ML17333A201
| ML17333A201 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/20/1995 |
| From: | Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1173F, GL-92-01, GL-92-1, NUDOCS 9512060297 | |
| Download: ML17333A201 (16) | |
Text
PRIGRITY 1
CELERATED RIDS PROCESSING)
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9512060297 DOC.DATE 95/11/20 NOTARXZED: NO DOCKET FACXL:50-315 Donald C.
Cook Nuclear Power Plant, Unit 1, Indiana M
05000315 50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Xndiana M
05000316 AUTH.NAME AUTHOR AFFILIATION FXTZPATRICK,E.
Indiana Michigan Power Co. (formerly Xndiana
& Michigan Ele RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
R
SUBJECT:
Forwards response to NRC GL 92-01,Rev 1,Suppl 1, "Reactor Vessel Structural Integrity."
DISTRXBUTXON CODE:
A028D COPIES RECEXVED-LTR ENCL SIZE:
TITLE: Generic Letter 92-01, Rev 1, Suppl 1 Responses (Reactor Vesse NOTES:
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1 RECIPIENT ID CODE/NAME HICKMAN,J NRR/DE/EMCB NUDOCS-ABSTRACT RES/DE/MEB NRC PDR COPXES LTTR ENCL 1
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U N
NOTE TO ALL"RIDS" RECIPIENTS:
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Indiana Michigan Power Company PO Box 16631 Columbus, OH 43216 lNQIANdL MICHIGAN PQWKR November 20, 1995 AEP:NRC 1173F Docket Nos.:
50-315 50-316 U ~
S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.
C.
20555 Gentlemen:
Donald C.
Cook Nuclear Plant Units 1 and 2
RESPONSE
TO NRC GENERIC LETTER 92-01, REV. 1, SUPPLEMENT 1 REACTOR VESSEL STRUCTURAL INTEGRITY
References:
(1) Letter from Indiana Michigan Power to the NRC AEP:NRC:1173E, dated August 16, 1995 (2) Letter from Wisconsin Public Service (WPS)
Corporation to the NRC, dated August 21, 1995 The Nuclear Regulatory Commission (NRC) issued Generic Letter(GL) 92-01, Revision 1,
Supplement 1 ~
"Reactor Vessel Structural Integrity" on May 19, 1995.
This supplement required that addressees identify, collect, and report any new data pertinent to analysis of the structural integrity of their reactor pressure vessels (RPVs) and assess the impact of this new data on their RPV integrity analysis.
Issues to be addressed included the requirements of section 50.60 of title 10 of the Code of Federal Regulation (10 CFR 50.60),
10 CFR 50,61, Appendices G and H to 10 CFR 50, (which encompass pressurized thermal shock and upper shelf energy evaluations) and the potential impact on low temperature overpressure protection limits or pressure-temperature limits.
The generic letter supplement requires licensees to provide the following information within 90 days:
(1) a description of those actions taken or planned to locate all data relevant to the determination of RPV integrity, or an explanation of why the existing database is considered complete as previously submitted.
Indiana Michigan Power responded to item (1) above in our 90 day response letter AEP:NRC:1173E dated August 16, 1995.
In that letter we committed to review additional RPV material data available in various other databases as noted in reference 1 and initiate confirmatory follow-up communications with sister plant q5120b02 K 05000 97 q51120
U.
S. Nuclear Regulatory Commission Page 2
AEP:NRC:1173F owners.
We also committed that, upon completion of the said
- review, we would respond to items 2, 3, and 4 of the GL supplement within six months of the date on the supplement.
The attachment to this letter is our six month response to items 2, 3,
and 4 of the GL supplement.
Review of the various industry databases for Cook Nuclear Plant was completed by ATI Consulting Company and AEPSC personnel.
This review identified no significant changes in the existing best-estimate chemistry data, which have been summarized in the attachment, Therefore, there are no necessary revisions to the evaluation of RPV integrity in accordance with the requirements of 10 CFR 50.60, 10 CFR 50.61, Appendix G and H to 10 CFR 50 nor impact on the LTOP or P-T limits.
Currently, some of the information from the databases is considered as interim data pending further verification.
