DCL-15-106, ASME Section XI Inservice Inspection Program Request for Relief NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements

ASME Section XI Inservice Inspection Program Request for Relief NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements

ML15246A244
Person / Time
Site:	Diablo Canyon
Issue date:	09/03/2015
From:	Allen B Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-15-106
Download: ML15246A244 (10)

Text

Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Vice President, Nuclear Services Mail Code 104/6 P. 0. Box 56 September 3, 2015 Avila Beach, CA 93424 805.545.4888 PG&E Letter DCL-15-106 Internal: 691.4888 Fax: 805.545 .6445 U.S. Nuclear Regulatory Commission 10 CFR 50.55a ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant (DCPP) Unit 1 and Unit 2 ASME Section Xllnservice Inspection Program Request for Relief NDE-FWNS-U1/U2 to Allow Use of Alternate Examination Volume Coverage Requirements

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.55a(g)(5)(iii), Pacific Gas and Electric Company (PG&E) hereby requests NRC approval of lnservice Inspection Request for Relief NDE-FWNS-U1/U2 for the Diablo Canyon Power Plant Unit 1 and Unit 2 third lnservice Inspection Interval.

Relief is requested from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, for examination coverage of Class 2 feedwater nozzle-to-vessel welds. The details of the proposed request are enclosed.

PG&E requests approval of NDE-FWNS-U1/U2 by September 3, 2016.

PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

If you have any questions or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720.

Sincerely, s1.7.n 5 ' 4a-rntt/4231/50033145 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator John P. Reynoso, NRC Acting Senior Resident Inspector Siva P. Lingam, NRR Project Manager Gonzalo L. Perez, Branch Chief, California Department of Public Health State of California, Pressure Vessel Unit A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

Enclosure PG&E Letter DCL-15-1 06 10 CFR 50.55a Request NDE-FWNS-U1/U2 Relief Request in Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

1. ASME Code Component(s) Affected Diablo Canyon Power Plant (DCPP), Unit 1 and Unit 2, ASME Section XI , Code Class 2 steam generator (SG) feedwater nozzle-to-shell welds (two welds) are affected:

Line Code Item Outage Description Weld Number Size Cat. No. Examined (inch)

Unit 1 Feedwater Nozzle-to-C-B C2.21 FW-13.03.01 16 1R17 Shell Weld Unit 2 Feedwater Nozzle-to-C-B C2.21 FW-13.03.01 16 2R17 Shell Weld

2. Applicable Code Edition and Addenda The DCPP Unit 1 and Unit 2 third intervallnservice Inspection (lSI) Program Plan is based on the American Society of Mechanical Engineering (ASME) Boiler and Pressure Vessel Code,Section XI, 2001 Edition, with 2003 Addenda.

3. Applicable Code Requirement ASME Section XI, Table IWC-2500-1, Category C-B, Item No. C2.21 requires that SG feedwater nozzle-to-shell welds be volumetrically examined once during the lSI interval. Essentially 100 percent of the inner one-third volume of the weld, and adjacent base material, is to be examined in accordance with the requirements of Appendix I, 1-2120. The applicable examination volume is defined by Figure IWC-2500-4(a) and the examination is performed per the rules of ASME Section V, Article 4, as supplemented by Table 1-2000-1 .

4. Impracticality of Compliance The Unit 1 and Unit 2 SG feedwater nozzle-to-shell weld configurations are such that essentially 100 percent coverage of the ASME Code required examination 1

Enclosure PG&E Letter DCL-15-1 06 volume from the outside diameter (OD) is not feasible, as determined during the third interval examinations conducted in the Diablo Canyon Unit 1 and Unit 2 seventeenth refueling outages (1 R17 and 2R17).

Background Information The DCPP replacement SGs shell and nozzle forgings are fabricated from SA-508 Grade 3 Class 2 material with a nominal shell thickness of approximately 3.50 inches. The feedwater nozzles intersect the shell cylinder at a right angle and are joined by a weld extending concentrically around the nozzle forging and through the full thickness of the shell. The weld joint design is an unequal depth double U - groove design with an included groove angle of 7 degrees. The nozzle-to-shell welds were made using high strength filler metals (i.e., E9018M, ER80S-2, and S3NiMo1) whose composition and mechanical properties are similar to the joined base metals.

