ML20207J151

Rev 0 to Response to Generic Ltr 88-01 & NUREG-0313,Rev 2

ML20207J151
Person / Time
Site:	Oyster Creek
Issue date:	08/10/1988
From:	Capodanno G, Covill D, Lorenzo R GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20207J092	List: ML20207J087 ML20207J151
References
RTR-NUREG-0313, RTR-NUREG-313 GL-88-01, GL-88-1, TR-050, TR-050-R00, TR-50, TR-50-R, NUDOCS 8808300263
Download: ML20207J151 (55)

Text

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6 a GPUN RESPONSE

.TO GENERIC LETTER 88-01 AND NUREG 0313, REV. 2 TR - 050 REV. O AUTHORS S $$

D. COVILL - E & D/ME 6 P/*'/re _

R.' LORENZO - MGR. %CEP APPROVALS 8 2 El G. R. CAPODANNO DIRECTOR, ENGINEERING & DESIGN l

i L, 2-4.re

_ . K. dRONEBERG - DIRECTOR ,

ENGINEERING PR ECTS '

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- $N '

R. W. KEATEN - DIRECTOR I QUALITY ASSURANCE G808300263 DR 880812 p ADOCK 05000219 PDC

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TABLE OF CONTENTS PAGE 1.0 Introduction 4 1.1 Purpose 1.2 Summary of Oyster Creek IGSCC Inspections.

1.2.1 10R and 11R Refueling Outages-Inspections / Repairs.

2.0 Sunnary of Specific Piping Systems 6 2.1 Recirculation System 2.1.1 System Description, Materials and Operations.

2.1.2 Background / History 2.1.3 12R Planned Scope of Work 2.1.4 GPUN Inspection / Improvement Program 2.1.4.1 Future Improvements 2.1.4.2 Future Inspection Program 2.2 Core Spray System 9 2.2.1 System Description, Materials and Operations 2.2.2 Background / History 2.2.3 12R Planned Scope of Work l 2.2.4 Future Inspection Program 2.3 Shutdown Cooling Sy em 10 2.3.1 System Deser an, Materials and Operations 2.3.2 Background /Hi.cory 2.3.3 12R Planned Scope of Work  ;

2.3.4 GPUN Inspection / Future Improvements '

2.3.4.1 Future Inprovements 2.3.4.2 Future Inspection Program 2.4 Reactor Water Clean-Up (RWCU) System 10 i 2.4.1 System Description Materials and Operations l 2.4.2 Background / History 2.4.3 12R Planned Scope of Work 2.4.4 GPUN Inspection / Improvement Program l s 2.4.4.1 Future Improvements 2.4.4.2 Future Inspection Program 2.5 Isolation Condenser System 12 2.5.1 System Description, Fbterials and Operation 2.5.2 Inside Containment 2.5.2.1 Background / History 2.5.2.2 12R Planned Scope of Work 2.5.2.3 GPUN Inspection / Improvement Program 2.5.2.3.1 Future Improvements 2.5.2.3.2 Future Inspection Program

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PAGE 2.5.3 Outside containment 2.5.3.1 Background / History 2.5.3.2 12R Planned Scope of Work 2.5.3.3 GPUN Inspection / Improvement Program 2.6 Closure Head Piping Welds! .

14 2.6.1 Description, Materials and Operation 2.6.2 12R Planned Scope of Work 2.6.3 Future Improvements 2.6.4 Future Inspection Program 3.0 Water Chemistry Control At Oyster Creek 15 3.1 GPUN Actions Taken 3.2 Water Chemistry Effect Data'from other Utilities 4.0 GPUN Technical Clarifications to GL 88-01 16 4.1 Comparison of 10R and 11R Inspection Methodologies 4.2 Post Stress Improvements 4.3 Inspection of Cast Materials 4.4 Implementation of Hydrogen Water Chemistry 4.5 Reactor Water Clean-Up System Scope Reduction 5.0 NUREG 0313, Rev. 2 21 5.1 NUREG 0313 scope  ;

5.2 GPUN Proposed Program j 5.3 Summary of Program Differences l 5 .1 Sample Expansion 6.0 Other Generic Letter 88-01 Responses Required 24 l

7.0 Suranary 27 8.0 References 27 9.0 Figures 28 s 10.0 Tables 33 TOTAL EFFECTIVE PAGES .55

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1.0 INTRODUCTION

1.1 Purpose - GPU Nuclear (GPUN) is required to perform. inspections'of reactor coolant systems (RCS) piping to' detect intergranular stress corrosion. cracking (IGSCC). For the upcoming 12R refueling outage, GPUN will meet the intent of ~ Generic Letter. 84-11[1],

which has been the' guideline for such-inspections until recently.

In January of this year, the Nuclear Regulatory Commission (NRC) issued Nuclear Regulatory Guide (NUREG) 0313 Rev. 2[2} , and this was supplemented by Generic' Letter (GL) 88-0l[3]. In accordance

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with the generic letter, a response is due within six months from the date received outlining future inspection program' plans including the next refueling outage. As explained earlier, 12R inspections will meet.the intent of 84-11. GPUN has reviewed the NUREG for full compliance during the 13R refueling outage. There are currently 473 welds at the Oyster Creek NGS which are within the scope of the NUREG. Full compliance during 13R would be extremely burdensome, result in extensive radiation exposures (apprx. 575 man-rem), and increase the outage schedule by approximately 3-4 weeks. Because of these concerns,.our excellent and unique history concerning IGSCC and the need to have a manageable work scope, reasonable personnel exposure levels, and a realiatic long-range plan agreed on with the NRC, GPUN will implement the long term program (defined within) in the 13R refueling outage. The GPUN Program meets the intent of the NUREG while reducing radiation exposures.

In summary, this document will delineate GPUN's planned program for the mitigation and inspection of IGSCC at the Oyster Creek Nuclear Generating Station. It will provide a history of the IGSCC inspections, repairs and mitigating actions to date, as well an identify our planned improvements to minimize the possibility fct IGSCC which is as follows:

" Implement Hydrogen Water Chemistry (HWC) during the' beginning of the Cycle 12 operating cycle (1989). Where HWC is effective, a factor of 2. reduction in inspection frequency has been applied.

• Stress improve all accessible /inspectable welds inside the drywell (except Reactor Water Clean-Up System) by the end of the 14R refueling outage (1992).

• Replace all Isolation Condenser condensate piping outside the drywell from the isolation condenser to the containment penetrations, the steam side piping on the 75' elev. from the containment penetration to the penetration of the 95' elev. floor and the drywell penetration piping with resistant materials during the 13R refueling outage (1990).

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• Corrosion Resistant Clad.(CRC: the welds in the Reactor Water Clean-Up System drywell penetrations and welds on the closure head of the reacto: during the 13R refueling outage (1990).

'* Replace other closure head piping'IGSCC susceptible welds with resistant materials during tee.13R refueling outage (1990).

. i This document will also identify the basis for CPUN's technical-clarifications as compared with NUREG 0313. The clarifications are as noted below, and will be discussed in detail in latet I sections of this document.

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• Acceptability of our 10R (1983/1984) inspections for I detecting IGSCC. ';

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• Use of visual inspection during. pressure testing for I casting-to-casting welds. l

• Reduce inspection fregency of stress improved welds. 'l

• Perform less than 100% UT inspection following stress improvement of small diameter welds.

• Limit scope of IGSCC inspections for the Reactor Water Clean-Up System (RWCU) to inboard of the outermost l containment isolation valves. l l

1.2 Summary of Oyster Creek IGSCC Experience l 1.2.1 10R and llR Refueling Outages - Inspections / Repairs In 1983 (10R), 31 Recirculation System welds were inspected to IEB 82-03f^], including three velds that were fluorescent y e penetrant inspected on the ID and dispositioned as geometry, and 3 core spray welds. No indications of IGSCC were detected [5). -i In 1984, a leak was detected from an Isolation Condenser return line weld during a pressure test of the-condenser tubes. This leak resulted in the inspection of over 150 s welds in the Isolation Condenser System (ICS) and the RWCU System. All the inspectable steam and condensate welds outside the drywell (127) were inspected. Twenty-seven welds in the ICS piping outside the drywell were found to contain indications of IGSCC. Nine were replaced thru spool piece change out and eighteen were repaired with full structural weld overlays (6).

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In the'following 11R refueling outage, inspections of Reactor Coolant System (RCS) welds were performed following the guidelines of Generic Letter 84-11. The inspection included 151 butt welds and all eighteen (ICS) weld overlays deposited in 10R. Three welds in the C-loop of the Recirculation System and one weld in the Isolation Condenser System (ICS) steam line had indications of IGSCC. The indication in the ICS weld was determined to have been a misinterpreted indication during 10R and not a "new" crack [7]. One Recirculation system weld was evaluated against the criteria of the then-draft Rev. 2 of NUREG 0313, and accepted as. stress improved (SI'd); the other two Recirculation system welds and the ICS weld were repaired with full structural weld overlays.

2.0 Summary of Speci.fic Piping Systems 2.1 Recirculation System 2.1.1 Fystem Description, Materials and Operations Oyster Creek's Recirculation System consists of 26" OD piping fabricated with Type 316 heavy wall stainless steel. The entire system experiences large flow rates of water at operating temperature and pressure while the reactor is operating. The system consists of five loops with piping of uniform dimensions. Unlike later vintage boiling water reactors (BWR's), each loop is segregated from the others and takes suction from the reactor vessel I annulus and discharges to the lower vessel area containing the diffuser.

2.1.2 Background / History IGSCC in this system was described earlier in Section 1.2.1.

During the 11R refueling outage of 1986, 64 of the system's.

welds were stress improved (SI'd) with the induction heating stress improvement (IHSI) method. All 64 welds were inspected following stress improvement and weld j overlay repair of two welds was performed.

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The Recirculation System safe-ends were overlayed on both the ID and OD before operation began in 1969(8]. This was performed because cracking was detected in several Type 316 components that were subjected to the final vessel heat treatment resulting in the sensitizing of these ,

components. The overlays (both ID and OD) were "low I carbon, high-ferrite" (as stated in the repair specifications) Type 308L weld metal. Howevet, since the safe-end to nozzle shop weld was performed with Inconel 182 weld metal, a portion of the ID is covered with 182 (see Figure 1).

