ML19312E080

Inservice Insp Exam Rept for 800112-0309.Describes Exams Performed During 1980 Refueling Outage

ML19312E080
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Site:	Maine Yankee
Issue date:	05/28/1980
From:	Maine Yankee
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ML19312E079	List: ML19312E078 ML19312E080
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Download: ML19312E080 (28)

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f INSERVICE INSPECTION EXAMINATION REPORT MAINE YANKEE ATOMIC POWER COMPAh"I JANUARY 12, 1980 THROUGH MARCH 9, 1980 8006030 /d

r O

PREFACE This summary report covers the Inservice Inspection during the 1980 refueling outage of the Maine Yankee Atomic Power Plant located at Wiscasset, Maine.

Included is an abstract of examinations performed with the conditions observed and corrective measures taken. These examinations were performed primarily by Peabody Testing Magnaflux of Hartford, Connecticut, with a limited number of inspections performed by Yankee personnel.

11

g.

s TABLE OF CONTENTS PAGE NIS-1 OWNER'S DATA REPORT l-15

SUMMARY

REPORT

1.0 INTRODUCTION

16 1.1 Examination Methods 16-17 1.2 Evaluation of Data 18 1.3 Examination Results 18 2.0

SUMMARY

OF EXAMINATIONS 19 2.1 Reactor Vessel 19 2.2 Pressurizer 19 2.3 Steam Generator No. 1 21 2.4 Piping 21 3.0 REACTOR COOLANT SYSTEM LEAK TEST 23

4.0 CONCLUSION

S 24 iii l

FORM NIS 1 OWNERS' DATA REPORT FOR INSERVICE INSP As requir:d by th7 Pr: visions of the ASME Code Rules Maine Yankee Atomic Power Company, Wiseasset,

1. Owner Maine (Name and Address of Owner)

Maine Yankee Atomic Power Plant, Wiscasset, Maine

2. Plant _

(Name and Address of Plant)

3. Plant Unit

1. 1

4. Owner Certificate of Authorization (if required)_

DPR--36

5. Commercial Service Date _12/29/72 6. National Board Number for Unit Reactor Vesse] 70RM

7. Components Inspected Component or Manufacturer Manufacturer Appurtenance or Installer State or National or installer Serial No.

Province No.

Board No.

R.P. Vessel C.E.

67106 N/A 20865 R.P. Head C.E.

67206 N/A 20865 Pressurizer C.E.

67601 N/A 20858 Stm. Gen. #1 C.E.

67501 N/A 20919 Piping S&W N/A N/A N/A Blowdotm Piping Mod.

M.Y.

N/A N/A N/A l

(2) information in items I through 6 on this data report is included on

n. x 11 in.,

and the number of sheets is recorded at the top of this form.

each sheet is.. umbered This form (E00029) may be obtained from the Order Dept., ASME,345 E. 47th St., New York N.Y

. 10017 Page 1 of 25 l

i

,o

FORM NIS 1 (back)

8. Examination Dates __1/12/80 to 3/9/80

9. Inspection Interval from 12/29/72 to 12/29/82

10. Abstract of Examinations. Include a list of examinations and a statement concerning status of work required for current interval.

See Pages 3, 4 and 5

11. Abstract of Conditions Noted See Page 6

12. Abstract of Corrective Measures Recommended and Taken See Page 7 We certify that the statements made in this report are correct and the examinations and corrective mea-sures taken conform to the rules of the ASME Code,Section XI.

Date 22 19 fc Signed es "4

/

By '//[i, >. [Ms A

~

mc

/

',/7

/

Owner e g,a r Certificate of Authorization No.(if applicable)

DPR-36 Expiration Date October 21,2008 CERTIFICATE OF INSERVICE INSPECTION 1, the undersigned, holding a valid commission issued by the National Board of Boiler a Pressure Vessel Inspectors and/or the State or Province of Massachusetts Hart ford, Ct.

and employed by HSB I&I Co.

or have inspected the components described in this Owners' Data Report during the period 1/12/80 to 3/11/80

, and state that to the best of my knowledge and belief, the Owner has performed examinations and taken corrective measures described in this Owners

• Data Report in accordance with the requirements of the ASME Code,Section XI.

By signing this certificate neither the inspector nor his employer makes any warranty, expressed or implied, concerning the examinations and corrective measures described in this Owners' Data Report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personalinjury or property damage or a loss of any kind arising from or connected with this inspection.