Therefore, the responses to items 2, 3, and 4 are considered our interim response, and we will continue to re-evaluate the new information for validation and notify the NRC of any significant changes should new data become available.
Sincerely, pm+
E.
E. Fitz atrick Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS.~i< DAY OF 1995 tary Publ c My Commission Expires:
pit Attachment
F. /
"r~
/p
~I
~f'
U.
S. Nuclear Regulatory Commission Page 3
AEP:NRC:1173F CC:
A. A. Blind G. Charnoff H. J. Miller NFEM Section Chief NRC Resident Inspector
- Bridgman J.
R. Padgett
ATTACHMENT TO AEP NRC 1173F
ATTACHMENT TO AEP:NRC:1173F Page 1
Indiana Michi an Power Com an Donald C
Cook Nuclear Plant Units 1 and 2
Generic Letter 92-01, Revision 1, Supplement 1
Interim Report on items 2, 3 and 4 Part 2
(six month response) of the generic letter required addressees to respond to the following items within six months:
(2) an assessment of any change in best-estimate chemistry based on consideration of all relevant data; (3) a determination of the need for use of the ratio procedure in accordance with the established Position 2.1 of Regulatory Guide 1.99, Revision 2,
for those licensees that use surveillance data to provide a basis for the reactor pressure vessel (RPV) integrity evaluation; and (4) a written report providing any newly acquired data as specified above and (1) the results of any necessary revisions to the evaluation of RPV integrity in accordance with the requirements of 10CFR50.60, 10CFR50.61, Appendices G and H to 10CFR Part 50, and any potential impact on the low temperature overpressure protection (LTOP) or pressure-temperature (P-T) limits in the technical specifications or (2) a certification that previously submitted evaluations remain valid.
Revised evaluations and certifications should include consideration of Position 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data.
Responses to each of the items listed above for both units are as follows.
UNIT 1
RESPONSE
TO ITEM 2 The reactor vessel for Cook Nuclear Plant unit 1 was fabricated by Combustion Engineering.
The material property data were gathered from:
a)
EPRI's surveillance capsule database (PREP3),
b) Westinghouse Owners Group's database (RPVDATA), c)
Combustion Engineering's Report "Reactor Vessel Group Records Evaluation
- Program, Phase 2 for the Donald C.
- Plates, Forgings, Welds and Cladding" (ABB-RVG), d) NRC's "Reactor Vessel Integrity Database" (RVID),
and e)
Oakridge National Laboratory's "Power Reactor Embrittlement Database" (PR-EDB), and were compared for the beltline materials, i.e., the base metal and
ATTACHMENT TO AEP:NRC:1173F Page 2
the weld metal.
For the base metal (plate B4406-3),
there is no data indicating variation of the chemistry.
Therefore, the search focused on the vessel beltline and surveillance capsule welds for unit 1.
The welds of interest are heats
- 1P3571, 13253/12008 and 13253.
Weld wire, heat 1P3571, exists in the circumferential weld of the unit 1 reactor vessel, which is part of the beltline material for PTS evaluation.
This material exists in welds of Kewaunee, Maine
- Yankee, and LaSalle reactor vessels.
This material also exists in the surveillance capsules at Kewaunee, Maine Yankee, and LaSalle 1
and possibly Hatch 1 surveillance capsules.
Cook Nuclear Plant surveillance capsule program does not contain 1P3571 weld wire material.
Using methodology previously approved by the NRC, the 1P3571 chemistry values for copper and nickel content in the unit 1
weld have historically been calculated as the average of "best-estimate" copper and nickel contents of the known sister plants.
The assessment'f all relevant data for heat 1P3571 shows the average of the 62 valid chemistry measurements to be 0.264 wtS Cu and 0.748 wtS Ni.
These data are shown in the attached Table 1 and are consistent with the information provided by Wisconsin Public Service (WPS) in their August 21,
- 1995, submittal for Kewaunee (reference
- 2) with small corrections noted in Table 1.