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Since we cannot locate the weld material chemistry. test reports nor the inspection reports or the recorded, as-deposited ferrite content, we consider it prudent to stress improve the safe-end to piping system welds. And, since the ID of the nozzle to safe-end welds were.overlayed with Incocel 182, we consider it prudent to stcess improve the nozzle to safe-ends welds. Approximately 23 man-rem of exposure per safe-end will be required to perform these tasks.

2.1.3 12R - Planned Scope of Work - During 12R, 10% of the 64 previously stress improved (SI'd) welds will'be inspected including the three cracked welds.  ;

i Also during 12R, both "C" loop recirculation safe-ends will potentially be SI'd and inspected, if the inspection ' '

technique can be qualified. The effort required to perform j this includes machining the OD cladding to provide a l surface finish and contour adequate for performing UT for j IGSCC. Then stress improving followed by a post process UT  !

inspection. The inspection area will include the nozzle to j sefe-end weld (the safe-end side of that weld, and-the l nozzle side of that weld for a distance of IT into the )

nozzle from the veld centerline) and the safe end-to-pipe l weld.

The technical basis for selecting these two safe ends is  !

that during the llR outage, GPUN stress improved 64 Recirculation System welds and inspected all of them after l j

v SI. The onl- Toop to contain welds with indications of  !

IGSCC were t:.rce in the C-loop. During the' overlay I

' cladding effort before operation, described above, access to tae vessel was provided by removing the elbow at the top of the C-loop vessel inlet riser. One of the replacement ,

welds (NG-C-23) was one of the three welds found to contain '

indications of IGSCC during 11R. We consider that these circumstances provide sufficient concern to warrant stress improving these two safe ends first.

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We. expect that treating these'two safe e~nds during 12R will provide much needed, useful information.for stress

' proving, machining, and inspecting the remaining ~eight safe ends in future outages.. Knowledge of the radiological environment and lessons learned during-the.12R outage will enable us to more efficiently perform this work in_the future. le. expect that time and exposure savings would_be substantial. If no-cracks'are found in the C-loop safe ends ic 12R, we consider that these results coupled with' the implementation of HWC in Cycle 12 provides adequate assurance that cracks will not initiate / grow in the other eight. safe ends. Should cracking be detected in the C-loop safe ends during 12R, the other eight safe ends will be-stress improv i and inspected during 12R.

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2.1.4 GPUN Inspection / Improvement Program 2.1.4.1 Future Improvements - The planning for the recirculation system includes HWC following 12R.

The remaining safe-ends in A, B, D, and E loops of the recirculation system will be stress improved and post process inspected, (depending upon the success with 12R inspections) in 13R and 14R. The four vessel inlet safe ends will be stress improved and inspected during 13R, and the four vessel outlet nozzles will be stress improved and inspected when the system is deconned during 14R.

2.1.4.2 Future Inspection Program - Beyond 12R, Oyster Creek's recirculation system will benefit from most welds having been stress improved (except-five casting to casting welds-(pump suction elbow to pump casing) and 8 safe-ends (16 welds)) and ICSCC mitigated by the introduction of Hydrogen Water Chemistry (HWC). HWC will aid in preventing IGSCC. Because IGSCC will be significantly mitigated by the two actions mentioned (See Sections 4.2 & 4.4 for. technical basis), a 50%

inspection of uncracked welds every ten years is considered sufficient (i.e., 10% every refueling outage). Repaired welds will be inspected at a rate of 50% in the outage following repair. Then all such welds will be inspected again within the next two refueling cycles. Recirculation system safe-end welds will be inspected immediately following stress improvement. The safe-erds will only be stress improved if an inspection technique and procedure can be qualitied.

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2.2 Core Spray System 2.2.1 System Description. Materials and Operations - The Core Spray System is designed only to be in use during a loss-of-coolant accident (LOCA) involving a. loss of reactor water inventory. Generally, the system has a temperature ,

environment of less than 200*F. Because'of thermal mixing, some piping and welds close to the reactor vessel exceed 200*F; therefore, boundaries of IGSCC susceptibility are I limited. The system consists of 6' inch and 8 inch diameter

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Typs 316 stainless steel pipe. Having two redundant loops, the system enters the reactor vessel through'two safe-ends I attached to two separate nozzles. The system has a total of 26 potentially IGSCC susceptible welds in the scope of l NUREG 0313 Rev.2.

l 2.2.2 Background / History - During the 10R outage, 3 Core Spray )

System welds were inspected for IGSCC. In 11R, 16 welds i were inspected,_of which 2 had been inspected in 10R. No l I

indications of IGSCC were detected.

2.2.3 12R Planned Scope of Work - During the upcoming 12R refueling outage, 9 Core Spray System welds will be )

inspected. This number includes the 4 safe-end welds  ;

associated with the system's 2 safe-ends. The 4-safe-end welds will only be stress improved if the inspection procedure is qualified. In addition to this, the safe-ends (only if capable of being inspected) and all IGSCC susceptible welds in the system will be stress improved.

Not all stress improved welds will receive a post-process inspection but a sufficient sample will be post process inspected (see Section 4.2 for Technical Basis). However, all safe-end welds will be inspected after stress improvement.

2.2.4 Future Inspection Program - All 26 welds within the scope of NUREG 0313 will be stress improved during 12R. Because of the IGSCC mitigating activities being performed on Core Spray, 20% of the system welds will be inspected per outage (100% of affected system welds to be inspected every 10 years). This is considered sufficient (See Section 4.2 for Technical Basis).

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However, initiation may be at temperatures as elevated as 350'F for very short periods of time. The system is composed partly of 14 inch diameter Type 316 stainless steel, schedule 80 pipe (14 welds total), and makes a transition to carbon steel inside the drywell before the -

2nd Containment Isolation Valve-(CIV).

2.3.2 Background / History - This system has shown no indications of IGSCC thus far. Two of the welds in' Shutdown Cooling were inspected in'10R. During llR 6 welds.were examined of which 2 were examined in 10R. The system contains a total of 14 susceptible welds.

2.3.3 12R Planned Scope of Work - It is planned'to inspect two welds in Shutdown Cooling during 12R. In addition, HWC, which will be implemented following the 12R refueling outage, will be a significant IGSCC mitigator for 9 of the 14 welds (see Section 4.4 for Technical Basis).

2.3.4 GPUN Inspection / Improvement-Programs 2.3.4.1 Future Improvements - The 14 welds will be stress improved and post process inspected during the 13R refueling outage.

2.3.4.2 Future Inspection Programs - Based on GPUN's plans to stress improve this system's piping and with the implementation of HWC, an inspection schedule of 10% of the system welds will be inspected per outage (i.e., 50% every 10 years). This is considered sufficient (see Sections 4.2 & 4.4 for technical basis).

2.4 Reactor Water Clean-Up (RWCU) System 2.4.1 System Description, Materials and Operations - The RWCU System is operating at virtually all times during power operation. It consists of 6 inch diameter Type 316 stainless steel pipe. Up to the inlet to the first non-regenerative heat exchanger and from the outlet (return) of the third regenerative heat exchanger, reactor coolant is above 200'F.

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2.4.2 Background / History - During the 10R refueling outage, 8 welds were inspected in the RWCU System. A. total of 10  ;

welds were inspected during 11R of which 2 had been i inspected in the 10R refueling outage. No IGSCC has been detected in the system to-date.  ;

2.4.3 12R Planned Scope of Work - It is planned that 10 RWCU ,

System 12R. welds inside containment will be inspected during In addition HWC, which will be implemented following the 12R mitigator.

refueling outage, will be a significant IGSCC I

2.4.4 GPUN Inspection / Improvement Program  :

2.4.4.1 Future Improvements - Since RWCU is operating during plant power operation, the system will )

derive full benefit from HWC. In addition, 5

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welds that are inside containment penetrations (unaccessible for UT inspection) will be  ;

internally clad with IGSCC resistant material in 13R. {

2.4.4.2 Future Inspection Program - 20% of RWCU welds inside containment will be inspected each outage (i.e., 100% every ten years).

The welds that are outside containment but inboard of the second containment isclation valve (CIV) will be inspected following the system decon that will occur in 14R.

The reason for this is that these welds would otherwise present a high personnel exposure problem (Appx. 133~ man-rem w/o decon and 27 man-rem w/decon). The possibility of future inspections on weldt between containment and the second isolation valve being performed visually such as (television cameras, leak detection, etc.)

is being investigated. This piping is currently walked down by Plant Operations daily. Should cracking be detected in welds inside the second containment isolation valve (CIV), sample expansion selection will consider welds out to the regenerative heat exchanger. The five (5) CRC'd welds inside the penetrations will be examined for leakage during pressure testing.

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2.5 Isolation condenser System 2.5.1 System Description, Materials and Operation - The Isolation Condenser System (ICS) is a standby, high pressure system for removal of fission product heat from the reactor vessel following a reactor trip and for isolation of the reactor from the main condenser. The system prevents overheating of the reactor fuel, controls the reactor pressure rise, and limits the loss of reactor coolant through the relief valves. During normal power operation, the system is fully.

pressurized; however, flow is prevented during the standby mode by one closed condensate return line isolation valve per loop. The system consists of 10", 12" and 16" Type'316 Stainless Steel on the steam side, and 8" and 10" Type 316 stainless steel on the condensate side. The ICS has -1 experienced numerous initiations in the years  !

since initial start up of Oyster Creek. In addition, isolation valve seat leakage has l resulted in portions of the system experiencing temperatures greater than 200'F for long periods l of time without the system actually operating. l These 2 operating scenarios may account for the extensive but localized IGSCC in the ICS {

condensate return line outside of containment. l 2.5.2 Inside Containment 2.5.2.1 Background / History - No IGSCC has been detected in this portion of the ICS. During.10R, 19 welds had been inspected. In 11R, 12 welds were )

inspected of which 4 had been examined during 10R. l l

2.5.2.2 12R Planned Scope of Work - During 12R, the 2 ICS safe-ends will be examined (a total of 4 welds) and 10 welds in piping will be inspected. The four safe-end welds will only be inspected and stress improved if the inspection procedure is qualified.

