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MONTHYEARHNP-08-045, Second Ten-Year Interval Inservice Inspection Program-Final Documentation Including Requests for Relief in Accordance with 10 CFR 50.55a2009-02-0505 February 2009 Second Ten-Year Interval Inservice Inspection Program-Final Documentation Including Requests for Relief in Accordance with 10 CFR 50.55a Project stage: Request ML0912701882009-05-19019 May 2009 Acceptance Review Regarding Relief Requests on Final Documentation for the Second 10-Year Inservice Inspection Program for Limited Coverage of Welds in Multiple Exam Categories Project stage: Acceptance Review ML0912704132009-07-21021 July 2009 Request for Additional Information Regarding Relief Request 2R1-018 on Final Documentation for the Second 10-Year Inservice Inspection Program for Limited Coverage of Welds in Examination Category B-F Project stage: RAI HNP-09-095, Response to Request for Additional Information Regarding Relief Requests 2R1-018, 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, & 2R2-011 for the Second 10-Year Inservice Inspection Program2009-09-24024 September 2009 Response to Request for Additional Information Regarding Relief Requests 2R1-018, 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, & 2R2-011 for the Second 10-Year Inservice Inspection Program Project stage: Response to RAI HNP-10-005, Response to Request for Clarifications Regarding Relief Request 2R1-0182010-02-11011 February 2010 Response to Request for Clarifications Regarding Relief Request 2R1-018 Project stage: Request ML1008204632010-04-0808 April 2010 Request for Relief 2R1-018 for the Second 10-Year Interval Inservice Inspection Program Plant Project stage: Other 2009-07-21
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Category:Letter type:HNP
MONTHYEARHNP-18-050, Request to Extend Reactor Vessel Surveillance Capsule Report Submission Date2018-09-17017 September 2018 Request to Extend Reactor Vessel Surveillance Capsule Report Submission Date HNP-18-004, License Amendment Request to Change Shearon Harris Nuclear Power Plant, Unit 1, Emergency Plan Emergency Action Level Scheme2018-08-13013 August 2018 License Amendment Request to Change Shearon Harris Nuclear Power Plant, Unit 1, Emergency Plan Emergency Action Level Scheme HNP-18-035, Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Lnservice Inspection Program for Containment, Third Ten-Year Interval2018-06-0404 June 2018 Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Lnservice Inspection Program for Containment, Third Ten-Year Interval HNP-18-049, Submittal of 1 0 CFR 50.54(q) Evaluation Form of Changes to Procedure EMP-420, Emergency Program Maintenance, Revision 192018-05-0303 May 2018 Submittal of 1 0 CFR 50.54(q) Evaluation Form of Changes to Procedure EMP-420, Emergency Program Maintenance, Revision 19 HNP-18-023, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes2018-05-0202 May 2018 Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes HNP-18-042, Annual Environmental (Non-radiological) Operating Report2018-04-30030 April 2018 Annual Environmental (Non-radiological) Operating Report HNP-18-032, Annual Radiological Environmental Operating Report2018-04-25025 April 2018 Annual Radiological Environmental Operating Report HNP-18-031, Annual Radioactive Effluent Release Report2018-04-25025 April 2018 Annual Radioactive Effluent Release Report HNP-18-047, Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval, Non-Proprietary Version of Calculation2018-04-20020 April 2018 Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval, Non-Proprietary Version of Calculation HNP-18-045, Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval2018-04-18018 April 2018 Submittal of Relief Request I4R-18, Reactor Vessel Closure Head Nozzle Repair Technique, Inservice Inspection Program, Fourth Ten-Year Interval HNP-18-044, Cycle 22 Core Operating Limits Report, Revision O2018-04-16016 April 2018 Cycle 22 Core Operating Limits Report, Revision O HNP-18-039, Supplement to License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses2018-04-13013 April 2018 Supplement to License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses HNP-18-021, Annual Report in Accordance with Technical Specification 6.9.1.22018-02-19019 February 2018 Annual Report in Accordance with Technical Specification 6.9.1.2 HNP-18-020, Supplement to Response to Request for Additional Information Regarding License Amendment Request for Spent Fuel Storage Pool Criticality Analyses2018-02-16016 February 2018 Supplement to Response to Request for Additional Information Regarding License Amendment Request for Spent Fuel Storage Pool Criticality Analyses HNP-18-019, Cycle 21 Core Operating Limits Report, Revision 12018-02-14014 February 2018 Cycle 21 Core Operating Limits Report, Revision 1 HNP-18-017, CFR 50.54(q) Evaluations2018-02-0505 February 2018 CFR 