ML20081J251
| ML20081J251 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/11/1991 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20081J245 | List: |
| References | |
| GL-88-01, GL-88-1, NUDOCS 9106210194 | |
| Download: ML20081J251 (46) | |
Text
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37 G.
Primary Coolant System Pressure Isolation Valves Specificati2D8
- 1. -Periodic leakage testing (a) on each valve listed in Table 4.3.1 shall be accomplished prior to exceeding 600 peig reactor pressure every time the plant is placed in the cold shutdown condition for refueling, each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, whenever the valve is moved whether by manual actuation or due to flow conditions, and after returning the valve to service after maintenance, repair or replacement work is performed.
H.
~ Reactor Coolant System Leakace 1.
Unidentified leakage rate shall be calculated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2.
Total leakage rate (identified and unidentified) shall be calculated at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
3.
A channel calibration of the primary containment sump 1
flow !;.tegrator and the primary containment equipment drain tank flow integrator shall be conducted at least once per 18 months.
I.
An inservice inspection program for piping identified in' NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in the generic letter or in accordance with alternate measures approved by the NRC staff.
Basen Data-is available relating neutron fluence (E>l.0MeV)_and'the change in the
- Reference Nil-Ductility Transition-Temperature (RTNDT).
The pressure-temperature (P-T). operating curves.(a),(b) and-(c) in Figure 3.3.1 were developed based on the results of testing and evaluation of specimens removed from the vessel after 8.38 EFPY of operation.
Similar testing and-analysis will be performed throughout vessel life to monitor the effects of neutron irradiation on the reactor vessel shell materials.
The_ inspection program will reveal problem areas should they occur, before a leakzdevelops.
In addition, extensive visual inspection for leaks will be made on critical systems. Oyster-Creek was designed and constructed prior to (at To satisf7LALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrat-t ing valve compliance with the leakage criteria.
- -NRC Order dated April 20, 1981.
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i OYSTER CREEK 4.3-2 Amendment No.
82,90,97,118,120, 151 9106210194 910611 PDR ADOCK 05000219 P
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1 ATTACHMENT 2
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TOPIC /1 REPORT i
No. 050' REVISION 3 i
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GPUN RESPONSE TO GENERIC LETTER 88-01 AND NUREG 0313, REV. 2 TR - 050 REV. 3 AUTHORS l? !"
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G. SLEkRV DIRECTOR, ENGINEERING & DESIGN 40A Date 2-Id-9/
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01J-0012.1 l
DOCUMENT NO.
[c r.,'J] Nuclear TR-05U REV. /3 TTTLE GPUN RESPONSE TO GENERIC LETTER 88-01 AND NUREG 0313 - REV. 2 REV
SUMMARY
OF CHANGE APPROVAL DATE 2
Provides (1) RWCU inspection plan for welds C.
Che 9 -6 /C outside the second containment isolation NVho R.
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l.orenzo valve, (2) sample expansion criteria for
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Laggart AI/. if.d,- fo system safe-ends, RWCU outside the second D. Covill containment isolation valve and system weld
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categories, (3) update of 12R inspections and bbb S'M repairs, (4) revised 13R inspection plan, and (5) revised 13R piping replacement and stress improvement plan. The content has been revised entirely.
3 (1) Defer stress improvement plan from 13R to
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14R.
l1 L. Lorenzo f// N/kd' (2) Defer closure head welds repair / replacement i. W.
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plans from 13R to 14R.
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TR - 050 Rev. 3 Page 2 of 43 TABLE OF COKTENTS PAGE 1.0L INTRODtICTION 4
1.1 ~ Purpose 5
1.2 Summary of Oyster Creek IGSCC Inspections 5
1.2.1 10R and 11R Refueling Outages-Inspections / Repairs 5
1.2.2 12R Refueling Outages - Inspections / Repairs 5
'1.3 Summary of Proposed 13R Inspection Plan 6
1.4 Summary of Planned IGSCC Mitigating Actions 6
2.0
SUMMARY
OF SPECIFIC PIPING SYSTEMS 7
2.1 Recirculation System 7
2.1.1 System Description, Materials.and operations.
7 2.1.2 Background / History 7
2.1.3 12R Scope-of Work 8
2.1.4 Future _ Improvement / Inspection Program 9
'2.1.4.1 Future Improvements' 9
2.1.4.2 13R and Future Inspection Program 9
2.2 Core Spray System-9 2.2.1 Syste.v Description, Materials and operations 9
2.2.2 Background / History 9
2.2.3 12R Scope of Work 10 2.2.4 Future Improvement / Inspection Program 10 2.3 Shutdown Cooling System 10 2.3.1-System Description, Materials and Operations 10 2.3.2 Background / History 10 2.3.3 - 12R Scope of Work 10 2.3.4 Future Inspection Program 10 2.4 Reactor Water Clean-Up (RWCU)1 System' 11 2.4.1 System Description, Materials and operations 11 2.4.2 Background / History 11 2.4.3 12R Scope of Work-11 2.4.4 Future Improvement / Inspection Program 12 2.4.4.1 Future Improvements 12 2.4.4.2?
13R and Future Inspection Program 12 2.5 Isolation condenser System 12 2.5.1 - System Description,: Materials and Operation 12-2.5.2 Inside Containment 12 2.5.2.1 Background / History.
12 2.5.2.2 12R Scope of Work 12 2.5.2.3 Future Improvement / Inspection Program 13 2.5.2.3.1
'13R and Future Improvements 13 2.5.2.3.2 Future Inspection Program 13 013-0012.3 a
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2.5.3 Outside Containment 13 2.5.3.1 Background / History 13 2.5.3.2 12R Scope of Work 13 2.5.3.3 Future Inspection Program' 14 2.6 Closure Head Piping Welds 14 2.6.1 Description, Materials and Operation 14 2.6.2 12R Scope of Work 14 2.6.3 Future Improvement / Inspection Program 15 3.0 WATER CHEMISTRY CONTROL AT OYSTER CREEK 15 3.1 GPUN Actions Taken 15 3.2 Water Chemistry Effect Data from other Utilities 16 4.0' GPUN TECHNICAL CLARIFICATIONS TO GL 88-01 17 4.1 10R and 11R Inspections 17 4.2 Post Stress Improvement Inspection 17 4.3 Inspection of Cast Materials 18 4.4 Implementation of Hydrogen Water Chemistry 19 4.5 RWCU Welds Outboard of the Second Insolation Valve 19 5.0 NUREG 0313, REV. 2 20 5.1 NUREG 0313 Scope 20 5.2 GPUN Proposed Program 21 5.3 Sample Expansion 21 6.0 OTHER GENERIC LETTER 88-01 RESPONSES REQUIRED 23 7.0
SUMMARY
25 l
8.0 REFERENCES
26 9.0 FIGURES 27 10.0 TABLES 32 i
013-0012.4
TR - 050 Rev. 3 Page 4 of 43
1.0 INTRODUCTION
1.1 Puroose GPU Nuclear (GPUN) is required to perform inspections of reactor coolant systems (RCS) piping to detect intergranular stress corrosion cracking (IGSCC).
For tha 12R refueling outage, GPUN planned to meet the intent of Generic Letter 84-11[1], which was the guideline for such inspections until recently.
In January 1988, the Nuclear Regulatory Commission (NRC) issued Nuclear Regulatory Guide (NUREG) 0313 Rev. 2(2), and this was supplemented by Generic Letter (GL) 88-01[3].
In accordance with the generic letter, GPUN's response was submitted on August 12,.1988, out-lining future inspection program plans including the 12R refueling outage.
In response to our 12R plans the NRC issued a letter (9) which took exceptions to our plan.
