View source for L-17-308, 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)

{{Adams
| number = ML17299A030
| issue date = 10/25/2017
| title = 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)
| author name = Bologna R
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000412
| license number = NPF-073
| contact person = 
| case reference number = L-17-308
| document type = Code Relief or Alternative, Letter type:L
| page count = 10
}}

=Text=
{{#Wiki_filter:FENOC '                                                                        Beaver Valley Power Station FirstEnergy Nucfear Operating Company P.O. Box 4 Shippingp ort, PA 15077 Richard D. Bologna 724-682-5234 Site Vice President Fax: 724-643-8069 Octobe r 25, 2017 L-17-308                                                          10 CFR 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 10 CFR 50.55a Request for Alternate Reactor Vessel Nozzle Flaw Depth Sizing Criteria (Request 2-TYP-4-RVSE-2)
In accordance with the provisions of 10 CFR 50.55a(z)(2), FirstEnergy Nuclear Operatin g Company (FENOC) hereby requests Nuclear Regulatory Commission (NRC) approva l of a proposed alternative to certain reactor vessel nozzle flaw depth sizing criteria for the Beaver Valley Power Station, Unit No. 2. The enclosure identifies the affected components, the applicable code requirements, and the description and basis of the proposed alternative.
FENOC requests approval of the proposed alternative by September 21, 2018 to support the fall 2018 refueling outage.
There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A.
Lentz, Manag er- Fleet Licensing, at (330) 315-6810.
Richard D. Bologna

==Enclosure:==

Beaver Valley Power Station, Unit No. 2, 10 CFR 50.55a Request 2-TYP-4-RVSE-2 cc:     NRC Region I Administrator NRC Resident Inspector NRC Project Manage r Director BRP/DEP Site BRP/DEP Representative

Enclosure L-17-308 Beaver Valley Power Station, Unit No. 2, 10 CFR 50.55a Request 2-TYP-4-RVSE-2 (8 Pages Follow)

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request No. 2-TYP-4-RVSE-2 Page 1 of 8 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 Code Class:                      Class 1 System:                          Reactor Coolant System (RCS)
In accordance with 10 CFR 50.55a(g)(6)(ii)(F)(1 ), the reactor vessel nozzle to safe-end dissimilar metal welds listed below are subject to examinations in accordance with American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-770-2, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds .... "
Component               Description of Weld                               Inspection Item 2RCS*REV21-N-24         Reactor Vessel Hot Leg Nozzle to Safe-End         A-2 2RCS*REV21-N-26         Reactor Vessel Hot Leg Nozzle to Safe-End         A-2 2RCS*REV21-N-28         Reactor Vessel Hot Leg Nozzle to Safe-End         A-2 2RCS*REV21-N-23         Reactor Vessel Cold Leg Nozzle to Safe-End        B 2RCS*REV21-N-25         Reactor Vessel Cold Leg Nozzle to Safe-End        B 2RCS*REV21-N-27         Reactor Vessel Cold Leq Nozzle to Safe-End        B As determined by FirstEnergy Nuclear Operating Company (FENOC), additional welds that may be volumetrically examined concurrent with the above-listed welds are the reactor vessel safe-end to pipe austenitic welds listed below. These welds are part of the risk-informed inservice inspection (ISi) program. The risk-informed ISi program is currently being revised for the fourth ISi interval in accordance with ASME Code Case N-716-1, "Alternative Classification and Examination Requirements, Section XI, Division 1." The components in the table below are therefore included in this request, and the inspection items are to be determined (TBD).
Component         Description of Weld                            Inspection Item 2RCS*001-F01      Reactor Vessel   Hot Leg Safe-End to Pipe     TBD 2RCS*004-F01      Reactor Vessel   Hot Leg Safe-End to Pipe TBD 2RCS*007-F01      Reactor Vessel   Hot Leg Safe-End to Pipe     TBD 2RCS*003-F04      Reactor Vessel   Cold Leg Safe-End to Pipe TBD 2RCS*006-F04      Reactor Vessel   Cold Leg Safe-End to Pipe TBD 2RCS*009-F04      Reactor Vessel   Cold Leg Safe-End to Pipe TBD

===2. Applicable Code Edition and Addenda===
ASME Code, Section XI, 2013 Edition with no Addenda.

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 2 of 8
: 3. Applicable Code Requirements Volumetric examinations of the identified welds must be conducted in accordance with Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems,"
Supplements 2, 10, and 14 of the ASME Code, Section XI, 2013 Edition.
Section 3.2(b) of ASME Code, Section XI, Appendix VIII, Supplement 2 states:
The RMS [root mean square] error of the flaw depths estimated by ultrasonics, as compared with the true depths, shall not exceed 0.125 in.
[inch] (3mm) [millimeters].
Section 3.3(c) of ASME Code, Section XI, Appendix VIII, Supplement 10 states:
Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3mm).
Section 3.3(d) of ASME Code, Section XI, Appendix VIII, Supplement 14 states:
Supplement 2 or Supplement 3 examination procedures, equipment, and personnel are qualified for depth-sizing when the flaw depths estimated by ultrasonics, as compared with the true depths, do not exceed 0.125 in.
(3mm) RMS, when they are combined with a successful Supplement 10 qualification.