50.54(q) Evaluations HNP-18-001, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors.2018-02-0101 February 2018 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Reactors. HNP-18-002, Response to Request for Additional Information Regarding a License Amendment Request Proposing Changes to Emergency Diesel Generator Technical Specifications Surveillance Requirements2018-01-22022 January 2018 Response to Request for Additional Information Regarding a License Amendment Request Proposing Changes to Emergency Diesel Generator Technical Specifications Surveillance Requirements HNP-18-003, Response to Request for Additional Information Regarding License Amendment Request for Spent Fuel Storage Pool Criticality Analyses2018-01-18018 January 2018 Response to Request for Additional Information Regarding License Amendment Request for Spent Fuel Storage Pool Criticality Analyses HNP-18-007, License Amendment Request to Incorporate Tornado Missile Risk Evaluator Into Licensing Basis - Supplement Regarding De Minimis Penetrations2018-01-11011 January 2018 License Amendment Request to Incorporate Tornado Missile Risk Evaluator Into Licensing Basis - Supplement Regarding De Minimis Penetrations HNP-18-006, Submittal of 14-Day Special Report for Accident Radiation Monitors2018-01-0404 January 2018 Submittal of 14-Day Special Report for Accident Radiation Monitors HNP-17-093, Supplement to License Amendment Request Proposing a New Set of Fission Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 32017-11-29029 November 2017 Supplement to License Amendment Request Proposing a New Set of Fission Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3 HNP-17-078, Response to Request for Additional Information Regarding a License Amendment Request Proposing Changes to Emergency Diesel Generator Technical Specifications Surveillance Requirements2017-11-27027 November 2017 Response to Request for Additional Information Regarding a License Amendment Request Proposing Changes to Emergency Diesel Generator Technical Specifications Surveillance Requirements HNP-17-082, Response to Request for Additional Information Regarding License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Limit Detailed In..2017-10-30030 October 2017 Response to Request for Additional Information Regarding License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Limit Detailed In.. HNP-17-084, Response to Request for Additional Information Regarding a License Amendment Request Proposing Changes to Emergency Diesel Generator Technical Specifications Surveillance Requirements2017-10-30030 October 2017 Response to Request for Additional Information Regarding a License Amendment Request Proposing Changes to Emergency Diesel Generator Technical Specifications Surveillance Requirements HNP-17-025, Fourth Interval Lnservice Inspection Plan, Third Interval Containment Lnservice Inspection Plan, and Fourth Interval Lnservice Inspection Pressure Test Plan2017-10-23023 October 2017 Fourth Interval Lnservice Inspection Plan, Third Interval Containment Lnservice Inspection Plan, and Fourth Interval Lnservice Inspection Pressure Test Plan HNP-17-072, License Amendment Request to Incorporate Tornado Missile Risk Evaluator Into Licensing Basis2017-10-19019 October 2017 License Amendment Request to Incorporate Tornado Missile Risk Evaluator Into Licensing Basis HNP-17-077, License Amendment Request Regarding Rod Control Movable Assemblies Technical Specifications2017-10-10010 October 2017 License Amendment Request Regarding Rod Control Movable Assemblies Technical Specifications HNP-17-076, Supplement to License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete Duplicate Reporting Requirements.2017-10-0202 October 2017 Supplement to License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete Duplicate Reporting Requirements. HNP-17-073, Supplemental Information for License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses2017-09-14014 September 2017 Supplemental Information for License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses HNP-17-062, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal2017-09-13013 September 2017 Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Submittal HNP-17-061, Supplement to License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses2017-07-20020 July 2017 Supplement to License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses HNP-17-008, License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses2017-06-28028 June 2017 License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses HNP-17-003, License Amendment Request for Emergency Diesel Generator Surveillance Requirements Regarding Voltage and Frequency Limits and the Voltage Limit for Emergency Diesel Generator Load Rejection2017-06-0505 June 2017 License Amendment Request for Emergency Diesel Generator Surveillance Requirements Regarding Voltage