L-11-321, CFR 50.55a Request for Alternative Examination Requirements for Reactor Vessel Safe-End Welds
| ML113620646 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley (NPF-073) |
| Issue date: | 12/27/2011 |
| From: | Harden P FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-11-321 | |
| Download: ML113620646 (14) | |
Text
FENOC FirstEnergy Nuclear Operating Company Paul A. Harden Site Vice President December 27, 2011 L-11-321 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 Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 724-682-5234 Fax: 724-643-8069 10 CFR 50.55a 10 CFR 50.55a Request for Alternative Examination R~guirements for Reactor Vessel Safe-End Welds In accordance with 10 CFR 50.55a, Nuclear Regulatory Commission (NRC) review and approval is requested for a proposed alternative to certain American Society of Mechanical Engineers (ASME) Code requirements associated with volumetric examinations of Beaver Valley Power Station Unit No.2 (BVPS-2) reactor vessel nozzle to safe-end dissimilar metal welds.
The affected components, the applicable ASME Code requirements, a description of the proposed alternative and basis for use is provided in the Enclosure. The alternative is proposed for use during the remainder of the current BVPS-2 10-year inservice inspection interval, which is scheduled to expire on August 28, 2018.
The proposed alternative is to be implemented prior to the fall 2012 BVPS-2 refueling outage. Therefore, FirstEnergy Nuclear Operating Company requests approval of the proposed alterative by August 31, 2012.
There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Supervisor - Fleet Licensing, at (330) 315-6808.
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Beaver Valley Power Station, Unit No.2 L-11-321 Page 2 of 2
Enclosure:
10 CFR 50.55a Request Number: 2-TYP-3-RVSE-1 cc:
NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request Number: 2-TYP-3-RVSE-1 Proposed Alternative in Accordance with 10 CFR SO.SSa(g)(S)(iii)
Page 1 of 12
--Inservice Inspection Impracticality--
1.0 ASME Code Components Affected
Code Class:
Class 1 System:
In accordance with 10 CFR 50.55a(g)(6)(ii)(F)(1), the reactor vessel nozzle to safe-end dissimilar metal (DSM) welds listed below are subject to volumetric examinations in accordance with American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-770-1, "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 Leg 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 program and are listed in the Examination Schedules as Examination Category R-A, Item R 1.11.
Component Description of Weld Inspection Item 2RCS*OO1-F01 Reactor Vessel Hot Leg Safe-End to Pipe R1.11 2RCS*OO4-F01 Reactor Vessel Hot Leg Safe-End to Pipe R1.11 2RCS*007 -F01 Reactor Vessel Hot Leg Safe-End to Pipe R1.11 2RCS*OO3-F04 Reactor Vessel Cold Leg Safe-End to Pipe R1.11 2RCS*OO6-F04 Reactor Vessel Cold Leg Safe-End to Pipe R1.11 2RCS*OO9-F04 Reactor Vessel Cold Leg Safe-End to Pipe R1.11
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 2 of 12 2.0
Applicable Code Edition and Addenda
Beaver Valley Power Station Unit No.2 (BVPS-2) In-Service Inspection and Repair/Replacement Programs:
BVPS-2 Ultrasonic (UT) Examination:
3.0
Applicable Code Requirement
ASME Code Section XI, 2001 Edition through 2003 Addenda ASME Code Section XI, 2001 Edition with no Addenda and Appendix VIII Supplements The volumetric examinations are to be conducted in accordance with Appendix VIII, "Performance Demonstration Initiative [POI] for Ultrasonic Examination System,"
Supplements 2 and 10 of the ASME Code Section XI, 2001 Edition with no Addenda.
Alternatives to Appendix VIII, Supplements 2 and 10, are ASME Code Cases N-695 (Supplement 10) and N-696 (combined Supplements 2 and 10). These Code Cases were approved for use by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Table 1, "Acceptable Section XI Code Cases."
Code Case N-695, "Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1," Paragraph 3.3(c) states, "Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS [root mean squared] error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)."
Code Case N-696, "Qualification Requirements for Appendix VIII Piping Examinations Conducted from the Inside Surface,Section XI, Division 1," Paragraph 3.3(d) states, "Supplement 2... 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."
4.0 Impracticality of Compliance An Electric Power Research Institute (EPRI) letter dated September 23, 2011 states, "To date, no vendor has been capable of meeting this criterion [an overall error less than or equal to a 0.125 inch RMS error (RMSE)]."
