Text

Relief Request 01-007
	ML021210076
Person / Time
Site:	McGuire
Issue date:	04/18/2002
From:	Barron H; Duke Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	01-007
	Download: ML021210076 (177)
	v • d • e

Duke Power Company A Duke Energy Company Wr Power.

McGuire Nuclear Station A Dukr EeW Compan MG01VP 12700 Hagers Ferry Rd.

Huntersville, NC 28078-9340 H. B. Barron Vice President, McGuire (704) 875-4800 OFFICE Nuclear Generation Department (704) 875-4809 FAX April 18, 2002 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001

Subject:

McGuire Nuclear Station, Unit 1 Docket No. 50-369 Relief Request 01-007 Pursuant to 10CFR50.55a(g)(5)(iii), Duke Energy Corporation requests relief from certain ASME Section XI Code requirements as described in the attached Relief Request No.01-007.

This relief request addresses cases of limited examination coverage from inspections performed during the end of fuel cycle (EOC) 14 for Unit 1. This request is applicable to the Second 10-year Interval Inservice Inspection Program Plan. The 1989 Edition of the ASME Section XI Code contains the applicable requirements.

The enclosed relief request describes for each specified case that the ASME Code requirement is impractical. Each specific instance is described in detail, including a basis for why a reasonable assurance of structural integrity exists.

Although, submittal of Relief Request No. 0 1-007 satisfies a previous identified commitment,1 there are no additional commitments associated with this relief request.

Please direct any questions regarding this request to Norman T Simms of Regulatory Compliance at (704) 875-4685.

Sincerely, H. B. Barron Enclosure 7AOL 7 1

Letter, H. B. Barron to NRC, Dated July 11, 2001, Inservice Inspection Report

U.S. Nuclear Regulatory Commission April 18, 2002 Page 2 of 2 xc w/enclosure:

Mr. L. A Reyes US Nuclear Regulatory Commision, Region II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 Mr. R. E. Martin, Project Manager (addressee only)

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08G9 Washington, D.C. 20555 S. M. Shaeffer Senior NRC Resident Inspector McGuire Nuclear Station

bxc w/o encl: R.K. Rhyne R. Branch G.J. Underwood D.E. Caldwell G.D. Scarboro R.D. Klein (MGO1MM)

N.T. Simms bxc w/ encl:

RCG Files Master File # 1.3.2.13 NRIA File/ELL

ENCLOSURE RELIEF REQUEST NO.01-007

Serial No.01-007 Page 1 of 59 Duke Energy Corporation McGuire Nuclear Station Unit 1 SECOND 10-YEAR INTERVAL REQUEST FOR RELIEF NO.01-007 Duke Energy Corporation has determined that conformance with certain ASME Section XI Code requirements is impractical.

Therefore, pursuant to 10CFR50.55a (g)

(5)

(iii),

Duke Energy requests relief from applicable portions of the code.

Included in this request are sixteen welds: one Examination Category B-D weld, two Examination Category B-F Welds, one Examination Category B-H weld, seven Examination Category B-J welds, one Examination C-B weld and four Examination Category C-F-I welds.

The McGuire Unit-i Inservice Inspection Plan was written to the requirements of the 1989 Edition of ASME Section XI, no addenda.

The items in this Request for Relief were performed during refueling outage EOC-14.

Code Case N-460 applies to the examinations performed during this outage.

I.

System / Component(s) for Which Relief is Requested:

Examination Category B-D:

Pressurizer Nozzle-to-Vessel Weld for Safety Nozzle to Upper Head.

ID Number Item Number IPZR-14 B03.110.004 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

Serial No.01-007 Page 2 of 59 Figure IWB-2500-7(b).

ASME Section V, Article 4, T

424.1, Examination

Coverage, 1989 Edition with no Addenda.

"The volume shall be examined by moving the search unit over the examination surface so as to scan the entire examination volume."

Due to part geometry and actual physical barriers, obtaining 100%

coverage of the required volume is not possible with the existing limitations.

III.

Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to scan 100% of the examination volume.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 1, (Pages 1-9) due to single sided access, the examination coverage was limited to 69.07%.

In order to achieve more coverage the weld would have to be re-designed to allow scanning from both sides.

V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

Serial No.01-007 Page 3 of 59 VI.

Justification for the Granting of Relief:

Examination Category B-D:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-7(b) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

In addition, this weld was examined during installation using volumetric and surface NDE methods.

Ultrasonic examination of this weld was conducted using personnel, equipment and procedures qualified through the PDI Program for ferritic pressure vessel welds.