Date

/V14 V D 19 T6 M

Commissions CT. 8, f 2 7 Inspe"ctor's S'ignature National Board, State, Province and No.

I Page 2 of 25

r 1.

Owner Maina Yrnku Atmic Pow 2r Crepany, Winemc2t, Meine 2.

Plant Maine Yankee Atomic Power Plant, Wiscasset, Maine 3.

Plant Unit #1 4.

Owner Certificate of Authorization DPR-36 5.

Commercial Service Date 12/29/72 6.

National Board Number for Unit Reactor Vessel 20865 10.

Abstract of Examinations COMPONENT CATEGORY NUMBER OF COMPONENTS EXAMINED EXAMINATION Reactor Vessel B-G-1 18 closure studs PT, UT B-G-2 8 flange bolts, nuts, flange VT ligaments of unused instru-mentation nozzle.

Pressurizer B-D 100% of 2 nozzle to vessel UT welds.

B-F 100% of I safe-end weld PT, UT B-F 100% of 3 safety / relief PT valve nozzle to safe end welds - special inspection, augmented ISI.

Page 3 of 25

c.

1.

Owner Main,Yrnk 2 Atomic Pow)r C*mpany, Wiccen"at, Maine 2.

Plant Maine Yankee Atomic Power Plant, Wiscasset, Maine 3.

Plant Unit #1 4.

Owner Certificate of Authorization DPR-36 5.

Commercial Service Date 12/29/72 6.

National Board Number for Unit Reactor Vessel 20865 10.

Abstract of Examinations COMPONENT CATEGORY NUMBER OF COMPONENTS EXAMINED EXAMINATION Steam Generator B-D 100% of primary inlet nozzle UT No. I and primary outlet nozzle including inner radii.

B-G-2 100% of 20 studs, nuts, VT ligaments of primary manway (inlet side)

B-H 7 feet of support skirt to UT vessel weld (integrally welded support)

Piping Pressure B-F 100% of 5 safe-end welds PT, UT Boundary Page 4 of 25

1.

Own3r Main 2 Y nkoa Atomic Powtr Crmp?ny, Winen91st, M:ine 2.

Plant Maine Yankee Atomic Power Plant, Wiseasset, Maine

• ~ ' -

3.

Plant Unit #1 4.

Owner Certificate of Authorization DPR-36 5.

Commercial Service Date 12/29/72 6.

National Board Number for Unit Reactor Vessel 20865 10.

Abstract of Examination COMPONENT CATEGORY NUMBER OF COMPONENTS EXAMINED EXAMINATION Piping Pressure C-E-1 100% of blowdown penetration PT, RT Boundary weld - baseline inspection of new installation 100% of 3 new support member PT, VT welds.

Installation baseline examination C-G 100% of 200 blowdown piping PT, VT socket welds - Baseline inspection of new installation Status of work required for current Interval:

This inspection completes the second period of the first interval.

The regenerative heat exchanger welds identified in the 1978 NIS-1 report are classified SC-II.

The 1978 report erred in defining these welds as SC-1 and therefore, subject to examination.

Page 5 of 25

'1.

Owner Main 7 Yrnkas Atemic Powar Crmpany Wicc m st, Meine 2.

Plant Maine Yankee Atomic Power Plant, Wiscasset, Maine 3.

Plant Unit #1 4.

Owner Certificate of Authorization DPR-36 5.

Commercial Service Date 12/29/72 6.

National Board Number for Unit Reactor Vessel 20865 11.

Abstract of Conditions Noted WELD / COMPONENT EXAMINATION METHOD INDICATIONS Valve PR-M-16 Visual at leak test Gland packing leaked during RCS leak test PZR NZL, PN-3 VT Narrow grinding PN-4 VT wheel marks on nozzle barrel -

maximum depth 1/8" and 1/16" length

• 1" Page 6 of 25

.1

1.

Own2r Maina Yrnket Atomic Power Ccep ny, Wincersst, M int 2.

Plant Maine Yankee Atomic Power Plant, Wiscasset, Maine 3.

Plant Unit #1 4

Owner Certificate of Authorization DPR-36 5.

Commercial Service Date 12/29/72 6.

National Board Number for Unit Reactor Vessel 20iS5 11.

Abstract of Corrective Measures Recommended and Taken WELD / COMPONENT CORRECTIVE MEASURES Valve.PR-M-16 Added packing PZR NZL - PN-3 Evaluated notch configuration PN-4 and found condition is within design considerations (See summary report for details).