===4. Reason for Request===
An Electric Power Research Institute (EPRI) letter dated August 28, 2017 regarding the vendor that will be performing the Beaver Valley Power Station, Unit No. 2 reactor pressure vessel nozzle examinations states:
Neither the procedure nor any individuals were successful at obtaining a 0.125" [inch] RMSE [root mean square error] as required to perform through wall sizing for ASME Section XI Appendix VIII Supplements 2 &
10 ....
And that:
        ... several vendors have attempted to obtain Performance Demonstration Initiative (POI) Supplements 2 & 10 qualifications from the inside surface of the inlet and outlet nozzles, which are connected to the Reactor Pressure Vessel (RPV).
Consequently, relief from the supplement 2, 10, and 14-specified 0.125 inch (in.) RMSE depth sizing accuracy requirements is necessary to support examinations when conducted remotely from the inside diameter of the RPV nozzles. The 0.125 inch RMSE depth sizing requirement is expected to be met when performing examinations from the outside diameter of the nozzles. However, performing examinations of the nozzles from the outside diameter would increase personnel total dose exposure by

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 3 of 8 approximately 0.5 REM. Furthermore, the ASME has issued two code cases (described below) that state examination procedures, equipment, and personnel are qualified if the RMSE does not exceed 0.250 inches for the RV nozzle piping thicknesses in this request.
Therefore, relief is being requested in accordance with 10 CFR 50.55a(z)(2) as a hardship without a compensating increase in quality and safety.
: 5. Proposed Alternative and Basis for Use In lieu of the 0.125 inch RMSE depth sizing accuracy requirement in ASME Code Section XI, Appendix VIII, supplements 2, 10, and 14, FENOC is proposing use of ASME Code Cases N-695-1, "Qualification Requirements for Dissimilar Metal Piping Welds, Section XI, Division 1," and N-696-1, "Qualification Requirements for Mandatory Appendix VIII Piping Examinations Conducted From the Inside Surface, Section XI, Division 1." FENOC proposes to use the depth-sizing criteria of ASME Code Case N-696-1 when examining the reactor vessel nozzle to safe-end dissimilar metal welds and reactor vessel safe-end to pipe austenitic welds. If only the reactor vessel nozzle to safe-end dissimilar metal welds are examined, FENOC proposes to use the depth-sizing acceptance criteria contained in ASME Code Case N-695-1. If flaws are detected, they will be dispositioned as identified in Section 5.3 of this relief request.
5.1 Proposed Depth Sizing Criteria FENOC proposes to use a vendor qualified for inner diameter detection and length sizing per ASME Section XI, Appendix VIII as applicable to the welds included in this request. Instead of the depth sizing requirements identified in Supplement 2, Supplement 10, and Supplement 14 of ASME Section XI Appendix VIII, depth-sizing qualification will meet the requirements of ASME Code Cases N-695-1 and N-696-1.
Paragraph 3.3(d) of Code Case N-695-1 states:
For qualifications from the inside-surface, examination procedures, equipment, and personnel are qualified for depth sizing if the RMS error of the flaw depth measurements, as compared to the true flaw depths, does not exceed 0.125 in. (3mm) for piping less than 2.1 in. (54 mm) in thickness, or 0.250 in. (6mm) for piping 2.1 in. (54 mm) or greater in thickness.
Paragraph 3.3(c) of Code Case N-696-1 states:
Supplement 2 examination procedures, equipment, and personnel are qualified for depth-sizing if the RMS error of the flaw depth measurements as compared to the true flaw depths, does not exceed 0.125 in. (3mm) for piping less than 2.1 in. (54 mm) in thickness, or 0.250 in. (6mm) for piping 2.1 in. (54 mm) or greater in thickness, when they are combined with a successful Supplement 10 qualification .
The FENOC inside diameter examination vendor has achieved the acceptance criteria

