ML19242B797
| ML19242B797 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/12/1979 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 7908090416 | |
| Download: ML19242B797 (48) | |
Text
/
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wisconsin Elecinc roara couvarr 231 W. MICHIGAN, P,0. BOX 2046. MILWAUKEE. WI 53201 July 12,1979 Mr. J. G. Keppler, Director Office of Inspection and Enforcement Region III U. S. NUCLEAR REGULATORY COMMISSION 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
DOCKET NOS. 50-266 AND 50-301 RESPONSE TO IE BULLETIN 79-13 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On May 25, 1979, the Office of Nuclear Reactor Regulation informed all PWR licensees of the occurrence of cracking in welds in the feedwater nozzle-to-piping welds at the D. C. Cook Unit 2 facility and requested infor-mation concerning similar welds at other PWR plants. Our letter of June 20, 1979, provided the requested information and stated our intent to inspect the steam generator feedwater nozzle welds of bott. Point Beach units during the week of July 1,1979. On June 25, 1979, we informed you by telegram of our intent to proceed with ultrasonic examination only of the feedwater line welds during hot shutdown of the Point Beach units beginning June 30, 1979.
Subsequently, NRC issued IE Bulletin 79-13 on June 25, 1979, which required radiographic examinations of feedwater nozzle-to-piping welds and adjacent pipe and nozzle areas within 90 days of the date of the bulletir.. Upon receipt of the bulletin and after identification of a packing leak in a Unit 2 ten-inch Residual Heat Removal System valve, it was decided to proceed with Unit 2 to cold shutdown on June 30 which then allowed repair of the RHR valve and provided the opportunity to perform radiographic and ultrasonic examinations of the feedwater line welds.
Eight Unit 2 feedwater piping welds have been inspected, including the weld connecting each three-inch auxiliary feedwater pipe to each main feedwater pipe.
Linear indications and several small cracks were found during radiographic and ultrasonic examinations of the Unit 2 welds.
These indications are described in our July 6,1979, 24-hour notification letter.
Representatives of Wisconsin Electric Power Corpany, Southwest Research Institute, and Bechtel Power Corporation met with NRR and IE personnel in Washington on July 6 to discuss the results of our inspections and the expected extent of the repairs.
The enclosure to this letter summarizes information presented at our meeting and presents additional feedwater piping system information requested by NRC personnel during the meeting, including results of all testing and metallurgical examination undertaken to date.
n'A M.-'J JUL 13 7979 7908090 g
- Mr. J.. G. Ke'pple t, Di rector July 12, 1979 As discussed in more detail in'the enclosure, the Unit 2 feedline welds are being repaired. While completing these repairs, the feedwater piping in the inspected area is being reconstructed in a manner similar to the original construc-tion. No significant changes are being made fion the original construction requirements and we have determined that the repair work does not constitute an unreviewed safety question.
A 10 CFR 50.59 review has been performed by the Manager's Supervisory Staff and the results of this review have been documented.
Further, metallurgical laboratory examinations of the welds which were removed for replacecent indicate that the welds did not present an unsafe condition and, had this knowledge been available earlier, the repairs would not have had to be made.
At tne completion of these repairs, baseline radiographic and ultra-sonic examinations of the affected welas will be performed.
These examinations will be the reference for the next inspection of these welds which is tentatively scheduled for the Spring 1980 refueling of Unit 2.
It is anticipated that the repair effort will be conpleted by July 15 and the unit will be returned to power generation by July 17, 1979.
As we discussed at the July 6 meeting, we intend to perform the inspec-tion of the Unit 1 feedwater piping welds during the scheduled Fall refueling outaae for Unit 1.
This outage is tentatively scheduled to begin on September 28, 1979.
It is our intention to continue to foll7w developments relating to feedwater line cracking and to review the experience of other operating PWR plants.
Very truly yours,
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Sol Burstein Exe tive Vice President Enclosures Copy to: Office of Inspection and Enforcement Division of Reactor Operations Inspection Mr. A. Schwencer, Chief Office of Nuclear Reactor Regulation E
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EtiCLOSURE Response to IE Bulletin 79-13 July 10,1979 This attachment provides Wisconsin Electric Power Company's response for Point Beach fluclear Plant Units 1 and 2 to flRC IE Bulletin fio. 79-13 entitled
" Cracking in Feedwater System Piping". This attachment contains analytical and operating infomation for both Point Beach Units and inspection results for Unit 2.
Unit 1 inspection results will be provided at a later date.
In addition to responses to each of the bulletin items (fa,nraphrased form), additional infomation is also provided as requested by the f;RC staff during our meeting of July 6, 1979. Also enclosed are simplified isometric sketches of each steam generator and adjacent feedwater piping; see Figures 1 thru 4 attachad.
1.
Examine feedwater nozzle-to-piping welds and piping supports.
RESP 0:lSE The Unit 2 welds identified on Figures 3 and 4 have been inspected and repaired.
Appendix A, attached, provides details of the radiographic (RT) and ultrasonic (UT) examinations on the main feedwater piping welds.
The RT evaluation and UT examination and evaluation were perfomed by Southwest Research Institute personnel.
Appendix B, attached, provides a sumary of indications found during other examinations, such as the steam generator feedwate" nozzle I.D.
surface, and the corrective actions that have occurred due to this inspection.
Appendix C, attached, provides an interim raport of the preliminary metallurgical evaluation of indications found in the "A" loop reducer.
The laboratory examinations were performed at the Southwest Research Institute.
Figures 1 thru 4 tabulate the design stress levels from the steam generator nozzle weld (data point 1) thru the check valve located closest to the steam generator.
The original piping analysis did not consider the reducer; the stress results for data point I are for a 16 inch diameter pipe. Thus, the actual stresses are less than those tabulated.