Stress improvement will be performed on the 2 ICG 1

safe-ends (4 welds) and on 9 additional steam side welds. Not all stress improved welds will receive a post-process inspection but a sufficient sample will be post-process inspected (see Section 4.2 for Technical Basis). However, all safe-end welds will be inspected after stress improvement. In addition, HWC, which will be implemented following the 12R refueling outage, will be a significant ICSCC mitigator for the condensate piping up to the second valve.

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2.5.2.3 GPUN Inspection / Improvement Program 2.5.2.3.1 Future Improvements - In addition to the 12R stress improvement work scope, 10 welds inside containment will be stress improved during 13R. In 14R, all of the welds not yet stress improved will be treated.

During 13R, the ICS containment i penetrations will be replaced with resistant material, thereby eliminating a number of uninspectable l welds and some susceptible welds.

l I l l 2.5.2.3.2 Future Inspection Program - The steam l side portion of the system piping will i have 20% of its welds inspected each l outage (i.e., 100% every 10 years)

(See Section 4.2 for Technical Basis),

while the condensate side. portion of the system piping will have 10% of its welds inspected each outage (i.e., 50% i j every 10 years) (See Section 4.2 & 4.4 '

for Technical Basis).

l l 2.5.3 Outside containment 2.5.3.1 Background / History - In 10R, 127 ICS welds outside  !

of containment were examined. In this case, a large sample was inspected because of detecting a i through-wall leak in an 8" dia. condensate return line weld. As a result of this ef fort, 18 welds were overlay repaired and 9 were replaced through spool piece changeout. During 11R, 58 previously inspected welds were examined. Of these, 18 were the 10R overlayed welds. One additional weld was found to contain indications of IGSCC, and as a result, was overlayed in 11R. This weld was i diagnosed during 10R as having a root geometry l indication. Seventeen (17) more welds that were installed as a result of spool piece repairs in l 10R were not included in the 11R sampling base.

2.5.3.2 12R Planned Scope of Work - During 12R, GPUN plans

, to inspect 19 welds in ICS outside containment.

Within this sample are 4 overlayed welds, of which I was overlayed in 11R.

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-2.5.3.3 GPUN Inspection / Improvement Program - All of the condensate piping outside the drywell, all of the steam piping on 75' elev. outside the drywell and the containment penetrations (uninspectable welds)

(See Section 2.5.2.3.1) will be replaced in 13R with ICSCC resistant material. There will be no uninspectable welds in the penetrations. As j category A welds, they will require inspection at a rate of 25% entry 10 years. j

.l 2.6 Closure Head Piping Welds  !

2.6.1 Description, Materials, and Operations There are three alloy steel nozzles with Type 316 stainless i steel weld neck flanges welded to them on the reactor l vessel closure head. The original flanges were subjected l to the closure head's final post weld heat treatment.

Thereby, the flanges were furnace sensitized. All three  !

flanges were replaced as part of the pre-operation repair effort described earlier. The nozzle weld preparation was buttered with Inconel 182, and the replacement flanges butt welded with Inconel 82. One 6" nozzle is a spare to which a blind flange is bolted. The other 6" nozzle is the inlet for the head spray line. The 4" nozzle is for vent piping.

2.6.2 12R Planned Scope of Work During the 12R refueling outage 2 of the closure head welds will be inspected.

2.6.3 Future Improvements We plan to improve all welds to Category A.by the end of the 13R outage. The three nozzle-to-flange welds will be corrosion resistant clad (CRC) with Inconel 82. The remaining welds will be replaced with materials resistant to IGSCC.

2.6.4 Future Inspection Program Applicable welds will be inspected in accordance with Generic Letter 88-01 for Category A.

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3.0 1 WATER CHEMISTRY CONTROL AT OYSTER CREEK 1

3.1 GPUN Actions Taken- The following actions have been or will be taken to mitigate the initiation and propogation of IGSCC at Oyster Creek:

• Implemented EPRI water chemistry guidelines (1984). )

• Commence hydrogen water chemistry (HWC) following 12R(Jan. 1989).

• Estabished a new chemistry laboratory with state-of-the-art-  !

equipment for analyses (1985). i Plugged a large number ot' leaking condenser tubes during 10R.

During Cycle 11, no tube leaks were evident. I

• Replace resin in the reactor water cleanup demineralizer at the onset of silica leakage.

• Perform air in leakage surveys. )

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• Replace resin in condensate polishers before onset of significant ionic leakage (1985). l Our efforts to improve improvements water chemistry have resulted in substantial over the last two operating cycles.

during Cycle 10 and Cycle 11 to-date has been ~ 0.1 pS/cm. Water conductivit consider that We this improved conductivity has substantially reduced the 1 potential for new crack initiation and has slowed the growth of existing, undetected cracks, if any.

will further reduce the potential for new cracks.The addition of HWC in Cycle 12 Pipe tests sponsored by EPRI & GE indicated a factor of improvement of 20 with respect to crack initiation.

We also consider that the lack of IGSCC in the shroud head bolts (creviced alloy performance 600) in the is an indication nf adequate water chemistry past.

In other BWR's, many of these bolts, identical in design to Oyster Creek's, were fe.ad to contain crevice-induced IGSCC, including some that were 100% cracked. The cause of this cracking has been attributed to the presence of a crevice, beneath the collar, and water conductivity. In the 11R outage, all 36 bolts were UT inspected at Oyster Creek. Nr, indications were detected.

control noted earlier are expected to provide significantAgain, improvement the imp in reducing the potential for new crack development.

3.2 Water Chemistry Ef fect Data From Other Utilities In-plant water chemistry studies performed at Dresden-2 and Peach Bottom-3 have shown the significance of improving normal water chemistry control (NWC), specifically for conductivity. This work, which is partially funded by EPRI, is being performed to show the improvement practices are followed. on materials performance when good chemistry The Dresden-2 data has been obtained while the plant was injecting hydrogen (HWC) and the PB-3 data was obtained under NWC conditions. Figures 2 and 3 show the average crack growth about 27 Ksl rates

- __for sensitized Type 304 specimens loaded to Q in.

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Figure 2. data for.PB-3 clearly shows the improvement in crack growth when water. conductivity is . reduced.. There is a factor of 4 improvement (from 96 to 24 mils /yr.) when the average. conductivity is reduced from 0.5 to 0.2 pS/cm. The Figure 3 PB-3 data shows the impact of two resin intrusions on growth while. the normal chemistry conductivity had been maintained at - 0.2 pS/cm.

Even with the two resin instrusions within a'short period of time apart (~ 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />), the average crack growth rate increased by only a factor of 2 (from 22.5 mils /yr. .to 44 mils /yr.) We would expect,that barring.furtherfupsets the crack growth. rate would return to its rate before the intrusions. These data were obtained from CAV specimens in autoclaves connected to the plant's RCS.

Similar results have been obtained in the laboratory. The crack growth rate for sensitized Type 304 at ~ 0.1 pS/cm conductivity was ~ 29 mpy while at ~ 0.45 pS/cm, the rate was ~ 240 mpy 1 (Figure 4). These growth rates compare very favorably with the NRC-calculated rate of 390 MPY (Appendix A.2 of NUREG 0313 Rev. 2). Figure 2's Dresden-2 data shows the impact of implementing HWC with low conductivity ( ~ 0.l'pS/Cm). The Figure 4 data shows similar results in laboratory conditions. We consider that the low average conductivity obtained at Oyster Creek over the last two cycles has significantly contributed to a reduced propensity for developing new cracks and a slow crack growth rate - 1 for potential existing cracks; therefore, even if cracks have initiated, growth will be slow and does not represent a safety concern. The implementation of HWC following the next outage-(12R),

will further improve the condition of affected stainless steel P Ii P ng.  !

l 4.0 GPUN TECHNICAL CLARIFICATIONS TO GL88-01 i

4.1 Coinparison of 10R and 11R Inspection Methodologies l Based upon the IGSCC personnel qualification and procedure j development for the 10R inspection and confirmation of the 11R  !

result *, the GPUN program takes credit for the 10R IGSCC inspections. During the 10R and 11R outages, inspections were performed to look for IGSCC in various systems at Oyster Creek.

While the inspections had differences, the results were consistent. The following is a summary of the 10R and 11R s inspections.

During 10R, inspections were performed by Magnaflux personnel using a qualified ICSCC inspection procedure. The GPUN procedure was qualified at Battelle Columbus Laboratories, and all inspectors were further qualified to the procedure on site at Oyster Creek utilizing EPRI supplied samples. The inspection was manual using a 45' shear probe supplemented when required with a 60'.

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During 11R, inspections were performed by GE using manual and automated techniques. The manual techniques were similar to those used in 10R. The probes used in llR were of an improved design using the WSY-70 technique. Personnel and procedures were qualified by the EPRI NDE Center. The automated inspections were performed using the GE Siaart system. This system included automated probe manipulation and automated data evaluation. Again, both the procedure and personnel were qualified at the EPRI NDE Center under the NRC/BWROG/EPRI Coordination Plan.

During the 10R' inspection, 181 welds were examined. Twenty-seven welds were characterized as having IGSCC and were either overlay repaired (18) or replaced (9). Three recirculation system welds were suspected of containing IGSCC and, based on further ID visual and liquid penetrant examinations, were dispositioned as geometry.

During the llR inspections, 169 welds were examined. Ninety-seven welds were previously examined during the 10R inspections. The three previously examined Recirculation System welds from 10R that were also examined on the ID were again confirmed as geometry. Two other welds (NG-C-23 and NE-1-27) which had indications that were dispositioned as geometry during the 10R inspection were dispositioned as containing IGSCC During 11R.

A comparison of the techniques used during 10R and 11R is presented in Table 1. Table 2 describes all welds inspected during 10R and llR and their status. Table 3 provides a suenary of welds inspected.

In conclusion, 98% of the repeated examinations from 10R were confirmed during ilR as not having IGSCC indications. Although the techniques employed during 10R and 11R were ret identical, the results were consistent. All indications detected during the 10R inspection were supplemenced by an RT examination and for those extreme uncertainty calls, three recirculation system welds were PT examined from the inner diameter. Additionally, if NDE could not clearly distinguish IGSCC from geometry, we conservatively dispositioned them as containing IGSCC.

