ML20238E508
| ML20238E508 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/31/1987 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18032A690 | List: |
| References | |
| NUDOCS 8801050196 | |
| Download: ML20238E508 (57) | |
Text
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REPORT OF PIPE INSPECTIONS, CORRECTIVE ACTIO!fS, AND ANALYSES BROWNS FERRY NCLEAR PLANT UNIT 2, CYCLE 5 OUTAGE Intro' duction l
Inspections for intergranular stress corrosion cracking (IGSCC) have been l
performed in accordance with the requirements of Generic Letter 84-11 on l
Browns Ferry Nuclear Plant (BFN) unit 2.
This report summarizes the inspections, findings, and consequent mitigative and corrective actions.
This report supersedes that which was transmitted to D.
R. Muller, of NRC, on March 10, 1986.
Since the completion of the efforts reported in that document, additional concerns have been raised and new inspection and repair technologies have been developed. Also, the protracted length of the cycle 5 outage made it possible for additional inspections and corrective actions to be undertaken.
A summary of the inspections and corrective and mitigative actions is given in the tables of appendix 1.
These activities are discussed in-depth in other appendices, as outlined below.
The past and present piping configurations are illustrated in figures 1 through 6.
Summarv of Activities Inspections Ultrasonic inspections for IGSCC were performed early in the outage (fall 1984) on the stainless steel velds, susceptible to IGSCC, in piping systems equal to or greater than 4 inches in diameter and operating at temperatures 0
over 200 F.
The piping systems inspected were those that are part of or connected to the reactor coolant pressure boundary; all accessible, l
inspectable welds located inboard of the second isolation valve were inspected in each system.
The head spray system was not inspected because it was scheduled to be-removed at the time the inspections were performed.
It was not a required safety system and had never been used. The stainless steel portion of the system was removed from service later in the outage.
l
'8801050196 871231 PDR ADOCK 05000260 g
_ Induction Heating Stress Improvement (IhSI) was performed early in 1985.
After the performance of IHSI, 25 percent of the original inspection workscope was reinspected. The 25-percent sample was selected from those welds which required recording and evaluation of an indication; additional welds needed to complete the sample were chosen from weld locations shown to have a high propensity for IGSCC.
l Inspections were performed later in the outage following the discovery of IGSCC in a previously inspected reactor water cleanup (RWCU) weld. Weld DSRWC-2-5 was found to be flawed during a section XI inspection that was performed in October 1986. This finding lead to the inspection of all RWCU welds which had not received post-IHSI inspections earlier.
. The recirculation inlet nozzle safe ends and core spray safe ends were inspected in June 1986.
All 10 of the recirculation inlet safe ends were found to have indications of crevice cracking in the thermal sleeve attachmer.t area. To correct this, the safe ends and riser piping were replaced, as discussed below.
Baseline ultrasenic and radiographic inspections were performed on the new welds resulting from the replacement.
As a result of the extended outage length and the safe end replacement, a second IHSI implementation was undertaken in May 1987 to treat the new welds and those welds which were dropped or excluded from the first work scope.
Post-IHSI inspections were performed on 100 percent of the welds treated during this second effort.
All level 2 and 3 ultrasonic testing (UT) inspectors who performed these IGSCC or post-IHSI examinations were trained and certified for IGSCC detection by EPRI and had therefore demonstrated competance in accordance with IEB 83-02.
Details and results of the inspections outlined above are reported in appendix 2.
Disposition of Flawed Welds IHSI The 100-percent inspection reconfirmed the existence of cracking which had been discovered during cycle 4 in two sweepelet-to-recirculation manifold welds (KR-2-14 and KR-2-36).
Small indications were also detected in another sweepolet-to-recirculation manifold weld (KR-2-41) and in one end cap-to-recirculation manifold weld (KR-2-37).
Each of these 4 welds was found to have relatively short end shallow indications, the longest detectec indication being 4 inches and the deepest detected indication being 26-percent throughwall. These four welds were treated by IHSI to produce a f avorable residual stress pattern and reduce their susceptibility to further IGSCC degradation.
Fracture mechanics evaluations were performed to justify continued operation with these flawed welds, and these evaluations are provided in appendix 3.
The evaluations were based on the post-IHSI indication sizes which differed somewhat from the pre-IHSI indication sizes, but not significantly.
It is noteworthy to mention that two more small indications were found in welo KR-2-36 after IHSI was performed, and these indications were also considered in the fracture mechanics evaluations. The evaluations were performed in accordance with ASME Section XI, IWB-3640, and the recommendations of
I l Generic Letter 84-11.
These conventional approaches were also supplemented by J
elastic-plastic fracture mechanics tearing instability analyses to evaluate the possible effects of low toughness welds.
The results of the evaluations for all 4 welds indicate that design basis safety margins are maintained in the welds, by a Inrge margin, considering the worst case effects of the--
observed flaws; and these margins are maintained indefinitely during the life of the plant because of the beneficial effects of the IHSI treatment, which is expected to inhibit further IGSCC propagation.
Thus, these 4 welds are still fully qualified for at least one fuel cycle because the evaluation results i
indicate that the design basis safety margins are maintained..These 4 welds (KR-2-14, KR-2-36, KR-2-37, and KR-2-41) will be reinspected and reevaluated for continued operation during the next (cycle 6) outage.
Ove rlays After the IHSI treatment was applied to GR-2-15, a weld which joins a 28-inch by 12-inch reducer to a 12-inch riser line, a throughwall leak was discovered on the reducer side of the weld. The geometry of the 12-inch end of the reducer renders this area uninspectable.
It is believed that an almost throughwall indication existed before applying IHSI and that the IHSI treatment caused this to open up.
Weld GR-2-15 was repaired per ECN P-5215 by j
the application of a full structural overlay which was required to have an effective thickness of 0.35-inch.
The overlay was designed in accordance with the recommendations of Generic Letter 84-11.
The structural justification for the everlay repair, including design details and an independent review of the design, is provided in Appendix 4.
The structural justification and independent review indicate that design basis safety margins are maintained by the overlay for at least two complete cycles of operation, even if a 360-degree throughwall flaw exists in the original weld.
Therefore, the overlay design on GR-2-15 is fully qualified for at least two complete cycles of operation. This is confirmed by the Unreviewed Safety Question Determination (USOD) which was processed under ECN P-5215.
The axial shrinkage resulting from the overlay on veld GR-2-15 was originally estimated to be 0.209-inch.
The original US0D in ECN P--5215 addresses the impact of this shrinkage on the recirculation inlet nozzles and dispositions the shrinkage of 0.209-inch as being acceptable.
On completion of the overlay, the axial shrinkage was measured to be 0.2811-inch, and the USOD was revised to address the impact of the measured shrinkage on the recirculation inlot nozzles. The revised USOD dispositions the shrinkage of 0.2811-inch as being acceptable.
The evaluations of sweepolet welds KR-2-14, KR-2-36, and KR-2-41, as presented in appendix 3, address the estimated axial shrinkage of 0.209-inch at veld overlay CR-2-15 by considering the resulting stresses at j
the sweepolet welds. These stresses are small in comparison to either the j
operational (sustained) loading stresses or the accident (short term) loading stresses. The measured axial shrinkage of 0.2811-inch is approximately 34 percent larger than the estimated shrinkage of 0.209-inch, and the resulting increase in the small shrinkage stresses at the sweepolet welds will be approximately 34 percent, assuming elastic behavior. A 34-percent increase in the shrinkage stresses is'very small in comparison to the operational stresses, and on observation of the stress intensity factors in appendix 3, it is easily seen that a 34-percent increase in shrinkage stresses will not invalidate the previously predicted zero crack growth for I
the balance of plant life.
Also, the stability calculations for welds KR-2-14, KR-2-36, and KR-2-41 include a safety factor of 2.773 on the shrinkage stresses, while a safety factor of 1.0 would be adequate because these stresses are secondary. The extra margin provided by the safety factor of 2.773 more than compensates for a 34-percent increase in shrinkage stresses; therefore, the results of the fracture mechanics evaluations in appendix 3 will not be invalidated by the larger measured shrinkage.
In October 1986, an indication of IGSCC was detected in RWCU pipe-to-elbow l
weld DSRWC-2-5.
The indication was approximately 4.5 inches long and
)
0.300 inches deep. This represented a crack length of 23 percent of the pipe j
circumference and a depth of 70 percent of the pipe wall thickness. The weld j
,was repaired per ECN P-0997 by the application of an overlay with an effective
]
thickness of 0.200-inch.
The overlay was designed in accordance with the requirements of the draft of NUREG-0313, revision 2.
The structural justification for the repair, including design details, is provided in appendix 5.
A special requirement was included in ECH P0997, revision 2, to ensure that
]
the axial shrinkage as a result of the overlay on weld DSRWC-2-5 is within acceptc.ble limits. The shrinkage was determined to be 0.401-inches by using
]
measurements which were taken at four equally spaced locations both before and after application of the overlay. Loads on the affected reactor pressure vessel nozzles, recirculation pump nozzles, piping, and supports as a result of the shrinkage of 0.401-inches were determined in BFN Piping Analysis j
N1-268-lR, revision 2.
These loads were then combined with the existing loads and the combination of the loads were then evaluated and determined to be acceptable. Therefore, the shrinkage of 0.401-inches is judged to be within acceptable limits.
Replacement The jet pump instrumentation nozzle safe ends and reducers were inspected as part of the 100-percent inspection, and extensive cracking was found in the safe ends and reducers near welds JP-2-1A and JP-2-1B.
These components were replaced per ECH P-0720 with a new design using 316 NG stainless steel.
A section of RWCU piping crntaining pipe-to-elbow weld JkWC-2-4 was also replaced. During the 100-percent inspection, an axial crack approximately 3/4-inch 3tng and 50-percent throughwall was found near that weld. The piping was replaced in accordance with MR A-183065, using 304 stainless steel as replacement material Six new welds were installed as the result of this replacement effort (see RWCU table in appendix 1).
At the time of the replacement, there was no IHSI contractor onsite, so heat sink welding was used to produce a favorable residual stress pattern and reduce the susceptibility of the new welds to IGSCC.
Later in the outage, crevice cracking was discovered in the thermal sleeve attachment area of the recirculation inlet safe ends.
After a study of several repair options, it wa6 decided to replace the safe ends with the improved noncreviced design.
Because the replacement material available included long tangent elbows, most of the riser piping was replaced along with the safe ends.
This new configuration reduced the total number of welds in the recirculation system by 10.
The replacement material was 316 NG stainless
-- steel.
Details of the safe end inspection and replacement are given in
" Safe End Replacement:
A Utility Perspective," a technical paper which was presented at the October 1987 EPRI Seminar on Pipe Repair and Replacenaent.
This paper is attached as appendix 6.
It should be noted that a destructive examination of several safe end samples is currently in progress to confirm and further characterize the cracking mechanism.
Preliminary results indicate that the thermal sleeves as well as the safe ends contair. extensive IGSCC.
Analysis of the recirculation piping was performed to evaluate the replacement of the recirculation inlet nozzle safe ends and associated piping. Data taken from this analysis revealed that the moments and stresses in welds GR-2-15, KR-2-14, KR-2-36, and KR-2-41 changed slightly as compared to the moments and stresses for the original safe end configuration.
Calculations vere then performed to reassess the adequacy of the overlay design on weld GR-2-15 fcr the new moments and stresses and the impact of the new moments and stresses on the structural integrity of flawed welds KR-2-14, KR-2-36, and KR-2-41.
The calculations demonstrate that the design basis safety margins are still maintained by the overlay on GR-2-15 even if a 360-degree throughwall flaw exists in the original weld and that the structural integrity of flawed welds l
KR-2-14, KR-2-36, and KR-2-41 will not be impacted by the new moments and stresses. These calculations are presented in appendix 7.
l Other Weld DSRHR-2-5A, which is in the residual heat removal (RHR) system, was radlographel during this outage because the baseline radiographs were not found onsite or in permanent storage. The radiographs revealed linear indications which are located 0.2 and 0.6-inch below the outside surface.
The indications appeared to be lack of fusion in the circumferential direction with no appreciable throughwall dimension. The indication which is located 0.2-inch below the surface could be detected with UT.
The UT revealed that this indication does not originate at the inside surface and that there is n3 evidence of appreciable throughwall depth. The indication which is located 0.6-inch be]ow the surface could not be detected by conventional IGSCC detection and sizing methodology; therefore, this indication does not originate at the inside surface, and there is no evidence of appreciable throughwall depth.
It was concluded from the radiographs and UT that the indications wert not IGSCC but instead were lack of fusion between layers with no appreciable throughwall dimension; therefore, weld DSRHR-2-5A was left "as is."
IHSI In order to provide a level of protection against IGSCC, IHSI was implemented on all accessible susceptible welds in the recirculation, RHR, RWCU, and core spray piping systems.
Two separate IHSI efforts were completed during the outage.
From December 1984 to March 1985, 149 velds were given IHSI treatments by the General Electric Company. This work scope did not include the recirculation inlet or outlet safe end welds or the core spray safe end welds.
Additionally, a RWCU and four recirculation welds were dropped from the work scope because of technical difficulties encountered during the treatment.
The startup schedule in effect at the time of this first IHSI effort called for completion of IHSI by the end of March; the job was therefore completed according to that schedule. The unexpected
- --__- -. - _. _ _ - - extension of the outage, however, presented an opportunity to perform IHSI on the welds omitted or dropped from the earlier work scope.
Also, IHSI was called for in the safe end and riser replacement plans to achieve an additional measure of IGSCC protection for the new welds.
For these reasons, a second IHSI effort was undertaken in May 1987 by NUTECH Engineers, Inc..The work scope for this job included the new recirculation inlet safe end and riser welds, recirculation outlet safe end welds, core spray safe end welds, and the recirculation and RWCU welds that were dropped from the previous work scope. With the end of this effort in June 1987, all IHSI commitments and plans for BFN unit 2 were completed. The details of both IHSI efforts are reported in appendix 8.
Other Issues Remaining Post-IHSI Inspections
References:
- 1) Letter from R. Gridley (TVA) to D. R. Muller (NRC) dated November 10, 1986.
2)
Letter from Marshall Grotenhuis (NRC) to S. A. White (TVA) dated March 26, 1986.
Item 2-E of Generic Letter 84-11 states that all welds treated by IHSI must be
" post treatment UT acceptance tested"; however, in the referenced correspondence it was agreed that TVA could perform post-IHSI inspections on less than 100 percent of the IHSI treated welds providing that the remaining welds were inspected at the next refueling outage. The inspections which have been performed during tne cycle 5 outage are detailed in appendix 2 and summarized in appendix 1.
The table in appendix 1 indicates which welds received post-IHSI volumetric examinations. Approximately.54 percent of the welds treated with IHSI were UT or RT examined following the IHSI treatment; therefore, the remaining 46 percent of the IHSI-treated welds will receive post-IHSI examinations during the cycle 6 refueling outage.
Leak Detection and Leakage Limits
References:
- 1) Letter from John A. Zwolinski (NRC) ta S.
A. White (TVA) dated August 26, 1987.
- 2) Memorandum from J. A. Domer (TVA) to D. R. Miller (NRC) dated December 15, 1986. of Generic Letter 84-11 gives the requirements for reactor coolant leakage detection systems.
Section B requires that for sump level monitoring systems, "the level shall be monitored at 4-hour intervals or less," and section C limits the outage time for inoperable leakage measurement instruments to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TVA has complied with these requirements by changing the technical specifications of all three BFN units. The technical specification changes were submitted to NRC in reference 2; and NRC accepted the proposed changes in reference 1.
BFN will monitor leakage in accordance j
with the referenced technical specifications. No additional actions are required in this matter.
j Section E requires the performance of a visual examination for leakage each time the containment is deinerted; however, recent statements made by Mr. G.
E. Gears, NRC, BFN Project Manager, indicate that this requirement has been retracted.
Improvements in IGSCC detection techniques have increased NRC's confidence in the utilities' knowledge of piping integrity sufficiently to make this requirement unnecessary. Additionally, DFN's improvement in drywell leakage detection, as discussed above, further reduces the need for extra visual inspections.
BFN will continue to perform system pressure tests in accordance with the ASME Section XI System Pressure Test Program.
Conclusion This report has summarized the actions taken at BFN unit 2 in response to the requirements of Generic Letter 84-11.
With the exception of the remaining post-IHSI examinations mentioned above, all required piping inspections have been completed. All stress improvement and piping replacement work also has been completed, where indicated. The two overlays which have been installed during this outage will be inspected in accordance with appropriate governing regulations.
Planned raitigation activities include the installation of hydrogen water chemistry and the replacement or corrosion resistant cladding of the penetration welds (see appendix 1).
These measures are being planned for implementation during or before the cycle 6 refueling outage.
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l Figure 1 Recirculation System Before Safe End Replacement Figure 2A Current Recirculation System, Loop A Figure 2B Current Recirculation System, Loop B Figure 3 Current Recirculation System, j
l (Isometric)
I Figure 4A Residual Heat Removal System Figure 4B Residual Heat Removal System (continued)
Figure 5 Core Spray System Figure 6 Reactor Water Cleanup System
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SUMMARY
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r r-APPENDIX 2 l
IGSCC EXAMINATION RESULTS 2678U
-~--~' ' ' ~ ~~
./.
FIRST IGSCC EXAMINATION EFFORT 1984 - 1985 SYSTEM:
Core Spray PRE:
IllSI Weld Noe t!cthod Size Scan Limitation Evaluated Results DCS-2-1 UT45/PT 12 No Scan 3 DCS-2-2 UT45 12 No' Scan 4 Satisfactnf DCS-2-3 UT45/PT 12 No Scan 3 or 4 Sa tis factav Sa tis factmy DCS-2-4 UT45 12 SatisfactM DCS-2-5 UT45 12 No Scan 4 DCS-2-7 UT45 12 Satisfacten DCS-2-10 UT45/PT 12 No Scan 3 Satisfactory Satisfactory DCS-2-11 UT45 12 No Scan 4 DCS-2-12 UT45/PT 12 No Scan 3 or 4 Satisfactory r
DCS-2-13 UT45 12 Satisfactog Satisfactog DCS-2-14 UT45 12 No Scan 4 DSCS-2-1 UT45 12 Satisfactoq.
DSCS-2-2 UT45 12 Satisfactog DSCS-2-9 UT45 12 Satisfactory DSCS-2-14 UT45 12 Satisfactmy DSCS-2-15 UT45 12 Satis facto:y TCS-2-401 UT45/PT 10 Satisfactmy TCS-2-403 UT45/PT 10 Satisfactory TCS-2-405 UT45/PT 12 Satis factmy No Scan 3, Scan 4 Satisfactrey limited @ 2:30 and 3:30 TCS-2-406 UT45/PT 12 No Scan 4 Sa tis factmy TCS-2-410 UT45/PT 12 No Scan 3 TCS-2-417 UT45/PT 10 Satisfaetery TSCS-2-418 UT45/PT 10 Sa tis fa ct.ecy Saticfactery TCS-2-421 UT45/PT 12 No Scan 3, Scan 4 Sa tis factocy limited @ 8:30 to 9:00 TCS-2-422 UT45/PT 12 No Sean 4 TCS-2-426 UT45/PT 12 No Scan 3 Sa tis factory Satistacte:y 6
e S
e
' BFN UNIT 2 CYCLE 5 OUTAE RPT.
SYSTEM:
Recirculation SERIES:
KR
~
PRE:
IHST-
~
Weld No.