- However, based upon the original piping analysis, there are no known weld locations in the feedvlater piping where either the themal stress or the summation of the seismic stresses exceed 90% of the code allowable stresses. Thus, no other welds have been examined.
The Unit 2 feedwater piping supports and restraints have been visually inspected for confomance to the design requirements as shown on Bechtel drawing M-2409, revision 7 which was.provided to fiRC by our letter of June 20, 1979. The supports appear to be in confomance with the original design requirements, but some additional verification of spring constants is still being pursued.
Sc;3 216 a
2.
Perform additional piping weld inspections.
RESP 0tiSE The Unit 2 inspections will be performed during the Spring 1980 refueling. The Unit 1 inspections will be performed during the Fall 1979 refueling.
3.
Effects of cracking indications on multiple nuclear unit facilities.
RES"Of4SE As shown in Appendix C, by the metallurgical examination, the in-dications discovered during the inspection af Unit 2 are gererally on the order of 0.043 inches deep on the inside surface of the weld. The difference between the piping mill nominal and minimum wall thickness is 0.117 inches for 18 inch, schedule 80 pipe (tmin is 0.820 inches) and 0.104 inches for 16 inch, schedule 80 pipe (tmin is 0.739 inches) per ASTM A106 -75a, Table A2.
In addition, the analytical minimum wall thickness required to withstand the operating pressure of 1050 psi (per Af1SI B21.1, paragraph 104.1.2, equation 3) is only 0.625 inches, including an 0.08 inch corrosion allowance.
The general indications discovered during the Unit 2 examination do not reduce the sound metal wall thickness below the mill minimum wall thickness or the analytically required minimum wall thickness and thus do not coastitute a safety hazard.
At full power with all feedwater heaters in use, the feedwater temperature entering containment is about 425 F.
In the event of a unit trip, inlet water temperatures may drop to about 80"F.
The resultant pipe metal temperatures would still be above the range of carbon steel brittle-to-ductile transition temperatures. Thus, the feedwater pipe system is not believed to operate in a non-ductile temperature environment.
During the July 6, 1979 meeting with f!RC the Unit 2 inspection results were discussed, with the laboratory investigations showing that the Point Beach fluclear Plant indications were not serious, and do not have any apparent safety significance.
On this basis, it was generally agreed that the Point Beach indications were not serious. The Unit 1 annual Fall refueling outage is tentatively scheduled for only a few days beyond the expiration of the 90 day inspection interval specifieJ in IE Bulletin 79-13, and, as discussed at the meeting, the Unit 1 inspections will be performed during that refueling outage.
4.
Report indications to NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.
RESPONSE
The existence of linear indicat.ans and cracks in the Unit 2 welds (as discussed in Appendices A and B) was positively identified late on July 5 by the Southwest Research Institute metallurgical laboratory.
fiRC was advised of this during cur meeting of July 6 and a written 24-hour notification was sent to flRC Region III on July 6.
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Provide a report identifying the inspection schedule, adequacy of applicable o.
operating and emergency procedures, and ability to detect feedwater leaks in?ide containment.
_Rr3PONSE The intital inspections required by Item 1 of the Bulletin have been completed for Unit 2.
Comparable Unit 1 inspections will be performed during the Fall 1979 refueling outage.
The operating and emergency procedures in use at the Point Beach Nuclear Plant are adequate to recognize and respond to a feedwater line break accident. The indications and symptoms of the accident are dependent upon the location and size of the break. A feedwater line break upstream of the fir.al check valve at the steam generator feedwater nozzle would result in a reactor trip. The reactor trip would probably be caused by a steam flow / feed flow mismatch coincident with a low water level in either steam generator or by low-low water level in either steam generator.
Upon the ' trip of the reactor, the plant operators would follow emergency operating procedure E0P-5A for an emergency shutdown of the reactor.
A smaller break upstream of the check valve cculd be of insuffi-cient magnitude to cause a sustained steam flow / feed flow wismatch.
In
~
such an eventuality, the reactor would be tripped on a two out of three low-low water level signal in either steam generator.
The low-low level signal in either steam generator would also automatically start the motor driven feedwater pumps. As discussed in Section 14.1.11 of the Final Facility Description and Safety Analysis Report (FFDSAR), the loss of normal feedwater flow caused by a feedwater line break does not result in any adverse conditions.
In the event of a feedwater line break in the short run of piping between the final check valve and the feedwater nozzle of the steam generator, the same reactor trip functions would occur and the plant operators would be presented with symptoms and indications essentially identical to those of the rupture of a steam pipe.
Plant E0P-2A, " Steam Line Break", addresses the operator action in the event of such a rupture.
Under certain circumstances, such as a feedwater pipe crack, the feedwater leakage may not be of sufficient magnitude to manifest itself in a reactor trip.
In such a situation there are several excellent leakage detection methods to reveal the presence of significant leakage levels in containment. The humidity detection instrumentation offers one means of detecting low level leakage.
Plots of containment air dew point variations above a bass-line maximum, established by cooling water temperature to the containment air coolers, can determine incremental leakage equivalent to 2 to 10 gpm. The sensitivity of this method is dependent upon the cooling water temperature, containment air temperature variation, and containment air recirculation rate.
A second leak detection method is ' based on the principle that the condensate collected by the containment cooling coil matches, under equilibrium conditions, the leakage of water and steam from systans within containment.
Measurement of the condensate drained from each of the fan cooler units can
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be made to determine condensate rate and thus leak rate.
Should a it.. accur, the condensation rate will increase above the previous steady state due to the increased vapor content of the air.
Tnis condensate from the fan coolers is piped to sump A in both containments.
A high iiquid level in the sump is alarmed in the control room.
It takes approximately 22 gallons to actuate the Unit 1 alarm and about 42 gallons for an alarm in Unit 2.
