ML19051A108
| ML19051A108 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 02/20/2019 |
| From: | Bologna R FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19051A107 | List: |
| References | |
| L-19-031 | |
| Download: ML19051A108 (10) | |
Text
WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 WHEN SEPARATED FROM ENCLOSURE D, THIS DOCUMENT CAN BE DECONTROLLED FENOC FirstEnergy Nuclear Operating Company Richard D. Bologna Site Vice President Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 724-682-5234 Fax: 724-643-8069 February 20, 2019 L-19-031 1 0CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Request Alternative Examination Frequency For Reactor Vessel Nozzle to Safe-End Welds (Request 2-TYP-4-RV-05, Revision 0)
In accordance with the provisions of 10 CFR 50.55a(z)(2), FirstEnergy Nuclear Operating Company (FENOC) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative examination frequency for Beaver Valley Power Station, Unit No. 2 reactor vessel nozzle to safe-end welds. Enclosure A identifies the affected components, the applicable code requirements, and the description and basis of the proposed alternative.
The currently scheduled examinations are to be performed during the spring of 2020 refueling outage. FENOC requests approval of the proposed alternative by February 28, 2020 (prior to the currently scheduled examinations) to permit implementation of the proposed alternative examination frequency.
The document L TR-SDA-18-054, Revision 0, entitled Technical Justification to Support the Extended Volumetric Examination Interval for Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds," was used as a basis for this request.
This document contains information proprietary to Westinghouse Electric Company LLC
("Westinghouse"). A non-proprietary version of this document is provided as Enclosure B.
An application for withholding proprietary information from public disclosure (letter CAW-19-4856), which includes an accompanying affidavit, proprietary information notice, and copyright notice is provided as Enclosure C. The affidavit signed by Westinghouse, the owner of the information, sets forth the basis on which the information contained in the proprietary version of document L TR-SDA-18-054 (Enclosure D) may be withheld from public disclosure by the NRC and addresses with
Beaver Valley Power Station, Unit No. 2 L-19-031 Page 2 specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the NRC's regulations.
Accordingly, it is respectfully requested that the information that is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the NRC's regulations.
Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-19-4856 and should be addressed to Camille T. Zozula, Manager, Infrastructure and Facilities Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2, Suite 259, Cranberry Township, Pennsylvania 16066.
There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Acting Manager Nuclear Licensing and Regulatory Affairs, at 330-315-6808.
Enclosures:
A.
Beaver Valley Power Station, Unit No. 2, 10 CFR 50.55a Request 2-TYP-4-RV-05, Revision 0 B.
LTR-SDA-18-054-NP, Revision 0, "Technical Justification to Support the Extended Volumetric Examination Interval for Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" [Non-proprietary]
C. Application For Withholding Proprietary Information From Public Disclosure D.
LTR-SDA-18-054-P, Revision 0, "Technical Justification to Support the Extended Volumetric Examination Interval for Beaver Valley Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" [Proprietary]
cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
ENCLOSURE A L-19-031 Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-05, Revision 0
[7 Pages Follow]
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-05, Revision 0 Page 1 of 7 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)
- Hardship Without a Compensating Increase in Quality and Safety -
- 1. ASME Code Components Affected Component Numbers:
Reactor vessel cold leg nozzle-to-safe-end dissimilar metal (OM) welds: 2RCS-REV21-N-23, 2RCS-REV21-N-25, and 2RCS-REV21-N-27 Code Class:
Class 1 Examination Category:
Class 1 Pressurized Water Reactor (PWR) Pressure Retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds Containing Alloy 82/182 (ASME Code Case N-770-2, Table 1)
Inspection Item:
B
- 2. Applicable Code Edition And Addenda American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV)
Code (ASME Code),Section XI, 2013 Edition with no Addenda.
3. Applicable Code Requirements
ASME Code Case N-770-2 (Reference A), Table 1, "Examination Categories, Class 1 PWR Pressure Retaining Dissimilar Metal Piping and Vessel Nozzle Butt Welds Containing Alloy 82/182," Inspection Item B, identifies the examination requirements for unmitigated butt welds exposed to cold leg operating temperatures greater than or equal to 525 degrees Fahrenheit (°F) and less than 580°F. For this inspection item, a volumetric examination is required every second inspection period not to exceed 7 years.
- 4. Reason For Request For the reasons described in the following paragraphs, compliance with the ASME Code Case N-770-2 requirement to volumetrically examine the reactor vessel cold leg nozzle-to-safe-end welds every second inspection period, not to exceed 7 years, results in an unnecessary hardship without a compensating increase in the level of quality and safety.
