ML20199D887
| ML20199D887 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/30/1998 |
| From: | Mary Anderson, Charles Brown, Galbraith S IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20199D859 | List: |
| References | |
| INEEL-EXT-98-01, INEEL-EXT-98-01156, INEEL-EXT-98-1, INEEL-EXT-98-1156, NUDOCS 9901200238 | |
| Download: ML20199D887 (74) | |
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INEEL/ EXT-98-01156 o
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November 1998
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Technical Evaluation Report on the
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Third 10-Year Interval inservice l
inspection Program Plan:
Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant, Unit 1, Docket Number 50-348 and Updated inservice Inspection Program Plan: Joseph M. Farley Nuclear Plant Unit 2, Docket Number 50-364 M. T. Anderson
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INEEL/ EXT-98 01156 November 1998 f
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Technical Evaluation Report on the 1
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Plan: Joseph M. Farley Nuclear Plant l
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INEELIEXT-98-01156 Technical Evaluation Report on the i
Third 10-Year Interval inservice Inspection Program Plan:
l Southern Nuclear Operating Company, l
Joseph M. Farley Nuclear Plant Unit 1, Docket Number 50-348 and Updated inservice Inspection Program Plan:
Joseph M. Farley Nuclear Plant Unit 2, Docket Number 50-364 i
M. T. Anderson, C. T. Brown, S. G. Galbraith, A. M. Porter i
Published November 1998 Idaho National Engineering and Environmental Laboratory Materials Physics Department l
Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 i
l Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JCN No. J2229 (Task Order A24) i l
4 ABSTRACT This report presents the results of the evaluation of the Joseph M. Earley Nuclear Plant Inservice Inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components and the Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program submitted May 28,1997, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. The Joseph M. Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components and the Joseph M.
Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program are evaluated in l
Section 2 of this report. The inservice inspection (ISI) plan is evaluated for (a) compliance
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with the appropriate editionladdenda of Section XI, (b) acceptability of examination i
sample, (c) correctness of the application of system or component examination exclusion
. criteria, and (d) compliance with ISI related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this l
report.
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l This work was funded under:
4 U.S. Nuclear Regulatory Commission j
JCN No. L2229, Task Order A24 Technical Assistance in Support of the NRC Inservice Inspection Program i
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SUMMARY
The licensee, Southern Nuclear Operating Company (SNC), prepared the Joseph M.
Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year for Class 1. 2, and 3 Components and the Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice inspection Program to meet the requirements of the 1989 Edition of the American Society of j
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI. The third 10-year interval for Unit 1 began December 1,1997, and will end on November 30,2007.
By a safety evaluation report (SER) dated March 20,1997, the NRC allowed Southern Nuclear Operating Company (SNC) to update the Unit 2 ISI Program approximately 44 months early to coincide with the required update of the Unit 1 program. The Joseph M.
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Farley Nuclear Plant Unit 2 UpdatedInservice Inspection Program will cover second and third interval examinations in the time frame from December 1,1997 through November 30, 2007 (updated interval). Therefore, Unit 2 is currently in the second 10-year interval and the third 10-year interval will begin July 30,2001.
l The information in the Joseph M. Farley Nuclear Plant Inservice inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components and the Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program submitted May 28,1997. was reviewed. The review included requests for relief from the ASME Code Section XI l
requirements that the licensee has determined to be impractical. As a result of this review, a request for additionalinformation (RAl) was prepared describing the information and/or clarification required from the licensee in order to complete the review. The licensee provided the requested information in a submittal dated April 6,1998. A second RAI was prepared and the licensee provided the requested information in a submittel dated July 13,1998.
Based on the review of the program plan (s), the licensee's responses to the Nuclear Regulatory Commission's RAls, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, no deviations fron, rQAory requirements or commitments were identified in the Joseph M. Farley Nuclear Plant inservice inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components and the Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program.
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CONTENTS ABSTRACT...
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SUMMARY
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- 1. I NT R O D U CTI O N......................
......... 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN
... 3 2.1 Documents Evaluated.......................
... 3 2.2 Compliance with Code Requirements
....... 3 2.2.1 Compliance with Applicable Code Editions
............ 3 2.2.2 Acceptability of the Examination Sample
.............. 5 2.2.3 Exem ption Criteria..................................... 5 2.2.4 Augmented Examination Commitments..
....... 5 2.3 Conclusion
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- 3. EVALUATION OF RELIEF REQUESTS
....7 3.1 Class 1 Components
.7 3.1.1 Reactor Pressure Vessel..
............... 7 3.1.1.1 Request for Relief No. RR-13 Revision 1, Examination Category B G-1, Item B6.10, Reactor Pressure Vessel (RPV)
Closure Head Nuts............................... 7 3.1.1.2 Request for Relief No. RR-16 Revision 1, Examination Category B D, Item B3.90, Roactor Pressure Vessel (RPV)
Nozzle-to-Vessel Welds............................ 8 3.1.1.3 Request for Relief No. RR-17 Revision 1, Examination Category B-A, Item No. B1.30, Reactor Pressure Vessel (RPV)
Shell to-Flange Weld.
............................11 3.1.1.4 Request for Relief No. RR-18, Use of Code Case N-521 Alternative Rules for Deferral of Inspections of Nozzle-to Vessel Welds, Inside Radius Sections, and Nozzle-to-Safe End Welds of Pressurized Water Reactor (PWR) Vessels 13 3.1.2 Pressurizer.....
............... 15 3.1.2.1 Request for Relief RR-6 Revision 1, Examination Category B-D, item B3.110, Pressurizer Nozzle-to-Vessel Welds......... 15 3.1.3 Heat Exchangers and Steam Generators......
. 17 3.1.3.1 Request for Relief RR-7 Revision 1, Examination Category B-F, item B5.70, Steam Generator Nozzle-to-Safe End Butt Welds
.. 17 3.1.3.2 Request for Relief RR 8, Revision 1, Examination Category B-D, item B3.140, Steam Generators (Primary Side) Nozzle Inside Radius Section.........
............ 19 3.1.4 Piping Pressure Boundary
.. 20 3.1.4.1 Request for Relief RR-9, Revision 1, Examination Category B-J, item B9.31, Branch Pipe Connection Welds NPS 4 or Larger. 20 iv t
r 3.1.5 Pump Pressure Boundary..
............25 3.1.6 Valve Pressure Boundary.
. 25 3.1.7 General
. 25 3.2 Class 2 Components
. 25 3.2.1 Pressure Vessels
... 25 3.2.1.1 Request for Relief RR-10 Revision 1, Examination Category C-B, item C2.22, Nozzle irsside Radius Section
. 25 3.2.1.2 Request for Relief No. RR-14, Revision 1. IWC-1220 Components Exempt From Examination...
. 26 3.2.2 Piping...
.. 30 3.2.2.1 Request for Relief No. RR-19, Use of Code Case N 524, Alternative Examination Requirements for 1.ongitudinal Welds l
in Class t end 2 Piping
... 30 3.2.3 Pumps
..... 31 3.2.3.1 Request for Relief No. RR-15, Revision 1. Examination Category C-C, item C3.30, integrally Welded Attachments on Charging Pumps
. 31 3.2.4 Valves
........ 33 3.2.5 General..
.. 33 l
3.2.5.1 Request for Relief No. RR 20, Revision 1, Use of Code Case N-509, Alternative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments...
.... 33 3.3 Class 3 Components
. 34 3.3.1 Pressure Vessels
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3.3.2 Piping.
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3.3.4 Valves
.. 34 3.3.5 General.....
.............. 34 3.4 Pressure Tests..................
..........................34 3.4.1 Class 1 System Pressure Tests.............
.....34 3.4.1.1 Request for Relief No. RR-21, Use of Code Case N-498-1, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems...
34 3.4.1.2 Request for Relief No. RR-22, Use of Code Case N-416-1, Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement items by Welding, Class 1, 2, and 3
. 37 3.4.1.3 Request for Relief No. RR-26, ASME Class 1, Small Diameter l
(s 1 inch), Reactor Coolant system (RCS) Pressure Boundary Vent and Drain Connections....................... 39 l
3.4.2 Class 2 System Pressure Tests...
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3.4.2.1 Request for Relief No. RR-24, Examination Category C-H, l
Hydrostatic Testing of Charging Pump Suction Piping From the CVCS Boric Acid Blender, Boric Acid Filter, and Chemical
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Mixing Tank.
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3.4.2.2 Request for Relief No. RR-28, Use of Code Case N-522, l
Pressure Testing of Containment Penetration Piping.
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3.4.2.3 Request for Relief No. RR-30, Examination Category CH, IWC-5222, Pressure Testing of Safety injection System Piping Segments Which Are Nonisolable From Class 1 Piping 46 3.4.3 Class 3 System Pressure Tests
... 48 3.4.3.1 Request for Relief No. RR-25, Class 3, IWA 5244(b) Buried Portions of Service Water System Piping
.49 3.4.3.2 Request for Relief No. RR-29, IWA-5244(b), (c), Concrete Encased Portions of Spent Fuel Pool Cooling System Piping Adjacent to Spent Fuel Pit.
.. 50 l
3.4.4 General
. 52 3.4.4.1 Request for Relief No. RR-23, Revision 1, IWA-5250(a)(2) l Corrective Measures for Bolted Connection..
... 52 3.5 General....
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3.5.1 Ultrasonic Examination Techniques 54 i
3.5.1.1 Request for Relief No. RR-1, Material Requirements for Calibration Blocks Used For Ultrasonic Examination of Heavy I
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l Wall Vessels
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3.5.1.2 Request for Relief No. RR-2, Notch Location Requirements for
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l Calibration Blocks Used for Ultrasonic Examination of Heavy Wall Vessels 56 3.5.1.3 Request for Relief No. RR-3, Hole Location Requirements for Calibration Blocks Used for Ultrasonic Examination of Heavy Wall Vessels
........... 57 3.5.1.4 Request for Relief No. RR-4, Dimensional Requirements for Notches Placed in Ultrasonic Calibration Blocks.,....... 59 3.5.1.5 Request for Relief No. RR-5, Curvature Differences Between Ultrasonic Calibration Blocks and the Components to be Examined.....
................. 60 3.5.2 Exempted Components......
........60 3.5.3 Other
..............61 3.5.3.1 Request for Relef No. RR-11, IWA-2610, Reference System for All Welds and Areas Subject to Volumetric and Surface Examination.
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3.5.3.2 Request for Relief No.12, Snubber Testing.
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- 4. CONCLUSION.
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- 5. REFERENCES
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TECHNICAL EVALUATION REPORT ON THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
SOUTHERN NUCLEAR OPERATING COMPANY, JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1, DOCKET NUMBER 50-348 AND UPDATED INSERVICE INSPECTION PROGRAM PLAN:
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2, DOCKET NUMBER 50-364
- 1. INTRODUCTION Throughout the service life of a water cooled nuclear power facility, its components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1,2, and 3 are required by 10 CFR 50.55a(g)(4) (Reference 1) to meet the requirements, except the design and access provisions and the p;eservice examination requirements, of the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, (Reference 2) to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120 month inspection intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein, and subject to Nuclear Regulatory Commi.ssion (NRC) approval. The licensee, Southern Nuclear Operating Company (SNC), has prepared the Joseph M. Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components (Reference 3) and the Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program (Reference 4) to meet the requirements of the 1989 Edition j
of the ASME Code,Section XI. The third 10 year intervai for Unit 1 began December 1, l
1997, and will end on November 30,2007. By a safety evaluation report (SER) dated l
March 20,1997 (Reference 5), the NRC allowed Southern Nuclear Operating Company (SNC) to update the Unit 2 ISI Program approximately 44 months early, to coincide with l
the required update of the Unit 1 program. The Joseph M. Far/ey Nuclear Plant Unit 2 1
4 UpdatedInservice Inspection Program (Reference 4) will cover second and third interval examinations in the time frame from December 1.1997 through November 30,2007 (Updated interval). Therefore Unit 2 is currently in the second 10-year interval and the third 10 year interval for Unit 2 will begin July 30,2001.
Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the Code requirements may be used when authorized by the NRC. The licensee must demonstrate either that the proposed alternatives provide an acceptable level of quality and safety, or that Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. Pursuant to 10 CFR 50.55a(g)(5)(iii), if the licensee determines that conformance with certain Code examination requirements is impractical for its f acility, the licensee shall submit informat;on to the NRC to support that determination. Pursuant to 10 CFR 50,55a(g)(6)(i), the NRC will evaluate the licensee's determination that Code requirements are impractical. The NRC may grant relief and may impose alternative l
requirements that it determines to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
The inf ormation in the Joseph M. Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components (Reference 3) and the Joseph M.
Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program (Reference 4) submitted May 28,1997, were reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. This review was performed using the standard review plans of NUREG-0800, Section 5.2.4, " Reactor Coolant Boundary inservice Inspections and Testing," and Section 6.6, " Inservice inspection of Class 2 and 3 Components" (Reference 6).
In letters dated February 12,1998 (Reference 7) and June 12,1998 (Reference 8), the NRC requested additionalinformation that was necessary to complete the review of the inservice inspection (ISI) program plan. The requested information was provided by the licensee via letters dated April 6,1998 (Reference 9) and July 13,1998 (Reference 10).
The Joseph M. Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components (Reference 3) and the Joseph M. Farley Nuclear Plant Unit 2 UpdatedInservice inspection Program (Reference 4) are evaluated in Section 2 of this report. The ISI program plans are evaluated for (a) complance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the NRC's previous reviews. The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI,1989 Edition. Inservice test programs for j
snubbers and for pumps and valves are being evaluated in other reports.
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- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to ISI activities. This section describes the submittals reviewed and the results of the review.
2,1 Documents Evaluated Review has been completed on the following information from the licensee:
Joseph M. Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year for Class 1, 2, and 3 Components dated May 28,1997 (Reference 3)
Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program dated May 28,1997 (Reference 4)
I Joseph M. Farley Nuclear Plant inservice Inspection (ISil Program Response to NRC Qucstions/ Comments UpdatedISIprograms, dated April 6,1998 l
(Reference 9) l Joseph M. Farley Nuclear Plant inservice inspection (ISI) Program Correction to Response to NRC Questions / Comments UpdatedISIprograms, dated April 27, 1998 (Reference 11)
Joseph M. Farley Nuclear Plant inservice Inspection (ISI) Program Response to NRC Questions / Comments UpdatedISIprograms, dated July 13,1998 (Reference 10) l l
2.2 Compliance with Code Requirements i
2.2.1 Compliance with Applicable Code Editions Inservice inspection program plans are to be based on Section XI of the ASME Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). The third interval at Joseph M. Farley Nuclear Plant Unit 1 began December 1,1997; therefore, the Code applicable t'o the third interval ISI program is the 1989 Edition. The third interval at Joseph M. Farley Nuclear Plant Unit 2 will begin July 30,2001. However the licensee received authorization in an SER dated March 20,1997 (Reference 5) to update the ISI Program 44 months early to coincide with the required update of the Unit 1 program. Therefore, the Code applicable to the remaining portion of the second interval and the third interval updated ISI Program is the 1989 Edition.
in accordance with 10 CFR 50.55a(c)(3),10 CFR 50.55a(d)(2), and 10 CFR 50.55a(e)(2), ASME Code cases may be used as alternatives to Code l.
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requirements. Code cases that the NRC has approved for use are listed in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability, (Reference 12) with any additional conditions the NRC may have imposed. When used, these Code cases must be implemented in their entirety. The licensee may adopt an approved Code case by providing written notification to the NRC. Published Code cases awaiting approval and subsequent
- listing in Regulatory Guide 1.147 may be adopted only if the licensee requests, and the NRC authorizes, their use on a case by-case basis.
The licensees ISI programs include the Code cases listed below. These Code cases either have been approved for use in Regulatory Guide 1.147 or are included as requests for relief.
Code Case N 3071 Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, When Examinations Are Conducted From the Center Drilled Hole Code Case N-416 Alternative Rules for Hydrostatic Testing of Repair or Replacement l
of Class 2 Piping j
Code Case N-416-1 Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement Items by Welding, class 1, 2, and 3.
(Evaluated in 63.1.4.2 of this report.)
Code Case N-432 Repair Welding Using Automatic or Machine Gas Tungsten-Arc Welding (GTA W) Temperbead Technique Code Case N-457 Qualification Specification Notch Location for Ultrasonic Examination of Bolts and Studs l
l Code Case N-460 Alternative Examination Coverage for Class 1 and 2 Welds i
' Code Case N-461 Alternative Rules for Piping Calibration Block Thickness f
Code Case N-463-1 Evaluation Procedures and Acceptance Criteria for Flaws in Class 1 Ferritic Piping that Exceed the Acceptance Standards of IWB-3514.2 Code Case *N-491 Alternative Rules for the Examination of Class 1, 2, and 3 and MC Components and Supports of Light Water Cooled Power Plants Code Case N-498 Alternative Rules for Ten Year Hydrostatic Pressure Testing for 7
Class 1 and 2 Systems Code Case N-498-1 Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems (Evaluated in 53.1.4.1 of this report.)
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Code Case N-509 Alternative Rules for the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments (Evaluated in 63.2.5.1 of this report.)
Code Case N 521 Alternative Rules for DeferralofInspections of Nozzle-to Vessel Welds, inside Radius Sections, and Mozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessels (Evaluated in 63.1.1.4 of this report.)
Code Case N-522 Pressure Testing of Containment Penetration Piping (Evaluated in 63.4.2.2 of this report.)
Code Case N-524 Altemative Examination Requirements for Longitudinal Welds in i
Class 1 and 2 Piping (6 valuated in 63.2.2.1 of this report.)
2.2.2 Acceptability of the Examination Sample j
inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1,2, and 3 components and their supports using sampling schedules described i
in Section XI of the ASME Code and 10 CFR 50.55a(b). Sample size and weld selection procedures have been implemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct.
i 2.2.3 Exemption Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). The exemption criteria have been applied by the licensee in accordance with the Code, as discussed in the ISI program plan, and appear to be correct, with the exception of those components specified in Request for Relief RR 14, 2.2.4 Augmented Examination Commitments In addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:
Reacter vessel examinations will be performed in accordance with the requirements of Regulatory Guide 1.150, Rev.1 (Reference 13).
The reactor coolant pump fly wheel will be inspected in accordance with plant l
l Technical Specifications.
1 The steam generator tubing will be inspected in accordance with Plant Technical Specification 4.4.6 and Regulatory Guide 1.83, Rev 1 (Reference 14).
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4 The main steam lines will be inspected in accordance with plant Technical Specification 4.4.11.3 and Branch Technical positions APCSB-3-1 and MEB-3-1.
2.3 Conclusion Based on the review of the documents listed in Section 2.1, no deviations from regulatory requirements or commitments were identified in the Joseph M. Far/ey Nuclear Plant Inservice Inspection Program Unit 1 Third Ten Year for Class 1, 2, and 3 Components (Reference 3) and the Joseph M. Farley Nuclear Plant Ursit 2 Updated Inservice Inspection Program (Reference 4). Note that this report does not include a review of the implementation of the augmented examinations,it merely records that the licensee has committed to perform them.