Briefly, these exceptions included:
(1) the exclusion of the kWCU piping welds outside of the second containment isolation valves; (2) taking credit for 10R inspections unless the examiners who took the requalification exame had passed on the first attempt; (3) the sample expansion criteria (unless further technical justification is provided) for the recirculation system safe-ends, isolation condenser piping outside the'second containment isolation valves and the RWCU piping outside the second containment isolation valves; and (4) the number of Category G welds remaining after 13R.
Subsequent to the receipt of the NRC letter (9), GPUN initiated two telecons (10, 11) to clarify the NRC requirements for the 12R and 13R inspections.
Based on these telecons, GPUN committed to revise its inspection plan (13), address the remainder of the
. category "G" welds by the end of 13R (13), and address the RWCU piping welds outside the second containment isolation valve six months after restart from the 12R outage (14).
On April 17, 1990, NRC issued a safety evaluation of the above mentioned GPUN's submittal and requested GPUN to submit revised inspection plans (18). _GPUN submitted this revised inspection plan on October 18, 1990-(20).
As a result of recent GPUN Management Meetings, the schedule for accomplishing the IGSCC improvement program has.been changed.
This document will delineate GPUN's revised IGSCC improvement program which includes (a) the revised IGSCC inspection plan for 13R and (b) the revised schedule of mitigating actions to minimize the possibility of IGSCC.
l-013-0012.5
TR - 050 Rev. 3 Page 5 of 43 1.2 Summaty_qf ovator Creek IGSCC Inspections 1.2.1 10R and 11R Refuelino Outages - Insnections/RenalIA In 1983 (10R), 31 Recirculation System welds were inspected to IEB 82-03[4), including three unids that were fluorescent dye penetrant inspected on the ID and dispositioned as geometry.
No indications of IGSCC were detected (5).
In 1984 (10R), a leak was detected from an Isolation Condenser Condensate return line weld during a pressure test of the condenser tubes.
This leak resulted in the inspection of over 150 welds in the Isolation condenser System (ICS) and the RWCU System.
All the inspectable steam and condensate Isolation Condenser welds outside the drywell (127) were inspected. Twenty-seven welds in the ICS piping outside the drywell were found to contain indications of IGSCC.
Nint sere replaced thru spool piece change out and elghteen were repaired with full structural weld overlays (6).
Three welds were destruc-tively examined, and the failure mechanism was concluded to be intergranular stress corrosion cracking.
In the 11R refueling outage, inspections of Reactor Coolant System (RCS) welds were performed following the guidelines of Generic Letter 84-11.
One hundred and sixty-nine (169) butt welds which included all eighteen (18) ICS weld overlays deposited in 10R were inspected.
Three welds in the C-loop of the Recirculation System and one weld in the Isolation Condenser System (ICS) steam line outside the Drywell bad 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).
The initial inspection sample as well as the increased sample inspection welds were chosen based upon the previous inspections and the difficulties associated with determining their final disposition as non-IGSCC.
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.
1.2.2 12R Refuelino Outaces - Insocetions/Repairg During the 12R refueling outage, 156 butt welds within the scope of Generic Letter 88-01 in the Recirculation, Shut-down Cooling, Reactor Water Cleanup, Core Spray, and Isolation Condenser systems, closure head piping, and six structural weld overlays (two in the Recirculation system and four in the Isolation Condenser system outside the drywell) were ultrasonically examined for IGSCC.
Six butt 013-0012.6
TR - 050 Rev. 3 Page 6 of 43
- welds contained UT indications with the characteristics of IGSCC.
Five (two Recirculation system and three in the J
Isolation condenser system outside the drywell) required wold overlay repair; one Recirculation system weld was analyzed and found to be acceptable for continued opera-tion without repair.
The six inspected overlays contained no indications of IGSCC in either the overlays or the outer 25% of the original pipe wall. Additionally, two welds (one 8" Core Spray, one 2" Reactor Head Cooling
[RHC)) were found to be leaking during the hydro test.
The core spray weld was overlay repaired; the head weld was cut off and re-welded. The RHC defect was destructively evaluated and found to be IGSCC.
1.3 Summary of Proposed 13R IneDection Plan During 13R refueling outage, eighty-four (84) pipe welds plus eight (8) safe-end welds will be inspected. Additionally, eighty l
(80) replacement welds (Category "A") will be UT baseline inspected.
These welds are listed in Table 2 through Table 14.
1.4 Summarv of Planned IGSCC Miticatino Actions 1.4.1 Implement Hydrogen Water Chemistry (HWC) during Cycle 12 (1990). As a result of implementing HWC, a reduced inspection frequency may be requested in the future, based on industry experience and BWR Owners Group's recommendations [19).
1.4.2 Replace the following pipes with IGSCC resistant material during the ~3R refueling outage (1991).
a.
all Isolation Condenser large boro piping outside the drywell (from the drywell penetrations to the iso-lation condensers)-that has been susceptible to IGSCC.
This will reduce the number of welds requiring inspection by sixty-seven (67).
In addition, the six outside containment isolation valves will be replaced.- All the new welds will be stress improved during 14R.
b.
all piping within the four (4) isolation condenser p
drywell penetrations and the two (2) Reactor Water Clean-up system drywell penetrations which contain welds that are not inspectable. This will eliminate thirteen (13) presently uninspectable welds.
1.4.3 Replace the Head Cooling Spray Nozzlw Assembly, the 4" tee and the flange of the reactor vent line with IGSCC resistant
(
material during 14R.
1.4.4 Stress Improve (SI) all accessible /inspectable welds inside the drywell (except Reactor Water Clean-Up System) by the end of the 14R refueling outage (1992).
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Rev. 3 Page 7 of 43 2,0
SUMMARY
OF SPECIFIC PIPIf10 SYSTEMS 2.1 Recirculation System 2.1.1 System Description. Materials and coerati2n1 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 in operation.
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 annulus and discharges to the lower vessel area containing the diffuser.
2.1.2 Backaround/ History IGSCC in this system was described earlier in Section 1.2.1.
During the llR refueling outage of 1986, 64 of the system's 89 welds were stress improved (SI'd) with the induction heating stress improvement (IHSI) method.
All 64 welds were inspected following stress improvement.
Among the 64 welds inspected, three (2) were found to have indications of IGSCC.
Two of the IGSCC affected welds were overlay repaired and one remained in the use-as-la condition.
Uninspected and non-stress improved welds include 20 safe-end welds and 5 uninspectable casting-to-casting welds.
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 carbon, high-ferrite" (as stated in the repair specifications) Type 308L weld metal.
However, since the safe-end to nozzle shop weld was performed with Inconel 182 wold metal, a portion of the ID is covered with Inconel 182 (see Figure 1).
Since we cannot locate the weld material chemistry test reports nor the inspection reports of the reconded, 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 Inconel 182, we consider it prudent to stress improve the nozzle to safe-ends welds.
Approximately 23 man-rem of exposure per safe-end will be required to perform these tasks.
013-0012.8
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TR - 050 Rev. 3 Page 8 of 43 2.1.3 12R - Scope of Work During 12R, 10% of the 61 previously otress Lmproved (SI'd) welds plus the three welds containing IGSCC were inspected.
The initial sample of 6 category "C" welds inspected in 12R had 2 with indications of IGSCC (NG-D-11 and NG-D-21).
As a result of this finding, an additional 6 welds required inspection.
In this additional sample, one weld (NG-D-18) had indications of IGSCC.
The methodology utilized to determine the initial welds as well as the first sample expansion was based upon previous inspections and the difficulties associated with determinfig their final dis-position.
Therefore, the remaining 49 category "C" welds were inspected during 12R.
No further IGSCC indications were detected. All 61 Category C welds were reinspected in 12R.
Also during 12R, both "C"
loop recirculation safe-ends were stress improved and inspected. The effort required to perform this included machining the CD cladding to provide a surface finish and contour adequate for performing UT for IGSCC.
