HNP-12-060, Relief Request I3R-09 Reactor Vessel Closure Head Nozzle Repairs Inservice Inspection Program - Third Interval, Request for Additional Information Response

Relief Request I3R-09 Reactor Vessel Closure Head Nozzle Repairs Inservice Inspection Program - Third Interval, Request for Additional Information Response

ML12139A407
Person / Time
Site:	Harris
Issue date:	05/18/2012
From:	Corlett D Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-12-060, TAC ME8523
Download: ML12139A407 (11)

Text

David H. Corlett Supervisor, Licensing/Regulatory Programs Harris Nuclear Plant Progress Energy Carolinas, Inc.

May 18, 2012 Serial: HNP-12-060 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400 / Renewed Facility Operating License No. NPF-63

Subject:

Relief Request I3R-09 Reactor Vessel Closure Head Nozzle Repairs Inservice Inspection Program - Third Interval, Request for Additional Information Response (TAC No. ME8523)

References:

1. Letter from D. H. Corlett to the U.S. NRC, Relief Request I3R-09 Reactor Vessel Closure Head Nozzles, Serial HNP-12-054 dated May 3, 2012 (ADAMS Accession No. ML12128A007)

2. Letter from A. T. Billoch Colón to C. L. Burton, Request for Additional Information Related to the Relief Request I3-09 for the Reactor Vessel Closure Head Nozzles Repair for the Third Inservice Inspection Program Interval (TAC No. ME8523), dated May 11, 2012 (ADAMS Accession No. ML12128A294)

Ladies and Gentlemen:

By Reference 1, Carolina Power & Light Company, doing business as Progress Energy Carolinas, Inc., submitted Relief Request I3R-09 for the repair of degraded reactor vessel closure head nozzle penetrations for Shearon Harris Nuclear Power Plant, Unit No.1. By Reference 2, the NRC provided a request for additional information (RAI) on that subject. The RAI response is provided in the enclosure to this letter.

P.O. Box 165 New Hill, NC 27562 T> 919.362.3137

U.S. Nuclear Regulatory Commission Page 2 HNP-12-060 This document contains no regulatory commitments.

Please refer any questions regarding this submittal to me at (919) 362-3137.

Sincerely,

Enclosure:

Relief Request I3R-09, Reactor Vessel Closure Head Nozzle Repairs, Inservice Inspection Program - Third Interval, Request for Additional Information Response cc: Mr. 1. D. Austin, NRC Sr. Resident Inspector, HNP Ms. A. T. Billoch Colon, NRC Project Manager, HNP Mr. V. M. McCree, NRC Regional Administrator, Region II

HNP-12-060 Enclosure Shearon Harris Nuclear Power Plant / Unit No. 1 Docket No. 50-400 / Renewed Facility Operating License No. NPF-63 Relief Request I3R-09 Reactor Vessel Closure Head Nozzle Repairs Inservice Inspection Program - Third Interval Request for Additional Information Response

U.S. Nuclear Regulatory Commission Page 1 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response

References:

1. Letter from D. H. Corlett to the U.S. NRC, Relief Request I3R-09 Reactor Vessel Closure Head Nozzles, Serial HNP-12-054 dated May 3, 2012 (ADAMS Accession No. ML12128A007)

2. Letter from A. T. Billoch Colón to C. L. Burton, Request for Additional Information Related to the Relief Request I3-09 for the Reactor Vessel Closure Head Nozzles Repair for the Third Inservice Inspection Program Interval (TAC No. ME8523), dated May 11, 2012 (ADAMS Accession No. ML12128A294)

By Reference 1, Carolina Power & Light Company, doing business as Progress Energy Carolinas, Inc., submitted Relief Request I3R-09 (RR) for the repair of degraded reactor vessel closure head nozzle penetrations for Shearon Harris Nuclear Power Plant, Unit No.1 (HNP). By Reference 2, the NRC provided a request for additional information (RAI) on that subject. The RAI response follows.

Question 1.

Page 6, Section 5.a. of the RR states that the interpass temperature will be determined by one of the following methods: (1) heat flow calculations or (2) measurement based on a test coupon. Discuss the exact measurement method that will be used and explain why that method is selected.

Response 1.

Option 1, the use of heat flow calculations, is used to determine a conservative maximum anticipated interpass temperature to ensure interpass temperature limits are not exceeded.