When examining from the inside diameter (10), the vendor attempts to meet the Supplement 1 0 (Code Case N-695) and combined Supplement 2 and 10 (Code Case N-696) required RMSE values for flaw depth sizing have been unsuccessful. Process enhancements including new delivery systems, new transducers, and software modifications have been implemented, but have not achieved the desired improvements
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 3 of 12 in performance. This result indicates the ASME Code acceptance requirement for flaw depth sizing is impractical for use with current 10 ultrasonic examination technology.
Additionally, the inability to achieve the ASME Code requirement for flaw depth sizing may be, in part, attributed to 10 interferences that include weld root and counterbore restrictions, or that the examination material consists of cast stainless steel.
5.0 Burden Caused by Compliance Compliance with the performance demonstration initiative qualification program without an alternative qualification requirement would necessitate significant modifications to the reactor coolant system welds. Alterations such as these may result in reduced structural integrity of the reactor coolant pressure boundary. Even with such modifications, the vendor depth sizing accuracy issue would not likely be fully addressed.
6.0 Proposed Alternative and Basis for Use As approved for use by the NRC in Regulatory Guide 1.147, FENOC proposes to use ASME Code Case N-696 to perform a combined Supplements 2 and 10 qualification when examining the reactor vessel nozzle to safe-end OSM welds and reactor vessel safe-end to pipe austenitic welds. If only the reactor vessel nozzle to safe-end OSM welds are examined, FENOC proposes to use ASME Code Case N-695 to perform a Supplement 10 qualification. However, FENOC proposes using alternative RMSE depth sizing criteria as compared to the values stated in Code Cases N-695 and N-696.
The FENOC 10 examination vendor has demonstrated the ability to depth size flaw indications in OSM welds with an RMSE of 0.189 inch instead of the 0.125 inch RMSE required by Appendix VIII Supplement 10 (Code Case N-695) and the RMSE of 0.245 inch for the combined Appendix VIII Supplements 2 and 10 qualification (Code Case N-696). The difference between the 0.245 inch RMSE and the Code Case N-696 required 0.125 inch RMSE would be added to the flaw depths determined during actual sizing of flaws. FENOC also proposes that if only the reactor vessel nozzle to safe-end OSM welds are examined, the difference between the 0.189 inch RMSE and the Code Case N-695 required 0.125 inch RMSE would be added to the flaw depths determined during actual sizing of flaws. Figures 1 and 2 provide representative sketches of a reactor vessel outlet nozzle to safe-end weld and an inlet nozzle to safe-end weld.
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 4 of 12 stainless Steel Weld 2.;625 Stainless Steel Pipe Stainless Steel Safe-End Alloy 82/182 Weld Alloy 182 Buttering Carbon Steel Nozzle
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Stainless Steel Cladding Figure 1: BVPS-2 Reactor Vessel Outlet Nozzle
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 5 of 12 Stainless Steel Weld (032.47 )
j Stainless Steel Pipe Stainless Steel Safe-End Alloy 82/182 Weld Carbon Stee I Nozzle t
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Alloy 182 Buttering i
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Stainless Steel Cladding Figure 2: BVPS-2 Reactor Vessel Inlet Nozzle
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 6 of 12 The nominal diameters 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 NominallD Nominal Outside Nominal Thickness Diameter (OD)
Reactor Vessel Outlet 28.97" 34.22" 2.63" Nozzle to Safe-End Reactor Vessel Outlet 29.20" 34.22" 2.51" Safe-End to Pipe Reactor Vessel Inlet 27.47" 32.47" 2.50" Nozzle to Safe-End Reactor Vessel Inlet 27.70" 32.47" 2.39" Safe-End to Pipe The BVPS-2 reactor vessel nozzle weld examinations would be completed using the I D applied PDI-qualified UT equipment, personnel, procedures and techniques qualified by the vendor. During the nozzle weld examination process, ID surface profile data would be recorded using an immersion UT process. Contour wedges that match the ID contour will be utilized when the examination is performed. This display is interactive with the actual UT data such that key flaw characterization information such as flaw depth sizing and flaw location can be compensated for by using the surface profile directly underneath the transducer. This profilometry software has the ability to calculate the areas where the water path under the transducer is greater than 1/32 inch; 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 ID examination. Based on BVPS-2 exams performed in 2008, the surface conditions of these nozzle to safe-end DSM welds are relatively smooth, and therefore, no lack of coverage would be expected. However, the surface geometry of the safe-end to pipe welds include weld root, counterbores, and fit-up variations, which result in loss of contact between the transducer and the surface. In the instances where UT volumetric examinations are limited by surface geometry, eddy current examinations would be used to characterize any surface-breaking flaws for both nozzle to safe-end DSM welds and safe-end to pipe welds. This ensures that 100 percent of the weld surfaces have been examined. A similar approach was utilized in the 2008 UT volumetric examinations.
