L-17-319, Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-4 Examination Requirements (Request No. 2-TYP-4-RV-02)
| ML17318A030 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/13/2017 |
| From: | Bologna R FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2-TYP-4-RV-02, L-17-319 | |
| Download: ML17318A030 (29) | |
Text
FENOC ::....._
Beaver Valley Power Station P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 Richard D. Bologna 724-682-5234 Site Vice President Fax: 724-643-8069 November 13, 2017 L-17-319 10 CFR 50.55a(z)(2)
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 Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-4 Examination Requirements (Request No. 2-TYP-4-RV-02)
Pursuant to 10 CFR 50.55a(z)(2), FirstEnergy Nuclear Operating Company (FENOC) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-729-4 volumetric and surface examination coverage requirements for certain Beaver Valley Power Station Unit No. 2 (BVPS-2) reactor vessel head penetrations. The proposed alternative provides reasonable assurance of structural integrity and is described in detail in the Enclosure.
FENOC requests approval of the proposed alternative by September 28, 2018 to permit implementation of the alternative during the twentieth BVPS-2 maintenance and refueling outage. The maintenance and refueling outage is scheduled to begin in October 2018.
Westinghouse Electric Company LLC Technical Report WCAP-16144, Revision 1, "Structural Integrity Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: Beaver Valley Unit 2," was used as a basis for this request.
Proprietary and non-proprietary versions of WCAP-16144, Revision 1, were previously submitted with a FENOC letter dated December 30, 2008 (Accession Number ML090020385).
Beaver Valley Power Station, Unit No. 2 L-17-319 Page 2 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, Manager - Fleet Licensing, at 330-315-6810.
Sincerely, Richard D. Bologna
Enclosure:
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 cc: NRC Region I Administrator NRC Senior Resident Inspector NRR Project Manager Director BRP/DEP Site BRP/DEP Representative
Enclosure L-17-319 Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 (26 pages follow)
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 1 of 26 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)
- Hardship Without a Compensating Increase In Quality and Safety -
1.0 ASME Code Components Affected
Component Numbers:
Beaver Valley Power Station Unit No. 2 (BVPS-2) Reactor Vessel (2RCS-REV-21)
Head Penetrations 1, 3, 5 through 34, 37, 38, 41, 42, 43, 45, 46, 48, 49, and 54 through 65 Code Class: 1 Examination Category: PWR [Pressurized Water Reactor] Reactor Vessel Upper Head Item Number: 84.20
==
Description:==
UNS N06600 nozzles and UNS N06082 or UNS W86182 partial-penetration welds in head 2.0 Applicable Code Edition And Addenda American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, 2013 Edition, with no addenda.
3.0 Applicable Code Requirements The Code of Federal Regulations 10 CFR 50.55a(g)(6)(ii)(D)(1) requires that examinations of the reactor vessel head be performed in accordance with ASME Code Case N-729-4 (Reference 8.1 ), subject to the conditions specified in paragraphs 10CFR50.55a(g)(6)(ii)(D)(2) through (4).
Paragraph 2500, "Examination Requirements," of ASME Code Case N-729-4 states, in part:
If obstructions or limitations prevent examination of the volume or surface required by Figure 2 for one or more nozzles, the analysis procedure of Mandatory Appendix I shall be used to demonstrate the adequacy of the examination volume or surface for each such nozzle. If Mandatory Appendix I is used, the evaluation shall be submitted to the regulatory authority having jurisdiction at the plant site.