3 t

i Page 7 of 25

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o FIGURE 104

1.0 INTRODUCTION

This report describes the inservice inspection performed during the 1980 refueling outage at the Maine Yankee Atomic Power Station, Wiscasset, Maine. The non-destructive examination procedures used for inservice inspection were in accordance with the 1974 edition of the ASME Boiler and Pressure Vessel Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components", up to and including the Summer 1975 Addenda; except that piping examinations and acceptance criteria for piping examinations were in accordance with the Summer 1976 Addenda. Additional examinations were as delineated by Plant Technical Specifications.

Inspections were conducted to the extent allowed by system design and the non-destructive testing technology employed. This report summarizes the areas examined, the type of examinations, test data results, evaluations, and repairs. Manual ultrasonic testing techniques were employed in conjunction with visual, liquid penetrant, and radiographic examinations.

1.1 Examination Methods All non-destructive examinations were performed in accordance with Yankee Atomic Electric Company procedures found in Engineering Guidelines Book III, " Inservice Inspection NDE Procedures". The examination procedures were reviewed and approved by personnel qualified to SNT-TC-1A, Level III; and conform to the requirements of ASME ~Section XI (S' 75 & S' 76) and applicable portions of ASME Section V (S' 73). The inservice inspection examinations were performed and evaluated by personnel qualified to the 1975 edition of SNT-TC-1A.

Page 16 of 25

The following procedures were utilized for the examinations:

PROCEDURE NUMBER REVISION TITLE YA-ISI-l 2

Inservice Inspection Program Requirements YA-VT-1 1

Visual Examination Procedure YA-PE-2 1

Liquid Penetrant Examination YA-RT-1 1

Radiographic Examination YA-UT-1 1

Ultrasonic Examination - General Requirements YA-UT-4 1

UT of Vessels - Nozzle to Vessel Welds YA-UT-5 1

UT of Integral Support Attachment Welds YA-UT-7 1

UT of Bolting YA-UT-ll 1

UT of Piping Dissimilar Metal Welds YA-UT-13 0

UT of Vessel - Nozzle Inner Radius YA-UT-14 0

UT of Piping - Base Material and Weld Heat Affected Zones The technique sheets used for these examinations are as follows:

TECHNIQUE SHEET NO.

REVISION SUBJECT UT-13, TS-7 0

Nozzle Inner Radius Examination Page 17 of 25 e

J

1.2 Evaluation of Data The examination results were reviewed by personnel qualified to SNT-TC-1A Levels II and III.

Indications were evaluated to the acceptance standards in the 1974 edition of the ASME Boiler and Pressure Vessel Code Section XI, up to and including the Summer 1975 Addenda, except that the acceptance standards for piping examinations are in accordance with the Summer 1976 Addenda.

1.3 Examination Results Summaries of all examinations that were performed are contained in Section 2.0 of this report. All detailed examination data, calibration records, material and equipment certifications, personnel qualifications,.

and procedures are maintained at the Plant site as part of the permanent Plant records.

Details of reportable indications and corrective measures can be found in the applicable parts of Section 2.0.

l Page 18 of 25 m

l 2.0

SUMMARY

OF EXAMINATIONS This section summarizes the inservice examination data for the 1980 refueling outage at Maine Yankee.

The results are summarized in accordance with the examination categories of the ASME Code Section XI.

2.1 Reactor Vessel B-G-1 Pressure Retaining Bolting > 2 Inch Diameter Eighteen reactor head closure studs; numbers 2, 7, 8, 10, 12, 14S, 20, 22, 24, 26, 28, 32, 34, 40, 46, 50, 52, and 54 were examined ultrasonically, and by liquid penetrant. There were no significant indications.

B-G-2 Pressure Retaining Bolting < 2 Inch Diameter Eight bolts and nuts on the reactor head instrumentation nozzle

1. 66 were visually examined while installed.

No recordable indications were observed.

2.2 Pressurizer B-D Nozzle-To-Vessel Welds Two nozzles, No. PN-8 and PN-15, were examined ultrasonically.

The examination included'100% of each nozzle-to-vessel weld.

No recordable indications were found.

B-F Nozzle-To-Safe End Welds _

The nozzle-to-safe end weld on safe end SE-N-3 (NZL PN-3) was 100%

exami ed by both ultrasonics and liquid penetrant. No recordable indications were found by either method.

