ML20114D068

Summarizes 841109 Meeting W/Northeast Utils & Westinghouse at Site to Observe & Evaluate Effectiveness of Ultrasonic Insp of Cast Stainless Steel Welds in Primary Piping Sys

ML20114D068
Person / Time
Site:	Millstone
Issue date:	11/21/1984
From:	Thomas Taylor Battelle Memorial Institute, PACIFIC NORTHWEST NATION
To:	Hum M Office of Nuclear Reactor Regulation
Shared Package
ML20107K577	List: ML20107K574 ML20114D068
References
NUDOCS 8501300669
Download: ML20114D068 (6)

Text

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OBallelle Pacific Northwest Laboratories P.o. Bon 999 Richland, Washington U.S. A. 99352 Telephone (509) 375-2838 Telen 15 2874 November 21, 1984 Martin Hum Division of Engineering Materials Engineering Branch Office of Nuclear Reactor Regulations Mail Stop 318 Nuclear Regulatory Commission ,

Washington, DC 20555

Dear Martin:

On November 9,1984, a meeting was held with Northeast Utilities and Westingho,use personnel,at the Millstone Unit 3 plant site.

The p.urpose of the visit was to observe and evaluate th,e effectiveness ultrasonic inspection of' cast stainless steel welds in the primary piping system. A list of personnel attending the meeting-is attached to this letter.

- Prior to the visit, NRC personnel requested inspections of four welds on Unit 3. Before witnessing examinations in the plant, a Westinghouse field inservice inspection team demonstrated detection of mechanical fatigue cracks. The demonstration samples were the same samples Westinghouse used in the experi-ments described in WCAP-9894. .

Demonstrations on Millstone Unit 3 Examinations of portions of four welds on Unit 3 were' witnessed.

The welds were located in the piping on the outlet side of the steam generator (i.e. , thb same pipeing that was observed at the Callaway plant site).

The results of the examinations that were witnessed showed:

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. Surface preparation of the welds examined was adequate for ultrasonic examination.

. A continuous backwall reflection was maintained during a O degree examination of the pipe except on a statically cast elbow l to pipe weld. A backwall reflection could not be maintained from the statically cast elbow.

8501300669 850118 PDR ADOCK 05000423 A PDR

' Martin Hum oBattelle Page 2 November 21, 1984

. The weld root and weld counterbore could be detected during a O degree examination. The examination indicated a very slight counterbore.

. The angled beam examination of the piping could not show

. indications from the counterbore and weld geometry. Prior to the angled beam examination, the Westinghouse inspector stated that he had physically crawled into the pipe to visually inspect the weld counterbore. The visual examina-tion showed the weld counterbore to be very slight and blended. If the weld counterbore was slight and blended, .

the counterbore may not be detectable during the angled beam examination.

. The ambient acoustic noise level was approximately 25-30%

full screen height.

. Examination from the ferritic nozzle side of the steam generator showed large indications from the inconel but-tering. , ,

Based on the demonstrations and discussions at Millstone Unit 37 I conclude:

1. The demonstrations at Millstone Unit 3 showed that adequate penetration of the piping material could be obtained using a O degree longitudinal examination. ,

2. [ equate examination of the statically cast elbow to pipe weld could not be demonstrated from the statically cast elbow side. Failure to maintain a backwall raflection during a O degree examination and failure to demonstrate adequate penetration during an angled beam examination are evidence cited to support this conclusion.

3. Weld root and weld counterbore indications could not be detected during angled beam examinations. Whether the lack of ultrasonic response was due to inadequate penetration or the blended condition of the counterbore could not be determined. However, based on the demonstration at Mill-stone Unit 3, adequate penetration of angled beam examina-tion could not be proven. -

4.- The large indications received from the inconel bu4tering may cause difficulties in interpretation for flaws near the buttered nozzle.

Martin Hum Page 3

)BaHelle November 21, 1984 Based on the demonstrations an'd discussions at Millstone Unit 3, the following recommendations are made:

1. A request for relief of examination of the statically cast elbow should be suggested to the utility.

2. Additional welds should be examined during the presence of a regional I&E inspector to demonstrate that an angled beam examination can detect,th,e presence of counterbore and weld root indications.

Sincerely, T. T. TAYL R Nondestructive Testing Section TTT:kw Attachment . . .

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\eano V. 2a E G e S INTEROFFICE CORRESPONDENCE em November 26, 1984 to B. A. Barna

• J. F. Cook auwa' TRIP REPORT - MILLSTONE UNIT NO. 3 CAST STAINLESS STEEL PIPING EXAMINATION, NOVEMBER 8 and 9, 1984 - JFC-18-84 While in Washington, D.C. on November 8, 1984, to meet with NRC personnel and plan for the Millstone meeting, it was learned that the examination of cast stainless steel (SST) would be discussed at the Advisory Committee on Reactor Safeguards (ACRS) Metal Components Subcommittee. Hence this ACRS meeting was also covered.

