ML22237A312
| ML22237A312 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 08/25/2022 |
| From: | Ameren Missouri, Union Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22237A310 | List: |
| References | |
| ULNRC-06762 | |
| Download: ML22237A312 (8) | |
Text
Enclosure to ULNRC-06762 August 25, 2022 ENCLOSURE OAR-I FORMS (2)
Page 1 of 8
E170.O1 00 CALLAWAY NUCLEAR PLANT FORM OAR-I OWNERS ACTIVITY REPORT Report Number 25 Plant Callawav Enemy Center, 8315 County Road 459, Steedman, MO 65077 Unit No. 1 Commercial service date 12/19/1984 Refueling outage no. 25 (if applicable)
Current inspection interval 4thInteal (IS1 Program Plan) fist, 2nd, 3rd, 4th, other) 3rd Current inspection period Period (ISI Proaram Plan)
(1st, 2nd, 3rd)
Edition and Addenda of Section Xl applicable to the inspection plans 2007 Edition with 2008 Addenda Date and revision of inspection plans SI Program Plan: 7-December-2020, Rev. 002 Edition and Addenda of Section Xl applicable to repair/replacement activities, if differentthan the inspection plans Same as inspection plan.
Code Cases used for inspection and evaluation: N-532-5, N-661-2 and N-71 6-1 (if applicable, including cases modified by Case N-532 and later revisions)
CERTIFICATE OF CONFORMANCE I certify that(a) the statements made in this report are correct; (b)the examinations and tests meet the Inspection Plan as requited by the ASME Code, Section Xl; and (c) the repair/replacement activities and evaluations supporting the completion of 25 (refueling outage number) conform to the requirements of Section XI.
/_ I, 133511 Signed _e S S l5JIVueldmg En Date 25 August 2022
, Owner or Owne?s Desn Tide CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and employed by Hartford Steam Boiler Inspection and Insurance Co. of Hartford, Connecticut have inspected the items described in this Owners Activity Report, and state that, to the best of my knowledge and belief, the Owner has performed all activities represented by this report in accordance with the requirements of Section XI.
By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the repair/replacement activities and evaluation described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising from or connected with this ins pection.
Commission NB 14725 Endorsements: C, I. N, R In (National Board Number and Endorsement)
Date 08/25/2022 (06/11)
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CALLAWAY NUCLEAR PLANT TABLE I ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Item Description Evaluation Description Category and Item Number NO new Recordable-Indications or flaws were found during the Reactor Vessel Examination. All flaws and/or Recordable Indications listed below can be demonstrated to have been identified in previous Reactor Vessel inspections. Phased-Array Ultrasonic techniques, changes in recording methodology and procedures have allowed for better identification, resolution and measuring of the flaws and RIs.
The ISI Reactor Vessel 10-Year examination found the following 5 All indications were evaluated as acceptable. The flaws Recordable Indications (RI): were evaluated using ASME Code table IWB-3514-2 B-MB1 .30 . 2-RY-1O1-121 (Flange to Vessel Weld) 1 RI; - alt (%) measured versus alt (%) acceptable as listed:
B-A/B1.12 . 2-RV-1Ol-122-B (Long Shell Weld @ 207°) 2 RI; -
B-A/B 1.12 . 2-RV-1O1-122-C (Long Shell Weld @ 326°) 1 RI; and
. 2-RV-101-121 1.5% v 4.4%;
B-D/B3.90 . 2-RV-1O5-U1-C (Inlet Nozzle F to RV Weld @ 247°) 1 RI. -
. 2-RV-101-12 2-B(1) 0.8% v 3.3%;
2-RV-101-122-B (2) 0.7% v 2.5%;
All Recordable indications (K(s) were compared with the 2004J1na1 report and were previously 2-RV-101-122-C 1.4% v 3.3%;
known Ms. No new Ri werejound, but resolution and technology altowedfor better identification, resolution and measuring ofthe Ris. In all instancesJwerJlaw indications have been recorded
. 2-RV-105-121-C 0.7% v 2.2%.
during the present examination (2022, RF25). This does not mean that the vesselfiaw population The previous JO-year examination results were reviewed prior to is changing as the indications are considered to be wetdingfabrication indications and have been examination and areas containing previously recorded indications there since manufacture orfabrication. It should be recognized that the examination performed were investigated during this recent examination.
in 2004 was different in technology, recording methodology, and procedure requirements. The main difference between the number ofindications recorded in the previous examination and the current examination is a difference in recording thresholds and use ofPhase-Array Ultra Sonic (PAUT) techniques.