The qualifications were conducted on samples with access to both sides of the weld.

Therefore, Duke Energy Corporation does not claim credit for a single sided examination.

This weld is located on the NC system line from the pressurizer upper head to one of the NC relief valves.

This weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e.,

embrittlement) associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation. A leak at this weld would result in the following:

a) Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased Pressurizer enclosure temperature.

This parameter is continuously monitored by the Operations via an OAC alarm, and is periodically monitored by the System Engineer.

c) Increased input into the Ventilation Unit Condensate Drain Tank (VUCDT).

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

Serial No.01-007 Page 4 of 59 d) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

e) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in the pressurizer enclosure or containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII. References.

Information for Examination Category B-D affected weld: Pages 1-9 cover this weld.

B03.110.004

Serial No.01-007 Page 5 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-F:

iC Steam Generator NPS 4" or Larger Nozzle-to-Safe End Butt Weld.

ID Number Item Number lSGC-Inlet-W5SE B05.070.005 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

Figure IWB-2500-8 (c).

ASME Section XI, Appendix

III, Paragraph 111-4420, 1989 Edition with no addenda.

"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."

III.

Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to cover the required examination volume from two beam-path directions.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 2, (Pages 1-4) due to single sided access, the examination coverage was limited to 75.00%.

In order to achieve more coverage the weld would have to be re-designed to allow scanning from both sides.

Serial No.01-007 Page 6 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective NDE methods available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined during installation using volumetric and surface NDE methods.

The most effective ultrasonic technique for the examination of dissimilar metal welds uses refracted longitudinal waves.

The longitudinal wave is preferred as the austenitic weld metal and buttering create highly attenuative barriers to shear wave ultrasound.

The longitudinal wave is less affected by these difficulties.

However, the longitudinal wave is affected by mode conversion when it strikes the inside surface of the safe end or pipe at any angle other than a right angle to the surface.

The calculations below shows that a 450 refracted longitudinal wave striking the inside surface of a pipe will produce a 22.90 refracted shear wave in addition to the normally expected 450 reflected longitudinal wave.

Sin'

= (sin 450 x Vs)

+ VL

= (0.707 x 0.123)

+ 0.223 Where:

sin-' is the shear wave angle V, is the shear wave velocity of the stainless steel safe end/pipe material in inches/gsec.

Serial No.01-007 Page 7 of 59 VL is the longitudinal wave velocity of the stainless steel safe/pipe end material in inches/Jlsec.

As shown in the graph below, the mode conversion process creates two sound beams of differing intensities reflecting off of the inside surface.'

At incident angles greater than 300 the shear wave will predominate.

However, the shear wave is attenuated and scattered by the austenitic weld metal and the layer of buttering. The examination sensitivity is degraded to such an extent that any examination using the second sound path leg is meaningless.

Therefore, the two-beam path direction coverage requirement is impractical.

In order to obtain the required two-beam path direction coverage, welds would have to be re-designed to allow scanning from both sides.

Reflected Sound Beam Energy In Steel on A Free Face 0.9700........

/ --

o0 o..

60 --

- L-wa aa, W

S0.5 0 0.............

S h e 0.3 0 0 -- - -

0.10 0 -- - -

0.000 0.00° 15.20*

30.00' 50.00, 60.00.

70.00' 80.00' 89.20*

90.00° L-Wave Incident Angle The examination of Category B-F dissimilar metal welds was conducted in accordance with the requirements of ASME Section XI, Appendix III to the maximum extent practical.

Refracted longitudinal wave search units 1 Firestone, F. A. : Tricks with the Supersonic Reflectoscope, J. Soc. Nondestructive Testing, vol. 7, no. 2 Fall 1948.

yve Energy ar wave Energy

Serial No.01-007 Page 8 of 59 were used in accordance with NRC Information Notice No.

90-30: Ultrasonic Inspection Techniques for Dissimilar Metal Welds, May 1, 1990.

This weld is located on the safe end inlet (Hot Leg) nozzle on the 1C Steam Generator. The weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e.,

embrittlement) associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation. A leak at this weld would result in the following:

a) Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased Steam Generator enclosure temperature.

This parameter is continuously monitored by the Operations via an OAC alarm, and is periodically monitored by the System Engineer.

c) Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

d) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

e) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in the Steam Generator enclosure or containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Serial No.01-007 Page 9 of 59 Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category B-F affected welds:

Pages 1-4 cover this weld.