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 DSM welds.
The demonstrated techniques were conducted from the ID surface. Each of these demonstrations used UT test procedures and equipment similar to those to be applied for the BVPS-2 weld examinations. Summary information on these demonstrations is provided below.
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 7 of 12 The first demonstration (non-blind) was conducted in 2002/2003 as part of an open procedure qualification for a Swedish nuclear power plant and was performed under the auspices of the Swedish Qualification Centre. The test sample, supplied by the power plant owner, consisted of six partial ring segments that when put together formed a 360 degree test piece. The inner diameter was 23 inches and the weld thickness was 3.3 inches. The materials of construction included a stainless steel clad SA508 Class 1A ferritic steel forging buttered with Inconel' and welded to a SA312 Type 316 stainless steel forging. The weld material was Inconel'. The 10 surface was smooth. Of the 25 defects within the segments, 10 were 10 surface-connected branch cracks confined to the weld and buttering. Details of these 10 surface flaws are provided in Table 2; this defect matrix includes both circumferential and axial cracks. Table 2 also provides the UT measured through-wall dimension for each of the defects and the RMSE value. It is noted that if the RMSE adjustment of 0.064 inches were added to each of the UT measurements, the adjusted RMSE value is 0.133 inches.
The second demonstration (non-blind) was conducted in 2007 as part of a weld inlay equivalency site specific demonstration and was performed under the auspices of EPRI.
This test sample was a full scale, 360 degree mock-up and is approximately 41 inches long, 27.5 inches 10 and 2.9 inches thick. It contained a DSM weld (Inconel' buttering and Inconel' weld metal) between a SA-508 ferritic steel forging (nozzle) and a 316 stainless steel forging (safe-end). An alloy 52 weld inlay was added to three of the four quadrants across the DSM weld. The inlay thickness for each of the three quadrants was 0.2 inch, 0.07 inch and 1 inch for the second, third and fourth quadrants, respectively. Four alternative planar flaws were added to each of the three inlaid quadrants. These four flaws are essentially identical to flaws in the POI 601 Series Practice Mockup. The 10 surface included a 10- 15 degree taper on the safe-end.
Details of the flaws are provided in Table 3; this defect matrix includes both circumferential and axial cracks. Table 3 also provides the UT measured through-wall dimension for each of the sixteen defects and the RMSE value. If the RMSE adjustment of 0.064 inches was added to each of the UT measurements, the adjusted value is 0.083 inches.
The third demonstration (blind) was conducted in 2010 as the initial test in a series of round robin examinations on RPV nozzle to safe-end welds. Examinations were conducted blind on six individual test coupons. Each test coupon consisted of a DSM Alloy 82/182 buttering weld between a ferritic steel forging and a stainless steel safe-end, and contained a single axial stress corrosion crack. Each test coupon was approximately 29 inches in diameter and was approximately 2.9 inches thick. The 10 surface was smooth. Each test coupon was destructively analyzed to determine the actual crack depth. Information on the flaws is noted in Table 4; this defect matrix includes only axial cracks. Table 4 also provides the UT measured through-wall dimension for each of the six defects and the RMSE value. If the RMSE adjustment of 0.064 inches was added to each of the UT measurements, the adjusted RMSE value is 0.125 inches.
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 8 of 12 Each of these test samples included variables common to DSM welds - multiple materials with different acoustic properties, and dendritic and coarse-grained microstructures. These two variables lead to inaccuracies in locating the UT response from a planar flaw extremity within the weld and buttering. Demonstrations 1 and 3 did not involve additional UT examinations to determine the 10 surface profile, which is used to compensate for beam propagation in the weld. Also different in each of these test samples is the type of planar flaw, the flaw morphology, and the actual dendritic structure of the weld and buttering. These differences between the demonstrations are factors in the variation in RMSE values. When the three data sets from Tables 2, 3 and 4 are combined, the RMSE is 0.092 inches. These data sets do not include the RMSE adjustment of 0.064 inches.
As evidenced by the combined RMSE for dissimilar and similar metal welds demonstrated in accordance with Code Case N-696, UT techniques can determine the through-wall size of planar flaws. The test samples consisted of surface conditions representative of shop and field weld configurations. While the RMSE value is not consistent with the established ASME Code standard, it still represents a reasonable measurement of flaw through-wall depth, albeit with a greater uncertainty. However, this uncertainty is to be factored into the flaw size which would be used in fracture mechanics analyses in accordance with the proposed alternative.
The BVPS-2 UT vendor has no future Appendix VIII, Supplement 2 and 10 or non-Appendix VIII programs involving 10 performance demonstrations scheduled.