Since sump contents removal requires manual operation, the operators would be alerted to an increase in total containment leakage rate. As with the containment humidity indica-tions, if an increased total containment condensation rate is apparent, a containment entry and inspection would be made to determine the source of the leakage.
6.
Submit a written report to NRC presenting the pipe weld inspection results.
RESPONSE
This submittal constitutes Wisconsin Electric's response to this IE Bulletin with respect to Item 1 for Point Beach Nuclear Plant Unit 2.
When the metallurgical examination report for Unit 2 is completed it will be forwarded to NRC. This is tentatively scheduled for mid-August 1979.
The Bulletin Item 2 results for Unit 2 will be provided after the Spring 1980 refueling.
The report for the Unit 1 inspections for both Items 1 and 2 of the bulletin will be provided within 30 days of the Unit 1 inspection.
This is expected to be in early November 1979, following inspections during the Fall refueling.
The following are additional items discussed at the July 6,1979 meeting with NRC staff.
7.
Steam Generator Feedwater Chemistry The Point Beach Nuclear Plant steam generator feedwater chemistry was originally based upon " phosphate control".
However, the units were converted to the "all volatile treatment" (AVT) method in September 1974 for Unit 1 and in November 1974 for Unit 2.
In Wisconsin Electric's letter of August 18,1978 (Steam Generator Operating History Questionnaire) to Mr. K. R. Goller of the NRC, the feedwater chemistry specification for AVT control is presented. Table 1 attached hereto is the same as the table attached to the August 18, 1978 letter. The feedwater chemistry is maintained in accordance with specifications for AVT control.
8.
Feedwater System Transients Significant operating events that would have caused thermal transients to the feedwater piping system are listed in Tables 2 and 3.
These tables have been developed from information contained in the ' Operations Section of the Monthly Semi-annual and Annual Operating Reports which are periodically submitted to the NRC.
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I AVT CONTROL, SECONDARY CHEMISTRY SPECIFICATIONS LII'.ITED NORMAI, STEAM CINEPATOR BLo*/CO*O:
NOR".AL POWER CPERATION PCNER o?IRATION*
<l5% POWER O?SRATION 12T LAYt#
pH 8.5 - 9.0
- 8. 5
' 9. 2 8.0 - 10.0 10.0 - 10 5 Catior. Cor.ductivity, p.hos/cm
<2.0
<7.0
<7.0 Anonia, ppm
<0.25
<0.5 SodLum, p;n
<0.10 40.5
<0.5 Chlorida, ppm
<0.15
<5.0
- Silica,
,r.p m
<l.0 Frac Erdroxide, ppm as CACO 3
<0.15 0.15 - 1.0**
<0.15
<1.0 Suspcnded Solids, ppm
<l.0 El:wdem F2 c, spm Continuous as kequired
.Maximu:a Available
<5.0
<1CO Oxygcn, ppb 73 - ygg Eyd:::ir.e, pps
?
TEE thlA'IER pH 8.8 - 9.2 (g Total Conductivity,12mhos,cm
<4.0 C:").0xygen, ppb
<5.0 Same As Normal Same as Norr.al As Required
- 1 Excess Hydra ine, pab 5.0
==
Power Operation Power Operation Ey Situatien
% Ccpper, pyh.
<5.0 Iron, ppb
<10 O,. yh=
gap.
P "G 7' CO:nENS ATE
-Z Cation Conductivity, v2.os/cm
<0.2 M
P Period of operation within these limits should not exceed two weeks
- Pericd of operation within these limits should not exceed twenty-four hours.
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TABLE 2 LIST OF SIGilIFICArtT OPERATIllG_
EVEliTS FOR FEEDWATER PIPIllG SYSTEM TRAi4SIEllTS (Based upon Operating Reports)
P0lf4T BEACH tluCLEAR. ANT U'11T 1 DATE EVENT 7/70 Hot functional test completed.
11/2/70 Initial criticality.
11/22/70 Turbine trip from 40% power.
11/29/70 Power escalation to 70% and unit trip.
12/4/70 Turbine and reactor trip from 425 MWe.
12/5/70,
Turbine trip from 70% power.
12/6/70 Unit trip from 70% power.
12/18/70 Unit trip from 80% power.
1/4/71 Auxiliary feedwater injection.
1/8/71 Turbine trip from 82% power.
1/9/71 Turbine trip from 70% power.
1/27/71 Unit trip from 90% power.
1/28/71 Turbine trip from 92% power.
2/3/71 Turbine trip from 90% power.
2/4/71 Unit trip from 90% power plus a turbine trip.
2/9/71
' Unit trip from 80% power.
2/26/71 Load runback from 450 MWe to 200 MWe.
5/29/71 Reactor trip from no-load.
7/2/71 Reactor and turbine trip.
8/29/71 Reactor and turbine trip.
595
<:31 TABLE 2 UtilT 1 - C0t4T'D DATE EVEllT 9/7/71 Load runback from 425 fWe to 260 MWe.
9/18/71 Load runback from 480 MWe to 390 IUe.
12/3/71 Reactor and turbine trip.
1/3/72 Reactor and turbine trip.
1/19/72 100% load rejection test / reactor and turbine trip.
2/12/72 Reactor trip during startup.
2/13/72 Load runback of 20%.
4/13/72 Load runback of 20%.
4/21/72 Reactor and turbine trip.
7/3/72 20% step increase in power.
9/11/72 Turbine and reactor trip from 99% power.
7/2/73 Reactor and turbine trip.
8/13/73 Reactor and turbine trip.
1/11/74 Reactor and turbine trip.
1/18/74 Reactor and turbine trip.
2/3/74 Reactor and turbine trip.
8/2/74 Reactor and turbine trip.
9/25/74 Reactor trip from 99% power.
10/4/74 Reactor and turbine trip.
2/27/75 Emergency shutdown with reactor and turbine trip.