These welds were then post process UT examined.
The inspection area included the nozzle to safe-end weld (the safe-end side of that weld, and the nozzle side of that weld for a distance of IT into the nozzle from the weld centerline) and the safe-end-to-pipe weld.
These welds contained no indications of IGSCC.
i l
The technical basis for selecting these two safe-ends was l
that during the llR outage, GPUN stress improved 64 Re-circulation System welds and inspected all of them after SI.
The only loop to contain welds with indications of IGSCC were three in the C-loop.
During the safe-end overlay cladding effort before operation, described above, l_
access to the vessel was provided by removing the elbow at tha 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 llR.
We consider that these circumstances provided sufficient concern to warrant stress improving these two safe-ends first.
Treating these two safe-ends during 12R has provided much needed, useful information for stress improving, 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 l
more efficiently perform this work in the future. We l
expect that future time and exposure savings will be l
substantial.
Since no indications of IGSCC were found 013-0012.9
TR - 050 Rev. 3 Page 9 of 43 in the C-loop safe-ends in 12R, we consider that these results coupled with the partial implementation of HWC in Cycle 12 provides adequate assurance that cracks will not initiate / grow in the other eight safe-ends.
During 12R, 3 D-Loop welds were dispositioned as containing indications of ICSCC (NG-D-11, 18, and 21).
A plug was removed from NG-D-11 during 12R for destructive evaluation.
Results of this analysis revealed that the crack tip was blunted and probably existed prior to IHSI in 11R [15).
Welds NG-D-11 and NG-D-21 were overlay repaired and NG-D-18 was left as stress improved.
2.1.4 Future Imorovement/Insoectiqn Procram 2.1.4.1 Future Imorovements - The planned mitigating actions for the recirculation system welds include implementing HWC following 12R, stress improvement, and post process inspection of the safe-ends in A, B,
D, and E loops of the recirculation system during 14R.
f 2.1.4.2 13R and Future Insoection Procram - The welds of recirculation system will be inspected in accordance with GL88-01 and Tables 2 and 5 of this report.
2.2 Core Sorav System 2.2.1 System Description. Materiale 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 ust r 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 limited.
The system external to the reactor vessel consists of 6 inch and 8 inch' diameter Type 316 stainless steel pipe. Having two redundant loops, the system enters the reactor vessel through two safe-ends attached to two separate nozzles. The system has a total of 27 potentially IGSCC susceptible welds in the scope of NUREG 0313 Rev.2.
2.2.2 Backaround/ History During the 10R outage, 3 Core Spray System wolds were inspected for IGSCC.
In 11R, 16 welds were inspected, of which 2 had been inspected in 10R.
No indications of IGSCC were detected.
013-0012.10 1
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-e%ei y-x TR - 050 Rev. 3-Page 10 of 43 2.2.3 12R Scope of Work During the 12R refueling outage, the system's safe-end welds and 19 of the 21 IGSCC susceptible piping butt welds were stress improved.
Initially,-11 core spray welds were inspected.
This number included the 6 safe-end welds associated with the_ systems 2 safe-ends.
Within this initial group of welds, there were no indications of IGSCC.
Not all stress improved welds received a post-process inspection, but it was assessed that a sufficient sample were post-process inspected (See section 4.2 for Technical Basis).
However, all safe-end welds were in-spected after stress improvement. Subsequently, during a pre-operational hydrostatic pressure tust, a leak was detected from core spray weld NZ-3-38.
Subsequene exami-nations identified the presence of indications of IGSCC contained in this weld.
This weld, NZ-3-38, was one of the 19 welds which were stress improved during 12R.
It was determined that NZ-3-38 was examined during 11R, but did not receive a post stress improvement inspection during 12R.
As a result of this event, an expanded sample group of eight (8) additional welds (NZ-3-39, 40, 42, 81, 84, 88, 89, & 90) were examined.
Within this group seven
-(7) were stress improved during 12R.
Within this population, two'(2) welds (NZ-3-40, 42) were pre-process inspected in llR, however'none were post-process inspected during 12R until this event. There were no indications of IGSCC detected within the expanded population.
NZ-3-38 was weld overlay repaired and returned to service.
2.2.4 Future Improvement / Inspection Procram 25 of 27. welds within the scope of NUREG 0313 were stress improved during 12R.
Stress iniproved but not previously inspected (11R or 12R) welds will be inspected during the 13R outage.
The system's safe-ends and pipe welds will be inspected in accordance with GL88-01 and Tables 3 and 7 of this report.
The two remaining welds will be stress improved in 14R.
2.3
-Shutdown Coolina System 2.3.1 System Description, Materials and coerations The Shutdown Cooling System is utilized intermittently, only during plant shutdowns where cooldown to below 212*F is required. However, initiation may be at temperatures as elevated as 350*F for very short periods of time'.
The system is composed 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 1
2nd Containment Isolation Valve (CIV).
013-0012.11
7, TR --050 Rev. 3 Page 11 of 43 2.3.2 Backcround/ History This system has shown no indications of IGSCC thus far.
Two of the welds in Shutdown Cooling were inspected in 10R.
During 11R, 6 welds were examined, of which 2 were examined in 10R.
The system contains a total of 14 susceptible welds.
2.3.3 12R Scope of Work Three welds in Shutdown Cooling were inspected during 12R with no indications of IGSCC.
In addition, HWC, which was 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 Future Imorovement/Insoection Procrams All the welds will be inspected in accordance with GL88-01 and Table 6 of this report. During 14R, the 14 welds will be stress improved and post process inspected.
2.4 Egaetor Water Clean-Up (PWCU) System 2.4.1 System Descriotion, 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 of the first non-regenerative heat exchanger and from the outlet (return) of the third regenerative heat exchanger, reactor coolant is above 200'F.
2.4.2 Hackaround/ History During the 10R refueling outage, five (5) welds were inspected in the RWCU System. A total of 10 welds inside containment were inspected during 11R of which 2 had been inspected-in the 10R refueling outage.
No IGSCC has been detected in the system to date.
i i
2.4.3 12R Scoce of Work l
10 RWCU System welds inside containment were inspected during 12R with no indications of IGSCC.
In addition, HWC, which was implemented following the 12R refueling outage, will be a significant IGSCC mitigator.
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Page 12 of 43 2.4.4 Future Improvement / Inspection Procram 2.4.4.1 Future Imorovements - Since RWCU is operating during plant power operation, the system wi)1 derive full benefit from HWC.
In addition, the piping that is inside containment penetrations (unaccessible for UT inspection) will be replaced with IGSCC resistant material in 13R.
The new piping will contain no butt welds inside the penetrations.
2.4.4.2 13R and Future IneDection Procram - All the RWCU welds located inside the second isolation valves will be inspected in accordance with GL88-01 and Table 8.
10% of the RWCU welds outboard of the second isolation valves will be inspected during 13R [18, 19),
2.5 leolation condenser svetem 2.5.1 System Description, Materials and Ooeration 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 iso-lation of the reactor from the main condenser.
The system prevents overheating of the reactor fuel, controls the re-actor 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 experienced numerous initiations in the years since initial start up of Oyster Creek.
2.5.2 Inside C:.itainment 2.5.2.1 Backaround / History - No IGSCC has been detected in this portion of the ICS.
During 10R, 18 welds had been inspected.
In 11R, 12 welds were in-spected of which 6 had been examined during 10R.
2.5.2.2 12R Scope of Work - During 12R, a total of 14 welds which include the 2 ICS safe-ends (a total of 4 welds) and 10 piping welds were inspected with no indications of IGSCC.
Stress improvement was performed on the 2 ICS safe-ends (4 welds) and on 9 additional steam side welds.