Because of the geometry of the Beaver Valley Power Station, Unit No. 2 (BVPS-2) reactor vessel cold leg nozzle-to-safe-end welds, the limited access that exists around the outside diameter of the cold leg nozzle piping, and the significant radiological dose
2-TYP-4-RV-05, Revision 0 Page 2 of 7 that would be received if the volumetric examinations were performed from the outside diameter of the nozzles, volumetric examination of the reactor vessel cold leg nozzle-to
-safe-end welds is performed from the inside diameter of the nozzles. To gain access to the inside diameter of the nozzles, it is necessary to remove the core barrel from the reactor vessel. Removal of the core barrel is a critical lift because of the tight clearances between the structure and the reactor vessel welded attachments, the weight of the core barrel, and the high dose rates that are present during the core barrel lifts.
In addition to the personnel safety hazards normally associated with critical lifts, critical lifts have the potential to result in damage to the assembly. The BVPS-2 core barrel weighs approximately 200,000 pounds. The bottom of the core barrel is secured within the reactor vessel by clevis inserts and core support lugs that interface with radial keys on the outside diameter of the core barrel. Because of the close tolerance fit between the clevis inserts, core support lugs, and the radial keys on the core barrel, precision lifts are required to remove and replace the core barrel. Every core barrel lift presents risk in that the reactor pressure vessel and reactor pressure vessel internals may be damaged during the lift.
Movement of the core barrel presents significant radiological risk. During core barrel lifts, strict radiological controls are observed to keep dose as low as reasonably achievable. Shielding is provided for the polar crane operator, and tag line holders remain behind concrete shielding to minimize dose. Remote cameras are utilized to allow station personnel to remotely direct the lift and align the core barrel prior to placement in the storage stand. During the lift, unrelated work is stopped and the only personnel allowed on the refueling deck are those personnel that are required to complete the lift.
Due to the BVPS-2 refueling cavity design, the highly radioactive core barrel rises out of the water approximately seven feet when it is moved between the reactor vessel and the storage stand. During movement of the core barrel in 2014, high radiation levels (210 milli-rem per hour) were observed in the BVPS-2 containment building, creating a risk of associated radiation exposure to workers.
Improvements in shielding and the use of cameras helped reduce the dose received by station personnel during movement of the core barrel for the 2014 reactor vessel examinations to 15.4 milli-rem. Although lower dose has been achieved, significant radiological risk to personnel moving the core barrel continues to be present.
An extension of the volumetric examination interval required by ASME Code Case N-770-2 is requested because of the personnel safety hazards, risk of equipment damage, and radiological risks presented by core barrel removal.
5. Proposed Alternative and Basis for Use
A one-time inspection interval of 9 years between volumetric examinations is proposed for the identified BVPS-2 reactor vessel cold leg nozzle-to-safe-end welds instead of the
2-TYP-4-RV-05, Revision 0 Page 3 of 7 maximum 7 year interval identified for Inspection Item B in Table 1 of ASME Code Case N-770-2. The volumetric examinations of the BVPS-2 reactor vessel cold leg nozzle-to-safe-end welds were last performed in the spring of 2014, during the 2R17 maintenance and refueling outage. In accordance with Table 1, Inspection Item B of ASME Code Case N-770-2, the volumetric examinations are required to be performed again by the spring of 2021 (2R21 ). With the proposed alternative, the next volumetric examination of the BVPS-2 reactor vessel cold leg nozzle-to-safe-end welds would be performed during the 2R23 refueling outage, which is currently scheduled for the spring of 2023.
Electric Power Research Institute (EPRI) Materials Reliability Program 2012 technical report: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349, Reference B) provides a technical basis for extending the re-examination interval of large diameter butt welds operating at cold leg temperatures from an interval of 7 years to an interval of 10 years. The basis for extending the re-examination interval for large diameter butt welds is that 1) there has been no service experience with primary water stress corrosion cracking (PWSCC) found in reactor vessel cold leg nozzle OM butt welds, 2) crack growth rates in reactor vessel cold leg nozzle OM butt welds are relatively small, and 3) the likelihood of cracking or through-wall leaks is very small in reactor vessel cold leg nozzle OM welds.
The MRP-349 re-examination interval extension is based on the resistance of Alloy 82/182 to cracking, as determined by crack growth rates in this material (that is, flaw tolerance). The flaw tolerance analyses performed to date have shown that the critical crack sizes in large-diameter butt welds operating at cold leg temperatures are very large. The analyses assume that a flaw has been initiated and use conservative inputs to determine the rate of crack growth. Assuming that a flaw initiates, the time required to grow to through-wall is in excess of 20 years in most cases analyzed. More recent analyses have been performed for the reactor vessel nozzles using through-wall residual stress distributions that were developed based on the most recent guidance.