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Examination Item Description Evaluation Description Category and Item Number The results of the Ultrasonic (UT) examinations performed during R25 revealed
- Embedded flaws are considered fabrication flaws. These 3 detectable inside surface connected planer flaws and 2 embedded indications two flaws were not recorded in the prior examination (2017) in the dissimilar welds listed below: and are included in this report for ftiture reference and both are evaluated as acceptable to ASME Code table IWB R-AfR1.20
- 2-BB-01-F202 (27-/2 Elbow to Loop #2 RPV Inlet Safe-End) - 2 3514-2 using a/t (%) measured versus a/t (%) acceptable:
Embedded. .
2-BB-0l-F202 (1) 2.7% v 9.6%;
R-A/R1.20
- 2-BB-01-F302 (27W Elbow to Loop #3 RPV Inlet Safe- End) 3 detectable
-
- 2-BB-0l-f202 (2) 2.7% V 9.7%.
inside surface.
- The inner surface was interrogated looking for service induced SCC using a qualified ASME Code,Section XI, Appendix VIII automated phased array procedure.
- Eddy current was used in the 2742 elbow to ioop #3 RPV inlet safe-end (inlet nozzle azimuth 247°) similar metal weld to corroborate the ultrasonic observation of ID connection.
- The Three ID surface connected indications detected in 27-y2,, elbow to loop #3 RPV inlet safe-end (inlet nozzle azimuth 247°) were evaluated as NOT acceptable to ASME Code table IWB-3514-2 by these examination techniques and required additional analysis.
- A review of previous inspection reports indicates these flaws were previously shown to be present at this particular location based on the 2004, 2007, 2013, and 2017 inspections. There has been no growth in the flaws over time, and it can be concluded that these flaws were there since the initial welding ofthe safe-end to the cold leg elbow with a stainless-steel weld. Thus, the flaws are acceptable as fabrication flaws and not service induced flaws.
- The ASME Section XI 1WB-3600 evaluation performed in this report was based on plant-specific loading, geometry, and material properties. The maximum allowable end-of-evaluation flaw sizes were calculated per the Appendix C methodology of the ASME Code. New a/t (%) measured versus a/t (%) acceptable:
- 2-BB-0l-F302 (1) 12.5% v 39%;
2-BB-0l-f302 (2) 24.4% v 73%;
- 2-BB-0I-F302 (3) 14.0% v 30%.
Page4 of 8
Examination Item Description Evaluation Description Category and Item Number N/A The results of the ultrasonic examination revealed two recordable indications Indication delta length and width measurements in the cladding of the Reactor Vessel: between the current examination technique and past examination campaigns for both clad patch locations are CladBarePatch@ 185 . .
within the accuracy ofthe current technique s scan sync
. Clad Bare Patch @ 303 interval of 0.2, thus indicating an unchanging The intent ofthe examination was to determtne f the two known clad indications were exhthzttng
. . . .. . condition.
any detectable changes to the exposed carbon steel (i.e., base material) In addition, the clad layer and clad to base material interface was to be examined for cracktng and cits-bonding Comparison of the depth measurements for clad patch respectively.
location 303° are within the calibration tolerance of the current technique, which would indicate a static condition.
for the clad patch at 187°, the delta between the current technique and past examination campaigns is greater than the calibration tolerance however it indicates a shallower depth than the previous measurement (2017).
This more than likely due to the origination point being provided from the thinnest area of the remaining clad and the slight concavity that surrounds the area.
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TABLE 2 ABSTRACT OF REPAIR/REPLACEMENT ACTIVITIES REQUIRED FOR CONTINUED SERVICE Code Class Item Description Description of Work Date Completed Repair/Replacement Plan Number Pipe Hanger/Support Removed, Clean, Inspected and Re-installed struts of 2-May-2022 14510153.351 pipe Hanger/Support BB-04-C004/241 . 14510153.371 The traceability of the replace pipe Hanger/Support nut and bolt could not be fully verified due to the inability to find RO 0249843. CR 202205301 documents the non-conforming condition. The replacement nut and bolt were removed from a Clamp Assembly (MEN: 6327335). The manufacture of the Clamp AssembLy, Bergen-Power Pipe Supports, provided a Certificate of Compliance based on the P0 581776 and manufacturers part number 2600-7.