B05.070.005

Serial No.01-007 Page 10 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-F:

IC Steam Generator NPS 4" or Larger Nozzle-to-Safe End Butt Weld.

ID Number Item Number lSGC-Outlet-W6SE B05.070.006 Ii.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

Figure IWB-2500-8(c).

ASME Section XI, Appendix III, Paragraph 111-4420, 1989 Edition with no addenda.

"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."

III.

Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to cover the required examination volume from two beam-path directions.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 2 (Pages 5-8) due to single sided access, the examination coverage was limited to 75.00%.

In order to achieve more coverage the weld would have to be re-designed to allow scanning from both sides.

Serial No.01-007 Page 11 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective NDE methods available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined during installation using volumetric and surface NDE methods.

The most effective ultrasonic technique for the examination of dissimilar metal welds uses refracted longitudinal waves.

The longitudinal wave is preferred as the austenitic weld metal and buttering create highly attenuative barriers to shear wave ultrasound.

The longitudinal wave is less affected by these difficulties.

However, the longitudinal wave is affected by mode conversion when it strikes the inside surface of the safe end or pipe at any angle other than a right angle to the surface.

The calculations below shows that a 450 refracted longitudinal wave striking the inside surface of a pipe will produce a 22.90 refracted shear wave in addition to the normally expected 450 reflected longitudinal wave.

Sin-1 =

(sin 450 x Vs)

+ VL

= (0.707 x 0.123)

+ 0.223 Where:

sin-' is the shear wave angle Vs is the shear wave velocity of the stainless steel safe end/pipe material in inches/gsec.

Serial No.01-007 Page 12 of 59 VL is the longitudinal wave velocity of the stainless steel safe/pipe end material in inches/ptsec.

As shown in the graph

below, the mode conversion process creates two sound beams of differing intensities reflecting off of the inside surface. 2 At incident angles greater than 300 the shear wave will predominate.

However, the shear wave is attenuated and scattered by the austenitic weld metal and the layer of buttering.

The examination sensitivity is degraded to such an extent that any examination using the second sound path leg is meaningless.

Therefore, the two beam path direction coverage requirement is impractical.

In order to obtain the required two-beam path direction coverage, welds would have to be re-designed to allow scanning from both sides.

Reflected Sound Beam Energy In Steel on A Free Face

0)

LU c

n-0.00' 15.20' 30.00' 50.00' 60.00' 70.00' 80.00' L-Wave Incident Angle 89.20' 90.00' The examination of Category B-F dissimilar metal welds was conducted in accordance with the requirements of 2 Firestone, F. A. : Tricks with the Supersonic Reflectoscope, J. Soc. Nondestructive Testing, vol. 7, no. 2 Fall 194.

L-wave Energy

--N--Shear wave Energy

Serial No.01-007 Page 13 of 59 ASME Section XI, Appendix III to the maximum extent practical.

Refracted longitudinal wave search units were used in accordance with NRC Information Notice No.

90-30: Ultrasonic Inspection Techniques for Dissimilar Metal Welds, May 1, 1990.

This weld is located on the safe end outlet (Cold Leg) nozzle on the IC Steam Generator. The weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e.,

embrittlement) associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation. A leak at this weld would result in the following:

a) Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased Steam Generator enclosure temperature.

This parameter is continuously monitored by the Operations via an OAC alarm, and is periodically monitored by the System Engineer.

c) Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

d) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

e) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in the Steam Generator enclosure or containment, but could not specifically identify this weld as the source of leakage.

A containment entry

Serial No.01-007 Page 14 of 59 would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category B-F affected welds:

Pages 5-8 cover this weld.

B05.070.006

Serial No.01-007 Page 15 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-H:

Pressurizer Integrally Welded Attachment for Support Skirt to Lower Head.

ID Number Item Number lPZR-SKIRT B08.020.001A II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

Figure IWB-2500-13.

This weld was examined using the ultrasonic method.

See Relief for Alternative 00

001, and NRC Safety Evaluation Report dated 08/23/01 in Attachment 7 (Pages 1-13) for a full explanation of the weld configuration issue.

III.

Code Requirement from Which Relief Is Requested:

This weld was examined to the maximum extent practical per the requirements of Request for Alternative 00-001.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 3 (Pages 1-10) the examination coverage was limited to 75.16%.

The entire examination volume was covered 100% from at least one direction.

Serial No.01-007 Page 16 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-13 could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined during installation using surface NDE methods.