Table 2: Defect Matrix for Open Procedure Qualification -ID Surface Connected Planar Flaws Only Flaw Description Orientation Ligament Truth Truth Through-UT - Measured (inches)
Length Wall Dimension through-Wall Dimension (inches)
Dimension (inches)
(inches)
Flaw 1.1 - Branch Circ / 9 0 skew /
0 0.70 0.24 0.34 Crack (in weld) 10 0 tilt Flaw 1.2 - Branch Circ / 0 0 skew /
0 0.87 0.36 0.44 Crack (in buttering) 9 0 tilt Flaw 1.3 - Branch Circ /1 0 skew /
0 1.19 0.48 0.38 Crack (in buttering) 0 0 tilt Flaw 1.4 - Branch Circ / 0 0 skew /
0 2.76 1.02 1.09 Crack (in buttering) 2 0 tilt Flaw 1.5 - Branch Circ / 10 0 skew /
0 2.78 1.33 1.42 Crack (in weld) 10 0 tilt
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 9 of 12 Table 2 (Continued): Defect Matrix for Open Procedure Qualification - ID Surface Connected Planar Flaws Flaw Description Flaw 2.1 - Branch Crack (in buttering/weld)
Flaw 2.2 - Branch Crack (in buttering)
Flaw 2.3 - Branch Crack (in weld)
Flaw 2.4 - Branch Crack (in buttering/weld)
Flaw 2.5 - Branch Crack (in buttering)
Abbreviated Terms Circ - Circumferential Ax -Axial Onl Orientation Ligament (inches)
Ax /10 0 skew /
0 8
0 tilt Ax / 8 0 skew / 0 0
0 tilt Ax / 11 ° skew /
0 10 0 tilt Circ / 8 0 skew /
0 0
0 tilt Circ / 0 0 skew /
0 9.5 0 tilt Truth Truth Through-UT-Measured Length Wall Dimension through-Wall Dimension (inches)
Dimension (inches)
(inches) 0.68 0.25 0.28 0.49 0.25 0.22 0.67 0.33 0.38 0.83 0.35 0.47 1.72 0.71 0.82 RMSE 0.084 RMSE with Vendor Tolerance Adjustment 0.133 Table 3: Defect Matrix for Open Weld Inlay Equivalency Test Demonstration -ID Surface Connected Planar Flaws Only Flaw Description Orientation Ligament Truth Truth Through-UT-Measured (inches)
Length Wall Dimension through-Wall Dimension (inches)
Dimension (inches)
(inches)
Flaw 1-Q2 (Thermal Circ / 8 0 skew /
0 1.80 0.339 0.37 Fatigue Crack) 0 0 tilt Flaw 1-Q3 (Thermal Circ / 8 0 skew /
0 1.80 0.339 0.42 Fatigue Crack) 0 0 tilt Flaw 1-Q4 (Thermal Circ / 8 0 skew /
0 1.80 0.339 0.36 Fatigue Crack) 0 0 tilt Flaw 2-Q2 (Thermal Circ / 0 0 skew /
0 2.63 0.35 0.34 Fatigue Crack) 0 0 tilt Flaw 2-Q3 (Thermal Circ / 0 0 skew /
0 2.63 0.35 0.33 Fatigue Crack) 0 0 tilt Flaw 2-Q4 (Thermal Ax / 11 ° skew /
0 2.63 0.35 0.37 Fatigue Crack) 0 0 tilt
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 10 of 12 Table 3 (Continued): Defect Matrix for Open Weld Inlay Equivalency Test Demonstration -ID Surface Flaw Description Flaw 3-Q2 (Alternative Planar Flaw)
Flaw 3-Q3 (Alternative Planar Flaw)
Flaw 3-Q4 (Alternative Planar Flaw)
Flaw 12-Q2 (Thermal Fatigue Crack)
Flaw 12-Q3 (Thermal Fatigue Crack)
Flaw 12-Q4 (Thermal Fatigue Crack)
Flaw 1 (Thermal Fatigue Crack)
Flaw 3 (Alternative Planar Flaw)
Flaw 12 (Thermal Fatigue Crack)
Abbreviated Terms Circ - Circumferential Ax -Axial Connected Planar Flaws Only Orientation Ligament Truth Truth Through-UT - Measured (inches)
Length Wall Dimension through-Wall Dimension (inches)
Dimension (inches)
(inches)
Ax / 11 ° skew /
0 0.50 0.374 0.40 0° tilt 0
0.50 0.374 0.37 Ax / 11 ° skew /
10° tilt 0
0.50 0.374 0.27 Ax / 11 ° skew /
0° tilt 0
3.50 0.815 0.81 Circ / 5° skew /
0° tilt 0
3.50 0.815 0.81 Circ / 5° skew /
0° tilt 0
3.50 0.815 0.77 Circ / 5° skew /
0° tilt Circ / 8° skew /
0 1.80 0.339 0.37 0° tilt Ax /11 skew /
0 0.60 0.374 0.44 0° tilt Circ / 5 skew /
0 3.50 0.815 0.79 0° tilt RMSE 0.035 RMSE with Vendor Tolerance Adjustment 0.083 Table 4: Defect Matrix for Blind Round Robin Program - Axial Flaws Onl J Flaw Description Orientation Ligament Truth Truth Through-UT-Measured (inches)
