ML20217A495

Forwards Relief Request 98-02 for Limited Exam Results for Three Listed Welds Inspected During end-of-cycle 10 Refueling Outage.Rev 0 to Calcualtion NDE-91-1 Encl

ML20217A495
Person / Time
Site:	Catawba
Issue date:	03/16/1998
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217A503	List: ML20217A495 ML20217A509
References
NUDOCS 9803250092
Download: ML20217A495 (14)

Text

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O *:Di Ikt-]esaa ,% - Cauwba NudarSurian 4800 Concord Road York, SC 29745 -

cm g, m ' (803) Ui425I onwE Via hsks (803)831'3426MX March 16, 1998 U.S. Nuclear Regulatory Commission

' Attention: Document Control Desk washington, D.C. 20555 j

Subject:

Duke Energy Corporation Catawba Nuclear Station, Unit 1 Docket Number 50-413 Request for Relief Number 98-02 Limited Weld Examinations Pursuant to 10 CFR - 50.55a (g) (5) (iii) , please find attached Catawba Request for Relief Number 98-02. This request for relief.is associated with limited examination results for three welds which were inspected during the Unit 1 end-of-cycle 10 refueling outage. These welds are the Reactor Vessel Upper Head to Flange Weld, the Excess Letdown Heat Exchanger Head to' Flange Weld, and the Auxiliary Feedwater Nozzle to Transition Ring Weld.

The attachment to this letter contains all technical information necessary in support of this request for relief.

If there are any questions associated with this request for relief, please call L.J. Rudy at (803) 831-3084.

Ver truly yo A

Gary R. Peterson )

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U.S. Nuclear Regulatory Commission Page 2

. March 16, 1998 xc (with attachment):

L.A. Reyes U.S. Nuclear Regulatory Commission Regional' Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 D.J. Roberts Senior Resident Inspector (CNS)

U.S. Nuclear Regulatory Commission Catawba Nuclear Station P.S. Tam NRC Senior Project Manager (CNS)

U.S. Nuclear Regulatory Commission Mail Stop O-14H25 Washington, D.C. 20555-0001-T

Request for Relief  !

i Serial No. 98-02 Page 1 of 7 DUKE ENERGY CORPORATION l STATION: CATAWBA NUCLEAR STATION UNIT 1 10-YEAR INTERVAL REQUEST FOR RELIEF NO. 98-02

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

ASME Section XI Code Class 1 Examination Category: B-A Pressure Retaining Welds in Reactor Vessel, ASME Section XI Code Class 2 )

Examination Category: C-A Pressure Retaining Welds in Pressure Vessels and ASME Section XI Code Class 2 Examination Category C-F-1 Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping.

Weld Number item Number 1RPV-WO8 801.040.001 l 1ELDHX-HD-FLG C01.020.003 1SGD-W261 C05.011.011 ,

11. Code Requirement:

ASME Section XI 1989 Edition; Examination Category: B-A Pressure ,

Retaining Welds in Reactor Vessel, Table IWB-2500-1, item Number B01.040; Examination Category: C-A Pressure Retaining Welds in Pressure  ;

Vessels, Table IWC-2500-1, item Number C01.020; and Examination Category C-F-1 Pressure Retaining Welds in Austenitic Stainless Steel or  ;

High Alloy Piping, Table IWC-2500-1, item Number C05.011 require a I volumetric examination of essentially 100% of the weld volume. Duke Energy l Corporation, with NRC approval, has adopted Code Case N-460 which '

defines " essentially 100%" as greater than 90% coverage.

1 e l l

o

Request for Relief Serial No. 98-02 Page 2 of 7 lil. Code Requirement from which Relief is Requested:

Relief is requested for the above identified Weld ID Numbers:

. Class 1 Reactor Vessel Upper Head to Flange Weld, from meeting the coverage requirements as defined in ASME Section XI, Figure IWB-2500-5, Examination Volume A-B-C-D.

. Class 2 Excess Letdown Heat Exchanger Head to Flange Weld, from meeting the coverage requirements as defined in ASME Section XI, Figure IWC-2500-1, Examination Volume A-B-C-D.

. Class 2 Auxiliary Feedwater Nozzle to Transition Ring Weld, from meeting the coverage requirements as defined in ASME Section XI, Figure IWC-2500-7, Examination Volume C-D-E-F and Appendix ill.

IV. Basis for Relief:

During the ultrasonic examination of the Reactor Vessel Upper Head to Flange Weld 1RPV-WO8 (Item No. B01.040.001) shown in Attachment 1, greater than 90% coverage of the required examination volume could not be obtained. Causes of those limitations are due to Flange Configuration and the proximity of the lifting lugs to the Flange Weld, which limits the ultrasonic coverage to 87.67% of the required volume in order to achieve greater than 90% coverage of the required volume, the lifting lugs would have to be removed. Ultrasonic examination from the inside surface of the head is not feasible because of the high radiation exposure.