11/16/75 Emergency shutdown with reactor and turbine trip.
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TABLE 2 UNIT 1 - CONT'D DATE EVENT 1/10/76 Load runback of 20%.
l 1/14/76 Reactor trip.
11/30/76 Load runback of 20% from 90% power.
2/21/77 Reactor and turbine trip.
4/5/77 Reactor and turbine trip.
1/7/78 Reactor and turbine trip.
2/9/78
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Reactor and turbine trip.
4/2/78 Reactor and turbine trip.
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TABLE 3 LIST OF SIGrtIFICAtiT OPERATIrlG EVEtiTS FOR FEEDWATER PIPIT;G SYSTEM TRAi4SIEtiTS (Based upon Operating Reports)
POINT BEACH llVCLEAR PLAtlT UillT 2 DATE EVEtlT 12/22/71 Normal plant heatup.
5/30/72 Initial criticality.
8/4/72 Turbine trip from 20% power.
8/18/72 Unit trip from 10% power.
12/7/72 '
Turbine trip from 20% power plus a turbine and reactor trip from 10% power.
1/14/73 Reactor and turbine trip.
2/18/73 Reactor and turbine trip.
3/8/73 Received 100% power license.
3/9/73 Reactor and turbine trip.
3/10/73 Reactor and turbine trip.
3/14/73 Reactor and turbine trip.; - 2.
3/24/73 Reactor and turbine trip.
3/26/73 Reactor and turbine trip.
3/30/73
' Reactor and urbine trips - 2.
4/8/73 Reactor and turbine trip.
5/30/73 Reactor and turbine trip.
6/19/73 Reactor and turbine trip.
n4 10/13/73 20% load runback.
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Gv 12/15/73 Reactor and turbine trip.
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i TABLE 3 UNIT 2 - C0 tit'D DATE EVENT 7/2/74 Reactor trip from 0% power.
12/27/74 Reactor and turbine trip.
2/11/75 Reactor and turbine trip.
2/24/75 Reactor trip from 10% power.
8/19/75 Reactor trip from about 10% power.
1/14/76 Manual unit trip.
2/21/76 Load runback of 20%.
4/8/76 Unit trip.
5/7/76 Load runback of 20%.
6/1 3/76 Load runback of 20% from 100% power.
9/3/76 Reactor trip.
1/12/77 Reactor and turbine trip.
6/28/77 Reictor and turbine trip.
7/7/77 Reactor and turbine trip.
1/10/78 Reactor and turbine trip.
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APPEl1 DIX A Wisconsin Electric Power Company Point Beach 14uclear Plant SUINARY OF RADIOGRAPHIC NID ULTRAS 0: llc EXAlliflATIONS FOR UNIT 2 iMIri FEEDWATER PIPIiiG WELDS July 1979 e
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p.- t RADIOGRAPilIC REVIEW t
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OnJuly2,1979,S.A.Wenk,SwRILevelIIIRT)reviewedthefollowingradio-graphs at the Point Beach Unit 2 Generating Station.
The radiography was per-formed by personnel from the Superior Industrial X-Ray Corp., and the sensitiv-ity was in compliance with the requirements of NRC IE Bulletin 79-13.la.
Radio-graphic technique details are shown on the Superior Industrial X-Ray review form, a copy of which is contained in each weld film packet.
The results of the July 2,1979, review are tabulated below:
Steam Generator A Inch Reducer-to-Pipe Weld 1._
At radiographic station marker #32, a 1/2-inch long linear indication, 1/2 inch from the edge of the center bead on the pipe side of the weld.
2.
At radiographic station marker #38, a 5/8-inch long linear indication on pipe side.
3.
At radiographic station marker #40, a 1/2-inch long linear indication on pipe side.
Steam Generator A Inch Nozzle-to-Reducer Weld 1.
Starting a' radiographic station marker #10 and extending to station marker #16 a continuous linear indication which is crack-like in appearance.
2.
Between radiographic station markers #42 and #43, there are two transverse linear indications each 1/4-inch long extending from the weld crown on the pipe side of the weld into the base metal.
Steam Generator B Inch Reducer-to-Pipe Weld 1.
Linear indication between radiographic station markers #2 and #8, 3/4 inch from root bead centerline on the pipe side.
2.
Linear indication between station markers #7 and #10, 3/4 inch from root bead centerline on the pipe side.
3.
Linear indication between station markers #12 and #14, 3/4 inch from root bead centerline on the pipe side.
4.
Linear indication between station markers #14 and #23, 5/8 inch from root bend centerline on the pipe side, very pronounced between station markers
- 19 and #23.
5.
Linear indication between station markers #46 and #49, 5/8 - 3/4 inch from root bead centerline on pipe side.
Steam Generator B Inch Nozzle-to-Reducer Weld' 1.
Linear indication between radiographic station markers #10 and #12 at the edge of the cap weld 1/2 inch from the centerline of the weld bead closest to the reducer.
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2.
Linear indication between station markers #22 and #24 in the same area as described in (1) above.
3.
Cluster of porosity at station marker #27, 3/8-inch long but separated.
4.
Linear indication between station markers #36 and #32 in the same area as described in (1) above.
5.
Linear indication between station markers #34 and #38 at the base of the bead on the reducer side.
6.
Linear indication between station markers #41 and #43, 3/8 inch from centerline of the bead described in.(1) above.
7.
Linear indication at station marker #42 + 3/4 inch extending into the redu:er transverse to the weld, the indication is 1/2-inch long.
8.
Suck-back or concavity of root bead between station markers #43 and #44.
9.
Linear indication between station markers #49 and #53 in the same atea described in (1) above.
10.
Transverse linear indication at station marker #56 + 1/2 inch. Due to nature of image, visual examination is reco= mended.