The "best-estimate" chemistry for this heat in Cook Nuclear Plant unit 1 is the mean of all the measured values for copper and nickel as specified in C.l.l of Regulatory Guide 1.99, Revision 2.'he corresponding chemistry factor is 205.3 deg F, as determined from the tables in the regulatory guide.
The current licensing basis "best-estimate" chemistry data for this weld heat are 0.28 wtS Cu and 0.74 wtS Ni, with a chemistry factor of 208.7 deg F.
This value of 208.7 is conservative with respect to the "best-estimate" mean chemistry value of 205.3 deg F determined from all the measured data.
- Thus, there is no need to update the existing licensing basis values for the circumferential weld in Cook Nuclear Plant unit 1.
Southern Nuclear Operating Company (SNOC) has identified two additional data points for chemistry values of copper'and nickel for weld wire 1P3571.
These points are based on the analysis of the surveillance weld material from Hatch Nuclear Plant unit l.
The validity of this chemistry information and the presence of this weld wire in the Hatch reactor vessel has yet to be confirmed by SNOC.
Pending validation of the Hatch 1 data, it has not been included in the calculation of average chemistry values for Cook Nuclear Plant unit 1.
The "best-estimate" average copper and nickel content from Hatch 1 are 0.28 wt% Cu and 0.76 wtS Ni. After validation, if it is included into the calculations, it will not significantly change the calculated average of copper and nickel contents.
ATTACHMENT TO AEP:NRC'1173F Page 3
The tandem weld wire 13253/12008, which is in the lower shell axial weld of Cook Nuclear Plant unit 1 vessel, also exists in Fermi 2, Fitzpatrick, Maine
- Yankee, and Fort Calhoun vessels.
For the tandem weld heat 13253/12008, there is only one chemistry measured data point available from a test at Wylie Labs, as shown in Table 2, with chemistries of 0.21 wt% Cu and 0.86 wtS Ni.
There is no known weld qualification data or other test data available for this weld heat.
The corresponding chemistry factor is 206.6 deg F as determined from the tables in Regulatory Guide 1.99, Revision 2.
The current licensing basis "best-estimate" chemistry data for Cook Nuclear Plant unit 1 axial welds are 0.28 wtS Cu and 0.74 wtS Ni, with a chemistry factor of 208.7 deg F.
This is conservative with respect to the "best-estimate" mean chemistry value of 206.6 deg F from the measured data.
- Thus, there is no need to update the existing licensing basis values for the axial welds in the Cook Nuclear Plant unit 1 vessel.
Weld heat 13253 exists in the Cook Nuclear Plant unit 1
surveillance weld specimens and not in the vessel welds.
Table 3
shows the measured data for weld heat 13253.
Chemistry measurements are available from the Cook Nuclear Plant unit 1 and Salem 2 surveillance capsule material.
These data do not have any effect on the welds in the Cook Nuclear Plant unit 1 vessel for pressurized thermal shock (PTS) evaluations.
RESPONSE
TO ITEM 3 Since Cook Nuclear Plant uses the average chemistry values from all three sister
- plants, there is no need for ratioing the weld chemistry factor for any of the vessel belt line weld heats.
- Thus, there is no ratio effect for the Cook Nuclear Plant unit 1
material.
RESPONSE
TO.ITEM 4 The chemistry factor limits of the controlling materials submitted in earlier letters AEP:NRC:0894M dated June 22,
- 1990, and AEP:NRC:0561D dated August 7,
- 1990, bound the new chemistry factors.
Therefore, there is no impact of the above evaluations on
UNIT
RESPONSE
TO ITEM 2 The reactor vessel for Cook Nuclear Plant unit 2 was fabricated by Chicago Bridge and Iron (CB&I) Company.
The material property data were gathered from RVID, RPVDATA, PREP3 and PR-EDB databases and were compared for the beltline materials, i.e, the base metal and weld metal.
ATTACHMENT TO AEP:NRC:1173F Page 4
In reviewing all the vessel materials, the limiting material for Cook Nuclear Plant unit 2 reactor vessel is plate heat number C-5556-2; there is no data indicating variability of chemistry data in this material.
