RNP-RA/13-0018, Relief Request (RR)-8 from Fifth Ten-Year Inservice Inspection Program Plan
| ML13080A258 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/13/2013 |
| From: | Wheeler S Duke Energy Carolinas |
| To: | Document Control Desk |
| References | |
| RNP-RA/13-0018 | |
| Download: ML13080A258 (8) | |
Text
Duke Energy Sharon A. Wheeler H. B. Robinson Steam Electric Plant Unit 2 Manager - Support Services Duke Energy 3581 West Entrance Road Hartsville, SC 29550 Serial: RNP-RA/13-0018 10 CFR 50.55a MAR 1 3 2013 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 RELIEF REQUEST (RR)-08 FROM FIFTH TEN-YEAR INSERVICE INSPECTION PROGRAM PLAN Pursuant to 10 CFR 50.55a(a)(3)(ii), Carolina Power & Light doing business as Progress Energy, requests relief from the ASME Code,Section XI, 2007 Edition through 2008 Addenda, Appendix VIII, Supplement 10. The requirements of ASME Section XI Appendix VIII, Supplement 10 require a vendor's Root Mean Square (RMS) error value to be 0.125" or less.
Progress Energy requests the use of an alternative to the ASME Code requirements pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(g)(5)(iii) on the basis that the code requirement is impractical in that no vendor has been able to achieve this value to date. The proposed alternative to the ASME Code is applicable for the remainder of the fifth 10-year Inservice Inspection (ISI) Interval at H. B. Robinson Steam Electric Plant (HBRSEP), Unit 2, which began on July 21, 2012, and is currently scheduled to end on February 19, 2022.
Relief request RR-08, enclosed, provides for alternative volumetric examinations of the reactor vessel hot leg and cold leg nozzle-to-pipe dissimilar metal welds performed from the inside surface during the HBRSEP fall 2013 refueling outage. To support inspections scheduled during this refueling outage Progress Energy requests approval of this relief request by August 1, 2013.
This letter contains no new regulatory commitments.
Serial: RNP-RA/13-0018 Page 2 of 2 If you have any questions concerning this matter, please contact Mr. Richard Hightower, Supervisor - Licensing/Regulatory Programs at (843) 857-1329.
Si cerel haron A. /heeler Manager - Support Services - Nuclear SAW/jsk Enclosure c:
Mr. V. M. McCree, NRC, Region II Ms. Araceli Billoch Col6n, NRC NRC Resident Inspector, HBRSEP Unit No. 2
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0018 Relief Request (RR)-08 From Fifth Ten-Year Inservice Inspection Program Plan Page 1 of 5
1.0 ASME Code Components Affected
Class 1 Reactor Coolant System Reactor Vessel Nozzle-to-Safe End Dissimilar Metal Welds Listed in Table 1 below.
Table 1 Nozzle to Safe-End Inspection Plan Code Case Weld No.
Summary Number N-770-1 Inspection Item Hot Leg Loop "A" DM 29.0" ID (Nom.)
107/01DM 35203 A-2 Weld Cold Leg Loop "A" DM 27.5" ID (Nom.)
107/41DM 36903 B
Weld Hot Leg Loop "B" DM 29.0" ID (Nom.)
107A/OIDM 37003 A-2 Weld Cold Leg Loop "B" DM 27.5" ID (Nom.)
107A/41DM 38703 B
Weld Hot Leg Loop "C" DM 29.0" ID (Nom.)
107B/OIDM 38803 A-2 Weld Cold Leg Loop "C" DM 27.5" ID (Nom.)
107B/41DM 40503 B
Weld This relief request was developed as an Engineering Change and is also applicable to the same welds identified in Table 1 for ASME Section XI, Category B-F, Item Number B5.10 but since the requirement to perform the examinations to ASME Code Case N-770-1 is endorsed by 10CFR50.55a, performing the ASME Code Case is an acceptable alternative to the requirements of ASME Section XI, Category B-F, Item Number B5. 10, therefore it is not necessary to request relief from ASME Section XI, Category B-F, Item Number B5. 10 in this relief request.
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0018 Relief Request (RR)-08 From Fifth Ten-Year Inservice Inspection Program Plan Page 2 of 5 Component materials and nozzle weld configurations are shown in Figure 1.
2" R. (apprx.)
Alloy Steel (RPV Nozzle)
Austenitic Stainless Steel '
(Safe End)
/
/i Austenitic Stainless Steel Clad i
3Q0 MACHINE THIS AREA TO 27-7/16" +1/16'/-0" DIA.
PRIOR TO X-RA Y OF WELD.
r
[*l I -z1/8" 11/"
Figure 1
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RAI13-0018 Relief Request (RR)-08 From Fifth Ten-Year Inservice Inspection Program Plan Page 3 of 5 2.0
Applicable Code Edition and Addenda
2.1 ASME Boiler and Pressure Vessel Code, Section Xl, 2007 Edition through the 2008 Addenda.
2.2 ASME Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1. Use of this Case is required by 10 CFR 50.55a(g)(6)(ii)(F).
3.0 Applicable Code Requirements 3.1 Code Case N-770-1, Table 1, Footnote (4) applies to volumetric examination of Inspection Items A-2 and B, and requires that "Ultrasonic volumetric examination shall be used and shall meet the applicable requirements of Appendix VIII."
Appendix VIII, Supplement 10, Paragraph 3.3(c) requires that "Examination procedures, equipment, and personnel are qualified for depth-sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm)."