Cy Cheng of NRC NRR made a presentation on cast SST examination and problems with changes in ultrasonic properties. The use of performance demonstrations for near-term operating license plants was discussed. Dr. Shewman, the subcommittee chairman, expressed concern over possible missed fabrication flaws. The non-uniformity in region practices relative to evaluating cast SST was discussed. Shewman ex-pressed concern again and subcommittee member Ward suggested that NRC NRR should have guidance for the regions. -

Westinghouse presented a discussion of their ultrasonic examination practices and the Westinghouse-EPRI owners group program on UT of Cast SST. This program, scheduled for 27 months,'has four phases: (1) test sample fabrication; (2) improved manual technique development; (3) automated inspection and data processing; and, (4) demonstration of. flaw detection and characterization cap:Lilities. Based on concerns expressed by Shewman, Westinghous. said they would consider adding fabrication flaws to the test samples. Jp to 75 specimens are anticipated.

On November 9 the performance demonstation at Millstone 3 was covered. An initial meeting was held with the NRC staff and consultants to discuss previous demonstration resul ts, the ACRS meeting of November 8, and the plans for

. the day. Following that a meeting was held with all involved to discuss plans for the demonstrations. Attachment I lists the attendees. It should be noted that some UT technicians present are not listed as they did not ,

fill out the roster.

Westinghouse demonstrated the UT equipment set up using their calibration block, a known cracked block and the Millstone calibration block. Their technique, which~ uses a 41' nominal refracted L-wave, showed a signal from the crack. Of the two callibration blocks (Westinghouse and Millstone) the Westinghouse block showed the higher calibration signal to noise ratio. The TS" l

instrument calilbrated for the Millstone block had a gain setting 6db higher than the instrument calilbrated for the Westinghouse block. The Westinghouse search unit uses a water-filled boot which facilitates coupling to slightly irregular surfaces, but may cause problems with angle variations. A -1 to +1* variation in incident angle will cause a 10* range of refracted angles.

Following the demonstrations, selected welds in the primary coolant piping near a steam generator inlet nozzle were examined. Two centrifigally-cast pipe to static-cast elbow welds and the nozzle to elbow weld were examined over representative portions. With straight beam examination, counter-bore in the pipe could be identified. No counter-bore in the static-cast elbows was detected. In most cases a clean back reflection in the elbows was not observed. The angle beam could not detect the counter-bore in the pipe.

In examining from the clad ferritic nozzle, multiple signals from the cladding were noted that could interfere with detection of flaw indications. However, the cast calibration block used provided more than the required sensitivity for examination from the ferritic nozzle. A clad-ferritic calibration block should be used to calibrate for the examinations from the nozzle side. All the weld surfaces and adjacent base metal were well-prepared for ultrasonic examination with essentially flat scanning surfaces.

Based on the observations, the examinations can be considered at best to provide limited coverage from the pipe side. Examinations from the static cast fitting sides appeared not effective. Thus,100% coverage of the code required examination volume is not being obtained.

To further evaluate and possibly develop an effective examination, use of a static cast calibration block or static cast mockup with known reflectors should be considered. It has been reported that different special transducers have been effective on some static cast material. Trial use of such transducers could be investigated. More information on the acoustic properties of the scan areas and examination volume, for each joint would be helpful.

If access were available to the ID, through transmission techniques could be used to verify penetration of the material.

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Attachment:

As Stated cc: L. S. Beller B. W. Brown

-M. R. Hum, NRC .

C. R. Mikesell J. A. Seydel l

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d Attachment 1 PSI DEMONSTRATION NOVEMBER 9, 1984 l Attendees j W. E. Terry, NUSCo Licensing R. G. Schmidt, NUSCO Reliability Engineering H. F. Conrad, NRC NRR J. M. Stankoski, NUSCo Reliability Engineering Tom Taylor, Battelle Robert- A. McBrearty, NRC Reg I J. P. Durr, NRC Reg I E. Doolittle, NRC Licensing Bernie Lefebure. Westinghouse Insp. Serv.

Mimi Weaver, Westinghouse Nuclear Safety Joe Enrietto, Westinghouse Materials Technology Charles M. Peterson, Westinghouse Nuclear Op. Div.

Walt Emerson, Stone and Webster, Lead Licensing Engineer D. C. Adamonis, Westinghouse NSID Inspection Service G. R. Perkins, NDE Engineering Consult.Inc. (PSI Consultant)

Robert W. Pritchard, NUSCo, PSI Project Engineer J. F. Cook, EG8G Idaho

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NOV 2 61984 Carolina Powerand Light Company ATTN: Mr. -E. E.' Utley

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i Executive Vice President Power Supply and Engineering and Construction 411 Fayetteville Street Raleigh, NC 27602 )

1 Gentlemen:

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SUBJECT:

MEETING

SUMMARY

- REPORT NO. 50-400/84-40 This letter refers to the technical meeting held at your request in the NRC Ragion II Office on November 2, 1984. This meeting concerned activities authorized by NRC Construction Permit No. CPPR-158 for the Shearon Harris .

facility.