The common weld material that is present in Cook Nuclear Plant unit 2 is weld wire heat
- S3986, which also exists in the reactor vessel welds of Brunswick 1, Brunswick 2, Peach Bottom 2, and Quad Cities 2.
Several weld qualification and supplemental tests were performed by CB&I for this heat number.
Surveillance weld specimen chemistry data is available from Brunswick 1, Cook Nuclear Plant unit 2, and Trojan power plants.
The assessment of all relevant data as noted in Table 4 shows the average of the ten valid chemistry measurements to be 0.055 wt% Cu and 0.937 wt% Ni.
By comparison, the copper content from the Cook Nuclear Plant unit 2
surveillance weld is exactly the same (0.055 wtS), and the nickel content of 0.97 wtS is slightly higher than this average.
Although this is not the limiting vessel
- material, the use of the ratio procedure for these weld chemistries would produce an adjustment factor equal to 1.0, therefore, there would be no effect.
RESPONSE
TO ITEM 3 The limiting beltline material for Cook Nuclear Plant unit 2
reactor vessel is the base metal plate, heat number C5556-2.
There is no data indicating variability of chemistry in this material.
Thus there is no ratio effect for the controlling material in Cook Nuclear Plant unit 2.
ESPONSE TO ITEM 4 The controlling material and the chemistry factor remain unchanged from the already docketed information.
Therefore, there is no impact of the above evaluations on the PTS, LTOP or P-T limits for Cook Nuclear Plant unit 2.
ATTAeHMZNT To.AZF:NRC:1173P Table 1 Summary ofProperties for Weld Wire Heat No. lP3571 Compilation ofMeasured ChemistriesPom all Data Sources Heat No.
Flux Type Flux Lot Pct. Cu Pct. Ni Source Reference 1P3571 1P3571 Linde 1092
'958 Linde 1092 3958 0.4 0.37 0.82 0.75 CB,WQ M1.42 C.-E 78-12 RSP CE,WQ M1A3 C-E 78-12 RSP 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571
'P3571 1P3571'P3571 1P3571 1P3571 1P3571 1P3571 1P3571 A'3571 Linde 1092 3958 Linde 1092 3958 Linde 1092 395S Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 0.18 0.74 KWE,SC "P" WCAP-13257 0.19 0.35 0.17 0.73 0.74 0.72 KWE,SC "P" KWE SC "P" KWE,SC "P" WCAP-13257 WCAP-13257 WCAP-13257 0.434 0.2 0.8 0.77 KWE,SC "V" WCAP-13257 KWE,SC Unirr.
WCAP-8107 0.15 0.19 0.19 0,34 0.54 0.71 0.67 0.72 KWE,Supp. "P" KWE,Supp. "P" KWE,Supp. "P" KWE,Supp. "R" WPS, 8/21/95 WPS, 8/21/95 WFS, 8/21/95 WFS, 8/21/95 0.17 0.18 0.186 0.196 0.209 0223 0.2 0.64 0.67 0.689 0.803 0.795 0.871 0.7 KWB,Supp. "R" WFS, 8/21/95 KWE,Supp. "R" WPS, 8/21/95 KWE,Supp. "S" WPS, 8/21/95 KWB,Supp. "S" WPS, 8/21/95 KWE,Supp. "S" WFS, 8/21/95 KWE,Supp. "S" WPS, 8/21/95 KWE,Supp. "V" WPS, 8/21/95 Q.066 Q.736 KWE,SC "R" WCAP-13257 0.207 0.769 KWB,SC "R" WCAP-13257 0.214, 0.816 KWE,SC "V" WCAP-13257 RPVDATAdeveloped by ATIConsulting Note: WQ = Weld Qualification, SC Surveillance Capsule, Supp.
Supplemental Test, Surv. = Surveillance Matl.
ATTACHMENT TO AEP: NRC:
7 3F TnMo I (cprtf
)'eat No.