In the Inner Diameter Temper Bead (IDTB) repair scenario, the maximum heat input of 32,200 Joules per inch with an average time of 1 minute between subsequent weld passes results in a calculated base material temperature increase of approximately 6° F. Based on AREVAs experience with 124 IDTB reactor vessel head nozzle repairs, the typical time between weld passes will be significantly greater than a minute as a result of weld sequencing, viewing previously deposited weld passes, completing paperwork, independent verifications, and routine equipment maintenance including tungsten electrode replacement.

Question 2.

Page 8, Section 5.b. of the RR states, in part, "...For this modification, the repair weld is suitable, except for the taper transition, for ultrasonic testing (UT) examination and a final surface examination can be performed..." Discuss the examination method(s) that will be used for the taper transition. If the answer is: No examination will be performed, discuss how the taper transition can be ensured of its structural integrity (i.e., no flaws).

U.S. Nuclear Regulatory Commission Page 2 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Response 2.

Liquid penetrant examination will be performed on the entire weld, including the taper transition. In addition, 70L and 45L axial ultrasonic examination scans looking down (see RR Figures 5 and 7) will interrogate the taper transition volume. The performance of the surface and ultrasonic examinations provides assurance of structural integrity.

Question 3.

Page 8, Section 5.b. of the RR states, in part, "...Approximately 70% of the weld surface will be scanned by UT. Approximately 83% of the RVCH ferritic steel heat affected zone (HAZ) will be covered by UT" Based on the aforementioned examination coverage, Figure 4 through Figure 8 provide the coverage for UT examination with the hashed lines.

Question 3.a.

Discuss the area and volume of the new weld that could not and will not be examined.

Response 3.a.

The repair weld produces a region that limits the examination volume. The downward aimed angle beam transducers (45L and 70L) are used to interrogate this area for defects (planar defects normal to the beam, cracking, lack-of-fusion, etc.). The UT is being performed in addition to the surface examinations. There is no portion of the repair volume that does not receive at least single direction ultrasonic coverage.

Question 3.b.

Discuss how the area and volume of the new weld that could not be examined can be ensured of structural integrity (i.e., no flaws).

Response 3.b.

Liquid penetrant examination will be performed on the entire surface area. In addition, the volume in question will be examined to the extent practical using the 70L and 45L (RR I3R-09, Figures 5 and 7) axial ultrasonic examination scans (looking down). There is no portion of the repair that does not receive surface liquid penetrant examination and at least single-direction ultrasonic coverage of the volume.

U.S. Nuclear Regulatory Commission Page 3 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Question 3.c.

If 70% of the new weld surface will be scanned by UT, clarify what is the percentage of coverage for the "volume" of the weld.

Response 3.c.

The actual volume examined will be calculated after the as-built dimensions of the weld are known and the examination is performed. It is anticipated that greater than 80% of the examination volume will obtain two-directional coverage.

Question 4.

Page 10, Section 5.d., last paragraph of the RR states that the stress on the postulated circumferential flaw has a margin of 1.43 with respect to the allowable stress, and the depth of axial flaw has a margin of 3.9 with respect to the allowable flaw depth. First paragraph on page 11 states that the fatigue crack growth is acceptable.

Question 4.a.

Provide the final crack size (length and depth) in the axial and circumferential direction at the end of 40 years and explain why the final crack size is acceptable.

Response 4.a.

• Axial flaw: final depth (af) is 0.1008 inch, since length/depth is 2, length = 0.202 inch.

• Circumferential flaw: the final flaw depth of the 360° circumferential flaw is 0.1002 inch.

The final crack sizes are acceptable based on ASME Code,Section XI, IWB-3640 flaw evaluations which demonstrate that the final flaw sizes satisfy the applicable Code acceptance criteria, as discussed below in the response to Question 4.b.

Question 4.b.

Explain how the margins on stress and flaw depth satisfy the required margins in the American Society of Mechanical Engineers (ASME) Code,Section XI, IWB-3600.

U.S. Nuclear Regulatory Commission Page 4 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Response 4.b.

For flaws in IDTB weld, the applicable section is IWB-3640. Following the procedures in IWB-3641 and acceptance criteria of IWB-3642 the flaw evaluation based on Appendix C is performed.

