catawba, units 1 and 2 - third ten-year inservice ... · "status of the ultrasonic examination...

' ( -, DUKE

ENERGY11

CNS-17-024

May 25, 2017

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject: Duke Energy Carolinas, LLC (Duke Energy) Catawba Nuclear Station, Unit 1 and 2 Docket Number 50-413 and 50-414 Relief Request # 17-CN-001

Tom Simril

Vice President

Catawba Nuclear Station

Duke Energy

CN01VP I 4800 Concord Road

York, SC 29745

o: 803.701 .3340

f: 803. 701 .3221

[email protected]

Third Ten-Year lnservice Inspection Plan , Limited Volumetric Examinations

Pursuant to 10 CFR 50.55a(g)(5)(iii), Duke Energy hereby requests NRC approval of the following relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI , Rules for lnservice Inspection of Nuclear Power Plant Components, 1998 Edition with 2000 Addenda. Please find enclosed the Relief Request #17-CN-001 associated with Category B-J pressure retaining welds in piping for branch pipe connection welds of NPS 4 or larger.

There are no regulatory commitments contained in this letter or its attachments.

If you have any questions concerning this material , please call Dustin Yang at (803) 701-3084.

Sincerely,

Tom Simril Vice President, Catawba Nuclear Station

Enclosure 1. Relief Requested in Accordance with 10 CFR 50.55a(g)(5)(iii) for Volumetric Examination of Class 1 Branch Pipe Connection Welds NPS 4 or Larger

www.duke-energy.com

Document Control Desk May 25, 2017

xc (with enclosure):

C. Haney Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257

J . D. Austin Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station

M. Mahoney (addressee only) Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop 0-8H4A Rockville , MD 20852

Enclosure 1

Duke Energy Carolinas, LLC

Catawba Nuclear Station, Units 1 and 2

Relief Request Serial #17-CN-001

Relief Requested in Accordance with 10 CFR 50.55a(g)(5)(iii) for Volumetric Examination of Class 1 Branch Pipe Connection Welds NPS 4 or Larger

Catawba Nuclear Station, Units 1 and 2 Relief Request #17-CN-001 Enclosure 1 Page 2 of 6

1.0 ASME Code Component(s) Affected

1.1 ASME Class 1, Category B-J Pressure Retaining Welds in Piping, Item B9.31 - Branch Pipe Connection Welds NPS 4 or Larger

1.1.1 Unit 1 Welds for Which Relief is Requested:

Weld #1 NC22-WN7 (Summary No. C1 .B9.31 .0001 ), 14" Diameter Pressurizer Surge Line to Reactor Coolant Loop 1 B Hot Leg

Weld #1 NC22-WN8 (Summary No. C1 .B9.31 .0002), 12" Diameter Residual Heat Removal Pump 1A Line to Reactor Coolant Loop 1 B Hot Leg

Weld #1 NC24-WN9 (Summary No. C1 .B9.31 .0003), 6" Diameter Safety Injection Pump 1 B Line to Reactor Coolant Loop 1 A Hot Leg

1.1.2 Unit 2 Welds for Which Relief is Requested:

Weld #2NC11-WN7 (Summary No. C2.B9.31.0001), 14" Diameter Pressurizer Surge Line to Reactor Coolant Loop 2B Hot Leg

Weld #2NC11-WN8 (Summary No. C2.B9.31.0002), 12" Diameter Residual Heat Removal Pump 2A Line to Reactor Coolant Loop 2B Hot Leg

Weld #2NC13-WN9 (Summary No. C2.B9.31.0003), 12" Diameter Residual Heat Removal Pump 2B Line to Reactor Coolant Loop 2C Hot Leg

1.1 .3 The above welds connect SA-182 F304N forged stainless steel branch piping nozzles to the centrifugally cast stainless steel , SA-351 CF8A, main coolant piping.