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4.2 Post Stress Improvement Inspections - We consid_r that for certain sizes of piping performing stress-improvement (SI) without 100%

immediate post-SI inspection is a prudent technical approach to mitigating IGSCC and that performing a 100% reinspection over the

~

following two outcges is not warranted. The concern with not performing immediate irspection is that, while SI places the inner half of the weld in compression, it also places the. outer half in tension. For large-diameter piping (> 12-inch), the residual stress neac mid-wall is largely compressive (~ - 15 Ksi). This is the reason that most ICSCC in large diameter piping appears to have been arrested near mid-wall. Changing the residual stress pattern such that the outer half is subjected to tensile stresses-could, in fact, result in continued through-wall growth of a previously arrested crack. Although we are not aware of any cases where this has happened, we consider this to be a valid technical concern. Therefore, for all 14-inch Shutdown Cooling welds which are stress improved, a post process inspection will be performed.

However, for smaller diameter piping (<12-inch), calculations show that a 10% through-wall (TW) crack will propagate to 80% TW within one operating cycle. The maximum crack depth allowed by Section XI is 60% for SMA and SA welds. This is a result of the presence of a linear TW residual stress that is tensile on the ID and compressive on the OD. Once the crack reaches mid-wall and enters the compressive reg:on, the applied stress intensity is too high for the compressive stresses to stop, or even retard, crack growth.

Stress improvement of smaller diameter piping will not make conditions for unacceptable crack growth worse. .For example, SI of 8-inch diameter piping with a 10% TW crack will essentially prevent further crack growth, whereas a 10% TW crack in an as-welded joint will grow to an unacceptable depth within one operating cycle.

After SI of an 8-inch diameter joint, a crack must be at least 72%

TW before it can grow. These comparisons were based on calculations using a 10 Ksi operating stress. Similar results were determined for 10-inch diameter pipe.

The major conclusion of this evaluation is that if a crack will not grow to an unacceptable depth within an operating cycle in the as-welded condition, the sanie crack would not grow to an unacceptable depth within an operating cycle if stress improved.

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I We are unaware of any instances of crack initiation or unacceptable i crack growth in a properly stress improved weldment. We are aware of..several instances of supposed new cracks in SI'd welds that were inspected in successive outages. In the first. outage, the j indications were not dispositioned as ICSCC; in the following outage, indications were dispositioned as IGSCC. However, reviews of these cases concluded that the interpretation of the first i outage's data was incorrect and should have been dispositioned as ICSCC. We are also not aware of any cases of IGSCC being detected in a stress improved weldment that contained no reportable / recordable indications in the previous outage inspection. l We consider that this evaluation shows that it is prudent to stress improve smaller diameter piping regardless of whether or not.100%

immediate post-SI inspectica is performed. SI will improve the condition of joints with no or shallow cracks and will not worsen ,

the condition of joints with deeper cracks. We do not consider it l technically prudent to avoid SI due to the ALARA penalty taken to perform 100% immediate post-SI inspection and the associated required weld crown reduction to facilitate the inspection.

Therefore, the GPUN Program does not require 100% post-SI inspection for welds <l2 inches in diameter (core spray and ICS ')

excluding safe ends). However, for welds which will be both SI and inspected during the same refueling outage, the inspection will be performed, in a majority of cases, after SI.

4.3 Inspection of Cast Material -There are currently no NDE methods  !

that comply with NUREG 0313 requirements which detect IGSCC in  !

as-cast Type 316 stainless steel. When an inspection technique is '

qualified for our configuration, we will consider adding these welds to our inspection program. There are also no reported cases of through-wall IGSCC in austenitic stainless steel castings. GPUN l does not consider that IGSCC in cast stainless steel is a generic  !

problem.  !

By the end of the 14R outage, only the 5 Recirculation pump suction elbow-to-pump casing (both are castings) welds will not have been stress improved. That is because heavy-walled piping should not be stress improved unless it can be adequately examined afterwards.

However, these 5 welds will be protected by HWC. For this reason, visual inspection during pressure testing for these 5 Recirculation ,

s System welds will be performed in lieu of UT inspection. l 1

4.4 Implementation of Hydrogen Water Chemistry l CPUN will be implementing Hydrogen Water Chemistry (HWC) following i the 12R (1988) refueling outage. Where HWC is effective in flowing I systems (Recirculation System and Reactor Water Clean-Up System), a j factor of two reduction in inspection frequency has been applied.

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Portions of systems that are stagnant during power operation will be afforded some level of additional protection by HWC.. The existence of turbulence and/or thermal ~ mixing would provide this benefit. Systems considered to-benefit in this,way include shutdown cooling and ICS condensate-return lines, all of which are attached to the recire, piping. For the GPUN Program, a factor of 2 reduction in inspection frequency has been applied.

4.5 Reactor Water Clean-Up System Scope Reduction The Reactor Water Clean-Up System piping outboa.rd of the outermost Containment Isolation-Valves will not be inspected or included in l NUREG 0313 scope for the following reasons:

4.5.1 Safety Significance of Reactor Water Clean-Up System - A pipe

! failure outboard of the outermost Containment Isolation Valves l (CIV) would not have nuclear safety consequences. The effects

( of postulated pipe failures outside containment on equipment l and structures necessary to shut down the reactor were 1

evaluated as described in Section 3.6.2.6 of the FSAR. A.

double-ended rupture of the RWCU piping outside the outermost CIVs would result in a low reactor coolant level and automatic isolation of the RWCU from the reactor. For this reason, j there would be no compromise in the abliity to safely shut down the reactor. The radiological releases for such a break would be within the secondary containment boundary and would not cause 10CFR20 limits to be exceeded.

l 4.5.2 Radiation Exposure Considerations l This piping is also a high radiation area. While decon will reduce exposure, there would still be a lot of dose expended in prepping the welds, inspection, scaffolding, etc. The man-rem exposure for performing inspections and weld crown l reduction (excluding scaffolding) is estimated at 343 man-rem i w/o decon and 69 man-rem w/decon. Scaffolding and insulation  !

removal and reinstalling is estimated at 200 man-rem w/o decon j and 40 man-rem w/decon. However, as noted earlier, should cracking be detected in the piping inside the second CIV, sample expansions may consider welds outside the second CIV to j the heat exchangers. l l

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5.0 NUREG 0313, Rev. 2 5.1 NUREG 0313 Scope For Oyster Creek, the scope of the NUREG applies to the following systems:.

System Extent No of Welds Recirculation Whole 89 Isolation Condenser Whole 189 Shutdown Cooling Stainless Steel Portion 14 greater than 200'F Core Spray. Stainless Steel Portion 26 greater than 200*F Reactor Water Clean-Up From Recirc. to inlet 147 side of non-regenera-I tive heat-exchanger-l and from outlet side of third regenerative l heat exchanger shell l to Recirc..

Closure Head Three nozzles to points

! at which diameter is less than 4" or changes ,

to carbon steel. 8 l

i l -

l Table 4 lists the approximate number of welds inciuded'in the scope of the NUREG for each system. The table also lists the number of uninspectable welds (and why).

l In summary, 473 welds (54 of them uninspectable) fall l within the scope of the NUREG. The number ;f welds in each system by category, as well as the number of inspections required for 13R and 14R by category, are listed in Tables 5, 7, 9, 11, 13, 15 and 17. The reason for identifying welds inside/outside the drywell for the Isolation Condenser system is that the drywell is a difficult area to  !

schedule work since it is normally the most congested area during an outage.

l s

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i I..__ ___ _ _ ._______._._.__..______._ _ ____.__ _ ._________.______._ __..._.. __ _

TR - 050' Rev. O Page 22.

5.2 GPUN Proposed Program - Tables 6, 8, 10, 12, 14, 16, 17 describe the categorization and inspection of welds before,

-during and after the 13R and 14R refueling outages. Tables 12, 15 and 16 indicate a dramatic reduction.in lower category welds because of replacement of a portion of the ICS piping outside of the drywell and the repairs to the RWCU penetrations and closure head welds. These tables also indicate the movement of many welds into category C due to stress improvement.

5.3. Summary of Program Differences The GPUN proposed program (See Table 18 for proposed inspec' on schedule) takes credit for all the technical clarifications listed in Section 4.0 and the replacement / repairs noted earlier. This reduces the total number of welds susceptible to IGSCC and within the scope of the NUREG from 473 welds to-333 welds. Also of the 333 welds, 60 welds will be category A "Resistant Material", and 33 will be uninspectable (this includes 28 safe-end welds which GPUN will attempt to make inspectable). As a result of this reduction and GPUN's technical clarifications, the quantity of welds to be inspected during the 13R refueling outage reduces from 260 to approximately 64 welds (excluding potential safe-end inspections). This reduction results in an overall' man-rem reduction of ~ 500 during 13R and an increase of ~ 70 w/o decon and ~ 6 w/decon for 14R. This results in a net reduction for the 13R and 14R outage for the GPUN Program of ~430 w/o decon and ~496 w/decon.

5.4 Sample Expansion - GL88-01 requires sample expansion if indications of ICSCC are detected in the initial sample. The additional sample size should be approximately equal to that of the initial sample of the category of weld in which IGSCC is detected, irrespective of sample and pipe size. If IGSCC is detected in the second sample, all welds in that category should be inspected. The saeple expansion requirements of GL 88-01 will be met as modified herein:

a) Recirculation System Safe-Ends If cracking is detected in the inlet safe-end welds in 13R, all eight remaining outlet safe-end welds will be inspected in 13R.

s b) Isolation Condenser Piping Outside Second Isolation Valve If indications of IGSCC are detected in the initial somple, the additional sample size will be approximately equal to that of the initial sample of the category of weld in which IGSCC was detected, but limited to the Isolation Condenser System outside the second isolation valve. If IGSCC is detected in the second sample, all welds in that category will be inspected within the Isolation Condenser System outside the second isolation valve.

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TR - 050 Rev. O Page'23.

c) RWCU Piping Because the RWCU System is the only system (excluding the.

Isolation. Condenser System outside the second isolation valve) which will not be stress improved, it'will be

-treated for sample expansions separately. JIhe RWCU System is also unique in that inspections outside the drywell are i very high exposures. '

Outside the drywell, there are 34 welds located in.the "valve nest." The average dose / weld for insulation removal, weld crown reduction, and inspection.is estimated at 8.5 Man-Rem. The remaining 78 welds to the 200*F transition are located.in the RWCU heat exchanger room.