Figure 2, as referenced by paragraph 2500, requires that the volumetric or surface examination coverage distance below the J-groove weld (Dimension "a") be 1.5 inches for incidence angle "8" less than or equal to 30 degrees, or 1 inch for incidence angle "8" greater than 30 degrees; or to the end [bottom] of the tube, whichever is less. These
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 2 of 26 coverage requirements are applicable to the BVPS-2 reactor vessel upper head penetrations as follows:
Incidence Angle, "8 Required Coverage, "a" Penetrations (degrees) (inches) 1 to 33 s 30 1.5 34 to 65 > 30 1.0 4.0 Reason For Request The bottom end of the BVPS-2 reactor vessel upper Figure 1: Reactor Vessel head control rod drive mechanism (CROM) penetrations Head Penetration are externally threaded, internally tapered, and have an ultrasonic corner shadow zone produced by the thread 1----o.625" -"-1 relief, precluding ultrasonic or eddy current data Alloy acquisition in a zone extending up approximately 1.45 J-wel d 600 inches from the bottom of each nozzle. For the majority "Down~iill" Tu be Side of the penetrations, these geometric limitations reduce OD ID the inspectable distance from the bottom of the J-groove weld fillet to the top of the thread relief to less than the required coverage Dimension "a" shown in Figure 2 of ASME Code Case N-729-4.
Figure 1 and Table 1 show the geometry of the BVPS-2 I 0~~
reactor vessel head penetrations and the attainable examination coverage (in inches) below the J-groove weld fillet on the downhill (limiting) side of each 125" penetration.
Based upon the measured values listed in Table 1, l l alternatives to the coverage requirements of ASME Code ~0.314" f I Case N-729-4 with respect to the volumetric and surface 20" examinations of Item 84.20 are necessary. Specifically, alternatives to the required 1.5 inches for reactor vessel ~ = Unexamined Region head penetrations 1, 3, and 5 through 33 (with incidence angles less than or equal to 30 degrees) and from the required 1 inch for reactor vessel head penetrations 34, 37, 38, 41, 42, 43, 45, 46, 48, 49, and 54 through 65 (with incidence angles greater than 30 degrees) are necessary.
The presence of thermal sleeves in the majority of the CROM penetrations prohibits dye penetrant testing of the tapered inside diameter (ID) surface of the penetration tube.
Dye penetrant testing of the outside diameter (OD) of the penetration tubes is difficult due to the threads and the need to properly clean the surface to provide accurate test results. As a result, performing dye penetrant testing on the bottom nozzle inside diameter area would require thermal sleeve removal and would result in significant radiation exposure to plant personnel. The estimated radiation dose rate in the under-head penetration tube area is 1.26 roentgen equivalent man (REM) per hour.
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 3 of 26 Table 1: Penetration Inspection Coverage Inspection Inspection Measured Coverage Measured Coverage Penetration Penetration Dimension A Obtainable Dimension A Obtainable Number Number (inches) Dimension B (inches) Dimension B (inches) (inches) 1 1.64 1.44 34 1.12 0.92 2 1.88 1.68 35 1.60 1.40 3 1.52 1.32 36 1.28 1.08 4 1.72 1.52 37 1.08 0.88 5 1.44 1.24 38 1.04 0.84 6 1.46 1.26 39 1.36 1.16 7 1.40 1.20 40 1.44 1.24 8 1.32 1.12 41 1.12 0.92 9 1.60 1.40 42 1.12 0.92 10 1.60 1.40 43 1.16 0.96 11 1.28 1.08 44 1.36 1.16 12 1.60 1.40 45 1.08 0.88 13 1.56 1.36 46 1.00 0.80 14 1.52 1.32 47 1.28 1.08 15 1.40 1.20 48 1.08 0.88 16 1.64 1.44 49 1.08 0.88 17 1.46 1.26 50 1.36 1.16 18 1.32 1.12 51 1.40 1.20 19 1.36 1.16 52 1.32 1.12 20 1.20 1.00 53 1.28 1.08 21 1.52 1.32 54 1.16 0.96 22 1.28 1.08 55 1.00 0.80 23 1.40 1.20 56 0.92 0.72 24 1.48 1.28 57 1.12 0.92 25 1.40 1.20 58 0.60 0.40 26 1.24 1.04 59 0.96 0.76 27 1.44 1.24 60 0.88 0.68 28 1.32 1.12 61 0.88 0.68 29 1.08 0.88 62 0.68 0.48 30 1.20 1.00 63 0.88 0.68 31 1.60 1.40 64 1.08 0.88 32 1.44 1.24 65 0.80 0.60 33 1.32 1.12 5.0 Proposed Alternative And Basis For Use As an alternative to the volumetric and surface examination coverage requirements shown as Dimension "a" in Figure 2 of ASME Code Case N-729-4, FENOC proposes the use of attainable ultrasonic examination distances shown as Dimension "B" in Table 1 of this request. Specifically, in lieu of the required 1.5 inches for incidence angles less than or equal to 30 degrees (penetrations 1, 3, and 5 through 33), and the required 1.0 inch for incidence angles greater than 30 degrees (penetrations 34, 37, 38, 41, 42, 43, 45, 46, 48, 49, and 54 through 65), the examination coverage recorded as Dimension "B" in Table 1 above will be obtained. For all other penetrations, the
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 4 of 26 required examination coverage Dimension "a" reflected in Figure 2 of ASME Code Case N-729-4 will be met or exceeded. Table 2 provides the scope of this request as a summary of the applicable N-729-4 examination coverage distances "a" (based upon incidence angle) and alternative (achievable) examination coverage for each penetration.