S. A. Wenk SwRI Level III - RT e
Cd5 Le L J/-
RADIOGRAPillC REVIEW On July 4,1979, S. A. Wenk, SwRI Level III RT, reviewed the following radio-graphs at the Point Beach Unit 2 Generating Station. The radiography was per-formed by personnel f rom the Superior Industrial X-Ray Corp., and the sensitiv-ity was in compliance with the requirements of NRC IE bulletin 79-13.la.
Radio-graphic technique details are shown on the Superior Industrial X-Ray review form, a copy of which is contained in each weld film packet.
Weld IB-16 A Side Pipe-to-90* Elbow 1.
Starting at radiographic station marker 0*and continuing through station marker #13 is a linear indication 7/16 inch from centerline of middle bead
\\
on the pipe side.
2.
Linear indication between station markers #6 and #8, 7/16 inch from center line of middle bead on the elbow side.
3.
Linear indication described in (1) above continuing from station marker
- 13 through station marker #26.
4.
Linear indication between station markers #16 and #17 in the center of the middle bead.
5.
Linear indication at station marker #37, 3/8 inch from center of middle bead on pipe side.
6.
Linear indication described in (1) above between station marker #44 and station marker 0.
Weld IB-16 B Side Pipe-to-90* Elbow 1.
Linear indication between station uarkers #7 and #13, 1/2 inch from center-line of middle bead on the pipe side.
2.
Linear indication described in (1) above for this weld continuing between station carkers #13 and #26.
3.
Crater at station marker #30, 5/8 inch from centerline of middle bead on elbow side.
d.
Linear indication between station markers #28 and #39 on elbow side of adddle bead.
5.
Linear indication between station markers #28 and #39, a continuation of the indication described in (1) above.
6.
Linear indication between station markers #39 and #49 on elbow side of middle bead.
7.
Porosity at station marker #49 in center of middle bead, code acceptable.
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The linear indications described above could possibly be I.D. machining marks, and should be verified by visual and penetrant examination.
S. A. Wenk SwRI Level III - RT O
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ULTRASONIC EXAMINATION OF STEAM GENERATOR SECONDARY FEEDWATER N0ZZLE WELDS AT POINT BEACH NUCLEAR GENERATING STATION NO. 2 On the first, second, cnd third of July 1979, the welds and adjacent base metal for the feeduater reducer-to-nozzle, pipe-to-reducer, and elbow-to-pipe welds, on the feedwater piping of the steam generators f or Point Beach Nuclear Generating Station No. 2 were examined in accordance with the SwRI Procedure 600-3, Rev. 46 insofar as was practicable.
This procedure required O' longi-tudinal and 45' and 60* shear wave examination from both sides of the w.1d.
Calibration was performed on Standard No.18-CS-X.688-CS-PTB.
This procedure requires, as e minimum, attenuation measurements, 0* longitudinal, 45* shear transverse to the pipe axis, and 45' and 60* shear examinations.
The basic calibration block used for these examinations was 18-CS-X.688-CS-PTB.
All three examination angles were used on each weld except as noted below.
All three examination angles was not performed on each weld. Exceptions were:
(1)
The nozzle-to-reducer and reducer-to-pipe weld on steam generator A were examined using 0* longitudinal mode only.
(The 45* and 60* shear wave examinations were not performed.)
(2)
The nozzle-to-reducer and reducer-to-pipe weld on steam generator B were examined using 0* longitudinal and 45* shear wave.
60* shear wave examinations were not performed.
The not performed examinations were at the perogative of the management of Point Beach Nuclear Generating Station No. 2 because of the decision to remove the reducer.
Evaluation of Ultrasonic Indications on Steam Generator B 18 Inch Reducer-to-Nozzle Ueld The 45' shear wave examinations from the downstream nozzle side of the weld showed only two indications exceeding 50 percent of the calibration.
These indications damped on the crown and were resolved as geometrical in nature, probably counter bore-to-crown reflections.
The 45' shear wave examinations from the reducer or upstream side of the weld disclosed a low level signal for 360* of the circumference with several indica-tions exceeding the 50 percent recording level. However, in accordance with Paragraph 8.0 of the ultrasonic procedure used, indications regardless of their amplitude which, in the opinion of the operator, are caused by non-geometric conditions should be recorded.
Precise locations of those portions of the signals which exceeded the mandatory 50 percent calibration were duly recorded. Resolution of these indications indicated that they were crack-like in nature not caused by geometry. These signals were also detected from the veld crown with the ultrasonic beam being directed from the nozzle toward the reducer.
(This information was utilized in resolving the indications.)
n}
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Figure 1 is a polar ordinate plot showing the relative position and amplitude of the ultrasonic signals as well as the reported radiographic indications.
It should be noted however, that the ultrasonic stations proceeded clockuise from top dead center when f acing the direction of flow and the radiographic station markers proceeded counterclockwise from top dead center facing the This difference has been corrected for all plots and the director of flow.
plots are clockwise f acing the direction of the flow.
The inch stations on the outside of the polar plot Figure 1 are the ultrasonic The radiographic station markers were corrected to agree station markers.
with the ultrasonic station markers and so plotted.
This provides concise correlation between the radiographic indications and the ultrasonic indications.
No recordable indications were noted on either 0* longitudinal examination or the 45" transverse examinations.
Evaluation of Ultrasonic Examination of Steam Generator B Feeduater 16 Inches Pipe-to-Reducer Weld The 45' shear wave examination from the reducer side of the weld indicated only of one point uhere the ultrasonic signal reached an amplitude of 50 percent the calibration amplitude. No signals were noted which appeared to travel for any significant length.
The 45' shear wave from the pipe side of the weld showed an indication which traveled for 360* of the weld circumference exceeding 50 percent of the cali-This indication was also detectable when bration level at several places.
examined fron the weld crown with the shear wave directed from the reducer to the pipe or opposite the mandatory scan.