The average dose / weld in this room is estimated at 2.5 Man-Rem. The dose estimates do not include scaffolding work; additionally, most of the insulation on this piping contains asbestos. It should be noted that GPUN does not consider the welds outside the second isolation valve within the scope of the Generic Letter (See Technical q Clarification 4.5).  ;

1 Therefore, for the above reasons the RWCU sample-expansion if IGSCC is detected within RWCU System will be only within RWCU and inside the second isolation valve. j d) Remaining Welds (Recirculation, Core Spray, Shutdown Cooling, Isolation Condenser Inside Drywell, and-Closure Head Piping)

We will meet the sample expansion requirements of GL 88-01. That is, for each category of weld, we will inspect l an equal number of welds in the second sample and, if cracking is detected in the second sample, all remaining welds in the applicable category will be inspected.

As previously discussed, we consider that the inspections performed in 10R, to IEB 82-03, were adequate. As such, .

Category D welds include those that were inspected in 10R but not re-inspected in 11R to GL 84-11.  !

l i

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TR - 050 Rev. O Page 24.

6.0 Other Generic Letter 88-01 Responses Required 6.1 The NRC Staff has proposed a change to the Technical Specifications (TS) to include a statement in the section on ISI that the Inservice Inspection Program for piping covered by the scope of NUREG 0313 will be in conformance with the NRC positions on schedule, methods and personnel, and sample expansion included in GL-88-01. The staff has also recognized that the Inservice Inspection and Testing sections may be removed from the TS in the future in line with the Technical Specifications Improvement Program. If this does come into existence, then this requirement would remain with the ISI section when it is included in an alternative document.

GPUN believes that TS changes should be viewed in light of the NRC Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors (52FR3788, 2/6/87). The intent of the Policy Statement is to improve the safety of nuclear power plants through the development of more operator-oriented TS, improvement of TS Bases, reduction of action statement induced plant transients and more efficient use of NRC and industry resources.

Since a TS for GL 88-01 requirements would not support the Proposed Policy, GPUN does not concur that a change to the TS is warranted.

In lieu of a TS change, GPUN proposes the inclusion of our response to GL 88-01 as part of our Inservice Inspection Program pending the final recommendation of the Technical Specifications Improvement Program.

6.2 The NRC Staff position in GL 88-01 is that leakage detection systems should be in conformance with position C of Regulatory Guide 1.45 "Reactor Coolant Pressure Boundary Leakage Detection Systems" or as otherwise previously approved by the NRC.

Leakage detection systems for Oyster Creek were reviewed by the NRC Staff during the Systematic Evaluation Program and the results were documented in Section 4.16.2 of Integrated Plant Safety Assessment Report for Oyster Creek, NUREG-0822 dated January, 1983. The actions identified in that report have been completed with the exception of the airborne particulate and gaseous radiation monitoring system (APGRMS). GPUN's recent submittal of July 1, 1988, states that installation of a new APGRMS will be completed during the operating cycle 12. The submittal also identifies that there are several leak detection methods available for unidentif ied leakage into the containment sump at Oyster Creek which operate on diverse principals.

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The normal method of monitoring unidentifledl leak rate is to j obtain flow Integrator readings f rom the containment sump pump

.j discharge'every four hourLperiod and calculate average flow rate.

~

-i Approximately 1~gpm can be measured'in a-four hour'. interval. This- 1 methodology.is identified in Oyster Creek Technical specifications. '

as.the primary method.of leakage measuremcnt.

~

- When the flow integrator is not available, the average' leakage. l rate can.be' calculated using'the known. volume between the high and.  ;

the low level' alarms for the sump'and-the time reouired to fill l the sump'between these levels. .)

A recorder available in the control room also provides continuous '

indication of.an estimated' unidentified leak rate to the  ;

containment sump by-utilizing a 'dif ferential pressure signal as a - l result of the sump level change.. The sensitivity of the recorder l is approximately 0.2 gpm. j

- Additionally, a timer available in the 480 volt switch' gear rood provides the run time of the containment-sump pumps. This run time along with the estimated flow rate of the sump pumps can 1 provide approximate leak rates. This methodology is utilized every four hours during power operation.

- Also, an annunciator will alarm in the control room if the time to fill the containment sump is too short an int < al. The time associated with this alarm is set to bring in the alarm if unidentified leak rate equals or exceeds 4 gpm. '

These methods provide quantitative indications of unidentified RCS ,

leakage inside containment and also provide assurance that I unidentified leakage can be detected and'av ntified during Cycle 12 operation pending operability of the ,.w-APGRMS.

The NRC Staf f position was further amplified in- GL 88-01 by additional criteria as follows:

1. Plant shutdown should be initiated for inspection and

.m etive action when, within any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or

!& any leakage detection system indicates an increase in

- m. of unidentified leakage in excess'of 2 gpm or its . ,

s . valent, or when the total unidentified leakage attains j a rate of 5 gpm or equivalent, whichever occurs first. For  !

sump level monitoring systems with fixed-measurement-interval methods, the level should be monitored at approximately 4-hour intervals or less.

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Unidentified leakage should include all leakage other than:

~

2.

r (a) leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered, and conducted to a sump or collection tank, or (b) leakage into the conteinment atmosphere from ources that are both specifically located and' known either not to intcrfere with the operation, of unidentified leakage monitoring systems or not to be from a throughwall crack in the piping within the reactor coolant pressure boundary,

3. For plants operating with an IGSCC Category D, E, F, oi s welds, at least one of the leakage measurement instrumaats associated eith each sump shall be operable, and the c- <-

time for inoperable instruments shall be limited to 24 tours, or inmediately initiate an orderly shutdown.

By Amendment 97 to Provisional Operating License No. DPk "S for Oyster Creek, the limiting conditions for operation ar' surveillance requirements were authorized for the Reactor Coolant System leakage. This amendment added two new definitions (identified and unidentified leakage) to TS Section 1.0; revised TS 3.3.D to include LCO's for the l containment sump flow monitoring system and the equipment i drain tank monitoring system; and added a new surveillance 4 section TS 4.3.H. This amendmer.t incorporated GPUN's  !

response dated September 8, 1983, to IE Bulletin 82-03.

On March 17, 1987, GPUN submitted Technical Specification Change Request #158 which adds additional c6ncervatism to these requirements by proposing to limit the unidentified leakage for the Reactor Coolant System to a maximum leak ate increase of 2 gpm within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period while operating at steady state power. As of this date, the NRC Staff has not completed their review of this proposed change. If approved, the TS would also address item 1 of the NRC Staff position.

s t

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6.3 GPVN plans to notify the NRC of any flaws identified that do not m(fst IWB-3500 criteria of Section XI of the Code for continued operation without evaluation, or a change found in the condition of the welds previously known to be cracked, and our evaluation of tne flaws for continued operation and/or repair plans. The NRC C review-shall not be a constraint for restart of the plant.

7.0 Summary - Even though GPUN is not in complete compliance with NUREG 03L3, the intent of the NUREG has been adopted. Oyster Creek has not experienced IGSCC to as great an extent as other BWR's. Through a balanced use of inspections and mitigating activities, GPUN will ensure that Oyster Creek does not have or develop a major IGSCC problem. Our approach is sensitive to a controllable outage work

• scope, reasonable personnel exposuee, and compliance in executing the long-range olan developed jointly by the NRC and GPUN.

8.0 Refereace

1. USNRC Generic Letter 84-11. "Inspections of BWR Stainless Steel Pipir.g," April 19, 1984.

2. NUREG-0313 Rev. 2. "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,"

, USNRC, January 1988.

3. USNRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," January 25. 1088. j

4. USNRC IEB No. 82-03 Rev. 1, "Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants," Octob*r 28, 1982.

5. GPUN Topical Report No. 012 dev. 1 "0yster Creek Recirculation System Piping Inspection Program." September 5, 1983.

6. GPUN Technical Data Report No. 580 Rev. 2. "Isolation Condenser System Piping Cracked Welds-Repair and Failure Analysis," 11-5-85.

7. GPUN Topical Report No. 039 Rev. O, "0yster Creek Cycle 11R Outage IGSCC Activities," 9-30-86.

8. Oyster Creek TSAR Amer *nts 29, 35, 36, 37, 40, 43 and 47. '

i _ _ __ - - - - - - - - -

o TR - 050 Rev. O Page 28.

9.0 Figures

1. Safe end configurat Un.

l

2. Comparisca of c ack growth rates.

3. Ecach Bottom Unit 3.

4 Effects of aqueous impurities on crack growth.

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TR - 050 Rev. O Page 29.

l 1

l 0/C Oesfon Configuration .

l l

Inconel 182 11 gget 3gg $fg yyp,3 gg,,,,73,y

. soul.

T ,,,,.. .

ry,. ii. ,

r- I c

son.i a.. ..

M IS2 s extended root type'308L clad overlay FIGURE 1 - SAFE-END CONFIGURATION I

- ,_ _ - - , ,.-_,.,.,,-..,---,,-,..,,.__,,,--,--,_-,-,.-.-._------.,,,-,--,,,_,,-__,.,,,--n-,,,,-,.--,_. --

TR - 050 Rev. O Page 30.

l 25 n 20 -

3 Avg cond -0.2 ps/cm E da Normal BWR 7 = 24 mils /yr 6 water chemistry a

(15 -

(Peach Bottom 3) p

.w

$ l O  ! Avg cond -0.3 s/cm E

f .da = 60 mils /yr dl  !

Hydrogen water checiistry l 5 - Avg cond -0.5 ps/cm (Dresden 2) l da Avg cond -0,1 ps/cm

- = 96 mils /yr da p dt - = <5 mils /yr 0 i d .  !

O 500 1000 1500 2000 2500 Test Time (hrs) .

Figure 2 4. Comparison of the crack growth rates of sensitized type 304 slainless steel spec 6 mens in normal water chemistry at Peach Bottom 3 and in hydrogen walor chemistry at Dresden 2.

s .

FIGURE 2 i

TR - 050 Rev. O Page 31.

PEACH BOTTOM UNIT 3 Sensitized 'sW 304 SS Crack Length (in.) Conductivity (uS/cm)

.754 2.5

.753 - ya = 2.39 2.57 x 10-4 in/h 2.0 et

.752 -

= 4.824.04 x 10-4 in.M - 1.5

.751 - -

1.0 Rosin Mn intrusion

• intrusion No.1 No. 2

.750 - -

0.5 L.

.749 I I I O 2100 2200 2300 2400 2500 2WX) 2700 2800 Time (h) u.