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- 3. EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the third 10-year inspection interval are evaluated in the following sections.
3.1 Class 1 Components 3.1.1 Reactor Pressure Vessel 3.1.1.1 Request for Relief No. RR-13 Revision 1. Examination Category B-G-1, item B6.10, Reactor Pressure Vessel (RPV) Closure Head Nuts.
Code Requirernent-Section XI, Table IWB-2500-1, Examination Category B-G-1, item B6.10 requires a surface examination of the Reactor Vessel Closure Head Nuts each 10-year interval.
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to perform a VT-1 visual examination in lieu of the required surface i
examinations. The licensee stated:
"In lieu of the 1989 Edition, Code-required surface examination, the subject RPV Closure Head Nuts will receive a VT-1 Visual Examination."
Licensee's Basis for Requesting Relief (as stated)-
l "The ASME,Section XI,1989 Addenda, Table IWB-2500-1, Category B-G-1, item l-B6.10, allows Visual Examination Method, VT-1 in lieu of the Surface Examination Method required by the 1989 Code."
Justification "The closure head nut configuration does not allow for an adequate Magnetic Particle examination. The MT method requires two-directional coverage to detect the surface flaws. The I.D. configuration permits examination in one direction only.
"Section XI Code personnelin ISI Optimization performed a survey on bolting which did not reveal any service induced cracking. This survey was then used as part of the technical basis for changing the code required examination for Category B-G-1, item B6.10. The 1989 Addenda and subsequent editions of ASME Code changed the examination requirement from a surface examination to a VT-1. The proposed December 3,1997 amendment to 10CFR50.55a issued by the NRC proposed the l
adoption of the 1995 Edition of ASME Section XI with Addenda through 1996. As a j
result, the NRC has recognized that VT-1 examinations of RPV closure head nuts provides an acceptable alternative to the 1989 Code required surface examinations.
Public health and safety will not be endangered; therefore, this request should be granted pursuant to the requirements of 10CFR50.55a(a)(3)(i)."
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Evaluation-The Code requires that a surface examination be performed on the reactor vessel closure head nuts each 10-year interval. The licensee has proposed to perform a VT-1 visual examination of the reactor pressure vessel (RPV) closure head nuts in lieu of the Code-required surf ace examination, it should be noted that all items in Examination Category B-G-1 except the reactor pressure vessel closure head nuts and the closure studs (when removed) require VT-1 visual examinations and/or volumetric examination (as applicable).
I Typical conditions that would require corrective action prior to putting closure head nuts back into service would include corrosion, deforened or sheared threads, deformation, and degradation (i.e., boric acid attack). The Code requires a surface examination for closure head nuts. Surface examination procedures are typically qualified for the detection of linear flaws (cracks) and have acceptance criteria specifying only rejectable linear flaw lengths. Acceptance criteria are not provided in the 1989 Edition of the Code, item B6.10, as they were in the course of preparation when the Code was published. Without clearly l
defined acceptance criteria, conditions that require corrective measures may not be
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l adequately addressed. The 1989 Addenda of Section XI addresses these problems by j
changing the requirement for the subject reactor pressure vessel closure head nuts from l
surface to VT-1 visual examination and providing appropriate acceptance criteria.
l Article IWB 3000, Acceptance Standards, IWB 3517.1, Visual Examination, VT-1, describes conditions that require corrective action prior to continued service for bolting and
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associated nuts. One of these requirements is to compare crack-like flaws to the flaw standards of IWB-3515 for acceptance. Because the VT-1 visual examination acceptance criteria include evaluation of crack-like indications and other conditions requiring corrective j
action, such as deformed or sheared threads, localized corrosion, deformation of part, and other evidence of degradation mechanisms,it can be concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure 1
head nut. As a result, the INEEL staff believes that VT-1 visual examination provides an acceptable level of quality and safety.
1 Conclusion-Based on the comprehensive assessment that the VT-1 visual examination i
l provides, and considering that the 1989 Addenda and later editions of the Code require l
only a VT-1 visual examination on reactor pressure vessel closure head nuts, it is concluded that an acceptable level of quality and safety will be provided by the proposed
(
alternative. Therefore, it is recommended that the proposed alternative, VT-1 visual examination, be authorized pursuant to 10 CFR 50.55a(a)(3)(i),
t 3.1.1.2 Request for Relief No. RR-16 Revision 1, Examination Category B D, item B3.90, Reactor Pressure Vessel (RPV) Nozzle-to-Vessel Welds d
Code Requirement -Section XI, Table IWB-2500-1, Examination Category B D, item B3.90 requires a volumetric examination of reactor pressure vessel (RPV) nozzle-to-vessel welds as defined by Figure IWB-2500-7(a) and (b). The examination volume includes 100% of 8
the weld length. Additionally,Section XI, Article 1-2100 requires that ultrasonic examination of vessel welds greater than 2 inches in thickness be conducted in accordance with ASME Code,Section V, Article 4. Article 4 requires two-directional coverage wherever feasible.
Licensee's Code Re//ef Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from examination coverage for reflectors oriented transverse to the weld for the RPV nozzle-to-vessel welds listed below.
Outlet Nozzles Inlet Nozzles Unit 1 ALA 1 -1100-17 ALA1-1100-18 ALA1-1100-19 ALA 1 -1100-20 ALA1-1100-21 ALA1-1100-22 Unit 2 APR1-110017 APR1-1100-18 APR1-1100-19 APR1-1100-20 APR1-1100-21 APR1-1100-22 I
Licensee's Basis for Requesting Relief (as stated) -
" Examination coverage and the basis for limitations for each type of nozzle are listed below:
" Inlet Nozzles-The required examination volume and associated weld configuration for the inlet nozzles is shown in Section XI, Figure IWB-2500-7(a), except that the inner radius is a smooth contour, not a protrusion as shown in the figure. Coverage and limitations for this configuration are:
Reflectors Parallel to the inlet Nozzle-To-Vessel Weld - Ultrasonic examinations will
=
l be performed from the nozzle bore using scans as allowed by T-441.3.2.2 Coverage from this direction is 99%.
l Reflectors Transverse to the Inlet Nozzle-To-Vessel Weld Ultrasonic examinations will be performed on the ID of the vessel wall and accessible portions of the l
adjoining nozzle using scans, directed clockwise and counterclockwise. Scanning l
cannot be performed on the curved portion of the nozzle inner radius. Coverage from this direction is 70%.
i Composite Coverage - Composite coverage is calculated as 84.5% based on the e
average of the two coverages listed above
" Outlet Nozzles - Tne required examination volume and associated weld configuration (barrel type nozzle with a protruding inner radius) for the outlet nozzles is shown in 9
Section XI, Figure IWB-2500-7(a). Coverage and limitations for this configuration are listed below.
1.
Reflectors Parallel to the Outlet Nozzle-To-Vessel Weld Ultrasonic examinations will be performed from the nozzle bore using scans, as allowed by T-441.4.2.
Coverage from this direction is 100%.
2.
Reflectors Transverse to the Outlet Nozzle-To-Vessel Weld - Ultrasonic examinations will be performed on the ID of the vessel wall using scans, directed clockwise and counterclockwise. The protruding inner radius prevents scanning i
on the nozzle. Coverage from this direction 53%.
3.
Composite Coverage Composite coverage is calculated as 76.5% based on the average of the two coverages listed above."
Justification "Various techniques have been evaluated including the use of additional angles; however, it was concluded that the techniques described above permits the maximum practical coverage to be obtained. Compliance with Code coverage requirements would necessitate refabrication of the RPV nozzles, which would be extremely expensive. Denial of this relief request would cause an excessive burden upon Southern Nuclear Operating Company because refabrication of the nozzles to perform the Code required examinations is impractical; therefore, approval of this relief request should be granted pursuant to 10CFR50.55a(g)(6)(i).
Licensee's Proposed Alternative Examination (as stated)-
" Ultrasonic examination of these welds will be performed to the maximum extent practical from the nozzle bore and from the RPV ID surface. No other examination will be conducted."
Evaluation-The code requires 100% volumetric examination for the subject nozzle-to-vessel welds. Complete examination coverage is not possible due to nozzle configuration, including nozzle curvature and the protruding inner radius portion of the outlet nozzles.
Therefore, the volumetric examination is impractical to perform to the extent required by the Code. To meet the Code requirements, the nozzle-to-vessel welds would require refabrication. Imposition of this requirement would create a considerable burden on the licensee.
j The licensee can complete a significant portion (84.5% composite coverage of the inlet nozzles and 76.5% composite coverage of the outlet nozzles) of the Code-required volumetric examinations. Therefore, existing patterns of degradation will be detected and reasonable assurance of the structuralintegrity of the subject nozzle-to-vessel welds will be provided.
10
~ - - - -..
b Conclusion-The volumetric examinations of the subject nozzle to-vessel welds in the Reactor Pressure Vessel are impractical to perform at Farley, Unit 1 and Unit 2, to the extent required by Section XI of the ASME Code because of the geometric configuration of the nozzles. Performing the ultrasonic examinations to the maximum extent practical from the nozzle bore and from the RPV inside surface will provide reasonable assurance of the continued structuralintegrity of the pressurizer nozzle to vessel welds. Therefore,it is recommended that relief be granted pursuant to 10CFR50.55a(g)(S)(i).
3.1.1.3 Request for Relief No. RR-17 Revision 1, Examination Category B A, item No.
B1.30, Reactor Pressure Vessel (RPV) Shell to-Flange Weld.
Code Requirement-Section XI, Table IWB-2500-1, Examination Category B-A, item B1.30, requires a volumetric examination of the reactor pressure vessel shell-to-flange weld. The l
applicable examination volume, shown in Figure IWB-2500-4, includes essentially 100% of the weld length.
Licensee's Code Re//ef Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from volumetric examination of the RPV shell to-flange weld to the extent required by the Code. Specifically, relief is requested from examination coverage for reflectors oriented tran /erse to shell-to-flange Weld ALA1-1100-1 for Unit 1 and Weld APR 1 1100-1 for Unit Licensee's Basis for." ~uesting Relief (as stated)-
" Examination coveroge and the basis for the limitations are listed below.
(Attachment 17-1 shows the Farley shell to-flange configuration)".
1.
Reflectors Parallel to the Shell to-Flange Weld-As allowed by T-441.3.2.2, ultrasonic examinations will be perforrned from the flange seal surface with the sound beam striking the examination volume at near-normalincidence to the weld fusion line. Essentially 100% coverage will be obtained during these examinations; therefore, Code requirements for reflectors oriented parallel to the weld will be met.
Reflectors transverse to the Shell-to-Flange Weld Automated examinations using a sled in contact with the RPV ID will be performed on this weld from the using transducers oriented to detect indications transverse to the weld; however, interference between the sled holding the transducers and the sharp ID taper will l
limit the examinations. The examinations will be performed in both the clockwise and counter-clockwise directions with a minimum of 47% of the examination volume being scanned. Discussions with Westinghouse examination personnel i
indicate that the flange configuration at Farley has a severe taper when compared to many other reactor pressure vessels they examine; therefore, the limited
{
coverage is appropriate.
i l
- a. Figures, drawings and attachments furnished with the hcensee's submittal are not included in this report.
f i
11
2.
Composite Coverage - Based on the average of the two scans listed above, composite coverage is calculated as 73.5%."
Justification
" Examinations conducted during the first interval were performed using a combination of manual examinations from the flange surface and automated " immersion" technique examinations from the RPV ID. During the " Immersion" technique examinations, the physical flange geometry had much less effect on coverage than it did with the
" contact" technique currently used by NDE vendors, since with an " immersion" technique the transducers are not in contact with the ID surface. Since NDE vendors have changed to contact techniques, total compliance with Code requirements would necessitate either refabrication of the RPV flange or for the NDE vendors to change equipment and techniques. Refabrication of the RPV to install a new flange or requiring an NDE vendor to obtain/ develop specialized automated inspection equipment for Farley would be very expensive. Denial of this relief request would cause an excessive burden upon SNC because refabrication of the RPV flange to perform Code required examinations is impractical; therefore, approval of this relief request should be granted pursuant to 10CFR50.55a(g)(6)(i).
" Additionally, the examination of the weld from the flange surface provides assurance that circumferential cracking exceeding acceptance standards have not developed.
From a technical standpoint, circumferential cracking is considered to be the more limiting case with service-induced axially oriented cracking considered a very unlikely scenario for this weld. Therefore, while the specified code coverage requirements (for axially oriented flaws) will not be met, the Code examination performed from the flange surface in conjunction with the limited automated examinations will provide reasonable l
assurance that the structuralintegrity of the weld is being maintained."
Licensee's Proposed Alternative Examination (as stated)-
"None. Southern Nuclear Operating Company will continue to work with the NDE vendor to evaluate techniques and equipment such that optimized coverage of this weld is obtained, to the extent practical."
Evaluation-The Code requires that the subject reactor pressure vessel shell-to-flange wcld be 100% volumetrically examined during the inspection interval. Due to the extreme flange taper, the ability to scan for indications transverse to the shell-to-flange weld is i
limited. It i's impractical to examine the subject welds to the extent required by the Code.
To obtain the complett coverage required by the Code, design modification of the RPV flange assembly or redesign of automated inspection equipment by the NDE vendors would j
be necessary. Imposition of this requirement would cause a considerable burden on the l
- licensee, i
The licensee has committed to continue to work with the NDE vendor to evaluate i
techniques and equipment that will optimize coverage of this weld to the extent practical.
The licensee calculates that the composite coverage achievable is 73.5% of the required 12
l >
examination volume. Based upon the percent of volumetric coverage obtainable, it is reasonable to conclude that a pattern of degradation,if present, will be detected. As a result, reasonable assurance of continued structuralintegrity will be provided.
Conclusion-Based on the above evaluation, it is concluded that obtaining the Code-required volumetric coverage is impractical for the reactor pressure vessel shell-to flange
- welds at Farley Units 1 and 2. frnposition of the Code requirements would cause a considerable burden on the licensee. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.1.1.4 Request for Relief No. RR 18, Use of Code Case N 521 Alternative Rules for Deferral ofInspections of Nozzle to Vessel Welds, inside Radius Sections, and l
Nozzle-to-Safe End Welds of Pressurized Water Reactor (PWR) Vessels Code Requirernent-Section XI, Table IWB 2500-1, Examination Category B-D, items B3.90 and B3.100, Examination Category B F, item B5.10, requires 100% volumetric l
examination of all reactor vessel nozzle-to-shell welds and nozzle inner radius sections and nozzle-to-safe end butt welds each inspection interval as defined by Figures lWB-2500-7 and IWB-2500-8. At least 25% but not more than 50% (credited) of the nozzles shall be examined by the end of the first inspection period and the remainder by the end of the j
inspection interval.
L l
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3), the licensee proposed to use the alternative requirements contained in Code Case N-521, Alternative Rules for Deferral of Inspections of Nozzle-to-Vessel Welds, Inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel, in lieu of the examination requirements of the 1989 Edition of ASME XI. The licensee stated:
" Southern Nuclear Operating Company proposes that, as an alternate to the timing of the examinations of the nozzle-to-vessel welds,inside radius sections, and nozzle-to-safe end welds required by the 1989 ASME Code, that for Farley Unit 1(Unit 2 as applicable) these examinations be accomplished according to the alternate requirements of Code Case N-521."
Licensee's Basis for Requesting Relief (as stated)-
"On August 9,1993, ASME issued Code Case N-521 which approved rules for j
deferring the examinations of Nozzle-to-Vessel Welds, Inside Radius Sections, and i
Nozzle-to-Safe End Welds of PWR reactor vessel to the end of the inspection l
interval'if certain conditions are met."
Justification 4
" Historically, pressurized water reactors have examined the reactor vessel outlet 4
nozzle-to-shell welds, their inside radius sections, and associated nozzle-to-safe end j
welds during the first inspection period in order to comply with the requirements of the l
ASME Section XI Code. The remaining nozzle examinations and the reactor vesselISI are completed later in the inspection interval. These examinations are performed 13
l l
remotely from the vesselinside surface using a submerged automated inspection tool.
The examinations are generally critical path during refueling outages, as they are performed when the vesselis de fueled and has water in the refueling canal. The similar reactor vesselinlet nozzles are accessible only when the core barrel is removed as when the 10 year reactor vessel ISI is being performed. To comply with the 1989 Code inspection schedule requires installing the automated inspection tool on the vessel a minimum of two times. This creates a hardship on Southern Nuclear Operating Company without a corresponding increase in safety or quality. Code Case N-521 permits an alternative examination schedule whereby the automated inspection tool will be needed only once per 10-year interval. This will provide an acceptable examination level that does not endanger the health and safety of the public. In addition, Southern Nuclear will realize significant opportunities for savings in contractor cost, critical path time, radiation exposure, and internal manpower requirements. Therefore, the use of Code Case N-521 should be granted pursuant to 10 CFR 50.55a(a)(3)(ii). All nozzles were examined during the third period, second interval, in order to maintain an approximate 10-year examination interval." (For Unit 2) "All nozzles are scheduled during the third period, second interval, in order to maintain an approximate 10-year examination interval."
Evaluation-The Code requires examination of at least 25% but not more than 50%
(credited) of the nozzles by the end of the first inspection period and the remainder by the end of the inspection interval. The licensee has requested authorization to use Code Case N-521 and defer examination of these areas until the end of the third 10-year interval.
Code Case N-521 states that examination of RPV nozzles, inner radius sections, and nozzle-to-sate end welds may be deferred provided (a) no inservice repairs or replacements by welding have ever been performed on any of the subject areas, (b) none of the subject areas contain identified flaws or relevant conditions that currently require successive i
inspections in accordance with IWB-2420(b), and (c) the unit is not in the first interval.
I The licensee, in response to the NRC's request for additionalinformation, confirmed that the above conditions have been met. In addition, the licensee has volumetrically examined (for Unit 2, will examine) all the subject areas during the third period of the second interval, thereby establishing a new sequence of examinations that will not exceed 10 years between inspections.
To comply with the examination schedule required by the Code it would be necessary to install the automated inspections toolin the Reactor Pressure Vessel a minimum of two j
times. Imp'osition of this requirement on Southern Nuclear Operating Company would cause a burden. Because the licensee repeated (for Unit 2, will repeat) the examinations at the end of the previous interval and will meet the conditions in the Code Case, the licensee's proposed alternative will provide an acceptable level of quality and safety since the maximum time of 10 years between inspections will not be exceeded.
Conclusion-Considering that the licensee has met (Unit 2, will meet) all of the conditions stated in Code Case N-521, has examined all of the affected areas during the third period of the second interval,and has established a new sequence of examinations such that the 14
~ _.. ~ - - - _ _
. - _. ~. _.