Method Size Scan Limitation Evaluated Results KR-2-1 PT 4
No d~ due to Sa tis f acto ry -
T configuration KR-2-2 UT45 28 Sa tis facto ry KR-2-3 UT45 28 No Scan 4 Satisfactory KR-2-4 PT 4"
No tfr due to sa tis f actory configuration-KR-2-11 UT45 28 No Scan 3 Satis fa ctory KR-2-12 UT45 22 No Scan 3 Sa tisfa ctory KR-2-13 UT45 22 Limited contact Sa tis factory Scan 4 KR-2-15 UT45 22 Sa tis facto ry KR-2-16 UT45 12 Sa tis f a ctory KR-2-17 UT45 12 Satisfactory KR-2-18 UT45 12 Satisfactory KR-2-19 UT45 22 Limited contact Satisfactory Scan 4 KR-2-20 UT45 22 Limi.ted contact Sa tis fa ctor*-
Scan 4 KR-2 UT45 12 Sa tis fa ctory KR-2-22 UT45 12 Satisfactory KR-2-23 PT 4
No UT due to satisfactory configuration KR-2-24 UT45 28 Satisfactory KR-2-25 UT45 28 No Scan 4 Sa tis f actory KR-2-26 PT 4.
No UT due to Satisfactory configuration KR-2-33 UT45 28 No Scan 3 Sa. tis f a cto ry KR-2-34 UT45 22 No Scan 3 Satisfactory KR-2-35 UT45 22 Limited contact Satisfactory Scan 4 KR-2-38 UT45 12 Sa tis factory KR-2-39 UT45 12 Sa tis f a ctory KR-2-40 UT45 12 Satisfactory KR-2-42 UT45 22 Sa tis facto ry KR-2-43 UT45 12 KR-2-44 UT45 12 Sa tis f a ctory Satisfactory-KR-2-45 UT45 28 Satisfactory KR-2-46 UT45 28 No Scan 3' Satisfactory KR-2-47 UT45 28 Satisfactory KR-2-48 UT45 28 Satisfactory KR-2-49 PT 4
No UT due to Satisfactory confir,uration KR-2-50 UT45 28 Satisfactory KR-2-51 UT45 28 Satisfactory KR-2-52 UT45 28 Sa tis f a ctory KR-2-53 PT
SYSTEM: Recirculation SERIES: GR PRE:
IHSI-Weld No.
Method Size Sean Limit'ation Evaluated Results GR-2-1 UT45 28 No Sc'an 3 Sa tis fa ctory GR-2-2 UT45 28 No Scan 4 Satisfactory GR-2-3 UT45 28 No Scan 3 -
Sa tis f actory GR-2-4 PT 4
No UT performed due Satisfactory to configuration GR-2-7 PT 4
No UT performed due Sa tis factory to configuration GR-2-8 UT45 28 No Scan 3 or 4 Satisfactory GR-2-9 UT45/PT 12 No Scan 3 Sa tis fa cto ry GR-2-10 UT45 12 Satisfactory GR-2-11 UT45 12 Sa tis fa cto ry GR-2-12 UT45 12 No Scan 3 Satisfactory GR-2-13 UT45 12 Sa tis factory GR-2-14 UT45 12 Satisfactory GR-2-16 UT45 12 Scan 3 limited due Satisfactory to weld buildup GR-2-17 UT45 12 Sa tis fa ctory GR-2-18 UT45 22 No Scan 3 Satisfactory GR-2 UT45 12 No Scan 3 Satisfactory.
GR-2-20 UT45 12 Satisfactory GR-2-21 UT45 12 Satisfactory GR-2-22 UT45 12 No Scan 3 GR-2-23 UT45 12 Satisfactory GR-2-24 UT45 12 Satisfactory Satisfactory GR-2-25 UT45 22 No Scan 4 Sa tis f actory
,GR-2-26 UT45 22 No Scan 3 Satisfactory GR-2-27 UI45 28 No Scan 3 Sa tis factory GR-2-28 UT45 28 No Scan 4 Satisfactory GR-2-29 UT45 28 No Scan 3 Sa tis f acto ry GR-2-30 PT 4
No UT performed due Sa tis factory to configuration GR-2-33 PT 4
No UT performed due Sa tis fa ctory to configuration GR-2-34 UT45 28 No Scan 3 or 4 Sa tis f a ctory GR-2-35 UT45 12 Scan 4 limited GR-2-36 UT45 12 Satisfactory GR-2-37 UT45 12 Sa tis fa ctory Satisfactory GR-2-38 UT45 12 No Scan 3 GR-2-39 UT45 12 Satisfactory GR-2-40 UT45 12 Satisfactory Satisfactory GR-2-41 UT45 12 No Scan 3 Satisfactory GR-2-42 UT45 12 Satisfactory GR-2-43 UT45 12 Satisfactory GR-2-44 UT45 22 No Scan 3 GR-2-45 UT45 12 Satisfactory GR-2-46 UT45 12 Satisfactory GR-2-47 UT45 12 Sa tis f a ctory Sa tis f a ctory BFN UNIT 2 CYCLE 5 OUTAGE RPT.
SYSTEM:
Recirculation SERIES:
GR PRE:
IHSI-k' eld No.
Method Size Scan Limitation Evaluated Results
~
GR-2-48 UT45 12 No Sca'n 3 Satisfactory GR-2-49 UT45 12 Satisfactory GR-2-50 UT45 12 Satisfactory GR-2-51 UT45 22 No Scan 4 Sa tis fa ctory GR-2-52 UT45 22 No Scan 3 Satisfactory GR-2-53 UT45 28 No Scan 3 Satisfactory GR-2-54 UT45 28 Satisfactory GR-2-55 UT45 28 No Scan 4 Satisfactory GR-2-56 UT45 28 No Sean 4 Satisfactory GR-2-57 UT45 28 No Scan 3 Satisfactory s
GR-2-58 UT45 28 No Scan 4 Satisfactory GR-2-59 UT45 28 Satisfactory GR-2-60 UT45 28 Satisfactory GR-2-61 UT45 28 Satisfactory GR-2-62 UT45 28 No Scan 4 Satisfactory GR-2-63 UT45 28 No Scan 3 Sa tis f a ctory GR-2-63A PT 4
No UT performed due Satisfactory to configuration GR-2-63B PT 4
No UT performed due Satisfactory to configuration GR-2-64 UT45 28 No Scan 4 Satisfactory j
i D
h
. BFN UNIT 2 CYCLE 5 0UTAGE RPT.
- e SYS*EM: Residual Heat Removal SERIES: DSRHR
~
PRE:
~~
~
Held No.
Method Size Scan Limitation Evaluated Results
~
DSRHR-2-1 UT45 24
~
DSPJIR-2-2 UT45 24 Sa tis fa ctory i
DSRHR-2-3 UT45 24 Satisfactory DSRHR-2-4 UT45 24 Satisfactory DSPJB-2-4A UT45 24 Satisfactory DSRHR-2-5 UT45 24 No Scan 4 Satisfactory DSPJm-2-5A UT45 24 No Scan 3, limited Sa tis factory Satisfactory Scan 4 DSRHR-2-6 UT45-24 DSRHR-2-7 UT45 24 Satisfactory DSRHR-2-8 UT45/PT 6
No Scan 4 Satisfactory DSRHR-2-9 UT45 20 Satisfactory DSRHR-2-10 UT45 20 Satisfactory DSRHR-2-11 UT45 20 Satisfactory Satisfactory s
4 6
4 e
t e BFN UNIT 2 CYCLE 5 0UTAGE RPT.
9
SYSTEM: Residual Heat Removal SERIES: DRRR
~
PRE:
Method Size Scan Limitation Evaluated Results DPER-2-2 UT45/PT 24 No Sc'an 4 Sa tis factory DRHR 1-3 UT45/PT 24 No Scan 3 or 4 DRHR-2 4 UT45/PT 24 Sa tis factory Sa tis fa cto ry DRHR-2 3 UT45 24 No Scan 4 Sa tis factory DRHR-2-6 UT45 24 No Scan 3 Sa tis fa ctory DRHR-2-7 UT45/PT 24 No Scan 4, 5, an'd 6 Sa tis factory DRHR-2-8 UT45 24 No Scan 3 Satisfactory DRHR-2-9 UT45 24 No Scan 4 Sa tis fa ctory DRHR-2-11 UT45/PT 24 No Scan 4 Satisfactory DRHR-2-12 UT45/PT 24 No Scan 3 or 4 DRHR-2-13
'UT45 24 Sa tis fa cto ry Satisfactory DRHR-2-14 UT45 24 No Scan 4 Satisfactory DRHR-2-15 UT45 24 No Scan 3 Sa tis factory DRHR-2-16 UT45 24 No Scan 4 Sa tis fa ctory DRHR-2-17 UT45 24 No Scan 3 Sa tis fa cto ry DRER-2-18 UT45 24 No Scan 4 Sa tis fa cto ry DRHR-2-19 UT45 20 No Scan 3 Sa tis facto ry DRHR-2-21 UT45 20 No Scan 4 Sa tis fa c to ry.
~~
DRHR-2-22 UT45 20 No Scan 3 Satisfactory DRHR-2-23 UT45 20 No Scan 4 Satisfactory
. TRER-2-191 UT45/PT 20 No Scan 3 Satisfactory l
M.
e W
S 1
l l
\\
. DFN UNIT 2 CYCLE 5 0UTAGE RPT.
____-____b
l l
SYSTEM: Reactor Water Cleanup SERIES:
DRWC and DSRWC PRE:
IHSI,
Weld No.
Method Size Sean Limitation Evaluat :d Re.sults
~
DRWC-2-1A UT45/PT 6
No Scan 3 and 4
~
DRWC-2-1 UT60 6
No Scan 3 Sa tis factory DSRWC-2-1 UT60 6
Sa tis factory Sa tis factory DRWC-2-2 UT60/PT 6
No Scan 3, limited Sa tis f actary
)
. Scan 4 from 11:30 - 12:30-DRWC-2-3 UT60/PT 6
No Stan 3, limited Sa tis facttry Scan 4 from 10:00 - 1100 and 5:30 -6:30 DSRWC-2-1A UT60 6
DSRWC-2-2 UT45 6
Scan 3 limited from Satisfactory Satisfactory 5:00 - 7:00 4
DSRWC-2-3 UT60 6
Scan 4 limited from Satisfactory 4:30 - 7:00 DSRWC-2-4 UT60 6
Scan 3 limited from Sa tis factory 5:00 - 7:00 DSRWC-2-5 UT60 6
DSRWC-2-6 UT60 6
Sa tis fa ctory Satisfactory DSRWC-2-7 UT60 6
Scan 4 limited from Sa tis factory 4:00 - 8:00 DRWC-2-5A UT45 6
No Scan 3 DRWC-2-5B UT45 6
No Scan 4 Ea tis factory Sa tis fa ctory
=
9 e
O 9
i i
i
. BFN UNIT 2 CYCLE 5 00 TACE RPT..
e SYSTEM:
Core Spray Post I}ISI Examination Weld Noe Method Size Scan Limitation Evaluated Results DCS-2-5 UT45 12 No Scan 4 Satis factory DCS-2-7 UT45 12 Sa tis f actory -
DCS-2-13 UT45 12 Satis factory e
em e.
A e
]
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l BFN UNIT 2 CYCLE 5 OtJTAGE RPT.
SYSTEM: Reactor Water Cleanuo Post IHSI Examination Weld No.
Method Size Sean Limitation Evaluated Results DRWC-2-1 UT60/PT 6
No Scan 3 Sa tis facto ry DRWC-2-1A UT45 6
No Scan 3 and 4 DRWC-2-2 PT 6
Satisfactory DRWC-2-3 PT 6
Satisfactory-Sa tis factory DRWC-2-5A UT45/PT 6
No Scan 3 Satisfactory DRWC-2-5B PT 6
Satisfactory DSRWC-2-1 UT60 6
Satisfactory DSRWC-2-2 UT45 6
Limited Scan 3-Satisfactory l
y 1
I 9
l l
5 6
f gpN UNIT 2 CYCLE 5 0UTAGE RPT.
L____l_________._________._-___________
SYSTEM:
Recirculation SERIES:
GR Post IHSI Examinations Weld No.
Method S_ize Scan Limitation Evaluated Results i
GR-2-1 UT45 28 No Scan 3 GR-2-4 PT 4
No UT due to Sa tis fa ctory Satisfactory configuration GR-2-7 PT 4
No UT due~to Satisfactory configuration GR-2-9 UT45/PT 12 No Scan 3 GR-2-11 UT45 12 Sa tis fa ctory Sa tis fac to ry.
GR-2-12 UT45/PT 12 No Scan 3 GR-2-14 UT45 12 Satisfactory GR-2-17 UT45 12 Sa tis factory Satisfactory GR-2-18 UT45/PT 22 No Scan 3 GR-2-19
'PT 12 Satisfactory GR-2-20 UT45 12 Sa t i s f a c to ry.
GR-2-22 PT 12 Satisfactory Sa tis factory GR-2-30 PT 4
No UT due to Sa tis fa ctory configuration GR-2-34 PT 28 GR-2-35 PT 12 Sa tis factory GR-2-38 PT 12 Satisfactory GR-2 PT 12
~~
Satisfactory GR-2-44 PT 22 Sa tis factory
.GR-2-45 PT 12 Satisfactory GR-2-48 PT 12 Satisfactory Sa tis fa c tory l
GR-2-51 UT45 22 No Scan 4 GR-2-53 PT 28 Satisfactory 1
GR-2-55 PT 28 Satisfactory GR-2-59 PT 28 Sa tis f actory GR-2-60 UT45 28 Satisfactory GR-2-62 UT45 28 No Scan 4 Satisfactory GR-2-63A PT 4
No UT due to Sa tis f a cto ry Satisfactory configuration GR-2-63B PT 4
No UT due to Satisfactory configuration GR-2-64 UT45 28 No Scan 4 Sa tis fa ctory l
l I
, BFN UNIT 2 CYCLE 5 0UTAGE RPT.
~
E_
e SYSTEM: Recirculation SERIES: KR Post IHSI Examination
~
~
Feld No.
Method Size Scan Limitation Ev-luated Results KR-2-1 PT 4
No UT performed due Sa t is fac tory to configuration KR-2-2 UT45 28 KR-2-3 PT 28 Satisfactory KR-2-4 No Re-exam 4
.}h) IHSI performed Sa tis fa c tory due to configuration KR-2-11 PT 28 KR-2-15 UT45 22 Sa tis factory KR-2-23 PT 4
No UT performed due Sa tis f a cto ry Sa tis factory to configuration KR-2-25 PT 28 KR-2-26 No Re-exam 4
Sa tis fa ctory No IHSI performed due to configuration KR-2-33 PT 28 KR-2-34 UT45/PT
- 2ss No Scan 3 Sa tis f a ctory KR-2-35 UT45 Sa tis f a ctory 22 Limited, contact Satisfactory Scan 4 KR-2-43 UT45 12 Sa t is f a ctory KR-2-45 UT45 28 Satisfactory KR-2-46 PT 28 Satisfactory KR-2-49 PT 4
No UT performed due Satisfactory to configuration KR-2-53 PT 12 No UT performed due satisfactory l
to configuration 4
I
~
i I
e 4
0 BFN UNIT 2 CYCLE 5 OUTAGE RPT.
- -----~~-
^
~
SYSTEM:
Residual Heat Removal Post IHSI Examination Weld No.
Method Size Scan Limitation Evaluated Results DRHR-2-4 UT45 24 DRHR-2-5 PT 24 Sa tis facto ry Satisfactory DRHR-2-6 UT45 24 No Scan 3 Satisfactory DRHR-2-9 PT 24 Sa tis facto ry DRHR-2-13 PT 24 Satisfactory DRHR-2-17 UT45/PT 24 No Scan 3 Sa tis fa ctory DRHR-2-18 PT 24
~
Satisfactory DRHR-2-19 PT 24 Sa tis fa ctory DSRHR-2-4 UT45 24 Sa tis facto ry DSRHR-2-5 PT/RT/UT45 24 No Scan 4 Satisfactory DSRHR-2-6 UT45 24 Satisfactory DSRHR-2-8 UT45/PT 6
No Scan 4 Sa tis f acto ry DSRHR-2-9 UT45 20 Satisfactory
~
i
\\
. BFN UNIT 2 CYCLE 5 0UT,WE RPT.
_ _ _ _ _ _ _ _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - ~ ~ ~ - ~ ~
Jet Pump Instrumentation Nozzle Velds k' eld No.
Method Scan Limitation Evaluated Results JP-2-1A UT45/PT
~
j Numerous linear indications
~
both circumferential and axially. Indicaticas were near throughwall. One throughwall indication was one inch long.
JP-2-1B UT45/PT Numerous linear indications both circrmferential and
~
axially. Indicatives were near throughwall.
, 2A UT45 No Scan 4 Sa tis (actory
' 2-2B UT45 No Sean 4 Sa tis f a cto r,f 3A UT45 Satisfactory
.'P-2 -3 B UT45 Sa tis factory 2-2-4A UT45 Sa tis factory 1
i JP-2-4B UT45 Sa tis factory 1
O b
i I
e l
1 l
l 1
! BFN UNIT 2 CYCII 5 0UTAGE RPT.
S'; TEM: Recirculation V '.umetric Indications
~
Weld numbers KR-2-14 and KR-2-36 had previous reported indications during the Unit 2, Cycle 4 Outage.
Outage by TVA persoonel qualified to IE Bulletin 83-02 for the det
- SCC.
The subject welds were re-examined hnd sized using advanced ultrasonic string techniques.
The results are as follows:
Outage 0' Clock Weld No.
Examined Throuthwall Length Position Scan No.
r2-2-14 U2/C4 20%
1.25" 1:30 Scan 3 side i.-2-14 U2/C5 Pre IHSI 15%
2.1" 1:30 Scan 3 side
, 14 U2/C5 Post IHSI 12.6%
2.1" 1:30 Scan.3 side 36 U2/C4 20%
1.25 7:30 Scan 3 side U2/C4 20%
1.25 4:00 Scan 3 side l
U2/C4 20%
1.25 4.30 Scan 3 side
- 36 U2/C5 Pre IHSI 10%
1.2 4:00 Scan 3 side U2/C5 Pre IHSI 10%
1.2 4:30 Scan 3 side U2/C5 Pre IHSI 26%
" 1.1 6:00 Scan 3 side U2/C5 Pre IHSI 22%
2.2 7:30 Scan 3 side U2/C5 Pre IHSI 10%
2.3 10.00 Scan 3 side FR-2-36 U2/C5 Post IHSI 15%
1.2 4:00 Scan 3 side U2/C5 Post IHSI 15%
1.2 4:30 Scan 3 side U2/C5 Post IHSI 15%
1.1 6:00 Sean 3 side U2/C5 Post IHSI 25%
2.2 7:30 Sean 3 side U2/C5 Post.IHSI 14%
1.0 8:00 Scan 4 side U2/C5 Post IHSI 10%
2.3 10:00 Sean 3 side 1
U2/C5 Post IHSI 17%
1.0 1:30 Scan 4 side l
Two additional indications were found af ter IHSI in the Scan 4 direc o' clock positions 1:30 and 8:00.
The 12:00 o' clock position was assumed from the outside radius of the nearest riser to elbow weld.
l l
I i
l t
I
i
~~
l KR-2-37 Indications are specified below with througbwall dimensions md l
lengths.
i Pre IHSI 0' Clock Position, Length
% Throuchwall Scan Direction 2:00 4"
~
20%
Sem 4 2:30 1"
20" Sem 4 10:00 2"
10%
Sem 3 Post IESI 2:00 4"
12". -
Sem 4 2:3G 1"
9%
Sem 4 10:00 2"
7%
Sem 3 10:30 2"
8% -
Sem 4 7:00 1"
4%
Sema3 The 12:00 o' clock position was assumed @ top center looking downstre the end cap.