Table 2: Scope of Request Alternative Alternative Pen. 9 a Relief Pen. 9 a Relief Coverage Coverage No.* (degrees) (inches) Requested No.* (degrees) (inches) Requested (inches) (inches) 1 0.0 1.5 1.44 Yes 34 33.1 1.0 0.92 Yes 2 8.7 1.5 N/A No 35 33.1 1.0 N/A No 3 8.7 1.5 1.32 Yes 36 33.1 1.0 N/A No 4 8.7 1.5 NIA No 37 33.1 1.0 0.88 Yes 5 8.7 1.5 1.24 Yes 38 33.1 1.0 0.84 Yes 6 12.4 1.5 1.26 Yes 39 33.1 1.0 N/A No 7 12.4 1.5 1.20 Yes 40 33.1 1.0 N/A No 8 12.4 1.5 1.12 Yes 41 33.1 1.0 0.92 Yes 9 12.4 1.5 1.40 Yes 42 37.3 1.0 0.92 Yes 10 17.6 1.5 1.40 Yes 43 37.3 1.0 0.96 Yes 11 17.6 1.5 1.08 Yes 44 37.3 1.0 N/A No 12 17.6 1.5 1.40 Yes 45 37.3 1.0 0.88 Yes 13 17.6 1.5 1.36 Yes 46 38.7 1.0 0.80 Yes 14 19.8 1.5 1.32 Yes 47 38.7 1.0 N/A No 15 19.8 1.5 1.20 Yes 48 38.7 1.0 0.88 Yes 16 19.8 1.5 1.44 Yes 49 38.7 1.0 0.88 Yes 17 19.8 1.5 1.26 Yes 50 38.7 1.0 N/A No 18 25.4 1.5 1.12 Yes 51 38.7 1.0 N/A No 19 25.4 1.5 1.16 Yes 52 38.7 1.0 N/A No 20 25.4 1.5 1.00 Yes 53 38.7 1.0 N/A No 21 25.4 1.5 1.32 Yes 54 40.0 1.0 0.96 Yes 22 27.0 1.5 1.08 Yes 55 40.0 1.0 0.80 Yes 23 27.0 1.5 1.20 Yes 56 40.0 1.0 0.72 Yes 24 27.0 1.5 1.28 Yes 57 40.0 1.0 0.92 Yes 25 27.0 1.5 1.20 Yes 58 42.7 1.0 0.40 Yes 26 28.6 1.5 1.04 Yes 59 42.7 1.0 0.76 Yes 27 28.6 1.5 1.24 Yes 60 42.7 1.0 0.68 Yes 28 28.6 1.5 1.12 Yes 61 42.7 1.0 0.68 Yes 29 28.6 1.5 0.88 Yes 62 42.7 1.0 0.48 Yes 30 28.6 1.5 1.00 Yes 63 42.7 1.0 0.68 Yes 31 28.6 1.5 1.40 Yes 64 42.7 1.0 0.88 Yes 32 28.6 1.5 1.24 Yes 65 42.7 1.0 0.60 Yes 33 28.6 1.5 1.12 Yes
- Penetration Number N/A in the table above means not applicable.