J FIGURE 3

TR - 050 Rev. O Page 32.

1 I

l CYCLIC

• / LOAD

+ % CONSTANT LOAD

.BR ~ -- ---

I i

.85 '- Cas < a wm M Acu N*2so4 < o., ,sfeu a o estCM N*2soa l g _

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5 1 -

.84 -

3 $ I u e _

.83 *

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- 21 i ~ g x .82 .

x

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$ i

- - - - -~ ~ ~ = 200 pph OX YGE N - - - - -- 2 3 - - 20 pph OX YGE N . - - ~

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5 .

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.80

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__ s - i

.78 * " ~

,77 _, , , -

1_ . s ,,

- s._ L = 1

• 1 = * ' ' I * * ~ A- - L-- l 0 500 1000 1500 2000 .

2500 TIMEthi j FIAure 4 Effeet of Aqueous impurit les on Crack Crowth of Sensit f red Type 104 Stainless Steel in 550*F (288*C) Water 1

1

j .

TR - 050 Rev.-0  :

Page 33. l 1

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10.0 Tables i

1. Inspection History Comparisons

2. Inspection History Per Weld

' 3. Welds Examined On IGSCC Affected Systems

4. Number of Welds in Scope of NUREG 0313, Rev. 2 1

5. Recirculation System - NUREG 0313

6. Recirculation System - GPUN Program l

7. Core Spray System - NUREG 0313

8. Core Spray System - GPUN Program I 9. Shutdown Cooling - NUREG 0313

10. Shutdown Cooling - GPUN Program

11. RWCU - NUREG 0313

12. RWCU - GPUN Program l

l 13. Isolation Condenser System Inside Drywell - NUREG 0313 i

14. Isolation Condenser System Inside Drywell - GPUN Program

15. Isolation Condenser System Outside Drywell - h0 REG 0313

16. Isolation Condenser System Outside Drywell - GPUN Program l 17 Closure Head - NUREG 0313 & GPUN Program

18. Comparison of NUREG 0313 & GPUN Program

19. Inspection Schedule Comparison >

l

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ATTRIBUTES 198310R INSPECTIONS 198611R INSPECTION laguistion Compliance ,

NRC issued Bulletin No. 82-03 dated

• NRC issued Generic Letter 84-11 October 14, 1982 "Stress Corrosion "Inspections of BWR Stainless Steel Cracking in Thick-Wall, Large Diameter, Piping" a reinspection program of j

Stainless Steel, Recirculation System piping susceptable to IGSCC from the 1

Piping at Bo111ag Water Reactor Plants" previous inspection performed under IE l Bulletin 82-03 and 83-02.

Personnel Qualifice Lion GPUN's Special Processes and Programs GPUN's Special Processes and Programs j

' , and Quality Control and the Nagnaflux and Quality Control and the General

, Corporation ultrasonic examiners Electric Corporation attended a attended a qualification session qualification session sponsored by

] sponsored by Battelle Laboratories EPRI at Charlotte, N.C. to qualify j Columbus, Ohio to qualify in the in the detection of IGSCC.

i detection of IGSCC.

Ntual cut-out specimens of km IGSCC taken from the Nine Mile Point Nuclear i Station's recirculation piping for use in gus11fication of Inservice Inspection Ultrasonic examiners.

All of the ISI Contractors Ultrasonic All of the Ultrasonic Examiners were Examiners selected far the Oyster Creek required to attend an IGSCC training .'

recirculation examinations were required session and qualify through practical to attend GPUN IGSCC training sessions examination by detecting known IGSCC '

and qualify through practical examination in welds specimens eight samples.

by detecting known IGSCC in weid specimens provided by EPRI four samples.

t l

" "* # 3 - - - - _ _ _ _ -- - - - - _ _ - - . _ - . - - - - - - - _ - - - - - _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

r 13 - 050 Rev. O .

TARI.E 1 Page 35. -

ATTRIBUTES 198310R INSPECTIONS 1986 IIR INSPECTION Comparison of welds examined during 97 welds .

both 10R and 11R - 97 welds ' 95 welds 2 welds No IGSCC No ICSCC IGSCC

'emparison of technique for 2 welds NG-C-23 subsequently found to have IGSCC NG-C-23 durl:g IIR Procedure 6130-QAP-7209.08. Rev. O GE UT 43. Rev. 6 FRR #2 J indications Root Geometry GE ESD UT-2. Rev. 2. FRR #1 2 Indications Geometry (RT Confirmed) 2 Indications IGSCC Se:rch unit 45* 45*

60* 60*

wsy-70 s11c-40 NE-1-27 ME-1-27 (GPUN) MTIS-008. Rev. 4 GPUN 6130-QAP-7209.08. Rev. ?

3 Indication Root Geometry 1 Indication Geometry (RT Confirmed) 2 Indication IGSCC Search unit 45* 45*

60*

wsy-70

• RTD-00 creeper

• 45* Spot Maximum Scanning Speed not to Exceed 6" per second not to . exceed 3" per second Minimum Search unit Overlap 251 501 Su rfcca .ndition Requirements NA Weid Crown Reduction s

Ol93v/3

TR - 050 Rev. O '

TABIE 1 Page 36. .

ATTRIBUTES 198310R INSPECTIONS 198611R INSPECTIUN TECHNIQUE < PROCEDURE UT PROCEDURE UT Detection GPUN Procedure MTIS-008. Rev. 4 GPUN Procedure 6130-QAP-7209.08. Rev. 2 Manual GPUN Procedure 6130-QAP-7209.08. Rev. O GE UT 1.30. Rev.12. FRR #3 Automatic GE UT 1.30. Rev.13. FRR fl GE UT 43. Rev. 6. FRR #2 GE UT 43. Rev. 7

lzing , ISI-2.9. Rev. O. Dev. 1 GPUN 6100-STD-7230.39. Rev. 0

, GE ESD UT-2. key. 1 FRR #1 GE ESD UT-2. Rev. 2 Ressits 181 Total Welds Examined 169 Total Welds Examined 27 Welds had confirmed ISSCC 4 Welds had confirmed IGSCC e

e C _m _ ____ _ _-_ _ _ _ _ _ - _ _ _ _ _ _ _ - - -

TR - 050 '

Rev, O Page 37.

TABLE 2 10R 11R ICSCC WELD S.I. INSP ICSCC? INSP IGSCC? DISPOSITION 1 NE-2-10 NO YES NO YES NO NA I 2 NC-B-12 IHSI YES NO YES NO NA 3 NE 'i +20 NO YES NO YES NO NA 4 NZ-3-82 NO YES NO YES NO NA 5 NE-2-21A NO YES NO YES NO NA 6 NE-1-1 NO YES NO YES NO NA 7 NE-2-23 NO YES NO YES NO NA 8 NE-1-5 NO YES EO YES NO NA 9 NC-B-23 IHSI YES NO YES NO NA i 10 NE-1-10 NO YES NO YES NO NA

! 11 NG-B-24 IHSI YES NO YES NO NA i 12 NE-1-21 NO YES NO YES NO NA l 13 NG-B-24A NO YES NO YES NO NA l 14 NE-5-13 NO YES NO YES NO NA l 15 NE-2-122 NO YES NO YES NO NA l l 16 NE-1-30 NO YES NO YES NO NA  ;

l 17 NE-2-24 50 YES NO YES NO NA i 18 NG-A-5 IHSI YES NO YES NO NA i 19 NE-2-27 NO YES NO YES NO NA 20 NG-A-6 IHSI YES NO YES NO NA i 21 NG-C-2 IHSI YES NO YES NO NA 22 NE-1-45 NO YES NO YES NO NA l 23 ND-10-9 NO YES NO YES NO NA l 24 NE-1-49 NO YES NO YES NO NA 25 NG-C-4 IHSI YES NO YES NO NA 26 NG-A-18 IHSI YES N0 YES NO NA 27 NE-2-51 NO YES NO YES NO NA 28 NE-1-56 NO YES NO YES NO NA 29 NE-2-53 NO YES NO YES NO NA 30 NG-A-25 IHSI YES NO YES NO NA ,

31 NG-C-9B IHSI YES N0 YES NO NA l 32 NE-1-64 NO YES NO YES NO NA ,

33 NE-2-79 NO YES NO YES NO NA l 34 NE-1-69 NO YES NO YES NO NA l 35 NG-C-12 IHSI YES NO YES NO NA l 36 NG-B-4 IHS! YES NO YES NO NA l 3T NE-2-83 NO YES NO YES NO NA l l

33 NE-2-1 NO YES NO YES NO NA 39 NC-2-85 NO YES NO YES NO NA ,

40 NE-2-7 NO YES NO YES NO NA l 41 NE-2-86 NO YES NO YES NO NA  !

42 NU-1-3 NO YES NO YES NG NA l 43 NG-C-22 IHSI YES No YES NO NA  :

! 44 NE-5-26 NO YES NO YES NO NA l 45 NG- IHSI YES NO YES YES WO 46 NE-1-t4 NO YES NO YES NO NA 47 NE-2-89 NO YES NO YES NO NA I 48 NE-1-27 NO YES NO YES YES WO I l 49 NE-2-92 NO YES NO YES NO NA i

l

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, i TR - 050 )

Rev. O Page 38.

