ML20151K471
| ML20151K471 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Oyster Creek |
| Issue date: | 11/06/1984 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Starostecki R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20151H203 | List:
|
| References | |
| FOIA-86-26 GL-84-11, NUDOCS 8411140012 | |
| Download: ML20151K471 (1) | |
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8 UNITED STATES h
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WASHINGTON, D. C. 20555 November 6, 1984
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Docket No. 50-219 MEMORANDUM FOR: Richard W. Starostecki, Director Division of Project and Resident Programs, Region I FROM:
Darrell G. Eisenhut, Director Division of Licensing, NRR
SUBJECT:
INSPECTION AND REPAIRS ON THE ISOLATION CONDENSER SYSTEM PIPING, TASK INTERFACE AGREEMENT 84-53, OYSTER CREEK
REFERENCE:
R.W. Starostecki memorandum to D.G. Eisenhut dated June 7, 1984; Oyster Creek and Millstone 1 Isolation Condenser and Recirculation system pipe cracks.
As requested, the technical and safety aspects of crack indications identified at Oyster Creek have been evaluated.
We have reviewed the submittals regarding the inspection and repairs to the Isolation Condenser (IC) system piping. All IC piping welds outside the containment and 15 IC piping welds inside the containment were 7
ultrasonically inspected. A total of 27 welds outside the containment were reported to show crack-like indi. cations. Of these, 9 welds were replaced and 18 welds were overlay repaired.
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We have concluded that the inspection and repairs were performed satisfactorily, and the plant can be safely returned to power. Subsequent reinspection of piping systems at the next refueling outage should be conducted in accordance with staff guidance provided by Generic Letter 84-11, dated April 19, 1984.
Our Safety Evaluation is enclosed.
This completes the NRR. review pursuant to TIA 84-53
&ZWj j[ Darrell g.' EishnhGt, Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosure:
Safety Evaluation L
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY g-i OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 INSPECTION AND REPAIRS OF ISOLATION CONDENSER SYSTEM PIPING
1.0 INTRODUCTION
During the current Oyster Creek refueling outage, au'gmented ultrasonic testing (UT) was performed on the recirculation system piping in accordance with IE Bulletin 82-03.
No intergranular-stress-corrosion-cracking (IGSCC) indications were reported.
In a hydrostatic testing of the "A" loop isolation condenser (IC) leakage from two small pin-holes was observed from the 8-inch return (condensate) line outside the containment near weld NE-2-12.
The IC system consists of two loops of steam (supply) lines (12" arrd 16") and condensate (return) lines (8" and 10").
All the piping is made of type 316 stainless steel materials.
Subsequently, all the IC piping welds'(124) outside the containment were ultrasonicall l
A total of 27 welds including 10 welds in the condensate lines (y inspected, nine welds in 8" lines and one weld in 10" line) and 17 welds in the steam lines (11 welds in 12" lines and. 6 welds in 16 line)' were reported to show crack-like indications. All reported. crack indicatio'ns were oriented.in the circumferential direction and located in the heat affected zones.
Fifteen welds in the IC system inside the containment were also ultrasonically inspected and no crack-like indications w4re detected.
2.0 _ DISCUSSION AND EVALUATION
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2.1 Inspection Qualified UT personnel from GPU Nuclear Corporation (GPUN) and Virginia Corporation of. Richmond (VCR) performed the ultrasonic examinations.
Crack detection, discr'imination and sizing were performed primarily by GPUN.
VCR performed only confirmatory crack depth sizing on welds showing crack-like indications.
The licensee indicated that one of the VCR UT personnel participatjng in the crack depth sizing took the UT sizing course given by EPRI at the NDE Center,. Charlotte, North Carolina and passed the examinations.
For some welds, the depth sizing results reported by the two g c::r 4
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teams did not always agree, with variations as much as 64% of the wall thickness.
Region.I has determined that the UFUN UT procedures, calibration standards, equipment and IGSCC detection capabilities were satisfactorily demonstrated in accordance with IE Bulletin 83-02, and that the same prdcedures and techniques were used in the UT examination.
Region I also I
indicated that all GPUN UT personnel conducting these inspections have received appropriate training in IGSCC inspection using service-induced IGSCC cracked thick-wall pipe specimens.
As will be shown later, the UT j
sizing data were not relied upon for any repair considerations.