Not all 013-0012.13
..m m
.m_
._m TR - 050 Rev. 3 Page 13 of 43 stress improved welds received a post process inspection but a sufficient. sample were post-process inspected (see Section 4.2 for Technical-Basis). However, all eafe-end welds were in-spected after stress improvement. In addition, HWC, which was implemented following the 12R refueling outage, will be a significant ICSCC mitigator for the condensate piping up to the second (normally closed) valve.
2.5.2.3 Future Improvement / Inspection Procram 2.5.2.3.1 Future Imorovements - During 13R, the piping within the four (4) ICS containment penetrations will be replaced.with resistant material with no butt welds inside the penetrations.
During 14R, all of the welds not yet stress improved will be treated.
2.5.2.3.2 13R and Future Inspection Procram -
All the welds will be inspected in accordance with GL88-01 and Table 9.
2.5.3 outside containment 2.5.3.1 Backoround/ History - In 10R, 127 ICS welds outside of containment were examined.
In this case, a large cample_was inspected because of detecting a through-wall leak in an 8" dia.
condensate return.line weld.. As a result of this effort, 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 diagnosed during 10R as having a L
root geometry indication.
Seventeen (17) more welds that were installed as a result of spool piece repairs in-10R were not included in the 11R sampling base.
2.5.3.2 12R Scope of Work - During 12R, GPUN inspected 37 welds in ICS outside containment. Within this sample, there are 4 overlayed welds (one i
was overlayed in 11R), 23 welds which were not inspected during 11R, and 10 welds inspected during 11R.
This inspection resulted in 3 welds requiring weld overlay repair.
Two were not
-013-0012.14 i
l
TR - 050 Rev. 3 Page 14 of 43 inspected during llR and one was inspected during 11R.
The sample expansion, as a result of the above indications, was in accordance with NUREG 0313, Rev.2 for the ICS Piping outside the Drywell.
2.5.3.3 13R and Future Imorovement /Tnspection Proorgm -
All the piping outside the drywell and inside the penetrations will be replaced in 13R with IGSCC resistant material. These welds will be inspected in accordance with GL88-01 and Table 12 of this report. All new welds outside second isolstion valves will be stress improved (where possible) and post process inspected during 14R.
2.6 Glosure Head Pipino Welds 2.6.1 Egserietion, Materials, and ooeratione There are three alloy steel nozzles with Type 316 stainless steel weld neck flanges welded to them on the reactor vessel closure head.
The original flanges were subjected to the closure head's final post weld heat treatment.
Thereby, the flanges were furnace sensitizad.
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" nozrle is a spare to which a blind flange le bolted. The other 6" nozzle is the inlet for the head spray line. The 4" nozzle is for vent piping.
There are four (4) other butt welds connected to the above nozzles which are susceptible to IGSCC.
Therefore, the closure head contains seven (7) welds within the boundary of NUREG 0313 Rev. 2.
2.6.2 12R Scoce of Work During the 12R refueling outage 2 of the closure head nozzle-to-flange welds were inspected with no indications l
of IGSCC.
However, subsequent to these examinations a leak was observed adjacent to a 2 inch butt weld within the head spray nozzle 6 x 2 inch reducer during the hydro test.
As a result of this observation a ring sample was removed con-l taining the defect and destructively analyzed.
The results l
of the analysis characterized the crack to be ICSCC [16).
l An expanded examination was conducted where two four-inch l
and one two-inch butt welds within the vent piping were L
ultrasonically examined to determine the presence of IGSCC.
The results of these examinations showed no indications of IGSCC were present. The head spray nozzle weld was repaired and returned to service.
013-0012.15
J TH - 050 Rev. 3 Page 15 of 43 2.6.3 D1JtitLyuture Inst,oetion ProantrD All the applicable welds will be inspected in accordance with Cenoric Letter 88-01 and Table 10.
The four piping welds will be replaced with ICSCC rosistant material (Refernneo Pars. 1.4.3).
3.0 NATXg_qlinillTRY CONTROL-_AI_QXJLTULQiggE 3.1 GPUN_Aglions Taken The fellowing actions have been or will be taken to mitigate the iritiation and propagation of 10 SCC at oyster Creeks Implemented EPRI water chemiscry guidelines (1984).
Comn.ence hydrogen water chemistry (ifWC) during Cycle 12 (1990).
Established a new chemistry laboratory with state-of-the-art equipment for analyses (1985).
Plugged a 11rge number of leaking condenser tubes during 10R.
During Cycle 11, no tube leaks were evident.
Replaced resin in the reactor water cleanup domineralizer at the onset of silica leakage.
Perform air in-leakage surveys.
Replace resin in condensate polishers before onset of significant ionic leakage (1985).
Our efforta to improve waier chemistry have resulted in substantial improvements over the last two operating cycles. Reactor Water conductivity during C,cle 10, Cycle 11, and Cycle 12 to date has bee.1 about 0.1 uS/cm.
We consider that' this improved conduc' ivity has substantially reduced the potential for new crack initiation and has slowed the growth of existing, undetected cracks, if any.
The addition of HWC in Cycle 12 will further reduce the potential for new cracks.
Pipe tests sponsored by EPRI & CE_ indicated a factor of improvement of 20 with respect to crack initiation following the introduction of appropriate additions of hydrogen.
We also consider that the lack of IGSCC in the shroud head bolts (creviced alloy 600) is an indication of adequate water chemistry performance in the past.
In other BWR's, many of these bolts, identical in design to Oyster Creek's, were found to contain crevice-induced IGSCC, including some that were 100% cracked.
The cause of this l
013-0012.16 i.
TR - OSO Rev. 3 Page 16 of 43 cracking has been attributed to the presence of a crevice beneath the collar, and water conductivity.
In the llR outage, all 36 bolts were UT insgmeted at Oyster Creek.
No indications were detected.
Again, the improvements in water chemistry control noted earlier are expected to provide significant improvement in reducing the potential for new crack development.
3.2 Water Chemistry Effect 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 on materials performance when good chemistry practices are followed.
Ti.e Dresden-2 data has been obtained while the plant was injecting hydrogen (HWC) and the PD-3 data was obtained under HWC conditions.
Figures 2 and 3 show the average crack growth ra es for sensitired Type 304 specimens loaded to about 27 Kei -
in.
Figure 2 data for PD-3 clearly shows the improvement in crack growth when water conductivity is reduced.
There is a factor of 4 i
improvement (from 96 to 24 mils /yr.) when the averA23 conductivity is reduced from 0.5 to 0.2 uS/cm.
The Figure 3 PG-3 data shows the l
impact of two resin intrusions on growth while the normal chemistry conductivity had been maintained at about 0.2 uS/cm.
Even with the two resin intrusions within a short period of time apart (about 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 further upsets, the crack growth rate would return to its rate before the intrusions. These data were obtained f rom CAV specimens in autoclaves connected to the plant's RCS.
Similar results have been obtained in the laboratory.
The crack growth rate for sensitired Type 304 at about 0.1 uS/cm conductivity was about 29 mpy while at about 0.45 uS/cm, the rate was about 240 mpy (Figure 4).
"hese growth rates compare very favorably with the NRC-calculated rato of 390 MPY (Appendix A.2 of NUREG 0313 Rev. 2).
Figure 2's Dresden-2 data shows the impact of implementinq.isc with low conductivity (about 0.1 us/cm).
The Figure 4 data shows similar results in laboratory conditions.
We consider that the low average conductivity obtainod at Oyster Creek over the last few cycles has significantly contributed to a reduced propensity for developing new cracks and a slow crack growth rate 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 during cycle 12 (1990), will furtner improve the condition of affected stainless steel piping.
013-0012.17
0 4
TR - 050 Rev. 3 Page 17 of 43 4.0 CPUN TECHNICAL CLARIFICATIONS TO OL81-9J 4.1 lQE_and 11R Inspections The NRC staff did not concur with GPUN's position for taking credit for the IOR outage examinations, because these examinations were performed prior to September 1985.