and Frequency Limits and the Voltage Limit for Emergency Diesel Generator Load Rejection HNP-17-040, Snubber Program Plan2017-06-0101 June 2017 Snubber Program Plan HNP-17-051, 10 CFR 50.54(q) Evaluation2017-05-24024 May 2017 10 CFR 50.54(q) Evaluation HNP-17-050, Transmittal of 10 CFR 50.54(q) Evaluation for Revision 21 to PEP-250, Activation and Operation of the Joint Information Center2017-05-24024 May 2017 Transmittal of 10 CFR 50.54(q) Evaluation for Revision 21 to PEP-250, Activation and Operation of the Joint Information Center HNP-17-041, Response to Request for Additional Information Regarding License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete2017-05-22022 May 2017 Response to Request for Additional Information Regarding License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106, and Delete Du HNP-17-033, License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 112017-05-22022 May 2017 License Amendment Request Proposing a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods That Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 HNP-17-049, Transmittal of 10 CFR 50.54(q) Evaluation and Revision 26 to PEP-110, Emergency Classification and Protective Action Recommendations.2017-05-18018 May 2017 Transmittal of 10 CFR 50.54(q) Evaluation and Revision 26 to PEP-110, Emergency Classification and Protective Action Recommendations. HNP-17-047, 10 CFR 50.54(q) Evaluation2017-05-18018 May 2017 10 CFR 50.54(q) Evaluation HNP-17-048, Transmittal of 10 CFR 50.54(q) Evaluation and Revision 28 to PEP-230, Control Room Operations.2017-05-18018 May 2017 Transmittal of 10 CFR 50.54(q) Evaluation and Revision 28 to PEP-230, Control Room Operations. HNP-17-046, 10 CFR 50.54(q) Evaluation2017-05-18018 May 2017 10 CFR 50.54(q) Evaluation HNP-17-024, Annual Environmental (Nonradiological) Operating Report2017-04-28028 April 2017 Annual Environmental (Nonradiological) Operating Report HNP-17-023, Annual Radioactive Effluent Release Report2017-04-28028 April 2017 Annual Radioactive Effluent Release Report HNP-17-034, Response to Request for Additional Information Regarding License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106..2017-04-25025 April 2017 Response to Request for Additional Information Regarding License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report, Delete Reference to Plant Procedure PLP-106.. HNP-17-035, Request for Pilot Plant Status and Fee Waiver to Implement Tornado Missile Risk Evaluator2017-04-12012 April 2017 Request for Pilot Plant Status and Fee Waiver to Implement Tornado Missile Risk Evaluator HNP-17-045, 10 CFR 50.54(q) Evaluation of Change in Emergency Plan Implementation Procedure PEP-241, Revision 7, Technical Support Center (TSC) Emergency Ventilation System Operation2017-03-30030 March 2017 10 CFR 50.54(q) Evaluation of Change in Emergency Plan Implementation Procedure PEP-241, Revision 7, Technical Support Center (TSC) Emergency Ventilation System Operation HNP-17-014, Annual Report in Accordance with Technical Specifications, Section 6.9.1.22017-02-16016 February 2017 Annual Report in Accordance with Technical Specifications, Section 6.9.1.2 HNP-17-015, Summary of 10 CFR 50.54(q) Evaluation2017-02-15015 February 2017 Summary of 10 CFR 50.54(q) Evaluation 2018-09-17
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Serial:. HNP-i,0-005 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO REQUEST FOR CLARIFICATIONS REGARDING RELIEF REQUEST 2R1-018
References:
- 1. Letter from D. H. Corlett to the Nuclear Regulatory Commission (SERIAL:
HNP-08-045), "Second Ten-Year Interval Inservice Inspection Program - Final Documentation Including Requests for Relief in accordance with-10 CFR 50.55a," dated February 05, 2009
- 2. Letter from D. H. Corlett to the Nuclear Regulatory Commission (SERIAL:
HNP-09-095), "Response to Request for Additional Information Regarding Relief Requests 2R1-018, 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, and 2R2-01 For The Second 10-Year Interval Inservice Inspection Program (TAC NOS. ME0608, ME0609, ME0610, ME061 1, ME0612, ME0613, ME0614, and ME0615)," dated September 24, 2009 Ladies and Gentlemen:
During the December 3, 2009, conference call between the NRC and members of the Harris Nuclear Plant (HNP) staff, the reviewers requested supplemental information as documented via a December 9, 2009, email from M. G. Vaaler, Project Manager. HNP submitted this original Relief Request as Serial: HNP-08-045 (Reference 1), with additional information provided as Serial: HNP-09-095 (Reference 2).
The Enclosure to this letter provides the requested additional information.
This document contains no new or revised Regulatory Commitments.
Please refer any questions regarding this submittal to me at (919) 362-3137.
Sincerely, D. H. Corlett Supervisor -Licensing/Regulatory Programs Harris Nuclear Plant Progress Energy Carolinas, Inc.