.The meeting provided Carolina Power and Light Company (CP&L) an opportunity to d:monstrate the ability of their ultrasonic examination procedure and equipment to detect, locate, and size actual crack flaws propagating from the pipe inside--

diameter in centrifugally cast stainless steel as discussed in the enclosed r: port. _,

It is our opinion that this meeting was informative and beneficial to both NRC and CP&L because of the exchange of information concerning inspection of cast stainless steel.

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will b2 placed ir 49C's . Public Document Room . unless you notify this office by telephone within 10 days of the date of this letter and submit written applica-tion to withhold information contained therein within 30 days of the date of the

letter. Such application must be consistent with the requirements of 2.790(b)(1).

Should you have any questions concerning this matter, we will b,e pleased to discuss them.

Sincerely,

'd Divid M.

rrelli, Chief s

f Reactor P ojects Branch I Division of Reactor Projects

Enclosure:

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Carolina Power a,nd Light' Company 2 POV 2 61984

Enclosure:

Inspection' Report No. 50-400/84-40

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- R. A.-Watson, Vice-President . '

Harris Nuclear Project

' R. M.. Parsons, Project General Manager

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r Report No.: 400/84-40 Licensee: Carolina Power and Light Company 411 Fayetteville Street Raleigh, NC 27602 Docket No.: 50-400 License No.: CPPR-158 Facility Name: Harris 1 Inspection Conducted: NRC Regional Office, Atlanta, Georgia Inspectors: \ m e 4 M.' ;A // ,Q O- 8 Y Date signed J. % Coley, Reactor K neer ,

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SUMMARY

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0n November 2,1984, a technical meeting was held with Carolina Power and Lt;ht '.:;mpany (CP&L), representatives in the Region II Office for the purpose of providing CP&L an opportunity to demonstrate the capability of their ultrasonic examination procedure and equipment to detect actual flaws and artifical reflec-tors in Region II's Centrifugally Cast Stainless Steel (CCSS) test specimens.

CP&L was successful in detecting the I.D. reflectors, including cracks. Region II's CCSS test specimens represented examination conditions much worse than CP&L will experience during the examination of the Shearon Harris wrought stainless steel reactor coolant loop piping. Therefore, CP&L's procedure IST-501, and ~ ultrasonic cquipment have demonstrated conservative detection capability for the preservice examination at the Shearon Harris facility.

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- o REPORTDEThILS

1. Person ..-

ContTted-Licensee Employees

• T. Bromback, Project Specialist, CP&L

• S. Pruitt, Senior Specialist, CP&L Other Organizations

• R. Saunders, Level III, Nuclear Energy Services (NES)

2. Technical Issues Involved CP&L Shearon Harris, and NES personnel identified in paragraph 1 above, visited the Region H Office on November 2, 1984, to demonstrate the capability of their ultrasonic examination procedure (IST-501) and equipment - ,

to detect, locate, and size actual crack flaws propagating from the pipe - - '

inside diameter in Centrifugally Cast Stainless Steel (CCSS). Although the Shearon Harris facility does not have the difficult-to-examine CCSS piping in their reactor coolant loop piping, CP&L will be using a refracted --

longitudinal wave, transducer. The refracted longitudinal wave transducer is the only. practical method of" inspecting difficult-to-penetrate weld volumes and most cast stainless steels. However, the refracted lon'gitudinal Wate '

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mode suffers significant mode conversion when interfacing with corner type reflectors such as I.D. pipe cracks. Therefore, CP&L wanted to test their procedure and equipment detection capability prior to the start of their preservice baseline examinations at the Shearon Harris Unit 1 facility.

P.2gic:r-II provided the CCSS specimens for CP&L's performance demonstration.

cest. All cracks and artificial reflectors were covered so that the licensee

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had no prior knowledge of their location. CP&L used a Krautkramer-Branson USL-38 instrument and dual 1 MHz by 1 inch transducers. The transducers had lucite wedges ground to fit the contour of the reactor coolant loop piping at a specific focal length.

NES performed the actual examination of the test specimens and Region II personnel provided guidance and monitored the ultrasonic instrument calibration and subsequent crack detection techniques used by the

. participants. After the participants had completed the examination of the weld samples, the cracks. and artificial reflector were revealed to the licensee by performing a liquid penetrant test on the sample.

3. Meeting Conclusions During the technical meeting at the Region II Office, CP&L successfully demonstrated the capability of their ultrasonic examination procedure and equipment to detect actual flaws in the volume subject to examination.

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TEXAS UTILITIES GENERATING COMPANY Log # TXX-4218 SKYWAY TOwtR + 4 00 NORTH OLIVE STREET, L.B. 81

• DALLAS, TEXAS TS301 File # 905.4 July 6, 1984 Mr. B. J. Youngblood, Chief Licensing Branch No. I Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION CONFIRMATORY ULTRASONIC EXAMINATION OF CAST STAINLESS WELD AT CPSES

Dear Mr. Youngblood:

. Enclosed are forty (40) copies of the Westinghouse Report entitl'ed,

" Demonstration of Ultrasonic Examination Techniques Applied to Welds in Main Coolant Loop Piping" dated June 28, 1984. This report is being transmitted in response to the B. J. Youngblood letter to t R. J. Gary dated February 24, 1984.