The Certificate of Compliance provides reasonable assurance that the replacement nut and bolt have the proper pedigree and will provide adequately support to the design function of the support. The degraded condition will be resolve by the regenerated traceability documentation resulting from actions developed in CR 202205301.
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E170O100 CALLAWAY NUCLEAR PLANT FORM OAR-I OWNERS ACTMTY REPORT ReportNumber ... .. ... ....,, ..,., V V .. V VVVVVVVVVVVVVVVV V V VVVVVVV 25 VVV V VVVVVVVVVVVVVVVVVVVVVVVVVVVVV V
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VVVVVVVVV VVVVVVV VVV VVVV VVVVVV Date and revision of inspection plans VV V.. V.. V VV VVVVV VIWEVProflram: 1O/1O/2O19Ey4jVVVV VVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVVV VVVVVVV VVVVVVVVVV VVVVV VVVV VVVVVVVVVVVV IWL Proiram: 07/23/201 8. Rev. 4 Edition and Addenda of Section XI applicable to repair/replacement activities, fdifferentthan the inspection plans Jnctionn Code Cases used for inspection and evaluation: ,
VVVVVVVVVVV VVVV an d VN 65 VVVVVV VV VVVVVVV VVVV (if applicable, including cases modified by case N-532 and later revisions)
CERTWICATE OF CONFORMANCE certify that (a) the statements made in this report are correct; (b) the examinations and tests meet the Inspection Plan as required by the ASME Code, Section Xl; and (c) the repair/replacement activities and evaluations supporting the completion of 25 frefuehng outage number) conform to the requirements of Section XI. tq73o i/l A ifI------,
Signed Date OvLét or Ownk Designee, Title CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and employed by Hartford Steam 3oilerlnspectzon and Insurance Co of Hartford, Connecticut have inspected the items described in this Owners Activity Report, and state that, to the best of my knowledge and belief, the Owner has performed all activities represented by this report in accordance withthe requirements of Section Xl.
By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the repair/replacement activities and evaluation described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising from or connected with this inspection.
Commission 1B 14725 Endorsements: C, I, N, R V 1__, (National Board Number and Endorsement)
Date 08/25/2022 (naf11t Page 7 of 8
CALLAWAY NUCLEAR PLANT TABLE 1 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Item Description Evaluation Description Category and Item Number L-B, L2.20 Tendon Wire CR 202103224 Containment Post Tension system tendon V39 wire did not meet the ultimate tensile strength acceptance criteria during the lab testing of the removed tendon wire. The lab test average was found to be 235.3 ksi whereas the ultimate tensile strength acceptance criteria is a minimum of 240 ksi.
Reviews of the original containment design parameters, the containment tendon system design basis, and the containment response to a postulated severe accident has demonstrated there is significant margin in the original design. The original specification for the tendon wire required a yield stress value of 192 ksi which results in a containment pressure capacity of 152 psig. The average yield for the wire samples that failed the ultimate strength requirements were 21 7.97 ksi. Inryco designed the tendon system using an average wire stress of I 56.9 ksi. Based on the final number of tendons installed, Bechtel calculated an actual wire minimum prestress of 127.16 ksi at design pressure. Bechtel also calculated a maximum containment pressure capacity of 152 psig vs.
a design pressure of 60 psig. The minimum tested yield for all 6 samples was 21 3.2 ksi. This is 1 1 % greater than the design basis yield stress of 192 ksi. The ultimate strength of the wire is not used as a design basis for the containment building tendons. Maximum design stress of 192 ksi is significantly less than the ultimate strength results from the sample wire, indicating that the tendon will not fail prior to the maximum design stress. The lift-off forces for all surveyed tendons were within or above the projected acceptance band at the 35th year surveillance point. This verifies that each of the tendons is operable by meeting its overall design function.
TABLE 2 ABSTRACT OF REPAIR/REPLACEMENT ACTIVITIES REQUIRED FOR CONTINUED SERVICE Code Class Item Description Description of Work Date Completed Repair/Replacement Plan Number None None None None None Page 8 of 8