2.0 Applicable Code Edition and Addenda

ASME Boiler and Pressure Vessel Code, Section XI, 1998 Edition with the 2000 Addenda

3.0 Applicable Code Requirements

3.1 The ASME Code, Section XI, IWB-2500, Table IWB-2500-1 Examination Category B-J, Pressure Retaining Welds In Piping, Item B9.31 applies to the welds identified in this request. For Item B9.31, a surface and volumetric examination is required once each interval.

3.2 Figure IWB-2500-11 is applicable and specifies that surface examination be performed on area A-B and a volumetric examination be performed on volume C-D-E-F in accordance with Figure IWB-2500-8.

3.3 IWA-2232 requires that ultrasonic examinations be conducted in accordance with Appendix I. 1-2220 requires that for welds in piping, ultrasonic examination procedures, equipment, and personnel used to detect and size flaws in piping welds shall be qualified by performance demonstration in accordance with Appendix VIII.

3.4 ASME Code, Section XI , Appendix VIII, Vlll-3110(c) specifies that the requirements of Appendix Ill, as supplemented by Table 1-2000-1 shall be met.


3.5 Relief is requested from the following requirements for the welds identified in Sections 1.1.1 and 1.1.2:

1. The requirement to perform a volumetric examination of volume C-D-E-F in accordance with Figure IWB-2500-8.

2. The requirement to perform a procedure demonstration in accordance with Appendix I, 1-2220.

4.0 Impracticality of Compliance

4.1 ASME Code, Section XI, Appendix 111, 111-2200 requires, in part, that "personnel who perform recording or determine which indications are to be recorded in accordance with 111-4510 shall have successfully completed the qualification requirements of lll-2200(a), for the procedure to be used for the examination. The qualification shall include demonstrated proficiency in discriminating between flaw indications and indications of geometric or metallurgical origin ."

4.1.1 Compliance with the above requirement is considered impractical for the following reasons:

1. During the previous (second) inservice inspection interval, procedure demonstration was attempted , as documented in Relief Request 04-CN-001 , Revision 1 (ADAMS Accession No. ML051230324). This demonstration revealed that none of the known flaws within the inner 1/3 volume of the weld were detectable in the cast stainless steel mock-up. Therefore, Duke Energy concludes that the requirements of Appendix Ill, 111-2200 cannot be met and that no effective ultrasonic examination of the inner 1 /3 volume (ASME Code Figure IWB-2500-8, Volume C-D-E-F) can be performed on these welds.

2. It is Duke Energy's opinion that procedure demonstration capable of reliably detecting known flaws within the inner 1 /3 volume of these welds is still not practical using available technology.

4.2 ASME Code, Section XI, Division 1, Appendix 111, 111-4420 requires that "The examination shall be performed using a sufficiently long examination beam path to provide coverage of the required examination volume in two-beam path directions. The examination shall be performed from two sides of the weld , where practicable, or from one side of the weld , as a minimum." ASME Code, Section XI , Division 1, Appendix Ill , 111-4430 requires that 'The angle beam examination for reflectors transverse to the weld shall be performed on the weld crown on a single scan path to examine the weld root by one half V path in two directions along the weld". These requirements cannot be met because the weld configurations do not allow scanning of essentially 100% of the required examination volume (Figure IWB-2500-8, Volume C-D-E-F) in two axial or circumferential directions. In order for scanning to meet the requirements of 111-4430, the weld joint would have to be redesigned , which is impractical.

4.3 The use of radiography to meet the volumetric examination requirement of the ASME Code, Section XI is considered impractical because access is no longer possible for film placement on the interior of the piping .


4.4 Selection of alternative welds for which essentially 100% of the required volume specified in Figure IW8-2500-8, Volume C-D-E-F is not possible because all of the 89.31 welds at both Units 1 and 2 connect forged stainless steel branch piping nozzles to the centrifugally cast stainless steel , SA-351 CF8A, main coolant piping. There are (2) NPS 4 branch connection welds at each unit that connect the nozzle to the outside surface of the pipe that would allow for limited volumetric examination of Figure IW8-2500-8, Volume C-D-EF, but these welds were not selected for examination because the estimated volumetric coverage obtainable for these welds is no greater than 20% of the required volume C-DE-F.