Two additional indications were observed a.f ter IHSI.
KR-2-41 lengths. Indications are specified below with throughwall dimensiens md
\\
Pre IHSI 0' Clock Position Leneth
% Thtcuchwall Scan Direction 6:00
.75
~
12",
Sc:a 3 8:00
.5 12*.
Sem 3 11:00 4.
12".
. Scan 3 l
P_ost IHSI 6:00
.75 19*.
Scan 3 i
i 8:00
.5 19%
11:00 4.
Scan 3 19%
Scan 3 e
W I
e BFN L?llT 2 CYCLC 5 0UTAG RPT.
a J
)
SYSTEM: Reactor Water Cleanup PRF.:
IHSI DRWC-2-4 One linear indication indicative of IGSCC was located on the pipe side of the weld at the 11:00 o' clock position.
i major dimension parallel to the pipe axis. Indication length was.75 inch long with 50'4 Throughwall loss was approximately POST-IHSI SYSTEM: Recirculation GR-2-15 Weld GR-2-15 was evaluated prior to IHSI as having no apparent disconuities Accessibility was limit.ed to the Scan 4 direction which had a weld buildup on the pipe side of the weld.
The forging side was limited in contact due to the configuration.
A throughwall linear indication was observed after IHSI as having a dimension of 1" at the 4:00 o' clock position on the forging side of the veld.
forging side due to limitations caused by the configuration.The i 4
. BFN UNIT 2 CYCLE 5 0UTAGE RPT.
Post IHSI Surface Examinations DRWC-2-1A Reactor Water Cleanup System Linear indicatica observed at 2:30 o' clock p,osition with dimensions.625 inches lcag located.21 inches away from the weld.
Indication removed by mechanical method, Acceptable Report 0407-C 7-2-85.
GR-2-8 Recirculation System Linear indication observed at the 8:00 o'cl'ock position.4 inches long located
.5 inches from the toe of the weld. Indication removed by mechanical method.
Acceptable Report R-0658-B 6/21/85, KR-2-12 Recirculation System Linear indication observed at the 6:
located.375 inches from the toe of the weld.00 o' clock position.125 inches long Indication renoved by mechanical method. Acceptable Report R-0543-B 3/15/85.
DSRER 2-5A Residual Heat Removal System Volumetric Indication The subject weld was radiographer during th6 Unit 2, Cycle 5 Outage because baseline radiographs was not found on site or in permanent storage.
During the evaluation process, a linear indication was noted in radiograph segment 3-4 using AA film.
The indication was re-radiographer using a more sensitive M film which revealed a new indication not previously identified using AA film.
The entire lengths of welds DSRRR-2-5A and DSRER-2-5 were re-radiographer using Kodak M film.
The results of weld DSRER-2-5 are acceptable however weld DSRER-2-5A had linear indications in the circumferential direction with an aggregate length of 45 inches (46 percent of the circumferential length of the weld affected). The indications are indicative of lack of fusion.
The parallax shot of one segment indicates the indications to be in a plane.6 inches below the OD surface in radiograph segments (1-2), (3-4), (4-5), (5-6), (6-1) and.2 inches below the CD surface in segment (2-3), (3-4).
The indication located at.2 inches below the j
surface (segments 2-3, 3-4) can be detected with special UT.
I
.6 inches below the surface cannot be detected by conventional IGSCCThe ind at detection and sizing UT methodology also indicating a small reflective
- surface, i.e., small defect.
G G BFN UNIT 2 CYCLE 5 0UTAGE RPT.
f j
IGSCC EXAMINATION 0F REACTOR WATER CLEAN UP SYSTEM - FALL 1986 Weld.
Nominal Pipe NDE Procedure Scan No.
Size (Inches)
Method No.
Performed Results (Comments)
DRWC-2-1A 6"
PT N-PT-01 5 & 6 only Acceptable UT45 BF-UT-17 due to valve DR'WC-2-2 6"
PT N-PT-01 Acceptable DRWC-2-2 6"
UT45 N-UT-25 No scan 4, (Post-IHSI)
DRWC-2-2 6"
UT60 N-UT-25 scan 3 limited DRWC-2-3 6"
PT N-PT-01 Acceptable DRWC-2-3 6"
UT45 N-UT-25
" (No scan 3, (Post-IHSI)
DRWC-2-3 6"
UT60 N-UT-25 scan 4,11mited 5:30 - 6:30)
DSRWC-2-1A 6"
UT45 N-UT-25 (Scan 3 limited A N7ptable DSRWC-2-1A 6"
UT60 N-UT-25 7:00 - 8:00)
(1 h -IHSI)
DSRWC-2-2 6"
UT45 BF-UT-17 Acceptable (Post-IHSI)
DSRWC-2-3 6"
UT45 BF-UT-17 Acceptable (Post-IHSI)
DSRWC-2-4 6"
UT45 BF-UT-17 Scan 3 Acceptable limited (Post-IHSI) 5 to 7:00 DSRWC-2-5 6"
UT45 BF-UT-17 Reject (4 1/2" linear indication)
Post-IHSI leading to overlay DSRWC-2-5-OL 6"
UTO N-UT-28 Acceptable (overlay).
DSRWC-2-5-OL 6"
UT70 N-UT-28 Acceptable (overlay)
DSRNC-2-6 6"
UT45 BF-UT-17 Acceptable (Post-IHSI; DRWC-2-7A E"
UT45 BF-UT-17 Acceptable (New weld-PSI)
DRWC-2-7A 6"
UT45 N-UT-25 Acceptable (New weld)
DRWC-2-7B 6"
UT45 BF-UT-17 Acceptable (New I
Weld-PSI)
DRWC-2-7B 6"
UT45 N-UT-25 Acceptable (New weld)
DSRWC-2-7X 6"
UT45 BF-UT-17 Acceptable (New weld-PSI)
DSRWC-2-7X 6"
UT45 N-UT-25 Acceptable (New weld)
DRWC-2-4X 6"
UT45 BF-UT-17 Acceptable (New Weld-PSI)
DRWC-2-4X 6"
UT45 N-UT-25 Acceptable (New weld)
DRWC-2-4X 6"
UT60 N-UT-25 Acceptable (New weld) 2678U J
IGSCC EXAMINATION OF REACTOR WATER-CLEAN UP SYSTEM - FALL 1986 (Continued)
Weld Nominal Pipe NDE Procedure Scan HL__
Size (Inches)
Method No.
Performed Results (Comments)
DRWC-2-4B 6"
UT45 N-UT-25 Acceptable (New weld)
DRWC-2-4B 6"
UT45 BF-UT-17 Acceptable (New DRWC-2-4A 6"
UT45 BF-UT-17 weld-PSI)
Acceptable (New weld-PSI)
DRWC-2-4A 6"
UT45 N-UT-25 Acceptable (New weld)
DRWC-2-5A 6"
PT N-PT-01 Acceptable (Post-IHSI)
DRWC-2-5A 6"
UT45 H-UT-25 No scan 3 Acceptable (Post-IHSI)
DRWC-2-5B 6"
PT N-PT-01 Acceptable (Post-IHSI)
DRWC-2-5B 6"
UT45 N-UT-25 No scan 4 Acceptable valve (Post-IHSI) 2678U
" POST-IHSI EXAMINATIONS OF REPLACEMENT PIPING AND OTHER WORK Weld Nominal Pipe NDE Procedure Scan E2x_
Reaion Size (Inches)
Method No.
Performed Results 2RA1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RA1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RAl AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RA1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RA1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RA1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RA1 SUR 14 PT NPT 01 Satisfactory 2 RAS IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RA5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RA5 IL AX 12 LO (UT)
NUT 45 Satisfactory 2RA5 AX 12 RL60 (UT)
. NUT 45 Scan 3 Satisfactory 2 RAS IL AX 12 RL60 (UT)
NUT 45 Scan 3 & 4 Satisfactory 2 RAS AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RA5 LS 12 S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 2RA6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RA6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RB1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RB1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RB1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RB1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RB1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RB1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RB1 SUR 14 PT NPT 01 Satisfactory 2RB5 IL CIR 12 RL60 (UT)
NUT 45 Scant 5&6 Satisfactory 2RB5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RB5 IL AX 12 LO (UT)
NUT 45 Satisfactory 2RB5 AX 12 RL60 (UT)
UUT 45 Scan 3 Satisfactory 2RB5 IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RB5 AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RBS LS 12 S45 (UT)
NUT 45 Scans 7, 8, Satisfactory 9, & 10 2RB6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RB6 AX 12
.S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RC1 AX 14 RL45 (UT)
UUT 44 Scan 3 Satisfactory 2RC1 CIR 14 S45 (UT)
NUT 44 Scan 5 & 6 Satisfactory 2RC1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2R'C1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RC1 AX 14 S45 (UT)
NUT 44 Sean 3 Satisfactory 2RC1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RC1 SUR 14 PT NPT 01 Satisfactory 2RCS IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RC5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RC5 IL AX 12 LO (UT)
NUT 45 Satisfactory 2RCS AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RCS IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RC5 AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RC5 LS 12 S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 2RC6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory
- 2678U
POST-IHSI EXAMINATIONS OF REPLACEMENT PIPZNG AND OTHER WORK (Continued)
Weld Nominal Pipe NDE Procedure Scan Ent_
Reaion SJze (Inches)
Method No.
Performed Results 2RC6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RD1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RD1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RD1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RD1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RD1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RD1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RD1 SUR 14 PT NPT 01 Satisfactory 2RD5 IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RDS CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RD5 IL AX 12 LO (UT)
NUT 45 Satisfactory 2RDS AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RDS IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RD5 AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RD5 LS 12 S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 2RD6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RD6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RE1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RE1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RE1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RE1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RE1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RE1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RE1 SUR 14 PT NPT 01 Satisfactory 2RES IL CIR 12 RL60 (UT) 1RE 45 Scans 5 & 6 Satisfactory 2RES CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RES IL AX 12 LO (UT)
NUT 45 Satisfactory-2RES AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RES IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RES AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RES LS 12 S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 2RE6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RE6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RF1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RF1 CIR 14 S45 (UJ)
NUT 44 Scans 5 & 6 Satisfactory 2RF1 AX 14 CR70 (LI)
NUT 44 Scan 4 Satisfactory 2RF1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RF1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RF1 AX 14 560 (UT)
NUT 44 Scan 3 Satisfactory 2RF1 SUR 14 PT NPT 01 Satisfactory 2RF5 IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RF5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RF5 IL AX 12 LO (UT)
NUT 45 Satisfactory 2RF5 AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RF5 IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RF5 AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RF5 LS 12 S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 t
POST-ZHSI EXAMINATIONS OF REPLACEMENT PZPING AND OTHER WORK (Continued)
Weld Nominal Pipe NDE Procedure Scan Ent.
Rgginn Size (Inches)
Method No.
Performed Results 2RF6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RF6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RG1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RG1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RG1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RG1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RG1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RG1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RG1 SUR 14 PT NPT 01 Satisfactory 2RG5 IL CIR 12 RL60 (UT)
, NUT 45 Scans 5 & 6 Satisfactory 2RG5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RGS IL AX 12
NUT 45 Satisfactory 2RG5 AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RGS IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RG5 AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RG5 LS 12 S45 (UT)
NUT 45 Scans 7, 8, Satisfactory 9, & 10 2RG6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RG6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RH1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RH1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RH1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RH1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RH1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RH1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RH1 SUR 14 PT NPT 01 Satisfactory 2RHS IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RHS CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RHS IL AX 12 LO (UT)
NUT 45 Satisfactory 2RHS AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RHS IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RHS AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RHS LS 12
.S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 2RH6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RH6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RJ1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RJ1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RJ1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RJ1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RJ1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RJ1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RJ1 SUR 14 PT HPT 01 Satisfactory 2RJ5 IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RJ5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RJ5 IL AX 12 LO (UT)
NUT 45 Satisfactory 2RJ5 AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RJ5 IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2678U i
_ _.. _.. _ _.. ~
POST-IHSI EXAMINATIONS OF REPLACEMENT PIPING AND OTHER WORK (Continuad)
Weld Nominal Pipe NDE Procedure Scan Ho,.
Recion Size (Inches)
Method No.
Performed Results 2RJ5 AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RJ5 LS 12 S45 (UT)
NUT 45 Scans 7, 8, Satisfactory 9, & 1P 2RJ6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RJ6 AX 12 S45 (UT)
NUT 47 Scans 3 & 4 Satisfactory 2RK1 AX 14 RL45 (UT)
NUT 44 Scan 3 Satisfactory 2RK1 CIR 14 S45 (UT)
NUT 44 Scans 5 & 6 Satisfactory 2RK1 AX 14 CR70 (UT)
NUT 44 Scan 4 Satisfactory 2RK1 AX 14 RL45 (UT)
NUT 44 Scan 4 Satisfactory 2RK1 AX 14 S45 (UT)
NUT 44 Scan 3 Satisfactory 2RK1 AX 14 S60 (UT)
NUT 44 Scan 3 Satisfactory 2RK1
_ SUR 14 PT NPT 01 Satisfactory 2RK5 IL CIR 12 RL60 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RK5 CIR 12 S45 (UT)
NUT 45 Scans 5 & 6 Satisfactory 2RKS IL AX 12 LO (UT)
NUT 45 Satisfactory 2RK5 AX 12 RL60 (UT)
NUT 45 Scan 3 Satisfactory 2RK5 IL AX 12 RL60 (UT)
NUT 45 Scans 3 & 4 Satisfactory 2RKS AX 12 S45 (UT)
NUT 45 Scan 4 Satisfactory 2RK5 LS 12 S45 (UT)
NUT 45 Scans 7, 8,
Satisfactory 9, & 10 2RK6 CIR 12 S45 (UT)
NUT 47 Scans 5 & 6 Satisfactory 2RK6 AX 12 545 (UT)
NUT 47 Scans 3 & 4 Satisfactory GR-2-4 04 RT NRT 01 Satisfactory GR-2-7 04 RT NRT 01 Satisfactory GR-2-8 28 RT NRT 01 Satisfactory
)
GR-2-15 OL --
12 PT NPT 01 Satisfactory GR-2-15 OL --
12 UT 45 BFUT 29 Satisfactory GR-2-15 OL --
12 UT 70 NUT 28 Satisfactory GR-2-30 04 RT NRT 01 Satisfactory GR-2-33 04 RT NRT 01 Satisfactory GR-2-34 28 RT NRT 01 Satisfactory KR-2-1 04 PT NPT 01 Satisfactory KR-2-1 04
.UT45 (UT)
NUT 25 Satisfactory KR-2-4 04 PT NPT 01 Satisfactory KR-2-4 04 UT45 (UT)
NUT 25 Satisfactory KR-2-11 22 RT NRT 01 Satisfactory KR-2-23 04 PT NPT 01 Satisfactory KR-2-23 04 UT45 (UT)
NUT 25 Satisfactory KR-2-26 04 PT NPT 01 Satisfactory KR-2-26 04 UT45 (UT)
NUT 25 Satisfactory KR-2-33 22 RT NRT 01 Satisfactory m-NlA(RI)
NUT 42 Scans 3, 4, Satisfactory 5, & 6 N1B(RI)
NUT 42 Scans 3, 4, Satisfactory 5, & 6 TCS-2-401 --
10 PT NPT 01 Satisfsetory TCS-2-401 --
10 UT45 NUT 45 SE Side Satisfactory only 1
i a
POST-IHSI EXAMINATIONS OF REPLACEMENT PIPING AND OTHER WORK (Continued)
Weld Nominal Pipe NDE Procedure Scan pion Recion Size (Inches)
Method' No.
Pe rforrned Results TCS-2-403 10 PT NPT 01 Satisfactory TCS-2-403 10 UT45 NUT 42 SE Side Satisfactory only TCS-2-417 10 PT NPT 01 Satisfactory TCS-2-417 10 UT45 NUT 42 Se Side Satisfactory only TSCS-2-418 --
10 PT NPT 01 Satisfactory TSCS-2-418 --
10 UT45 (UT)
NUT 45 SE Side Satisfactory only DRWC-2-5A --
06 UT45 (UT)
. NUT 25 No scan 3 Satisfactory DRWC-2-5A --
06 UT60 (UT)
NUT 25 Supplemental Satisfactory Exam DSRWC-2-5 OL--
06 UT NUT 28 Satisfactory DSRWC-2-5 OL--
06 UT70 (UT)
NUT 28 Satisfactory N5A (CS) TSC 10 UT45 NUT 38 Satisfactory NSB (CS) TSC 10 UT45 NUT 38 Satisfactory l
2678U
_____.,__,,.___,._--__-s-
APPENDIX 3 ANALYSIS OF FLAWED WELDS KR-2-14, KR-2-36 KR-2-41, AND KR-2-37 FOR INDUCTION HEATING STRESS IMPROVEMENTS
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June 13, 1985
~~p SIR-85-008
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.tr. J. E. Wilson L29 8 5 0 G 18 NM Tennessee Valley Authority e
1270 C%stnut St. Tower 11
" @7 Chattanooga, TN 37401
- i
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Ubject:
e Analysis of Flawed Welds for Induction Heating Stress Improvement
'i
- Browns Ferry Nuclear Plant, Unit 2
.aar Ed:
'aclosed please find five copies of the initial issue (Revision 0) of our eport on the subject analysis.
We have incorporated all the comments.
.,ransmitted by your letter of May 3, as well as those resulting from our own internal review prccess.
additional ~ analysis, this should complete our efforts on this i
'always it has been a pleasure working with you and the TVA staff on this As project.
On a related topic, Tony Giannuzzi has just completed a study for EPRI on IHSI k effectiveness in retarding IGSCC crack growth (EPRI Project T303-1).
- inal report was submitted to EPRI in May His iPRI will be distributing it to utility members shortly., and it is our understand
- f use to you in your NRC deliberations on the subject welds.This report may be f
of the ccmplete program package for our computer p l
purpose in sending this is to see if the new documentation package resolves My the problems you were having running the tearing instability option. In this regard, I refer you to Section 5.7 of the User's Manual, and to Sample Problem Ill, which has been revised from the preliminary version in your Beta Test Manual.
I also call your attention to the extensive QA verification effort we have conducted, which is documented in the pc-CRACK verification repo I
The curret trend in the U.S. Nuciear Industry is uward increased use of Fracture F.'echan h s to relax loss-of-coolant accident design criteria, and apply leak-tefore-treak criteria instead (Ref:
NUREG 1061, Volume 3 Nov.
1984, and Voise 5, Apr il 1985).
for ntilities tc implement these criteria changes. We have no to four U.S. clienm Pacific Gas and ti d.-ic - 2 copies Northern States Power Coc - 3 copies New York PNer Authority - 2 copies Northeast Utilities Service Co. - 1 copy, 3150 ALMAge gyPESSWAY SWTE 228 o sam m e e== - ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ~ ~
f r,.-
- ~
f.I'.;"p
/'.
itj*/Page2 SIR-85-008
, ;q%'
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$ '.f whom are electric utility companies also.and we are in the final sta The feedback we have recrived
'!,h, 4garding our training course, and the pc-CRACK program itself, has beemvery
'~ Q.
psitive.
I would be happy to provide the names of the pc-CRACK users at 0 these companies so you can obtain their impressions directly.
g
!.If, based on this additional data, you decide to accept our proposal II -
- c. -
>UT
' 043P (dated March 19,1985), then TVA shopld issue us a purchase order per
'.:Id that proposal, and keep the enclosed notebook and diskettes as your initial
~
topy.