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 5 of 26 Appendix I of ASME Code Case N-729-4 provides the analysis procedure for the evaluation of an alternative examination area or volume to that specified in Figure 2 of the ASME Code Case if impediments prevent the examination of the complete zone.
As discussed previously, the BVPS-2 reactor vessel head CROM penetrations are externally threaded and internally tapered, precluding ultrasonic or eddy current data acquisition of the complete zone required by the ASME Code Case for the penetrations listed in Section 1 of this request.
Section 1-1000 of ASME Code Case N-729-4 requires that for alternative examination zones that eliminate portions of the Figure 2 examination zone below the J-groove weld, the analyses shall be performed using at least the stress analysis method (1-2000) or the deterministic fracture mechanics analysis method (1-3000) to demonstrate that the applicable criteria are satisfied. In support of this request, the techniques of both 1-2000 and Method 1 of 1-3200 were performed and are included in Reference 8.2, previously provided to the NRC with request 2-TYP-3-RV-02.
The analyses and evaluations presented in Reference 8.2 were developed in 2008 based on the requirements of ASME Code Case N-729-1 and remain applicable to BVPS-2 for the fourth 10-year inservice inspection interval. Reference 8.2 was developed in accordance with the analysis procedure for alternative examination area or volume definition contained in Appendix I of ASME Code Case N-729-1. The requirements of Appendix I of ASME Code Case N-729-1 are identical to the requirements of Appendix I of ASME Code Case N-729-4. The calculation inputs for the stress analysis and fracture mechanics analysis contained in Reference 8.2 are expected to remain the same throughout the fourth 10-year inservice inspection interval.
There have been no major modifications to the operating limits for BVPS-2 since 2008 that would affect these calculation inputs. Because the calculation inputs used in Reference 8.2 have not changed and are not expected to change throughout the fourth 10-year inservice inspection interval, the technical basis contained in Reference 8.2 remains applicable for the BVPS-2 fourth 10-year inservice inspection interval.
5.1 Stress Analysis in Accordance with 1-2000 Section 1-2000 of ASME Code Case N-729-4 requires a plant-specific analysis to demonstrate that hoop and axial stresses remain below 20,000 pounds per square inch (tensile) over the entire region outside the alternative examination zone but within the examination zone defined in Figure 2. Analyses were performed for five different CROM geometries, including the outermost row (at 42. 7 degrees angular position from the reactor vessel centerline), rows at 40.0 degrees, 38.7 degrees, 25.4 degrees and the center location. The actual achievable ultrasonic examination coverage dimensions were used to define the alternative examination zone. A summary of bounding penetration geometries and examination coverage distances are shown in Table 3.
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 6 of 26 Table 3: BVPS-2 Bounding Analyses and Minimum Inspection Coverage Below the J-Groove Weld Penetration Analyzed Minimum Achievable Reference 8.2 Nozzle Nos. Penetration Inspection Coverage Figure No. Bounded by the Nozzle Distance Below the Bottom Analyzed Nozzle (degrees) of the Weld (inches)
A-1 1 0 1.08 A-1 (Downhill and 2-17 0 1.08 Uphill)
A-2 (Downhill) 18-45 25.4 0.84 A-3 (Uphill) 3.16 A-4 (Downhill) 46-53 38.7 0.80 A-5 (Uphill) 4.32 A-6 (Downhill) 54-57 40.0 0.72 A-7 (Uphill) 4.76 A-8 (Downhill) 58-65 42.7 0.40 A-9 (Uphill) 4.88 The stress analysis methodology and conclusions are in Section 5 of Reference 8.2.
The hoop stress distribution plots for the analyzed geometries are provided in figures A-1 through A-9 (Note that in all cases the hoop stresses during steady state operation dominate the axial stresses. Sections 5.3 through 5.5 of Reference 8.2 contain additional information). The minimum inspection zones shown in figures A-1 through A-g represent the minimum achievable inspection coverage distances for each bounding configuration as shown in Table 3 above.