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 4 of 8 of 0.250 inch RMS in regards to through-wall sizing for both Code Cases N-695-1 and N-696-1.
5.2 Description of Volumetric Examination Figures 1 and 2 provide representative sketches of the reactor vessel outlet nozzle to safe-end weld and the reactor vessel inlet nozzle to safe-end weld.
The nominal diameter and thickness dimensions for the reactor vessel outlet and inlet nozzle to safe-end welds, as well as the safe-end to pipe welds, are provided in Table 1.
Table 1. Nominal Diameter and Wall Thickness Dimensions Weld Description      Nominal Inside      Nominal Outside           Nominal Diameter (in.)       Diameter (in.)      Thickness (in.)
Reactor Vessel              28.97                 34.22                2.63 Outlet Nozzle to Safe-End Reactor Vessel              29.20                 34.22               2.51 Outlet Safe-End to Pipe Reactor Vessel              27.47                32.47                2.50 Inlet Nozzle to Safe-End Reactor Vessel             27.70                 32.47                2.39 Inlet Safe-End to Pipe During the nozzle weld examination process, inside diameter surface profile data is recorded using an immersion ultrasonic examination (UT) process. Contour wedges that match the inside diameter contour are utilized during the examination. The display is interactive such that the actual UT data (key flaw characterization information such as flaw depth and flaw location) can be compensated for by using the surface profile directly underneath the transducer. The profilometry software is able to calculate the areas where the water path under the transducer is greater than 1/32 inch, and this information is used to calculate examination volume coverage where detection scans are limited. The 1/32 inch value is generally considered for OD examinations but offers a conservative reference for the inside diameter examination. Based on exams performed in 2008, 2012, and 2014, the surface conditions of the nozzle to safe-end dissimilar metal welds are relatively smooth , and no lack of coverage is expected.
There are no limitations for axial scans for circumferential flaws, but the same profilometry software discussed earlier is used to detect areas where limitations for UT exist during the circumferential scans for axial flaws.
FENOC's examination vendor has participated in three non-ASME Code required performance demonstrations associated with depth sizing of planar flaws in dissimilar metal welds. The demonstrated techniques were conducted from the inside diameter

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 5 of 8 surface. The demonstrations used UT test procedures and equipment similar to those that will be applied for the Beaver Valley Power Station, Unit No. 2 weld examinations.
Summary information on these demonstrations was provided in a FENOC letter dated December 27, 2011 (Accession No. ML113620646).
5.3 Flaw Disposition If a flaw or flaws are detected and are measured as less than 50 percent through-wall in depth, they will be sized in accordance with the personnel, procedures, and equipment qualified to meet the requirements of ASME Code Cases N-695-1 and N-696-1.
If a flaw or flaws are detected and they are measured as greater than 50 percent through-wall depth, and the flaws will be left in service without mitigation or repair, flaw evaluations will be submitted to the NRC for review and approval prior to reactor startup. In addition to the information normally contained in flaw evaluations, this evaluation will include:
: 1. Information concerning the mechanism that caused the flaw.
: 2. Information concerning the inside surface roughness/profile in the region of the inspection.
: 3. Information concerning areas in which the UT probe may lift off from the surface of the pipe/weld.
In conclusion, the proposed alternative of ASME Code Cases N-695-1 and N-696-1 to ASME Code, Section XI, Appendix VIII, supplements 2, 10, and 14 depth-sizing requirements assures that the subject welds would be fully examined by vendor procedures, equipment, and personnel qualified by demonstration. FENOC has determined that the proposed alternative provides reasonable assurance of the structural integrity and leak-tightness of the subject welds, and that complying with the ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
: 6. Duration of Proposed Alternative The proposed alternative is requested for use during the fourth 10-year inservice inspection interval, which will begin on August 29, 2018, and is currently scheduled to end on August 28, 2028.
: 7. Precedent A similar request was approved for Salem Nuclear Generating Station, Unit Nos. 1 and 2 for their fourth 10-year inservice inspection interval. The FE NOC request for BVPS-2 also proposes to utilize ASME Code Cases N-695-1 and N-696-1 as an alternative to ASME Section XI, Appendix VIII requirements for RPV nozzle examinations from the inside diameter. The Nuclear Regulatory Commission staff letter authorizing the alternative is referenced on the next page:

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 6 of 8 Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Safety Evaluation of Relief Request SC-14R-171 Regarding the Fourth 10-Year Interval of the lnservice Inspection Program (CAC NOS. MF9348 and MF9349), dated August 17, 2017 (Accession No. ML17219A186).

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 7 of 8 Alloy 82/182 Weld Stainless Steel Weld 2.625                           Alloy 182 Buttering Carbon Steel Nozzle l
Stainless Steel t
(~28.97)
Safe-End t           ( 0.22 }
Stainless                                    Stainless Steel Pipe                                   Steel Cladding Figure 1: BVPS-2 Reactor Vessel Outlet Nozzle (dimensions are in inches)

Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RVSE-2 Page 8 of 8 Alloy 82/182 Weld Stainless Steel Weld                                               Alloy 182 Buttering Carbon Steel
{(2I32.47)                                     Nozzle
         !                                               f
{027.47)            l Stainless Steel Safe-End                                  ( 0.22)
Stainless                                       Stainless Steel Pipe                                      Steel Cladding Figure 2: BVPS-2 Reactor Vessel Inlet Nozzle (dimensions are in inches)}}