We con arrange for the training program at your convenience..
If, however, you decide not
~-
to buy pc-CRACK, or "to defer your decisitm any longer, I must ask, in fairness to our other customers, that you return this manual, plus all copies of the Beta Test materials we previously sent you, within three weeks of this date.
Jn either case, I thank you very much for serving as a Beta-Test Site forthis 2
oroduct, and I look forward to continuing our long and valued association
- ith TVA.
~~
e ry r 7 ours,
/
I
.)
Q /f P. C. Riccardella e
/dih
(
Enclosures 1.
Report SIR-85-008, Revision 0 (5 copies) 2.
pc-CRACK Package cc:
Frank Novak, WSI w/1 copy of Report SIR-85-008 l
(Enclosure 2 is filed in Applications Engineering Group).
O O
hTHUCTUnla INTIGRITY NINC
_____m.________-
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Fracture Mechanics Evaluation of Recirculation System Piping Welds in 0
Browns Ferry Unit 2 Nuclear Power Plant 1
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Report No.: SIR-85-008
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Revision 0 SI Prcject No: TVA-06 June 14, 1985 i
' Fracture Mechanics Evaluation of Recirculation System Piping Welds in Browns Ferry Unit 2 Nuclear Power Plant 0
0
(
lC Prepared by:
' []
I U Struttural Integrity Associates San. Jose, California i
P.repared for:
Tennessee Valley Authority i
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Prepared by:
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Date:
A. Y. K dd
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k fe W/W Date: b//3/W n.,.
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Reviewed and
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a Approved by:
M/661
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Date:
6/Gt&T
'P. C. RiccaPoella ' V Project Manager r
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u STRUCTURAL INTEGRITY a-ASSOCIATESINC
e REVISION CONTROL SHEET
(
~
SECTION
]
PARAGRAPH (S)
DATE REVISION
~
REMARKS All All 6/14/85 0
Initial Issue
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i D STRUCTURAL INTEGRITY l
ASSOCIATES,INC L.
TABLE OF CONTENTS
(
J Page LIST OF TABLES tii LIST OF FIGURES iv
1.0 INTRODUCTION
1-1 2.0
SUMMARY
OF INSPECTION RESULTS 2-1 i
3.0 EVALUATION METHODOLOGY.
3-1 3.1 Applied Stresses 31 3.1.1 Stresses Due to Operational Loadings 3-2 3.1.2 Residual Stresses.
~
3.2 Stress Intensity Factors 3-3 3.3 Crack Growth 3-5 3.4 Allowable Flaw Size (IWB-3640) 3-6 3.5 Allowable Flaw Size (EPFM).
3-7 3.5.1 3600 Part-Through-Wall Cracks in Tension.
3-7 3.5.2 Through-Wall Cracks in Tension 3-9 3.5.3 Limitations-for J-Controlled Growth 3-10 3.5.4 Stress-Strain Laws Considered
. 3-11 3.5.5 Weld Metal Toughness Data.
3-14 3.5.6 Critical Flaw Size Determination 4-1
[I 4.0 EVALUATIONS AND RESULTS 4-1 4.1 Weld KR-2-14 4-1 4.2 Weld KR-2-36 4-2 4.3 Weld KR-2-41 4-4 4.4 Weld KR-2-37 4-5 5.0 OISCUSSION AND. CONCLUSIONS 5-1
6.0 REFERENCES
6-1 e
e t
j i
g
^ STRUCTURAL INTEGRITY ii ASSOC 1;GF_ SINC 1
~
LIST OF TABLES k
Table Pace 2-1 Upper Bound Crack Sizes and Worst Case Crack Locations.
. 2-2 3-1 f
Summary of Applied Stress at Weld KR-2-14
. 3-16 i
3-2 Summary of Applied Stress at Weld KR-2-36.
3-17 3-3 Summary of Applied Stress at Weld KR-2-41.
~
. 3-18 3-4 ' Allowable End-of-Evaluation Period Flaw Dep'th to Thickness Ratio for Circumferential' Flaws --
a I
Normal Operating (Including Upset and Test)
'i Conditions
. 3-19
3-5 F for a Circumferentially Cracked Cylinder in i
Tension (t/Rj = 1/10)
. 3-20 3-6 h1 for a Circumferentially Cracked Cylinder in Tension (t/R i = 1/10)
. 3-20
~~ 3-7
( [1 Ff6raCircumferential,Through-WallC[ack in a Cylinder of t/R = 1/10 in Tension.
. 3-21 3-8 h1 for a Circumferential Through-Wall Crack in j
a Cylinder of t/R = 1/10 in Tension.
. 3-21 i
3-9 Material Stress-Strain Properties of Base and Weidment Materials Used in Analysis.
. 3-22 3-10 Three Commonly ~ Used Ramberg-Osgood Constants for Weldment Materials
. 3-23 3-11 Welding Processes for 5tainless Steel Pipe
.'3-24
~
l-i_
f t
STRUCTURAL in INTEGRITY ASSOCLKfESINC =
j i
LIST OF FIGURES 6
Figure Page F13wlhdic5tionsinWeldKR-2-14.
2-1 2-3 2-2 Flaw Indications in Weld KR-2-36.
2-4 2-3 Flaw Indications in Weld KR-2-41. '.
2-5 2-4 Flaw Indications in Weld KR-2-37.
2-6 i
3-1 Comparison of Measured-and Computed Residual Axial Stresses Along Inner Surface of a Welded, IHSI Treated 12-Inch Sweepolet 3-27 3-2 Comparison of Measured and' Computed Residual
-f Circumferential Stresses Along Inner Surface
~
of a Welded, IHSI Treated 12-Inch Sweepolet 3-28.
I 3-3 i
Computed Residual Stresses at Weld Centerline for IHSI Treatment of a 12-Inch Sweepolet 3-29 1
3-4 Computed Residual Stresses 0.75 inch (1.9 cm)
'j from Weld Centerline. 0.25 inch (0.64 cm) from Co.il Centerline, for IHSI Treatment of..a 12 Inch Sweepolet 3-30 L
3-5 Post-1HSI Residual Stress Distribution.
3-31 3-6 Analytical Model for Post-IHSI Residual Stress Calculation.
3-32
'3-7 Post-1HSI Residual Stress Distribution.
3-33 3-8 Magnification Factors of Circumferential Crack in a Cylinder (a/t = 0.1) 3-34 349 Stress Corrosion Crack Growth Data for Sensitized Stainless Steel in BWR Environment (Ref. 7) 3-35 3-10 Comon Assumptions Used to Estimate Circumferential Cr'ack Growth 3-36 3-11 Average Effective Circumferential Crack Growth l
Rate As a Function of Operation Periods Used l
in Calculation of Time Between Inspections i
3-37 L
3-12 Tearing Modulus Concept for Stable Crack Growth.
3-38 l
3-13 Circumferentially Cracked Cylinder in Tension 4
3-39 I
STRUCTURA iv INTEGRITY ASSOCIATESIN
~
LIST OF FIGURES (continued) i
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Fi ure Pace 3.4 Through-Wall Flawed Cylinder Under Reme:e Tension 3-40 3-15 -Ramberg-Osgood Characterization Stress-Strain Curves 3-41 3-16 Compilation of Material Toughness J-T Curves (from Data of Refs. 17 to 21).
3-42 3-17 Lower Bounds of J-T Data for Wrought Stainless Steel Base Metal and for Stainles.s Steel Weld Metal from TIG, SMAW and SAW Welding Processes 3 43 3-13 Effect of Ernst Correction on Lower Bound Weld and Base Plate J-T Curves 3-44 2)
Lower Bound J-T Reference Curves for use in Elastic-Plastic Fracture Mechanics Analysis of Austenitic Stainless Steel Pipes.
3-45
' iO Material J-R Curve Derived from Lower Bound' J-T Diagram for SAW/SMAW Weldment Material 3-46
- 21 Material J-R Curve from Lower Bound J-T Diagram for SAW/SMAW Weldment Material (Expanded Scale).
3-47
(
4-1 Stress Intensity Factor Versus Crack Depth for Weld KR-2-14 4-8
-?
Predicted Stress Corrosion Crack Growth for Observed Ultrasonic Flaw Indication - Weld KR-2-14.
4-9 4-3 Comparison of Predicted Crack Growth with Allowable Flaw Size Limits - Weld KR-2-14 i.
4-10 4-4 Stress Intensity Factor Versus Crack Depth for Weld KR-2-36 4 11 4-5 Predicted Stress Corrosion Crack Growth for Observed Ultrasonic Flaw Indication - Weld KR-2-36.
4-12 4-6 Comparison of Predicted Crack Growth with Allowable Flaw Size Limits - Weld KR-2-36
. '4-13 4-7 Stress Intensity Factor Versus Crack Depth for Weld KR-2-41'.
4 14 4-8 Predicted Stress Corrosion Crack Growth for Observed Ultrasonic Flaw Indication - Weld KR-2-41.
4-15 1
STRUCTURAL i
v
~,
INTEGRITY ASSOCIATES,1NC W
_ _ _ _ _. _ _ - _ - - - - - - - - - - - - - - - - - ^ ^ - - - - - - ~
l LIST OF FIGURES (continued)
Fi
.e
(
Page 4'
Comparison of Predicted Crack Growth wid Allowable Flaw Size Limits - Weld KR-2-.21.
4-16 4-10. Stress Intensity Factor Versus Crack Depth '
j for Weld KR-2-37 4 17
- 4. 1 Predicted Stress Corrosion Crack Growth for
{
Observed Ultrasonic Flaw Indication - Weld KR-2-37.. 4-18 4-12 Comparison of Predicted Crack Growth with Allowable Flaw Size Limits - Weld KR-2-37.
4-19 I
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.DSTPJCTURAL
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vi UiTIGRITY ASSOCIATESINC
- - - - - - - - ~ ~ ~ ~
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1.0 (i
INTR 0gUCTION 1
During t.he 1984L8S outage at the Browns Ferry Unit 2 Nuclear Power Plant, ultrasonic (UT) examination of the recirculation system piping produced indications at four weld joints which are believed to result from inter-granular stress corrosion cracking (IGSCC).
Similar indications have been observed at a number of other Boiling Water Reactors (BWRs) in the U.S.
and overseas.
These four welds have'been evaluated to demonstrate t' heir acceptability in a
accordance with ASME Section XI requirements, supplemented by the recom-j mendations of NRC Generic Letter 84-11. The welds were.also analyzed using Elastic Plastic Fracture Mechanics Tearing Instability methodology to account for possible effects of low toughness weld metal.
All of the welds were treated by induction heating stress improvement (IHSI) ~ to inhibit further IGSCC propagation.
1
~
Structur'al Integrity Associates.(SI) was contracted by the Tennessee Valley k.j Authority (TVA) to perform the evaluations of the four weld joints.
i This report documents the results of the analyses, which demonstrate that design basis safety margins are maintained in these welds, considering worst case interpretation of the UT indications.
Section 2 of this report sumarizes the inspection results.
Section 3 describes the flaw evaluation methodology used to evaluate the welds, and Section 4 preseats the evaluation results. Section 5 presents the conclusion of the evaluation regarding the continued, safe operation of the plant.
~
GW l-a I
1 1-1 STRUCTURAL s
INTEGRITY ASSOCIATESlhC O
- 20
SUMMARY
OF INSPECTION RESULTS
(,_
After a thorough in-service inspection of the recirculation and associated stainless sti!el~ piping systems, IGSCC-like indications were found in three ring header-to-sweepolet welds and one ring header to-end cap weld.
[
Figures 2-1 to 2-4 provide a weld-by-weld summary of these indications, including indication sizes detected after IHSI treatment *.
All the indications are
~
circumferential1y oriented, and have been conservatively assumed to be cracks or crack-like for purposes of this evaluation.
Upper bound crack dimensions and worst case positions with respect to the applied stresses were used in the crack growth calculations and are. tabulated in Table 2-1.
I 1
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- Some changes in indication sizes occured between the pre-and post-IHS1 inspections of these welds, but they were not significant, i--
2-1 STRUCTURAL INTEGRITY ASSOCIATESINCc
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TABLE 2-1
(
Upper Bound Crack Sizes and Worst Ca:e' Crack Locations 1
~
l Crack Depth Crack Length Worst Applied Stress Weld No.
(% Wall Thickness)
(inches)'-
Location (degrees)*
KR-2-14 12.6 2.1 80**
KR-2-36' 25 2.2 80
- KR-2-41 Ig 4
60**
KR-2-37 12 5
any position t
- h 0 is along the 9
- 00 direction and 90 is along the 12:00 l
direction in Figures 2-1 to 2-4.
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the highest stress location (see Section 3.1).
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STRUCTURAL 26 INTEGRITY ASSOCIATESINC
3.0 EVAll!AT10N METHODOLOGY 3.1 Applied Stresses Two major types of stresses are considered in this evaluation, stresses due to oper1tional loa' dings and residual stresses.
The operational stresses include pressure, deadweight, shrinkage due to weld oterlay repair of weld GR-2-15, thermal, and seismic. Residual stresses were evaluated considering the beneficial effects of the IliSI treatment which was performed on these.
I welds.
These-stresses are described in more detail in the following sections.
~
j 3.1.1 Stresses Due to Operational Loadings
~
J
)
The applied moments on sweep-o-lets KR-2-14, KR-2-36 and KR-2-41.were provided by TVA (Ref.1). Due to the complexity of the sweep-o-let geometry, the -stress due to these moments varies in a non-linear f ashion along the
}
azimuthal location of the weld between the ring header and the sweep-o-let.
Also, the stress varies with distance from the cr atch region of the sweep-o-let (Ref. 2). The highest stress location in a sweep-o-let is at the crotch
(
I}
region, and the stress decays rapidly as one moves away from that region.
l Since the weld seam between the sweep-o-let and the ring header is somewhat removed from the crotch region,.it does not see the full stress concentration attributable to the crotch region, but the stresses are still higher than the nominal bending stresses caused by the pipe applied moments.
A. stress concentration methodology for such a sweep-o-let weldment location was developed in Reference 2, and is used in this evaluation to obtain~ the appropriate stresses.
Table 31 to 3-3 present the applied stresses due to the various applied loadings for welds MR-2-14, KR-2-36 and KR-2-41 calculated in accordance with the Reference 2 methodology.
As azimuthal angle increases from Oo (longitudinal section) to 900 (transverse sections), the stress concen-
~
tration factor on bending moment increases from 1 to 3.
The corresponding stress concentration f actor for pressure incr5ases from 0.6 to 1.
These f
inside surf ace concentration factors also include local through-wall bending e
stress effects on the moment terms. At the weld location., the magnitude of s
l N STRUCTURll
~
3-1 INTEGRITY ASSOC 1KIEsIN
/
the out' side surface stress is approximately two-thirds of the. inside surface strers'and is compressive (Ref. 2).
b-Tables 3-1 to 3-3 also give the resultant through-wall membrane and bending:
stresses, and the ASME Code stress ratio (Pm + P )/Sm for normal and faulted' b
conditions.
The maximum normal condition membran.e stress is seen to be 15.4 ksi for all welds and locations, which corresponds to a maximum stress ratio of 0.91.
Note that this stress ratio conservatively includes thermal-expansion arid weld overlay shrinkage stresses as,a primary stress term.
The maximum faulted condition stress ratio is.only negligibly higher (0.925) and, therefore, normal conditions will govern the allowable flaw size calculations.
3.1.2 Residual Stresses Residual stresses are known to play a significant role in IGSCC. A favorable f
residual stress pattern can arrest further crack growth, while an unfavor-able one can accelerate crack growth.
A general survey of available l
' analytical and experimental results was performed to ' establish the most Appropriate residual stress profile for use"in the subsequent crack growth kp
- analysis, in evaluating the indications two representative post-lHSI residual stress distributions, one for the sweep-o-let welds and one for the end cap weld, are considered.
f Rybicki, et al (Ref. 3) have presented extensive analytical ~ results on
(
induction heating of welded stainless steel pipes. These analytical results cover a wide range of piping welds and fittings. Also Ishikawajima-Harima Heavy Industries (IHI) Company in Japan did an in-depth study to qualify and verify the IHS1 process for boiling water reactor piping (Ref. 4).
Figures 3-1 to 3-4 present computed and experimentally measured residual stresses for a sweep-o-let weld with IHSI treatment.
t Figures 3 1 and 3-2 present the inner surface stresses, circumferential and longitudinal, versus distance from the weld centerline.
}'
The measured stresses compare very favorably to the analytical results. As shown in the figures, the surface' residual stress in a 12-inch sweep-o-let is about 20 to'40 ksi compressive.
l-figures 3-3 and 3-4 present through-wall analytical results, from two l
different finite element modcIs.
They give about the same magnitud$T E TU m
.3-2 INTEGRITY ASSOCIATESINC
surface stress, as compared to each other and to those in Figures 3-1 and 3-
~
~
2.
The two finite element models also give similar through-wall residual
{'
stress patterns.
f-Finally, Figure 3-5 presents test data on 12 inch sweep-o-let weld to a 22 inch pipe, from Reference 5.
The test data give only inside and outside surf ace stresses, but at three different a'zimuthal angles; 00, 450, 900 Since no through-wall test data are available from that test, linear through-wall stress profiles are assumed.
Of the three angle s examined, the inside surface stress at 00 had the least compressive stress, but all three through-wall stress profiles are.similar.
Also, the surface stresses agree
~
reasonably with the previous ana* ytical and experimental results.
Thus, for f
conservatism, the 00 residual stress distribution of Figure 3-5 was used for sweep-o-lets in this evaluation.
i Figures 3-6 and 3-7 present an analytical model and IH5I residual stress results for a 16-inch end cap (Ref. 4).
No results, either analytical or experimental, are available for a 22 inch end cap. Therefore, the 16 inch end cap analytical IH51 residual stress is assumed for the 22 inch end cap in the
( ')
recirculation system.
Figure 3-6 presents the finite element model identifying all the dimensions, boundary conditions and length of the heating coil.
Figure 3-7 presents the inside and outside surface stresses as a function of the distance.from the weld centerline.
Near the weld centerline, the compressive surface stresses are on the order of 20 ksi.
Since no through-wall data are available, a linear through-wall stress profile is assumed for the subsequent end cap crack growth analysis.
3.2 Stress Intensity Factors Pipe dimensions used in this analysis are as follows (Ref.1):
1
(
22 Inch Pipe 1? Inch Pipe Outside Diameter (in.)
I 22 12.75 Inside Diameter (in.)
19.75 11.592 Pipe Wall Thickness (in.)
1.125 0.579 i
3-3
" STRUCTURAL INTEGRITY ASSOCIATESINC
(
An analytical mo_ del of a 3600 circumferential cra.ck in a cylinder of radius to thickness ratio of 10:1 (Ref. 6) was used for the fracture mechanics growth evaluation.
The applied loading ccnsists of piping loads due to shrinkage, dead weight, pressure, thermal expansio 1, and seismic. and the residual stress distributions discussed in Section 3.2.
For the piping loads, the loading consists of the piping stresses tabulated in Tables'3-1 to 3-3 for the sweep-o-lets, or just internal' pressure stress. for the 22 inch
^
end cap (5.662 ksi axial, 11.324 ksi. circumferent'ial). - The post-IH51 residual stress distribut-ions are given in Figures 3-5 and 3-7.
For purposes of the fracture mechanics analysis, the axial stress distribu-tions from these loading cases have been expressed in terms of third degree.
i.
polynomials of the form:
Ij a=A 0 + A x + Apx2 + A x3 1
3 (1)
(t where cris axial stress in the units of ksi, x is the distance from the inside surface, and A0-A3 are the coefficients resulting from the curvefit.