Harris Nuclear Plant P. 0. Box 165 New Hill, NC 27562 14v04
HNP- 10-005 Page 2 DHC/kms
Enclosure:
Response to Supplemental Questions on Relief Request 2R 1-018 cc: Mr. J. D. Austin, NRC Sr. Resident Inspector, HNP Mr. L. A. Reyes, NRC Regional Administrator, Region II Mr. R. Winegarden, Harris Plant Authorized Nuclear Inservice Inspector Ms. M. G. Vaaler, NRC Project Manager, I-TNP
Enclosure to SERIAL: HNP-10-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO SUPPLEMENTAL QUESTIONS ON RELIEF REQUEST 2R1-018 Summary Carolina Power & Light Company, now doing business as Progress Energy Carolinas, Inc.,
(PEC) submitted Relief Request (RR) 2R1-018 for the Harris Nuclear Plant (HNP). The proposed RR is for the second 10-year inservice inspection (ISI) interval, in which the licensee adopted the 1989 Edition of the ASME Code,Section XI, no Addenda, as the Code of Record.
Per the conference call on December 3, 2009, between the NRC and members of the Harris staff, the following supplemental information was requested in regard to relief request 2R1-018.
Request 1: Please verify that the welds in RR 2R1-018 are a part of the reactor pressure vessel (RPV)
Response: The Inlet Nozzle Dissimilar Metal (DM) welds are located off the Reactor Vessel at the Elbow to Nozzle, as depicted in the following:
SERIAL: HNP-09-095 (Reference 2):
Enclosure 2, page 8 of 10, Reactor Vessel drawing, 1-ISI-RV-1 Enclosure 2, page 9 of 10, Reactor Vessel drawing, 1-ISI-RV-A600 These examinations are performed in conjunction with the Reactor Vessel examination using the same equipment and nomenclature.
Request 2: Please verify that the inspections were conducted from the inner diameter of the weld (ID)
Response: All examinations were performed from the inner diameter (ID) of the weld, as provided in the following:
SERIAL: HNP-08-045 (Reference 1), Section 5.0:
".... volumetric examination for the subject welds at HNP is restricted due to geometric surfaces (inner diameter surface counter-bore and root configuration) which limit accessibility and make the 100 percent volumetric examination impractical for these areas."
SERIAL: HNP-09-095 (Reference 2): "ID" typed in "Examination Surface" boxes in the following:
Enclosure 2, Attachment A, Analysis Log # SE-95-1 Enclosure 2, Attachment B, Analysis Log # SE-215-1 Enclosure 2, Attachment C, Analysis Log # SE-335-1 Page 1 of 4
Enclosure to SERIAL: HNP-10-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO SUPPLEMENTAL QUESTIONS ON RELIEF REQUEST 2R1-018 Request 3: Please verify that the eddy current inspection was a full coverage exam Response: The Eddy Current examinations were full coverage, i.e. 100% of the ID surface of the required examination volume per ASME Section XI Figure IWB-2500-8 examination surface E-F, as presented in:
SERIAL: HNP-09-095 (Reference 2): "100" for "CCW" and "CW" in "ET" columns in the following:
Enclosure 2, Attachment A, RPV Coverage Estimate Breakdown, Weld No. RVNOZCl-N-05SE Enclosure 2, Attachment B, RPV Coverage Estimate Breakdown, Weld No. RVNOZB l-N-03 SE Enclosure 2, Attachment C, RPV Coverage Estimate Breakdown Weld No. RVNOZA1-N-01SE Request 4: Please expand upon the description of the difference(s) between the 1996 and 2006 weld inspections Response: During the first interval, the UT examinations were performed by a different vendor using different equipment, resulting in different coverage limitations. The transducers used by the vendor in 1996 had a 3/8" by 1" footprint, while the transducers used by the 2006 vendor had a 22mm x 22mm footprint. In 2006, each of the transducers was individually gimbaled within the examination sled to maximize compliance with the ID surface of the nozzles.
Additionally, this 2006 combination of technique and equipment was qualified to ASME Code,Section XI, Appendix VIII, as administered by the EPRI PDI, while the method utilized in 1996 was not ASME Code, Appendix VIII, PDI qualified.
There were no supplemental Eddy Current examinations performed during the first interval examinations. For 2006, an Eddy Current examination with 100% surface coverage was performed.