Page 19 of 25

Visual Indications on Nozzle Barrels Visual indications of grind outs from a narrow wheel were found on NZL PN-3 and PN-4 The indications were found when performing an augmented liquid penetrant examination of the NZL safe ends.

The indication on PN-3 la located in the nozzle barrel region at the 8 o' clock position.

It is approximately 1" long, 5/8" wide, and 1/8 to 1/16 inch deep. The sides and edges are not blended to code tapers (3:1); however, there are no sharp corners or fillets.

The indications on PN-3 are located at the 4 and 5 o' clock positions on the nozzle barrel. They are 1 inch long, approximately 5/8 inch wide and 1/16 and 1/8 inch deep respectively.

The sides and edges are not blended to code tapers, however, there are no sharp corners or fillets.

Minimum wall calculations for this area show a.149" minimum design with 1.27" minimum furnished. Reduction of the wall size to 1.12 will not reach minimum design or cause pressure / temperature failure.

The gauge configuration is not that of a sharp cornered notch, but rather a depression made from a 1/8 inch thick grinding wheel, probably 4 inch in diameter.

The wheel was worked as the grinding was performed, so no sharp edges are present and sides are tapered. The tapers are approximately 1 to 1 on the sides and greater than 4 inch radius on the ends.

The design report for the pressurizer did not require a fatigue analysis of the safety valve nozz1'es since the exclusion requirements of ASME Section III were satisfied.

Page 20 of 25

In the bending direction a conservative approximation of the grind area is a circumferential groove 0.125 inches deep. This results in a stress concentration factor of approximately two (2).

An-increase in bending stress by a factor of two still does not impact the Code exclusion on fatigue.

In the hoop stress direction, the actual nozzle stress is approximately one fif th of the allowable value per Code.

In the hoop direction the effect of the shallow grind area is negligible; however, a significant stress concentration could be applied without affecting the original Code analysis.

Finally, since no credit has been taken for area reinforcement in the nozzle the area compensation calculations are unaffected.

'Page.21 of 25

2.3 Steam Generator No. 1 B-D Nozzle-To-Vessel Welds Primary inlet and outlet nozzles were ultrasonically examined. All areas of the nozzle to vessel welds were examined, except as limited by insulation studs. No recordable indications were observed.

B-H Integrally Welded Vessel Supports, The support skirt to vessel weld was examined ultrasonically starting under the centerline of the cold leg nozzle, CW from above, for a length of seven (7) feet. No recordable indications were observed.

B-G Pressure Retaining Bolting < 2 Inch Diameter Twenty studs and nuts were visually examined while installed on S.G.

1. 1 hotleg manway. No recordable indications were observed.

2.4 Piping B-F Pipe-To-Safe End Welds Five dissimilar metal welds were 100% examined by ultrasonics and liquid penetrant. Ultrasonic examinations were performed from both the ferritic and austenitic sides of the weld. The five safe end welds examined were:

RC-5-64, RC-46-78, RC-48-14, RC-18-25, and DRH-6-104.

There were no recordable indications observed during the ultrasonic or liquid penetrant examinations.

Page 22 of 25

's L

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C-E-1/C-G Pressure-Retaining Welds in Piping New installation welds on a 2" blowdown piping system were 100%

examined by the liquid penetrant method and visual method.

All indications were mechanically removed and rewelded as required.

Final LP inspections were performed on all repairs prior to doing a preservice hydro. These examinations provide documented baseline data for the construction / modification of the blowdown system as required by Maine Yankee QA Program and ASME Section IX.

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l' 3.0 ' SYSTEM PRESSURE LEAK TEST The reactor coolant system was given a leak test at its normal operating pressure. Visual examinations were performed after the system had been under pressure for four hours. This test was performed as required by ASME Section XI Article IWB-5000 and examination categories B-E and B-P.

Numerous minor leakages were detected at packing glands and gasketed closures while increasing the system pressure to the normal operating pressure. All excessive leakages were corrected prior to commencing the final system leak test.

On the final examination the packing on valve PR-M-16 was found to be leaking again.

It was determined that the leak rate was unacceptable for normal operation and that more packing would be required. Packing was added and the condition corrected.

No other leakages (other than normal controlled leakages) were detected on the final reactor coolant system leak test.

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4.0 CONCLUSION

S r

The examinations performed during this inspection outage completes

- the inservice inspection requirements for the second inspection period of the first interval.

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