The hoop stress distribution plots in figures A-1 through A-9 demonstrate that in all cases, the stresses remain below 20,000 pounds per square inch (tensile) over the entire region outside the alternative examination zone but within the examination zone defined in Figure 2 as required by 1-2000 of N-729-4.
5.2 Deterministic Fracture Mechanics Analysis in Accordance with 1-3200, Method 1 In addition to the stress analysis detailed above, a fracture mechanics analysis was performed in accordance with Method 1 of 1-3200 to demonstrate that a potential axial crack in the unexamined zone will not grow to the toe of the J-groove weld prior to the next scheduled examination. Because previous penetration repairs have been required on the BVPS-2 reactor vessel head, the re-examination frequency is every refueling outage (every 18 months) per ASME Code Case N-729-4.
The complete fracture mechanics analysis is provided in Section 6 of Reference 8.2, and was performed using input from the previously discussed stress analysis and bounding penetration geometries. The results of the analysis are shown as flaw tolerance charts which can be used to determine the minimum required inspection
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 7 of 26 coverage to ensure that any flaws initiated below the weld in the region of the penetration nozzle not being inspected would not reach the bottom of the weld before the next inspection (figures 6-12 through 6-20 of Reference 8.2). These crack growth projections are also provided in figures 8-1 through 8-9 of this enclosure.
In accordance with Method 1 of 1-3200, the crack growth calculations performed to produce the flaw tolerance charts assume the initial upper extremity of the through-wall flaw to be at or within the bottom edge of the alternative examination zone, and the lower extremity to be located on the penetration nozzle where either the inside or the outside surface hoop stress becomes compressive. Inside and outside surface hoop stress was averaged and applied along the entire length of the assumed through-wall crack. The stress intensity factor was calculated using the standard expression for an axial through-wall crack in a cylinder. A crack growth rate determination was made in accordance with Appendix O of the 2004 Edition ASME XI.
The resulting flaw tolerance charts in figures 8-1 through 8-9 demonstrate that a postulated through-wall flaw at the bottom edge of the proposed alternative examination zone will not grow to the toe of the J-groove weld within the 18-month inspection interval. In all cases, the crack growth predictions show greater than four years of full power operation required to grow the postulated flaw to the toe of weld. Additionally, the initial upper extremity locations of axial through-wall flaws assumed in figures 8-1 through 8-9 are conservative based on a review of the achievable inspection coverage zone in Table 2 of this request because all the assumed upper crack extremities are located within the achievable inspection zone.
The above proposed alternative examination zones for the BVPS-2 CROM penetrations listed in Section 1 of this request are supported by Appendix I of ASME Code Case N-729-4 utilizing both the stress analysis criteria of 1-2000 and the deterministic fracture mechanics criteria of 1-3200, and thus provide reasonable assurance of structural integrity.
6.0 Duration Of Proposed Alternative The duration of the proposed alternative is the BVPS-2 fourth ten-year inservice inspection interval scheduled to begin on August 29, 2018 and end on August 28, 2028.
- 7. 0 Precedent A similar request 2-TYP-3-RV-02 was previously approved by the NRC (Reference 8.3) to use the proposed alternative examination coverage distances during the BVPS-2 third inservice inspection interval. The third interval is scheduled to end in 2018.
Request 2-TYP-4-RV-02 proposes the same alternative examination coverage distances (shown in Table 2 of this request) as those proposed in request 2-TYP-3-RV-02, and is supported by the same analyses and analysis inputs as discussed in the last paragraph of Section 5.0.
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 8 of 26 8.0 References 8.1 ASME Code Case N-729-4, "Alternative Examination Requirements for PWR
[pressurized water reactor] Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," Approval Date June 22, 2012.
8.2 Westinghouse Technical Report WCAP-16144-P, Revision 1, "Structural Integrity Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: Beaver Valley Unit 2," dated December 2008.
8.3 NRC Letter "Beaver Valley Power Station, Unit No. 2 - Relief Request No.
2-TYP-3-RV-02 Regarding the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-1 Examination Requirements (TAC No.