The subject welds were examined in 1R 17 and 2R 17 to the extent practicable using a combination of 35, 45, and 60 degree angled shear waves and zero degree longitudinal waves. The 35 and 45 degree angles were used for radial-out examinations in order to achieve the maximum possible coverage of the Code specified examination volume (Note: 60 degrees is not able to interrogate the Code examination volume in the radial-out direction due to the restricted setback due to the nozzle boss configuration). Forty-five and sixty degree angles were used for radial-in and circumferential scan examinations. No flaws were detected in any of the examinations of the subject welds.

The following table summarizes the exam volume coverage attained for both welds in each of the four scan directions and a combined average value.

Although the entire Code exam volume was interrogated by the zero degree longitudinal wave scan, coverage values are based on 35, 45, and 60 degree angles since they would be expected to detect service induced planar flaws emanating from the inside surface. The sketches that are presented at the end of this Enclosure illustrate the coverage for each of the inspection angles and directions used to determine coverage values.

Radial-In Radial-Out Circ. Scan Circ. Scan Steam Scan Scan Exam Volume Exam Volume Combined Generator Volume Volume CoverageCW Coverage CCW Coverage2 Coverage Coverage 1 Direction Direction 1-3 (Unit 1) 100%> 39%> 100%> 100°/o 84%>

2-1 (Unit 2) 100°/o 17%> 100%> 100°/o 79o/o 1

Combined coverage average for 35 and 45 degree angles.

2 The reported combined coverage value is an equal weighted average of the coverage values from each of the four scan directions.

2

Enclosure PG&E Letter DCL-15-1 06 Impracticality The compound curvature of the nozzle forging boss to shell contour transition zone and the forging design diameter constitute geometric restrictions that preclude full examination volume coverage from the outside surface. The coverage limitations are associated with the radial-out oriented scans.

An inherent design characteristic of the flange type nozzle configuration is that there is often insufficient setback distance for the radial-out scan beams to cover the entire Code specified exam area at the inside surface.

5. Burden Caused by Compliance "Essentially 100 percent" coverage of the exam volume from the outside surface would require redesign of the SG to move the weld farther back from the nozzle reinforcement or eliminate the weld by integrally incorporating the nozzle into the shell. Either of these two modifications would effectively result in performing major redesign and rework or replacement of the entire SG to accommodate full coverage of the exam are~ as specified by Code.

Performing examinations from the inside diameter of the SGs would require accessing the secondary side of the generators which involves substantial effort to remove the manway cover, making provisions for personnel access into a confined space, and work in a high risk foreign material exclusion area.

These efforts required to attain a small incremental increase in coverage would incur increased personnel radiation exposure and an increase in personnel safety risk due to work in a difficult to access and highly constrained work environment without a commensurate increase in examination effectiveness.

6. Proposed Alternative and Basis for Use PG&E proposes that the alternative ultrasonic examinations conducted to the maximum extent practicable from the outside surface provide reasonable assurance that the structural integrity of the subject welds remains intact.

The 1R 17 and 2R 17 examinations were implemented to the extent practicable using manual scan techniques and small footprint search units in effort to attain the greatest possible coverage of the required examination volume. The volume examined on both of the subject feedwater nozzle-to-shell welds includes the weld and surrounding base material near the inside surface of the weld joint, which are typically the highest stress regions and where degradation would likely manifest, should it occur.

The radial-in and circumferentially oriented angle beam scans fully interrogated the ASME XI Code exam volume, whereas, the radial-out scans covered a portion of the volume. Examination of ferritic materials from a single side of the weld has been demonstrated as effective in studies (Reference 1) and by 3

Enclosure PG&E Letter DCL-15-1 06 successful single side ferritic ASME XI, Appendix VIII qualifications per the Performance Demonstration Initiative program. Therefore, it is expected that the ultrasonic techniques employed on the DCPP feedwater nozzle-to-shell welds would have detected structurally significant flaws if extant within the examination area.

The 1R 17 and 2R 17 ultrasonic examinations with combined coverage values of approximately 84 percent and 79 percent for the selected Unit 1 and Unit 2 subject welds, respectively, provide reasonable assurance that the structural integrity of these welds remains intact and provide an acceptable level of quality and safety.