The staff did agree, however, to consider those 10H inspections if GPUN could demonstrate that the examiners had passed the requalification tests on the first attempt. GpUN has reviewed the records, and cannot demonstrate that the 10R examiners took the requalification exam.
Dased on this fact, CPUN will not take credit for the 10R inspections.
However, we note that the 10R inspections did, in fact, detect IGSCC in 10 welds and " suspected
- IGSCC in 9.
Of these, cracking was verified destructively in 3, non-destructively by ID PT examination of replaced wolds in 5, and one suspect IGSCC call was destructively verified as not being IGSCC.
Therefore, while these examinations were performed by examiners not considered qualified, it is clear that they were effective in detection of 10 SCC.
4.2 Post Stress Improvement Inspections We consider 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 outager is not warranted.
The concern with not performing immediate inspection 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 near mid-wall is largely compressive (about - 15 Kei). This is the reason that most IGSCC in large diameter piping appears to have been arrasted near mid-wall.
Changing the residual stress pattern such that the outer half is subjected to tensile strenses 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% of pipe wall thickness.
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 region, the applied stress intensity is too high for the cc.mpressive stresses to stop, or even retard, crack growth.
013-0012.18
TR - 050 Rev. 3 Page 18 of 43 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 j
operating cycle.
After SI of an 8-inch diameter joint, a crack must be at least 724 TW before it can grow.
These comparisons were based on calculations using a 10 r,si 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 same crack would not grow to an un-acceptable depth within an operating cycle if stress improved.
We are unaware of any instances of crack initiation or unaccept-able crack growth it, a properly stress improved weldment. We are aware of several instances of supposed new cracks in SI'd welds that vere inspected in successive outages.
In the first outage, the indications were not dispositioned as IGSCC; in the following outage, indications were dispositioned as Ioscc.
However, reviews of these cases concluded that the interpretation of the first outage's data was incorrect and should have been dispositioned as IGSCC. 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.
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 inspection 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 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 <12 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 Insoection of cast Material GPUN does not consider that IOSCC in cast stainless steel is a generic problem because there are no reported cases of through-wall IOSCC in this material. The UT techniques developed to date can only detect cracks with site larger than 50% through-wall thickness in centrifugal cast stainless steel (CCSS).
The grain structure in the CCSS is much more uniform than that found in statical cast stainless steel.
It is the lack of uniformity that 013-0012.19
.~.
l 1
TR - 050 Rev. 3 Page 19 of 43 affects the ability to perform a reliable examination for small cracks such as IGsCC. Therefore, we consider that there is no NDE technique available that would reliably detect ICSCC and meet the requirements of Gb 88-01 and NUREG 0313, Rev. 2 [19).
EPRI is working on developing screening methods to determine acoustic characteristics of castings and the resultant methods for performing reliable examinations.
GPUN will develop plans to inspect casting welds when acceptable techniques are available.
Since no meaningful results would be generated by the current NDE methods, ultrasonic inspection of these casting welds during 13R would not be in the interests of reducing radiation exposure.
By the end of the 14R outage, only the 5 Recirculation pump suction elbow-to-pump casing (all are castings) welds will not have been stress improved.
Tnat 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 these reasons, visual inspection during pressure testing for these 5 Recirculation System welds will be performed in lieu of UT inspection.
4.4 Imniementation of Hydrocen Water Chemistrv GPUN has implemented Hydrogen Water Chemistry (HWC) during Cycle 12 (1990). Where HWC is effective in flowing systems (Recirculation System and Reactor Water Clean-Up System), a factor of two reduc-tion in inspection frequency will be applied upon approval by the NRC.
Portions of systems that are stagnent during power operation will also be afforded additional protection by HWC.
The existence of 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 will be applied in these cases upon approval by the NRC.
4.5 RWCU Welds outboard of the Second Isolation Valve We have determined from a review of the piping stress report, piping fabrication and installation records, and past inspection history thats (a) the operational piping stresses are predominately higher, approximately by a factor of 2, inboard of the second isolation valve.
Therefore, welds residing inboard have a higher propensity for ICSCC than those outboard of the second valves (b) there is no documentation that weld 'aoaire exist within the welds outboard of the second valve w)3:h could increano the IGSCC propensity over those welds residing inaoard of the second valve. Additionally, the welds located outboard of the 013-0012.20
TR - 050 Rev. 3 Page 20 of 43 second isolation valve received similar RT during construction as that of inboard welds; (c) there is no evidence of significant piping material chemistry difference between the piping inboard and outboard of the second isolation valve.
Therefore, the propensity for IGSCC in regard to alloy composition is relatively the same; (d) the results of our previous augmented examinations conducted during the 11R and 12R outages show no evidence of IGSCC of RWCU welds inboard of the second valve.
Based on the above dincussion, the results of the previous augmented examination would bound the conditions of the RWCU welds located outboard of the second isolation valve.
Therefore, we believe it is prudent to implement a visual inspection progra.n f or wolds located outboard of tbo second isolation valve during the hydro test.
However, in response to the NRC Staf f's generio concern for kWCU welds outboard of the second isolation valve, GPUN will inspect 10% of these welds in the 13R [19).
Should indications of IGSCC be found in theos wolds, a required sample expansion as discussed in paragraph 5.3.2 would be implemented.
5.0 KV1mG 0313. REV. 1 5.1 Et! PEG 0313 Sp_godor Oypter Cregh The scope of the NUREG applies to the following syst6ms:
Eystem Entta(
Ng of Mytjg Recirculation Wholo 89 Isolation Condenser Whole 189 shutdown cooling Stainicas Steel Portion 14 Core Spray Stainless Steel Portion 27 greater than 200'F Reactor Water Clean-Up From Rectre, to in1.at 136 side of non-regenerative heat exchanger and from outlet side of third re-generative heat exchanger shell to Recirc.
Closure Head Three nozzles to points 7
at which diameter is less than 4" or changes to carbon steel.
Table 1 lists the approximate number of welde included in the scope of the NUREG for each system both before and after 13R.
013-0012.21
TH - 050 j
R3v. 3 Page 21 of 43 l
In summary, 462 welds (26 of them are not ins pectable ) fall within the scope of the NUREG.
Among the 26 uninspect.able welds, 19 welds will be eliminated through pipe spools replacement during 13R.
Therefore, only 7 welds will not be inspectable after 13R.
The number of welds in each system by category, as well as the number of inspections required for 13R by category, are listed in Tables 2 thru 14.
The reason for identifying welds inside/outside the drywell for the Isolation Condenser System and the Reactor Water clean-up System is that the drywell is a difficult area to schedule work since it is normally the most congested area during an outage.
5.2 GPUP Proposed Proag33 Tables 2, 3, 4, 5,
6, 7,
8, 9, 10, 11, 12, 13, 14 and 15 describe the categories and inspection of welds before, during and after the 13R refueling outage.
Tables 8, 9, 10 and 12 indicate a dramatic reduction in lower category welds because of the replacement of the ICS piping and the replacement of the RWCU penetration piping.
5.3 Samole Expansion GL88-01 requires sample expanolon if indications of IOSCC are de-tected in the initial sample.
The additional sample size should be approximately equal to that of the initial sample of the cate-gory of weld in which IGSCC is detected, irrespective of sample and pipe size.
If ICSCC is detected in the second sample, all welds in that category ohould be inspected.
The sample expansion requirements of GL 88-01 will be partially met as modified herein:
5.3.1 System Safe-Ende It is proposed that should flaws be detected in safe-end welds of a specific category, as defined in CL88-01, then l
an equal number of safe-end welds be examined within that category in the expansion sample.