Should you have any questions in this matter please contact me  !

directly.

Sincerely, H. C. Schmidt

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A DIVISION OF TEKAS t!TELETIES ELECTRIC COMPANY

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1 COMANCHE PEAK UNITS I AND II DEMONSTRATION OF ULTRASONIC EXAMINATION TECHNIQUES APPLIED

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TO WELDS IN MAIN COOLANT LOOP PIPING -

D. C. ADAMONIS JUNE 28, 1984 1

APPROVEDi . 44, T. R. Mager, Man er Metallurgical an NDE Analysis l l

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PLANT ENGINEERING DIVISION P. O. BOX 355 l PITTSBURGH, PENNSYLVANIA 15230 l

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This report has been reviewed and checked.

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TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

1 2.0 DEMONSTRATION DETAILS 2 2.1 Calibration Block 2 2.2 Calibration Procedure 2 2.3 Fatigue Cracked Weld Mock-up Demonstration 2 2.4 Field Weld Demonstration 3

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3.0 CONCLUSION

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LIST OF ILLUSTRATIONS

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1 Loop 2 Reactor Coolant Pipe /TBX-1-4200 7 2 Loop 4 Reactor Coolant Pipe /TBX-1-4400 8 l

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LIST OF APPENDICES Appendix Title Page A List of Attendees - Comanche Peak Cast 9 i Stainless Main Coolant Loop Piping UT Demonstration, March 20-21, 1984 1

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1.0 INTRODUCTION

i The Comanche Peak Nuclear Power Station, near Glen Rose. Texas was~ the site of a meeting on March 20 and 21, 1984 to discuss issues relating to the inspectability of welds in cast austenitic stainless steel main coolant lo'op

- piping via ultrasonic techniques and to provide a demonstration of capabilities in that area. ' A list of attendees is attached as Appendix A.

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[ The preservice examination of the Comanche Peak Unit I main coolant loop piping welds was conducted in the Fall of 1982. Portions of these examinations were observed by USNRC Region IV inspectors. The NRC inspectors reported that adequate material penetration could not be  ;

verified because only sporadic back reflections were identified during ,

longitudinal wave examinations and the increased gain used during angle beam examinations saturated the CRT display such that no indications in the first half of the pipe thickness could be identified or evaluated.

These conclusions were documented in NRC Inspection Report 50-445/82-19.

A demonstration concerning ultrasonic testing of cast stainless steel was conducted at the Callaway Nuclear Power Station on January 25, 1984 during a meeting among the NRC Nuclear Reactor Regulation and Region III staffs, Callaway and Wolf Creek personnel, and representatives from Westinghouse.

The NRC attendees at that meeting concluded that a valid ultrasonic examination of cast austenitic piping was possible.

4 In light of the apparent differences between the conclusion reached at the Calla'way meeting and the findings of the Region IV Comanche Peak report, the NRC requested that Texas Utilities address the technical issues identified in Inspe~ction Report 50-445/82-19. It was determined that this could best be accomplished by a confirmatory examination or demonstration, similar to that performed at Callaway, on a minimum of three welds including weld Joint #13 on Comanche Peak Unit I isometric drawing TBX-1-4200. -

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( i This report will document results of the demonstrations which were conducted on two main coolant loop weld mock-ups containing mechnically induced _

fatigue cracks, one Unit I main coolant loop weld, and four Unit II main coolant loop welds.

2.0 DEMONSTRATION DETAILS 2.1 Calibration Block The calibration block used to establish system sweep and distance amplitude calibration for demonstrations on the cracked weld mock-ups and all five

. field welds was identified as TBX/2, HT Cl488. The block is ASTM A-351, Grade CFBM centrifuga11y cast stainless material; approximately 2.2 inches thick and contains 3/16-inch diameter side drilled holes at depths of 1/4T, 1/2T, and 3/4T. This block was used for calibrations for the Unit I preservice examinations.

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2.2 Calibration Procedure

' Sweep and distance-amplitude calibrations for demonstrations on the cracked weld mock-ups and the Unit I and II field welds were established on side drilled holes in the TBX/2 calibration block per Westinghouse procedure 151-206, Revision O. The ultrasonic test system consisted of a Sonic Mark I portable ultrasonic instrument, a 1.0 inch diameter,1.0 MHz, straight beam search unit, and a nominally 40' refracted longitudinal wave,1.0 MHz, water .

column search unit.

2.3 Fatigue Cracked Weld Mock-Up Demonstration The demonstration of crack detection in cast austenitic stainless steel

~ weldments ' involved examination of two weld mock-ups from the WCAP-9894 study, " Reliability of Ultrasonic Test Method for Detecting Natural Fatigue Cracks in Centrifuga11y Cast Stainless Steel Pipe". The weld samples were machined from two ring weldments joining 32-inch OD, SA351, CFBA, type 304 m

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centrifuga11y cast stainless steel pipe fabricated specifically for the WCAP-9894 program. Both samples are on the order of two inches thick, four inches wide, and sixteen inches long. .