For the circumferential flaw, the stress margin is calculated per Article C-5000 of ASME Code Section XI.

The stress margin:

St/m = 1.43 where m is the membrane stress, St =mc/SFm, where mc is the critical membrane stress, and SFm is the safety factor of 2.7 per C-2620 For axial flaws, the calculated stress ratio (SFm h/f) is 0.519 and the nondimensional flaw length is 0.211. Thus the allowable flaw size (a/t) determined from Table C-5410-1 of ASME Code Section XI is 0.75 and allowable flaw depth is 0.395 inch. Thus the allowable flaw size margin, aallow/af= 3.9.

The margins of 1.43 for circumferential and 3.9 for axial flaws exceed the required margins of the ASME Code; therefore, the flaw evaluations demonstrate that the required margins of IWB-3600 are satisfied.

Question 4.c.

The RR states that the fracture toughness margins for flaws have been shown to be acceptable. Discuss the fracture toughness margins.

Response 4.c.

The fracture margin calculation includes the required safety factors and hence the required margin is only 1.0. Thus the calculated margins, 1.43 for circumferential flaws and 3.9 for axial flaws, are acceptable.

Question 5 Page 11. Section 5.e. of the RR discusses the worst-case flaw in the J-groove weld.

Question 5.a.

Describe the worst-case flaw.

U.S. Nuclear Regulatory Commission Page 5 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Response 5.a.

A flaw in the J-groove weld cannot be sized by currently available nondestructive examination techniques. It is conservatively assumed that the as-left condition of the remaining J-groove weld includes flaws extending through the entire Alloy 82/182 J-groove weld and butter material. It is further postulated that the dominant hoop stresses in the J-groove weld would create a situation where the preferential direction for cracking would be radial. A radial crack in the Alloy 82/182 weld metal would propagate by primary water stress corrosion cracking (PWSCC), through the weld and butter, to the interface with the low alloy steel head material, where it would blunt, or arrest. Any growth of the postulated as-left flaw into the low alloy steel head would be by fatigue crack growth under cyclic loading conditions.

Question 5.b.

Explain why the proposed nozzle repair design configuration is considered to be acceptable for 30 years of operation after the weld repair.

Response 5.b.

Linear-elastic (LEFM) and elastic-plastic (EPFM) fracture mechanics analyses were used to demonstrate that the remaining worst-case as-left J-groove flaw would be acceptable for 30 years of service. Although the postulated flaw did not satisfy ASME Code Section XI IWB-3612 for all transient loading conditions, LEFM analysis

• determined that the uphill side of the reactor head penetration was the worst case position for the postulated flaw,

• calculated the final flaw size by fatigue crack growth, and

• identified the controlling service conditions for evaluation by EPFM.

For normal and upset conditions, the controlling loading condition was identified to be a reactor trip, for which it was shown, using safety factors of 1.5 on primary loads and 1.0 on secondary loads, that the applied J-integral (0.785 kips/in) was less than the J-integral of the low alloy steel head material (2.473 kips/in) at a crack extension of 0.1 inch. For emergency and faulted conditions, the controlling loading condition was a large loss of coolant accident, for which it was shown that with safety factors of 1.5 on primary loads and 1.0 on secondary loads that the applied J-integral (2.359 kips/in) was less than the J-integral of the low alloy steel head material (2.474 kips/in) at a crack extension of 0.1 inch. Flaw stability during ductile flaw growth was easily demonstrated for both loading conditions using safety factors of 3.0 and 1.5 for the reactor trip and 1.5 and 1.0 for the large loss of coolant accident.

U.S. Nuclear Regulatory Commission Page 6 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Question 6 Discuss whether all the flaws identified in Nozzles No. 5, 17, 38, and 63 will be removed as a result of the proposed repair. If not, justify how the nozzles with remnant flaws can be ensured of structural integrity so as not to affect the function of the control rod drive mechanism and the structural integrity of the reactor vessel head.

Response 6 It is believed that the flaws that have been detected by UT have been removed when the lower portion of the nozzle was machined away from the J-groove weld. However, as discussed in the response to question 5, flaws are postulated to exist in the remaining portion of the J-groove weld and shown in the evaluation to be acceptable for 30 years of service.