There is inadequate accessibility of the inside surface (surface C-D) of the Pressurizer Support Skirt Weld to perform the required surface examination.

Therefore, an ultrasonic examination will be used to inspect the inner examination surface from the skirt's exterior surface.

The ultrasonic procedure and the basic calibration block will conform to the requirements of ASME Section XI, Appendix I, 1989 Edition, and ASME Section V, Article 5,

1989 Edition.

This is the weld joining the pressurizer support skirt to the pressurizer lower head.

This weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e.,

embrittlement) associated with neutron bombardment.

This weld joins the pressurizer support skirt, a non pressure boundary component, to the lower pressurizer head.

Therefore, the weld serves no pressure boundary function. However, if a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation. A leak at this weld would result in the following:

Serial No.01-007 Page 17 of 59 a) Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased Pressurizer enclosure temperature.

This parameter is continuously monitored by the Operations via an OAC alarm, and is periodically monitored by the System Engineer.

c) Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

d) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

e) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in the pressurizer enclosure or containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

Serial No.01-007 Page 18 of 59 VIII.

References:

Attachments 3 & 7 Information for Examination Category B-H affected welds:

Pages 1-10 & 1-13, respectively, cover this weld.

B08.020.001A

Serial No.01-007 Page 19 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Weld for Reactor Coolant System.

ID Number Item Number 1NCIF-1-7 B09.011.007 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8 (c),

Examination Volume C-D-E-F.

For welds joining cast austenitic materials:

ASME Section XI, Appendix III, Paragraph 111-4420, 1989 Edition with no addenda.

"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."

III.

Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to examine 100% of volume C-D-E-F of cast stainless steel welds from two beam path directions.

Serial No.01-007 Page 20 of 59 IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 4, (Pages 1-4) due to single sided access, and weld crown taper the examination coverage was limited to 33.20%.

In the case of the above listed pipe to pump weld, austenitic weld metal characteristics and single sided access caused by the component geometry prevents two sided coverage of the examination volume.

The welded component configuration would have to be re-designed to allow scanning from both sides of the weld over the required examination volume.

V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8 (c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined during installation using volumetric and surface NDE methods.

This is a weld on the "A" cold leg of the reactor coolant system to the "A" Reactor Coolant Pump Outlet Nozzle.

This weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e., embrittlement) associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering

Serial No.01-007 Page 21 of 59 evaluation. A leak at this weld would result in the following:

a) Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

c) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

d) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

Serial No.01-007 Page 22 of 59 VIII.

References:

Information for Examination Category B-J affected welds:

Pages 1-4 cover this weld.

B09.011.007

Serial No.01-007 Page 23 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Weld for Reactor Coolant System.

ID Number Item Number 1NC-3087-WI B09.011.008 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8(c),

Examination Volume C-D-E-F.

For welds joining cast austenitic materials:

ASME Section XI, Appendix III, Paragraph 111-4420, 1989 Edition with no addenda.

"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."

III. Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to examine 100% of volume C-D-E-F of cast stainless steel welds from two beam path directions.

Serial No.01-007 Page 24 of 59 IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 4 (Pages 5-8),

due to the proximity of a pipe restraint, the examination coverage was limited to 85.50%.

In the case of the above listed piping weld, austenitic weld metal characteristics and single sided access caused by the pipe restraint prevents two sided coverage of the examination volume.

The welded component configuration would have to be re designed to allow scanning from both sides of the weld over the required examination volume.

V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined during installation using volumetric and surface NDE methods.

This weld is located on the "A" cold leg of the reactor coolant system near the reactor vessel nozzle.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation. A leak at this weld would result in the following:

a) Increased containment humidity.

This parameter is indicated in the control room and is monitored

Serial No.01-007 Page 25 of 59 periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

c) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

d) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

Serial No.01-007 Page 26 of 59 VIII.

References:

Information for Examination Category B-J affected welds:

Pages 5-8 cover this weld.

B09.011.008

Serial No.01-007 Page 27 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Weld for Reactor Coolant System.

ID Number Item Number INClF-107 B09.011.061 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8(c),

Examination Volume C-D-E-F.

For welds joining cast austenitic materials:

ASME Section XI, Appendix III, Paragraph 111-4420, 1989 Edition with no addenda.

"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."

III.

Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to examine 100%

of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 4, (Page 9-15) the examination coverage was limited to 60.85%.

This is an elbow to nozzle weld where access is limited to the pipe side only.

In the case of the above listed elbow to nozzle weld, austenitic weld metal characteristics and single

Serial No.01-007 Page 28 of 59 sided access caused by the component geometry prevents two sided coverage of the examination volume.