I The mock-up identified as WELD 7 DW1 contains a fatigue crack with a depth equal to 10%* of the specimen thickness (0.20 in, crack depth). The mock-up identified as WELD 6 OV 1 contains a fatigue crack with a depth equal to 14%* of the specimen thickners (0.28 in. crack depth). Both fatigue cracks were induced mechanically via three-point bending and extend over the entire weld length (4").

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Both cracks were detectable at repeatable positions along the crack lengths with signal-to-noise ratios on the order of 2.5 to 1. This demonstration indicated that the 40' refracted longitudinal wave technique is capable of detecting mechanically induced fatigue cracks in these materials.

2.4 Field Weld Demonstrations Ultrasonic examination technique demonstrations were conducted on a. total of five field welds, one in Unit I and four in Unit II. All demonstrations -

consisted of a sample of each weld to establish general noise levels and evidence of penetration with both the straight beam and 40' refracted longitudinal wave search units.

Unit I The weld selected for field demonstrations in Unit I was the loop 2 reactor vessel inlet nozzle safe end-to- 271" x 22' cast elbow weld identified as weld #13 on Comanche Peak isometric drawing TBX-1-4200, attached as Figure 1.

Source data for the cast elbow is provided below: ,

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Item Material / Heat Manufacturer /Date f

271".x 22' Elbow SA351, CF8A Breda Fucine Meridionali, Ht. 3-3249-1620, Ser. #4 Bari, Italy /1976

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The field demonstration conducted on this particular weld confinned findings J noted in the NRC report. The surface condition of the elbow, adjacent to the weld, appaared to be the major factor contributing to the examination difficulties. The condition of the surface was, for the most part, as-cast.

4 As such, it is typical of cold leg 22' elbows in other four loop plants. The j configuration of the 22' elbow did not allow the supplier to machine the OD j surface for a significant distance as measured from the edge of the weld prep.

Complete coupling of the straight beam search unit was difficult, thus a con-sistent backwall reflection could not be maintained during scanning. The CRT display during 40' refracted longitudinal wave angle beam scanning was satu-rated with noise over to about one-half the calibrated sweep length. This condition was not considered extremely significant as the inner 1/3 of the pipe thickness constitutes the required examination volume per Section XI of i the ASME Boiler and Pressure Vessel Code,1980 Edition. Aoevidenceofthe 1 - counterbore was detected via angle beam from the elbow side of the weld.

! A continuous back reflection was maintained during straight beam scans on the nozzle safe end side of the weld. Evidence of the counterbore was de-tected during 40' refracted longitudinal wave; examinations from the safe end

( of the weld.

Unit II .

l The four welds selected for demonstrations in Unit II were in the loop 4 crossover leg, identified as welds #5, #6, #7, and #8 in Comanche Peak isometric drawing TCX-1-4400, attached as Figure 2. Weld #5 joins the steam generator nozzle to a 31" 1.D. x 40' cast elbow, weld #6 jo. ins the 31", I.D. x l 40* csst albow to a 4'6-7/8" length of 31" 1.D. centrifuga11y cast pipe, weld

1. 7 joins the 4'6-7/8" length' of 31" I.D. pipe to a 31" I.D. x 90* cast elbow and weld #8 joins the 31" 1.D. x 90* elbow to a 3'5-3/4" length of 31" I.D.

centrifuga11y cast pipe. Source data for these components of the piping systen are provided below:

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Item Material / Heat Manufacturer /Date 31" 1.D. x 40* SA351 CF8A/ Breda Fucine Meridionali, Elbow Ht. 3-3612-0762 Ser. #12 Bari, Italy /1977 31" 1.D. x 4'6-7/8" SA351. CF8A/ Sandusky Foundry, Pipe Ht. 156375, Pc. 2* Sandusky. Ohio /1978 31" I.D. x 90* SA351,CF8A/ Breda Fucine Mendionali, Elbow Ht. 3-3729-1939 Ser. #18 Bari. Italy /1976 31" I.D. 3'5-3/4" SA351, CF8A/ Sandusky Foundry, Pipe Ht. 156375, Pc. 2* Sandusky, Ohio /1978

• - Both pipe lengths cut from same heat of pipe.

In contrast to the results from weld #13 in Unit I, consistent backwall

' reflections were easily maintained during straight beam scanning, noise levels on CRT displays during 40' refracted longitudinal wave examinations were significantly lower, and ID geometry in the form of a counterbore response was consistently detected when scanning weld #5 from both sides.

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Counterbores adjacent to the other three welds w re not detected because of access limitations. Slightly more access thougt., would have made i detection of the counterbores adjacent to the other welds possible.

Surface condition adjacent to these welds was better than that noted for

. weld #13 of Unit I. In particular, tfie elbow OD surfaces were machined I by the supplier for a greater distance as measured from the edge of the

[ weld prep.

3.0 CONCLUSION

S (1) Demostrations perfomed on Unit I, loop 2, weld #13 confimed the l observations of NRC inspection report 50-445/82-19. Surface condition of the elbow appeared to be the major factor contributing to diffi-i' culties in obtaining consistent backwall echoes during straight beam scanning and high noise levels during the 40' refracted longitudinal wave scans of this particular weld. This is due to the configuration of the 22' elbow which did not allow the supplier to machine the OD

( surface for a significant distance as measured from the edge of the I weld prep.