Figure 2 is a polar ordinate plot The showing the position and relative amplitude of the ultrasonic signals.
radiographic indications were also plotted to show a cecrelation between the ultrasonic and radiographic indications.
No reportable indications were noted during cirher the 45* transverss scan or the 0* longitudinal scan.
Meetinns, Discussions, and Further Examinations A meeting was held between EwRI, the Superior Industrial X-Ray, and Point Beach Nuclear Cencrating Station management on the morning of 2 July 1979.
The results of the examinations perforced on the night and morning of 1 and 2 July were discussed.
to the management of Point Beach During the discussion it was pointed out Nuclear Generating Station that although the indications were indicative of crack like defects, more data would be necessary to positively identify the source of the ultrasonic signals as eminating frcm cracks.
However, the in the opinion results of the radiographic and ultrasonic examiantions were, sufficiently conclusive of Point Ecach Nuclear Cencrating Station management, This was to indicate the necessity for mechanical removal of the reducer.
performed during the af ternoon and evening of 2 July 1979.
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Results of Ultrasonic Examinatlon of the 16 Inches Elbow-to-Pine Veld on Steam Generator B on 3 July 1979 ultrasonic examinations were performed on the elbow-to-pipe weld for both steam generators A and B.
During the 45" shear wave examina-tion it was again noted that there was a singular signal trackable for 360*
of the pipe circumference.
However, in most cases this signal did not exceed 50 percent of the calibration amplitude.
The one point at which the signal exceeded 50 percent of the calibration amplitude is clearly indicated in Figure 3.
This plot is from the data taken from the pipe side of the weld.
Figure 3 is a polor ordinate plot showing the respective positions for the ultrasonic and radiographic examinations.
No recordable indications were noted on the 60* shear wave examination from the pipe side of the weld.
No recordable indciations were noted using che 45* shear wave from the elbow side of the weld.
No recordable indications were noted udring the 45* transverse examination of the weld nor were any reportable indications noted during the O' longitu-dinal examination of the weld.
Results of Ultrasonic Examinction of the 16 Inches Elbow-to-Pipe h' eld on
-~
Steam Generator A The 45* shear wave examination of the pipe side of the elbow-to-pipe weld disclosed an ultrasonic signal preceding 360* around the circumference cf the weld.
Figure 4 is a polar ordinate plot showing the position where the ultrasonic signal exceeded 50 percent of the calibration amplitude as well as the positional location of the radiographic indications.
60* shear wave examination from the pipe side of the weld showed 2 places where the signal amplitude exceeded 50 percent of the calibration arplitude.
It was noted that the 60* shear uave signal was not trackable through the full 360 degrees.
An ultrasonic signal was also detected using 45* shear wave from the top of the veld crown toward the pipe ride of the weld.
Thus, confirming the absence of geometry as a source for the reficcted signal.
The 45' shear wave examination of the elbow side of the weld showe? a singular indication 360* around the circumference of the weld.
This signal was less than 50 percent of the calibration amplitude for the majority of the cir-cumference. Those areas where the signal amplitude exceeded 50 percent of the calibration amplitude were noted.
The 60* shear wave examination of the elbow side of the weld showed that the signal ampliutdc exceeded 50 percent of the calibration aeplitude in six places. The signal was not discernable except in the area adjacent to those areas exceeding 50 percent of the calibration aeplitude.
These locations were noted on Figure 5.
No recordable indications were found on the 45*~ transverse and 0* longitudinal scans of the weld.
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Supplementary ultrasonic examinations were performed af ter comparing data with radiographic examinations particularly in the areas where radiographic examina-tions disclosed transverse indications. These areas were located and ultrasenic examinations indicated presence of low amplitude signals, however, these signals were not detected during a circumferential or transverse 45" examination.
They were observed caly when the ultrasonic transducer was placed at a high skew angle on the base naterial adjacent to the veld or on the veld crown itself.
This information is included as supplementary data and was not plotted for reporting purposes.
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1 APPEllDIX B 7
Wisconsin Electric Power Company Point Beach fluclear Plant
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Su!>. MARY OF SURFACE I:lDICATI0 tis At40 CORRECTIVE ACTIUM5 FOR UtilT 2 FEEDWATER PIPING WELDS 9
July 1979
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1.0 Steam Generator "A" Feedwate. Pipeline 1.1 Feedwater Nozzle Numerous pits were observed on the inside surface of the feedwater nozzle. These wert ground out and two areas were repair welded.
These areas were each approximately 1/8" x 5/8".
The pits were generally shallow, with one or two about.040 inches deep. The rest were less than.020 inches deep. Most were on the bottom half of the nozzle within about 2b" of the edge.
On a relative basis, the "A" nozzle had less pitting than the "B" nozzle.
After repair of the nozzle, a dye penetrant examination of the outside and inside of the nozzle was perforned with acceptable resul ts. The weld preparation was then completed per Detail II in Phillips-Getschow welding procedure IA-MA 13, Revision 07-07-79.
Prior to welding to the reducer, the nozzle weld preparation was radiographed with acceptable results.
1.2 Reducer The weld preparation on the new reducer was made somewhat differently than on the original reducer. The counterbore was machined to a depth of ik" in order to remove its end from the area of the weld.
The transition angle frcm the counterbore to the inside of the reducer was machined to an angle less than ten degrees and a small radius was ground where the counterbore met the slope. The reducer ends were inspected with dye penetrant, ultrasonics and radiography with acceptable results.
1.3 Auxiliary Feedwater Connection Visual examination of the three-inch branch connection weld from the inside of the pipe indicated a lack of penetration in the root pass of the weld. This connection was cut off completely and the weld redone. The three-inch pipe was shortened approximately 3/8" during this operation. A radius was ground on the inside edge of the penetration in the main feedwater piping.
