I ,D J 4 j M , %c. ' 0 s - i 4 0 3 ,e > M % g R 2 0 4 4'a RE 0 A [ 2 HV l @/*9 4 1 3 s L 2 2 3 D A 2 AV / ) '2 9 4 t. ?'1 L ~ g s L k .6 3' 8 1 . + ' I' 2" ,t- ,.4 7 5 8 i; % - s 's[ t i i 9e 0 C / 0 4 l WE + d HV R L - A %3 s .A V L i - g EN - P4 1 Q 6 2 L ME 2 AN 8 a x? TU ST l' A,7 I l / g 6 o d0 6 7 9 .9ji L o D T '6 g .X 7_ b 6 1 6 5 l 3 2 ] Y '.1 1 3 1 2 v.- 1 w 3 u eH ]g oNic 5 c4 8 R o - F h i. U D C g A pusQ i lll !jjl{l l'

i e a 8 l 7 l 6 l 5 4 4 l 3 l 2 l 1 '= ~ -7 - ~-8 - EL 74 -6 ft Q REACTOR 9 Ou N, g o D P-268 7 44 - A-l* 606A l4 -2 A 3 O -4 A F3 f* .g f($.,D d l p,', 4R -B -2' 9[g -9A, C 5 WP-N N gw N3 % -6 Ai4R B-2A gp g *%% 6 ^M. a C' ' / 9 4 [ o'OcA44 , - 2* 42 % ?. *(~ t

g CBA i A4-B-20 g.g.

g, O .$' k jW DGSML12a METAL WILD < 41> DC ~ '89 . INSOE 'e l BLOCKWAL 9, ,5ji ' .p -4%~. C M f s -10;- "B -18 NO Uu w 3N s 21 C'OC 14 \\ ( REACTOR C .~ -17 T A'< @ B-II - w a -3 DRYwELL <k, ocf- - A* ' ~ F \\; z A K@f 44A dt6

T.

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ML20213F263

Contents

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SUMMARY

SUMMARY

2.0 INTRODUCTION

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SUMMARY

References:

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ML20213F263

Text

SUMMARY

SUMMARY

2.0 INTRODUCTION

SUMMARY

SUMMARY

References:

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=

9 m.=me*===

= = = =

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