. -.....,... ~. ~
. - _ -... - ~ _ _. ~. - - - -.
l-time between examinations will not exceed 10-years, the licensee's proposed alternative l
will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(i). The use of Code Case N 521 should be authorized for the third 10 year interval at Farley Unit 1 and for the time frame associated with the Unit 2 Updated Interval, or until the Code Case l
is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee may continue to use the Code Case with the limitations, if any, listed in the Regulatory Guide 1.147, 3.1.2 Pressurizer 3.1.2.1 Request for Relief RR-6 Revision 1, Examination Category B D, Item B3.110, Pressurizer Nozzle-to-Vessel Welds i
Code Requirement-Section XI, Table IWB-2500-1, Examination Category B D, item B3.110 requires a 100% volumetric examination of the pressurizer nozzle-to vessel welds as defined by Figure IWB 2500-7(b).
4 l
j
' Licensee's Code Relie/ Request-Pursuant to 10 CFR 50.55atg)(6)(i), the licensee requested relief from examining 100% of the Code-required volume of the pressurizer l
nozzle-to-vessel welds. Specifically, two-directional coverage cannot be obtained for the required examination volume of the following welds:
Unit 1 Unit 2 l
ALA1-2100-9 APR1-2100-9 ALA1-2100-10 APR1 2100-10 ALA1-2100-11 APR1-2100-11 ALA1-2100-12 APR1 2100-12 l
l ALA12100-13 APR1-2100-13 ALA1-2100-14 APR1 2100-14 i
Licensee's Basis for Requesting Relief (as stated)-
" Coverage listed below is the minimum projected coverage based on the following scans: 0-degree for laminar reflectors: 45,60, and 70-degree for reflectors parallel j
to the weld (when scanning clockwise and counter-clockwise on the weld / base material).
Unit 1 ALA1-2100-9 through -13
}.
Unit 2 APR 1 -2100-9 through -13 l
i 5
15
4 "These welds are accessible from the head side; however, the geometric configuration of the nozzles (attachment 6-1)* limits coverage from the nozzle side. Composite coverage is 75%.
Unit 1 ALA 1 -2100-14 Unit 2 A PR 1 -2100-14 "The pressurizer surge line nozzle-to-vessel weld is located on the bottom head with a ring of heater penetrations situated around the weld. While there is a small annular space remaining between the weld and the heater penetrations, coverage from the head is limited. Ultrasonic coverage from the nozzle is also limited due to nozzle configuration (Attachment 6-1). Composite coverage is 65%."
l Justification "Various techniques were evaluated during previous examinations such as using additional angles and bouncing the ultra-sound off of the clad surface; however, none were proven to be of practical use. Southern Nuclear Operating Company (SNC) will continue efforts to optimize coverage of these welds during future examinations.
" Compliance with Code coverage requirements would necessitate refabrication of the pressurizer nozzles, which would be very expensive. Denial of this relief request would cause an excessive burden upon SNC because refabrication of the nozzles to perforrn the Code required examinations is impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i)."
Licensee's Proposed Alternative Examination las stated)-
" Perform an ultrasonic examination of the nozzle to vessel welds to the maximum extent possible and perform a supplemental surface examination."
Evaluation-The licensee included a sketch in the request for relief detailing the nozzle configuration. The geometric configuration of the pressurizer nozzles is such that the Code-required 100% coverage cannot be achieved and is, therefore, impractical. To examine the welds in accordance with the requirements, the pressurizer nozzles would have to be redesigned, fabricated, and installed.
The licensee states that, for Welds ALA(APR)1-2100 9 through -13, 75% composite coverage can be obtained and, for Weld ALA(APR)12100-14,65% composite coverage can be obtained. The licensee evaluated various techniques during previous examinations, including the use of additional angles and bouncing the ultra-sound off of the clad surface; however, the licensee determined that none of the techniques proved to be of practical use. The licensee will perform a supplemental surface examination to provide additional confidence in the integrity of the subject welds.
- b. Figures, drawings and attachments furnished with the hcensee's submittal are not included in this report.
16
_ _ _ _ _.. - - ~.
Attaining the examination composite coverages stated above and performing surface I
examinations of the nozzles, coupled with the licensee's commitment to continue efforts to optimize coverage of these welds during future examinations, will provide reasonable assurance that the nozzle-to-vessel welds have not developed inservice-related flaws exceeding the Code requirements.
Conclusion-The volumetric examinations of the pressurizer nozzle to-vessel welds are impractical to perform at Farley, Unit 1 and Unit 2, to the extent required by Section XI of the ASME Code because of the geometric configuration of the nozzles as well as interference from other pressurizer components. The proposed alternative, to perform the ultrasonic examinations to the maximum extent possible and perform a supplemental surface examination, will provide reasonable assurance of the continued structuralintegrity of the pressurizer nozzle to vessel welds. Therefore, it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).
3.1.3 Heat Exchangers and Steam Generators 3.1.3.1 Request for Relief RR-7 Revision 1, Examination Category B-F, item B5.70, Steam l
Generator Nozzle-to Safe End Butt Welds l
Code Requirement-Section XI, Table IWB-2500-1, Examination Category B-F, item 85.70 l
requires both 100% volumetric and surface examination of the Steam Generator dissimilar metal nozzle-to-safe end butt welds as defined by Figure IWB-2500 8.
Licensee's Code Re//ef Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from examining 100% of the Code-required volume of the Steam Generator nozzle (with weld deposited safe-end) to pipe welds listed below.
Unit 1 l) nit 2 ALA1-4100-4CM A PR 1 -4100-4DM ALA 1 -4100-5 DM APR 1 -4100-5 DM ALA1-4200-4DM APR1-4200-4DM ALA1-4200-5DM APR1-4200-5DM j
ALA1-4300-4DM APR1-4200-4DM ALA1-4300-5DM APR1-4200-5DM 4
Licensee's Basis for Requesting Relief (as stated)-
"The complex configuration of these welds is shown in Attachment 7-1*.
Complete examination of each of these welds would require access from both sides of the C.
Figures, drawings and attachments furnished with the hcensee's s#-:..ol are not included in this report.
17
~,
weld; however, examination is limited both by nozzle geometry and the presence of the external weld-deposited clad overlay at the safe end to nozzle interface.
Ultrasonic examination can be conducted from the pipe side (up to and including the majority of the weld surface); however, the OD counterbore, the weld geometry, and the weld-deposited clad overlay may prevent complete scanning from this direction. Examination from the nozzle side is severely limited by the rough, as-cast, surface condition of the nozzle and the presence of the weld-deposited overlay. Composite coverage is 50%."
Justification "Various techniques have been evaluated such as bouncing the ultra-sound off of the inside surface; however, they are not practical for use on centrifugally cast stainless steel pipe. Southern Nuclear Operating Company (SNC) will continue efforts to optimize coverage of these welds during future examinations.
" Compliance with code coverage requirements would necessitate refabrication of the Steam Generator nozzles, which would be very expensive. Denial of this relief request would cause an excessive burden upon SNC because refabrication of the nozzles to perform the code required examinations is impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i)."
Licensee's Proposed Alternative Examination-None. Code required exams will be performed to extent practical.
Evaluation-The licensee included a sketch in the request for relief showing the nozzle geometry and the presence of the as-welded clad overlay at the nozzle-to-safe end interface. The geometry of the nozzle, including the clad overlay is such that the volume of the steam generator nozzle to safe end welds cannot be completely examined. The steam generator nozzle design, therefore, makes the Code-required examination impractical. To examine the welds in accordance with the requirements of the Code, the Steam Generator Nozzles would have to be refabricated and installed. Imposition of the Code requirements would result in a significant burden on the licensee.
The licensee has determined that the composite examination coverage obtainable is 50%. Various techniques have been evaluated however, none of the techniques are practical for centrifugally-cast stainless steel pipe. The licensee's proposed alternative is to perform'the ultrasonic examination to the maximum extent practical (possible).
l Southern Nuclear Operating Company has also committed to continue efforts to optimize coverage of these welds during future outages.
[
The proposed alternative, combined with the Code-required surface examination and continued efforts to optimize coverage, will provide reasonable assurance that the steam generator nozzle-to-safe end butt welds have not developed inservice-related flaws exceeding the Code requirements.
18
_ _. _ _.m O
l Conc /usion-The volumetric examinations of the steam generator nozzle to safe end butt welds are impractical to perform at Farley, Unit 1 and Unit 2, to the extent required by Section XI of the ASME Code due to the geometric configuration of the nozzles and the as-welded clad overlay at the nozzle-to-safe end interface. The proposed alternative, to perform the ultrasonic examinations to the maximum extent practical (possible), coupled with the Code required surface examination, will provide reasonable assurance of the continued structuralintegrity of the subject steam generator nozzle-to-safe end butt welds.
Therefore, it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).
3,1.3.2.
Request for Relief RR 8, Revision 1, Examination Category B D, item B3.140, Steam Generators (Primary Side) Nozzle Inside Radius Section 1
Code Requirement-Section XI, Table IWB-25001, Examination Category B D, item B3.140 requires 100% volumetric examination of the steam generator (primary side) nozzle inside radius sections as defined by Figure IWB-2500-7.
Licensee's Code Re//e/ Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from performing volumetric examination of the steam generator's primary side inlet and outlet inner radius sections identified below:
i Unit 1 Unit 2 ALA1-3100-IR-1 APR1-3100-IR-1 ALA1-3100-IR-2 A PR 1 -3100-IR-2 l
ALA1-3200-IR-1 APR1-3200 IR-1 l
ALA1-3200 IR-2 APR1-3200-IR-2 ALA13300-IR-1 APR1-3300 IR-1 ALA1-3300-IR-2 APR1-3300-IR-2 Licensee's Basis for Requesting Relief (as stated)-
"The steam generator primary side nozzles are integrally cast as part of the channel head; therefore, no welds exist which require volumetric examination. The steam generator nozzle inner radius section cannot be volumetrically examined from the outside of the nozzle or channel head because the rough, as-cast contact surface is not suit'able for ultrasonic coupling and the geometrical configuration (Attachment 8-1) requires an excessively long test metal distance resulting in high ultrasonic attenuation. The areas inside of the nozzles and channel head are covered with cladding in the 'as welded' condition; therefore, meaningful volumetric examination cannot be performed from the as-welded surface. Even with the proper preparation of the inside surface for volumetric examination, an adequate examination of the area of interest (base metal just belove the cladding) could not be achieved due to the resulting ultrasonic response at the clad-to-base metalinterface."
4 19 4
Justification "The steam generator nozzle sections were not designed for examination of the inside radius csing ultrasonic methods. Compliance with Code requirements would necessitate refabrication of the steam generator nozzles, which would be very expensive. Denial of this relief request would cause an excessive burden upon SNC because refabrication of the noezles to perform the Code required examinations is impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i)."
Licensee's Proposed Alternative Examination (as stated)-
"The inside surface of each steam generator primary side nozzle inner radius section will be visually examined. The examination area willinclude the inner radius surface region shown in Section XI, Figure IWB-2500-7, to the extent practical."
j l
Eva/uation-The steam generator nozzles in Farley Unit 1 and Unit 2 were not designed for ultrasonic examination of the inner radius from the external surface. The nozzle geometry and the as-cast surface of the steam generator channel head, along with the long test metal distance resulting in high ultrasonic attenuation, impede the ultrasonic examination of the inner radius region from the external surface. The areas inside of the nozzles and channel head are covered with cladding in the "as-welded" condition; therefore, meaningful volumetric examination cannot be performed from the inside surface. The steam generatcr materials, fabrication, and nozzle design; therefore, make the Code-required examination impractical. To meet the Code requirements, the steam generator 1
nozzles would require refabrication and installation. Imposition of the Code requirements would result in a significant burden on the licensee.
The licensee has proposed to perform a (VT-1) visual examination of the nozzle inside radius sections. The INEEL staff believes that such an examination will be capable of detecting rust on the surface of the stainless steel cladding, and will provide reasonable assurance that significant patterns of degradation do not exist and that the inner radius sections have not developed inservice-related flaws exceeding the Code requirements.
Conclusion-The volumetric examination required by Section XI of the ASME Code for the nozzle inside radius sections in the steam generators is impractical to perform at Farley Unit 1 and Unit 2 because of the component geometry and the as-cast surface of the steam generator heads, along with the excessively long test metal distance that results in high ultrasonic attenuation. The proposed alternative, to perform a visual examination of ti.e inside surface of each steam generator primary side nozzle inner radius section, will provide rea' onable assurance of the continued structural integrity of the subject steam s
generator nozzle inner radius regions. Therefore,it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i), provided the visual examinations is a VT-1.
}
f 3.1.4 Piping Pressure Boundary 3.1.4.1 Request for Relief RR-9, Revision 1, Examination Category B-J, item B9.31, Branch Pipe Connection Welds NPS 4 or Larger 20
O i
Code Requirement-Section XI, IWB-2500-1, Examination Category B-J, item B9.31 requires both 100% volumetric and surface examinations of the Class 1 branch pipe connection welds nominal pipe size 4 inches and greater as defined by Figures lWB-2500-l 9, -10, -11.
Licensee's Code Re//e/ Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from examining 100% of the Code-required volume of pressure retaining branch connections located on the centrifugally cast stainless steel, main loop piping welds listed below:
Unit 1 Unit 2 ALA 1 -4100-20BC APRI-4100-15BC ALA1-4100-22BC APRI-4100-16BC ALA1-4200-15BC APRI-4200-15BC l
ALA1-4300-16BC APRI-4200 21BC APRI-4300-15BC i
I Licensee's Basis for Requesting Relief (as stated)-
" Composite coverage in this relief request is calculated by Southern Nuclear Operating (SNC) using the average coverage of four scans: (1) pig side coverage for reflectors oriented parallel to the weld seam, (2) branch connection side coverage for reflectors oriented parallel to the weld seam. (3) clockwise coverage on the weld crown for reflectors oriented transverse to the weld sean., and (4) counter-clockwise coverage on the weld crown for reflectors oriented transveise to the weld seam.
Unit 1
" Welds 4100-20BC,4200-15BC, and 4300-16BC have configurations such that 2.4" to 3.9" thick stainless steel branch connections are " set-in" the centrifugally cast stainless steel main loop piping and then welded. (See " set-in" sketch on Attachment i
9-1).
Coverage is described below.
Unit 2 L
" Welds 4100-1LC,4100-16BC,420015BC and 4300-15BC have configurations such that 2.1" to 4.6" thick stainless steel branch connections are " set-in" the l
centrifugally cast stainless steel main loop piping and then welded. (See " set-in" sketen on Attachment 9-1)".
Coverage is described below.
i 1.
" Pipe Side Coverage for Parallel Reflectors - Due to the severe attenuation properties of the cast stainless steel material used in the main loop piping, meaningful data from the main run of pipe.is only obtainable utilizing a % node 3
4 e
- d. Figures. drawings and attachments furnished with the hcensee's subrnittal are not included in this report.
L 21 4
examination,45' refracted longitudinal (RL) wave technique. Coverage is determined to be 80% of the weld volume from the pipe side (one beam direction).
l 2.
" Branch Connection Coverage for Parallel reflectors - Scans are not performed l
from the branch connections side due to very. limited coverage; therefore, coverage is OE The basis for this determination is detailed below.
Scanning from the branch connection side would require bouncing a shear wave through metal paths (from the transducer to the examination volume) of 7" to 11" (6" to 13" Unit 2) of stainless steel which would significantly attenuate the ultrasonic energy reaching the branch connection / weld interface. Significant attenuation would then be obtained at the weld interface and the shear wave would not effectively penetrate into the cast stainless material. (See Attachments 9-2 and 9-3)* Additional composite coverage would be 4% to 7% if the scanning is performed.
Obtaining this minimal Coverage would require fabrication of three new F-304 stainless steel calibration blocks of non-standard diameter and thickness (approximately 10.1" OD by 2.4" thick,17.6" OD by 3.6" thick, and 19.4" OD by 3.9" thick (Unit 1)). (approximately 9.5" OD by 2.1" thick,18.6" OD by 3.9" thick, and 20.6" OD by 3.9" thick (Unit 2)).
l 3.
" Clockwise Coverage on Weld Crown for Transverse Reflectors - Scanning clockwise on the weld crown for transverse reflectors in the weld root will provide little, if any, meaningful coverage due to the curvature of the weld. While scanning was performed on the weld crown during the second interval, coverage plots indicated that the root of the weld was not effectively reached; therefore, coverage is determined to be OE (See Attachment 9-2 and 9-3).
4.
" Counter-Clockwise Coverage on Weld Crown for Transverse RcMectors -
[
Scanning counterclockwise on the weld crown for transverse reflectors in the weld root will provide little, if any, meaningful coverage due to the curvature of
[
the weld. While scanning was performed on the weld crown du ing the second interval, coverage plots indicated that the root of the weld was not effectively reached; therefore, coverage is determined to be OE (See Attachments 9-2 and I
9 3).
"C' mposite Coverage Using the method described above, the composite 5.
o coverage is calculated to be 20% SNC concludes that examinations for this configuration are performed to the maximum extent practical. Performance of additional examinations (from the branch connection side) of very limited coverage, reduced effectiveness, and with the necessity of f abricating non-
- e. Figures, drawings and attachments furnished with the licensee's submittal are not included in this report.
22 s
, - - - - ~
,+
l i
standard calibration blocks is considered to be a burden, with little compensating increase in the level of safety or quality.
" Weld ALA1-4100-22BC (Weld APR1-4200-21BC Unit 2) has a configuration such that the SA-351, CF8A cast stainless steel "sweepolet" is welded into the cast stainless steel piping. (See "sweepolet" sketch on Attachment 91). Coverage is described below.
1 1.
" Pipe Side Coverage for Parallel Reflectors - 100% of the weld volume is i
examined from the pipe side using the % node,45' refracted longitudinal (RL) wave technique described above (one beam direction).
2.
" Branch Connection Side Coverage for Parallel Reflectors - Coverage from the branch connection side (second beam direction)is 50%. Limitations are due to the combination of cast material and curved configuration.
l 3.
" Clockwise Coverage on Weld Crown for Transverse Reflectors - Approximately 100% of the required code coverage is obtained for this weld configuration.
4.
" Counter-Clockwise cow rage on Weld Crown for Transverse Reflectors -
Approximately 100% of the required code coverage is obtained for this weld configuration.
l l
l 5.
" Composite coverage Composite coverage for the second intervalis calculated to be 87.5%. SNC concludes that examinations for this configuration are performed to the maximum extent practical."
Justification "The geometric configuration of the branch connections prevents ultrasonic or radiographic examination of the welds to the extent required. Primary cracking mechanisms for these welds is considered by the nuclear industry to be stress-corrosion cracking, thermal fatigue cracking, or mechanically induced fatigue cracking.
Each is discussed below.
" Stress Corrosion Cracking - In a low oxygen, PWR primary system water environment there has never been any evidence of stress-corrosion cracking in 304 stainless steel.
i l
" Thermal Fatigue Cracking - Thermal fatigue cracking previously occurred in an FNP primary system branch line; with the cracking initiated by thermal stresses related to stratification. This cracking is located away
}
from the subject branch connection welds, With the subject welds located on the main run of piping, there should be sufficient turbulence and mixing present such that thermal stresses sufficient to initiate cracking in the welds would not be present.