These analyses have shown that the flaw tolerance of these locations is high and postulated circumferential flaws will not reach the maximum ASME allowable depth in less than 10 years. Specifically, Figure 5-4, "Circumferential Flaw PWSCC Crack Growth at the RV [Reactor Vessel] Inlet nozzle OM Welds," of MRP-349 shows that a flaw with a depth of 15 percent of the wall thickness would not grow to the maximum allowable ASME Code flaw size in less than 10 years of continued operation. This conclusion was not representative of a single plant but was determined for the most limiting wall thickness in the Westinghouse PWR fleet combined with the limiting piping loads from another plant in the Westinghouse PWR fleet. The inputs used to generate the results in Figure 5-4 of MRP-349 are judged to be conservative for BVPS-2. In particular, the BVPS-2 cold leg nozzle temperature may vary between 528.5°F and 543.1 °F, and is conservatively lower than the temperature of 565°F assumed in Figure 5-4.
The technical basis for the interval extension discussed in MRP-349 is also based on probabilistic analyses that are used to determine the likelihood of a flaw initiating and growing through-wall. Analyses have been performed to calculate the probability of failure for Alloy 82/182 welds using both probabilistic fracture mechanics and statistical
2-TYP-4-RV-05, Revision 0 Page 4 of 7 methods. Both approaches have shown that the likelihood of cracking or through-wall leaks in large-diameter cold leg welds is very small. Furthermore, sensitivity studies performed using probabilistic fracture mechanics have shown that even for the more limiting high temperature locations, more frequent inspections than required by ASME Code Section XI, such as that in EPRI Materials Reliability Program 2008 technical report: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139, Reference C) or ASME Code Case N-770-2, have only a small benefit in terms of risk.
Though past service experience may not be an absolute indicator of the likelihood of future cracking, it does give an indication of the relative likelihood of cracking in cold leg temperature locations versus hot leg temperature locations. While there is a significant amount of PWSCC service experience in hot leg locations, the number of indications in large-diameter butt welds is still small relative to the number of potential locations. Also, indications in hot leg locations have been detected before they were a safety concern.
Therefore, if hot leg PWSCC is a leading indicator for cold leg PWSCC, and the higher frequency of inspections will be maintained for the hot leg locations, it is reasonable to conclude that a moderately less rigorous inspection schedule would be capable of detecting any cold leg indications before they become large enough to be a concern.
Ultrasonic testing (UT) and eddy current testing (ECT) of the BVPS-2 reactor vessel hot leg nozzle-to-safe-end welds was performed during the fall 2018 refueling outage, and no recordable indications were observed. The absence of any recordable indications during the fall 2018 hot leg examinations provides added assurance that the one-time extension of the cold leg OM weld examinations by approximately 36 months provides an acceptable level of quality and safety. Also, the significantly higher operating temperature of the hot leg nozzles would be expected to initiate PWSCC flaws before flaws initiated in the cold leg nozzles.
During the spring 2014 BVPS-2 refueling outage, UT examinations were performed on the BVPS-2 reactor vessel cold leg nozzle-to-safe-end OM welds, and achieved 100 percent coverage. The UT examinations were performed in accordance with ASME Code Section XI, Appendix VIII, using performance demonstrated methods, with a minimum detectable flaw depth of 0.25 inch (10 percent of the wall thickness). No recordable indications were identified during the 2014 UT examinations. No recordable indications were observed during in-service UT examinations that were performed on the BVPS-2 reactor vessel cold leg nozzle-to-safe-end welds in 1996 and 2008.
The ECT method used for the 2014 examinations of BVPS-2 reactor vessel cold leg nozzle-to-safe-end OM welds was qualified in accordance with the same qualification procedure and practical trials used by the South Texas Project, Unit No. 2 (STP) and discussed in the June 30, 2016 NRC safety evaluation (Reference D). 100 percent coverage of the OM weld ID surface was obtained during the 2014 BVPS-2 reactor vessel nozzle ECT examinations. The OM welds are shop welds and have an essentially flat surface across the OM weld volume. The ECT inspection procedure used during the BVPS-2 spring 2014 examinations required that an indication with a minimum depth of 0.04 inch and a minimum length of 0.25 inch be recorded. No
2-TYP-4-RV-05, Revision 0 Page 5 of 7 recordable indications were observed during the 2014 ECT examinations of the reactor vessel cold leg nozzle-to-safe-end welds. Additionally, the ECT method used during the 2014 examinations was capable of detecting axial flaw lengths shorter than 0.25 inch.
The 2014 ECT data was re-evaluated by a Level Ill examiner, and it was determined that no indications were observed during the 2014 examinations with axial flaw length sizes larger than 0.16 inch.
A flaw tolerance evaluation for reactor vessel inlet nozzle OM welds described in MRP-349 showed that either a 25 or 50 percent inside surface weld repair performed during the initial weld fabrication process is a significant contributor in producing limiting PWSCC growth results.