During the ultrasonic examination of the Excess Letdown Heat Exchanger l

Head to Flange Weld 1ELDHX-HD-FLG (Item No. C01.020.003) shown in Attachment 1, greater than 90% coverage of the required examination volume could not be obtained. Single sided weld configuration due to flange ,

geometry and the proximity of vent and drain branch connections reduced the l coverage to 38.52%. In order to achieve greater than 90% coverage of the required volume, the weld configuration would have to be redesigned and the i branch connections removed to allow for access from both sides of the weld. j During the ultrasonic examination of the Steam Generator 1D Auxiliary Feedwater Nozzle to Transition Piece, Weld 1SGD-W261 (Item Number C05.011.011) shown in Attachment 1, greater than 90% coverage of the i required examination volume could not be obtained. Singled sided weld configuration reduced the coverage to 75%. In order to achieve greater than 90% coverage of the required volume, the weld configuration would have to be redesigned to allow access from both sides of the weld.

)

Request for Relief Serial No. 98-02 l Page 3 of 7 V. Alternate Examinations or Testing:

No additional examinations are planned during the current interval for Weld ID Numbers 1RPV-WO8,1ELDHX-HD-FLG, and 1SGD-W261. The use of i radiography as an attemate volumetric examination method for Weld ID {

Numbers 1RPV-WO8,1ELDHX-HD-FLG, and 1SGD-W261 is not practical due to component thickness and geometric configurations. Other restrictions making radiography impractical are the necessity to use double wall techniques due to inaccessibility of the ID surface and physical barriers prohibiting access for placement of source, film, number bands, etc. Duke Energy Corporation will continue to use the most current ultrasonic techniques available to obtain maximum coverage for future examinations of these Weld ID Numbers.

VI. Justification for the Granting of Relief:

i Reactor Vessel Upper Head to Flanae Weld Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWB-2500-5, Examination Volume A-B-C-D for Weld ID Number 1RPV-WO8 (Item Number B01.040.001) could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity. For results of the examinations, reference Attachment 2, Pages 1 through 19.

The Reactor Pressure Vessel (RPV) Upper Head to Flange Weld (Weld Number 1RPV-WO8) is by definition not in the beltline area of the RPV; therefore, it is not subject to fluence levels equal to or greater than 1 E17 n/cm2. RPV materials not in the highly irradiated beltline region are not prone to negative material property changes (i.e., embrittlement) associated with fuel reactivity neutron bombardment. Based upon 10 CFR 50.55a, the ASME Code Section XI 1975 and 1989 Editions require essentially 100%

RPV weld volumetric examinations of beltline welds during every inspection interval. The RPV Upper Head to Flange Weld does not meet the requirements of a beltline weld due to a significantly lower fluence exposure, resulting in far less potential degradation of ductility. The Catawba Nuclear station Unit 1 RPV was fabricated by the Rotterdam Dockyard Company and is free from unacceptable fabrication defects. Rotterdam performed rigorous state-of-the-art RF V inspections following fabrication to ensure no significant flaws existed.

The flange configuration and location of the RPV lifting lugs in the proximity of the RPV Upper Head to Flange Weld prevents obtaining 100% volumetric

Request for Relief .

Serial No. 98-02 Page 4 of 7 examination coverage; therefore, the 100% examinations are impractical.

Removal of the lifting lugs and/or ultrasonic examination from the inside i

surface of the head are not viable attematives and would create an undue burden on Duke Energy Company.

Excess Letdown Heat Exchanaer Head to Flance Weld Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWC-2500-1, Examination Volume A-B-C-D for Weld ID Number 1ELDHX-HD-FLG (Item Number C01.020.003) could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity. For results of the examinations, reference Attachment 3, Pages 1 through 12.

The Excess Letdown Heat Exchanger is not primarily used in power operations except as a backup to the normal service ' Letdown Heat Exchanger'. The Excess Letdown Heat Exchanger is located within the Containment Vessel and during power operations the heat exchanger is not accessible for visual inspections except during refueling outages or forced shutdowns. WI 4en not in service during power operations and in ' standby readiness,' the Excess Letdown Heat Exchanger is isolated on the tubeside and static pressure is approximately 50 psig.

If a leak were to occur at the weld in question, Channel Barrel Circumferential Weld to the Channel Flange Bolting Ring, there are several periodic tests and '

evaluations that are performed by established procedures that should identify the leakage for prompt OPS /ENG evaluation.