5.0 Proposed Alternative

5.1 As an alternative , Duke Energy performed a best effort ultrasonic examination using 70° and 60° refracted longitudinal wave search units manufactured by RTD to interrogate the outer 2/3 of the weld volumes and the adjacent base material. EPRI Report TR-107481 , "Status of the Ultrasonic Examination of Reactor Coolant Loop Cast Stainless Steel Materials," identify these transducers as having the highest signal to noise ratio when looking for the upper extremities of deep cracks. The calibration was performed using an existing calibration block made of SA-351 CF8A centrifugally cast stainless steel with axially and circumferentially oriented side-drilled holes. The coverage obtained during examination of the alternative volumes is documented in Enclosure 2.

5.2 Duke Energy also performed the following examinations during the 3rd lnservice Inspection Interval:

1. System leakage testing and VT-2 visual examinations in accordance with IW8-2500, Table IW8-2500-1 , Examination Category 8-P for all of the welds listed in Sections 1.1.1 and 1.1.2 of this request.

2. Surface examinations (liquid penetrant) in accordance with IW8-2500, Table IW8-2500-1, Examination Category 8-J, Item 89.31 , for the following welds:

a. Weld #1 NC24-WN9 (Summary No. C1 .89.31.0003)

b. Weld #1 NC22-WN7 (Summary No. C1 .89.31 .0001)

c. Weld #2NC13-WN9 (Summary No. C2.89.31.0003)

Note: Surface examinations were not performed on the other welds listed in Sections 1.1.1 and 1.1.2 of this request because Duke Energy had implemented Code Case N-663 during the 3rd Interval in lieu of performing the surface examination of these welds.

6.0 Justification for Granting Relief:

6.1 Current UT technology is not capable of reliably detecting or sizing flaws in cast austenitic stainless steel. Duke Energy will remain cognizant of industry efforts to improve the ultrasonic examination of cast stainless steel piping and will consider such technology when it is commercially available and proven by demonstration that the technology can discriminate between geometry and cracks.


6.2 These branch connection welds were fabricated by Southwest Fabricating and Welding Company in accordance with the requirements of ASME Section Ill , 1974 Edition with no addenda , and radiography was performed in the fabricator's shop where access was available from the inside of the welds for placement of the radiographic film. A liquid penetrant examination of the weld root pass and accessible surfaces of the finished weld surfaces was also performed during fabrication.

6.3 Preservice nondestructive examinations were performed on these welds prior to initial plant operation 1.

6.4 lnservice nondestructive examinations have been performed on these welds during the First, Second, and Third 10-Year lnservice Inspection Intervals, and no flaws were detected during these examinations 1.

6.5 lnservice nondestructive (liquid penetrant) examinations were performed on a sample of these welds during the Third 10-Year Inspection Interval, and no flaws were detected during these examinations.

6.6 In addition to the proposed alternative in 5.0 above, there are other activities which provide a high level of confidence that any leakage greater than 1.0 gpm, if it were to occur through these areas/welds , would be detected:

1. Leakage would be detected by the Reactor Coolant System (RCS) leakage calculation, which is performed at least once every three days under procedure PT/1, 2/A/4150/01 D, "NC System Leakage Calculation". This RCS leakage calculation is a requirement of Technical Specification 3.4.13, "RCS Operational Leakage".

2. Leakage is also evaluated in accordance with Technical Specification 3.4.15, "RCS Leakage Detection Instrumentation", and can be detected through one or more of the following :

• The Reactor Building atmosphere particulate radioactivity monitor

• The containment floor and equipment sump level monitors

• The incore instrument sump level alarm

• The containment ventilation unit condensate drain tank level monitor

6. 7 Duke Energy believes that the alternative examinations detailed in 5.0 provide reasonable assurance of the continued structural and leak-tight integrity of these Reactor Coolant System pressure boundary welds. Any significant RCS leakage through these areas/welds would be detected by leakage detection monitors and alarms described above.

1 Preservice and inservice examinations performed on several of these welds prior to the 3rd lnservice Inspection Interval were described in Relief Request 04-CN-001 , Revision 1 (ADAMS Accession No. ML051230324 ).