TABLE 2 10R 11R IGSCC l WELD SI INSP ICSCC? INSP ICSCC? DISPOSIT10N 50 NE-1-42 NO YES NO YES NO NA 51 ND-1-3 NO YES NO YES NO NA 53 NE-2-104 NO YES NO YES NO -NA 54 NE-1-53 NO YES NO YES NO NA 55 NG-D-2 IHSI YES NO YES NO NA I 56 NE-5-8 NO YES NO YES NO 'NA l 57 NE-2-94 NO YES NO YES NO NA l 58 NG-B-2 IHSI YES NO YES NO NA '

59 NE-2-95 NO YES NO YES NO NA 60 NE-2-6 NO YES NO YES NO NA i 61 NG-D-4 IHSI YES NO YES NO NA 62 NU-2-2 NO YES NO YES NO NA 63 NG-D-5 IHSI YES NO YES NO NA 64 NE-1-26 NO YES NO YES NO NA 65 NG-E-24 IHSI YES NO YES NO NA 66 NE-1-43 NO YES NO YES NO NA 67 NG-E-22 IHSI YES NO YES NO NA 68 NG-A-24 IHSI YES NO YES NO NA I 69 NG-E-13 IHSI YES NO YES NO NA 70 NG-B-5 IHS! YES NO YES NO NA 71 NC-D-11 IHSI YES NO YES NO NA I 72 NE-1-9 NO YES NO YES NO NA ,

73 NG-E-6 IHSI YES N0 YES NO NA 74 NG-A-14 IHSI YES NO YES NO NA 75 NG-E-4 IHSI YES NO YES NO NA 76 NE-2-101 NO YES NO YES NO NA -

77 NE-1-66 NO YES NO YES NO NA 78 NE-1-36 NO YES No YES NO NA l 79 NG-3-2 IHSI YES NO YES NO NA 80 NE-2-17 NO YES SUS YES NA WO i 81 NE-1-29 NO YES YES YES NA WO l 82 NE-2-80 No YEd SUS YES NA WO i 83 NE-2-8 NO YES YES YES NA WO l

84 NE-2-98 NO YES YES YES NA WO i 85 NE-1-32 NO YES YES YES NA WO ,

86 NE-1-11 NO YES YES YES- NA WO I 87 s NE-1-46 NO YES YES YES NA WO 88 NE-1-20 NO YES SUS YES NA WO 89 NE-1-51 NO YES YES YES NA WO i I

90 NE-1-25 No YES YES YES NA WO 91 NE-2-4 NO YES SUS YES NA WO 92 NE-2-28 NO YES YES YES NA WO 93 NE-1-54A NO YES YES YES NA WO 94 NE-2-103 NO YES SUS YES NA WO 95 NE-1-13 N0 YES YES YES NA WO 96 NE-1-2 NO YES SUS YES NA WO 97 NE-2-91 NO YES SUS YES NA WO I

I l

1Tt - 050 ,

Tnly 3 Rev. 0 .

WELDS EXAMINED ON IGSCC AFFECTED SYSTEMS M*

• Inspectable 0313 10R 11R Total Welds Inside/Outside Inside/Outside 1st Ten Year Period Recirculation 84 31 67 (3) 11 Isolativa Condenser 184 15 125 (27) 12 58 (1) 10

Sh tdown Coollag , 14 --- --- --

6 1

4 Cican-up Demin 142 ---

10 10 i

2-Core Spray 26 --- --- -- 16 60 l

i TOTAL 450 46 135 79 90 87 j ( ) Indications of IGSCC i.

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i Ol8 3v/25

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TR - 050

.Rev. O Page 40

TABLE 4 NUMBER OF WELDS IN SCOPE OF NUREG-0313. REV. 2 BEFORE 13R -l Inspectable Welds

. Total Uninspectable Inside Outside Inside Outside System Welds Welds Drywell Drywell 2nd Valve 2ad Valve l Recire. 89 25 (Note'1) 64 0 64L 0 f I

RWCU 147 5 (Note 2) 30 112 46 96 i  !

CS 26 4 (Note 3) 22 0 22 0 SDC- 14 0 14 0 14 -0 ,

l IC 189 20 (Note 4) 40 129 40 129 i l

Closure Head 8 0 8 0 8 0 TOTAL 473 54 178 241 194 225 Note 1:

5 casting-to-casting welds, and 20 safe-end welds due to as-welded OD c4 adding (5 safe-end to pipe / fitting welds inspected from one side only).

Note 2:

Welds inside penetrations (will be CRC'd in 13R).

Note 3:

4 safe-end welds due to as-welded OD cladding.

Note 4 8 welds inside penetrations, 4 flued head-to-valve welds, 2 l casting-to-casting-welds, 2 saddle welds (will be replaced or clad when ICS l piping outside drywell is replaced). and 4 safe-end welds due to as-welded OD  ;

1 cladding.

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i 1

8 16641/30 l

TR - 050  :

Rev. O Page 41 TABLE 5 NUREG 0313 RECIRCULATION. SYSTEM QUANTITY OF WELDS NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R J

A 0 0 0 0 0 B 0 0 0 0 0 C 61 15 61 15 61 D 0 0 0 0 0  :

E 3 1 3 1 3  !

i 0 F 0 0 0 0 c 0 0 0 0 TOTAL 64 16 64 16 64 Man-Rem exposure for inspections for 13R = Approx. 13 Man-Rem Man-Rem exposure for inspections for 14R = Approx. 13 Man-Rem w/o decon and 7  !

Man-Rem w/decon,

] j Notes: )

l 1. Table above excludes the t'ive recirculation pump suction elbow-to-pump casing (casting-to-casting) welds and 20 safe-end welds due to as-welded OD cladding.

2. Assumes inspection frequency reduced by a factor of 2 due to HWC t implementation.

i J

3. Table excludes planned inspection and stress improvement of 20 $

recirculation system safe-ends. GPUN is planning on attempting to machine the OD cladding, inspect and stress improve the four C-loop  !

safe-end welds during 12R. If successful the 8 remaining vessel inlet '

safe-end welds will be inspected and stress improved in 13R (~ 89 ,

Man-Rem), and the 8 remaining vessel outlet safe-ends in 14R (~89 i Man-Rem w/o decon and ~45 Man-Rem w/decon).

16641/31 l 5

s

. TR - 050 Rev. O Page 42 l

TABLE 6 j GPUN PROGRAM RECIRCULATI0h SYSTEM QUANTITY OF WELDS l l

NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS l INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO CATECORIES 13R INSPECTIONS 14R INSPECTIONS 15R A 0 0 0 0 0 B 0 0 0 0 0 j C 61 6 61 6 61 I O O 0 0 0 0 I E 3 1 3 1 3 F 0 0 0 0 0 G 0 0 0 0 0 TOTAL 64 7 64 7 64 Man-Rem exposure for inspections for 13R = Approx. 6 Man-Rem Man-Rem exposure for inspections for 14R = Approx. 6 Man-Rem l Notes: j l

1. Table above excludes the five recirculation pump suction elbow-to-pump casing (casting-to-cascing) welds and the 20 safe-end welds due to as-welded OD cladding. I l

l

2. Assumes inspection f requency reduced by f actor of 2 due to HWC l implementation.  !

3. Table excludes planned inspection and stress improvement of 20 recirculation l system safe-ends. GPUN is planning on attempting to machine the OD l cladding, inspect and stress improve the four C-loop safe-end welds during l 12R. If successful the 8 remaining vessel inlet safe-end welds will be inspected and satress improved in 13R (~ 89 Man-Rem), and the 8 remaining vessel outlet safe-ends in 14R (- 89 Man-hem w/o decon and ~45 Man-Rem l w/decon). l I

16641/32 l

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. TR - 050

( Rev. O Page 43 TABLE 7 NUREG 0313 CORE SPRAY SYSTEM i

QUANTITY OF WELDS l

l NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO CATEGORIES 13R INSPECTIONS 14R ' INSPECTIONS ISR A 0 0 0 0 0 B 0 0 0 0 0 C 22 20 22 11 22 D 0 0- 0 0 0 E O O O O O F 0 0 0 0 0 G 0 0 0 0 0 TOTAL 22 20 22 11 22 Man-Rem exposure for inspections and weld crown reduction for 13R = Approx. 6 Man-Rem l

l Man-Rem exposure for inspections for 14R = Approx.1 Man-Rem l Notes:

1. Table excludes 4 safe-end welds due to as-welded OD cladding.

16641/33

n .- . - .

TR - 050  !

Rev. O Page 44 l

TABLE 8 GPUN PROGRAM '

CORE SPRAY SYSTEM QUANTITY OF WEL 0 NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS i INSPECTION. PRIOR TO 13R PRIOR TO 14R PRIOR TO CATECORIES 13R INSPECTIONS 14R INS PE'.:TIONS 1SR A 0 0 0 0 0 B 0 0 0 0 0 C 22 9* 22 5 22 D 0 0 0 0 0 E O O O 0- 0 F 0 0 0 0 0 3

G 0 0 0 0 0 TOTAL 22 9 22 5 22 )

3 ,

• All welds stress improved during the 11R refueling outage. This inspection ,

'i (9) will result in all welds having been inspected for IGSCC at let.at once. '

l Man-Rem exposure for inspections and weld crown reduction for 13R = Approx. 5 i Man-Rem t

i Man-Rem exposure for inspections for 14R = Approx. O.4 Man-Rem

)

Notest 1

l 1. Table excludes 4 safe-end welds due to as-welded OD cladding.

i

! 2. Table excludes planned inspection and stress improvement of the_4 safe-end welds in 12R. GPUN is planning on attempting to machine OD cladding, inspect and stress improve the four safe-end welds during 12R.

1 1

16641/34

' TR - 050 Rev. 0 .i Page 45 .

i I TABLE 9 .

l NUREG 0313 l

SHUTDOWN COOLING ,

i QUANTITY OF WELDS-  ;

NUREG TOTAL WELDS TOTAL WELDS' TOTAL WELDS _ l' INSPECTION- PRIOR TO 13R PRIOR TO 14R- -PRIOR TO CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R A 0 0 0 0 0 B 0 0 0. 0 0-4 i C 0 0 14 7 14 ,

i >

l D 9 9* 0 0 :0 ,

l E O O O O :0  ;

i ,

) F 0 0 0 0 0

{

' G 5* 0 5 0 0 -

i c i i 14*

TOTAL 14 14 7 14 l a

i

• During the 13R refueling outage 14 welds will be stress improved and -

1- post-process inspected.

] Man-Rem exposure for inspections for 13R = Approx. 23 Man-Rem l

Man-Rem exposure f7r inspections for 14R = Approx. 2 Man-Rem i l

4 s

1 16641/35 1

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TR - 050 Rev. 0  :

Page 46  ;

e TABLE 10 CPUN PROGRAM j SHUTDOWN COOLING QUANTITY OF WELDS j' NUREG TOTAL WELDS TOTAL WELDS 10TAL WELDS 1iSPECTION PRIOR TO 13R PRIOR.TO 14R PRIOR TO (

CATECORIES 13R INSPECTIONS 14R INSPECTIONS 15R  !

A 0 0 0 0 0 4

B 0 0 0 0 0 -

l C 0 0 14 1 14 i D 9 9* 0 0 0 E O O O O O t F 0 0 0 0 0  !