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1 (2) Demonstrations performed on four welds in the Unit II, loop 4, cross-over leg. demonstrated the ability to penetrate the welds and adjacent base material on both sides of the welds. Consistent backwall echoes were maintained during straight beam scanning and noise levels 'during 40' refracted longitudinal wave scans were significantly lower than noted during scans of weld #13 in Unit I. This improvement is attrib-uted to the fact that surface conditions of the elbows adjacent to these welds are good. These particular elbow OD surfaces were machined by the supplier for a greater distance as measured from the edge of the weld prep because of the 16nger tangent configuration.

(3) Demonstrations performed on two centrifuga11y cast pipe weld mock-ups indicates the 40' refracted longitudinal wave technique is capable of detecting mechanically induced fatigue cracks. in these materials.

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• APPENDIX A LIST OF ATTENDEES

. COMANCHE PEAK

. CAST STAINLESS MAIN COOLANT LOOP PIPING UT DEMONSTRATION

March 20-21, 1984 i

i Name Company D. C. Adamonis Westinghouse NTD i

{, M. Blew Westinghouse Support Services -

K. V. Cook Oak Ridge National Laboratories ,

I

,- R. Dacko Texas Utilities Generating Company i

. D. Davis Texas Utilities Generating Company J. F. Enrietto Westinghouse NTD J. Keller Texas Utilities Generating Company a W. M. McNeill US Nuclear Regulatdry Comission W. G. Paul Texas Utilities Generating Company

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T. Taylor Pacific Northwest Laboratories D. Tomlinson US Nuclear Regulatory Comission sf K. Waida Westinghouse Support Services T

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\ D APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV '

NRC Inspection Report: 50-445/82-15 Docket: 50-445 Category A2 Licensee: Texas Utilities Generating Company -

- 2001 Bryan Tower .

Dallas, Texas 75201 Facility Name: Comanche Peak, Unit 1 1.] , Inspection at: Comanche Peak Steam Electric Station Inspectionconductdd: September 7-13, 1982 l Inspecto': r iv e 6sb -

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. v T.r. r. winson, xeactor 2nspactor, engineering section cate 1

Reviewed: 8 7 M.r.c e... < r. -

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s. r. westerman, LaieT, xeacwr Pro. lect '

uate Section A . .

Approved: k $1! Y s u.a# '9ihPl27 D. M. Hunnicutt,.Lnier, engineering section - <

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q. Inspection Summary

! Inspection Conducted September 7-13. 1982 (Report 50-445/82-19)

Areas Inspected: Routine unannounced inspection of construction activities including a site tour review of procedures review of quality records,

- cbservation of work in, progress and review o,f isometric drawings of components

'. and. piping examined during the Unit 1 preservice inspection. Also-examined J. were the licensee actions taken in response to IE Information Notice No. 82-34,

" Welds in Main Control Panels." This inspection involved 36 inspector-hours by

[5 one NRC inspector.

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Results: Within the three areas inspected, no violations or deviations were loentified.

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DETAILS *

1. Persons Contacted Principal Licensee Emolovees

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R. G. Tolson, Site QA Supervisor 4: "C. T. Brandt, QC S ervisor - Mechanical / Civil ti R. A. Perry, Q ity, gineer Preservice Inspection 4 "R. M. Kissinger, Pro.1ect Civil Engineer, TUSI

'j Other Personnel .

1

• D. Gulling, Preservice Inspection Coordinator, Westinghouse

• N. Bollingmo,! Level II Inspecter, Westinghouse 1

J. Delbusso, level I Inspector, Westinghouse 2

1 1; ",enotes these attending the exit interview on September 10, 1982. '

4 The NRC inspector also contacted other licansee and contractor personnel 9 during the course of the inspection.

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2. Site Tour ~ '

' j] The NRC inspector toured the Units 1 and 2 reactor building, auxiliary

,3. buildings and one warehouse to observe work in pragress, inspect completed work inspect received materials, and observe general L[ housekeeping co,nditions. ,

a Within the areas inspected, no violations or deviations were. identified, s 3. Iollowuo on Information Notice 82-34 A .

1- Information Notice 82-34, dated August 20, 1982, was sent~to all holders

.t M of a power reactor operatin license or construction permit as early notification of a potential significant problem. Inspections at three

F vendors facilities disclosed numerous welding practices not in accordance with the American Welding Society (AWS) standards and several quality

' assurance practices not in compliance with the vendors procedures.or NRC requirements.

]

a d The NRC inspector, accompanied by a TUGC0 QA/QC supervisor, toured the -

Unit I control room and performed a visual inspection of the welding fi W- inside eight of eleven installed control panels. Not all welds could be sj inspected as cables had been installed and the panels were energized. It N was apparent that TUGC0 had previously performed an examination as the 4

QA/QC supervisor knew the~ location of several weld discrepant conditions L and readily pointed them out to the'NRC inspector.