The welded component configuration would have to be re-designed to allow scanning from both sides of the weld over the required examination volume.

V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined during installation using volumetric and surface NDE methods.

Current ultrasonic technology is not capable of consistently detecting and sizing flaws on the far side of an austenitic weld for configurations common to U.S. nuclear applications.

To demonstrate that the best available technology was applied, PDI provides a best effort qualification instead of a complete single side demonstration.

PDI Performance Demonstration Qualification Summary (PDQS) for austenitic piping shows that single sided examination is performed as a best effort.

Therefore, the far side of the austenitic weld, which can only be accessed from one side, will be listed as an area of no coverage.

This weld is located on the "A" cold leg of the reactor coolant system at the nozzle from the NI system, downstream of INI-60.

This weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e.,

embrittlement) associated with neutron bombardment.

If a leak were to occur at the weld in question,

Serial No.01-007 Page 29 of 59 there are methods by which the leak could be identified for prompt Engineering evaluation. A leak at this weld would result in the following:

a)

Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

c) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

d) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

Serial No.01-007 Page 30 of 59 VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category B-J affected welds:

Pages 9-15 cover this weld.

B09.011.061

Serial No.01-007 Page 31 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Welds for Reactor Coolant System.

ID Number Item Number INClF-3613-3092 B09.011.069 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8 (c),

Examination Volume C-D-E-F.

III.

Code Requirement from Which Relief Is Requested:

Relief is sought from the requirement to examine 100%

of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 4, (Pages 16-22) the examination coverage was limited to 60.11%.

This is a pipe to nozzle weld where access is limited to the pipe side of the weld only.

In the case of the above listed nozzle to pipe weld, austenitic weld metal characteristics and single sided access caused by the component geometry prevents two sided coverage of the examination volume.

The welded component configuration would have to be re-designed to allow scanning from both sides of the weld over the required examination volume.

Serial No.01-007 Page 32 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

In addition, this weld was examined during installation using volumetric and surface NDE methods.

Current ultrasonic technology is not capable of consistently detecting and sizing flaws on the far side of an austenitic weld for configurations common to U.S. nuclear applications.

To demonstrate that the best available technology was applied, PDI provides a best effort qualification instead of a complete single side demonstration.

PDI Performance Demonstration Qualification Summary (PDQS) for austenitic piping shows that single sided examination is performed as a best effort.

Therefore, the far side of the austenitic weld, which can only be accessed from one side, will be listed as an area of no coverage.

This weld is located at the nozzle from the pressurizer surge line to the "B" loop reactor coolant hot leg.

This weld is not exposed to significant neutron fluence and is not prone to negative material property changes (i.e.,

embrittlement) associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation.

A leak at this weld would result in the following:

Serial No.01-007 Page 33 of 59 a)

Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

c) Increase in unidentified reactor coolant leakage.

This parameter would be exhibited during performance of the reactor coolant leakage calculation, which is required by Technical Specifications to be performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (McGuire normally performs this calculation every 24 Hrs).

The unidentified leakage specification in Technical Specification 3.4.13.1 is 1 gpm.

d) Other indicators such as containment radiation monitors EMF-38, 39, and 40, the containment floor and equipment sump levels.

Note: The above parameters would be used to identify a leak in containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

Serial No.01-007 Page 34 of 59 VIII.

References:

Information for Examination Category B-J affected welds:

Pages 16-22 cover this weld.

B09.011.069

Serial No.01-007 Page 35 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Weld for Safety Injection System.

ID Number Item Number INIIF-643 B09.011.207 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8(c),

Examination Volume C-D-E-F.

III.

Code Requirement from Which Relief Is Requested:

Relief is requested from the requirement to examine 100% of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume as could not be achieved.

As shown in Attachment 4, (Pages 23-26) the examination coverage was limited to 60.34%.

This is a pipe to valve weld where access is limited to the pipe side of the weld only.

In the case of the above listed pipe to valve weld, austenitic weld metal characteristics and single sided access caused by the component geometry prevents two sided coverage of the examination volume.

The welded component configuration would have to be re-designed to allow scanning from both sides of the weld over the required examination volume.

Serial No.01-007 Page 36 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

In addition, this weld was examined during installation using volumetric and surface NDE methods.

Current ultrasonic technology is not capable of consistently detecting and sizing flaws on the far side of an austenitic weld for configurations common to U.S. nuclear applications.