(*
The stress intensity factor for a circumferential crack in a cylinder of radius to thickness ratio of 10:1 can be expressed as follows (Ref. 6):
K1 = 6 (A F01+
ApF - + ha3 AF12+
.A F )
(2) 3 34 where F, F, F, and F i
1 2
3 i
4 are magnification factors and a is crack depth as shown in Figure 3-8.
i-For the linear elastic fracture mechanics portion of the analysis, the stress intensity factors can be calculated independently j
for piping stress and post-lH5I residual stress distributions, and the
~
resultant stress intensity factor is the superposition of the two. loading l-cases.
I t
l
^ STRUCTURAL 3g INTEGRITY ASSOCIATESINC 1
3.3 Crack Growth
(
A large body of laboratory data exist on stress corrosion crack growth ratas for sensitized stainless steels in simulated BWR environments. These dzca are sumarized in Figure 3-9, taken from Reference 7.
These data we'e obtained using fracture mechanics type specimens with different crack sims and loadings, which can be characterized by the crack-tip stress intensity ~
factor K.
The data represent a wide variation in material sensitization, as well as levels of dissolved oxygen in the water.
While subject to sore f
1 criticism because the simulated water chemistry in these ' tests did mt
. contain levels of impurities (chlorides, sulfates, etc.) that could exist in j
operating BWRs, the "best estimate" curve of Figure 3-9 is widely believed ta 1
provide a reasonably conservative bound' of stress corrosion crack grow:fi-rate for weld sensitized 204 stainless steel in BPR environments. This cure can be described by a power law representation of the form:
j da/dt = 2.27 x 10-8(K)2.26 (3) d
{
h where a is the crack depth in units of inches, t is time in units of hours, and K is the stress intensity facter in units of Ksi 8 Crack growth analyses typically make use of one of the two assumptions
~ illustrated in Figure 3-10 regarding crack.cngth extension, self-similar crack growth or constant aspect ratio crack growth. The former assumes that the incremental crack extension is the same at all points on the crack frer:c,.
While the latter assumes that the ratic of depth to length remains constant ~
during crack extension.
Considering field and laboratory experience with circumferential crack extension, it appears that the self-similar assumption may underpredict crack length versus time, while the constant aspect ratio assumption overpredicts.
Recent work by Gerber (Ref. 8) under contract to EPRI provides a new approsch for addressing circumferential crack extension which is more' technicall'y defensible than the above self-similar or constant aspect ratio approaches.
This approach utilizes data generated in a laboratory stress corrosion test of a 26 inch diameter welded pipe specimen at Battelle Pacific Northwest
^STH11CTURAL 3-5 INTIURITY t'
ASSCCIATESINC i
l
~
~
Labora ories (Ref. 9). IGSCC was induced in this pipe through loading to a
{
high applied stress in a simulated BWR environment, which was accelerated by the use of graphite wool to create an artificial crevice.
Crack growth l
occurred an~d was monitored both during operation and at several scheduled shutdown intervals for the test. A number of small^ cracks initiated early in l
[
the test, the length of which was periodically measured and the initiation of new cracks was noted and their lengths subsequently tracked as well. At the completion of the test, there were a total of 63 cracks with a combined length of 32.57 inches.
The average effective circumferential crack extension observed in this test
)
is presented in Figure 3-11.
This rate includes both growth of existing
{
cracks as well as new defects initiating and contributing to the effective
{
crack growth rate in each inspection interval.
Examination of Figure 3-11
}
suggests that an average effective circumferential crack growth rate of 0.5 mils / hour shou?d give a reasonably conservative estimate.
Thus, 0.5 mils / hours was used as the crack length growth rate in this report. It should be pointed out, however, that although this is_an average effective rate, it
{
is based on a laboratory test in which the local environment, load and cycles were all intentionally modified to accelerate IGSCC relative to actual plant conditions. Test and analytical data (Reference 4) have also shown that the
,[
IHS1 will suppress not only crack initiation but also crack propagation for
. cracks in both the length and depth directions.
3.4 Allowable Flaw Size (IWB-3640)
Based on detailed calculations presented in References 10 and 11, allowable flaw sizes for various levels of primary and applied loading (Pm + P ) have
)
b
_ been specified in ASME Section XI, IW3-3640 (Ref. 12).
A tabulation of
. allowable flaw sizes as a function of applied load is given in Table 3-4, which is taken directly from Section XI, IWB-3640.
Note that this table permits very large defects in some cases (as great as 75% of pipe wall) and
~
does not include consideration of any stress other than primary, notably
~
secondary and peak stresses from the design stress report as well as any weld residual stresses or misalignment / fit-up stresses which might exist from construction.
The argument for this exclusion is that, given the extremely I
I-STRUCTURAL i
3-6 INTEGRITY ASSOQATES.INC
___._____._____m_
([
high dur.tility of austenitic stainless steel, these strain controlled effects will self-relieve after a small amount of plastic deformation and/or
~
~
stable crack _ extension, and will have little or no impact on the loads and flaw sizes needed to cause unstable crack propagation or pipe rupture.
However, some recent fracture toughness data may invalidate the above argument, at least for some classes of austenitic weld metal (Ref.13).
To account for possibility of low ductility weld metal, secondary stresses from stress report were also included in the IWB-3640 allowable flaw size determinations in this report, although it is not required by the ASME Code.
L h
It is important to note that the very low measured toughness occurred only in s
~
a small percentage of the materials addressed in Reference 13, and may be of q
cnly limited concern from a probabilistic viewpoint.
Indeed, most IGSCC observed to date has been restricted to weld heat af fected Zones, which should exhibit the high toughness attributed to base material. Also, the low l
toughraess data to date has been limited to flux types of weldments (submerged.
~~
arc or shielded metal arc), which are not 'used in current construction
- )
practice nor in weld overlay repairs of pipe cacks. Nevertheless, to address these possible concerns, the analysis procedure used in this report includes j
thermal expansion affects as a primary stress condition in determining allowable flaw size from Table 3-4.
L 3.5 Allowable Flaw Size (EPFM)
Methodologies from References 14 and 15 are also used in this report to calculate applied J and T values for circumferential through-wall or part-through-wall cracks in pipes as functions of applied loading. Details of the
~
~
~
methodology used are provided below. These computed, applied J and T values I
)
are then compared to a J/T material curve on a J/T stability diagram (as in i
Figure 3-12) to provide a second means of determining allowable flaw size.
1-I i
3.5.1 3600 Part-Through-Wall Cracks in Tension I
As shown in figure 3-13, consider a cylinder with an inner radius Ri, outer radius Ro, and wall thickne s s t t
R o Rj, containing an internal i
=
I_
_^ STRUCTURAL INTEGRITY 3-7
/MIATES.lNC
l 4
axisymmetric part-through crack of depth a. e* denotes the far field uniform
(
t "sile stress and c = t-a the uncracked ligament.
A radius to thickness
- io (R /t) of 10 is used in this report, which corresonds approximately to r
i tne Schedule 80 piping used in service. The elastic-plastic formulae for J-applied in this case have been obtained from Reference 14 and are as follows:
Jappl = de + Jp C
=f1 (a,Rj/Ro)(P /E') + a c 2
o co C (a/t) hi (P/P )'"I (4)
~ '
e o
where:
ad f (a, Rj/Ro) =
1 e r(Ro-Ri )2 I
I E' = E/(1-v2)
E = Young's Modulus
\\
v: Poisson's Ratio c :-Yield Stress l\\
o c = Yield Strain o
a,n= Material constants of Ra.aberg-Osgood Model iL
~
e = a + [1 + (P/P )2]
[( n" ) (~a
) /(20)3 a
o K = c g* ri a F
~
Po = 2/g 3 o
[Ro2 - (Ri+a)2]
c
.P = a"r(R 2 - Rj )
2 o
i F = function given in Table 3-5 h1 = function given in Table 3-6 i_
For materials with n values between 1 and 10 but not exactly as provided in Table 3-6, the corresponding h1 values can be calculated by interpolation.
The non-dimensional tearing modulus, Tapp1 is calculated by:
l Tappl = (-
-) (d Jappt/da)
L c2 o
(5) e l
l STRUCTURAL 3-8 INTEGRITY ASSOCIATES INC l
?
where Tappl is the applied tearing ' modulus from' loading and.all the other f,~
(
c.intities are defined in the same manner as those in. Equation (4).-
.?
T::e applied tearing modulus can be determin Numerically by applying a
~
i finite difference scheme on the above definition.of Tappli e 9.
I I
I)
Tappi * (
2)
(6)L where da is a small crack length increment.
" 5. 2 Through-Nall Cracks in Tension a second case, consider a cylinder containing a circumferential through-11 crack of length 2a, and subjected to remote uniform tent, ion as shown in -
.gure 3-14 In this figure, R denotes the mean radius, t the wall thickness,.
the total angle span of the crack, and 2a = 2RY the total length of the Eack.
2b = 2::R is the pipe circumference, and P is the applied load, t.oad
- s applied by a uniform stress field at its. ends ~given by
\\\\
cr* = P/(2rRt)
(7) jain, a radius to thickness ratio of 10 is used to approximate service
. ping.
For a Ramberg-Osgood material, the elastic-plastic J-integral estimation has been obtained from Reference 15 and is given as follows:
Jappl
- Je + dp Jappl
- f1 (a, ( ) (P /E) + a cr 2
e co C (a/b) b.(P/P )
-(8).
o l
o where:
f(a,f)=aF/(4R22) 2 1 e t
E = Young's modulus tro = Yield stress co = Yield strain o,n = material constants of Ramberg-Osgood Model
^ STRUCTURAL f
INTEGRITY ASSOCIATES.INC.
3-9
_ _ _ - _ _ - - - - - - = - - -
'ae = a + [1 + (P/Po) ]
[(n" )(cr } 3/IO*)
K = cr"d n a F-
{
~,Po = 2 cro Rt [n 2 sin-1(1/2 sin Y )]
F = function given in Table 3-7
~
h1 = function given in Table 3-8 Interpolation is again used for materials with n values between I and 7, but F t exactly as provided in Table 3-8.
The non-dimensional tearing modulus Ta' pl.can be evaluated by differen-p ttation of Japp1 in the same manner as described in Sect' ion-3.5.1 for part-t ough-wall cracks.
i I.
i
.3 Limitations for J-Controlled Growth
- . order for the above tearing modulus stability concept to be valid, certain I
pitations on the theory must be checked. These limitations are necessary
' ensure that the incremental crack growth and non-proportional loading I
tne in the immediate vicinity of the crack -tip are sufficiently small to
(
,Jstify use of the J-integral in the analysis of crack growth, a condition I
which is defined in Reference 16 as "J-controlled growth". These conditions e generally satisfied, in the large scale yielding range, if the uncracked jament of the cracked cross-section (b) is sufficiently large to satisfy
. e following criteria:
e = h h >> l (9) and P=h n 1 (10)
While there are no generally accepted rules for how much greater than I these pgameters must be to ensure J-controlled growth, a value of W =5 to 10 t
j i; suggested as adequate for Equation (9) and a value of P=25 has been used
~
in a number of sources for Equation (10).
These parameters are thus l
calculated in the J-T analyses which follow, and compared to the above values a
to provide an assessment of the validity of the calculations.
P STRUCTUnqL 3
3-10 INTEGRITY ASSOCIATESINC
3.5.4 Stress-Strain Laws Considered i
(..
Thc crimary material stress-strain law used in this report is based on test data for stainless steel weldments and base metal reported by Westinghouse,
~
(Ref. 17). Figure 3-15 illustrates thcse stress-strain data at operating I
1 temperature and their Ramberg-Osgood representations. A complete tabulatien of ;,aterial tensile properties and corresponding Ramberg-Osgood parameters.
fr these materials are listed in Table 3-9 at both 750F and 5500F.
The weldment material properties at 5500F are used in this report as the primary basis for the J-T analysis results.
Hb ver, since studies have shown J-T analyses to be extremely sensitive to t6
- pecific material stress-strain law characteristics used in the anal-
~
y.
Ramberg-Osgood constants have been obtained for a number of different I
s
.nless steel weldment materials. Table.3-10 lists three commonly used I
d i sets for weldment material at 5500F. As a parametric study, allowable
' size results have been calculated using all three data sets.
3 5.5 Weld Metal Toughness Data i ta on the elastic-plastic toughness properties of austenitic sta(nless
~
el welds are presented in References 17 to 21 in the fem of J-resistance s
c
- es, J-T values, and/or tabulated JIc and J-Resistance curve slope values.
These data have been used to determine lower bound J-T material toughness curves for comparison with applied values to assess crack instability cad allowable flaw size by the J-T method.
The effects and.
results of welding process have been considered in establishing lower bound toughness properties for such stability evaluations.
A cornpilation of applicable material toughness J-T curves from the above f
references is shown in Figure 3-16. These curves represent wrought stainless
~
stael base metal toughness, along with the toughness of stainless steel weld mt bl representing submerged arc welding (SAW), stick or shielded metal arc
~
welding ISMAW), and gas tungsten arc welding (GTAW) or tungsten inert gas
,l
,TIG) welding processes. With the exception of the data from Reference 20, the J-T curves in Figure 3-16 were derived from J vs. crack extension (Aa) i l
curves (J-R curves) and the following equation.
s
^ STRUCTURAL INTIGRITY -
3-11 ASSCCIATESINC g
.m._.__-__--
E dJ T=
(
r f
'da where T'=
tearing modulus E' = E/(1 -v2)
E elastic modulus = 30,000,000 psi
=
V Poisson's ratio = 0.3
=
l-cf = material flow stress = 50,000. psi' I
.1J/da = slope of J-R curve at specific us E :ause of the absence of tensile properties.in many cases, the above values or. flow stress and elastic modulus were assumed throughout.
J-P. curves and-sile properties were not available from Reference 20, so J-T curves yere 1
~
. ed directly from this reference.
\\l
.F pre 3-17 presents lower bounds of the toughness data for the various j
rtegories cf material in Figure 3-16. It can be seen that there are distinct 1
,1(ferences in lower bound material toughness _between the wrought base metal
('
g'.
.:nd the various weld metals.
The base metal is' the toughest material
!)
(largest values of J and T), with the SAW and SMAW weld metals being the least i
~3 ugh, and the TIG weld metal having intermediate toughness.
. shown in the previous figures, SAW and SMAW welds possess lower toughness than TIG weld metal.
Differences' in the welding processes, primarily heat
~
input differences and the use of flux versus inert gas shielding, can be used k
to. explain such toughness differences. Key features of the TIG (GTAW), SMAW 1
and SAW processes (Ref. 22) are given in Table 3-11, along with the l
relationship to weld metal toughness. Essentially, SAW is a high heat input, L
flux-shielded process which can result in rciatively coarse microstructure
)
and relatively heavy slag /non-metallic inclusion contents.
Such ' micro-i structures would tend to give reduced toughness (Ref. 23). In comparison, TIG ic,a low heat input process with inert gas shielding' rather than flux.
l~
T".erefore, TIG welds should be of superior toughness. SMAW has intermediate neat input and shielding by gas and molten flux 'from the electrode covering.
l Ihus, SMAW welds are expected to have intermediate toughness - lower than TIG and slightly higher than SAW.
Figure 3-16 generally illustrates this-j expected trend.
STEUCTURA '
3-12 INTEGRITY ASSOCIATESIN H l
N
~
Lcwer bound J-T toughness curves for use in this analysis were derived from p.
{i d r,a of Figure 3-16 and 3-17.
~
Essentially, the' data were divided into three
- gories based on the preceding toughness discussions:. base material, TIG-c w !d metal, and SAW and SMAW weld metal.
For each of these categories, the 1:wer bound curves at low J' values were corrected for specimen size effects, 1 merged with the lower bound surves at higher J values from Reference 20
- a conservative manner.
To. account for the effects of specimen size and geometry in the small srecimen data of References 17,18,19 and 21, a modified' J approach, 'known j
a Jm, was used along with a modified T, Tm (Ref. 24 and 25).
Reference 24-
- 59ws that Jm and Tm can cortelate data for test situations in which th
~
- itions for 'J-controlled crack growth described in Section 3.5.3 are ssly violated.
c This approach is applied here to adjust the lower bound e.;.a of Figure 3-17.
j ia Reference 23, Jm is computed for the compact tension specimen as I
311ows:
(i.i a[
7 Jm = J +
JP da F
(12) o
~
7 = (1 + 0. 76 (-
))
(13) 1.
where Jp is the nonlinear part of the deformation theory J, b is the remainin ligament, W is the specimen width, a and a are the initial and extended crack o
lengths respectively, and 7 is as defined above.
is evaluated as:
Also, from Reference 24, Tm L.
Tm = T + 1
- 7 Jp l
L U(4 b
(14)
P~ the preceding equations for Jm and Tm, J and Jpmust be defined as I-f anctions of crack extension for each material evaluated.
Such definitions
. ave been obtained from the power law curve-fits of the'J-R data of the lower l-7 STRUCTURAL 3-13 INTEGRITY ASSOCIATESINC L_-____-_-____-____-________.
~
1
~'
~
(.
bou ' materials in Reference. 17,18,19 and 21. ' The resulting. values of Jm-Tm Sr the lower bound curv(; for each material category of Figure 3-17 are
~
il, atrated 'in. Figure 3-18.
In each case the Jm-Tm curve branches upward I~
from the respective J-T curve at a prescribed poin.t on the curve.
~
Fi ally, lower bound Jm - Tm curves of Figure 3-18 are then faired into the-I J, large Aa data of Reference 20 to obtain tfje lower bound J-T curves.
h l~
s,own in Figure 3-19 for each material. Again, it can be seen in Figure 3-s 19 that three distinct levels of material toughness exist:. the toughest-base meirial, the intermediate toughness - TIG welds, and the lowest toughness Sb and SMAW welds.
There lower bound curves of Figure 3-19 are employed in j
t' report to determine predicted fracture stresses for the subject pipe I
W 3 using elastic-plastic fracture mechanics analyses.
I J
brder to make crack growth corrections to applied J-integral values, a r frence J-R curve was derived from the lower bound J-T curve in Figure 3-1, Since the tearing modulus (T) is a function of the slope of the J-R curve (p
i ;/da), the reference J-R curve in Figure 3-20 was obtained by integrating
~e lower bound J-T curve in Figure 3-19.
The lower data points at small aa.
- - Figure 3-20 represent the raw data from the unmodified J-R curve for the j
l er bound material toughness. Figure 3-21 shows an expanded scale of this I
1
? Aa regime comparing the raw data (FUC-9) with the extrapolated Jm data
~1: can be seen that, in validating the raw data, Jm gives significant toughness advantages over the deformation J (raw data).
3.5.6 Critical Flaw Size Determination l
The above J-integral estimation methods and material data are then used to establish allowable flaw sizes for the subject welds for comparison to the allowable flaw sizes for these welds based on ASME _ Section XI, IWB-3640
)
methodology discussed previously (Section 3.4).
The basic technique.is ilM strated in Figure 3-12. The intersection of the applied and material J.
J T curves in this figure yields a critical value of J for predicted instability of the weld.
This critical value of - J uniquely defines a
~
i Eitical stress for a given flaw size, or conversely, a critical flaw size for a given stress level.
The later definition is used here.