Catawba Nuclear Station, Units 1 and 2 Relief Request #17-CN-OO 1 Enclosure 1 Page 6 of 6

7.0 Duration of Proposed Alternative

This request is proposed for the Catawba Units 1 and 2, 3rd lnservice Inspection Intervals identified below.

Catawba Nuclear Station, Unit 1 Third lnservice Inspection Interval :

Start Date: June 29, 2005 End Date: June 29, 20162

Catawba Nuclear Station, Unit 2 Third lnservice Inspection Interval:

Start Date: October 15, 2005 End Date: October 15, 20162

8.0 References

8.1 Letter from Dhiaa Jamil , Duke Energy to NRC, dated April 21 , 2005, Request for Relief Number 04-CN-001 , Revision 1, (ADAMS, Accession No. ML051230324)

8.2 ASME Boiler and Pressure Vessel Code, Section XI , 1998 Edition with the 2000 Addenda

8.3 EPRI Report TR-107481 , "Status of the Ultrasonic Examination of Reactor Coolant Loop Cast Stainless Steel Materials"

8.4 "NUREG/CR-6594 "Evaluation of Ultrasonic Inspection Techniques for Coarse-Grained Materials"

2 The Third Interval dates were adjusted in accordance with IWA-2430(d)(1 ).

Catawba Nuclear Station, Units 1 and 2

Relief Request #17-CN-001

Enclosure 2

Ultrasonic Volumetric Coverage Computations

Catawba Nuclear Station, Units 1 and 2 Relief Request #17-CN-OO 1 Enclosure 2 Page 2 of 6

Cl.89.31.0001 1NC22..WN7

Pipe Surtace 1

100%Axial Coverage Obtained fromS1

11 .7%CWICCW Coverage Obtain

' I I

A

B

.394 sq. in coverage I 3.37 sq. in total volume = 11 .7% coverage CW/CCW

F E

D =Missed CW I CC1N Coverage

Pipe Surface 1 (S1) is the 1 B Reactor Coolant System Hot Leg

Nozzle Surface 2 (S2) is the 14" Diameter Pressurizer Surge Line Nozzle

No Volumetric Coverage is Claimed for Figure IWB-2500-8, Volume C-0-E-F

Volumetric Coverage for Volume A-B-0-C is Calculated Below:

Nozzle Surface2

Clockwise (CW) Circumferential Coverage Claimed=

Counterclockwise (CCW) Circumferential Coverage Claimed =

Axial Coverage Claimed from S1 =

11 .7%

11.7%

100.0%

The examination procedure requires that axial scans be performed only from one s·ide , if access from both sides is not possible. Access from the nozzle side was not possible, so the aggregate coverage calculations are based on ( 1) axial and (2) circumferential scans.

Aggr egate Coverage Claim ed f or Volume ABDC = [11.7 + 11.7 + 100.0]/3 = 41.1%


C l.R'l..ll.0002 + l "Ol-V'l'.N8

I" x 1"

60" -- --1-----

c 2.74 :.q. in cuvu~ I 3.37 =<i- in tclal vol~ •81.3%=-go6 _

F

B

D • INHCd S1 Axial c.o.en.ge

E

1s• of Rupture Reolr:llnl ~ Weld e<igtn = 40' 6l.S'to l.eOQln x IOO'Y. \tlun111 • 37.0)(, Length x 111.3% ~ = 93 01' 81 ~ O>l.«11g41

- o·--i.-- =:....--+<;> B

117"!.DN/~ 1 Pipe ~1

Ccwe<age Ol>lalned D • 11.issedc:"/V I CCINc.....aga

L 394sq111c;:o...,.age1 c~,-----_.......,, o 3.37 sq 10 IW (v.l\.rne : • 11.7%~ F:

- --0 - -


Nozzle Surface 2 (S2) is the 12" Diameter Residual Heat Removal Pump 1A Line Nozzle

No Volumetric Coverage is Claimed for Figure IWB-2500-8, Volume C-D-E-F

Volumetric Coverage for Volume A-B-D-C is Calculated Below:

Clockwise (CW) Circumferential Coverage Claimed = 11 .7%

Counterclockwise (CCW) Circumferential Coverage Claimed = 11 . 7%

Axial Coverage from S1 was partially limited by a rupture restraint that obstructed examination for approximately 37.5% of the weld length .