Flux Type
.Flux Lot Pct. Cu Pct. Ni Source Reference IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP357f IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 IP3571 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 4
Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 '958 Lindc 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 0.19 0.17 0.17 0.22 0.43 0.24 0.22 0.23 0.2 0.22 0.22 02 022 02 0.22 0.2 0.22 0.23 0.21 0.21
.0.21 0.37 0.35 0.34 0.67 0.61 0.51 0.73 0.78 0.74 0.8 0.79 0.75 0.75 0.79 0.76 0.83 0.74 0.73 0.73 0.8 0.82 0.8 0.8 0.78 0.8 0.76 0.73 KWE,Supp. "V" WPS, 8/21/95 KWB,Supp. "V" WPS, 8/21/95 KWB,Supp. "V" WPS, 8/21/95 KWE,Surv. Test WPS, 8/21/95 KWB,Surv. Test WPS, 8/21/95 KWE,Surv. Test WPS, 8/21/95 KWE,Surv. Test WPS, 8/21/95 KWE,Surv. Test WPS, 8/21/95 LSI,SC 443 LSI,SC 444 LSI,SC 447 GE-NE-A166-1294-Rl GE-NE-A166-1294-Rl GF NE-A166-1294-Rl LSI,SC 44F GE-NE-A166-1294-Rl LSI,SC 44LD GE-NE-A166-1294-Rl LSI,SC 44M LSI,SC 44U LSI,SC 45D LSI,SC 45E LSI,SC 45K GE-NE-A166-1294-Rl GE-NB-A166-1294-Rl GE-NE-A166-1294-Rl GF NE-A166-1294-Rl GB-NE-A166-1294-Rl LSI,SC 45M GF NE-A166-1294-Rl LSI,SC Unirr.
WPS, 8/21/95 MY,Supp. C04-01 WPS, 8/21/95 MY,Supp. C04-02 WPS, 8/21/95 MY, Supp. C04-03 WPS, 8/21/95 LS I,SC 44A GE-NF A166-1294-Rl RPVDATAdeveloped by ATE Consulting 2
Note: WQ Weld Qualification, SC = Surveillance Capsule, Supp.
Supplemental Test, Surv.
Surveillance Matl.
ATTACHHENT TO AEP: NRC:
73F Table 1 (Cont.)
Heat No.
Flux Type Flux L'ot Pct. Cu Pct. Nl Source Reference 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 1P3571 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 Linde 1092 3958 0.33 0.78 0.33 0.77 0.31 0.78 0.32 0.78 0.32 0.78 Q.32 0.78 0.3 0.76 0.38 0.8 Q.53 0.81 0.52 0.8 0.25 0.66 0.33 0.705 0.25 Q.7 0.36 0.78 0.432 0.745 Q.356 Q.728 MY,SC Unirr.
CR-75-269 MY,Supp. C04-04 WPS, 8/21/95 MY,Supp. C04-05 WPS, 8/21/95 MY,Supp. CQ4-07 WPS, 8/21/95 MY,Supp. C04-08 WPS, 8/21/95 MY,Supp. C04-09 WPS, 8/21/95 MY,Supp. C04-IQ WPS, 8/21/95 MY, Supp. C04-11 WPS, 8/21/95 MY,Supp. C04-13 WPS, 8/21/95 MY,Supp. CQ4-14 WPS, 8/21/95 MY,Supp. C04-15 WPS, 8/21/95 MY,SC 253 deg WCAP-12819 MY,SC 253 deg WCAP-12819 MY,SC 263 deg BCL-585-21 MY,SC 263 deg BCL-585-21 MY,SC 263 deg BCI 585-21 Avg. Cu =0.264 Avg. Ni~0.748 Note:
~
denotes corrections to the values submitted by Wisconsin Public Service Corp.
- 1) nickel content was changed from 0.78 to 0.73 wt'/o
- 2) nickel content was changed &om 0.70 to 0.705 wt/o
~ ~>> 3) one entry with.Cu = 0.33, Ni= 0.71 for surveillance capsule 263o was deleted as itwas a duplicate entry.
CE ~ Combustion Engineering, KWE= Kewaunee, LS1 LaSalle 1, MY Maine Yankee WCAP ~ Westinghouse Report, WPS = Wisconsin Public Service Corp.,
GE-NE-A166-1294-Rl ~ GE Nuclear Report, BCL~ Battelle Columbus Labs Report CR-75-269 ~ Effects Technology Capsule Report RPVDATAdeveloped by ATE Consulting
ATTACHMENT TO AEP:NRC:1173P Table 2 Summary ofProperties forWeld Wire Heat No. 13253/12008 Compilation ofMeasured ChemistriesPom allData Sources Heat No.