I
^ STRUCTURAL 3-14 INTEGRITY ASSOCIMES.lbC 3
en
_________.__.___._____m.__.
._____...m_
___._2_m
.m.
m m____._
o
At this point, it must be noted that the'IWB-3640 tables for permissible flaw s.as were developed based on an inherent safety f actor of 2.773 on stress or 1
i to net section collapse of the cracked cross-section (Ref.11). Thus, in
~
ceder to provide a consistent basis for comparison, the ' applied loading on each pipe weld to be evaluated (Ref.1) is multiplied by a factor of 2.773 hefore applying the J-T critical flaw size determination described above.
,.lowable sizes 'for 3600 part-through-wall cracks and finite length, tnrough-wall cracks, defined in this manner, are thus used as end-poirts to prescribe a second allowable flaw size locus for the subject welds, with the same safety margins, but under the assumption of lower bound, flux-type
),
'erial toughness from Figure 3-18.
The new allowable flaw size loci are
' constructed by drawing a smooth curve parallel to the corresponding
. ~ awable locus from IWB-3640 between the two end points.
Since.such a tical flaw size determination potentially reflects less than full' limit
.ad ductility in the pipe cross-section, it is also appropriate to include
- bal secondary stress terms (such as thermal expansion) in the above
$liedloading.
L
- ritical flaw size loci have been determin[d in this manner for the four.
d welds with UT indications in Section 4.0 of this report.
They are then
- ompared to the IWB-3640 based allowable flaw sizes,' as well as to the-
~
};erved flaw sizes, plus predicted IGSCC crack propagation during subse-
- 2nt operation.
l 1
i L.
l L.
^ STRUCTURAL 3~15 INTEGRITY ASSOLWESINC s
~
TABLE 3-1
(.
Suasarv of Acpliec Stress at Weld G-2-14 BR4CH nonCNTE 10 0 Myy Mir CAS!$ (TT-LES) (FT-LISI 12' PIE DW
-2222 6560 PRECSUEE:
1150 PS!
TiffML 16119 -47473 T)As 0.579
!N CSE-n 244l 2467-CD=
12.75 IN CEE-!Y 3190 3506
!=
64.5 INit!
SSE-IY 4!34 4115 SM=
16950 PSI SSE-!Y 4683 5173
/
k 94AEE 5241
-3835 STESES MRML TD MLD DLE TD imCH 199TS (PSI)
AW5LE (D!5EE)
O 10 20 30 40 50 60 70 80 90 1
5
'."A3 IDERT 1 1
1.05 1.17 1.24
!.42
- 1. 9 2.!2 2.69 3
l 5AES PES 0.6 0.61 0.64 0.67 0.69 0.75 0.66 0.E?
0.95 1
A 4ISSURE E97.150 7723.769 810;.626 B483.434 E736.722 9496.437108S9.2411269.101M3.2212661.91 INE!E DW 1:20.465 1273.700 1352.664 1478.472 1469.513 1563.674 1742.8 2 1709.483 166
(!
S M ;CI i)EJ.% 88!:.1669218. 729791. 479 10702.55 10779. 93 11:22. 78 12622. 76 123'E. 23 MIKASE713.4883 871.96791054.1501293.2714:4.9:41711.906 2162.15 2459.2210916.369 292',.Z9 CSE 1111.2551276.2o21472.6761722.541 1E90.5% 2152.900 262.220 OSM.752 32?4.383 SSE 1722 19S6.595 2214.070 2710.513 :949.868 !!6).615 4112.037 4520.207 5!52.7 %
i MS3JRE 7597.!!0 77:3.769 E!03.626 6482.434 6736.722 9496.43710829.3 !!I69.1 MSHE DW
-81;.643 -B49.139 -901.776 -985.643 -992.542 -1042.44 -l!61.89 -ll38.*S -!!!0.15 -F26.7?0 SUPJ CE THERML -5238.12 -6145.E3 -6527.65 -71!!.70 -7124.62 7fA9.19 -6415.
951NKA6E-475.658 *21.311 -702.767 -862.238 -969.949 -1141.27 -1441.41 -
CBE -740.837 -850.B61 -991.784 -1159.22 -1260.39 -14:3.93 -1755.48 -19 3 SE
~
-1152 -1324.39 -152?. a -1907.00 -1965 '2 -2241.07 -2741.52 -3011.57 -3C.15 PEES'JI + CW + Trdr.*L + Sm!*ASE teomHE 9391.506 96!7.841 10!!6.67 10729.33 !!024.00 11929.66 !!644.03 14026.S ENDINS 8971.782 9470.360 10165.24 11:29.43 11436.39 12166.14 13773.93 13728 FESSLRS+ tw + THEmL + 03E +9fRINKA6E feffsh!
9576. 716 9830. 556 10:32.12 11019. I B ! !!!9.10 12238. 64 14082. 90 14509.00 102.16 in;9. 4 (PM+P!!!Sa 0.564999 0.579973 0.612514 0.650099 0.669973 0.724994 0.830549 0.2 2 998 0.005791 0.907 PFESSJE !w + THEWL + SSE +SHR!W.GE
.Tf5t.&
9679.506 9942. 940 10519. 02 11191.13 !!!! 5. 48 12499. o! 14329. 41 147tc.:3 1 (Pedi-lE4 0.571062 0.596959 0.6205910.65965; 0.67tj79 0.7:4869 0.8409: 0.8719?5 0.9:
4064 0.9:4!3
^ STRUCTURAL 3-16 INTIGRITY -
/ECCIATESINC
1
[
TABLE 3-2
~
(!
Scu.ary cf feplied Stress at held KR-2-d'
\\
.t tFD Of C ENTE r
1.D;C Phy M:2 i
EASES (FT-LES) (FT-LES)
I g2. p;pg W
1922
-35??
EEES$E=
1150 PS!
'IFNL 12!35
- 2810 T}K 0.579
!N OSE-IY 5097 599' 00=
12.75 IN CEE-!Y 5050 4797 2=
64.5
!)t843 ESE-IY 7355 6454 SM. 16950 PSI
~~ S-I Y 7524 7042 i
SF CSE 325 146 4
STIISIES HOFA*L TO WELD Id 70 ERIKH P30iS (PS!?
AN6LE (D'EREE) 0 10 20 00 40 50 60 70 80 90 W
.pD P3'EWT 1 1
1.05 1.17 1.24 1,42 1.8 2.12 2.69 3
j ASES PRES 0.6 0.61 0.64 0.67 0.69 0.75 0.64
- 0. E9
- 0. 9'.
1
?
k.EELcE M97.150 7723.769 8102.626 S45:.484 8736.722 9496.4I710089.24112 INit E N
658.4156 710.%91778.C.617 E76.:275 '10.4412 989.9481 1149.99011E9.763154.83610
(,j SWiCE 'DT.L 7220.465 510.B37 7951.50 5664.09: 6495.519096.99410090.23 9E5.714 9476.2 5
- -RING 5E27.16279 !7.24978 48.51520 62.59478 73.99592 ?0.565'? !!B.70: 140.1507172.B682181.:?5; O!E 2008.552 2!05.959 2659.E2 2139.542 3412.614 3334.660 4E0.51 5217.176 59:9.294
{
ISE 2S?2.976 019,868 !3 3.6214540.564 4944.926 5442.647 6909.E54 7605.00 B679.970 C4.553 1
'II3)RE 7597.1D 7723.769 8103.626 E4B;.454 8736.722 9496.G710839.2411269.1012028.E212661.91 a
CLTi31tE N
-42.945 -473.672 -518.707 -584.218 -606.960 -659 C65 -766.660 -793.175 -B36.55 SW;CE TMNL -481!.64 -5007.25 -!'%1.16 -5776.05 -5796.80 -6064.66 -6726.85 -6550.47 -6317.91 -4603.
SkRING5E-18.1085 -24 E3:1 -32.!434 -41.9279 -49. 306 -60.3773 -79.!'48 -93.4338 03E
-1!39.03 -1537.23 -1772.16 -209;.02 -2275.07 -2591.24 -3167.00 -3478.11 -3959.52 + 3775.62 SSE
-1921.11 -221;.24 -2559.12 -3027.04 -3296.61 -3761.76 -4606.56 -5070.02 -5736.64
'.,".~6.27 PM!!Ji + b + TETMt + EHR!nt4E PEP.9F#if 8914. B24 9100. 210 9566. 681 10094. 03 1020. 00 11192. 68 12722. 41 13128. !7 FENDINS 6589.372 6832.2:5 7:15.272 8002.E4 E066.407 6421.257 9445.01E 9296.!S7 90 PFEESJE+ W + THEFNL + CBE +SHFlutfi tP2mE
'249. 5349484. 520 10009. 97 10607. 29 10919. 77 11240. 49 1 Z74.16 13997. 90 143:6.
(PM+fB)/!9 0.545698 0.55?*58 0.590558 0.65793 0.644175 0.698554 0.800?.:5 0.622% 0.852S9 I PFESS$5 3 + THEWL + SSE +SHRl'tt4E PE=PEl.C 9395.220 9653.521 10206.4610640.79 !!!?4.15 !?tM.13 !!9:
4.05 14395.88 15202.91 15407.12 (F?'*F F 1.9 0.554296 0.569:29 0.6021510.630574 0.659"42 0.71'.818 0.E22068 0.049:14 0.902%6 0.
f
+
i STRUCTUIML
- 3. n INTEGRITY
/ ASSOC 1/UESINC
4 I
TAf>LE 3-3 (1
Euw.uv of Anlied Stress at held tR 2 41 J
BRtKH 7 MENT!
~
1514 PNy M2r
~
"ASES (FT-L251 (FT-LRS) 12' P UT N
-33:1 2905 PREES' E:
!!!0 PS:
f J
TIE P A -6n2 42091 THX=
- 0. 5 79 (N
03E-IY 6591 45!4 00=
12.75 IN TE-lY 2745 2952
!=
64.5 IMit!
I SSE-IY E226 57 86 Srt: 16950 P5I L
n E-!Y 3?B9 4155
,.5 }* AGE 40 19 I
i
!T71SSES NORPf4. T0 hE'D DLE TO MtKH PDOTS (PS11
)
AN5LE (CE5 REEL 6
10 20 00 40 50 60 70 80 92 str Jto M0 RENT 1 1
1.05 1.17 1.04 1.42 1.8 2.12 2.69 7
- ASES PRES 0.6 0.61 0.64 0.67 0.69 0.75 0.E4 0.E9
- 0. ?!
'ESSLCE 7597.150 77:3.769 2102.6:6 B4B.434 8736.72 9496.4!7'!02??.2411769.101:0:9.2 j
(
IE!!!
IW 555 M 39 654.5 M : 770.5049 925.:437 1021.475 1191.021 1465.863 1637.049 1901.1:4 1859.162 i
.1
' SL'FftC THEMAL 7829.02; 7928.017 2175.S42 8667.624 E432.002 E512.464 9004.321817?.199 6994.646 3768.22 MIND 5E3.5M353 4.77:446 6.110:18 7.925 03 9. E95311.016214.781017 :.33:7 21.!6566 22.22553 O!E 1370.604 1651.396 1976.11 2404.E6B 2666.362 3140.434 3941.150 4454.024 5:
41.5875:10.790
}
SSE 164?.436 2215.693 2640.!03 :202. 2 1 2 66.694 4159.162 5:04.006 5864.599 6679.3 i
t "II3'JE 7597.150 7723.769 8103.626 B483.484 8736.722 9496.43710889.2411269.
ItJTS1 E Im
-370.232 -436.20 -51!.669 -616.B:9 -680.983 '-787.347 -977.242 -1091.49 -1767.4 SUPfACE TIENAL -5219.34 -52M.47 -5450.56 -5778.41 -56:5.33 -5674.97 -6002.8 ShRikG5E-2.!5659 -3.120:9 -4.10637 -5.2901: -6.19290 -7.54774 -9.85473 -11.*92 05E
-91!.7:6 -1100.93 -1317.40 -1602.24 -1790.90 -209.62 -2627.4; -2969.34 -3494.!9 -3472.E6 ESE
-12!2.99 -1477. 12 -1760.20 -21:4.90 -2377.79 -2772.!0 -3469.47 -3909.73 -4:
86.20 -4541. 9 PFESS'If + DW + T.9% + SHRIEASE EfBRA5E L.
B?95.!!4 915.022 9595.711 10093.61 10314.8511'13.9012636.74 12?o8.071M1 BECIN5 6989.922 7156.263 7460.4 2 9000.669 7890.639 8087.:39 6737.472 8194.643 PFESStFE+ tw + THEF.% + G:E +SHRlWieE 1.WhE 9222.569 9400.254 9925. 063 10494. 42 10762.57 116I7.!! 112?3. 60 13650. 41 (PMPUIN 0.544163 0.5627 0.52'549 0.6!ESD 0.634960 0.686567 0,784:S 0.00034 0.E48729 0.00308 PFISSJ!. 'm + THEFM. + SSE +SMRINWi Trin9 9 03.;32 9524.!04 100 % 76 10617.34 10909.09 11206.93 1 % 04.11 !!005.51 14659.92 147:2.94
\\
IM % SM 0.542972 0.56190" 0.59220 0.6263910.643616 0.696574 0.796702 0.819204 0.664 1
STRUCTURAL 3 18 INTEGRITY ASSOCIATESINC L
~~
T 1
(
i.
~
TABLE 3-4 ALLOWABLE END-OF. EVALUATION PERIOD FLAW DEPTH 2 TO THICKNESS RATIO FOR CIRCUMFERENTIAL FLAWS - NORMAL OPERATING (INCLUDING UPSET AND TEST) C
/
- P, Ratio of Flaw Length. /,, to Ptpe Cartum+enmce (Note (3))
e.
03
^
Ib e (2)]
0.0 0.1 0.2 03 0.4 er Mor,
.5 (4).
(4)
(4)
(4)
(4)
(4)
.4 0.75 0 40 0.21 0 15 (4)
(4) 1.3 0.75 0.75 0.39 0.27 0 22 0.19 1.2 0.75 0.75 0 56 0 40 0.32 0 27 I
1.1 0.75 0.75 0 73 0.51 0 42 0.54
}
1.0 0 75 0 75 0.75 0 63 0.51 0 41 09 0.75 0 75 0 75 0.73 0 59 0 47 l
^
. 0.8 0.75 0.75 0.75 0.75 0t8 0.53 0.7 0.75 0.75 0.75 0 75 0 75 0 Sa es 0.6 0 75 0.75 0.75 0.75 0 75 0 63-a (Ig NO' 3.
(1: an depth = J.for a surface flaw 2J.for A Swbigef 4Cf Cao
- = nomer%Il tNCkness
.wat intrepol4L40n is perm.6lible.
8 (2
- primary membrant Strell f.
- rimary be%ng stress
.,,....,,,,,,,,,,.,, i,,,,.,,.n. u o m.....,, s, n.., u n (3)
. *f ffnCf 04,ed 06 nomina' F Cf 0 Amtter.
f (4) re 3 J514.3 shall be used.
l_
~
STRUCTURAI, 3-19 INTEGRITY ASSOCIATES.lbC.
~
l
~
.k TABLE 3..
~
. F for a Circumferentially Cracked Cylinder in Tension (t/Ri = 1/10),
a/t 1/8 1/4 1/2 3/4.
1 F
1.19 1.32 1.82 2.~4 9 r
1
[
TABLE 3-6
~
h1 for a Circumferentially Cracked Cylinder.
in Tension (t/Ri = 1/10) d a/t n=1 n=2 n=3 n=5-n=7 n=10
}
~1/8 4.00 5.13 6.09 7.69 9.09.
- 11.,1 1/4 4.17 5.35 6.09 6.93 7.30 7.41 1/2 5.40 5.90 5.63 4.51 3.49 2.47 3/4 5.18 3.78 2.57 1.59 1.31 1.10 l'
STRUCTURAL 3-20 INTEGRITY assccatsme
it TABLE 2 7
(!.
F for'a Circumferential, Throujh-Wall Crack in a Cylinder of t/R = 1/10 in Tension d
~
U a/b 1/16 1/8 1/4 1/2 i
F 1.077 1.259 1.802 4.209 a
TABLE 3-8 l
h1 for a Circumferential Through-Wall Crack 1
in a Cylinder of t/R = 1/10
,j in Tension
( ;\\
a/b n=1 n=2 n=3 n=5 n=7
- 1 1/16 2.979 3.967 4.655 5.576 6.104 1/8 3.221 4.157 4.708 5.163 5.102.
1/4 3.677 4.159
'4.032 3.238 2.605 1/2 3.091 2.220 1.713 1.137 0.816 i
(.
)
i
[
^ STRUCTURE 3 2:
-INTEGRITY ASSOCIATESINC
_._ i _ _ '; I
TABLE 3-9 Material Stress-Strain Properties of Base and Weldment
(
Materials Used in Analysis b
304 750F 304 5500F-TIG 750F TIG 5500F T.e strain at Pm 0.546*
0.347 0.299 0.103
- s. ess at Pmax 149,380 88,650 121,E90 70,100 a
30.7 17.3 13.63 2.83 n
1.92 2.49 4.00 11.S4 cro 38,200 24,800 65,900 53,900 I
F EeofFit 0.166-0.883 0.04-0.888 0.114.299 0.022-0.114 i
s
(
i)
YS+
43,000 24,800 68,900 53,900 TS 86,000 62,600 00,500 63,400
{
Elong. %
80.3 45 55 28
% RA 81 70 69 69 Diametral gage
+
Cross-head measurements o
0.4" Gage length
,E = 30 x 106 psi e
0.3
^ STRUCTURA 3-22 INTEGRITY ASSOCIATESIN 4
r'
}
Table 3-10 (i
nree Commonly used Ramberg-Osgood Constants for Weldment Materials cro a
n
^
l-Primary Curve 53.9
- 2.83 11.84 3
Alternate Curve A 44.8 3.39-6.89 J
Alternate Curve B 49.4 9.0 9.8 1
t
(
~
i
.i t
l-I
~
j L
N STRUCTURAL <
3-23 INTIGRITY ASSOCIATES.INC I G
g
{'
TAB'LE 3-11 WELDING PROCESSES-FOR STAINLESS STEEL' PIPE SUBMERGED ARC WELD (SAW) s AlfT0MATIC PROCESS ARC BETWEEN BARE METAL CONSLliABLE ELECTRODE (WIRE) AND WORKPIECE ARC SHIELDED BY GRANULAR AND FUSIBLE FLUX WHICH BLANKETS HOLTEN WELD METAL HIGH WELD DEPOSITION RATE AND SPEED DISADVANTAGES SLAG MUST BE REMOVED-AFTER EACH~
y PASS TO AVOID ENTRAPMENT IN WELD METAL HIGH HEAT INPUT CAN GIVE SLOW j
COOLING RATES AND C0 ARSE,. LOW TOUGHNESS MICROSTRUCTURE gj PICKUP FROM THE FLUX CAN CHANGE COMPOSITION OF-DEPOSIT RISK OF MICROFISSURING
~
USED FOR MOST SHOP WELDS - NOT IN FIELD RELATION TO WELD METAL TOUGHNESS RELATIVELY HEAVY SLAG / INCLUSION CONTENT-C0 ARSE MICROSTRUCTURE HIGH HEAT INPUT CAN GIVE HIGHER TERRITE CONTENTS THE AB0VE CAN LEAD TO REDUCED TOUGHNESS - PROBABLY THE LOWEST FOR THE. WELD PROCESSES CONSIDERED HERE STRUCTURAL i
3-24 INTEGRITY l
ASSOCIATESINC
r TABLE 3-11 (Continued)
{>
SHIELDED METAL ARC WELD (SMAW)-
i~,
MANUAL PROCESS Ai,C BETWEEN FLUX-COVERED CONSUMABLE ELECTRODE AND.
f-WORKPIECE
'~
~
SHIELDING BY GASEOUS SHIELD AND MOLTEN FLUX OR ~ SLAG FROM E!_ECTRODE COVERING MOST VERSATILE PROCESS - POSITIONS, P
ETC.