23 i
I
" Mechanically Induced Fatigue Cracking - Fatigue cracking initiated by mechanical means such as vibration is accounted for in the design of the branch connections. However, in the event that unusual vibration remained undetected and subsequently produced cracking, the cracking would most likely have initiated on the outside of the weld and been detected with the required surface examination.
"Overall the potential for cracking in these branch connection welds is low. The low potential for cracking in these welds in conjunction with the partial volumetric examination and complete surface examination performed should provide reasonable assurance of the continued structuralintegrity of these welds. Compliance with Code coverage requirements would require that the branch connection configurations and portions of the main loop piping be redesigned, fabricated, and installed which would be extremely expensive. Denial of this relief request would cause an excessive burden upon SNC because refabrication of the branch connections to perform the Code required examinations is impractical, therefore, approval of this relief request should be granted pursuant to 10 CFR 50.55a(g)(6)(i)."
Licensee's Proposed Altemative Examination (as stated)-
"None. Code required ultrasonic examinations are performed to the extent practical."
r Evaluation--in the request for relief, the licensee included a sketch displaying the branch connections'. geometric configuration. This configuration is such that the Code-required 100% coverage cannot be achieved and is, therefore, impractical. To examine the welds in accordance with the requirements of the Code, the brecch connections, and i
portions of the main loop piping, would have to be redesigned, facCcated, and installed.
Imposition of the Code requirement would result in a significant burden on the licensee.
The licensee states that for Unit 1 Welds 4100-20BC,420015BC, and 4300-16BC, and for Unit 2 Welds 4100-15BC,4100-16BC,4200-15BC and 4300-15BC a 20%
composite examination coverage can be achieved. For Unit 1 Weld ALA1-4100-22BC and Unit 2 Weld APR1-4200-21BC an 80% composite examination coverage can be achieved. The volumetric examination of the welds to the extent practical, in conjunction with the Code-required surf ace examination, will provide reasonable assurance that the subject branch connections have not developed inservice-related flaws exceeding the Code requirements.
Conclusion'-The volumetric examination required by Section XI of the ASME Code for l
l the subject branch connections is impractical at Farley, Unit 1 and Unit 2, because of the branch connections' geometric configurations. The proposed alternative, to perform the examinations to the extent practical, along with the Code-required surface examination, will provide reasonable assurance of the continued structuralintegrity of the subject branch connections. Therefore,it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).
t 24
-. - - - --. -._ ~ -.
4 3.1.5 Pump Pressure Boundary No relief requests.
3.1.6 Valve Pressure Boundary No relief requests.
3.1.7 General r
j i
No relief requests.
l 3.2 Class 2 Components 3.2.1. Pressure Vessels I
3.2.1 M Request for Relief RR-10 Revision 1, Examination Category C-B, item C2.22, Wozzle inside Radius Section Code Requirement-Section XI, Table IWB-2500-1 Examination Category C B, Item C2.22 requires 100% volumetric examination of the nozzle inside radius sections of
)
nozzles greater than one-half inch nominal wall thickness in Class 2 vessels as defined by Figure IWB-2500-4(a) or (b).
Licensee's Code Re/lef Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from performing the Code-required volumetric examination of the Steam Generator outlet nozzle ins sa radius section.
Licensee's Basis for Requesting Relief (as stated)-
"The steam outlet nozzle is manufactured from a solid forging with seven (7) holes, each 8 % inches in diameter, dalled through the forging (Attachment 10-1)' to provide flow restriction. The geometry of this nozzle with the drilled flow restrictor holes does not have an inner radius and; therefore, no meaningful examination can be performed."
Justification "As shown in Attachment 10-1, the steam outlet nozzle was designed with an internal multiple hole type flow restrictor. This design does not use a radiused nozzle as described in Figure IWC-2500-4, but instead has severalinner radii, corresponding to each hole. As a result, the design of the nozzle is not applicable to the Co'a requirement and compliance with the Code requirement is not practical; therefore, this relief request should be granted pursuant to 10 CFR l
50.55a(g)(6)(i)."
(
- f. Figures. drawings and attachments furnished with the licensee's submittal are not included in this report.
25
1 Licensee's Proposed Alternative Examination las stated)-
"None" Evaluation-As shown in the sketch attached to the relief request, the steam outlet l
nozzle was designed with an internal multiple hole type flow restrictor. This design does i
not use a radiused nozzle as described in Figure IWC-2500 4, but instead has several individualinner radii, corresponding to each hole. Therefore, the Code requirement does not apply to the design of the nozzle and is, therefore, impractical.
Conclusion-The Code requirement does not apply to the design of the nozzle and is, thus, impractical. Therefore, it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).
3.2.1.2 Request for Relief No. RR-14, Revision 1, IWC-1220 Components Exempt From Examination Code Requirement-Section XI, IWC-1220, Components Exempt from Examination, contains the exemption criteria for Class 2 components.
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to implement IWC-1220, from the 1989 Addenda of ASME Section XI, thereby I
exempting components from volumetric and surface examinations on vessels (including small diameter vessels fabricated essentially from piping) and their supports in Class 2 piping 4 inch nominal pipe size (NPS) and smaller in systems other than Residual Heat Removal (RHR), Emergency Core Cooling (ECC), or Containment Heat Removal (CHR) systems or portions thereof. The specific components are as follows:
Letdown Delay Tanks Regenerative Heat Exchanger Excess Letdown Delay Tanks Excess Letdown Heat Exchanger Letdown Heat Exchanger Seal Water Return Filter Seal Water Heat Exchanger Volume Control Tank Reactor Coolant Filter Letdown Reheat Heat Exchanger The licensee stated:
" Code required pressure tests will be performed to assure that an acceptable level of safety and quality is maintained for the applicable components."
Licensee's Basis for Requesting Relief (as stated)-
" Exemption criteria have been added to the 1989 Addenda of ASME Section XI (and subsequent editions / addenda) to allow the exemption of vessels, pumps, valves and their connections in piping NPS 4 and smaller (excluding high pressure 26
4 9
safety injection). The December 3,1997 amendment to 10CFR50.55a proposed the adoption of the 1995 Edition of ASME Section XI with Addenda through 1996 (which contains this exemption criteria). Additionally, this exemption criteria is contained in Code Case N-408-2, Alternative Rules for Examination of Class 2 l
Piping, which has received NRC approval for use in Regulatory Guide 1.147.
Therefore, the NRC has specifically recognized the use of this exemption criteria.
"The intent of the change to the exemption criteria is to allow exemption of a component that is connected to exempt piping, provided that failure of the component would not produce a leak greater than the flow through the exempt piping. Each of the components listed above (except for the Regenerative Heat Exchanger and the Volume Control Tank) have single inlet and outlet process piping and, therefore, meet the specific requirements of the exemption. The Regenerative Heat Exchanger and the Volume Control Tank are discussed below:
l
" Regenerative Heat Exchanger - The Regenerative Heat Exchanger has inlet and outlet piping on both the shell side and the tube side. The inlet and outlet piping for j
the shell side is 3" NPS: therefore, a crack or defect in the Regenerative Heat Exchanger shell would not produce a leak greater than would be produced by the loss l
of the corresponding inlet or outlet piping. Similarly, a crack or defect in the l
Regenerative Heat Exchanger tubing (which has no volumetric or surface examination requirements due to the size of the individual tubes) would not produce a leak greater than would be produced by the loss of the corresponding inlet or outlet piping. To produce a leak greater than that produced by the loss of a 4" NPS line would require failure of the shell and failure of multiple heat exchanger tubes, which is not considered a credible inservice failure. Therefore, the intent of the exemption is maintained. Additionally, use of this exemption for this component would eliminate unnecessary examinations located in high dose rate areas. Previous dose rate surveys for the Unit 1 Regenerative Heat Exchanger examinations indicate a contact dose rate of approximately 2800 mrem /hr.
l
" Volume Control Tank (VCT) - The VCT has the following influent / effluent lines.
1.
Normal Letdown Line - This 3" line enters the top of the tank and terminates with a spray nozzle.
2.
Outlet to Charging Pump Suction - The outlet to the charging pump suction header is a 4" line.
3.
Inlet from the CVCS Relief Valve Discharge Header - This 3" line is a common header for three relief valves and will not have flow unless one of the relief valves has lif ted.
4.
Alternate Return from the Seal Water Heater Exchanger - This 3" line enters the top to the tank and terminates with a spray nozzle. This path can be manually j
valved in when the excess letdown line is being used as a backup for the normal l
27
letdown line or when routing the seal return to the VCT gas space when degassing the Reactor Coolant System.
5.
Gas Supply / Return Lines These lines are gas filled only.
"After evaluation of the function of these lines, SNC concludes that the failure of the VCT would not normally produce a leak greater than that produced by the loss of a 4" NPS line. This conclusion is based on the following:
A.
The normal process is for the 3" normalletdown line to discharge to the VCT and for the charging pumps to take suction from the VCT via a 4" suction line.
The 4" suction line can be isolated frorn the tank with two motor operated valves.
B.
The alternate return from the seal water heat exchanger is normally valved out-of-service because the flow path is normally directed to the charging pumps suction piping. This line can only be used by manually repositioning the valves and should not be considered when applying the 1989 Addenda exemption criteria.
C.
Relief valves are not normally open and discharging into the tank while letdown is functioning. This is an infrequent event and should not be considered when applying the 1989 Addenda exemption criteria.
D.
The gas lines have no effect on water leakage from the tank."
hstification "Use of the later Code editionladdenda exemption criteria to exempt the above specified components should have no adverse affects on the existing level of safety and quality, and relief should be granted pursuant to the requirements of 10CFR50.55a(a)(3Hi). Denial of this relief request would require continued personnel radiation exposures to perform examinations not deemed necessary by later Code editions or by the NRC through approval of Code Case N 408-2."
Evaluation-The licensee has requested to use the exemption criteria of IWC-1222 of the 1989 Addenda in lieu of the exemption requirements of the Code of record. In accordance with the 1989 Code, piping NPS 4 and smaller is exempt from examination, but connected components are not. In the 1989 Addenda of Section XI, IWC-1222 was revised to exempt vessels, pumps and valves, and their connections in piping NPS 4 and smaller, with the following note. "In piping is defined as having a cumulative inlet and a cumulative outlet pipe cross-sectional area neither of which exceeds the nominal OD cross-sectional area of the designated size." In other words, a component connected to exempt piping'is exempt if, upon failure, it would not produce a leak greater than the volume flowing through the exempt piping. This exemption is also contained in Code Case N 408-2 Alternative Rules for Examination of Class 2 Piping,Section XI, Division 28
~-
-~ - - - - -.. -. _ -. - -
F 1, which has been approved for general use in Revision 11 of Regulatory Guide 1.147,
/nservice Inspection Code Case Acceptability-ASME Section XI, C; vision 1, The change in the Code described above parallels the logic used for the exemption of Class 1 systems. Specifically, IWB-1220(b)(2) exempts " components and their connections in piping in 1-inch nominal pipe size and smaller", where "in piping" is defined as having one inlet and one outlet pipe, each of which is 1-inch NPS or smaller.
The discrepancy between Class 1 and 2 systems was recognized by the Code committee, which patterned the exemption criteria for Class 2 in the 1989 Addenda after existing exemption requirements for Class 1 systems.
The licensee states that each of the subject components listed, with the exception of the Regenerative Heat Exchanger and Volume Control Tank, meet the specific requirements of the exeraption.
The exemption criteria in the 1989 Addenda for the subject components do not distinguish between tube side and shell side piping of heat exchangers. In the licensee's response to the NRC request for additionalinformation, dated April 6,1998, the licensee stated that the Regenerative Heat Exchanger has 3" NPS inlet and outlet lines on the shell side, and 3" NPS inlet and outlet lines on the tube side. Because the subject component contains two 3" inlets and two 3" outlets, the cumulative inlet and cumulative outlet pipe cross-sectional area exceeds the nominal cross-sectional area of the designated size (4"). Therefore, the INEEL staff believes that the Regenerative Heat i
Exchanger does not meet the exemption criteria described in IWC-1222 in the 1989 Addenda of ASME XI and that it should be examined to the requirements specified in IWC-2500.
The Volume Control Tank has three 3" inlet lines and one 4" outlet line. The licensee, after evaluation of the function of these lines, concluded that the failure of the VCT would not norma //y produce a leak greater than that produced from the VCT 4" suction line. The licensee's evaluatio-ncluded information concerning the ability to isolate the suction line, the fact that ne alternative return line from the seal water heat exchanger is normally valved out of service, and that the 3" iniet line from the CVCS Relief Valve Discharge header is a common header for three relief valves and will not have flow unless one of the relief valves has lifted. The exemption criteria in the 1989 addenda for the subject component do not distinguish piping that is isolated, normally valved out of service, or used only after activation or relief valves from other piping.
Therefore, the INEELstaff believes that Volume Control Tank does not fall within the exemption criteria in IWC-1222 in the 1989 Addenda of ASME XI.
Conclusion-The licensee's proposed alternative, to use the exemption criteria of the l
1989 Addenda and subsequent editions / addenda, will provide an acceptable level of quality and safety for all of the subject components except the Volume Control Tank and
}
the Regenerative Heat Exchangers. Therefore,it is recommended that the licensee's proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(i) for all of the subject components exception of the Volume Control Tank and Regenerative Heat Exchangers.
29
4 3.2.2 Piping 3.2.2.1 Request for Relief No. RR-19, Use of Code Case N 524, Alternative Examination Requirements for longitudinal Welds in Class 1 and 2 Piping Code Requirement-Section XI, IWC-2500-1, Examination Category C-F-1, Items C5.12 and C5.22, and Examination Category C-F-2, items C5.52 and C5.62 require surf ace and volumetric examination of the longitudinal weld for a distance 2.5t from its intersection with a circumferential weld, where t is the thickness of the pipe examined.
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the alternative requirements contained in Code Case N-524 in lieu of the Code requirements for all Class 2 piping. The licensee stated:
" Southern Nuclear Operating Company will comply with the requirements of ASME Section XI, Code Case N-524 as follows: (a) When only a surface examination is required, examination of longitudinal piping welds will be performed on those portions of the welds within the examination boundaries of the intersecting circumferential welds: (b) When both surface and volumetric examination are required, examination of longitudinal piping welds will be performed on those portions of the welds within the examination boundaries of intersecting circumferential welds provided the following requirements are met. (1) Where longitudinal welds are specified and locations are known, examination requirements shall be met for both transverse and parallel flaws at the intersection of the welds and for that length of longitudinal weld within the circumferential weld examination volume; (2) Where longitudinal welds are specified but locations are unknown, or the existence of longitudinal welds is uncertain, the examination requirements shall be met for both transverse and parallel flaws within the entire examination volume of intersecting circumferential welds."
Licensee's Basis for Requesting Relief (as stated)-
" Code Case N-524, approved August 9,1993, by the ASME Boiler and Pressure Vessel Code Committee Addresses alternative requirements for surface and volumetric examination requirements of longitudinal piping welds. By implementing the provisions of this Code Case, personnel radiation exposure, outage examination time, and costs can be significantly reduced at Plant Farley."
Justification "The proposed alternative testing requirements have been evaluated by the ASME Code Committee and have been deemed acceptable for determining the pressure boundary integrity of the affected components. The proposed alternative requirements, in accordance with the Code Case, will provide reasonable assurance that unallowable inservice flaws have not developed in the subject welds or that they will be detected and repaired pHor to retuin of the reactor to service. Thus an acceptable level of quality and safety will have been achieved and public health and 30
safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examination in lieu of the Code requirements."
l Eva/uation-ASME Section XI requires the examination of Class 2 longitudinal piping welds for a distance of 2.5 times the pipe thickness, measured from the intersection of the circumferential weld. The licensee's proposed alternative is to examine only the l
portions of the longitudinal weld within the examination area of the intersecting circumferential weld in accordance with Code Case N-524.
l Longitudinal welds are produced during the manufacture of the piping, not in the field as is the case for circumferential welds. Consequently, longitudinal welds are fabricated under strict manufacturing standards, which provides assurance of structuralintegrity.
l These welds have also been subjected to the preservice and initial inservice l
examinations, which provide additional assurance of structural integrity. No significant I
loading conaitions or material degradation mechanisms have been identified that specifically relate to longitudinal seam welds in nuclear plant piping. The most critical region of the longitudinal weld is the portion that intersects the circumferential weld.
l Since this region will be examined during the examination of the circumferential weld, the licensee's alternative provides an acceptable levei of quality and safety.
Conclusion-Use of Code Case N-524 for Class 2 piping provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative, to use Code Case N-524 for Class 2 piping, be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case N 524 should be authorized for the third 10-year interval at Farley Unit 1 and for the time frame associated with the Unit 2 Updated Interval, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. Af ter that time, the licensee must follow the conditions, if any, specified in the regulatory guide.
3.2.3 Pumps 3.2.3.1 Request for Relief No. RR-15, Revision 1, Examination Category C-C, item l
C3.30, Integrally Welded Attachments on Charging Pumps l
Code Requirement-Section XI, Table IWC-2500-1, Examination Category C-C, item C3.30 requires a surface examination of 100 percent of the areas of each welded l
attachment of Class 2 pumps defined by Figure IWC-2500-5.
l.icensee's Code Re//e/ Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from examining 100% of the Code-required surface of the charging j
pumps integrally welded attachments listed below.
i Unit 1 & 2 ALA2-5100 CS-1, CS-2, CS-3, and CS 4 31
_ _ _ _ _ _.. _ _. - _ _ _ _ _ _ _ _ _.. -. _. _. - - ~
1 ALA2-5110-CS-1, CS-2, CS-3, and CS-4 ALA2-5120-CS-1, CS-2, CS-3, and CS-4 Licensee's Basis for Requesting Relief (as stated)-
"Due to the component configuration, location, and support design, approximately 20 percent of each integrally welded attachment is inaccessible for examination."
Justification "The drawing in Attachment 15-18 shows that the component configuration, location, and support design are such that the Code-required surface of the charging pump integrally welded attachments cannot be fully examined. Compliance with Code coverage requirements wot/dd necessitate obtaining charging pumps with a modified design, which would be very expensive. Denial of this relief request would cause an excessive burden upon SNC because modification of the pumps to perform the Code required examinations is impractical; therefore, approval should be granted pursuant to 10CFR50.55a(g)(6)(i)."
Licensee's Proposed Alternative Examination (as stated)-
"None The surface examination will be performed to the maximum extent possible.
Eva/uation-The drawings included in the licensee's request for relief show that the support design, component configuration, and location are such that the Code-required surface of the charging pump integrally welded attachments cannot be fully examined These conditions make the Code-required examinations impractical. To examine the integral attachments in accordance to the Code requirements would require design modifications of the supports. Imposition of the Code requirements would result in a
]
significant burden on the licensee.
The licensee can compiete approximately 80% of the Code required examination.
Performing the surface examination to this exent will provide reasonable assurance that the subject integral welded attachments have not developed inservice-related flaws exceeding the Code requirements.
Conclusion-The surf ace examination of the charging pump in. grally welded attachments, to the extent required by the Code, is impractical at Farley, Units 1 and 2 because of the component configuration, location of the integral attachments, and the design of the supports. SNC will perform the surface examinations to the maximum extent possible, which will provide reasonable assurance of the continued structural integrity of the subject branch connections. Therefore,it is recommended that relief be i
granted pursuant to 10CFR50.55a(g)(6)(i).