Question 7 Section 6 of the RR states that the proposed repair has a design life expectancy of 14.8 effective full-power years (EFPYs). Describe in detail or submit the analysis that results in the design life of 14.8 EFPYs. Section 5.e. states that the nozzle repair design is considered to be acceptable for 30 years. Explain the discrepancy between the two different design life expectancies.

Response 7 The 14.8 EFPY life is based on PWSCC of the remaining Alloy 600 nozzle. Abrasive water jet machining (AWJM) will create a compressive stress layer (at least 0.003 inches thick) on the surface of the Alloy 600 nozzle in areas adjacent to the IDTB weld and at the roll transition location where elevated tensile stresses may be present. Since the stresses created by the AWJM process are compressive, PWSCC is not expected in this layer. An undetected flaw 0.002 inches deep (twice the maximum particle depth of the AWJM abrasive material) was assumed, which leaves a compressive stress layer 0.001 inches thick. General corrosion of a 0.001 inch thick compressive layer was estimated to take 12.5 EFPY. Once the compressive stress layer is removed by general corrosion, it was assumed that PWSCC would initiate immediately. It was estimated to take 2.3 EFPY for the PWSCC crack to propagate to 75% of the original Alloy 600 nozzle wall thickness. Therefore, the total estimated life of the repair is 14.8 EFPY.

The 30 year life is predicted based on the as-left J-groove flaw evaluation. The 14.8 EFPY is based on a separate PWSCC evaluation in the exposed original Alloy 600 nozzle. The overall acceptable life of the repair design is based on the most limiting life predicted amongst the weld anomaly analysis, the as-left J-groove analysis and the PWSCC evaluation of the original Alloy 600 nozzle, which is 14.8 EFPY.

U.S. Nuclear Regulatory Commission Page 7 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Question 8 Discuss the mockup for the proposed repair in general and specifically for the welding and inspection qualification.

Response 8 AREVA, in support of 124 similar repairs, has performed many qualifications using mockups since the initial inner diameter temper bead (IDTB) control rod drive mechanism nozzle repairs at Oconee Nuclear Station in 2001. During these repair evolutions, the site crew performs training on mockups for each of their respective specialties, i.e., machinists train on machining mockups, welders train on welding mockups, and non-destructive examination (NDE) personnel train on NDE mockups. At least one mockup exercise, from start to finish, is performed just prior to deployment to ensure the crew understands the entire repair process and their role in the process. This training occurs at AREVAs Lynchburg, Virginia training center with environmental conditions similar to those expected in the field.

An IDTB weld repair NDE mockup was fabricated to replicate expected configuration. It contains a series of electrical-discharge machining (EDM) notches at the triple point to simulate the triple point anomaly at various depths into the nozzle wall and cracking at the IDTB weld to low alloy steel interface. It also contains flat bottom holes drilled from the mockup outer diameter so that the hole is normal to the surface to simulate under bead cracking, lack of bond, and lack of fusion.

An Inconel calibration block is used and contains a series of EDM notches at nominal depths of 10%, 25%, 50%, and 75% deep from both ID and OD surfaces in both the axial and circumferential orientation. The block also contains 1/4T, 1/2T, and 3/4T deep end holes and side drilled holes that are used for calibration.

This is the same mockup used for the procedure qualification for the Davis Besse CRDM nozzle repairs in 2010.

Question 9 The RR stated that ASME Code Case N-638-1 will be implemented.

Question 9.a.

Clarify whether Shearon Harris, Unit 1, has already adapted ASME Code Case N-638-1 (approved by the NRC in Regulatory Guide (RG) 1.147, Rev. 15, for use with conditions) during the third 10-year Inservice Inspection interval.

U.S. Nuclear Regulatory Commission Page 8 of 8 HNP-12-060 Enclosure Relief Request I3R-09 RAI Response Response 9.a.

Yes, HNP adopted ASME Code Case N-638-1 in the Third Interval Inservice Inspection Program submittal to the NRC as HNP-08-038 (ADAMS Accession No. ML081330463).

Later revisions of the code case have not been adopted.

Question 9.b.

If the answer to RAI-9.a. is: No, then clarify why ASME Code Case N-638-4 (approved by the NRC in RG 1.147. Rev. 16, for use with conditions) that has superseded the previous versions of that code case is not being used.

Response 9.b.

Not applicable. The response to question 9.a. was yes.