Should flaws be detected in this expansion sample, all safe-end welds in
(
that npecific category will be oxamined.
I l
5.3.2 RWCU Ploina It is proposed that 10% of the welds outboard of the second valves be includod in the 13R UT inspection sample.
If indications of IGSCC are found in these welds, GPUN will approach the NRC Staff on the disposition of these welds and any plans for expansion. Also, we tentatively plan to inspect a further 10% of RWCU welde during future refueling-outages, pending results of 13R inspections and subject to possible inspection reduction as may be allowed by the use of UWC.
013-0012.22 t
7
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TR - 050 Rev. 3 Page 22 of 43 5.3 1 Remainino Welds (Recirculation. Core Sorav. Shutdown Coolina. Isolation Condenser Inside and Outside Drywell.
id closure Head Plo1D21 It is proposed that should flave be detected in welds of a specific category, as defined in GL88-01, then an equal number of welds be examined within that category in the expansion sample.
Should flaws be detected in this expansion sample, all welds in that specific category will be examined.
5.3.4 Samole Exoansion for " Suspect" Welds
" Suspect" welds are those that required extensive examination in order to disposition these welds as not containing indications of IGSCC.
For example, in 11R, there were eight recirculation system welds that required additional examinations that led to the conclusion, at that time, that the indications were characteristic of root condition, geometry, and/or counterbore.
Experience at other plants showed that these conditions often resulted in subsequent examinations concluding that the indications were really characteristic of IGSCC.
Therefore, our initial sample of six welds in 12R was selected from the eight " suspect" welds from 11R.
Two of the six were dispositioned in 12R as containing IGSCC. There were no substantial changes in location or signal characteristics from the 11R to 12R examinations.
Since our sample expansion required examination of additional six welds, the two remaining 11R " suspect" welds plus four others were examined. One of the two remaining 11R " suspect" welds was determined to contain IGSCC. This finding resulted in examination of the remaining 49 Category C Recirculation system welds.
No additionni indications of IGSCC were detected.
Based upon the above sequence, we consider that our approach to weld selection was cased on sound engineering and EPRI-qualified !!OE examJ ners' judgement. Therefore, we consider that the following sample expansion requirement for " suspect" welds to be technically sound and ALARA conscious:,
During each outage, a separate group of " suspect" welds, if there is any, will be estabittheJ.
Sample expansion of these " suspect" welds would not be vraquired since all the
" suspect" welds will be reinspected ter IGSCC during the following outages until they can be determined not to have indications of IGSCC.
In the event that the " suspect" welds are determined to be free of IGSCC, they will reenter the normal sampling plan.
013-0012.23
TR - 050 Rev. 3 Page 23 of 43 Currently, there are no " suspect" welds that require reinspection during 13R.
6.0 QTHER gtfiKRLQ_LETIgfLD.B-01 REstgligtg_EgqqlEED 6.1 The NRC Staf f has proposed a change to the Technical Specifica-tions (TS) to include a statement in the section on ISI that the In-service 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 In-service Inspection and Testing sections may be removed from the TS in the ruture 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 arternative document.
To comply with the requirements of GL88-01, GPUN will submit a TS Change Request consistent with this response during the upcoming refueling outage, scheduled for completion in the Spring of 1991.
6.2 The NRC Staff position in CL 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, NUREC-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). CPUN's recent submittal of July 1, 1988, states that installation of a new APGRHS will be completed during the operating cycle 12.
The submittal also identifies that there are several leak detection methods available for unidentified leakage into the containment sump at Oyster Creek which operate on diverse principles.
The normal method of monitoring unidentified leak rate is to obtain flow integrator readings from the containment sump pump discharge every four hour period and calculate average flow rate.
Approximately 1 gpm can be measured in a four hour interval.
This methodology is identified in Oyster Creek Technical Specifications as the primary method of leakage measurement.
When the flow integrator is not available, the average leakage rate can be calculated using the known volume between the high and the low level alarms for the sump and the time required to fill the cump between these levels.
013-0012.24
TR - 050 Rov. 3 Page 24 of 43 A recorder available in the control room also provides con-tinuous indication of an estimated unidentified leak rate to the containment sump by utilizing a differontial pressure signal as a result of the sump level change.
The sensitivity of the recorder is approximately 0.2 gpm.
Additionally, a timer available in the 480 volt switch gear room provides the run time of the containment sump pumps.
This run time along with the estimated flow rate of the sump pumps can provide approximate leak rates.
This methodolog) 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 interval.
The time associated with this alarm is set to bring in the alarm if unidentified luak rate equals or exceeds 4 gpm.
These methods provide quantitative indications of unidentified (CS leakage inside containment and also provide assurance that un-identified leakage can be detected and quar.tified during cycle 12 operation pending operability of the new APCRMS.
The NRC Staff position was further amplified in GL 88-01 by additional critoria as follows:
1.
Plant shutdown should be initiated for inspection and cor-rective 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 less, any leakage detection system indicates an increase in rate of unidentified leakage in excess of 2 gpm or its equivalent, or when the total unidentified leakage attains a rate of S gpm or i
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.
2.
Unidentified leakage should include all leakage other thans 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' containment atmosphere from sources that are both specifically located and known either not to interfere with the operations of unidontified leakage j
monitoring systems or not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.
013-0012.25
m TR - 050 Rev. 3 Page 25 of 43 3.
For plants operating with an ICSCC Category D, E, r, or a welde, at least one of the leakage measurement instruments associated with each sump shall be operable, and the outage time for inoperable instruments shall be limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or immediately initiate an orderly shutdown.
By Amendment 97 to Provisional Operating License No. DPR-16 for Oyster Creek, the limiting conditions for operation and curveillance 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 Leo's for the containment sump flow monitoring system and the equipment drain tank monitoring system; and added a new surveillance section TS 4.3.H.
This amendment incorporated GPUN's response dated September 8, 1983, to IE Bulletin 82-03.
On March 17, 1987, GPUN submitted Technical Specificatior, Change Request #158 which adds additional conservatism to these requirements by proposing to limit the unidentified leakage for the Reactor coolant System to a maximum leak rate 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. On May 23, 1989, the NRC staff approved this request as Amendment 133.
This amendment addresses item 1 of the NRC Staff position.
6.3 OPUN plans to notify tne NRC of any flaws identified that do not meet IWD-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 the flaws for continued operation and/or repair plans. OPUN will obtain NRC approval of the evaluations and/or repaire prior to restart.
[
7.0 EMMMARY In response to the NRC staff's concern that all Category G welds, as a minimum, be inspected no later than the end of 13R, GPUN has revised its inspection plan.
Even though CPUN is not in complete compliance with NUREG 0313, this revised plan is sensitive to the NRC concern while maintaining a controllable outage work scope and minimizing radiation exposure.
Except for the recirculation inlet nozzles and the RWCU piping outside of the second isolation valve, the proposed plan will be in compliance with the requirements of GL-88-01 and NUREG 0313, Rev. 2.
l l
l l-013-0012.26
TR - 050 Rev. 3 Page 26 of 43 8.0 gitgBIngig 1.
USNRC Generic Letter 84-11, " Inspections of DWR Stainless Steel Piping," April 19, 1984.
2.
NUREG-0313 Rev. 2,
- Technical Report on Material Selection and Processing Guidelines for DWR 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, 1988.
4.
USNRC IED No. 82-03 Rev.
1,
" Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at DWR Plants," October 28, 1982.
5.
GPUN Topical Report No. 012 Rev. 1,
" Oyster 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,
" Oyster Creek Cycle 11R Outage IGSCC Activities," 9-30-86.
8.
Oyster Creek FSAR Amendments 29, 35, 36, 37, 40, 43 and 47.
9.
NRC Letter dated October 18, 1988.
10.
NRC/GPUN Telecon - 10/25/88.