Request 5: Please provide a more detailed description and/or explanation of the limitations of the weld inspection technique(s) used Response: These inspections were performed from the ID surface of the Elbow to Nozzle examination area, an area that is not accessible for a direct visual inspection to determine existing obstructions resulting in the limited coverage. Discussions with an industry expert found that in an elbow to nozzle configuration, a counterbore would typically be required to better provide for ID mismatch. In addition, the root may have some type of protrusion that Page 2 of 4
Enclosure to SERIAL: IHNP-10-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO SUPPLEMENTAL QUESTIONS ON RELIEF REQUEST 2R1-018 could limit the examination, although this does not appear likely due to the complete coverage for the supplemental eddy current examination.
The examinations were limited in the circumferential direction only, specifically in areas of counterbore and/or root. The qualified inspection procedure used has no limitations due to surface configuration for axial scanning. The axial direction scans, looking for the more critical circumferential flaws, achieved 100% coverage.
The ID configuration of the Elbow to Nozzle examination area was profiled with an immersion zero degree transducer. The data from this profiling was used to determine areas where circumferential scanning transducer contact was compromised to an extent greater than the EPRI guideline of 1/32" of the contact area of the transducer.
Request 6: Please clarify the wave mode(s) and angles used for the inspection Response: All the examinations were performed to a qualified PDI procedure. The Inlet Nozzle Dissimilar Metal (DM) welds were examined using a 70 ° Longitudinal (refracted L) wave dual probe, as presented:
SERIAL: HNP-09-095 (Reference 2):
Enclosure 2, page 3 of 10, "70' Longitudinal Wave Dual Scan,"
Component ID RVNOZC l-N-05SE Enclosure 2, page 4 of 10, "70' Longitudinal Wave Dual Scan,"
Component ID RVNOZBl-N-03SE Enclosure 2, page 5 of 10, "70' Longitudinal Wave Dual Scan,"
Component ID RVNOZA1-N-01SE Request 7: Please provide an additional explanation of the eddy current technique used in this inspection (reference: V.C. Summer submittal dated June 20, 2006)
Response: The eddy current technique used was a surface examination technique. Per the examination vendor Wesdyne, a Westinghouse NDE Company, this exam was used to supplement the UT in areas of limited contact due to its smaller contact area (0.25"dia vs. 0.87" x 0.87" square for the UT transducers) and assisted in the characterization of any potential surface breaking ID connected PWSCC flaws, a concern for this component.
The eddy current technique parameters utilized:
- Up to two plus point probes applied circumferentially on the ID surface in scan increments of 0.80 inches (for axial flaws) and 0.25 inches axially.
- Automated systems for data collection and analysis.
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Enclosure to SERIAL: HNP-10-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 RESPONSE TO SUPPLEMENTAL QUESTIONS ON RELIEF REQUEST 2R1 -018 The target flaw size for the eddy current procedure is 0.28 inches long, which is well within the ASME Code linear flaw acceptance standard of 0.45 inches long for austenitic materials (as defined for the outside surface in the ASME Code Tables).
As provided by Wesdyne, the eddy current technique was developed to augment the ultrasonic examination method and to provide increased sensitivity at the ID surface. The eddy current technique was first used in the VC Summer reactor vessel primary nozzle examinations of 2000.
The procedure was refined after its first use in 2000 by applying it to the VC Summer hot leg dissimilar metal weld section removed from service. The removed section had a number of primary water stress corrosion cracking flaws along with non-relevant indications resulting from metallurgical interface and surface geometry. Using these actual flaws and geometric conditions in the removed section to refine the technique, Wesdyne developed reliable flaw-screening criteria which allowed for the successful use of the procedure in the VC Summer 2002 and 2003 examinations.
Since that time, the technique has been successfully blind tested at the Swedish NDT Qualification Centre (SQC Kvalificeringscentrum AB) under the program, "Qualification of Equipment, Procedure and Personnel for Detection, Characterization and Sizing of Defects in Areas in Nozzle to Safe End Welds at Ringhals Unit 3 and 4," Hakan Soderstrand 7-10-03. The important qualification parameters for Eddy Current in the SQC blind tests were as follows:
-Defect types: fatigue and stress corrosion cracks, surface initiated
-Tilt: +/- 10 degrees; Skew: +/- 10 degrees
-Detection target size: IDSCC 6mm (0.25 inches) long
-Flaw Location: within 10mm (13/32 inch)
-Length of the planar flaw within a 70% confidence interval: +/-9mm (3/8 inch)
-False call rate: less than or equal to 20% for the personnel qualification tests The technique has also been used to supplement examination of portions of the relevant near-surface volumes during more than 20 domestic pressurized water reactor nozzle-to-pipe examinations conducted by Wesdyne.
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