A weld repair search was performed for the BVPS-2 reactor vessel cold leg nozzle alloy 82/182 welds. Based on the reactor vessel fabrication contract deviation notices, and construction radiographic testing (RT) and ultrasonic testing (UT) examination results, it was determined that no significant weld repairs were performed on these welds during original plant construction.
A plant-specific flaw tolerance evaluation (L TR-SOA-18-054-P, provided in Enclosure 0) was performed for the BVPS-2 reactor vessel cold leg nozzle-to-safe-end welds. The purpose of the plant-specific flaw tolerance evaluation was to determine the largest axial and circumferential flaw sizes that could remain in service and not grow to an unacceptable size prior to the next proposed inspection. The evaluation considered both the stress caused by pipe reaction loads and the residual stresses remaining from welding activities. ASME Code Section XI, paragraph IWB-3640 and Appendix C provided the evaluation guidelines and procedures for calculating the maximum allowable end-of-evaluation period flaw sizes. The plant-specific evaluation assumes a 360 degree inside surface weld repair with a repair depth of 50 percent through the OM weld thickness, consistent with the recommendations provided in EPRI Materials Reliability Program 2010 technical report: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287, Reference E).
The BVPS-2 flaw tolerance evaluation documented in L TR-SOA-18-054-P concluded that for a 9 effective full power year (EFPY) period, the largest allowable initial flaw sizes for a circumferential flaw were 0.8952 inch in depth and 8.9520 inches in length.
The largest allowable circumferential initial flaw sizes were then compared with the minimum flaw sizes that were detectable during the reactor vessel cold leg nozzle examinations performed during the spring 2014 refueling outage. Since the minimum detectable flaw depth during the 2014 UT exams was 0.25 inch and the largest allowable initial circumferential flaw depth calculated in the BVPS-2 flaw tolerance evaluation was 0.8952 inch, it was determined that there were no circumferential flaws present in the spring of 2014 that exceeded the largest allowable initial circumferential flaw size.
The BVPS-2 flaw tolerance evaluation contained in L TR-SOA-18-054 also concluded that for a 9 EFPY period, the largest allowable initial flaw sizes for an axial flaw were 0.1004 inch in depth and 0.2008 inch in length. The 2014 ECT minimum detectable flaw
2-TYP-4-RV-05, Revision 0 Page 6 of 7 depth was 0.04 inch and the 2014 ECT data was re-evaluated to determine that there were no indications present that were greater than 0.16 inch in length. It is therefore concluded that there were no axial flaws present in 2014 that exceeded the largest allowable initial axial flaw sizes that were determined in the BVPS-2 reactor vessel cold leg nozzle flaw tolerance evaluation.
Because no axial or circumferential flaws exceeding the largest allowable initial flaw sizes were present during the spring 2014 reactor vessel nozzle examinations, it is technically justified to extend the volumetric examination of the BVPS-2 reactor vessel cold leg nozzle-to-safe-end welds from the 2R21 refueling outage to the 2R23 refueling outage currently scheduled for the spring of 2023 (a period of 9 years between examinations).
6. Duration of Proposed Alternative
The duration of the proposed alternative is the fourth 10-year inservice inspection interval at BVPS-2 that began on August 29, 2018 and ends on August 28, 2028.
- 7. Precedent Relief from reactor vessel cold leg nozzle weld examination requirements (request RR-ENG-3-20) was previously approved by the NRC for the South Texas Project, Unit No. 2 as described in Reference D. Like request 2-TYP-4-RV-05 for BVPS-2, request RR-ENG-3-20 requested a one-time extension of the interval between volumetric examination of ASME Code Class 1 reactor vessel cold leg nozzle-to-safe-end dissimilar metal butt welds of approximately two operating cycles beyond the required examination interval of ASME Code Case N-770, Inspection Item B. The butt welds at South Texas Project, Unit No. 2 are made of alloys containing Alloy 82/182, and are exposed to the cold-leg temperature of 563 °F during normal plant operation. Unlike the South Texas Project, Unit No. 2, there is no plan to implement a stress improvement process for hot and cold leg nozzle welds at BVPS-2 due to the low probability of crack initiation.
- 8. References A. Code Case N-770-2, '~lternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of listed Mitigation ActivitiesSection XI, Division 1," approved June 9, 2011.
B. Electric Power Research Institute (EPRI) Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349), 2012 technical report.
C. EPRI Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139), 2008 technical report.
D. Letter from Robert J. Pascarelli (NRC) to Mr. G. T. Powell (South Texas Project Nuclear Operating Company), "South Texas Project, Unit 2 - Request for Relief No.
2-TYP-4-RV-05, Revision 0 Page 7 of 7 RR-ENG-3-20 For Extension of the Inspection Frequency of the Reactor Vessel Cold-Leg Nozzle to Safe End Welds With Flaw Analysis," dated June 30, 2016.
(Accession Number ML16174A091 ).
E. EPRI Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287), 2010 Technical Report.