11.
NRC/GPUN Telecon - 11/30/88.
12.
GPUN GL88-01 Response - 8/12/88.
13.
GPUN GL88-01 Response 1/31/89.
14.
GPUN GL88-01 Response 11/16/89.
15.
General Electric Report No. 89-178-002 " Evaluation of Core Sample NG-D=11 from Oyster Creek Nuclear Generating Station."
16.
Generai Electric Report No. 89-178-007 " Metallurgical Examination of 2-inch Head cooling Line Nozzle Weld from Oyster Creek Generating Station."
17.
GPUN Letter to NRC No. 5000-90-1891 " Oyster Creek Nuclear Generating Station Docket No. 50-219, IGSCC Inspection Plan -
l RWCU" dated 2/21/90.
[
l 013-0012.27
TR - 050 Rev. 3 4
Page 27 of 43 10.
NRC Letter dated April 17, 1990.
19.
GPUN Response Letter to NRC No. 5000-90-1938 dated 6/7/90.
20.
CPUN Letter tc NRC No. 5000-90-1982 dated 10/18/90.
l 6
9.0 (1,g2858 1.
Safe-end configuration.
2.
comparison of crack growth rates.
3.
Peach flottom Unit 3.
4.
Effects of aqueous impurities on crack growth.
P c
f
'013-0012.28
TR - 050
=
Rev. 3 Page 28 of 43 la 16ue two aufttil 6 til6
~
etMnaebt (ePt IElll 4 lt!'
~
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, g g maim.._
e est als (stitual 41
- 1.616 muit man '
- wlits Iwe aufttli 8. lit ummenL, l.436 es (asstatut 61 signaamd (art latill.4m useinge IEW1 me unseses 00I OI
- - 4 le i I APle e SM nie(Wf tNat 41
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f u/s mensant e.tas ele q anlietas ihm 4
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'],
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i tems(AW 182 4 la s u f an intel
~ame/>.
fause mssuag mA/
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/
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W X
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,s
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flu une le* ma to. nee s to.ene e tuost l
las.flon terv Qufttfl 74182 e 362 er 61 aldmAm4 (wt IElllI4 I/g e
-1e ant 4 ic i iwta
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ses nAneie6 me-i t/ 46 - em 1Afl l@ Ustina P9148 le flata omf ataleb l
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FIGURE 1 - RECIRCULATION SAFE-END CONFIGURATION 013-0012.29 l
TR - 050 Rev. 3 Page 29 of 43 25 ~
N 7
Avg cond -0 2 ps/cm
~
E
- 24
- O'# '
Y ml BM weler chemetry g
(P6ech Gottom 3) 15
[
Avg cond -0 3 pucm y. 0 10 Hydrogen walet chenustry 4I A$w"-05#cm gv9 c g,,fe, 5
-7' b = <5 mds/yr 0
0 500 1000 1500 W
N Test Time (hts)
Figure 24. Competison of the crock growth tales of sensdued type 304 steentens steel specimens in poemal wales chamastry et Peach Bottom 3 and 6n hydrogen watef chemesley se Ovenden 2.
FIGURE 2 013-0012.30
TR - 050 Rev. 3 Page 30 of 43 1
\\
PEACH SOTTOM UNIT 3 Seewtlaed Type 304 84 Crack Length (In.)
Conductmty ( 5/cm)
.754 2.5 A = 2.36 2.57 a 104 nA 20 i
.753 at A e 4.32 4.04 x 104 inA-1.S
.752
'a 10
.751 entrusion intrusion No.1 No. 2 05
.750 l
0
.740 2100 222 220 24(D 2000 M
2700 2000 Time (h)
FIGURE 3 013-0012.31
e TR - 050 Rev. 3 Page 31 of 43 CYCtec J
/LOAO k % constamtLOAo
..s
- n' N
- 7808
< e i.44u i e esica n.p.
g
.85 I
1 g
.ei
.e3 It 9
3 1
.s2
~
1 3essanonvoam - - - - -
.. as sem ouveem -. <
g h
yee s n o s v oam - -.
88
_f a
f__
1 so
.79 A
~
.79 1-1
-4.
L,,, 4 a
a See 1n08 ISBs 2880
?$00
' 7 7 i _ a_
f man Ina FIGURE 4.
EFFECT OF AQUEOUS IMPURITIES ON CRACK GROWTH OF SENSITIZED TYPE 304 STAINLESS STEEL IN SCO' (288'C) WATER 013-0012.32-
. -.. -. -...-. -.. - - - - - =, - -.. _ ~ --. -.
TR - 050 Rev. 3 Page 32 of 43 10.0 TABLES 1.
Number of Welds in Scope of NUREO 0313, Rev. 2 2.
Recirculation Safe-End Welds 3.
Core Spray Safe-End Wolds 4.
Isolation Condenser Safe-End Welds 5.
Recirculation Pipe Welds 6.
Shutdown Cooling Pipe Wolds 7.
Core Spray Pipe Welds 8.
Reactor Water Clean-up Inside Second Isolation Valve 9.
Isolation Condenser Pipe Welds (Incide Second Isolation Valve) 10.
Closure Head Pipe Welds i
11.
Reactor Water Clean-up (Outside Second Isolation Valve) 12.
Isolation condenser Pipe Welds (Outside Second Isolation Valve) 13.
Safe-End Helds 14.
Summary of Total Inspectable Pipe Welds 15. - Inspection Schedules for BWR Piping Weldments Kev to Inspection Cateaories A.
Resistant Materials B.
Nonresistant Material SI within 2 years of operation C.
Nonresistant materials SI after 2 years of operation and Post Process Inspected.
D.
Nonresistant material - Inspected - No SI.
C/D. Nonresistant materials SI af ter 2 years of operation and not Post Process Inspected, but has been inspected for IGSCC during 11R or 12R.
This category will be treated as Category "D".
C/G. Nonresistant materiale SI af ter 2 years of operation and not Post Process Inspected and not inspected for ICSCC during 11R or 12R.
This category will be treated-as Category "G".
E..
Cracked - Overlayed or SI F.
Cracked - Inadequate - No Repair G.
Nonresistant - not inspected l
l
~013-0012.33 i
TR - 050 Rev. 3 Page 33 of 43 l
TABLE 1 NUMBER OF WELDS IN SCOPE OF NUREG-0313. REV. 2 BEFORE AND AFTER 118 INSPECTABLE WELDS UNIN-TOTAL IN-INSIDE OUTSIDE TOTAL SPECTABLE SPECTABLE INSIDE OUTSIDE 2ND ISO.
2ND ISO.
SYSTEM WELDS WELDS WELDS DRYWELL DRYWELL VALVE VALVE B
A B
A B
A B
A B
A B
A B
A Recire.
89 89
. 5
- 5 84 84 84 84 0
0 84 84 0
0 RWCU 136 131
- 5 0
131 131 30 30 101 101 46 46 85 85 CS 27 27 0
0 27 27 27 27 0
0 27 27 0
0 l
SDC 14 14 0
0 14 14 14 14 0
0 14 14 0
0 IC 189 114 16 2
173 112 44 44 129 68 44 48 129 64 Closure l
Head 7
7 0
0 7
7 7
7 0
0 7
7 0
0 TOTAL 462 382 26 7
436 375 206 206 230 169 222 226 214 149 l
casting-to-casting welds.
Welds inside penetrations (will be eliminated in 13R).
8 welds inside penetrations,.4 flued head-to-valve welds, 2 casting-to-casting-welds, 2 saddle welds (ICS piping outside drywell will be replaced during 13R).
As After 13R B: Before 13R 1
013-0012.34
TR - 050 Rev. 3 Page 34 of 43 TABLE 2 RECIRCULATION SAFE-END WELDS NUREG TOTAL WELDS.