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s Lack of' fusion, undercuttin aoparent incomolete -

^ welding, and weld wire rem g, excessive weld spatterants attached t the anomolies noted. hot all of these conditions were noted on eacn weld-or each panel.

No specific action or response was recuired of the licensee at the time

..a the Information Notice was issued. Trie licensee is, however, presently evaluating tne reportability of this matter under the provisions of M 10 CFR Part 50. 55(e). Until future actions are taken by the licensee, this W will be considered an unresolved item.

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4. Preservice Insoection - Unit 1 Q .

1 The NRC inspector reviewed the Westinohouse " Examination Program Plan for Vi Comanche Peak Nuclear Power Plant Unit 1 - Preservice examination 1 Program." Each Class I and Class 2 component requirinc examination and

$. - the type and extent of examination to be performed was clearly specified.

j Exceptions to the reouired examinations were identified and tne reason for J- each was reterenced in tne program plan. A i licensee and the American Nuclear Insurers,porovai Inc. sionatures by the Lj both parties had reviewed and approved the progra,m.(ANII) indicated that 3: The NRC inspector reviewed the personnel qualification records for I

sixteen of the inspectors involved with the preservice inspection. Eight Level I and eight Level Il personnel folders indicated that each inspector .

had sufficient experience and specialized education to. satisfactorily perform the examinations required. Each inspector's file also contained

% records of satisfactory visual acuity and color discrimination tests within the last year. Informal interviews with four of these inspectors indicated Ji that each.has a thorough knowledge of the inspection methods used and the procedures governing the examinations.

]

The NRC inspector.verfied that six of the ultrasonic instruments in use 9 displayed valid and current calibration stickers. The material certifi-d cations for one batch of ultrasonic couplant (Sonotrace 40 Batch #8124);

7 j

four batches of penetrant material (81LO54, 81J116 81LO71,50A032);four batches of cleaner (82A080, 81M038, 82D053, 81H066}; and six batches of A developer (81JO98, 82A007, 81M001, 820056 80B014 80E111) were reviewed 3 and found to meet the requirements for.res,idual su,lfur and halogens.

y 3 In addition, the NRC inspector reviewed Westinghouse Procedures OPS-NSD-101, i and 151-206 for adequac d 151-11,'ISI-47,151-70,E the reguirements of ASM B&PV Code, Section TheseXI.yprocedures and for compliance cover- to

!1- inservice. inspection utilizing magnetic particle licuid penetrant, and

?!i vitrasonic examinations performed on ASME Class i anc Class 2 M components. Each procedure contained the personnel and equipment i requirements, calibration requirements, component surface condition, j component temperature, evaluation and reporting requirements.

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The NRC inspector witnessed the 0*, 45', and 60' ultrasonic exanination

.' of one circumferential weld joint on Steam Generator No. 3. This was 1 i

. identified as weld No. 8 on isometric drawinc TBX-2-110. The examination-

.D was performed in accordance with Procedure 151-47 bv two inspectors d certified to the Level I and Level 11 reouirements 'of SNT-TC-1A. The original calibration of the ultrasonic system was not obeserved

'! NRC inspector verified that postinsoection amolitude calibration, but the Q was within 2 decibels of the recorde' d calibration and that there had been check

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g f no sweep shift. The NRC inspector observed. two other certified inspection personnel perform the ultrasonic examination of welded joint No.13 as shown on isometric drawing TBX-1-4200. This is the elbow-to pipin -

t e.; on the loop 2 cold leg of the reactor coolant system (RCS) pipino.g weld

.; Although the postexamination calibration check was within the established limits and ~

q the screen presentation was good, the two examination scans on the RCS a pipino could pot be verified as havino adequate material penetration. With R. the ultrasonic instrument sensitivity increased from calioration gain

j. settino to the scan gain settino, only a sporadic back reflection could be W identified. Aceounte lonoitudir.a1 wave penetration is nor ally cauced by W -

tne cresence or a:sence o? the far-surface back reflection.

fj the ' low transducer frecuency and elevated amoli1!ude, due to the increasedAo 7 gain for the examination, saturated the cathode rav tube (CRT) screen for a

accroximately half of the sweep range.

4 c'ations in the first half of the piping thickness 'could be identified orWith th evaluated. Fourteen joints on each of the four RCS loops were examined utilizing a O' longitudinal beam and a 41' refracted longitudinal beim.

1 Conversations with several of the inspection personnel indicated that these i

conditions were common to all of the RCS piping welds. The four loops were I fabricated from centrifugally cast stainless steel which is notoriously

'L difficult to ultrasonically inspect due to its extremely large and irregular j~; grain structure. Although differences of opinion as to the validity of ultrasonic inspection results have been expressed by many cognizant organi-

>]- ' . zations and indivuduals, this remains the only inservice volumetric exami-nation possibic. Prior to this preservice ultrasonic examination, each of d the piping joints was radiographically inspected and found to be acceptable.

1 Subsequent radiography will not be possible due to the lack of access to the y inside of the i No magnetic 9 particle M NRC inspec(MP) p pe and the wall thickness of the tor. MP and LP had been completed on all Class 1 and Class 2 com d) components prior to the beginning of this inspection .