Axial Coverage Claimed from S1 = 93.0%

The examination procedure requires that axial scans be performed only from one side, if access from both sides is not possible . Access from the nozzle side was not possible, so the aggregate coverage calculations are based on ( 1) axial and (2) circumferential scans.

Aggr egate Coverage Claimed for Vo lume ABDC = [11.7 + 11.7 + 93.0]/3 = 38.8%


CI.119.3Ul003 l O+WN9

c 2.8.\ sq. inCO'l<!nlgQ/'i'-• --~'<-----t--"f' 137 sq. In l<ltll YClkJl1W

&&3%CICNC!fllge :

F E

c li>-+-- ->r--

~: ~ ::.u-~ : •11 7".l.cct- :

D

~· [

..l:>----~~--~ F E


Nozzle Surface 2 (S2) is the 6" Diameter Safety Injection Pump 1 B Line Nozzle



Clockwise (CW) Circumferential Coverage Claimed=

Counterclockwise (CCW) Circumferential Coverage Claimed=

Axial Coverage from S1 was partially limited by the weld geometry.

Axial Coverage Claimed from S1 =

11.7%

11 .7%

84.3%

The examination procedure requires that axial scans be performed only from one side, if access from both sides is not possible. Access from the nozzle side was not possible, so the aggregate coverage calculations are based on (1) axial and (2) circumferential scans.

Aggregate Coverage Claimed for Volume ABDC = [11.7 + 11.7 + 84.3]/3 = 35.9%


Summary #C2.B9.31 .0001 (Component ID #2NC11-WN7)

Summary #C2.B9.31 .0002 (Component ID #2NC11 -WN8)

Summary #C2 .B9.31 .0003 (Component ID #2NC13-WN9)

Axial Coverage from S1 :

Area A-B-D-C = 2.05" x 1.55" = 3.18 in

Unscanned Area= 2.05" x 0.70"/2 = 0.72 in2

Area A-B-C-D Coverage Obtained =

[(3.18 in2- 0.72 in2}/3. 18 in2

] x 100% =77.4%

1 55· J

2NC11-WN7:

A I I

! ' I

! ' I cl-- ·-·-· _.\_. _ . c , I

\ I

' ' F I 2.05" IE _.,. ____ __,._

Pipe Surface 1 (S1) is the 2B Reactor Coolant System Hot Leg Nozzle Surface 2 (S2) is the 14" Diameter Pressurizer Surge Line

2NC11-WN8:

Pipe Surface 1 (S1) is the 2B Reactor Coolant System Hot Leg Nozzle Surface 2 (S2) is the 12" Diameter Residual Heat Removal Pump 2A Line

2NC13-WN9:

Pipe Surface 1 (S1) is the 2C Reactor Coolant System Hot Leg Nozzle Surface 2 (S2) is the 12" Diameter Residual Heat Removal Pump 2B Line

Volumetric Coverage (Applicable for Welds 2NC11-WN7, 2NC11-WN8 and 2NC13-WN9):




Clockwise (CW) Circumferential Coverage Claimed = 0%

Counterclockwise (CCW) Circumferential Coverage Claimed = 0%

Note: No circumferential coverage could be obtained due to the weld geometry.

Axial Coverage Claimed from S1 = 77.4%

Note: Axial Coverage from S1 was partially limited by the weld geometry.

The examination procedure requires that axial scans be performed only from one side, if access from both sides is not possible. Access from the nozzle side was not possible, so the aggregate coverage calculations are based on (1) axial and (2) circumferential scans.

Aggregate Coverage Claimed for Volume ABDC = [O + 0 + 77.4]/3 = 25.8%

catawba, units 1 and 2 - third ten-year inservice ... · "status of the ultrasonic examination...

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