TOTAL WELDS j
INSPECTION PRIOR TO 13R AFTER l-CATEGORIES 13R INSPECTIpl{S 13R I
C-12R/HF 4
0 4
l D/HF 0
0 8
j 0/ilF 16 8
8 l
TOTAL 20 0
20
!!F - HWC Flowing TABLE 3 CORE SPRAY SAFE-END WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSf1CTIONS 13R C-12R/NP 6
0 6
TOTAL 6
0 6
NP - No HWC Protection 013-0012.35
_-.... - -... ~... - - - _ - ~. -. - -. -.. - - -. -. ~ ~. _.
o TR - 050 Rev. 3 Page 35 of 43 TABLE 4 ISOLATION CONDENSER SAFE-END WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATECORIES 13R INSPECTIONS 13R C-12R/NP 4
0 4
TOTAL 4
0 4
NP-No HWC Protection TABLE 5 RECIRCULATION PIPE WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTICN PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R C-11R/HF 58 0
58 E/HF 6
3 6
UNINSPECTABLE 5
0 5
TOTAL 69 3
69 HF - HWC Flowing From Vessel Outlet to Vessel Inlet l
l.
013-0012.36 l
6
=
e-e-
w
,ws---,-
..--w-
y y
._,...-r.------4~-
,er--,
e..-,,v--,-y,-,-,
e
TR - 050 Rev. 3 Page 36 of 43 TABLE 6 SHUTDOWN COOLING PIPE WELDS NUREO TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R D/NP 4
3 5
D/HS 3
9 O/NP 1
1 0
c/HS 4
4 0
TOTAL 14 11 14
!!S - HWC stagnant with protection based on thermal mixing of the recirculation connection point NP - No HWC Protection l
l l
l 013-0012.37 l ':
l
a..-
.. - - - _ - - - --. -.. ~
_ _ ~. -.... -
e e
TR - 050 Rev. 3 Page 37 of 43 TABLE 7 CORE SPRAY PIPE WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R C-12R/NP 8
0 14 C/D-12R/NP 7
3 4
]
C/G/NP 3
3 0
D/NP 2
2 2
E/NP 1
0 1
TOTAL 21 8
22
-NP - No HWC Protection TABLE 8 REACTOR WATER CLEAN-UP PIPE WELDS (INSIDE SECOND ISOLATION VALVE)
NUREO TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R A
0 0
4 D/HF 20 10 42 G/HF 26 26 0
'UNINSPECTABLE 5
0 0
TOTAL 51 36 4G HF - HWC Flowing from recirculation sup7 y to return connection points-l l
l Note:
All Category "A" welds installed in 13R will be baseline inspected (4 welds) l l
013-0012.38 l
t
TR - 050 Rev. _ 3 Pago 30 of 43 TABLE 9 ISOLATIO!I CO!!DEllSER PIPE WELDS (IllSIDE SECOrlD ISOLATIOt1 VALVE) fiUREO TOTAL WELDS TOTAL WELDS IllSPECTIO!!
I'RIOR TO 13R AFTER CliTIGORTES 13R I_flSPECI1Q11S llR _
A/ils 0
0 12 C-12R/!iP O
O O
C/D-12R/ lip 5
4 1
D/ils 15 6
23 I
D / tlP 2
2 0
0/lis 12 12 0
G/f4P 2
2 0
C/0/flP 4
4 0
Ull!!1SPECTABLE 14 4
2 TOTAL 54 34 46 flP - flo liWC Protection 11S.- !!WC stagnant with protection based on thormal mixing at the recirculation connection point tiote : 1) All Category "A" wolds inst.alled in 13R will be baselino inopoeted (12 wolds) s l
013-0012.39 l
. 4 e
TR - OSO Rev. 3 Page 39 of 43 TABLE 10 CLOSURE HEAD PIPE WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO
.A AFTER CATEGORIES 13R INSPECTIONS DR D/NP 4
0 7
G/NP 3
3 0
TOTAL 7
3 7
NP - No HWC Protection TABLE 11 REACTOR WATER CLEAN-UP PIPE WELDS (OUTSIDE SECOND ISOLATION VALVE)
NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R D/EF 0
0 9
O/HF 85 9
76 TOTAL 85 9
85 HF - HWC Flowing from recirculation conrection pointa supply to return 1
013-0012.40 t
e o
TR - 050 Rev. 3 Page 40 of 43 TABLE 12 ISOLATION CONDENSER PIPE WELDS (OUTSIDE SECOND ISOLATION VALVE)
NUREG TOTAi 4 ELDS TOTAL WELDS INSPECTION PRIO., TO 13R AFTER CATEGORlKE 13R INSEECTIONS 13R A/NP O
O 64 D/NP 59 34 0
E/NP 22 5
0 G/NP 48 23 0
UNINSPECTADLE 2
2 0
TOTAL 131 64 64 NP - No llWC Protection Note All Category "A" wolds installed in 13R will be basolino inspected (64 welds)
TABLE 13 SAFE-END WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES llR IEffE_QIlpf4S 13R C-12R/HF 4
0 4
C-12R/NP 10 0
10 D/HF 0
0 8
G/HF 16 8
8 TOTAL 30 8
30 HF - IfWC Flowing NP - No HWC Protection 013-0012.41
e e
TR - 050 Rev. 3 Page 41 of 43 e
TABLE 14 GPUN PitOGRAM
SUMMARY
OF TOTAL INSPECTABLE WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER ChlsslQEl%S 1)R llibf1CllQNS llit A
0 0
80 C-11R/lir 58 0
58 C-12R/llr 4
0 4
C-12R/NP 18 0
32 C/D-12R/NP 12 7
5 D/llr 20 9
59 D/HS 20 7
32 D/NP 71 5
14 E/flF 6
3 6
E/NP 23 0
1 G/HF 127 36 84 G /IIS 16 14 0
l G/NP 54 4
0 l
C/G/NP 7
7 0
TOTAL 436 92 375 l
IIF - !!WC Flowing IIS - IIWC Etagnant - with protection via thermat mixing NP - No IIWC Protection Note:
All Category "A" welds installed in 13 A vill be bameline inspected (80 welds).
013-0012.42
TR - 050 Rev. 3 Page 42 of 43 TABLE 15 INSPECTION SCHEDULES FOR Ph1 E] PING WELDEENtf (3)
DESCRIPTION IGSCC
' JPF^ TION GPUN PROPCJED OF WELDMENTS NQIES CATECOPY ELP A JEEDf1LE ENTENT/SQHT M E Resistant Materials A
25' ave y 10 ys ' cm S AM*;
(at letn' 129
.a 6 y0Pte)
Nonresistant Matis.
(1)
B 50% ev,ry 10 res a SAME SI within 2 years (at lea s'. 9 5% in 6 of operation years)
Nonrenistant Matis.
(1)
C A11 next 2 re-SAME SI after 2 years fueling cycles, then of Operation All every 10 yrs (at least 50% in 6 yrs)
Nonresistant Matl.
(1)
D All every 2 refuel-SAME No SI ing cycles Cracked (1)
E 50% next outage, then SAME overlayed or All every 2 refuel-SI ing cycles Cracked F
All every refueling SAhE Inadequate outage No Repair Nonresistant (2)
G All
- f. ext refuelir.g All by end of 13R outage Not Inspected outage except 8 recirculation system safe-ends welds and RWCU welds located outboard of the second isolation valva (See para. 2.0).
013-0012.43
... ~.
. _ - ~
o;.+
e TR - 050
,a Rev. 3 e
Page 43 of 43 IABLE 15 i cont.._ )
Notegs (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)
Welds that are not UT inspectable should be replaced, " sleeved," or local leak detection applied.
RT examination or visual examination for leaks may aloa be considered.
I.
i 1
013-0012.44
--