The radiographic examinations were also perfonned to assist the discrimination between the root geometry and cracks. The ultrasonic examinations identified 27 (22%) IC welds (class 2) outside the containment that contained crack-like indications.
Of these, eight welds were classified as " suspect" because these welds could not be confirmed as cracks or classified as geometric reflectors.
These suspect welds were conservatively treated as cracked welds.
One of the' suspect welds was later replaced and no cracks were found on that weld by penetrant examination. Of the 27 cracked welds, 9 welds were rep. laced and 17 welds were overlay repaired.
2.2 Failure Analysis Two welds (NE-2-12 and NE-2-13) from the condensate line loop "A" and two welds (NE-1-15, loop A and NE-1-61, loop B) from the supply line were removed for failure analysis. General Electric's (GE) Turbine Technology Laboratory, GPUN's contractor, evaluated welds NE-1-15, NE-2-13, and the bottom half of NE-2-12 (containing the leaker).
Brookhaven National Laboratory (BNL),
the NRC's contractor, evaTuated. weld NE-1-61 an'd the top half of. weld NE-2-12. GE performed liquid penetration examination on the inside surface of weld NE-2-13 which is a " suspect" weld.
The weld revealed no crack indications.
Both GE and BNL used scanning electron microscopy and conventional metallography to study three cracked welds (NE-2-12, NE-1-15 and NE-1-61)the fai)ure mode in the They reported that all the cracks in these welds were intergranular, covered with heavy oxides and located adjacent to the weld bead.
The IC condensate return lines are normally stagnant and at ambient temperature because the condensate return isolation valves are closed during normal operating condition. ThestaffgenerallfdoesnotexpectIGSCCto occur at ambient temperature because the initiation of IGSCC is a temperature-dependent process.
However, the licensee reported that during the early operating years, the IC system was u. sed quite frequently (at least 33 times),
and extensive leakage through the condensate return isolation valves was observed at least seven times during the period of 1976 to 1980.
The steam leakage through the isolation valves elevated the temperature of the condensate return lines and thus, accelerated the initiation of the cracks in the condensate return lines.
Based..on the reported operating history, e.
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the staff believes that the cracking probably occurred during the period when i
extensive leakage through the isolation valves was observed. This is also consistent with the observed heavy o.xides on the fracture surfaces, which indicates that the cracks were not initiated recently.
Therefore, the staff considers that the cracking in the condensate return lines could be unique to Oyster Creek.
2.3 Repair GE performed the weld overlay design of the 18 cracked welds for the j
licensee.' The repair overlay was designed to meet the ASME Code Section XI IWB 3640 requirements and to provide a. full structural reinforcement of the cracked weld.
The overlay thickness was calculated based on a pressure of 1090 psi (corresponding to the technical specificatibn limit for the opening of the electro-mechanical relief valves), and the maximum dead weight (3.3 ksi) and seismic (5.1 ksi) stresses enveloping all 18 cracked welds.
The overla'y design is independent of the crack size as determined by the ultrasonic examination because the cracks in the repaired welds were assumed to be fully circumferential and extended through the original pipe wall. The designed. minimum overlay thickness for various pipe sizes ranged from 0.25 inch to' 0.40 inch, which did not include the thickness of the first layer that passed the Penetrant Test (PT) and the ferrite number test.,
Radiography tests Were performed on each finished weld overlay to ensure the :tructural and bonding integrity of the overlay.
The licensee has replaced 'the piping of nine welds (five 16" welds, two 12" welds and two 8" welds) which showed crack-like indications in the IC system.
The replacement material.for the 8" and 16" piping was purchased to type 316.
stainless steel with carbon content not to exceed 0.05%, and t.he material of the 12" replacement piping was purchased to nuclear grade type. 316 stainless steel.
The nucle'ar grade type 316 stainlesp2 steel material for 8" and 16" pipe sizes was not available to meet the need date. The licensee reported that the replacement piping was upgraded to meet the requirements in ASME Section III subsection NC (class 2) and the welding process of low heat input was used during fabrication.
2.4 Evaluation'
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The staff has reviewed ~the licensee's submittals regarding the ultrasonic examination results, metallography evaluation, and weld overlay designs of the IC system piping at Oyster Cr~eek to support continued service for one fuel cycle with 18 overlay repaired welds and 9 replaced welds in the IC piping system.