ME0349)," dated September 28, 2009 (Accession Number ML092640111).
Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 9 of 26 Figure A-1 Hoop Stress Distribution Below the Weld Downhill and Uphill Side (0° CRDM Penetration Nozzle) 10.oao - - - - - - - - - . - - - - - - - - . - - - - - - - - . - - - - - - - - . - - - - - - - r - - - - - - - r - - - - - - - - - ,
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 10 of 26 Figure A-2 Hoop Stress Distribution Below the Weld Downhill Side (25.4° CRDM Penetration Nozzle) 10.om:1 -.------ -------- ----~- ------ -,----- ----.-- ------ T----- ----,
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 11 of 26 Figure A-3 Hoop Stress Distribution Below the Weld Uphill Side (25.4° CROM Penetration Nozzle) 70.000 - - - - - - - - - - - - - - - - - - - - - - - - - - - . - - - - - - - . - - - - - - - - - - - - - - - -
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 12 of 26 Figure A-4 Hoop Stress Distribution Below the Weld Downhill Side (38.7° CRDM Penetration Nozzle) 80,000 .....--------.. ...------- - . . . . . - - - - - - - - - r - - - - - - - - - r - - - - - - - - -~ - - - - - - - - ,
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 13 of 26 Figure A-5 Hoop Stress Distribution Below the Weld Uphill Side (38.7° CRDM Penetration Nozzle) 6fl .OC111 I ! I I I
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 16 of 26 Figure A-8 Hoop Stress Distribution Below the Weld Downhill Side (42.7° CROM Penetration Nozzle) 8{}10[J[) .---===: : : ; - - - . - - - - - - - - - , - - - - - - - , - - - - - - - - - , - - - - - - - - - -
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 18 of 26 Figure 8-1 Through-Wall Axia l Flaws Located in the Center CRDM (0.0 Degrees) Penetration, Uphill and Downhill Side 2_5 --,----,---.---.----,----,----,----,---.---.----,----,----,---.---.---.----,----,----,---,---.---.----,----,----,---.---.----,----,----,----,---,---.---.----,----,---,---,---,---,---.----,---,---,---,---.---.----,---.-,....--,
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 19 of 26 Figure B-2 Through-Wall Axia l Flaws Located in the 25.4 Degrees Row of Penetrations, Uphill Side 2.5 ~-,---,---,--,----.--.------.--,...----r--, ...----,---.--,-----.--,----.--,-----.--.--- --r--,...----r---.-~--.-~-,---,----.--,---- r--.-----r---.r----r---.---,----.--~--.
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 20 of 26 Figure 8-3 Through-Wall Axial Flaws Located in the 25.4 Degrees Row of Penetrations, Downhill Side 2.5
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 21 of 26 Figure B-4 Through-Wall Axial Flaws Located in the 38.7 Degrees Row of Penetrations, Uphill Side I I I I I I I I I I I I I I I I I I I I I I I I
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 22 of 26 Figure 8-5 Through-Wall Axial Flaws Located in the 38. 7 Degrees Row of Penetrations, Downhill Side 2.5 -,----,--..,....----,--,---,--..---,--..--....,....-.---........-.----.----,.----.---,-..,....----,--,------,--..---,--..---,--..--....,....-.---........---,.----.----,.----.---,--r-- ,-~--.--..,..........._____,
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 24 of 26 Figure 8-7 Through-Wall Axial Flaws Located in the 40.0 Degrees Row of Penetrations, Downhill Side 2_5
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Beaver Valley Power Station, Unit No. 2 10 CFR 50.55a Request 2-TYP-4-RV-02 Page 25 of 26 Figure 8-8 Through-Wall Axia l Flaws Located in the 42.7 Degrees Row of Penetrations, Uphill Side 2_5 -.----.--..----.----,-- -.----,---r--r---,,---- ,---.---r----.--..----, ----,---,---.---.----.- ----,--,------.--.----. ----,---.----,---r---.- --,.----,---.---r----.-
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