1 4
- g. Figures, drawings and attachments furnished with the hcensee's submittal are not included in this report.
32 l
3.2.4 Valves No relief requests.
3.2.5 General 3.2.5.1 Request for Relief No. RR-20, Revision 1, Use of Code Case N-509, Alternative Rules for the Selection and Examination of Class 1, 2. and 3 Integrally Welded Attachments Code Requirement-Section XI, Examination Categories B-H, B-K-1, and C-C, requires volumetric or surface examination of 100% of the non-exempt integrally welded l
attachments. Examination Categories D-A, D 8, and D-C require visual examination of I
100% of the non-exempt integrally welded attachments.
l Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the alternative requirements contained in Code Case N-509 in lieu of the specified Code requirements. The licensee stated:
"As an alternate, Code Class 1, 2 and 3 examinations will be performed in accordance with Code Case N-509, except that, SNC will ensure that the sample size specified for piping, pumps, and valves will be a minimum of 10% of each of the items."
(
Licensee's Basis for Requesting Relief (as stated)-
"On November 25,1992, ASME issued Code Case N-509 which approved a set of alternative rules for the selection and examination of Class 1,2, and 3 L
Integrally Welded Attachments."
l Justification
" Code Case N-509 provides an alternative sampling plan which will retain an acceptable level of quality and safety for Class 1,2, and 3 Integrally Welded Attachments. Since approval was granted by ASME, the alternative requirements should be technically acceptable for determining flaws: therefore, permission should be granted pursuant to 10 CFR 50.55a(a)(3)(i). By implementing the alternative examinations; cost savings, personnel radiation dose, and outage time can be realized by SNC,"
Evaluation-The licensee has proposed to apply the requirements of Code Case N-509 as an alternative to the Code requirements for the examination of integrally welded attachments on Class 1,2, and 3 piping and components. Thc licensee has also committed to supplement the Code Case with a minimum examination sample of 10% of allintegral attachments to non exempt Class 1,2, and 3 components. Considering that l
most of the Code examination requirements are based on sampling to ensure the detection of service-induced degradation, extending the sampling philosophy to the integral attachment welds will provide an equivalent level of quality and safety.
33 l
. - ~. -.
Therefore, it is concluded that the alternative, to use Code Case N-509 with the minimum sample size of 10% of all non-exempt Class 1, 2, and.3 integrally welded attachments, will provide an acceptable level of quality and safety.
Conclusion--Based on the evaluation above, it is concluded that the use of Code Case N-509 for the subject components provides an acceptable level of quality and safety.
Therefore, it is recommended that the licensee's proposed alternative, to use Code Case N 509, with the commitment to ensure that the sample size specified for piping, pumps, and valves will be a minimum of 10% of each item, be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case N 509 should be authorized for the third 10-year interval at Farley Unit 1 and for the time frame associated with the Unit 2 Updated Interval, or until the Code Case is approved for general use by reference in Regulatory I
Guide 1.147. Af ter that time, the licensee must follow the conditions, if any, specified in the regulatory guide.
3.3 Class 3 Components 3.3.1 Pressure Vessels No relief requests.
3.3.2 Piping No relief requests.
3.3.3 Pumps No relief requests.
3.3.4 Valves No relief requests.
3.3.5 General 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests 3.4.1.1 Request for Relief No. RR-21, Use of Code Case N-498-1, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems I
34 l
Code Requirement-The components with the following Section XIitem numbers are required to be hydrostatically tested once each inspection interval in accordance with paragraphs IWB-5222, IWC-5222, and IWD-5223, as applicable.
Examination Cateaory item Number B-E B4.11, B4.12, B4.13, B4.20 B-P B15.11, B15.21. B15.31, B15.E 1, B15.61, B15.71 C-H C7.20, C7.40, C7.60, C7.80 l
D-A D1.10 D-B D2.10 i
1 D-C D3.10 l
Licensee's Proposed Alternative--Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the alternative requirements contained in Code Case N-498-1 in lieu of the required Code examination. The licensee stated:
" Southern Nuclear Operating Company will comply with the pressure testing requirements of ASME Section XI Code Case N-498-1 for the listed Code item numbers."
Licensee's Basis for Requesting Relief (as stated)-
"ASME Section XI item number components are required to be subjected to a hydrostatic test once each inspection intervalin accordance with paragraphs IWB-5222, IWC-5222, and IWD-5223 as applicable."
Justification "The proposed altemative testing requirements have been evaluated by the ASME Code Committee and the NRC and have been deemed acceptable for determining the pressure boundary integrity of the affected components. Implementation of pressure testing in accordance with the subject Code Case will ensure an acceptable level of quality and safety, does not decrease the margin of public health and safety and is thus authorized pursuant to 10 CFR 50.55a(a)(3)(i). By implementing the alternative examinations, cost savings, personnel radiation dose, and outage time can be realized by Southern Nuclear Operating Company at Plant Farley."
l Evaluation-The Code requires that a system hydrostatic test be performed once per intervalin accordance with the requirements of IWA-5000 for Class 1,2, and 3 pressure-retaining systems. In lieu of the Code required hydrostatic testing requirements, l
the licensee has requested authorization to use Code Case N-498-1, Alternative Rules for 1
35
10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems, dated May 11, 1994.
The system hydrostatic test, as required by Section XI, is not a test of structural integrity of the system but rather an enhanced leakage test. (Reference 15) Under hydrostatic test conditions, piping components receive only a small increase in pressure over the nominal operating pressures. Piping dead weight, thermal expansion, and seismic loads present far greater challenges to the structural integrity of a system than does the loading presented by the hydrostatic test pressure. Consequently, the Section XI hydrostatic pressure test is primarily regarded as a means to enhance leak detection during the examination of components under pressure, rather than as a method of determining structuralintegrity.
I Code Case N-498 Alternative Rules for 10-Year System Hydrostatic Testing for l
Class 1 and 2 Systems, was previously approved for general use on Class 1 and 2 systems in Regulatory Guide 1.147, Rev.11. For Class 3 systems, Code Case Revision N-498-1 specifies requirements identical to those for Class 2 components (for Class 1 and 2 systems, the alternative requirements in N 498-1 are unchanged from N-498). In lieu of 10 year hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-498-1 requires a VT-2 visual examination at nominal operating pressure and temperature in conjunction with a system leakage test performed in accordance with paragraph IWA-5000.
Class 3 systems do not normally receive the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommon, Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a review of Class 3 system failures requiring repair," the most common causes of failure are erosion-corrosion (EC), microbiologically-induced corrosion (MIC), and general corrosion. In general, licensees have implemented programs for the prevention, detection, and evaluation of EC and MIC; therefore, Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms.
Conclusion--Considering that Code Case N-498 was found to be an acceptable alternative for Class 1 and 2 systems, and that Class 3 systems receive inspections l
commensurate with their function and expected failure mechanisms, the licensee's proposed alternative, to use Code Case N-498-1, should provide an acceptable level of quality and safety. Therefore,it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case N 4981 should be authorized for the third 10-year interval at Farley Unit 1 and for the time frame associated with the Unit 2 Updated Interval, or until the Code Case is approved for 1
i
- h. Documented in licensee Event Reports and the Nuclear Plant Reliability Data System 1
databases.
l 36
}
. ~.
- ~ - -..- -.-._
t general use by reference in Regulatory Guide 1.147 After that time, the licensee must follow the conditions, if any, specified in the Regulatory Guide.
l 3.4.1.2. Request for Relief No. RR-22, Use of Code Case N-416-1, Alternative Pressure Test Requirement for We/ded Repairs or Installation of Replacement items by Welding, Class 1, 2, and 3 Code Requirement-Section XI components require a hydrostatic pressure test after welded repairs or installation of replacement items by welding, as noted in IWA 4000.
Licensee's Proposed Alternative--Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the alternative requirements contained in Code Case N-416-1 in lieu of the required Code examination. The licensee stated:
" Southern Nuclear Operating Company will comply with the pressure testing requirements of ASME Section XI Code Case N 416-1 for welded repairs or l
installation of replacement items by welding. In addition to the alternative rules of Code Case N-416-1, SNC proposes to augment the alternative tests by performing an additional surface examination on the root pass layer of butt and socket welds on the pressure retaining boundary of Class 3 components."
Licensee's Basis for Requesting Relief (as stated)-
"ASME Section XI Code Case N-416-1 was issued on February 15,1994. This Code Case has been approved by the NRC staff for use at Plant Farley during the second interval and at other plants, but has not been formally endorsed by l
inclusion in NRC Regulatory Guide 1.147."
l Justification i
"The proposed alternative testing requirements have been evaluated by the ASME Code Committee and the NRC and have been deemed acceptable for determining the pressure boundary integrity of the affected components. Implementation of pressure testing in accordance with the subject Code Case will ensure an acceptable level of quality and safety, does not decrease the margin of public health and safety and is thus authorized pursuant to 10 CFR 50.55a(a)(3)(i). By implementing the alternative examinations, reduction in costs, personnel radiation dose, and outage time can be l
realized by Southern Nuclear Operating Company at Plant Farley."
l Evaluation-Section XI of the Code requires that a system hydrostatic test be performed in accordance with IWA 5000 after repairs by welding on the pressure-retaining boundary. The licensee proposes to implement the alternative to hydrostatic pressure tests contained in Code Case N-416-1 for Code Class 1,2, and 3 repairstreplacements.
In addition, the licensee will supplement the pressure test with an additional surface examination on the root pass layer of Class 3 repair / replacement welds or welded areas.
Code Case N 416-1 specifies that the welds be nondestructively examined in accordance with the applicable subsection of the 1992 Edition of Section Ill. This Code l
37
. ~ -. _ _ -. __. _. - ~. _. -.
4 Case also allows VT-2 visual examination at nominal operating pressure and temperature in conjunction with a system leakage test, in accordance with paragraph IWA-5000 of the 1992 Edition of Section XI. Comparison of the system pressure test requirements of the 1992 Edition of Section XI to those of the 1989 Edition of Section XI, the latest Code edition referenced in 10 CFR 50.55a, shows that:
The test frequencies and pressure conditions are unchanged; The hold times either remained the same or increased; The terminology associated with the system pressure test requirements for
. all three Code classes has been clarified and streamlined; and The nondestructive examination (NDE) requirements for welded repairs remain the same.
Hydrostatic testing subjects the piping components to only a smallincrease in pressure over the design pressure and, therefore, does not present a significant challenge to pressure boundary integrity. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leak detection during the examination of components j
under pressure rather than as a measure of the structuralintegrity of the components.
Following welding, the Code requires volumetric examination (depending on wall' thickness) of repairs or replacements in Code Class 1 and 2 piping components, but only requires a surface examination of the final weld pass in Code Class 3 piping. There are no ongoing NDE requirements for Code Class 3 components except for VT-2 visual examination for leaks in conjunction with the 10 year hydrostatic tests and the periodic pressure tests. However, the INEEL staff believes that the examinations required by Code Case N-416-1 are appropriate for Class 3 systems when (1) a surface examination is performed on the root pass layer of butt and socket welds, and (2) a system pressure test at nominal operating pressure is performed.
Conclusion-Considering that Code Case N-416 h9s been accepted by the NRC and that a supplemental surface examination will be performed on the root pass for Class 3 systems,it is concluded that the licensee's proposed alternative will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative, to use Code Case N-416-1 with a supplemental surface i
examination on the root pass layer of butt and socket welds, be authorized pursuant to l
10 CFR 50.55a(a)(3)(i). The use of Code Case N-416-1 should be authorized for the l
third 10-year interval at Farley Unit 1 and for the time frame associated with the Unit 2 Updated Interval, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. Af ter tl.at time, the licensee must follow the conditions,if any, i
specified in the regulatory guide, f
38
. - - - -. - -.. ~. ~...l 3.4.1.3 Request for Relief No. RR 26, ASME Class 1, Small Diameter (s 1 inch), Reactor Coolant system (RCS) Pressure Boundary Vent and Drain Connections Code Requirement-ASME Section XI, lWB 2500-1, Examination Category B-P, item Numbers B15.51 and B15.71, requires the system hydrostatic test to include all Class 1 components within the system boundary.
l Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform the Class 1 System Hydrostatic Test with the vent and drain valves in the closed position. The licensee stated:
"The RCS vent and drain connections will be visually examined with the isolation valves in the normally closed position each refueling outage for leakage and evidence of past leakage during the ASME XI Class 1 System Leakage Test (IWB-5221).
l "The RCS vent and drain connections will also be visually examined with the isolation valves in the normally closed position during the 10-year ISI pressure test (IWB 5222 and Code Case N-498-1). This examination will be performed with the RCS at nominal operating pressure and at near operating temperature after satisfying i
the required 4-hour hold time."
Licensee's Basis for Requesting Relief (as stated)-
"These connections are equipped with manual valves which provide for double isolation of the reactor coolant system (RCS) pressure boundary. These valves are generally maintained closed during all modes of operation and the piping outboard of the first isolation valve is, therefore, not normally pressurized. The proposed alternative provides an acceptable level of safety and quality based on the following.
1.
ASME Section XI Code, paragraph IWA-4400, provides the requirements for hydrostatic pressure testing of piping and components after repairs by welding to the pressure boundary. lWA 4400(b)(5) excludes component connections, piping, and associated valves that are 1 inch nominal pipe size and smaller from the hydrostatic pressure test requirement after welded repairs. Therefore, requiring a hydrostatic test and visual examination of these s1 inch diameter RCS vent / drain connections once each 10-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME XI Code.
2.
The non-isolable portion of the RCS vent and drain connections will be pressurized and visually examined as required. Only the isolable portion of the vent and drain connections is not pressurized.
3.
A typical vent / drain connection includes two manual valves separated by a short j
pipe nipple which is connected to the RCS via another short pipe nipple and a j
half coupling. All connecticns are typically socket-welded and the welds j
received a surface examination after installation. The piping and valves are nominally heavy wall (Sch.160 pipe and 6000# valve bodies). The vents and 39
drains are not subjected to high stresses or cyclic loads, and the design ratings are significantly greater than RCS operating or design pressure.
4.
The Technical Specifications (TS) require RCS leakage monitoring during normal operation. Should any of the TS limits be exceeded, then appropriate corrective actions, which may include shutting the plant down, are required to identify the source of the leakage and restore the RCS boundary integrity.
" Additionally, SNC believes that there are also potential personnel safety and ALARA issues associated with pressurizing these connections. These issues are as follows:
1.
ASME Code Case N-4981 is currently used at FNP to perform this test.
Pressure testing these connections to the outboard valve requires the inboard isolation valves to be opened and subjects the valves and piping to RCS nominal operating pressure and near operating temperature. Opening the inboard valve at these conditions is contradictory to the requirement for double isolation of the RCS and thus creates the possibility for safety concerns for personnel performing visual examination of the connections.
l 2.
Performing the test with the inboard valves open requires several man-hours to position the valves for the test and then to restore them after the test is complete. All of these valves are located in close proximity of the RCS main loop piping thus requiring personnel entry into high radiation areas within the i
containment. Based on previous outage data it is estimated that dose associated with valve alignment and realignment would be approximately 1.2 man-Rem per test.
3.
Since this test would be performed near the end of an outage, when all RCS work has been completed, the time required to open and then close these vent / drain valves could impact the outage schedule."
Justification
" Requiring a hydrostatic test and visual examination of these s 1 inch diameter RCS vent / drain connections once each 10-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME XI Code. The added radiation exposure and potential for outage impact associated with opening the valves'is not considered justifiable, since the proposed alternative visual examinations (in conjunction with the TS monitoring requirements for RCS leakage) should provide assurance that the RCS pressure boundary, associated with these connections, is bein0 maintained at an acceptable level of quality and safety. Denial of this relief request results in the potential for outage schedule impacts and radiation exposure without a compensating increase in the level of quality and safety; therefore, the proposed alternative should be granted pursuant to the j
requirements of 10CFR50.55a(a)(3)(ii)."
40
9 Evaluation-The Code requires a system hydrostatic test, once per interval, of all Class 1 components within the RCS system boundary. The licensee has requested relief from hydrostatic test requirements for the subject line segments. The hne segments include two manually operated valves separated by a short pipe nipple that is connected to the RCS via another short pipe nipple and half coupling. This line configuration provides double isolation of the RCS system. Under normal plant operating conditions the subject line segments would see RCS temperatures and pressures only if leakby occurs from the inboard valve. To perform the Code-required test it would be necessary to manually open the inboard valves to pressurize the line segments. Pressurization by this method would defeat the RCS double isolation and may cause safety hazards for the personnel performing the examination duties. Typically these lines / valves are close to the primary and secondary RCS piping. Manual actuation (opening and closing) of these valves is estimated to expose plant personnel to 1.2 man-Rem per test. Therefore, the Code-required system hydrostatic test on these line segments presents a hardship on the licensee. The licensee's proposed alternative is to visually examine the isolation valves in the normally closed position each refueling outage for leaks and evidence of past leakage during the system leakage test. Also, the RCS vent and drain connections will be visually examined with the isolation valves in the normally closed position during the 10-year ISI pressure test. Visual examination of the isolation valves in the closed position each refueling outage and during the 10-year pressure test will provide reasonable assurance that structuralintegrity is maintained on the subject line segments.
Conclusion-Performing the Code-required system hydrostatic test on the subject line segments at Farley Units 1 and 2 would cause a significant burden that would not be compensated by a significant increase in quality and safety. Furthermore, the hcensee's proposed alternative will provide reasonable assurance that the subject line segments structuralintegrity will be maintained. Therefore, it is recommended that relief be granted pursuant to 10CFR50.55a(a)(3)(ii).
3.4.2 Class 2 System Pressure Tests 3.4.2.1 Request for Relief No. RR-24, Examination Category C-H, Hydrostatic Testing of Charging Pump Suction Piping From the CVCS Boric Acid Blender, Boric Acid Filter, and Chemical Mixing Tank Code Requirement-ASME Section XI, IWC-2500-1, Examination Category C-H requires pressure testing of Class 2 components in accordance with IWC-5221 and IWC-5222.
Licensee's Proposed A/ternative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed performing VT-2 visual examinations, once each inspection period at normal operating pressure, on the following components.
Unit 1 1.
2 inch diameter charging pump suction piping from the boric acid blender between check valve 01E21V211 and locked closed Class 2/ Class 3 interface valve 01E21V212.
41
2.
2 inch diameter charging pump suction piping from the CVCS boric acid filter between check valve 01E21V210 and normally closed Class 2/ Class 3 interface valve 01E21V264.
3.
1 inch diameter charging pump suction piping from the chemical mixing tank between check valve 01E21V187 and normally closed Class 2/ Class 3 interface valve 01E21V186.
I Unit 2 1.
2 inch diameter charging pump suction piping from the boric acid blender between check valve 02E21V211 and locked closed Class 2/ Class 3 interface valve O2E21V212.
2.