Note that volumetric examinations of the Reactor Vessel nozzle-to-safe end dissimilar metal welds are also required to be performed in accordance with Section XI, Appendix VIII, as required by 10 CFR 50.55a(b)(2)(xv) and 10 CFR 50.55a(b)(2)(xvi).
4.0 Impracticality of Compliance Since 2002, the nuclear power industry has attempted to qualify personnel and procedures for depth-sizing examinations performed from the inside surface of dissimilar metal and austenitic stainless steel butt welds in PWR piping. As of March 4, 2013, no domestic or international vendor has met the applicable root mean square (RMS) error requirement of ASME Section XI Appendix VIII, Supplement 10.
The examination vendor that Progress Energy intends to use for performing these examinations has not been able to meet the applicable RMS error requirement identified in Appendix VIII, Supplement 10. The examination vendor that Progress Energy intends to use for performing these examinations has an RMS error of 0.212".
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0018 Relief Request (RR)-08 From Fifth Ten-Year Inservice Inspection Program Plan Page 4 of 5 5.0 Proposed Alternative and Basis for Use Progress Energy proposes to use the following alternative for flaw depth sizing when dissimilar metal welds are examined from the inside surface:
- 1. Examinations shall be performed using ultrasonic (UT) techniques that are qualified for flaw detection and length sizing using procedures, personnel and equipment qualified by demonstration in all aspects except depth sizing.
- 2. Flaw(s) detected and measured as less than 50% through-wall in depth shall be adjusted by adding a correction factor of the RMS error of 0.087 inches to the depths of any measured flaws based on the vendors demonstrated RMS of 0.212".
Eddy Current (ET) examination shall be using to confirm whether any detected flaws are surface breaking.
- 3. If any inner diameter surface-breaking flaws are detected and measured as 50%
through-wall or greater, Progress Energy shall repair the indications or shall perform flaw evaluations and shall submit the evaluations to the NRC for review and approval prior to reactor startup. These flaw evaluations shall include the following:
- a. Information concerning the mechanism which caused the flaw.
- b. Information concerning the surface roughness/profile in the area of the pipe/weld required to perform the examination, and an estimate of the percentage of potential surface areas with UT probe "lift-off."
All other ASME Code,Section XI, requirements for which relief was not specifically requested applies, including the third party review by the Authorized Nuclear Inservice Inspector.
Because compliance with the applicable requirements is impractical, this request is submitted pursuant to 10 CFR 50.55a(g)(5)(iii). Progress Energy believes that the proposed alternative provides reasonable assurance that flaws detected during examination will be sufficiently sized to disposition in accordance with acceptance standards of ASME Code Case N-770-1.
6.0 Duration of Proposed Alternative The proposed alternative to the ASME Code is applicable for the remainder of the fifth 10-year Inservice Inspection (ISI) Interval at HBRSEP, which began on July 21, 2012, and is currently scheduled to end on February 19, 2022.
United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/13-0018 Relief Request (RR)-08 From Fifth Ten-Year Inservice Inspection Program Plan Page 5 of 5 7.0 Related Industry Relief Requests A similar request was approved for use at McGuire Nuclear Station, Unit 2 in NRC letter dated September 24, 2012 (ADAMS Accession No. ML12258A363).
8.0 References 8.1 2007 Edition through 2008 Addenda, ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components."
8.2 2007 Edition through 2008 Addenda, ASME Code, Section Xl, Appendix VIII, Supplement 10.
8.3 ASME Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class I PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1.
8.4 EPRI Policy/Procedure Directive 03-01: Criteria for Issuing Documentation of Depth Sizing Errors That Exceed the 0.125-inch RMS Appendix VIII Criteria.
8.5 EPRI Materials Reliability Program Letter MRP 2012-046, dated November 26, 2012,
Subject:
Assessment of Effect of the Depth-Sizing Uncertainty for Ultrasonic Examinations from ID Surface of Large-Bore Alloy 82/182 and Austenitic Stainless Steel Butt Welds in PWR Primary System Piping, Revision 1.
ATTACHMENT 12 Sheet I of I Event Date: 3/12/2013 Event Summary: (AR# 594576)
While working on Radiation Monitor R-7 (Incore Instrumentation Area Monitor) in the Control Room, FIN I&C Techs inadvertently turned R-32A (CV High Range Area Monitor) off.
Why this is significant:
This rendered the R-32A Radiation monitor inoperable. Due to the inoperability, the Plant entered an unplanned 30-day Action Statement (ITS 3.3.3) until the monitor was returned to service.
Behaviors I Error Traps I Flawed Defenses I Latent Organizational Weaknesses Encountered:
" Technicians were focused primarily on R-7 and the adjacent ratemeter R-8 (Drumming Room Area Monitor), and did not use proper situational awareness along with risk mitigation strategies to ensure the equipment located below R-7 was not affected.
" The Two-Minute rule primarily focused on industrial safety and the equipment in the immediate area and did not include potential for inadvertent bumping of R-32A or R-32B.
How could this event have been prevented?
A more robust pre-job brief and two minute rule that included the potential for affecting adjacent equipment located below the work area.
" Better peer checks during the R-7 removal activity to insure tools and body position did not make unintended contact with plant equipment.
Immediate Corrective Actions I Human Performance Area(s) for Improvement:
Upon discovery, Operations entered the 30 day Action Statement associated with ITS 3.3.3, performed testing, and returned R-32A to service. A HURB was conducted on 3/13/13 to capture and share lessons learned.
What are the lessons learned?
- When working on and around sensitive equipment, pre-job briefs and two minute rules should include "out of the box" thinking to include prevention of inadvertent bumping or contact with adjacent equipment