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The staff has reviewed GE's weld overlay designs for the 18 IC welds showing crack-like indications.
The overlays were designed to have a full structural i
strength and met all the repair guidelines in Generic Letter 84-11.
Because cf.the current concerns regarding the conservatism of the ASME Code Section j
XI IWB 3640 limits, the staff performed an independent limit load analysis t
to evaluate the design safety margin that will be present in the GE's weld overlay designs. An enveloped calculation based on weld NE-2-80 in the condensate return line was performed.
Weld NE-2-80 has the thinnest overlay design (0.25 inch) and the largest thermal expansion stress (11.27 ksi).
In the limit load analysis, the staff used a reduced flow stress of 45.5 ksi _(corresponding to half of the ASME Code allowed yield stress plus tensile stress for 316 stainless steel at a
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temperature of 550*F) and included in the safety margin calculation the thermal stresses from the espansion (11.27 ksi) and overlay shrinkages (2.1 ksi). The "i" index of stress : intensification factor was not considered in the thermal stresses.
The thermal expansion stress for weld NE-2-80 was reported in a recent analysis of the IC system by MPR Associates, Inc. for the licensee, whi'h was conservatively calculated based on a design c
temperature.
The shrinkage stress was calculated based on the actual displacement measurements before and after repair.
The staff.'s limit load ~
calculations have shown that there is a safety factor of 5.2 on the bending stresses (8.25 ksi) which includes the primary (dead weight and seismic stresses) as well as the secondary (thermal, expansion and shrinkage stresses) bending stresses.
Therefore, the staff agrees with the licensee's conclusion that the continued operation of Oyster Creek for one fuel cycle with the 18 overlay repaired welds.in the IC syst m is j.ustified because the Code required structural safety. margin in the 18 overlay repaired welds would be maintained.
During this refueling outage, the licensee replaced eight welds showing i
crack-like indications in the IC system.
Two 12" welds were replaced wi,th the nuclear grade type 316 stainless steel.
Nucl' ear grade 316 stainless ste 1 is considered to be not susceptible to IGSCC under normal BWR environment. The other six welds were replaced with conventional type 316 stainless steel material with carbon content not over 0.05%.
Stainless steel piping with carbon content over 0.02% is considered to be susceptible to IGSCC in normal BWR environment because such materials are prone to sensitization when heated to elevated temperatures.
It is known that an incubation period is required to initiate the IGSCC.
The length of 'the incubation period depends on the environment, stress, and the degree of sensitization of the materials., Based on the BWR operating experiences, the staff does not expect significant cracks to be generated in the conventional austenitic stainless steel within a period of one. fuel cycle.
Furthermore, the piping was replaced by using a low heat input welding process to minimize the sensitization in the heat-affected zones.
Therefore, the staff concludes that the six welds replaced with conventional type 316 stainless steel naterials are acceptable for continued service of one fuel cycle.
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m The staff noted that the licensee relied on radiography tests (RT) to
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confirm the structural and bonding integrity of the repair overlays.
This is not consistent with the present industry practice of using UT methods.
RT.is generally not as sensitive as UT in detecting the lack of bonding, lack of penetration and particularly, the small cracks in the overlays.
UT is considered more sensitive than RT and should be used to confirm overlay integrity.
However the staff acceptance of the RT results at this time is based on the following considerations:
(1) The overlay repaired IC welds are all class 2 welds located outside the containment. Monitoring. of such welds for potential leakage can easily be made during normal operation.
(2) The NRC, Region I, has confirmed that the overlay repairs were properly performed in accordance with qualified procedures consistent with ASME Code Section XI requirements.
Based on Region I's observations and the generally. good experience with overlay repairs, the staff does not anticipate any. major deficiencies in the structural and bonding integrity of.the weld overlays applied at Oyster Creek.
(3) The. licensee has agreed to ultrasonically inspect each overlay repaired weld during the next refueling outage to confirm the integrity of the overlays.
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3.0 CONCLUSION
The staff has concluded th'at the 0yster Creek It system piping has been inspected and repaired in accordance with all current staff guidelines, and that the plant can be safely returned to operation until the next refueling outage.
4.0 ACKNOWLEDGMENTS
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W. Hazelto'n' and W. Koo prepared this safety evaluation.