2 inch diameter charging pump suction piping from the CVCS boric acid filter between check valve 02E21V210 and normally closed Class 2/ Class 3 interface valve O2E21V264.
3.
1 inch diameter charging pump suction piping from the chemical mixing tank between check valve 02E21V187 and normally closed Class 2/ Class 3 interface valve O2E21V186.
The licensee stated:
"The subject piping segments will be visually examined, VT-2, once each inspection period in conjunction with the CVCS pressure tests while at normal operating pressure with the valves in their normal position."
Licensee's Basis for Requesting Relief (as stated)-
"These short piping segments (generally 12" in length or less) are not pressurized during normal modes of system operation. Additionally, there are no test connections provided between the check valves and the normally closed valves to allow for pressure testing using a test pump. Therefore, there is no practical method of pressurizing the piping segments without disassembly of the check valves (to q
remove the disc) or without modification of the system piping to provide test connections."
Justification "These segments of piping represent an extremely small portion of the charging pump suction lines subject to pressure tests. The proposed alternate examination will provide reasonable assurance that structural integrity will be maintained.
Pressure testing these piping segments would require either component disassembly or a system modification, which for these extremely short segments of piping,is considered an undue burden without a compensating increase in the level of quality and safety. Therefore the proposed alternative should be granted pursuant to 10CFR50.55a(a)(3)(ii)."
42
i 9
Eva/uotion-Orawings submitted by the licensee show that the design of the suction piping incorporates check valves that prevent flow from the test fill point to the specified boundary valves. The position of the check valves, therefore, precludes pressurization of these portions of piping to Class 2 requirements. These portions of 1-inch and 2-inch diameter piping would require either component disassembly or a system modification to perform the Code required hydrostatic test. The increase in plant safety would not.
compensate for the burden placed on the licensee by imposition of this requirement.
The licensee has stated that a visual examination (VT-2) will be performed on the subject piping segments once each inspection period in conjunction with the CVCS l
pressure tests while at normal operating pressure with the valves in their normal position.
I This will provide reasonable assurance of continued inservice structuralintegrity of the subject piping segments.
Conc /usion-The hydrostatic test required by the Code for the subject Class 2 lines presents a burden for Farley, Units 1 and 2, because of the position of the check valves.
Imposition of the requirements on Southern Nuclear Operating Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed visual examination during the pressure tests at normal operating pressure. This examination will provide reasonable assurance that the structuralintegrity of these lines is maintained. Therefore,it is recommended that relief be granted pursuant to 10CFR50.55a(a)(3)(ii).
3.4.2.2 Request for Relief No. RR-28, Use of Code Case N-522, Pressure Testing of Containment Penetration Piping Code Requirement-ASME Section XI, Table IWC-2500-1. Category C-H, requires pressure testing of all Class 2 pressure-retaining components in accordance with IWC-5221 and IWC-5222.
Licensee's Proposed Allemative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the alternative requirements contained in Code Case N-522, Pressure Testing of Containment Penetration Piping. The licensee stated:
"Farley Nuclear Plant will perform Appendix J testing on the penetrations listed in Table 1 commensurate with the alternatives included in Code Case N-522. Leakrate testing of the subject penetrations will be performed at peak calculated containment pressure in accordance with 10CFR50 Appendix J, Option B. The tests will be scheduled per Appendix J, Option B."
In the July 13,1998, submittal, the licensee verified that the tests will be performed at peak calculated containment pressure and that the test procedures willinclude methods for detection and location of throughwallleakage in containment isolation valves and pipe segments between the CIVs.
The subject penetrations are:
43
9 Penetration Function 12 Containment Cooling & Purge 13 Containment Cooling & Purge 29 Si Accumulator to RWST 31 Waste Processing
[
33 Containment Building Sump 47 Service Air 48 Instrument Air 49 SI Accumulator Fill Line 54 Containment Cooling & Purge
[
55' Containment Cooling & Purge 61A Containment Combustible Gas Control 618 Containment Combustible Gas Control 62 Waste Processing 63 N2 Supply to SI Accumulator 64A Reactor Coolant System i
66 Containment Combustible Gas Control 67 Containment Combustible Gas Control 70 Containment Cooling & Purge 71 Containment Leak Rate Test Connection 72 Containment Leak Rate Test Connection 78 Containment Building Sump j
79 Service Air (Unit 2) 82 Demineralized Water Supply 95 Spent Fuel Pool Cooling 97B instrument Air 103 Containment Combustible Gas Control f
Licensee's Basis for Requesting Relief (as stated)-
"These line segments are safety-related and subsequently included in the scope of
[
ASME Section XI only because they function as part of the containment boundary l
{
44
- ~.
-.. - - - - -. - _ - ~ -.
- - - - - -. - _.. - ~. - ~ _ _ - _ - _ -
1 l
l and are relied upon for containment integrity. Therefore,it is logical to test these i
segments using containment test criteria in lieu of the Section XI pressure testing requirements. Code Case N 522 allows Appendix J testing of the subject penetrations as an alternative to the Code-required Category C-H pressure tests."
Justification "The ASME XI Code Committee evaluated the proposed alternative testing requirements included in Code Case N-522 and determined that the alternatives are acceptable for ensuring the integrity of the subject Class 2 penetrations. Pursuant to the requirements of 10CFR50.55a(a)(3)(i), the implementation of the alternative testing ensures an acceptable level of quality and safety and does not decrease the margin of public health and safety. Additionally, implementation of the alternative examinations will reduce costs, personnel radiation exposure, and outage time."
Evaluation-The Code requires that a VT-2 visual examination be performed during system pressure testing for all Class 2 pressure-retaining piping, including those segments that penetrate primary containment. As an alternative, the licensee proposed to implement the requirements of Code Case N-522, Pressure Testing o/ Containment l
Penetration Piping. Code Case N-522 specifies that 10 CFR 50, Appendix J testing may be used as an alternative to Section XI pressure tests, for certain containment i
l penetration piping, i
The subject piping is fabricated and designated as Class 2 only because it penetrates i
the primary reactor containment and is considered an extension of the containment vessel. Since the piping on either side of these penetrations is non-classed, the requirements of Appendix J are more appropriate than those of Examination Category C-H. Appendix J pressure tests verify the leak tight integrity of the primary reactor containment and of systems and components that penetrate containment by localleak rate and integrated leak rate tests.
The Class 2 containment isolation valves (CIVs) and connecting pipe segments must withstand the peak calculated containment internal pressure. The INEEL staff believes that the pressure retaining integrity of the CIV's and connecting piping and their associated safety functions may be verified with an Appendix J test conducted at the peak calculated containment pressure. The licensee has proposed to perform the Appendix J testing at no less than the peak calculated containment pressure and will use procedures and techniques capable of detecting and locating through-wallleakage in the l
pipe segments between the CIV's.
Appendix J, Option A-Prescriptive Requirements, requires that three Type A tests be performed at approximately equalintervals during the 10 year ISIinterval, with the third test being done while shutdown for the 10-year plant ISI. Option A also requires Type B and C tests be performed during each refueling outage, but in no case at intervals greater than 2 years. This is more frequent than the periodic pressure tests required by ASME l
Section XI.
45 i
s
Appendix J, Option B-Performance Based Requirements, allows a licensee to perform Type A, B, and C tests at frequencies related to the safety significance and historical performance of the system's isolation capabilities. This could, in effect, allow only one test to be performed during the 10-year ISI interval. However, the staff's position, as stated in Regulatory Guide 1.163 Performance-based Contamment Leak-Test Program l
(Reference 16), is that the licensee is to establish test intervals of no greater than 60 l
months for Type C tests because of uncertainties (particularly unquantified leakage rates for test f ailures, repetitive / common mode failures, and aging ef fects) in historical Type C component performance data. While this five-year limit results in an increased time between testing over that required by Section XI (forty months). it is believed that Appendix J tests are more appropriate and provide reasonable assurance of the continued operability of containment penetrations. Therefore, the INEEL staff believes that the test frequencies associated with Appendix J, Option A (Type A, B or C) or Option B (Type C) Tests are commensurate with the Code-required pressure test frequencies.
The licensee has stated that the tests will be scheduled per Appendix J. Option B.
Therefore, the INEEL staff recognizes that the licensee test frequencies may exceed the 60 month interval for a Type C test as stated in Regulatory Guide 1.163. Based on the above analysis and information submitted, and since the licensee has not defined a specific test frequency for the Option B examinations, the INEEL staff believes that an acceptable level of quality and safety will be provided for a period of no greater than 60 months.
Conclusion-The licensee proposed the use of Code Case N-522 Based on the above analysis and information submitted, the INEEL staff believes that an acceptable level of quality and safety will be provided by the licensee's proposed alternative for a period of 60 months. Therefore, it is recommended that the licensee's proposed alternative to
)
implement Code Case N-522 be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for a limited time not to exceed 60 months from the start date of the third 10-year interval at Farley Unit 1, and from the start date of the Farley Unit 2, Updated interval.
3.4.2.3 Request for Relief No. RR-30, Examination Category CH, IWC 5222, Pressure Testing of Safety injection System Piping Segments Which Are Nonisolable From Class 1 Piping Code Requirement-ASME XI, Table IWC-2500-1, Category C-H requires pressure testing of Class 2 pressure-retaining components in accordance with IWC-5221 and IWC-5222.
Licensee's Code Re//e/ Request-Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee requested relief from performing the required hydrostatic pressure test at the required pressure for the following piping segments.
46
_______-.-------J
Line Number Descriotion 2" CCB-30 Between motor operated Valve Q1E21V068 (O2E21V068 for Unit 2) 3" CCB-30 and check valves 01E21V078A (02E21V078A for Unit 2), B & C (Hot Leg Safety injection) 2" CCB-31 Between motor operated Valve 01E21V072 (02E21V072 for Unit 2) 3" CCB-31 and check valves 01E21V079A (O2E21V079A for Unit 2), B & C (Hot Leg Safety injection) 2" CCB-22 Between motor operated Valve 01E21V063 (02E21V063 for Unit 2) 3" CCB 22 and check Valves 01E21V066A (02E21V066A for Unit 2), B & C (Cold Leg Safety injection) 2" CCB-21 Between motor operated Valve 01E21V016 (02E21V016 for Unit 2) 3" CCB-21 A & B and check Valves 01E21V062A (02E21V062A for Unit 2), B
& C (Cold Leg Safety injection) 6" CCB-29 Between motor operated Valve 01E11V044 (O2E11V044 for Unit 2) 10" CCB-29 and check Valves 01E21V076 (02E21V076 for Unit 2) A & B (RHR l
Hot Leg injection)
Licensee's Basis for Requesting Relief (as stated)-
"The system design did not include provisions for isolating these Class 2 piping segments from the Class 1 system to support Code-required pressure testing. There is no practical method of pressure testing the subject piping segments, to the Class 2 pressure test requirements, without also pressurizing the associated Class 1 reactor coolant system (RCS) piping and components to the same pressure. The boundary valves between the Class 1 and Class 2 piping (class break valves) consist of check valves for each of the subject segments. Therefore, these check valves allow injection to the RCS, but provide no pressure boundary between Class 2 and Class 1. The safety injection lines are provided with manual globe valves immediately upstream of the class break check valves. These valves are utilized for flow balancing of the safety injection flow rate to each RCS leg and are locked in a throttled position. Closing the valves to provide a boundary for pressure testing would require additional performance of the safety injection flow balance test which could significantly impact the outage duration. The RHR Hot Leg injection lines are not provided with any valves upstream of the check valves that can be used to l
establish the Class 2 pressure test boundary.
"The only practical method to pressure test these piping segments consists of i
performing a test concurrent with the Class 2 system 10-year pressure test. With i
the Class 1 system at nominal operating pressure (approximately 2235 psig), these Class 2 piping segments ccn be pressurized, using a test pump or jumpers, to approximately the same pressure while ensuring that the boundary check valves remain closed."
I 47
Justificatior, "There are no system components that can be used to provide a test boundary between these Class 2 piping segments and the Class 1 RCS. Therefore, pursuant to 10CFR50.55a(g)(6)(i), there are no prcctical means of performing the Code required pressure tests without modifying the existing systems or subjecting the entire Class 1 RCS to the higher Class 2 pressure. The proposed alternative pressure test < nd visual examinations should provide adequate assurance of the pressure boundary integrity of the subject Class 2 piping segments."
Licensee's Proposed Attemative Examination (as stated)-
"The Class 2 piping segments wis be included within the VT-2 visua! examination boundary during the Class 1 system leakage test (IWB-5221) each refueling outage.
" Additionally, once each inspection interval, the piping segments will be pressurized to approxirrately RCS nominal operating pressure in conjunction with the Class 1 10-year interval pressure test. The Class 1 10-year pressure test is performed in conjunction w.th Relief Request RR 21 which utilizes ASME Code Case N-498-1.
The piping will be VT 2 visualinspected while the piping is at approximately RCS nominal operating pressure after the required 4-hour hold time."
Evaluation-The designs of the subject systems do not provide adequate shutoff boundaries to prevent over pressurization of the adjacent Class 1 piping. The design of these lines makes the Code-required hydrostatic tests impractical. To perform the examinations in accordance with the Code, the piping and valves must be modified. The increase in plant safety would not compensate for the burden placed on the licensee by imposition of the requirement.
The licensee's proposed alternative is to include the subject Class 2 pipireg segments l
in the VT-2 visua! examination boundary during tne Class 1 system leakage test each refueling outage. In addition, the licensee will pressurize the piping segments to approximately RCS nominal operating pressure once each inspection interval in conjunction with the Class 1 10 year interval preswre test. The proposed alternative will provide reasonable assurance that unallowable inservice f!aws have not developed in the subject portions of *.he piping.
Conclusion-The hydrostatic teat required by the Code for the subject Class 2 piping segments is impractical at Farley Units 1 and 2 br-,se the lines are unisolatable from the associated Class 1 lines. Imposition of the Code requirement on Southern Nuclear Operating Company would cause a considerable burden without an increase in safety
[
significantly above that provided by the proposed testing. The licensee's proposed alternative will provide reaso.~.able assurance of the continued structural integrity of the L
subject piping segments. Therefore,it is recommended that relief be granted pursuant to 10CFR50.55a(g)'5)(i).
3.4.3 Class 3 System Pressure Tests 48
'4 3.4.3.1 Request for Relief No. RR-25, Class 3,IWA 5244(b) Buried Portions of Service Water System Piping Code Requirement-ASME Section XI, Paragraph IWA-5244(b) requires that in redundant systems where the buried components are nonisolable, the visual examination (VT-2) shall consist of a test that determines the change in flow between the ends of the buried components, in cases where an annulus surrounds the buried components, the areas at each end of the buried components shall be visually examined for evidence of leakage in lieu of the flow test.
Licensee's Proposed Alternative--Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed the use of a visual examination of the ground surface in lieu of the flow test as required by IWA-5244(b) for the buried portions of piping in the Service Water System.
The licensee stated:
"A visual inspection of the ground surface above the buried piping will be conducted once per period while the system is in operation. This visualinspection will identify any areas that appear to be abnormally wet. Any areas reported to be abnormally wet will be evaluated to determine the need for additional examinations and corrective action."
Licensee's Basis for Requesting Relief (as stated)-
"The Service Water System contains several hundred feet of buried piping extending from the pond intake structure to the auxiliary and turbine buildings. The system is comprised of two primary supply trains which branch to provide cooling water to numerous safety related loads located throughout the plant. While each train supply header is provided with instrumentation that can measure the flow rate prior to entering the underground piping, the downstream branch lines are not provided with sufficient flow instrumentation to allow the required comparison of flow. Therefore, a test that determines the change in flow in either train is not practical."
l Justification "The Service Water System is in operation during all safety related modes of plant operation. The performance of numerous components which are provided with l
cooling water from the Service Water System are routinely monitored by operations personnel. Any degradation in the performance of the supplied components would result in actions to determine the cause.
"A system leak in the buned supply piping of sufficient size to affect componett l'
would be detected during the proposed alternative examinations. Visible evidence of cooling performance would result in visible conditions in the surface areas which leakage in the surface areas above the buried piping would result in appropriate corrective actions to resolve the problem. This examination will provide reasonable assurance of the continued inservice integrity of the buried piping.
- lmplementation of the Code requirement at Farley Nuclear Plant would either require installation of instrumentation or design modifications to allow access to the 49
e piping. Implementation of these changes would be an undue burden that would not be compensated by a significant increase in safety above that provided by the proposed alternative. Public health and safety will not be endangered by allowing the proposed alternative to be performed in lieu of the Code requirement, therefore, it is requested that the proposed alternative be authonzed pursuant to 10CFR50.55a(a)(3)(ii)."
l l
Evaluation--Portions of the Class 3 piping in the service water system are buried. The design of the service water system does not provide access to the portions of buried piping to perform the Code-required visual examination during hydrostatic testing. The system design, therefore, makes the Code-required visual examination a burden to perform. In order to oerform the Code-required visual examination in accordance with the requirements, the service water system would require installation of instrumentation or design modifications to allow access to the piping. TSe increase in plant safety would not compensate for the burden placed on the licensee that would result from imposition of the requirement.
The licensee's proposed alternative examination is to perform a visualinspection of the ground surface above the buried piping once per period while the system is in operation The visualinspection will be used to identify any areas that appear abnormally wet. This examination will provide reasonable assurance of continued inservice structuralintegrity of the buried piping.
Conclusion-The VT-2 visual examination required by Section XI of the ASME Code during hydrostatic testing of the subject Class 3 service water system piping is a burden to perform at Farley, Unit 1 and 2, because the piping is buried with no access for visual examinations referenced in IWA-5244. Imposition of the requirement on Southern Nuclear Operating Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed alternative.
The proposed alternative will provide reasonable assurance that structuralintegrity of the Class 3 piping is maintained. Therefore, it is recommended that relief be authorized pursuant to 10CFR50.55a(a)(3)(ii).
3.4.3.2 Request for Relief No. RR-29, IWA-5244(b), (c), Concrete Encased Portions of Spent Fuel Pool Cooling System Piping Adjacent to Spent Fuel Pit Code Requirement-ASME XI, paragraph IWA-5244(b) requires that in redundant systems where buried components are nonisolable, that the visual examination VT-2 shall consist of a test that determines the change in flow between the ends of the buried components. IWA-5244(c) requires that in nonredundant systems where the buried components are nonisolable, such as return lines to the heat sink, the VT-2 shall consist only of verification that the flow during operation is not impaired.
Li ensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)ni), the licensee pronosed to visually examine the piping sections adjacent to the concrete emsed p; ping section in lieu of the flow test as required by IWA4244(b) and (c). The licensee stated:
'0 5
l J
f I
"The piping sections immediately adjacent to the concreted encased sections will be examined to determine any evidence of material degradation or potentialleakage during system inservice pressure tests, each examination period."