Structural reinforcement (a Weldolet) is being used for thi? connection to meet the requirements of B31.1.
1.4 Elbow to Pioe Weld The outside surface of the elbow to pipe weld was ground flat, polished and blended into the pipe in order to facilitate radiography.
No outer diameter indications were observed.
Linear indications were observed in the radiographs.
Also two tranverse (perpendicular to the weld centerline) indications, one approximately k" deep by 3/8" long and another 1/8." away that was 1/8" deep by
" long were noted. These indications were repaired.
The inside surface of the weld was then ground and polished and examined with y Q"
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A linear indication seven to eight inches long was observed near the top of the pipe on the pipe side of the weld.
In the same plane, near the bottom of the pipe, a machining grove was noticed approximately four to five inches long.
Both were removed by grinding.
After the inside of the weld had passed a dye penetrant examination, a radiograph revealed an intermittent linear indication from the 12 o' clock to 3 o' clock position facing upstream.
This was found to be a combination of lack of fusion and a line of slag against one side-wall and another line of slag between weld beads.
These were apparently left during the original submerged arc welding. Two areas, one approximately three inches long e-d one approximately four inches long were ground completely through the weld to remove the slag.
The weld was then repair welded.
2.0 Steam Generator "B" Feedwater Pipeline 2.1 Feedwater Nozzle, Numerous pits were observeu on the inside surface of the feedwater nozzle. These were ground out and repair welded.
The repair welding covered approximately 25 square inches at the bottom of the nozzle extendinc from the edge to 24" to 3"-inward. The pitting was generally worse on this nozzle than on the "A" nozzle.
Several pits were approximately 1/8" deep.
A dye penetrant examination of the inside and outside of the mzzle revealed three or four linear indications in the inside and eight on the outside. These were shallow and were ground out.
After repair of the nozzle, a dye penetrant examination of the inside and outside of the nozzle was performed with acceptable results. The weld preparation was then completed per Detail II in Phillips-Getschow's welding procedure, IA-MA-13, Revision 07-07-79.
Prior to welding to the reducer, the nozzle weld preparation was radiographed with acceptable results.
2.2 Reducer This new reducer was machined and inspected the same as the "A" reducer.
2.3 Auxi.liary Feedwater Connection Visual examination of the three-inch branch connection weld from the inside of the pipe revealed a hole in the weld approximately one-quarter inch deep. This connection was cut off completely and the weld redone. Dye penetrant checks of the three-inch pipe end prior to rewelding revealed several indications so the pipe was shortened approximately lk".
A radius was ground on the inside edge of the penetration in the main feedwater piping.
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Structural reinforcement (a Weldolet) is being used for this connec-tion to meet the requirements of B31.i.
2.4 Elbow to Pipe Weld The outside surface of the elbow to pipe weld was ground flat, polished and blended into the pipe in order to facilitate radio-graphy. No outer diameter indications were observed. The radio-graphs revealed several linear indications.
The inside of the weld was then ground, polished and dye penetrant examined.
A three to four-inch long indication near the top and a five to six-inch long indication near the bottom of the pipe on the pipe side of the weld were ground out.
Also observed and ground out were two spots of '
porosity. Minimum wall thickness was maintained and no repair welding was required.
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J APPEfiDIX C Wisconsin Electric Power Company Point Beach fiuclear Plant IflTERIf4 REPORT OF THE liEThl. LURG 1 CAL EVALUATIO:( OF Ui4IT 2 i'A" IMIttTLEouATER PiPI::G PEDUCER July 1979 A
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INTERIM REPORT METALLURCICAL ANALYSIS OF POINT BEACH 18-IN. TO 16-IN. REDUCER 1.0
' Introduction On July 5, Southwest Research Institute was supplied one 18-in. to 16-in, reducer, removed from the A steam generator of the Point Beach Unit 2 Generating Station.
The reducer was removed by cutting through the 18-in.
reducer to nozzle and the 16-in. reducer to pipe welds.
In general, the cut passed through the weld crown at the 0.D. and in the heat-affected zone of the root pass at the I.D. on the reducer side of the weld.
The weld fusion line at the 1.D. was not present over most of the circumference.
2.0 Spec _imen Preparation To date, only the 18-in, nozzle to reducer weld has been examined.
Radiographs performed at Point Beach by Superior Industrial X-Ray Corp.
and interpreted by Mr. S. Wenk of SwRI showed a crack-like indication from station marker #10 to #16, 63* - 100*.
Two transverse linear indi-cations, each 1/4 in. long, wera also detected at station markers #42 and
- 43, -265* running f rom the weld crown into the reducer base metal.
Before removing any material from the reducer, ultrasonic inspection was performed to more accurately position the defect causing the RT indi-cations.
Thin examination confirmed the existence of a flaw which was clearly present from stations 110 to #21 (62* - 130*) and #23 to #25 (144* -
157*).,
The flaw was positioned approximately 3/8 in. from the end of the reducer, that is, in the vicinity of the transition from the counter-bore to full reducer I.D.
Only the section from stations #8 to #26 was examined using ultrasonics.
Af ter location of the flaw giving rise to the RT indications, a ring 1-1/2' in to 2 in, wide was cut from the 18-in. end of the reducer. No lubricant was employed during this or any subsequent cutting operation.
The ring was then cut in half, through stations f!0 and #28, 0 and 180*.
Each half was examined visually using a stereoscopic microscope at mag-nifications up to 50X.
No crack could be unambiguously identified at the counter-bore tran-sition.
In the zone from station #10 to #20 the oxide was unusually rough and porous in appearance. Over most of the circumference the oxide at the counter-bore transition was not noticeably different from that in other regions.
Randomly distributed pit-like areas were present, but were not grouped linearly as at stations #10 - #20.
Cracks could be positively iden-tified at the root pass fusion line at stations #37 - #37.5, #42- #45, and
- 45.5 and #49 (-230" and 260* - 310*).