Date:
September 20, 1984 6
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NUCLEAR REGULATORY COMMISSION
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REGION I
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MEMORANDUM FOR:
Darrell G. Eisenhut, Director, Division of Licensing, NRR FROM:
Richard W. Starostecki, Director, Division of Project and Resident Programs, Region I
SUBJECT:
. CONTRIBUTION TO SAFETY EVALUATION REPORT (SER) 0YSTER CREEK ISOLATION CONDENSOR PIPING REPLACEMENT 1
AND REPAIR
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Region I has reviewed nondestructive examination records and procedures, has observed completed weld overlay repairs and has reviewed the licensee's evalu-ation of NDE results. We find that these activities were done by qualified personnel in accordance with. qualified procedures and methods.
The observations and review involved 46 technical staff hours.
The principal technical reviewer was R. A.. McBrearty.
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g Richard W.
tarostecki, Director i
Division of Project and Resident Programs
Enclosure:
Contribution to SER - Oyster Creek O
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.a CONTRIBUTION TO SAFETY EVALUATION REPORT OYSTER CREEK -- DOCKET NO. 50-219 v:;-.
NONDESTRUCT1VE ' EXAMINATION (NDEl During a hydrostatic test of the isolation condenser system, the licensee detected a leak emanating from a through wall crack in an 8-inch diameter pipe weld. Subsequent NDE of 140 welds defined the extent.of the system cracking.
Licensee evalua' tion of the NDE results indicated that certain whlds must be 1
replaced'and that others could be repaired by the application of weld overlay.
The initial NDE was performed by Magnaflux Quality Services, pipe replacement and repair by the General Electric Company and the subsequent NDE by Magnaflux Quality Services.
The evaluation and dispositior) of examination results was done by the licensee.
The Region has reviewed the NDE procedures, selected radiographic flims, ultra-sonic test data, NDE personnel qualification / certification records and the li-
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censee's eval-vation and disposition of NDE results.
We have verified that all required examinations were done in accordance with qualified procedures, using qualified personnel and concur with the evaluation and disposition of NDE results.
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TASK DATE:
fitdOs N8k TASK INTERFACE AGREEMENT TAC #: T 4j@
PROBLEM: OYSTER CREEK - Isolation Condenser Pipe Cracks LEAD 0FFICE:
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NRR REGION JOINT
REFERENCES:
Memo to DEisenhuyt from RStarostecki dated 6/7/84 subject: Oyster Creek and Millstone 1 Isolation Condenser and Recirculation Systems Pipe Cracks: Transfer of Lead Office and Request for Assistance ACTION PLAN:
NRR: Assume lead responsibility for resolution of isolation condenser pipe crack indications (and other pipe crack indications, should they be identified as a result of related inspections). In accordance with EDO memo of 12/08/82 responsibility for plant-specific problems will be transferred to NRR.
1.
Evaluate the technical and safety aspects of crack indications identified at Oyster Creek. Evaluation should address licensee's inspection program and findings, the licensee's plans for operation, j
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and the repair program. (MTEB) l 2.
Review the results of sample analysis being performed by Brookhaven Lab.
(MTEB) i 3.
Provide evaluation memo /SE input to PM.
(MTEB)
Completion Date: Prior to 7/11/84 to support plant restart.
R-I Inspect site inspection and reoair activitie.s, as appropriate. Provide information to NRR for evaluation of repair.
i NRR: Designate Lead Project Manager to assign TACs pnd coordinate correspondence, meetings, and reports /(ORB #5 J. Lombardo).--Prepare response to RI for DL signature.
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OFFICE C0ORDINATORS:
i ary Holahan (x27415)
R. Vollmer (x27270)
APPROVED.
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(x27817)
(x (I&E)
W
. Starostecki
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(Region I
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F. Mi a
492) i cc:
V. Stello,ROGR E. Conner,R-I H. Thompson,DHFS R. Wessman,DL Regional Adms.
T. Elsasser,R-I T. Speis, DST G. Holahan,DL J. Heltemes,AE0D R. Baer, IE R. Mattson,DSI W. Hazelton N. Grace,IE E. Rossi, IE D. Eisenhut,DL B.D. Liaw J. Taylor,IE E. Case, NRR R. Vollmer DE T. Novak E. Jordan,IE F. Miraglia,DL G. Lainas A-IW
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