Licensee's Basis for Requesting Relief (as stated)-
" Portions of the subject pipe lines adjacent to the Spent Fuel Pool were encased in concrete during plant construction without an annulus to facilitate visual examinations. The two 10" lines, HCC-105 and HCC-108, are suction lines from the Spent Fuel Pit and the 8" line, HCC-107, is the common return line from the i
Spent Fuel Pool Heat Exchangers that discharges below the water levelin the Spent i
l Fuel Pool. All three piping segments normally experience only low pressures and relatively low temperatures. The design did not include any flow measuring instrumentation on the suction piping or any means of observing flow discharging i
from the return line; therefore, a costly design modification would be required to i
meet the Code requirements."
l Justification
"..The proposed alternative visual examinations should provide adequate assurance
)
of the pressure boundary integrity of these low pressure concrete encased piping sections. An acceptable level of quality and safety will continue to be maintained and public health and safety will not be endangered by approving this relief request; therefore, approval should be granted pursuant to 10CFR50.55a(6)(3)(ii)."
Evaluation-The subject piping in the spent fuel pool cooling system cannot be visually examined because it was encased in concrete during plant construction. The design of the spent fuel poo. :ystem does not include any flow measuring instrumentation on the suction piping, or any means of observing flow discharging from the return line.
Therefore, the design of the system makes the Code requirements difficult to perform.-
In order for the licensee to meet the requirements established by the Code, desiga modifications would be required. The increase in plant safety would not compensate for the burden placed on the licensee that would result from imposition of the requirement.
The licensee will visually examine, each examination period, the piping sections immed.ately adjacent to the concrete encased sections to determine any evidence of
{
material degradation or potentialleakage during system inservice pressure tests. The l
licensee's performance of this examination will provide reasonable assurance that
{
structuralintegrity of the encased piping is maintained.
i Conclusion'-The visual examination requirements as specified in the Code for the subject Class 3 spent fuel pool cooling piping presents a burcen at Farley Units 1 and 2, because the piping is encased in concrete. Imposition of the requirement on the Southern Nuclear Operating Comnany would cause a burden that would not be compensated significantly by an increase in safety above that provided by the licensee's proposed alternative to perform a visual examination on the piping sections immediately adjacent to the concrete encased sections on a periodic basis. The proposed alternative will provide reasonable assurance that structuralintegrity of the Class 3 piping is maintained. Therefore,it is recommended that relief be authorized pursuant to 10CFR50.55a(a)(3)(ii).
51
3.4.4 General 3.4.4.1 Request for Relief No. RR-23, Revision 1, IWA 5250(a)(2) Corrective Measures for Bolted Connections Code Requirement-Section XI. lWA-5250(a)(2) requires that if leakage occurs at a bolted connection in ASME Section XI components, the bolting shall be removed, VT-3 examined for corrosion, and evaluated in accordance with IWA-3100.
Licensee's Proposed A/remative-Pursuant to 10 CFR 50.55ata)(3)(i), the licensee proposed alternative corrective actions for leakage at bolted connections in lieu of the requirements defined in IWA-5250(a)(2). The licensee stated:
"When leakage is identified at bolted connections by visual, VT-2 examination during system pressure testing, an evaluation will be performed to determine a.e l
susceptibility of the botting to corrosion and to assess the potential for failure. The evaluation will, at a minimum, consider the following factors:
1.
Bolting materials 2.
Corrosiveness of process fluid leaking 3.
Leakage location 4.
Leakage history at connection or other system components 5,
Visual evidence of corrosion at connection (while connection is assembled) 6.
Service age of bolting materials "When the pressure test is performed on a system that is in service or that Technical Specifications require to be operable, and the bolting is susceptible to corrosion, the evaluation shall address the connection's structural integrity until the next component / system outage of sufficient duration. If the evaluations conclude that the system can perform its intended safety related function, the bolt with the most apparent degradation will be removed (if necessary to assess the condition of the bolt) and a visual VT-1 examination will be performed when the system / component is taken out of service for sufficient duration to accomplish other system maintenance activities (and whon the component is not required to be operable per Technical Specifications).
"For bolting that is susceptible to corrosion, and when the initial evaluation indicates that the component cannot, with a reasonable degree of assurance, perform its safety ~related function until the next component / system outage of sufficient duration, the bolt with the most apparent degradation will be reme /ed (if necessary to assess the condition of the bolt) and a visual, VT-1 examination will be performed.
"When a bolt is removed for a VT-1 examination, the results will be evaluated in accordance with IWA 3100(a) as follows:
l l
52 l
l.
!L_._._.
l
" Class 1 Acceptance Standards (IWB-3517) will be used to determine the acceptability of Class 1,2 and 3 bolting (excluding the evidence of coolant leakage near bolting)."
Licensee's Basis for Requesting Relief (as stated)-
" Removal of pressure retaining bolting at mechanical connections for visual, VT-3 examinations may not be the most prudent course of action when leakage is detected. The Code requirement to remove, examine, and evaluated bolting in this situation does not allow Southern Nuclear Operating Company the option of considering other factors which may indicated the acceptability of the mechanical joint. Other factors that may be considered when evaluating bolting acceptability include, but are not limited to: bolting materials and service age, corrosiveness of leaking fluid, leakage location, leakage history, and extent corrosion. Satisfying Code requirements without regard to these factors may:
Increase the radiological dose to workers for leaks that are not a challenge to operational nor structural limits.
Unnecessarily cause the plant a delay in startup.
Unnecesserily require the plant to be shutdown to remove bolting on unisolable components."
Justification "The proposed alternative is a conservative and technically sound engineering approach that will provide an acceptable level of quality and safety for leaking bolted connections. Therefore,it is requested that the proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(i)."
Evaluation-in accordance with IWA 5250(a)(2), if leakage occurs at a bolted connection, the bolting must be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100. In lieu of this requirement, the licensee has proposed to evaluate the bolting to determine its susceptibility to corrosion. The proposed evaluation will consider, as a minimum, the bolting material, corrosive nature of l
the process fluid, leakage location and history, service age of the holting materials, and visual evidence of corrosion at the assembled connection. Based on the items included in the evaluation process, the INEEL staff believes that the licensee's alternative is a sound engineeri.9 approach. In addition,if the initial evaluation indicates the need for a more detailed analysis, the bolt with the most apparent degradation will be removed, VT-l 1 visually examined, and eve'uated in accordance with IWA 3100(a). The VT-1 i
examination critetia are more stringent than the simple corrosion evaluation described in IWA-5250. For these reasons, raasonable assurance of the operational readiness of the bolted connection will be provided.
Conclusion-The INEEL staff concludes that the licensee's proposed alternative to the requirements of IWA-5250(a)(2)is a conservative and technically sound engineering approach. As a result, significant patterns of degradution will be detected and an acceptable level of quality and safety will be provided. Therefore, it is recommended 53
that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
3,5 General 3.5.1 Ultrasonic Examination Techniques 3.5.1.1 Request for Relief No. RR 1, Material Requirements for Calibration Blocks Used For Ultrasonic Examination of Heavy Wall Vessels Code Requirement-Section XI,1989 Edition, Article 1-2000, paragraph I-2100 requires that ultrasonic examinations of vessel welds in ferritic materials greater than 2-inches thick be performed in accordance with Article 4,Section V. Paragraph T-441.1.2.1, Article 4, requires that the material from which calibration blocks are fabricated be of the same material specification, product form, and heat treatment as one of the materials being joined, Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed the use of calibration blocks that comply with the material requirements of the 1974 edition of ASME Section V. The licensee stated:
l
" Ultrasonic examination will be conducted using the above calibration blocks which l
comply with the material requirements of paragraph T-533, Article 5,Section V of l
the 1974 ASME Code. This paragraph requires the block to be of similar metallurgical structure and the same or an equivalent P-number grouping as the finished component."
The following calibration blocks will be used:
I l
Calibration Block #
Examinations APR-6 Steam generator channel head-to-tubesheet welds ALA-RV-1 Reactor vessel lower head-to-lower shell weld and all lower head welds ALA-RV-3 Reactor vessel top head welds ALA-RV-5 Reactor vessel flange-to-shell weld, vessel to shell welds, and nozzle l
inside radius i
licensee's Basis for Requesting Relief (as stated)-
i "During fabrication of the Farley Unit-1 (Unit-2) reactor pressure vessel and i
steam generators, the ultrasonic calibration blocks used to perform 1
examinations by the vessel manufacturer were fabricated to the requirements of i
A5ME Section Ill. When ASME Section XI was issued for inservice inspection, 54 4
4
the existing Section ill blocks were unacceptable for use. The blocks were refabricated to the applicable Section XI requirements, but materials from the vessel were no longer available and similar materials, SA-336 clad and SA-508 clad, were substituted."
lustification
" Ultrasonic examinations for preservice and almost two 10-year intervals of plant operation have been conducted using these calibration blocks. The use of new calibration blocks would not be consistent with the historical data. Ultrasonic examinations using these existing calibration blocks have proven to be adequate and will continue to assure that an acceptable level of safety and quality is being maintained. Therefore, relief should be granted pursuant to the requirements 10 CFR 50.55a(a)(3)(i)."
j Evaluation-In a submittal to the NRC on October 5,1989 (Reference 13) the licensee provided evaluation results to demonstrate the suitability of the alternative calibration block materials. The following is a summary of the results of the evaluation:
APR6: The SA-336 Cl F-1 calibration block and the SA-216 Gr. WCC ar'd SA 508 Cl. 2 or 3 materials examined have the same product form, are similar in chemistry, -
and have received equivalent postweld heat treatments and are classified as P-3.
ALA-RV-1: The SA-508 Cl. 2 calibration block and the SA-533 Gr. B material examined are similar in chemistry, and have received equivalent postweld heat treatments, and are classified as P-3.
ALA RV-3 and ALA-RV-5: The SA-508 Cl. 2 calibration blocks and the SA-533 Gr. B Cl. 2 materials examined are similar in chemistry, and have received equivalent postweld heat treatments and are both classified as ?-3.
Factors essential for acoustical compatibility were also considered in the evaluation. The material examined and the calibration blocks' material were found to be of similar nature to ensure acoustical compatibility. The velocity and attenuation differences among the materials were immeasurable and were determined to be equivalent.
While the Code requirements have not been explicitly met, the intent of the Code-to ensure acoustic compatibility has been met. The calibration blocks in question have been und for examinations since the plant was built. Therefore, continued use of these calibration blocks will provide consistent examination results. The licensee has demonstrated that the calibration blocks are equivalent in acoustic nature to the components being examined and that the alternative proposed would provide an acceptable level of quality and safety.
55
Conchdion-An acceptable level of quality and safety will be maintained by using the subject calibration blocks. Therefore,it is recommended that the proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(i).
3.5.1.2 Request for Relief No. RR 2, Notch Location Requirements for Calibration Blocks Used for Ultrasonic Examination of Heavy Wall Vessels s
Code Requirement-Section XI,1989 Edition, Article 1-2000, paragraph 1-2100 requires that ultrasonic examination of vessel welds in ferritic materials greater than 2-inches thick (t) be performed in accordance with Article 4,Section V. Figure T-441.1, Article 4, requires that notches be 2% t deep with a tolerance of + 10% -20% and located a 4
minimum of 3 inches from the edge and t/2 from the end of the block.
l Licensee's Proposed Alternative--Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed the use of calibration blocks with minor variations in dimensional requirements.
The licensee stated:
" Ultrasonic examinations will be conducted using the above calibration blocks which contain minor variations in 1989 Code specified dimensions."
The following calibration blocks will be used:
Calibration Block #
Examinations APR-7 Pressurizer welds ALA-RV 1 Reactor vessellower head to-lower shell weld and alllower head welds ALA RV-3 Reactor vessel top head welds ALA-RV-5 Reactor vessel flange to-shell weld, vessel to shell welds, and nozzle inside radius Licensee's Basis for Requesting Relief (as stated)-
" Calibration blocks ALA-RV-1, ALA-RV-3, and ALA-RV 5 have notches located 2-inches from the edge of the blocks. The ID notch in calibration block APR 7 is 0.033; inch deeper than required and the ID notch in ALA-HV-5 is 0.010-inch deeper than required. The ID notch in ALA-RV-5 is also 0.39-inch less than the required t/2 distance tom the end of the block."
Justification
" Figure T-542.2.1, Article 5,Section V,1989 Edition defines requirements for a similar calibration block; however, the clearance required from the ends of the 2-percent notch is 2-inches instead of the 3-inches required by Figure T-441.1, Article
- 4. The figure also requires that notches be t/2 from the end of the block. The 56 l
O distance of these notches from the block edge /end have been found to have no detrimental effect on calibrations. Any inaccuracy in calibration caused by the slight variation in depth of the notches is within the accuracy of the ultrasonic testing technique and is significantly less than the 2dB or 20% correction allowed on the distance-amplitude correction (DAC) curve, in order that future ultrasonic data may be compared with historical data, the existing calibration blocks should continue to be used. Ultrasonic examination using these calibrat on blocks has proven to be i
adequate and will continue to assure that an acceptable level of safety and quality is being maintained."
Eva/uation-The subje':t calibration blocks have been used for examinations since the plant was constructed. Therefore, continued use of these calibration blocks would provide con;istent examination results. The licensee has determined that the current notch locations have had no detrimental effect on calibrations. They have also determined that any inaccuracy in calibration caused by slight variations in notch depth is within the accuracy of the ultrasonic testing technique and is significantly less than the 2 dB or 20% correction allowed on the distance-amplitude correction (DAC) curve. The licensee has demonstrated that continued use of subject calibration blocks will provide an acceptable level of quality and safety.
Conclusion-The proposed alternative, to use the calibration blocks with minor variations i
to the 1989 Code requirements, will provide an acceptable level of quality and safety.
Therefore,it is recommended that the proposed attemative be authorized pursuant to 10CFR50.55a(a)(3)(i).
3.5.1.3 Request for Relief No. RR-3, Ho!e Location Requirements for Calibration Blocks Used for Ultrasonic Examination of Heavy Wall Vessels Code Requirement-Section XI,1989 Edition, Article 1-2000, Paragraph l-2100 requires
.that ultrasonic examination of vessel welds in ferritic materials greater than 2-inches thick be performed in accordance with A.'ticle 4,Section V. Figure T-441.1, Article 4, requires that aligned side-drilled holes be located a minimum distance of t/2 from the end of the block, where t is the block thickness. Nonaligned holes are required to be located a minimum of 1.5 inches from the end of the block. Side-drilled holes are required to be 1/4t,1/2t, and 3/4t 0.125-inch deep. Nonaligned holes are to have a minimum separation of t/4.
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee l.
proposed to use calibration blocks with variations in hole location, depth, and/or separation from the requirements for calibration blocks used to perform ultrasonic
}
examinations on the following vessels greater than 2-inch thickness:
i Calibtstion j
Block #
Examinations APR 5 Steam generator welds i
57 i'
..-.. - - - - - - - - -. - -. ~ -. -
l l
Calibratinn Block #
Examinations l
APR-6 Steam generator channel-head-to-tubesheet weld l
APR 7 Pressurizer welds AL A-RV-1 Reactor vessel lower head-to-lower sheli weld and alllower head welds l
ALA-RV-3 Reactor vessel top head welds t
- ALA-RV-5 Reactor vessel flange-to-shell weld, vessel to shell welds, and nozzle l
inside radius i
i The licensee stated:
" Ultrasonic examinations will be conducted using the above-mentioned calibration blocks,"
i Licensee's Basis for Requesting Relief (as stated)-
l
' The aligned holes in ALA-RV-5 are 0.39-inch less than the required distance of t/2 from the edge. The non-aligned holes in ALA RV 1, APR-5, APR-6, and AF R-7 are 0.25,0.065,0.05,0.103, and 0.04 inches deeper than required, ret pectively, in calibration block ALA-RV-5, the 1/4t and 1/2t holes have 0.45-inc7 less separation than required and the 1/4t and 3/4t holes have 0.20-inch I
less snaration than required."
Justification "The intent of the dimensional requirements is to eliminate interferences caused by other reflectors on the edge of the block. However, experience performing
- calibrations using these blocks has proven completely satisfactory. Continuing to use these blocks will permit the comparison of future data with historical data.
Ultrasonic examination using these calibration blocks has proven to be adequate and will continue to assure that an acceptable level of safety and quality is being maintained. Therefore, relief should be granted pursuant to the requirements of 10 CFR 50.55ata)(3)(i)."
Evaluation-The subject calibration blocks have been used for examinations since the plant was constructed. Therefore, continued use of these calibration blocks would
[.
provide consistent results as compared with those of previous examinations. The
}-
licensee, through successful calibrations using the subject calibration blocks, has demonstrated that the minor variations from the dimensional requirements of the hole locations have not diminished the ability to satisfy calibration requirements. Therefore, 4
j the continued use of the subject calibration blocks will provide an acceptable leve' of quality and safety.
f Conclusion-The proposed alternative, to use the calibration blocks with minor variations l
from the 1989 Code requirements, will provide an acceptable level of quality and safety.
f 58 i
i
.,u,_,
i Therefore, it is recommended that the proposed alternative be authorized pursuant to 10CFR50.55a(s)(3)(i).
3.5.1.4 Request for Relief No. RR-4, Dimensional Requirements for Notches Placed in Ultrasonic Calibration Blocks Code Requirement-Section XI,1989 Edition, Article 12000, Paragraph I 2200, Appendix
}
Ill, Paragraph 111-3430 requires that basic calibration blocks contain notches that are at least 1.0-inch long and 0.104t-0.009t + 10% -20% in depth where "t" is the block 2
j thickness.
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55ala)(3)(i), the licensee
)
proposed the use of ALA-21, ALA-23, ALA-26, and ALA-28 calibration blocks with minor variations in dimensional requirements. The licensee stated:
" Ultrasonic examinations will be conducted using the above calibration blocks which e
contain minor variations in 1989 Code specified dimensions."
{
Licensee's Basis for Requesting Relief (as stated)-
i "ALA-21 and ALA-23 have notch depths of 0.02t (2%t). TheI.D.
]
Circumferential and the O.D. axial notches in ALA-26 are 0.062-inch less than the required length. The I.D. axial notch in ALA-28 is 0.014-inch less than the 2
required depth and the two O.D. notches are 0.002-inch less than the required i
depth."
l 4
l Justification i
"These calibration blocks have been used throughout plant life. The use of smaller notches (than those required by Code) would only provide a more sensitive examination, and experience performing calibrations using these blocks has been completely satisfactory. Continuing to use these blocks will permit the comparison of future data with historical data and will continue to provide an acceptable level of quality and safety."
Evaluation-The subject calibration blocks contain notches that are smaller than those required by the Code. The undersize notches in the calibration blocks would provide a more sensitive examination. The subject calibration blocks have been in use since the plant was constructed. The continued use of these calibration blocks will provide a more consistent comparison with previous examinations. The licensee has also demonstrated the ability to soccessfully calibrate using the subject calibration blocks. Therefore, the continued use of the subject calibration blocks will provide an acceptable level of quality and safety.
Conc /usion-The proposed alternative, to use the calibration blocks with minor variations from the 1989 Code requirements, will meet or exceed the intent of the Code requirements as well as provide an acceptable level of quality and safety. Therefore,it is recommended that the proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(i).