Potential Failure Consequences A failure of the feedwater nozzle-to-shell weld could result in a loss of feedwater to a SG. Depending on the size of the postulated break (leak) the specific consequences will vary. At the smallest end of the break size spectrum, the feedwater system would be capable of maintaining SG level through normal makeup. Larger break sizes would result in depressurization of the SG and loss of heat transfer capability. The worst case consequence would occur if the nozzle-to-shell weld was to suffer 360 degree circumferential cracking. In this case, the break is bounded by the feedwater line break assumed in the DCPP design basis analysis.

Essentially no change to overall plant safety is expected due to implementation of the proposed alternative in lieu of the Code requirement. This assumption is based on the effectiveness of ultrasonic examination on ferritic material as previously described, and little or no historical occurrence of large service induced planar flaws in this type of weldment.

7. Duration of Proposed Alternative Relief is requested for the remainder of the DCPP Unit 1 and Unit 2 third lSI intervals. The DCPP Unit 1 third inspection interval nominally ended on May 6, 2015. The DCPP Unit 2 third inspection interval is nominally scheduled to end on March 12, 2016. Actual end dates of the interval are dependent on the completion dates of the 19th refueling outages for each unit, in accord with ASME Section XI, paragraph IWA-2430(d)(1).

As stated in the relief request, the third interval for Unit 1 nominally ended May 8, 2015, the 30th anniversary of the commercial operation date for the unit. However, per Section XI paragraph IWA-2430(d)(1 ), "Each inspection interval may be reduced or extended by as much as one year." Paragraph IWA-2430(d)(3) states, "That portion of an inspection interval described as an inspection period may be reduced or extended by as much as one year to enable an inspection to coincide with a plant outage." For Unit 1, the interval is being extended past the nominal end date to November 6, 2015, to coincide with the dates of 1R 19, and may be further continued if necessary until May 8, 2016.

4

Enclosure PG&E Letter DCL-15-1 06 Therefore, this submittal is timely. All subject examinations or their approved alternatives would be credited for the third inspection interval only.

For alternative requests based on impracticality, submittals must be made no later than 12 months after the expiration of the interval for which relief is sought per 10 CFR 50.55a(g)(5)(iii)/(iv).

8. Precedents This request is essentially the same as DCPP request NDE-25.2R8 approved in an NRC letter dated September 29, 1999, for the second lSI interval. It is also similar to Relief ISI-6 for Calvert Cliffs Nuclear Power Plant (TAC numbers MA-9404 and MA-9405). ISI-6 was approved by the NRC in a letter dated July 27, 2001.

9. References

1. P.G. Heasler and S.R. Doctor, 1996. Piping Inspection Round Robin, NUREG/CR-5068, PNNL-1 0475, U.S. Nuclear Regulatory Commission, Washington, DC 5

Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration RSG 1-3 & 2-1 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 45° and 60° Radial-In Scans Nozzle Forging 100% Coverage 45° and 60° Radial-In Scans

~ = Volume Examined 6

Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration Steam Generator Shell RSG 1-3 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 35° and 45° Radial-Out Scans 35° and 45° (35° Position Illustrated)

Nozzle Forging 67°/o Coverage 35° Radial-Out Scan 11 °/o Coverage 45° Radial-Out Scan

~ = Volume Examined With 35°

• = Volume Examined With 35° & 45° 7

Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration Steam Generator Shell RSG 2-1 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 35° and 45° Radial-Out Scans 35° and 45° (35° Position Illustrated)

Nozzle Forging 34°/o Coverage 35° Radial-Out Scan 0°/o Coverage 45° Radial-Out Scan

~ = Volume Examined With 35° 8

Enclosure PG&E Letter DCL-15-1 06 DCPP NDE-FWNS-U1/U2 Coverage Illustration Steam Generator Shell RSG 1-3 & 2-1 Feedwater Nozzle-to-Shell Weld Representation at Nozzle Top-Dead-Center Coverage 45° and 60° Clockwise and Counter-Clockwise Scans Nozzle Forging 100% Coverage 45° and 60° Clockwise and Counter-Clockwise Scans

~=Volume Examined 9