To demonstrate that the best available technology was applied, PDI provides a best effort qualification instead of a complete single side demonstration.

PDI Performance Demonstration Qualification Summary (PDQS) for austenitic piping shows that single sided examination is performed as a best effort.

Therefore, the far side of the austenitic weld, which can only be accessed from one side, will be listed as an area of no coverage.

This is a pipe to valve (1N170) weld located on the outlet side of the valve on the Emergency Core Cooling System (ECCS) Cold Leg Injection line.

This weld is not exposed to significant neutron fluence and is not prone to embrittlement associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation.

A leak at this weld would result in one or more of the following:

Serial No.01-007 Page 37 of 59 a)

Increase in outleakage from the associated Cold Leg Accumulator (CLA) tanks.

Level in these tanks is continuously monitored and alarmed in the control room and is maintained within limits established in Technical Specification 3.5.1.2.

The fill frequency for these tanks is also trended by the Safety Injection System Engineer.

b) Increased Steam Generator enclosure temperature.

This parameter is continuously monitored by the Operations via an 0AC alarm, and is periodically monitored by the System Engineer.

Note: The above parameters would be used to identify a leak in the containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category B-J affected welds:

Pages 23-26 cover this weld.

B09.011.207

Serial No.01-007 Page 38 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Weld for Safety Injection System.

ID Number Item Number 1NIIF-645 B09.011.219 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-l, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8 (c),

Examination Volume C-D-E-F.

IlI.

Code Requirement from Which Relief Is Requested:

Relief is requested from the requirement to examine 100% of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 4, (Pages 27-30) the examination coverage was limited to 60.34%.

This is a pipe to valve weld where access is limited to the pipe side of the weld only.

In the case of the above listed pipe to valve weld, austenitic weld metal characteristics and single sided access caused by the component geometry prevents two sided coverage of the examination volume.

The welded component configuration would have to be re-designed to allow scanning from both sides of the weld over the required examination volume.

Serial No.01-007 Page 39 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

In addition, this weld was examined during installation using volumetric and surface NDE methods.

Current ultrasonic technology is not capable of consistently detecting and sizing flaws on the far side of an austenitic weld for configurations common to U.S. nuclear applications.

To demonstrate that the best available technology was applied, PDI provides a best effort qualification instead of a complete single side demonstration.

PDI Performance Demonstration Qualification Summary (PDQS) for austenitic piping shows that single sided examination is performed as a best effort.

Therefore, the far side of the austenitic weld, which can only be accessed from one side, will be listed as an area of no coverage.

This is a pipe to valve (1N170) weld located on the inlet side of the valve on the ECCS Cold Leg Injection line.

This weld is not exposed to significant neutron fluence and is not prone to embrittlement associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation.

A leak at this weld would result in one or more of the following:

Serial No.01-007 Page 40 of 59 a) Increase in outleakage from the associated CLA tanks.

Level in these tanks is continuously monitored and alarmed in the control room and is maintained within limits established in Technical Specification 3.5.1.2.

The fill frequency for these tanks is also trended by the Safety Injection System Engineer.

b)

Increased inputs to the containment floor and equipment sumps.

Note: The above parameters would be used to identify a leak in the containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category B-J affected welds:

Pages 27-30 cover this weld.

B09.011.219

Serial No.01-007 Page 41 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category B-J:

NPS 4" or Larger Piping Circumferential Weld for Safety Injection System.

ID Number Item Number INIlF-280 B09.011.228 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWB-2500-1, lists the following requirements for the Examination Category as shown below:

ASME Section XI, Figure IWB-2500-8(c),

Examination Volume C-D-E-F.

III. Code Requirement from Which Relief Is Requested:

Relief is requested from the requirement to examine 100% of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 4, (Pages 31-38) the examination coverage was limited to 60.50%.

This is a pipe to valve weld where access is limited to the pipe side of the weld only.

In the case of the above listed pipe to valve weld, austenitic weld metal characteristics and single sided access caused by the component geometry prevents two sided coverage of the examination volume.

The welded component configuration would have to be re-designed to allow scanning from both sides of the weld over the required examination volume.

Serial No.01-007 Page 42 of 59 V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWB 2500-8(c) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

In addition, this weld was examined during installation using volumetric and surface NDE methods.

Current ultrasonic technology is not capable of consistently detecting and sizing flaws on the far side of an austenitic weld for configurations common to U.S. nuclear applications.

To demonstrate that the best available technology was applied, PDI provides a best effort qualification instead of a complete single side demonstration.