Following this examination, specimens for metallographic polishing were removed at station #10, #15, #22, #28, and #43 (63*, 94*, 138*, 180*,
and 270").
Sections 1/2 in. wide were removed adjacent to stations #10, rei 1
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- 15, and #43 for fractographie examination.
These sections were notched in the region of the counter-bore transition, cooled in liquid nitrogen, and broken open.
This procedure causes the remaining sound material to cicave, thereby facilitating identification of the subcritical crack.
3.0 Results The structure cf the reducer in all sections was normal for mild steel, Figure 1.
No microstructural abnormalities, such as islands of martensite, weld porosity, or lack of fusion, were evident in the veld or he't-affected a
Zone.
Oracking was found in all the sections taken. The depth of the largest crack and the number of cracks in each section are given in Table I.
In a.11 cases, multiple cracking was observed, Figures 2 through 4.
Although cracks were observed in all areas of the counter-bore, from the weld fusion line to the original reducer I.D., cracking with depths greater than 0.015 in, was restricted to the weld fusion line and counter-bore transition region.
Of the two, the counter-bore transition zone contained the deeper cracks, Figure 3.
tbny of the cracks nucleated at grooves created during the coarse machining operation used to counter-bore the reducer.
The mouths of most of the deeper cracks were wide and oxide filled, creating a pit-like appearance.
However, several small cracks, and one larger crack in Fig-ure 3, were not associated with large pits.
This indicates that pits form after cracks have nucleated, rather than being a precursor of cracking.
The presence of wide pits at the mouths of larger cracks, and the absence of such pits at small cracka may be interpreted as evidence that the larger cracks have been present for a considerable length of time.
The extent of oxidation along crack walls, which is considerable even in the vicinity of crack tips, Figures 2, 3, and 7, also suggests that the crack growth rate is small.
The operating conditions experienced by the reducer, 450*F f a pure water, do not normally cause rapid oxidation, and thus a considerable length of time would be required to cause the degree of oxidation observed.
Figure 5 shows the fracture surf ace of specimens cut at stations #10 -
10.5 and #43 - 43.5.
The black zone near the edge of the sample is the oxi-dized inse rvice crack, the silver area, laboratory cleavage fracture.
Both It is samples were broken in the vicinity of counter-bore transition zone.
of interest to note that one specimen was notched and impacted at room temperature.
In this case, the inservice crack blunted considerably, but no crack extension occurred.
Fractography has only been performed on specimens in the as-fractured condition. No cleaning or oxide removal has yet been attempted.
Figure 6 shows a typical region near the crack tip.
The surface is intergranular in appearance, with grain size comparable to that of the base metal, Figures 1 and 7.
This suggests that the mode of cracking is intergranular.
This ob-servation should not be interprcted as definitive since the possibility re-mains that the grains imaged are of the thick oxide often present near the crack tip, Figure 7, rather than of base metal.
This point will be clarified 1]ib
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3 TABLE I NWIBER AND DEPTIl 0F CRACKS Maximum Depth Number of Cracks Section (in.)
Deeper than 0.02 in.
- 10 0.018 8
- 15 0.043 13
- 22 0.014 4
v28 0.008 3
- 43 0.042 7
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Energy dispersive x-ray analysis was performed on all fracture surfaces.
Iron was the only major constituent of the spectrum, with very small copper and sulphur peaks. With extended count times, peaks due to phosphorus, zinc, nickel, and titanium could also be resolved.
These results are consistent with normal operating conditions, since Admiralty Brass condenser tubes and Inconel-600 steam generator tubes are present in the system.
No indica-tion of contamination of the system ty chlorides or sodium hydroxide could be found.
It is postulated that cracking is the result of corrosion fatigue.
The morphology of cracking is similar to that sometimes observed in con-ventional steam plants subjected to fatigue loadings.(1)
Such failure could be intergranular at low cyclic stresses and transgranular at higher stresses, similar to the behavior observed for sensitized austenitic stain-less steels in pure water environments.(2)
The origin of the cyclic stress is not immediately apparent.
The cir-cumferential orientation of the cracks indicates that the primary stress is axial, rather than a hoop stress.
Further analysis is necessary to deter-mine if cracking is symmetrical about the vertical axis.
Such a finding would indicate that reverse bending in the horizontal plane could be the primary cause of f ailure.
The preliminary findings reported here suggest that this is unlikely.
Thermal stresses, caused by cold water eddy currents, can generate sufficient cyclic stress to cause corrosion fatigue.
The stress in such a situation is biaxial in nature.
Unless an axial bias is imposed, thermal yclic stress is expected to cause both axial and circumferential cracking.
Further analysis is required to substantiate whether axial cracks are present.
4.0 conclusiors Linear indications in radiographs of the 18-in. reducer to nozzle weld aren'were caused by the presence of oxide-filled cracks up to 0.043 in. deep.
Multiple crack nucleation occurred at stress concentrators, such as the weld fusion line, machining grooves, and the counter-bore transition zone.
All cracking was restricted to the reducer internal diameter.
No 0.D. cracks were observed.
Oxide-filled pits are associated with deeper cracks.
All cracks were difficult to detect visually because of this thick oxide.
Cracks were shallow, less than 0.043 in., and appeared to have been present for a lengthy period.
The brittic to ductile transition temperature of the reducer material appears to be below room temperature,
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5.0 References 1.
Barer, R. D., " Boiler and Turbine Component Failures,"
Meta 11orraphy in Failure Analysis, Plenum Press, New York, 1977.
2.
Shoji, T.,
- Isc, T.,
Takahashi, H.,
and Susuki, M.,
Corrosion, Vol. 34, p 366, 9/78.
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HIGHER MAGNIFICATION OF CRACK SHOWN IN FIGUR5 4.
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