59
l 3.5.1.5 Request for Relief No. RR-5, Curvature Differences Between Ultrasonic l
Calibration Blocks and the Components to be Examined l
Code Requirement-Section XI,1989 Edition, Article 12000. Paragraph I-2200, Appendix lil, Paragraph Ill 3410 requires that basic cal;bration blocks be make from material of the same nominal diameter as the pipe to be examined.
Licensee's Proposed Alternative--Pursuant to 10 CFR 50.55af a)(3)(i), the licensee i
proposed the use of calibration block ALA-25, which has minor variations in dimensional requirements. The licensee stated:
l
" Ultrasonic examinations will be conducted using the above calibration block which l
contains a minor variation to 1989 Code specified dimensions."
l Licensee's Basis for Requesting Re/ief (as stated)-
"ALA-25 is a 14-inch diameter calibration block which is used to examine the 14 inch diameter Feedwater line. The weld connecting the feedwater piping to l
the steam generator, is approximately 16-inches in diameter."
l l
Justification "This calibration block has been used throughout plant life. Curvature differences between ALA-25 and the welds to be examined are minor and experience i
performing calibrations using this block has been complete ly satisfactory.
l Continuing to use the block will permit the comparison of future data with historical L
data and will continue to provide an acceptable level of quality and safety.
Therefore, permission to use this calibration block should be granted pursuant to 10 CFR 50,55ata)(3)(i)."
Evaluation-The subject calibration blocks have been used for examinations since the plant was constructed. Therefore, continued uce of these calibration blocks will provide l
consistent results when compared with previous examinations. The licensee, thrcugh l
successful calibrations using the subject calibration blocks, has demonstrated that the diameter variation has not diminished the ability to satisfy caEbration requirements.
Therefore, the continued use of the subject calibrat on blocks will provide an acceptable level of quality and safety.
Conclusion-The pro ased alternative, to use the calibration block with a 14 inch diameter, will provide an acceptable level of quality and safety. Therefore, it is recommended that the proposed alternetive be authorized pursuant to 10CFR50.55a(a)(3)(i).
3.5.2 Exempted Components No relief requests 1
60 i
. - -. -. -. ~ -
t 3.5,3 Other No relief requests 3.5.3.1 Resquest for Relief No. RR 11, IWA 2610, Reference System for All Welds and Areas Subject to Volumetric and Surface Examination.
Code Requirement-Section XI, Paragraph IWA-2610, requires that a reference system be established for all welds and areas subject to surface or volumetric examination.
Each such weld and area shall be located and identified by a system of reference points.
The system shall permit identification of each weld, location of each weld center line, and designation of regular intervals along the length of the weld.
Licensee's Proposed Alternative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed applying the required markings as the inservice examinations are performed.
The licensee stated:
"Each Class 1 and 2 piping weld undergoing a volumetric and/or surface examination will receive the required markings as the inservice examinations are performed."
Licensee's Basis for Requesting Relief (as stated)-
"For an operating plant, performing actual masking of welds in order to identify each weld centerline, length locations, etc., would require many man hours of radiation exposure. Many of the welds are insulated and, as such, many man-hours of radiation exposure would be required to remove and reinstallinsulation l
just to facHitate marking.' Also many man-hours of radiation exposure would be involved in locating and marking the wtads."
"To establish a weld reference system for all welds subject to surface and I
volumetric examination, regardless if the weld is scheduled for examination, would require many man-hours of radiation exposure without a compensating increase in safety. Marking the welds as defined in the Alternate Examinations will provide the needed assurance of traceability of the piping welds and repeatability of the examinations."
Evaluation-The Code requires a reference system for e.Il weld areas subject to surface or volumetric examination. This system shall permit identification of each weld, the location of each weld centerline, and marking at regular intervals along the length of the weld. For operating plants, establishing a welrt reference system for all welds and areas subject to surface or volumetric examinations is a major effort. In some cases, due to i
inaccessibility and/or severe radiological conditions, welds may not be able to receive the
?
reeuired identifications and markings. Therefore, the Code requirement for establishing a j
weld reference system for all welds subject to examinations in the absence of j
examination presents an unnecessary burden for an operating plant. The many tasks j
involved in establishing the reference system, e.g. locating the welds, removing j
insulation, marking the welds and reinstalling insulation, would require many manhours of 1
l 61 4
,m..
.f-..
_.y
,w-
radiation exposure. The increase in plant safety would not compensate for the burden placed on the licensee that would result from imposition of the requirement for all welds.
The licensee's alternative, to mark each required Class 1 and 2 piping weld, as the examinations are pcrformed will assure that the intent of the Code requirement will be achieved.
Conc /usion--Marking all welds and areas subject to surf ace or volumetric examinations, as required by the Code, in the absence of inspection places a dif ficulty on Farley, Units 1 and 2, because they are operating plants. Imposition of the requirement on Southern Nuclear Operating Company would cause a significant burden that would not be compensated by an increase in quality and safety. The licensee's proposed alternative, to perform the required markings as the examinations are performed, will provide reasonable assurance that traceability of the piping welds and repeatability of examinations will be achieved. Therefore, it is recommended that the proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(ii).
3.5.3.2 Request for Relief No.12, Snubber Testing Note: Request for Relief No.12 is considered part of the Inservice Tesi (IST) Program and is, therefore, not included in this evaluation. This request for relief regarding snubber testing will be evaluated by the Mechanical Engineering Branch of the NRC.
1 62
4 9
i
- 4. CONCLUSION Pursuant to 10 CFR 50.55a(g)(6)(i), it has been determir.ed that certain inservice j
examinations cannot be performed to the extent required by Section XI of the ASME Code. In the cases of Relief Requests RR 6, RR-7, RR-8, RR-9, RR-10, RR-15. RR-16 RR-17, and RR 30 the licensee has demonstrated that specific Section XI requirements are impractical. It is, therefore, recommended that relief be granted as requested.
Granting relief will not endanNr life, property, or the common defense and security and is otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Pursuant to 10 CFR 50.55afa)(3),it is concluded that for Relief Requests RR-1, RR-2, RR-3, RR-4, RR-5, RR-11, RR-13, RR-18, RR-19, RR-20, RR-21, RR-22, RR-23, RR-24, RR-25, RR-26, and RR-29 the licensee's proposed alternatives will (a) provide an acceptable level of quality and safety, or (b) Code compliance will result in hardship or unusual difficulty without a compensating increase in safety. It is recommended that the proposed alternative be authorized.
Pursuant to 10 CFR 50.55a(a)(3)(i),it is concluded that for Reliet Request RR-14 the licensee's proposed alternative will provide an acceptable level of quality and safety for the subject components with the exception of the Volume Control Tank and the l
Regenerative Heat Exchanger. Therefore, it is recommended that the licensee's proposed l
alternative be authorized for all of the subject components with the exception of the
{
Volume Control Tank and Regenerative Heat Exchangers.
t Pursuant to 10 CFR 50.55a(a)(3)(i),it is concluded that for Relief Request RR-28 the licensee's proposed alternative will provide an acceptable level of quality and safety for a period of 60 months. Therefore, it is recommended that the licensee's proposed l
alternative be authorized for a limited time not to exceed 60 months from the start date of the third 10-year interval at Farley Unit 1, and from the start date of the Farley Unit 2, updated interval.
This technical evaluation has not identified any practical method by which Southern Nuclear Operating Company can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the existing Joseph M. Farley Nuclear Plant Units t
1 and 2. Compliance with all of the Section XI examination requirements would necessitate redesign of a significant number of plant systems, procurement of replacement components, installation ei the new components, and performance of I
baseline examinations for these components. Even after the redesign efforts, complete j
compliance with the Section XI examination requirements probably could not be j
achieved. Therefore, it is concluded that the public interest is not served by imposing provisions of Section XI of the ASME Code that have been determined to be impractical.
1 j
Southern Nuclear Operating Company should continue to monitor the development j
of new or improved examination techniques. As improvements are achieved, Southern a
y 63
)
Nuclear Operating Company should incorporate these techniques in the ISI program plan examination requirerr ents.
Based on the re'tiew of the Joseph M. Farley Nuclear Plant inservice Inspection Progrem Unit 1 Thiro' Ten Year For Class 1, 2, and 3 Components (Reference 3), the Joseph M. Farley Nuctcar Plant Unit 2 Updated Inservice Inspection Program (Reference 4), the Southern Nuclear Coerating Company's response to the NRC's request for additional information, and the recommendations for granting relief from the ISI i
examinations that cannct be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified with the exception of Relief Request RR-14.
l 64
t
- 5. REFERENCES 1.
Code of Federal Regulations, Title 10. Part 50.
2.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1,1989 Edition.
3.
Joseph M. Farley Nuclear Plant inservice Inspection Program Unit 1 Third Ten Year For Class 1, 2, and 3 Components, submitted May 28,1997.
4.
Joseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program, i
submitted May 28,1997.
5.
Letter, dated March 20,1997, Herbert N. Berkow (NRC) to D,. N. Morey (SNC) containing NRC SER.
j 6.
NUREG-0800, Standard Revicw Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 5.2.4, " Reactor Coolant Boundary Inservice inspection and Testing," and Section 6.6, " Inservice inspection of Class 2 and 3 Components,"
July 1981.
7.
Letter dated February 12,1998, J. I. Zimmerman (NRC) to D. N. Morey (SNC) containing request for additionalinformation.
8.
Letter dated June 12,1998, J. l Zimmerman (NRC) to D. N. Morey (SNC) containing request for additionalinformation.
9.
Letter dated April 6,1998 to Document Control Desk (NRC), containing response to the NRC RAI dated February 12,1998,
- 10. Letter dated July 13,1998 to Cocument Control Desk (NRC), containing response to i
the NRC RAI dated June 12,1998.
- 11. Letter dated April 27,1998 to Document Control Desk (NRC), containing correction to response to the NRC Ouestion/ Comments dated April 27,1998.
- 12. NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, Revision 11, October 1994.
- 13. NRC Regulatory Guide 1.150. Reactor Pressure Vessel Beltline Weld Examinations, Revision 1, February 1983.
- 14. NRC Regulatory Guide 1.83, inservice Inspection of Pressurized Water Reactor Steam Generator Tubes, Revision 1, July 1975.
- 15. S.H. Bush and R. R. Maccary, " Development ofIn-Service Inspection Safety Philosophy for U.S.A Nuclear Power Plants," ASME,1971.
- 16. NRC Regulatory Guide 1.163, Performance-based Containment Leak-Test Program, September 1995.
65
+
4 NRC Form 335 U.S. Nuclear Regulatory Comminica
- 1. REPORT NUMBER NPCM 1102
- Assigned by NRC, Ado Vol., Supp., Rev., and 3201.3202 Addendum Numbers, if any)
(
BIBLIOGRAPHIC DATA SHEET INEEI/ EXT-98-Oll56
- 2. TITLE AND SUBTITLE
- 3. DATE REPORT PUBLISHED Technical Evaluation Report on the Third 10-Year Interval Inservice and Updated Month Year hspection Program Plan:
November 1998 Southern Nuclear Operating Company, Joseph M. Farley Nuclear Plant
- 5. AUTHOR (S)
- 6. TYPE OF REPORT M. T. Anderson Technical C. T. Brown
- 7. PERIOD COVERED (Inclusive Dates)
S. G. Galbraith A. M..orter
- 8. PERFORMING ORGANIZATION NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Connission, and mailing address; if contractor, provide name and mailing address)
Idaho National Engmeering and Environmental Laboratory Materials Physics Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415
- 9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above";if contractor, provide NRC Division, Office or Region, U.S.
N" clear Regulatory Commission, and mailing address)
Civil and Geosciences Branch Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington,D.C. 20555
- 10. SUPPLEMENTARY NOTES
- 11. ABSTRACT (200 words orless)
Tids report presents the results of the evaluation of the Joseph M. Farley Nuclear Plant Inservice Inspection Program Unit I nird Ten Year For Class I,2, and 3 Components and the kseph M. Farley Nuclear Plant Unit 2 Updated Inservice Inspection Program submitted May 28,1997, including the requests for relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. The Joseph M. Farley Nuclear Plant Inservice Inspection Program Unit 1 Third Ten Year For Class 1,2, and 3 Components and the Joseph M. Farley Nuclear Plant Unit 2 are evaluated in Section 2 of this report. The inservice inspection (ISI) plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.
- 13. AVAILABILITY STATEMENT
- 12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report)
Unlimited
- 14. SECURIFY CLASSIFICATION (This page)
Unclassified (This report)
Unclassified
- 15. NUMBER OF PAGES 16 PRICE
c Page 1 of 3 Joseph M. Farley Nuclear Plant. Units 1 & 2 Third 10-Year ISI interval TABLE 1
SUMMARY
OF RELIEF REQUESTS Relief -
Number Component Category No.
Volume or' Area to be Examined i Licensee Proposodi Request System or ~
Exam item Required Method '
. Alternative Relief Request Status'
' RR I Cahbration Article i Par.
Cahb.ation Blocks Volumetric Continue Use of Existing Authorned Blocks 2000 1-2100 Material Requirements examination Cahbration Blocks RR 2 Cahbration Article i Par.
Cahbration Blocks Volumetric Continue Use of Existing Authorized l
Blocks 2000 12100 Notch Locations examination Cahbration Blocks i
RR 3 Cahbration Article i Par.
Cahbration Blocks Volumetric Continue Use of Existing Authorized Blocks 2000 1-2100 Hole location, depth, and/or examination Cahbration Blocks seperation requirements j
RR 4 Cahbration Article 1 Par.
Cahbration Blocks Volumetric Continue Use of Existing Authonted Blocks 2000 1-2200 Notch dimensional requirements examination Cahbration Blocks t
RR-5 Cahbration Article 1 Par.
Cahbration Blocks Volumetric Continue Use of Existing Authorized Blocks 2000 1-2200 Curvature requirements examination Cahbration Blocks I
RR G Pressonier BD B3.110 Pressuriier Nozzle to Vessef Welds Volumetric Volumetric Examination to Authorized Examination maximum extent possible,
}
perform supplemental l
surface examination RRO ft.1 Steam BF B5.70 Primary Side Volumetric ar;d Perform Volumetric Authorized Generator Nozzle to Safe-End Welds Surface examination to the l
Examination maximum extent practical.
Continue performance of surface examination.
i RR-8 R.1 Steam B 3.140 Primary Side Intet and Outlet Nozzles Volumetric Perform Visual Examination Authorized i
Generator BD Inner Radius Sections to the extent practical of the inner radius section of the nozzles RR-9 R.1 Class 1 Piping B-J B9.31 Branch Connection Pressure Retaining Volumetric and Perform Volumetric Authorized Welds in Piping Surface examination to the l
Examination maximum extent practical.
Continue performance of surface examination.
RR 10 Steam C-B C2.22 Norile inside radius section of outlet Volumetric None. No inner Radius Autha.! zed R.1 Generator nozzles Exists
Page 2 of 3 Joscph M. Farley Nucitar Plant, Unita 1 & 2 Third 10-Year ISIInterval TABLE 1
SUMMARY
OF RELIEF REQUESTS Relief -
Ucensou Proposed -
Request.
System or Exam item -
Number Component Category No.
~ Volume or Area to be Examined '
' Required Method l Altemative :
Relief Request Status RR-11 Class 1 and 2 IWA-Weld Reference System for Class 1 Volumetric and/or Welds receive required Authorized piping 2610 and 2 piping welds Surface markings as inservice inspection is performed j
RR 13 Reactor Vessel BG1 B6.10 Reactor Vessel Closure Head Nuts Surface VT-1 Visual Examination Authorized R.1 RR 14 Multiple IWC-Vessel Welds Volumetric and/or Appiv exemption Criteria of Authorized R.1 Vessels 1220 Surface 1989 Addenda RR-15 Charging CC C3.30 Integrally Welded Attachments Surface Perform Surf ace Exams to Authorited maximum extent possible R.1 Pumps Rf t-16 Reactor Vessel BD B3.90 Inlet and Outlet Nozzle-to-Vessel Volumetric Volumetric Examinations Authorized R.1 Welds will be performed to the maximum extent practical f rom the nozzle bore HR-17 Reactor Vessel B-A B 1.30 Shell to Flange Weld Volumetric Volumetric Examinations Authorized will be performed to the R.1 maximum extent practical RR-18 Reactor Vessel BD B3.90 Nozils to-Vessel Welds Volumetric on a Code Case N-521 Authoriicd I
BF B3.100 Noizie inside Radius minimum of 25%.
B5.10 Nozzle-to-Safe End Butt Welds but not more than 50% credited by the end of the first i
inspection period RR-19 Class 2 Piping C-F-1 C5.12 Longitudinal Welds Volumetric and Code Case N-524 Authorized C5.22 Surface C-F-2 C5.52 C5.62 l
RR-20 Class 1, 2, and B H B-K-1, Alternative rules of selection of Volumetric or Use of Code Case N-509 Authorized I
R.1 3 Integrally C-C, integrally welded attachments surface Welded D-A, D-B.
examination Attachments D-C 4.,
m
- - - ~ - _- m
Page 3 of 3 Joseph M. Farley Nucl:ar Plant, Units 1 & 2 Third 10-Year ISI interval TABLE 1
SUMMARY
OF RELIEF REQUESTS Relief Request System or -
' Exam
. item 1 Licensee Proposed.
Number Component.
Category No.
Volume or Area to be Examined -
j Required Method.
. - Altemative Relief Request Status -
~
RR-21 Class 1, 2, and B E. B P.
Systern Hydrostatic Tests of Class 1 Use of Code Case N-498-1 Authorized 3 Piping and C-H, D A, 2, and 3 Systerns Components D B, D C.
RR 22 Repairs and NA NA Pressure Tests Following Repairs and Use of Code Case N-416-1 Authorized Replacements Replacements HR-23 Class 1, 2, and IWA-5250 Bolted Connections VT-3 Perform an Evaluation to Authorized R.1 3 piping (a)
Determine Need for Boft Removal HR-2 4 Clurging Pump C-H IWC-Suction Piping VT-2 System VT-2, once each inspection Authorized 5222 Hydrostatic Test at NOP RR-25 Service Water IWA.
Buried portion of piping Flow dif ferential Visual examination of Authorized System Piping 5244(b) ground surface above buried piping RR-26 Reactor BP B 15.51 Vent and Drain Connections VT-2 System Visual examination for Authorized Coolant B17.71 Hydrostatic Test evidence of leakage each refueling outage with System valves in normally closed position RR-28 Containment C-H IWC-Piping, Components, and Valves System Pressure Use of Code Case N-522 Authorized for 60 Isolation 5221
- Test, months System IWC-System 5222 Hydrostatic Test RR-29 Spent Fuel Pool IWA-Concrete encased piping Flow ditferential Visual examination of Authorized Cooling System 52444 bl.
adjacent piping during IWA system inservice tests, 5244(ct, each examination period RR-30 Saf ety injection C-H IWC-Class 2 piping segments System Pressure Piping segments will be Authorized System 5221
- Test, included within the VT-2
!WC-System visual examination 5222 Hydrostatic Test boundary during the Class 1 system leakage test.
.