)

In the 'Out of Service' mode, the Heat Exchanger is isolated from the Reactor Coolant System (NC), and if a leak were to occur at the subject weld, the water accumulated on the Containment floor would flow to the Containment Floor & Equipment (CF&E) Sump where a monitoring program is in place to evaluate changes in sump level. As part of the above sump monitoring program, alarms in control room will alert Operators of a high rate change in sump level, and a start of each shift Operators note the Containment Sump  !

level. Additionally, Heat Exchanger extemalieakage would be detected by  !

EMF monitors 38,39, & 40 within Containment by the presence of airbome  !

radiological activity from the NC systern leakage. With water on the floor in I Containment there would be an increase in Containment humidity measured ,

l by gauge read in Control Room and an increase in Ventilation Unit Condensate Drain Tank (VUCDT) level, which is monitored, also, for changes  ;

to evaluate potential Containment piping / equipment leakage.

i Since the Heat Exchanger is normally not in service, but in standby, the Heat Exchanger and weld joint in question are not exposed to NC system design  !

Request for kelief Serial No. 98-02 J Page 5 of 7 stress conditions of 2200 + psig except when Heat Exchanger is valved into  ;

(

service to support NC system letdown cooling functions. If a leak were to occur in the subject weld joint while Heat Exchanger becomes part of the NC system and subject to the NC system leakage program regulations, which )

basically allows < 1 gpm unidentified leakage maximum per station technical specifications, if NC system unidentified leakage > 1 gpm, power decrease ,

must commence to shutdown to find and correct leak. If a leak were j suspected externally from the Excess Letdown Heat exchanger while in service, control room operated valves could be actuated to isolate the Heat Exchanger and stop the leakage to maintain NC system inventory and Reactor protection. Also, available for leak detection are the CF&E sump monitoring program, Containment EMF's 38,39 & 40 for radiological release detection and humidity measurement in Containment and ventilation condensate increase in the VUCDT.

Auxiliary Feedwater Nozzle to Transition Rina Weld Although the examination volume requirements as defined in ASME Section XI 1989 Edition, Figure IWC-2500-7, Examination Volume C-D-E-F for Weld ID Number 1SGD-W261 (ltem Number C05.011.011) could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity. For results of the examinations, reference

! , Pages 1 through 5.

l' If a leak were to occur at the weld in question (Steam Generator 1D Auxiliary Feedwater (CA) Nozzle to Transition Piece, Weld 1SGD-W261), there are methods by which the leak could be identified for prompt Engineering  ;

evaluation. i The CA nozzles are at final feedwater pressure and temperature. In Modes 1  !

(Power </= approximately 17%),2, and 3, all feedwater is provided to the i steam generators via the CA nozzle (CF-to-CA Bypass Flow). In Mode 1 (Power >/= approximately 17% power), a small amount of feedwater (Tempering Flow) is directed to the CA nozzle of the steam generators to keep the nozzles at final feedwater temperature to reduce the thermal shock to the nozzles associated with CA system operation and transfer of feedwater flow to the nozzles during unit shutdowns. A leak at the CA nozzle would result in the following:

1) Increased S/G enclosure temperature. This parameter is monitored periodically by the Containment Ventilation System (VV) Engineer per the associated " Engineering Support Program".

Request for Relief Serial No. 98-02 Page 6 of 7

2) 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 (WL) Engineer and

' Reactor Coolant System (NC) Engineer per the associated " Engineering Support Program".

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 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 for each refueling outage. This walkdown should identify any leak at the weld in question.

1 Conceming 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 unuvailable to maintain steam generator levels. Very small leaks (< 1 gpm) would have no discemible effect on CA system operation. Leaks that approach 5 gpm would need to be evaluated for system operability effects.

Based on these evaluations, it is Duke Energy Corporation's position that the limited coverage will not endanger the health and safety of the general public.

Duke Energy Corporation will continue to perform UT examinations to the extent practical usina procedures and personnel qualified in accordance with ASME Section XI, Appendix 1, and Section V, Article 4, for the Reactor Vessel Closure Head and Appendix Ill 1989 Edition for the Excess Letdown Heat Exchanger and for the Auxiliary Feedwater Nozzle to Transition Ring Weld.

1 I

i

Request fsr Relief I

Serial No. 98-02 Page 7 of 7 -

Vll. Implementation Schedule:

These examinations will continue to be scheduled in accordance with the requirements of ASME Section XI for future inspection intervals at Catawba Nuclear Station, Unit 1.

Evaluated By: -

• Date __

v vv v NDE Level 111 Review By: #s, ,

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7 Reviewed By: (I)IlS(M1) Date 3 lT)% '

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Attachment i Description Table Attachment 2 UT Examination Data H01.040.001 Attachment 3 UT Examination Data C01.020.003 Attachment 4 UT Examination Data C05.0ll.011

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