PDI Performance Demonstration Qualification Summary (PDQS) for austenitic piping shows that single sided examination is performed as a best effort.

Therefore, the far side of the austenitic weld, which can only be accessed from one side, will be listed as an area of no coverage.

This is a pipe to valve (INI88B) weld located on the outlet side of the valve located on a 10" line connected to the "ID" Safety Injection Accumulator Tank on the ECCS Cold Leg Injection line.

This weld is not exposed to significant neutron fluence and is not prone to embrittlement associated with neutron bombardment.

If a leak were to occur at the weld in question, there are methods by which the leak could be identified for prompt Engineering evaluation.

A leak at this weld would result in the following:

Serial No.01-007 Page 43 of 59 a) Increase in outleakage from the associated CLA tanks.

Level in these tanks is continuously monitored and alarmed in the control room and is maintained within limits established in Technical Specification 3.5.1.2.

The fill frequency for these tanks is also trended by the Safety Injection System Engineer.

b)

Increased inputs to the containment floor and equipment sumps.

Note: The above parameters would be used to identify a leak in the containment, but could not specifically identify this weld as the source of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category B-J affected welds:

Pages 31-38 cover this weld.

B09.011.228

Serial No.01-007 Page 44 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category C-B:

Nozzle-to-Shell (or Head) Weld for ID Steam Generator Auxilliary Feedwater Nozzle to Steam Drum.

ID Number Item Number ISGD-W259 C02.021.008 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWC-2500-1, lists the following requirements for the Examination Category as shown below:

Figure IWC-2500-4 (a).

ASME Section V, Article 4, Paragraph T-424.1 states: "The volume shall be examined by moving the search unit over the examination surface so as to scan the entire examination volume."

III.

Code Requirement from Which Relief Is Requested:

Relief is being sought from the requirement to scan the entire examination volume.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 5, (Pages 1-4) the examination coverage was limited to 75.00%. This is a

ferritic nozzle to shell weld where access is limited to the vessel shell side only.

In order to achieve more coverage the welded component configuration would have to be re-designed to allow scanning from both sides.

V.

Alternate Examinations or Testing:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number

Serial No.01-007 Page 45 of 59 referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VI.

Justification for the Granting of Relief:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWC 2500-4(a) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

The qualifications were conducted on samples with access to both sides of the weld.

Therefore Duke Energy Corporation does not claim credit for the full volume when a single sided examination is performed.

In addition, this weld was examined during installation using volumetric and surface NDE methods.

This weld is located on the Auxiliary Feedwater nozzle on the 1D Steam Generator.

This weld is not exposed to significant neutron fluence and is not prone to embrittlement associated with neutron bombardment.

If a leak were to occur at the weld in question [Steam Generator (CA) Nozzle],

there are methods by which the leak could be identified for prompt Engineering evaluation.

A leak at a CA nozzle would result in the following:

a) Increased containment humidity.

This parameter is indicated in the control room and is monitored periodically by Operations and also the Containment Ventilation System Engineer.

b)

Increased S/G enclosure temperature.

This parameter is continuously monitored by the Operations via an OAC alarm, and is periodically monitored by the System Engineer.

c) Increased input into the VUCDT.

This parameter is monitored continuously by Operations via an OAC alarm and also periodically by the Liquid Radwaste System Engineer and Reactor Coolant System Engineer.

Note: The above parameters would be used to identify a leak in the steam generator enclosure, but could not specifically identify the CA nozzle as the source

Serial No.01-007 Page 46 of 59 of leakage.

A containment entry would be required to identify the exact source of the leakage.

Also, a containment walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

This walkdown should identify any leak at the weld in question.

Concerning the consequences of a leak at the CA nozzle (affects on CA system operation): Any leakage would result in a portion of the CA flow bypassing the steam generator, and therefore being unavailable to maintain steam generator levels.

Very small leaks

(< 1 gpm) would have no discernible effect on CA system operation.

Leaks that approach 5 gpm would need to be evaluated for system operability effects.

McGuire has specific Safety Analysis for accidents where minor and major main feedwater system pipe breaks are postulated.

These Safety Analyses demonstrate compliance with requirements of 10CFR50.

Replacement or re-design of any of these Class 1 or Class 2 nozzles is not a viable alternative.

Duke Energy believes the amount of coverage obtained for these examinations provides reasonable assurance of the continued structural integrity of the subject welds.

Also the CA nozzles are equipped with thermal sleeves to limit thermal shock due to auxiliary feedwater injections.