DISADVANTAGES SLAG BLANKET - SOURCE OF INCLUSIONS VISIBILITY IMPAIRED BY SLAG SLAG REMOVAL BETWEEN PASSES IS NECESSARY f
MOISTURE PICF.UP IN ELECTRODES
~
LOW DEPOSITION EFFICIENCY USED FOR REPAIRS AND FOR.CERTAIN PORTIONS OF FIELD AND SHOP WELDS
('g RELATION TO WELD METAL TOUGHNESS INTERMEDIATE TO HEAVY INCLUSION j
C0m ENT L
INTERMEDIATE HEAT INPLTI AND DILUTION EXPECT INTERMEDIATE TOUGHNESS i_
IL
^ STRUCTURAL 3-25 INTEGRITY ASSOClKrESINC.,
m_
m
~
TABLE 3-11
~
(Concluded)
(
GAS TUNGSTEN ARC WELD-(GTAW), OR TUNGSTEN INERT GAS (TIG)
ALTIOMATIC OR MANUAL PROCESS ARC BETWEEN NONCONSINABLE ELECTRODE '(TUNGSTEN) AND T
l WORKPIECE - FILLER METAL (WELD WIRE) CAN BE ADDED T0 WELD POOL - SHIELDED BY INERT GAS (ARGON OR HELIUM)
MULTI-POSITION, HIGH QUALITY WELD, BLIT LOW DEPOSITION RATES NO FLUX USED - N0 SLAG-INSIGNIFICANT CHANGES IN FILLER COMPOSITION DURING DEPOSIT - LOW PICKUP OF CONTAMINANTS
~
USED MOSTLY FOR FIELD WELDS, SOME SHOP WELDS, AND A_L,L WELD OVERLAYS t
RELATION TO WELD METAL TOUGHNESS i
REDUCED HEAT INPlfr THROUGH PULSING GIVES FINER,
{
TOUGHER MICROSTRUCTURE AND POTENTIALLY LOWER i
FERRITE
~
(
q NO SLAG-METAL REACTIONS AND RESULTANT NONMETALLIC-I INCLUSIONS INSIGNIFICANT PICXUP 0F CONTAMINANTS THE ASCVE CAN LEAD TO THE HIGHEST TOUGHNESS FOR THE WELD PROCESSES C,ONSIDERED HERE l
L
)
e
)
i._
1
.I 1
g STRUCTURAL 3-26 INTIGRITY ASSCCIATES,1NC !
9 t
c1~
t
~
2 j~
t
'O COMPUTED 6" -
gy pgpg,
STRESSES O.7, H 2" -
MEASURED STRESSES 4
"*"/)
L#e W
o O' AZIMUT" o
+
D 90* AllMUT,4 I
O*A21WUTH PE 20
- R = 6.375" 10 0 Cm
-3
-2
-l i
2 3
i 3
0 O
3
- 1.0
-O S O
O.5 1.0 I
.k h
DISTANCE FROM WELD CENTERLINE
,,,oo e -20 Hm 200 o O
O g-40 3
o
- -30 0 J
C (1
,r er O
o 8.60 o
4*
m E
O f
~
~
WELO
-80 FUSION LINES 600 W
i PlPE SIDE SWEEPOLET SIDE L
Figure 3-1.
Comparison of P.easured and Computed Residual b ial Stresses Along Inner Surface of a Welded, lHS! Treated I?. inch o
Sweepolet j
^
STRUCTURAI, 3-27 INTEGRITY ASSOC 1/GESINC
.___.___m.-.
J
(r
~
O
.C W3 O
e u'
e v
C N
o o
O a
6 i
6 6
N g
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q
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]
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>=
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i n
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1
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e e
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og 2
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p qc
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y
!! d 5 5 5 -
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nows aosvisia m
e ycc-tisssi seia og EU NM #
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- O I W 2 E M
W > w> w O
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l j
n.
i STRUCTUngI, _
3-29 INTEGRITY i
ASSOCIATESINC l
.l
m l
E" h
8 4
c g
0 0
n i
0 0
t 5
t 5 e e
- 2.
, 0 h 4
o S,
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8 A
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+
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)
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(
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(
8
[
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me 9
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t cr n
5 D
i i g
0 3
4
(
g 3
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t n n
- 9. l E
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=
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a
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R no S
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l trt ic T T T 2
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~
30 0
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/
30 20 $_
jy 90o l
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+
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-20
~ 30
/
40
-30 50
-40 1D 00 figure 3 5.
p05t-lHS! Residual Stress Distribution STRUCT M 3 33 INTEGRITY ASSOCIATESINC-w
-.-_.-_.-._~:_--a--_-._
O_,
D D
3 I
D O
0 I
0 T
T r
A T
A T
e A
A t
S S
nl S
S S
S ea E
S E
S Cn m
R E
R E
di T
R T
R S
T S
T l
e S
S W
L L
A P
A P
n I
O I
O rr X
O X
O f
m A
H A
H e
c n
a t
o o
o is 0
D
=
c o
q O
/
o
' 7 l
R
.~
7'g
.a
\\
ois O',
t
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e.m g\\
t T
a, R
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L P
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t t
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y *'--
5 G
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T
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s 0
i
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f e'
o
?,
oN l
e #
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N-1T
' o i
- f. '.
.. 'o-
'0' 3
~-*
s 0
0 0
0 4
3 2
0
,0 0
0 1
1 1
4
_ h: y. =
- La I
~
g AH:d C
gL~
MZ8 i'
i 44 a
.I a
j p
- A e
- A., a2, 4_,3 F*A I'.
\\
g 2.
i i
R
/n 7
-/-
o
?-
z*
rz
- ab
.5 F3 I
I
+-- t -+-
4 v
i l
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c
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c, j s.-
t l,*
a,. r t h).g
- r, t ig l
rg*F7 u
i 3
0
. '3 o
.4
.6 Fra etie a13ia:a.,e e Thren;h *ct a:1 (a /.
d Figure 3-8.
Magnification factors of Circ:derential
' ~
Crack in a Cylinder (a/t = 0.1)
I I._
STEUCTURAL l
3-34 INTEGRITY ASSDCIATESINC
_ - - _ - - - - -.. _... - _. - - - - -. - - - - _ - - - ~ - - _. -. _. - -. - - - - _ _ - _. _ _ _. _. _, _
_.._._.x_.-_-.-_-.
i
~
i l
(
r.
(
so'3
~
Upper Bound (Turnace Sensitizi d) 7 da/dt : 5.65 x10-9(K)3.07
/
O 10*4 Best Estimate (Weld Sensitizec )
g
~
.A da/dt : 2.27x10-6(g )2.26
=
7 Ef t la'a t t a ' 9 9 w '.8 7
2
- 0 2 as==
g 4-1 a1 (,er's eCl 1116 *.
6
/
/
/
/
A St h s'*ill: a' lik?'8 2
- C 2 as
- C '"I # ' * # b' # CI ~ I8 t 2
to-5
/
O s r i.'.r s a t,, w'.
- 2. -
erw-o sCt tiis n 3
/
y
.h..MelTertDifggatg,gy g)
Y g
c4. n,3 322 :.,. u.
h Y
O sa=istirse AT iive..
- e i
7 y
2
@ Ct. f ttb s
@ ma=C CL.4 ear. Cg%to
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E[ALUATI0f15ANDRESULTS
' 4.0
('
4.1 Weld KR-2-14 Input to the flaw evaluation for this weld was as 'follows:
Indication Length = 2.1 inches Indication Depth = 0.1418 inch
~
Pipe 0.D. = 12.75 inches /22 inches
(
(Riser circumference used in normalizing crack length for conservatism),
Pipe Wall Thickness = 1.125 inches Applied Stresses.(Table 3-1) l Pressure + DW + Thermal + Shrinkage = 14.80 ksi membrane,.13.88 ksi bend Pressure + DW + Thermal + Shrinkage + Seismic = 15.38 ksi membrane f
i Residual Stresses (Figure 3-5)
Figure 4-1 provides applied stress intensity factor versus crack depth data k
for the two load. cases used in the evaluation (t ping loads and IHSI residual i
stress).
Assuming the indication to be IGSCC, these stress intensity curves were used to perform post-lHS! IGSCC crack growth estimates and the resulting crack growth predictions are illustrated in Figure 4-2. The analysis results -
in no predicted track growth for the balance of plant life.
~
The allowable end-of-cycle flaw size was determined in accordance with ASME Section XI, Article IWB-3640, and using the J-T procedure described in Section 3.5.
The results are illustrated in Figure 4-3 in terms of allowable flaw depth versus length.
Note that, although not required by IWS-3640, thermal expansion stresses have been included in the evaluation to account for the possible effects of low toughness weldment material.
Also, in
!~
accordance with the recommendations of NRC Generic Letter 84-11, a maximum-allowable flaw size of 2/3 of the IWB-3640 limit (shown as a dashed line in Figure 4-3)isusedtoaliowforuncertaintyinflawdepthsizing.
l t
^ STRUCTURAL 4-1 INTEGRITY ASSOCIAIESINC L-
~
Also hown in Figure 4-3 are allowable flaw size curves cal' ulated by
-(
elastic-plastic fracture mechanics (EPFM) for the three different sets of c
Ramberg-Osgood constants of Table 3-10.
It is seen that the EPTM results yield srmewhat more conservative allcwable flaw size, but compare favorably to the 2/3 of IWB-3640 limit.
~
Referring to Figure 4-3, it is seen. that the 2/3 - of IWB-3640 limit is satisfied indefinitely by the analysis, since 'no crack propagation is predicted.
To add further assurance, the IGSCC crack growth analysis has been repeated assuming various initial flaw sizes ranging upward from the observed UT depth.
No crack propagation is predicted in the post-IHS!
(
condition for initial crack depths up to 0.414 inch, or 37% of the pipe wall.
It is also noteworthy that, given the relatively short length of ' the observed indication (5.3% of circumference), it would not tad to rupture of the pipe joint.even if the above crack growth or initial f k i size. estimates are significantly in error.
Leak-before-break is clearly-the appropriate hypothetical failure mode for this indication.
On the basis of the above evaluation, it is concluded that continued s
operation of the plant with this weld, considering the observed indication and the IH5I treatment which has been applied, will not lead to a reduction in plant safety margins, or a plant operational concern.'
l 4.2 Weld KR-2-36 Input to the flaw evaluation for this weld was as'follows:
Indication Length = 2.2 inches Indication Depth = 0.2813 inch Pipe 0.D. = 12.75 inches /22 inches (Riser circumference used in normalizing crack length for conservatism)
Pipe Wall Thickness = 1.125 inches STRUCTURAL 4-2 INTEGRITY ASSOCIATESINC
i 1
Applied Stresses (Table 3-2)
{
{,
Pressure + DW + Thermal + Shrinkage = 13.85 ksi membrane, 9.09 ksi bending
{
Pressure + DW + Thermal + Shrinkage + Seismic = 14.97 ksi membrane j
Residual Stresses (Figure 3-5)
Figure 4-4 provides applied stress intensity factor versus crack depth data for the two load cases used in the evaluation (piping loads and IH5I residual
}
stress). Assuming the indication to be IGSCC, these stress intensity curves l
l were used to perform post-IHSI IGSCC crack growth estimates and the resultirg crack growth predictions are illustrated in Figure 4-5. The analysis results in no predicted crack growth for the balance of plant life.
d The allowable end-of-cycle flaw size was determined in accordance with ASME Section XI, Article IWB-3640, and using the J-T precedure described 'in Section 3.5.
The results are illustrated in Figure 4-6 in terms of allowabte flaw depth versus length.
Note that, although not required by IWB-3640,
{
thermal expansion stresses have been included in the evaluation to account for the possible effects of low toughness weldment material.
Also, in accordance with the recormtendations of NRC Generic Letter 84-11, a maxicum allowable flaw size of 2/3 of the IWB-3640 limit (shman as a dashed line in Figure 4-6) is used to allow for uncertainty in flaw depth sizing.
Also shown in Figure 4-6 are allowable flaw size curves calculated by-elastic-plastic fracture mechanics (EPFM) for the three different sets of Ramberg-Osgood constants of Table 3-10.
It is seen that the EPFM results-yield somewhat more conservative allowable flaw sizes, but compare favorably to the 2/.3 of IWB-3640 limit.
t -
Referring to Figure 4-6, it is seen that the 2/3 of IWS-3640 limit is predicted to be satisfied indefinitely by the analysis, since no crack N
propagation is predicted. To add further assurance, the IGSCC crack growth s
analysis has been repeated assuming various initial flaw sizes ranging upward from the observed UT depth. No crack propagation is predicted in the post-IH5! condition for initial crack depths up to 0.612 inch, or 54% of the
^ STRUCTURAL i
4-3 INTEGRITY i
MSOCIATESINC
,.s
T pipe wall.
It is also noteworthy that, given the relatively short length of
_l the observed indication (5.5% of circumference), it would not lead to rupture of the pipe joint even if the above crack growth or initial flaw size
~
estimates are significantly in erro'r.
Leak-oefore-break is clearly the appropriate hypothetical failure mode for this indication.
I*
On the basis of the above evaluation, it is concluded that continued operation of the plant with this weld, considering the observed indication aad the IHS1 treatment which has been applied, will not lead to a reducticn in plant safety margins, or a plant operational concern.
4.3 Weld KR-2-41 Input to the flaw evaluation for this weld was as follows:
Indication Length = 4 inches-Indication Depth = 0.2138 inch Pipe 0.D. = 12.75 inches /22 inches (Riser circumference used in normalizing crack length for conservatism)
Pipe Wall Thickness = 1.125 inches l
Applied Stresses (Table 3-3)
Pressure + DW + Thermal + Shrinkage = 12.64 ksi membrane, 8.74 ksi bending Pressure + DW + Thermal + Shrinkage + Seismic = 14.47 ksi membrane Residual Stresses Zero and Post-lHS1 (Figure 3-5)
~
Figure 4-7 provides applied stress intensity f actor versus crack depth _ data.
for the two load cases used in the evaluation (piping loads and IHS1 residual stress). Assuming the indication to be IGSCC, these stress intensity curves were used to perform post-lHS1 IGSCC crack growth estimates and the resulting crack growth predictions are illustrated in Figure 4-8. The analysis results in no predicted crack growth for the balance of plant life.
I t.
I 4-4 STRUCTURAL i e
INTEGRITY ASSOCIATESINC J-_-____ _ _ _ _
l 1
~
The allowable end-of-cycle flaw size was determined in accordance with Section XI, Article IWB-3640 and using the J-T procedure described in Se 3.5.
[
The results are illustrated in Figure 4-9 in terms of allowable flaw depth versus-length.
Note that, although not required by IWB-3640, thermal expansion stresses have been included in the eva-luation~.to account for the i
1 possible effects of low toughness weld material.
Also, in accordance with the recommendations of NRC Generic Letter 84-11, a maximum allowable c 7
size of 2/3 of the IWB-3640 limit is used to allow for uncertainty in crack depth sizing.
Also shown in Figure 4-9 are allowable flaw size curves calculated by elastic-plastic fracture mechanics (EPFM), for the three different sets of Ramberg-Osgood constants of Table 3-10.
It is seen that the EPFM results yield somewhat more conservative allowable flaw sizes, but compare f to the 2/3 of IWB-3640 limit, j
Referring to Figure 4-9, it is seen that the flaw is predicted to remain at
<d its present size indefinitely, and thus satisfy the allowable flaw size limit by a large margin for the balance of plant life..To add further assurance,
~ l the IGSCC crack growth analysis has been repeated assuming various initial flaw sizes ranging upward from the observed UT depth.
No crack propagation is predicted in the post-lHSI condition for initial crack depths up to 0 6S4 inch or 61% of the pipe wall.
~
On the basis of the above evaluation, it is concluded that continued operation of the plant with this weld, considering the observed indication will not lead to a reduction in plant safety margins, or a plant operationa concern,.
4.4 Weld KR-2-37 Input to the flaw evaluation for this weld was as follows:
Indication Length = 5 inches Indication Depth = 0.135 inch-I 4-5 STRUCTURAL INTEGRITY ASSOCIATES.INC
.__._____._._-_.___m__
___.m_. __
Pipe 0.D. = 22 inches -
{l Pipe I.0. = 19.75 inches Pipe Wall Thickness = 1.125 inches Applied Stresses Pressure = 5.622 ksi
~
Residual Stresses (Figure 3-7)
Figure 4-10 provides applied stress intensity factor versds crack depth data
. 'for the two load cases used in the evaluation (pressure'and IHSI residual stresses).
Assuming the indication to be IGSCC, these stress intensity curves were used to perform IGSCC crack growth, estimates for both cases, and t'.e resulting crack growth predictions are illustrated in Figure 4-1.
The analysis case results in no predicted crack growth for the balance of plant life.
__ The allowable end-of-life flaw size was determined in accordance with ASM
{
Section XI, Article IWB-3640, and using the J-T procedure described in
?
Section 3.5.
The results are ' illustrated in Figure 4-12 in terms' of allowable flaw depth versus length.
Also, in accordance with the recom-mendations of NRC Generic Letter 84-11, a maximum allowable crack size of 2/3 of the IWB-3640 limit is used to allow for uncertainty in crack depth sizing.
Also shown in Figure 4-12 are allowable flaw size curves calculated by elastic-plastic fracture mechanics (EPFM) for the three different sets of Ramberg-Osgood constants of Table 3-10.
It is seen that the EPFM results yield less conservative allowable flaw sizes in this weld.
Referring to Figure 4-12,. it is seen that the flaw is predicted to remain at its present size indefinitely, and thus satisfy the allowable flaw size limit by a large margin for the balance of plant life.
To add further assurance, the IGSCC crack growth analysis has been repeated assuming various initial flaw sizes ranging upward from the observed UT depth.
No crack propagation j
is predicted in the post-Ifl51 condition for crack depths up to 0.81 inches,
~
~
or 727. of the pipe wall.
l L
STRUCTURAL 4-6 INTEGRITY ASSOCIATES 1NC-
p
?
l 4
. On the basis of the - above evaluation, it. is concluded that continued.
I 1.
operation of the plant with this weld,'considering the observed indication,
)
will not lead to a reduction in plant safety ma' gins, or a plant operational l
l concern.-
t i
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b 4-1 I
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STRUCTURqL 47 INTEGRITY ASSOCIATESINC
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// Circumference Ficure 4-3.
Comparison of Predicted Crack Growth with' l
' Allowable flaw Size Limits - Weld KR-2-14 t
1 STRUCTURAL i
4 10 INTEGRITY ASSOCIATESINC.
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Comparison of Predicted Crack Growth with Allowable flaw Size Limits - Weld KR-2-36 I_
~
4-13 STRUCTURAL INTEGRITY i
ASSOCIATES INC
94
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-165CCIATESINC I
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figure 4-9.
Comparison of Predicted Crack Growth with Alloweble flaw Size Limits - Weld KR-2-41 s
4-16 STRUCTURAL
.~
INTEGRITY ASSOCIATESINC
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& Post-lHS! Flaw Growth Prediction O
~
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1~
figure 4-12.
Comparison of Predicted Crack Growth with Allowable Flaw Size Limits - Weld KR-2-37 STRUCTURAL 4-19 INTEGRITY ASSOCIATESINC
^
w-___-_.--_-_--_____--____-.--.--_.