Length Wall Dimension through-Wall Dimension (inches)
Dimension (inches)
M1 (Stress Axial 0
Not provided 1.18 1.18 Corrosion Crack (SCC>>
M2 (SCC)
Axial 0
Not provided 0.37 0.18
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 11 of 12 Table 4 (Continued): Defect Matrix for Blind Round Robin Program - Axial Flaws Only Flaw Description Orientation Ligament Truth Truth Through-UT - Measured (inches)
Length Wall Dimension through-Wall Dimension (inches)
Dimension (inches)
M3 (SCC)
Axial 0
Not provided 0.66 0.56 M4 (SCC)
Axial 0
Not provided 0.61 0.37 M5 (SCC)
Axial 0
Not provided 0.16 0.09 M6 (SCC)
Axial 0
Not provided 0.90 0.63 RMSE 0.17 RMSE with Vendor Tolerance Adjustment 0.125 In addition, the BVPS-2 examinations will utilize UT analysts that have achieved consistent through-wall depth sizing results within a RMSE of 0.189 inch for OSM welds, and a RMSE of 0.245 inch for combined dissimilar metal and austenitic stainless steel welds. These UT analysts are trained on the POI demonstrated procedure using test data from non-POI performance demonstrations and other investigations on cracked samples. A subset of these UT analysts will be used for the BVPS-2 examinations.
FENOC considered the use of a site-specific mockup for 10 RMSE demonstrations; however, based on current UT examination technology, there would be no benefit to design and build a mockup for BVPS-2 to attempt further vendor 10 RMSE demonstrations.
As stated by EPRI, no vendor has demonstrated the ability to achieve the required 0.125 inch RMSE value. The proposed alternative assures that the subject welds would be fully examined by vendor procedures, equipment and personnel qualified by demonstration in all aspects except depth sizing. For depth sizing, the proposed addition of the numeric difference between the required and demonstrated achievable sizing tolerance to any flaw that is required to be sized compensates for the potential variation. FENOC has determined that the proposed alternative provides an acceptable level of quality and safety, pursuant to the provisions of 10 CFR 50.55a(g)(5)(iii).
7.0 Duration of Proposed Alternative The proposed alternative shall be utilized during the remainder of the BVPS-2 third 1 O-year inservice inspection interval, which began on August 29, 2008, and is currently scheduled to expire on August 28, 2018.
Beaver Valley Power Station, Unit No.2 10 CFR 50.55a Request 2-TYP-3-RVSE-1 Page 12 of 12 8.0 Precedent
- 1. Braidwood Station, Units 1 and 2 - Third 10-Year Inservice Inspection Interval Relief Request 13R-08, Alternative Requirements to ASME Section XI Appendix VIII Supplements 2 and 10), Examinations of Class 1 Pressure Retaining Welds Conducted from the Inside Surface In Accordance with 10 CFR 50.55a(a)(3)(i), (TAC Nos. ME6024 and ME6025), April 11, 2011.
- 2. R.E. Ginna Nuclear Power Plant - Safety Evaluation for Relief ISI-04 for the Fourth 10-Year Inservice Inspection Interval (TAC No. ME5120), April 12,2011.
- 3. Beaver Valley Power Station, Unit Nos. 1 and 2 - Safety Evaluation BV3-RV-2, Inservice Inspection Relief Request (TAC Nos. MD1137 and MD1138),
October 2, 2006.
- 4. Joseph M. Farley Nuclear Plant, Unit 1, and Vogtle Electric Generating Plant, Units 1 and 2 - Safety Evaluation of the Relief Request ISI-GEN-ALT-06-02 (TAC Nos.
MD2482, MD2483 and MD2484), September 29,2006.
9.0 References Electric Power Research Institute letter to FENOC, "Summary of WESDYNE International, LLC Supplements 2 & 10 Depth Sizing Results Obtained from the Inside Surface," September 23, 2011.