3~ The NRC. inspector randomly selected a sample from the preliminary 9 inspection results sheets for review. This sample included records for -

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d and ultrasonic inspections. Each C-M magnetic data sheet identified particle, theliquidcomponent penetrant,being examined, the inspectio t" used, identification and level of the inspection personnel, and identifi-cation of the materials or instrumentation used. For all inspections the J temperature of the ites examined was recorded and for ultrasonic inspe,ction, the temperature of the calibration standard was also recorded. . Calibration .

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,1 data sheets were included for each examination performed and each identified q.

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the standard used and the indication amplitudes achieved. The NRC inspector reviewed the records for approximately 200 ultrasonic . inspections and approximately 400 surface examinations. The records for each were complete, tnorough, and easily traceable to the individual. welds inspected.

Within the areas inspected, no violations or deviaticns were identified.

5. Unresolved Items ,

et .

Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or i

deviations. One unresolved item identified curing this inspection is discussed in paragraph 3.

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6. Exit Interview -

t with those persons An listedexit interview in paracraoh 1. was At thiscon, ducted interview, the NP.~ September 10, 19*2, inspecto

}y tne scope of tE.is' inspection and the findings.

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j- ENCLOSURE 5 t* 1 '

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PROPOSED'SER SECTIONS FOR

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4 SUPPLEMENT 1 TO NUREG-1031 e

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MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 DOCKET NO. 50-423 a

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• l Safety Evaluation Report By the Office of Nuclear Reactor Regulation

!c for Northeast Utilities ~

' Millstone Nuclear Power Station Unit 3 '

Docket No. 50-423 L

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. 9.3.2B Post-Accident Sampling System (NUREG-0737, II.B.3) n t.:

O Introduction 7.j j'i In our safety evaluation, we determined that the applicant met ten of the eleven NOREG-0737','It'em II.'B.3 criteria. By letter dated August 16, 1984, 1

tj the applicant provided additional information on the plant specific pro-cedure to estimate core damage.

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.1 Evaluation i

.i The applicant provided a procedure for estimating core damage during accident conditions. Core damage estimates are based on utilizing post-accident

-, sampling system measurements of fission product concentrations in primary q

t. coolant and in containment. Additional procedures are provided for confirm-ing core damage based on measured hydrogen concentration in containment and i for estimat'ing the' extent of core damage based on containment radiation-

'1 monitors. Reactor coolant system pressure and in-core thermocouple are used to establish if there has been adequate core cooling. This meets Criterion (g j (2)andis,therefore, acceptable.

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Conclusion a'

. Based on the above evaluation, we conclude that the applicant's post-accident j sampling system meets all the requirements of Item II.B.3 of NUREG-0737 and

}. is, therefore, acceptable.

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. RADIOLOGICAL ASSESSMENT BRANCH SUPPLEMENT SAFETY EVALUATION REPORT MILLSTONE UNIT 3 i

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- l '4 Section 12.3.2 Shielding i

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As required in NUREG-0737, II.B.2, " Design Review of Plant Shielding which may be used in Postaccident Operations", the applicant has provided

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j a radiation and shielding design review that identifies the location of vital areas and equipment in which personnel occupancy may be unduly limited i or safety equipment may be unduly degraded by radiation during operations following an accident resulting in a degraded core.

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,. DISCUSSION AND CONCLUSION

'! 5 j' ' The plant shielding design report was reviewed to evaluate the ability to 17~ have access to vital areas necessary to operate esser.tial systems required I; after a LOCA with significant core damage.

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<1 fj Vital areas which require continuous or frequent occupancy in order to con-Q '

trol, monitor, and evaluate the accident were identified. In addition, NNECO

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L1-a identified potential maintenance activities that might become necessary during

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recovery and detemined when after an accident such maintenance would be possible.

For the vital areas the licensee has provided a p'erson-rems time, distance and personnel occupancy study. The vital areas are the Auxilary Building, Main Steam

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Valve Building, Hydrogen Recombiner Building, Fuel Building, Sample Analysis

!1 l Area (Millstone 1 & 2 Chem. lab.), Control Room and Technical Support Center.

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p-Calculatiorsof source terms and estimated post-accident dose rates used L

for shielding design are based on Regulatory Guides 1.4, 1.7 and the

,; guidelines of General Design Criteria 19. The licensee has provided I-

-j " Radiation" maps that show access routes to post-accident vital areas, to be used as an administrative guide in the control of access and reduction j of personnel exposure during the course of an accident.

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Systems containing high levels of radioactivity in a post accident

]. environment were identified but found to be either irrelevant or ti negligible contributors of radiation dose following an accident.

-t NNECO's poit-accident access and shielding study for Millstone 3 shows that no personnel will be exposed to post-accident doses

, greater than GDC 19 dose' rate guidelines of 5 rem whole body or its 1

- equivalent to any part of the body for the duration of the accident.

4 1 '

Il On the basis of our review, we have concluded, that the licensee has 1

perfomed a radiation and shielding design review for vital area access f

i in accordance with Action Plan' Item II.B.2. .

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