, 5.0 DISCUSSION AND CONCLUSIONS
(
This report presents fracture mechanics flaw e.;aluations for four welds in
~
the Browns _J_erry_ Unit 2 recirculation piping system (three sweep-o-let to ring header welds, and one rino header to end cap. weld).
The four welds contained relatively small, crack-like indications.
These welds, along with the other, uncracked welds in the plant, were treated by Induction Heating Stress Improvement (IHSI) to produce a favorable residual stress pattern and thus reduce their susceptibility to IGSCC degradation.
The flaw evaluations were based on the post-lHSI' indication sizes, which differed somewhat from the pre-lHS1 inspections, but not significantly.
The evaluations presented in this r port were performed in accordance with ASME Section XI, IW8-3640 and the recommendations of tiRC Generic Letter 84-
- 11. These conventional approaches were also supplemented by Elastic Plastic Fracture Mechanics Tearing Instability analyses to account.for the possible ef fects of low toughness weld metal. The results of the analyses for all four
,3
-~
welds indicate that design basis safety margins are maintained in the welds,
~
by a 13rge margin, considering the worst case effects of the observed flaws; and that these margins are maintained indefinitely during the life of the plant, due to the beneficial effects of the IHSI treatment, which is expected to inhibit further IGSCC propagation. It is also noteworthy that all of the
" indications had circumferential lengths less than 10% of pipe circumference Thus, even in the event of large uncertainties in UT depth sizing or crack growth predictions, the governing failure mode would still be leak-before-i break.
On the basis of these factors, it is concluded that the inspection results and corrective actions taken should not result in any reduction in design-basis safety margins or ' increase in the probability of a pipe rupture at the plant.
One final point of significance is that the lHS) treatments, which were performed on a large per.centage of the remaining untracked welds, should I
,I 5-1 STRUCTURAL
\\
INTEGRITY
'\\
ASSOCiMESINC
greatIyreducetheprobabilityoffutureIGSCCinthesewelds.
7 j'
Thus, it is reasonable to expect that the plant will operate for a long period of time
(
with no further degradation due to IGSCC, anc mo reduction in leak-before-break margins re-lative to plants with piping not susceptible to IGSCC.
1 l
.i (il 1
j 1
l l
t I-1 5-2 STRUCTURAL I
INTEGRITY ASSOCIATESINQ Q--__-----__-.-
~
6.0 REFERENCES
1.
Transmittals from E. Wilson, TVA, Jan. 29, 1985 and May 3, 1985.
7
~
i 2.
SI Report, " Design Report for Recirculation. Piping Sweep-o-lets Repair and Flaw Et aluation, Browns Ferry Nuclear Power Plant, Unit 1", SIR ~
006, Sept. 1984.
3.
EPRI Report NP-2662-LD, Induction Heating of Welded BWR Pipes", December 1982." Comput 4
EPRI Report, NP-81-4-LD, " Residual Stress Improvement by Means of Induction Heating", March 1981.
y 5.
BWROG IGSCC Research Program Status Report' presented by T. Umemoto an A. Tanaka, " Application of Induction heating Stress Improvement to Pipe Branches", December 9, 1980.
6.
- Buchalet, Fracture Mechanics,1974", ASTM STP-590, pp. 385-402 H.,
" ASTM 8th National Symposiu::
on 7.
NUREG 1061, " Investigation. and Evaluation of Commission, March, 1984
~~
8.
"G~uidelines for Flaw Evaluation and itemedial Actions for Stainless
(
}-}
Steel Piping Susceptible to IGSCC", Final Report for EPRI Project T303-1, Report No. SIR-84-005, April 13, 1984.
9.
- Bickford, R.
L., et al, " Nondestructive Evaluation Instrument Sur-veillance Test on 26-Inch Pipe", EPRI NP-3393, January, 1984 10.
Ranganath, S., and Norris, D.M., " Evaluation Procedure and Acceptance Criteria for Flaws in Austenitic Steel Piping", Draf t No.
10, Sub-committee on Piping, Pumps, and Values of the PVRC of the WRC, July 1983.
f 11.
Ranganath, S., Mehta, H. S., and Norris, D.M., " Structural Evaluation of Flaws in Power Plant Piping", ASME PVP-Vol. 94, Circumferential Cracks in Pressure Vessels and Piping - Vol. I, pp.91-116, 1984 12.
ASME Boiler and Pressure Vessel Code,Section XI, 1983.
13.
ASME Section XI ' Meeting Minutes, May 25, 1984
- 14. Kumar, V.,
et al.,
"An Engineering Approach for Elastic-Plastic Fracture Analysis", EPRI NP-1931, July,1981.
- 15. Kumar, V.,
et al.,
.I EPRI NP-3607, Aug.,1984" Advances in Clastic-Plastic Fracture Analysis",
- 16. Hutchinson, J.
W.,
and Paris, P.
'., " Stability Analysis of J-C Controlled Crack Growth", in Elastic-Plastic Fracture. ASTM 668, American Society for Testing and Materials,1979, pp. 37-64.
STRUCTURAI 6-1 INTEGRITY ASSOCIAIEINC
\\
~
17.
Westinghouse Test Data, presented by J. Landes'at the meeting of ASME
(-
Boiler & Pressure Vessel Code Section XI, Task Group on Piping Flaw Evaluation, San Antonio, Texas, April 23. 1984.
18.
Gudas,-4.P., and Anderson, D. R., "J -R Curve Characteristics of Piping 1
Material and Welds", NSRDC, presented at U,5. NRC 9th Water Reactor Safety Research Information Meeting, Washington, D.C., Oct. 29, 1981.
- 19. NSRDC Test Data, presented by M. Vasslieros at the meeting of ASME Boiler & Pressure Vessel Code Section XI, Task Group on Piping Flaw Evaluation, San Antonio, Texas, April 23,1984.
20.
Paris, P.C., Brunetti, J. V., and Cotter, K. H., "The Effect of Lar9e Crack Extension on the Tearing Resistance of Stainless Stael Piping Materials", Presented at the CSitI Specialist Meeting on " Leak-Before-Break in Nuclear Reactor Piping Systems", Sept. 1-2, 1983 Monterey, CA.
21.
McCabe, D. E., Westinghouse letter to J. F. Copeland, Stainless Pipe Weldment Tests, Aug. 29, 1984.
- 22. Metals Handbook - Ninth Edition. Volume 6 - Welding Brazing, and Soldering, American Society for Metals, Metals Park, Ohio, c. 1983.
23.
- Tetelman, A.S.,
and McEvily, Jr.,
A.
J.,
Fracture of Structural Materials, John Wiley & Sons, Inc., New York, c. 1907, pp. 212-222.
(l) 24 Landes, J.
D., et al, " Elastic-Plastic Methodology to Establish R-Curves and Instability Criteria", Sixty Semi-annual Report, Jan. 1, 1982 to June 30, 1982, EPRI Contract No. RP 1238-2, Aug. 4, 1952.
I Ernst, H." A., " Material Resistance and Instability Beyond J Controlled 25.
I Crack Growth", presented at the Second International Spposium on Elastic-Plastic Fracture Mechanics. Philadelphia, PA, Oct. 1981.
I-I_
!I-L
? STRUCTURAL 6-2 INTEGRITY ASSOCIATES.INC
em APPENDIX 4 STRUCTURAL JUSTIFICATION FOR THE OVERLAY REPAIR ON WELD GR-2-15
.'~%. /.
=
STRUCTURAL JUSTIFICATICM FOR THE OVERLAY RE ON WELD CR-2-15 r
Overlav Sizine Calculations s
Weld overlay sizing calculations were performed based on a 360 0
through-wall circu=ferential crack in the 12-inch end of the 28 X 12-inch reducer. The thickness at this joint is 0.579 inch.
The resultant overlay is 0 35-inches thick and is depicted in Figure 1.
is in addition to the seal weld which id applied over the crackThe 0 35-inch thickn
,..~"
first weld layeE ~that clears dye-penetrant and the testing (PT) inspection.
Axial stresses at this joint are given as:
Pressure 6,321 psi
=
Dead' Weight
' =
1,990 psi Seismic 6,000 psi
=
Ther=al Expansion =
14,000 psi The primary stresses include pressure, dead weight, and seismic str thus the resultant stress is 14 311 psi.
The allowable stress, S, at 3
the design te=perature of 575cr is 16,675 psi.
m The pri=ary stress ratio, (Pe + P )/S is about 0.858, which results b
m, in an allowable flaw depth to thickness ratio, a/t of 0.495 ror a 360 crack, frc=_ASME Section XI, Table IWE-3641-1.
Therefore, the unrepaired joint it unacceptable; however, an overlay repair of 0.35-inch thickness results in several effects which render the repaired joint acceptable a/t ratio is reduced from 1 to 0.6232, and the primary stress ratio is The A
For this stress ratio of 0.5348, an allowable a/t ratio obtained from IWB-3641-1 for a 3600 l
a, is deter =ined to be 0.6155-inch deep. crack, and the allowed crack depth,
_Faticue Crack Greuth Consideration of fatigue crack growth during service is required to that the original 3600 allowed 0.6155-inch depth. crack of 0.579-inch depth will not extend past show the Thus, the allowance for fatigue crack growth is 0.0365 inch.
Axial stresses at this joint for heatup/cooldown cycles include pr and ther:al stresses, thus the resultant stress is 20: 321 psi.
- essure, This stress can be reduced by the unoverlaid-to-overlaid-thickness 0.6232, and this reduced stress is 12,665 psi.
- ratio, Program was used to compute the stress intensity factorThe EPRI DRIVE Computer 0.579-inch deep flaw having a stress of 12,665 psi.
, K, for a 3600,
- e intensity factor is approximately 38 ksi Vinen.
The resulting stress A weld metal fatigue crack growth curve is assumed equal to the upp of solution annealed Type 304 for DVR environments, as shewn in the er bound attached figure from EPRI Report NP-2423-LD.
corresponding to K:38 kai Vinch is about 4 x 10-4The crack grewth rate in/ cycle. Because 0
.o.=*
e
i
(
changes in K are negligible for small enounts of crack growth, it is
~
estimated that it would take 91 heatup/rooldown cycles to use up the 0.0365 inch allowance for fatigue.
, Conclusion Based on a conservative esticate of 10' heatup/cooldown cycles per year, it would take about 9 years for the crack to extend from 0.579 inch to the limit of 0.6155 inch. Thus, the joint is suitahle 'for service with the weld overlay for at least 2 fuel cycles which is the maximum that is currently accepted by NRC.
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..,CCIATES ASSOCIATES,INC.
.r r w
_ e
) Frecerics Copeiand.Ph D.
Thornas L Gerber. Ph D.
Anthony 1 Ctannu=1. Ph D.
Anthony N Muccarci. Ph D.
Peter C Ricca celta.Ph D.
l PCR-85-032 March 27, 1985 I
l 1
4 Mr. James E. Wilson Tennessee Valley Authority 1420 Chestnut Street Tower 11 1
Chattanooga, TN 37401
)
Subject:
Independent Review of the Overlay Repair for ' Weld GR-2-15' Browns Ferry, Unit 2
Dear Ed:
(
Our independent review of the overlay repair on weld GR-2-15 shows that the
(
weld overlay design on the subject weld is adequate.
Highlights of tite review for the subject weld overlay are summarized as follows:
l
. Axial stresses at this joint were calculated and tabulated in Table 1. Resulting stresses are very close to those used in the TVA analysis.
We concur with your approach of not using Code stress indices, as this is the standard approach used on all Browns Ferry, Unit 1 overlays, as well as those at most other plants.
. Based on the stresses in Table 1, a minimum thickness of 0.31 inch is required for the overlay (Table 2). The 0.35 inch thick designed overlay provides an extra 0.04 inch allowance for fatigue crack growth.
. Stress intensity factor for a 0.929 inch thick cylinder with a 0.579 inch deep, 3600 circumferential crack was calculated to be 35.1 Ksi 6n (Figure 1) which is compatible with 38 Ksi/In~given in the
~"
design analysis.
. The fatigue curve used in adequate and the 4x10-4 the design analysis was judged to be in/ cycle crack growth rate was reconfirmed.
3150 ALM ADEN EXPRESSWAY SUITE 226. SAM IOSE.C ALIFORNI A 95f ts o mormmmm o wmE ER$8mFmur~7
Page 2 PCR-85-032
. Allowable flaw size after the 0.35 inch overlay repair was evaluated and tabulated in Table 3.
It was also reconfirmed that fatigue crack growth from more than 90 heatup/cooldown cycles can be tolerated within the extra 0.04.i nc h thickness allowance.
Additional margin on cycles could also be obtained by taking credit for part of the first weld overlay layer if needed.
Should you have any further questions, please call me.
Very 9 J1
- ours, j'.
l l
/
P. C. Riccardella
/sl enc.
cc:
Frank flovak Welding Services, Inc.
G e
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i STRUCTURAL J
INTEGRITY usccmesme j
TABLE 1 Calculation of Applied Stresses TVA-06 WELD GR-2-15 Pressure =1150 psi DD=12.75 inches 2.as4.5 - in*+3 LOAD f1::
11y 11:-
11b A:: i a l Sig CASE (ft-1bf) (ft-lbf) (ft-lbf) (ft-lbf)
(psi)
PRESSURE 6~30,96 DW 3G1.00 139.00 10661.00 10661.91 1937.61 TE1 3162.00 701.00 12590.00 12617.49 2347.44 TEC 6621.00 77762.00 3232.00 739.'.2.77 13706.30 O B E->: y 1129.00 5317.00 31487.00 31932.77 5940.90 i
ODE y::
549.00 2486.00 12560.00 12000.66 2~32.03 1
1 l
l S S E->: y 1613.00 7922.00 46071.00 475~.5.76 88i3.06 SSE-y:
786.00 3509.00 17060.00 10:09.30
$337.78 i
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TABLE 2 Weld Overlay Sizing
~
pcCRACK STRUCTURAL INTEGRITY ASSOCIATES, INC.
VERSION
- 1. 0, AFRIL 1985 SAN JOSE, CA (400)970-8200-i WELD OVERLAY SIZING l
l OVERLAY SIZING FOR CIRCUMFERENTIAL CRACK:-
TVA-06, WELD GR-2-15 UALL THICKNESS = 0.5790 STRESS RATIO =
0.8550 L/ CIRCUMFERENCE O.O O.1 0.2 0. '"
O. 4' O.5--11.0 f~INAL A/T G.7500 0.7500 0.7500 O.7500 0.7500 0,6514 GVERLAY THICKNESS 0.1970 0.1930 0.1930 0.1950 0.1920 0.3099 e
9 e
t e
d9 u_________
_____.-_.______._____m_.___._-.
\\
TABLE 3 Allowable Flaw Size for Pipes with 0.35" Weld Overlay pcCRACM STRUCTURAL INTEGRITY ASSOCIATES, INC.
VERSION 1.O, APRIL 1995 SAN JOSE, CA (408)978-8200 CRITICAL FLAW SIZE EVAL'JATION CRITICAL FLAW SIZE FOR CIRCUMFERENTIAL CRACK:-
TVA-06, WELD GR-2-15
. ALL TH I CI: NESS =
0.9290 STRESS RATIO =
0.5320 L/ CIRCUM
.0
.1
.2
.3
.4
.5-->1.0 ALLOWABLE A/T O.7500 0.7500 0.7500 O.7500 O.7500 O.6635 O
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-49
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APPENDIX 5 STRUCTURAL JUSTIFICATION FOR THE OVERLAY REPAIR ON WELD DSRWC-2-5 1
1 l
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1 l
~_ _ ~ _ _ _ _ __ _
t Q TE pf4AL Vl4 OmL v i
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'870404 002
'O Att: J. M. Castltrein 00CwtNT NvusEn C. E. Anderson F43 BFN 5-165 SB-K WR CEB87081 Fnou PAGEIOF 1 K. S. Seidle W9C126 C-K 7YPE OF 00CUMENT DATE
'~
L,,_j REQUEST NEED OATE ur Ann vnig RELEASE REF.OIR NE2 REFERENCED DOCUMENTS AVAILASLE IN ONE OF THE RIMS SYTTEMS ATTACHMENT TO THis OlR DOCUMENT 10ENTIFYING NUMSER DOCUMENT ATTACHMENT NUMSER
- 1. QIR NEB 87101 129 87032A 938
~
- 2. DNE CAILL*LATION B41 870404 001 (CEB-CQS-281)
SUBJECT Design Requirunent for overlay thickness at weld DSRhC_2-5 1
SYSTEM $ AFFECTED
$Y$T " '
Reactor Water Cleanup 09 OUALITY INFORM ATION RE00ESTED / RE LEASED
'Ibe reference 1 Quality Infonmtion Release (QIR) provided design requironents for a full structural-weld overlay at weld IERWC-2-6.. EngineerLv calculations that establish the minimum overlap thickness have been prepsd and doctmented as refetulee 2.
Per these calculations, the mininrn required weld nrtal deposition in Step B of the overlay pmcess (see ref.1) is 0.~200 inches.
It is reccmnended that both the ref 1 QIR and this QIR'be referenced on the design drawing.
}
1 1
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l THE STEP A LAYER SHA LL AL SO BE L IQUID ' PENETRANT EXAMINEC PRIOR 70 BEGINNING THE STRUCTUR AL OVERLAY (STEP B),
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Cv84
~
STRESS INDICES FOR USE WITH EQUATIONS IN NS-3650 Appucable for 0,/t s 100 for C or K Indices and 0,/t s 50 for 8 Indices j
Intemal Pressure Moment Landing Thermal Leading
[ Note (1))
[ Note (1))
,,.r,.
1 B,
C, Kg 8,
C, K,
C, C',
K,
}
Piping Products and Joints [ Note (2)]
[ Note O)][ Note O)]
[P.ote O)][ Note O))
' [ Note O)]
Notes
]
i Straight pipe, remote from welds 0.5 1.0 1.0 1.0 1.0 1.0 1.0 LO.
(4) or other discontinuities
,,. ;,g,g,p,.g g,,,,
, ("..
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Q5kNNMh,' '
Y
. M,k b i M.. N2hk" TI
- (1) For the calculation of pressure and moment loads and specialinktructions reg
- Ni 3i,j.
r';, t p
- W ' W r!/.;b^,,U--g Ncy { '+.',.':. f)#
(2) For definitions, appucability, and specific restrictlens, see NS-3683.
7
-0) For special Instructions regarding the use of these ind6ces for welded products, intersecting welds,Tabutting "er eut.of-round, n.0.)O / products, su N S-3683.2.. a.Wi.- ?,@
~ '
~
' ' /M t
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c 16) See NS-3683.4(b), Gltth Butt Welds." C" g,.g.
., 5
-[(.
J
,d'I.):
' D) See N B-M83.4(c), Glath Fplet Welds.4W.
'y, (9) See N B-3683.5(b), Transitions Within a 1:3 Sloper. jr.
~ hi.
y' (8) See N B-M83.5'a), N B-4250 Transitions.Q/I..
g 1
. /7S.,
IM' (10) See NS-3683.6, Concentric and Eccentrle Reducers.d ad "o. 6 w,,,
MP e rk & p
,Q f." M
/-(*#.
'. i2y p
. 7%j u'gge (11) See NS-%83.7, Curved Pipe or Butt Wevding Elbows.'See also NBJ3641.2(a) and NS %83.2(b).,9t'b,
g d 6 ?.L:
(12) See NB.%83.8, Branch Connections per N B-%43.Su also NS-%83.Md).g.Spli. WC'.C' A, d Mp s.i, j
i '.'.".
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PT - Penetrant examination RL - Refracted longitudinal wave S - Shear wave l
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