G 5 5* O O .O  ;

. TOTAL 14 14* 14 1 14

• During the 13R refueling outage 14 welds will be stress improved and post  ;

, process inspected. i l

Man-Rem exposure for inspections and stress improvement for 13R = Approx. 23 Man-Rem l

Man-Rem exposure for inspections for 14R = Approx. 1 Man-Rem i

i Note: Table takes credit for reducing inspection frequency by a factor of 2

for 14R due to HWC.

j i

)

l l 16641/36

TR - 050 Rev. O i Page 47 i I

I TABLE 11 '

N1JREG 0313 '

! REACTOR WATER CLEAN-UP SYSTEM ~

-QUANTITY OF WELDS  !

l N11 REG TOTAL WELDS TOTAL WELDS TOTAL WELDS i INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO i INSPECTIONS 15R  !

CATEGORIES 13R INSPECTIONS 14R A 0 0 4 0 4 l

[

l B 0 0 0 0 0 l t

C 0 0 0 0 0 I

D 29 7 138 35 138

+

E O O O O O l

l F 0 0 0 0 C l

C 113 113 0 0 0 l TOTAL 142 123* 142 35 142 I

• 4 welds will be replaced with ICSCC resistant material during 13R. i 3

Man-Rem exposure for inspections and weld crown reduction for 13R = Approx. 451 i w/o decon and 90 w/decon

{

Man-Rem exposure for inspect. ions for 14R = Approx. 28 w/o decon and 6 w/decon i I

Note: (1) Table above excludes 5 inaccessibic welds in drywell penetration (

which will be CRC'd with IGSCC resistant material in 13R.

(2) Table assumes inspection f requency reduced by a f actor of 2 due to j HWC implementation. l l

16641/37 i

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• 1 TR - 050 i Rev. O Page 48  !

TABLE 12

.GPUN PROGRAM REACTOR WATER CLEAN-UP SYSTEM QUANTITY OF WELDS NUREG TOTAL WELOS TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R A 9 0 4 0 4 B 0 0 0 0 0 l

C 0 0 0 0 0 D 26 3 29 7 42 ,

t E O O O O O F 0 0 0 0 0 G 20 7* 13 13** O TOTAL 66 10 46 20 46

• 4 welds will be replaced with ICGCC resistant materials during 13R.

l ** Assumes 13 welds outside drywell, but inside CIV in 14R with decon.

l Man-Rem exposure for inspections and weld crown reduction for 13R = Approx. 7 l Man-Rem I l  ;

l Man-rem exposure for inspections and weld crown reduction for 14R = Approx. '

l 109 w/o decon and 22 w/decon l l

Notes (1) Table above excludes 5 inaccessible welds in drywell pen dration which will be CRC'd with IGSCC resistant material in 13R.

(2) Table assumes inspection frequency reduced by a factor of 2 due to HWC implementation.

l (3) The 13 welds outside the drywell and inboard of the outermost containment isolation valve will be inspected when RWCU is deconned l (14R). '

(4) Table excludes 96 welds outside 2nd isolation valve (See section 4.5 for basis).

16641/38 l

l l

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\

f TR - 050 -

Rev. 0-Page 49 TABLE 13 NUREG 0313 ISOLATION CONDENSER SYSTEM INSIDE DRYWELL l

QUANTITY OF WELDS 1

i NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS '

INSPECTION PRIOR TO 13R PRIOk TO 14R PRIOR TO

! CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R ~ '

A 0 0 4 0 4 B 0 0 0 0 0 C 9 8 19 11 36 D 10 4* 17 17* O 1

E O O O O O l 1

F 0 0 0 0 0

)

G 21 19* O O O i i

l TOTAL 40 31 40 28 40

• During 13R, 10 welds will be stress improved 4 welds replaced with IGSCC resistant materials and 8 inaccessible penetration welds removed. During 14R, 17 welds will be stress improved.

Man-Rem 3xposure for inspections, weld crown reduction and stress improvement for 13R = approx. 77 Man-Rem.

Man-Rem exposure for inspections and stress improvement for 14R = approx. 42 Man-Rem w/and w/o decon.

Notes:

s

1. Table excludes 4 safe-end welds due to OD cladding, and 8 inaccessible welds in penetrations.

1 16641/39 l

TR - 050 Rev. O Page 50 TABLE 14 GPUN PROGRAM ISOLATION C0h".>ENSER SYSTEM INSIDE DRYWEL1, QUANTITY OF WELDS NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R A 0 0 4 0 4 B 0 0 0 0 0 C 9 0 19 0 36 D 22 0 17 6 0  :

E 0 0 0 0 0  !

F 0 0 0 0 0 i i

G 9 9 0 0 0 i TOTAL 40 14* 40 6* 40 1

• During 13R, 10 welds will be stress improved 48 welds replaced with ICSCC resistant materials and inaccessible penetration welds removed. During 14R.  ;

17 welds will be stress improved. l 1

Man-rem exposure for inspections, weld crown reduction and stress improvement  !

f or 13R = approx. 38 man-rem. i Man-rem exposure for inspection . weld crown reduction and stress improvement for 14R = approx. 39 man-rem.

Notes:

1. Table takes credit for reducing inspection frequency (for condensate side piping) by a f actor of 2 for 14R due to HWC.

2: Table excludes 4 safe end welds due to OD cladding, and inaccessible welds in penetrations.

16641/40

TR - 050 Rev. O' Page 51 TABLE 15 NUREG 0313 ISOLATION CONDENSER SYSTEM OUTSIDE DRYWELL QUANTITY OF WELDS NUREG TOTAL WELDS TOTAL WELDS ~ TOTAL WELDS INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO CATEGORIES 13R -INSPECTIONS 14R INSPECTIONS 15R A 0 0 41 2 41 B 0 0 0 0 0 0 0 0 0 0 0 D 51 22 44 22 .44 E 19 5 10 5 10 F 0 0 0 0 0 C 59 32 0 0 0 TOTAL 129 59* 95 29 95 All condensate piping and the portion of the steam piping on 75' elev. up to the first weld on 95' elev. will be replaced with IGSCC resistant material during 13R.

13R exposure for weld crown reduction and inspection is 10 Man-Rem and for 14R is 2 Man-Rem.

NOTE: Table excludes 2 uninspectable saddle welds, 4 flued head-to-valve welds and 2 valve-to-valve welds. These welds will be replaced or CRC'd during 13R with resistant materials. j 16641/41

e

'TR - 050 Rev. O Page 52 L

i P

TABLE 16 GPUN PROGRAM ISOLATION CONDENSER SYSTEM OUTSIDE DRYWELL QUANTITY OF WELDS NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO

, CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R A 0 0 41 2 41 B 0 0 0 0 0 0 0 0 0 0 0 D 97 4 44 4 44 E 19 1 10 1 10 F 0 0 0 0 0 G 13 5 0 0 0 >

4 TOTAL 129 10* 95 7 95 i

• All condensate piping and the portion of the steam piping on 75' elev. up to the first weld on 95' elev. will be replaced with IGSCC resistant material during 13R.

i Man-Rem exposures for inspections and weld crown reduction for 13R = approx. 2 l Man-Rem and for 14R = approx. 1 Man-Rem.

NOTES: Table excludes 2 uninspectable saddle welds. 4 flued head-to-valve ,

welds and 2 valve-to-valve welds. These welds will be replaced or CRC'd l during 13R with resistant materials. I 1

, l 1

l 16641/42 a

, .-.m.. . . . _ _ _ , - . - - ,_ .___- _,_ . . , _ . , . . _ , _ - - - . - - _ . _-m-m .- _.--._.m ~ . , _ - y_m - . _ , , , _ . , . - . - . ,

. TR - 050 Rev. O '

Page 53~

i TABLE 17 NUREG 0313 & GPUN PROGRAM

' CLOSURE HEAD WELDS ,

4 QUANTITY OF WELDS i

NUREG TOTAL WELDS TOTAL WELDS TOTAL WELDS

! INSPECTION PRIOR TO 13R PRIOR TO 14R PRIOR TO '

CATEGORIES 13R INSPECTIONS 14R INSPECTIONS 15R A 0 0 8 0- 8 B 0 0 0 0 0  ;

C 0 0 0 0 0 D 2 0 0 0 0 E 0 0 0 0 0 F 0 0 0 0 0 G 6 . 0 0 0 0 1 TOTAL 8 0* 8 0 8 i

l

• All welds will be repaired / replaced with ICSCC resistant materials-during 13R.

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16641/43 1J a

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

i TR - 050 Rev. O Page 54

, TABLE 18 INSPECTION SCHEDULES FOR BWR PIPING WELDMENTS DESCRIPTION ICSCC INSPECTION CPUN PROPOSED OF WELDMENTS NOTES CATECORY EXTFNT/ SCHEDULE EXTENT / SCHEDULE Resistant Materials A 25% every 10 years SAME (at least 12% in 6 years)

Nonresistant Matls. (1) B 50% every 10 years SAME SI within 2 years (at least 25% in 6 of operation (1) years)

Nouresistant Matls. (1) C All next 2 re- 50% every 10 years (HWC),

i SI after 2 years fueling cycles then: 100% every 10 yrs. other-of Operation (1) All every 10 yrs wise, Note 1 does not (at least 50% in 6 yrs apply. For welds 12 i inches in diameter or less.

Nonresistant Matl. (1) D All every 2 refuel- All every 5 refueling ing cycles cycles (HWC)

Cracked (1) (2) E 50% next outage then: SAME overlayed or All every 2 refuel-SI ing cycles Cracked (2) F All every outage SAME Inadequate No Repair Nonresistant (3) G All next outage All by end of 13R outage except recirculation system safe-ends.

Not Inspected 1664i/45 J

TR - 050 *-

Rev. 0 /

Pa:e 55 .

P t TABLE 9 (Cont'd) ,

Notes:

1) All welds of non-resistant material should be inspected after a stress improvement process'as part of the process. Schedules shown should be followed after this initial inspection.

(2) See requirements for acceptable weld overlay reinforcements and stress improvement mitigation.

(3) Welds that are not UT inspectable should be replaced; "sleeved," or local leak detection applied. RT examination or visual excmination for leaks may also be considered.

16641/46

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