Flux Yyyc Flux Lot 'ct. Cu Pct. Ni Source Reference 13253/12008 Linde 1092 3791 13253/12008 Linde 1092 3774 13253/12008 Linde 1092 3714 13253/12008 Linde 1092 0.21 0.86 CE,WQ CE,WQ CE,WQ Wylie Labs C-E 78-12 RSP C-E 78-12 RSP C-E 78-12 RSP
'Avg. Cu =0.21 Avg. Ni~0.86 Note:
WQ Weld Qualification, CE = Combustion Engineering
~ ~
RPVDATAdeveloped by ATIConsulting
ATTACHMENT TO AEP:.NRC-l>3F Table 3 Sun1mary ofProperties for Weld Wire Heat No. 13253 Compilation ofMeasured ChemistriesPom allData Sources Heat No.
Flux Type Flux Lot Pct. Cu Pct. Ni Source Reference 13253 13253 13253 13253 13253 13253 13253 13253 13253 13253 Linde 1092 3774 Linde 1092 3833 Linde 1092 3724 Linde 1092 3791 Linde 1092 3791 Linde 1092 3774 Linde 1092 3833 Linde 1092 3833 Linde 1092 3833 Linde 1092 3774 0.27 0.74 0.23 0.71 0.283 0.732 0.267 0.728 0.244 0.734 0.247 '.728 CE,WQ CE,WQ CE,WQ CE,WQ CK1,SC SA2,SC SA2,SC SA2,SC SA2,SC SA2,SC C-E 78-12 RSP C-E 78-12 RSP C-E 78-12 RSP C-E 78-12 RSP WCAP-8047 WCAP-8824 WCAP-13366 WCAP-13366 WCAP-13366 WCAP-11554 Avg. Cu ~0,257 Avg. Ni~0.729 Note:
WQ Weld Qualification, SC ~ Surveillance Capsule, CE Combustion Engineering CK1 = D. C. Cook 1, SA2 Salem 2, WCAP Westinghouse Report
~ ~
RPVDATAdeveloped by ATIConsulting
ATTACHMENT TO AEP:NRC.
73F Table 4 Summary ofProperties for Weld Wire Heat No. S3986 Compilation ofMeasured ChemistriesPom allData Sources Heat No.
Flux Type Flux Lot Pct. Cu Pct. Ni Source Rcfcrcnce S3986 S3986 S3986 S3986 S3986 S3986 S3986 S3986 S3986 S3986 Linde 124 Linde 124 Linde 124 Linde 124 Linde 124 Linde 124 Linde 124 Linde 124 Linde 124 Linde 124 0934 0934
- 0934 0934 0934 0934 0934 0934 0.055 0.96 BW1, SC 0.051 0.06 0.98 Bwl, SC 0.81 CB&ICTR ¹337 SR-BNP1-1005-001 SR-BNP1-1005-001 CP&LLetter 0.06 0.9 CB&ICTR ¹337C CP&LLetter 0.05 0.06 0.05 0.055 0.051 0.06 0.97 CK2,SC 0.93 TRO,SC 0.97 TRO,SC WCAP-8512 WCAP-8426 WCAP-10861 0.96 CB&ICTR PT¹200(S)
CP&LLetter 0.97 CB&ICTR PT¹200(T)
CP&LLetter 0.92 CB&ICTRPT200A CP&LLetter Avg. Cu ~0.055 Avg. Ni~0.937 Note:
SC Surveillance Capsule, CB&I Chicago Bridge &Iron, CTR~ Certified Test Report BW1 Brunswick 1, CK2 D. C. Cook 2, TRO Trojan, CP&L Carolina Power &LightCo.
SR-BNP 1-1005-001 = GE Nuclear Report, WCAP Westinghouse Report
~
~
RPVDATAdeveloped by ATIConsulting