McGuire operates the CA nozzles consistent with the stress and fatigue qualifications provided by the Manufacturer (BWI).

VII.

Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

Serial No.01-007 Page 47 of 59 VIII.

References:

Information for Examination Category C-B affected weld:

Pages 1-4 cover this weld.

C02.021.008

Serial No.01-007 Page 48 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category C-F-i:

Piping Circumferential Welds for Safety Injection System.

ID Number Item Number 1NIIF-167 C05.011.113 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWC-2500-I, lists the following requirements for the Examination Category as shown below:

Figure IWC-2500-7 (a) requires 100% of examination volume.

III.

Code Requirement from Which Relief Is Requested:

Relief is requested from the requirement to examine 100% of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 6, (Pages 1-6) the examination coverage was limited to 59.82%.

This is a stainless steel penetration to elbow weld where access is limited to the elbow side of the weld only.

In order to achieve more coverage the weld would have to be re-designed to allow scanning from both sides.

V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Serial No.01-007 Page 49 of 59 Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Examination Category C-F-I:

Although the examination volume as defined in ASME Section XI 1989 Edition with no addenda, Figure IWC 2500-7(a) could not be covered, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity.

This weld was examined using procedures, personnel and equipment qualified through the Performance Demonstration Initiative (PDI).

In addition, this weld was examined during installation using volumetric and surface NDE methods.

Current ultrasonic technology is not capable of consistently detecting and sizing flaws on the far side of an austenitic weld for configurations common to U.S. nuclear applications.

To demonstrate that the best available technology was applied, PDI provides a best effort qualification instead of a complete single side demonstration.

PDI Performance Demonstration Qualification Summary (PDQS) for austenitic piping shows that single sided examination is performed as a best effort.

Therefore, the far side of the austenitic weld, which can only be accessed from one side, will be listed as an area of no coverage.

This weld is located on the "A" Train ECCS Cold Leg Injection supply from the ND system.

This weld is not exposed to significant neutron fluence and is not prone to embrittlement associated with neutron bombardment.

If a leak were to occur at the weld in question, the leak could be identified for prompt Engineering evaluation.

A leak at this weld would result in external leakage from this weld and would be exhibited on the floor of the Aux. Building pipe chase.

Operations perform surveillance in this area monthly for ECCS venting and would notice any leakage exhibited from this weld.

Also, a walkdown is performed when the unit reaches Mode 3 (full temperature / pressure) during the unit shutdown and startup for each refueling outage.

The

Serial No.01-007 Page 50 of 59 following additional walkdowns are performed on this piping each refueling outage:

a)

An inservice inspection walkdown is performed which verifies that no external leakage exists on the piping including this weld.

b)

A leakage walkdown is performed on ND system piping outside containment.

These walkdowns should identify any leak at the weld in question.

VII. Implementation Schedule:

Duke Energy Corporation will continue to use ultrasonic examination procedures to obtain maximum coverage to the extent practical of the item number referenced in Section I of this Request for Relief.

No additional ultrasonic examination is planned during the current interval for this weld.

VIII.

References:

Information for Examination Category C-F-I affected welds:

Pages 1-6 cover this weld.

C05.011.113

Serial No.01-007 Page 51 of 59 I.

System / Component(s) for Which Relief is Requested:

Examination Category C-F-I:

Piping Circumferential Weld for Safety Injection System.

ID Number Item Number INIlF-293 C05.011.120 II.

Code Requirement:

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition Table IWC-2500-1, lists the following requirements for the Examination Category as shown below:

Figure IWC-2500-7 (a) requires 100% of examination volume.

III.

Code Requirement from Which Relief Is Requested:

Relief is requested from the requirement to examine 100% of volume C-D-E-F.

IV.

Basis for Relief:

During the ultrasonic examination of this weld, 100%

of the required examination volume could not be achieved.

As shown in Attachment 6, (Pages 7-16) the examination coverage was limited to 61.30%.

This is a stainless steel pipe to valve weld where access is limited to the pipe side of the weld only.

In order to achieve more coverage the weld would have to be re-designed to allow scanning from both sides.

V.

Alternate Examinations or Testing:

No additional examinations are planned during the current interval for this weld.

Because of the configuration, Radiography would not provide any additional coverage.

Serial No.01-007 Page 52 of 59 Duke Energy Corporation will use the most effective ultrasonic techniques available to obtain maximum coverage for future examinations of this weld.

VI.

Justification for the Granting of Relief:

Examination Category C-F-l: