ML18003A905

Forwards Responses to Draft SER Open Items 37,39 & 43 Re Fracture Toughness,Exemption from Testing Four Heats of Manual Metal Arc Electrodes & Compliance W/Charpy V-notch Impact Orientation Requirements,Respectively

ML18003A905
Person / Time
Site:	Harris
Issue date:	06/14/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
LAP-83-181, NUDOCS 8306170224
Download: ML18003A905 (9)

Text

'REGUI.ATOR NFORMATION DISTRIBUTION 'M ('R IDS)

'ACCESSION NBR;8306170224 DOG DATE: 83/06/14 NOTARIZED: NO DOCKET

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.'FACIL:50 400 'Shearon Hanris Nuclear iPower 'Planti .'Uni,t ii "Carolina 05000400

!50 401 'Shear on Harris'.Nuclear 'Power IPlanti 'Uni;t. 2'q 'Cat ol ina 05000401

'AUTH',NAME'UTHOR'AFFILIATION MCDUFFIE~ M< A, 1Gar ol ina iPower . 8 Li ghti~Co ~

'REiC Iip ~ NAME- REC IP IDENT'FF ILIATION DENTONiH.R< Of f ice, of Nucl ear 'Reactor 'Regul-ationi Dir actor~

.SUBJEGT; .For wards responses to draft 'SER OperI.Items 37i39 8 43"r e.

fracture'<toVghnessiexemption *from ~testing. four heats of manual metal arc electrodes 8 compliance w/Charpy V-notch impact ori entatati'on~~requir ementsinespecti vel y.

DISTRIBUTION >CODE: Bpp 1'S >COPIES IRECEIVED: LTR .

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RECIPIENT iCOP IES 'RECIPIENT ~COPIES ID,'CODE/NAME~ L'TTR 'ENCL "ID 'CODE/NAME L'TTR ~ENCL

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.NRR/DL/SSPB "0 NRR/DSI/AEB 26

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CMLt, Carolina Power & Light Company SERIAL: LAP-83-181 JUN 14 1g83 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS ~ 1 AND 2 DO(XET NOS. 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT RESPONSES MATERIALS ENGINEERING BRANCH

Dear Mr. Denton:

Carolina Power & Light Company (CP&L) hereby transmits one original and forty copies of the responses to the Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) CP&L Open Items 37, 39, and 43.

Carolina Power & Light Company will be providing responses to other Open Items in the DSER shortly.

Yours very truly, M. A. McDuffie Senior Vice President Engineering & Construction PS/ce (68590NH)

Attachment CC Mr. N. Prasad Kadambi (NRC) Mr. Wells Eddleman Mr. G. F. Maxwell (NRC-SHNPP) Dr. Phyllis Lotchin Mr. J. P. O'Reilly (NRC-RII) Mr. John D. Runkle Mr. Barry J. Elliot (NRC-MTEB) Dr. Richard D. Wilson Mr. Travis Payne (KUDZU) Mr. G. 0. Bright (ASLB)

Mr. Daniel F. Read (CHANGE/ELP) Dr. J. H. Carpenter (ASLB)

Chapel Hill Public Library Mr. J. L. Kelley (ASLB)

Wake County Public Library 8306170224 830614 PDR ADOCK 05000400 E

411 Fayettevllle Street ~ P. O. Box 1551 o Raleigh, N. C. 27602

Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) 0 en Item 37 (DSER Section 5.3.1.2 a e 5-13)

To demonstrate compliance with the fracture toughness requirements of Paragraph IV.A.1 of Appendix G, 10 CPR 50:

as Indicate whether there are any Reactor Coolant Pressure Boundary (RCPB) heat-affected zones which require Charpy V-notch (CVN) impact testing per paragraph NB-4335.2 of the 1977 ASME Code. Provide CVN impact test data for these heat affected zones and their adjacent weld materials. Also, for those weld materials, indicate the weld seam location, weld parameters (process type, slag pool depth, electrode feed rate and oscillation, current, voltage, and slag conductivity) and heat treatment received by the welds and test coupons.

b. Indicate that there are no ferritic RCPB base metals other than in vessels which require fracture toughness testing to NB-2300 of the ASME Code. If there are ferritic RCPB base metal other than in vessels which require fracture toughness testing to NB-2300 of the ASME Code, provide CVN impact and drop weight data for all materials which will be limiting for operation of the reactor vessel.

~Res ense

a. There are no RCPB heat-affected zones which require CVN impact testing per paragraph NB-4335.2 of the 1977 ASME Code.

b. There are no ferritic RCPB base metals other than in vessels which require fracture toughness testing to NB-2300 of the ASME Code.

I Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) 0 en Item 39 (DSER Section 5~3~ 1 ~2 a e 5-13)

To justify an exemption from testing the four heats of manual metal arc electrodes used in nonbeltline region welds of Uni.ts 1 and 2 and one inlet nozzle of Unit 2, provide fracture toughness data from heats from similar materials that demonstrates the reference nil-ductility transition temperature, RT , for the manual metal arc welds and inlet nozzle forgings are less than 10 $ . The representative material must be fabricated from the same material specification and processed to an equivalent metallurgical heat-treated condition as the untested materials ~

~Res ense The inlet and outlet nozzles from both units were supplied by the same vendor and were fabricated from the same material specification and processed to the same heat-treated condition. As shown in Tables 5.3.1-2 and 5.3.1-3, all the nozzles exhibited an RT DT of 0 F or lower except for inlet nozzle Heat No.

406M-8 from Unit 2 whic was estimated to have an RTNDT of 10 F. Since the estimated RTNDT of 10 F is higher than the actual RTNDT obtained on the other eleven nozzles, exemption from testing the one inlet nozzle is considered to be justified.

A review of six metal arc electrodes used in fabricating welds in the Unit 1 and 2 vessels using the same material specification and processed to an equivalent metallurgical heat-treated condition showed that all welds exhibited an RTNDT of -20 F or less. Since this representative material exhibited RTND values of -20 F or less, exemption from testing the four heats of manual metal arc electrodes used in nonbeltline region welds with estimated RTNDT of 10 F is considered to be justified.

s Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) 0 en Item 43 (DSER Section 5.3.1.3 a e 5-19)

To demonstrate compliance with the Charpy V-notch (CVN) impact orientation requirements of Paragraph III.B.1 of Appendix 0, 10 CFR 50 for reactor vessel base material outside the beltline:

a. Indicate whether all reactor vessel base metal were CVN impact tested to the orientation requirements of NB-2322 and test requirements of NB-2330 of the Summer 1972 Addenda of the 1971 ASME Code.

b. Indicate which reactor vessel base metal were not CVN impacted tested to the requirements of NB-2322 and NB-2330 of the Summer 1972 Addenda of the 1971 ASME Code.

Ce Indicate the method of determining the reference nil-ductility transition temperature, RTNDT, for each base metal identified in (b).

d. Provide CVN impact test data which demonstrate for each material (A508, Class 2 or A533B, Class 1) type that the criterion in (c) is adequate for determining transverse CVN impact properties from longitudinal-oriented CVN impact specimens.

~Res onse as All reactor vessel base metal, except one inlet nozzle from Unit 2, was CVN impact tested to the orientation requirements of NB-2322 and test requirements of NB-2330 of the Summer 1972 Addenda of the 1971 ASME Code.

b. Inlet nozzle Heat No. 406M-8 from Unit 2 was the only base metal not CVN impact tested to the orientation requirements of NB-2322 and test requirements of NB-2330 of the Summer 1972 Addenda of the 1971 ASME Code.

c~ NUREG-75/087 Branch Technical Position MTEB 5-2 was used to determine the RT for the inlet nozzle identifed above. RTNNT was determined to be PF since the nil-ductility transition temperature, TNNT, was 10 F and CVN impact tests in the longitudinal orientation were greater than 50 ft-lbs and 35 mils lateral expansion at 40 F and were estimated to be greater than 50 ft-lbs and 35 mils at a temperature 20 F higher for transverse oriented specimens.

d. Data provided in Tables 5.3.1-2 and 5.3.1-3 on eleven other nozzles produced by the same supplier demonstrate that the criterion of MTEB 5-2 which was used is adequate for determining the transverse CVN impact properties for longitudinally oriented CVN impact specimens.

(6859PSAce)

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