ML19263D348

Forwards Inservice Insp Examination Rept, 781021-1216

ML19263D348
Person / Time
Site:	Yankee Rowe
Issue date:	03/15/1979
From:	Groce R YANKEE ATOMIC ELECTRIC CO.
To:	Office of Nuclear Reactor Regulation
References
WYR-79-31, NUDOCS 7903270482
Download: ML19263D348 (22)

Text

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Telephone 617 366-9011 TwK 710-390-0739 YANKEE ATOMIC ELECTRIC COMPANY s.a.2.1

{b bl 20 Turnpske Road Westborough, Massachusetts 01581 9

WYR 79-31 March 15, 1979 United States Nuclear Regulatory Commission Washington, D. C.

20555 Attention: Office of Nuclear Reactor Regulation

Reference:

License No. DPR-3 (Docket No. 50-29)

Subject:

Yankee Rowe Inservice Inspection Report In accordance with Article IWA-6000 of Section XI of the ASME Boiler and Pressure Vessel Code, we hereby submit the Yankee Rowe Inservice Inspection Examination Report. This report describes the inservice examina-tions performed during the period October 21, 1978, through December 16, 1978.

We trust this information is acceptable to you; however, should you have any questions, please contact us.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY

[N

'W Robert H. Groce Licensing Engineer RTT/em Enclosure 7903276<-tTs>

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9

INSERVICE INSPECTION EXAMINATION REPORT Yankea Atomic Electric Company Yankee Nuclear Power Station t

October 21, 1978 through December 16, 1978 O

Preface This summary report covers the inservice inspection of Yankee Nuclear Power Station during the period October 21, 1978 through Dece=ber 16, 1978.

Included in this report is the Form NIS-1 as required by the provisions of ASME Section XI and a sum =ary report of the examinations performed, conditions observed, and corrective measures taken.

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e Table of Contents Page i

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NIS-1 OWNER'S DATA REPORT

SUMMARY

REPORT 5

1.0 INTRODUCTION

5 1.1 Examination Methods 7

1.2 Examination of Data 7

1.3 Examination Results 8

2.0

SUMMARY

OF EXAMINATION 1

8 2 1 Reactor Vessel Data 9

2 2. Pressurizer Data l

10 2 3 Steam Generator Data 11 2.4 Pump Data 12 2.5 Feed and Bleed Heat Exchangers 13 2.6 Piping Data 14 2.7 Hangers and Support Data 15 2.8 Valve Data 18 30 SYSTEM PRESSURE LEAK TEST 18

4.0 CONCLUSION

S 11

FORM NIS-1 OWNERS' DATA REPORT FOR INSERVICE INSPECTIONS As required by the Provisions of the ASME Code Rc!cs

. Owner Yankee Atomic Electric Company. 20 Turnoike Road. Westborn, MA 019R1 (Name and Address of Owner)

Yankee Nuclear Power Station, Rowe, Massachusetts 01367 (Name and Address of Plant)

3. Plant Unit. Yankee Rowe

4. Owner Certificate of Authorization (if required) DPR-3

5. Commercial Service Date 7/1/61

6. National Board Number for Unit Reactor #NB 2396U

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

Province No.

Board No.

Reactor B&W 610 0011 NA 2396 U Pressurizer B&W 610 0011 NA 2397 UE 22 Steam Generator W

NA NA 404E (Z) 1 Piping S&W NA NA NA Note: Supplemental sheets in form of lists, sketches, or drawings may be used provided (1) size is 8% in. x 11 in.,

(2) information in items 1 through 6 on this data report is included on each sheet, and (3) each sheet is numbered and the number of sheets is recorded at the top of this form.

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

10017 111

FORAi NIS-1 (back)

8. Examination Dates 10/21/78 to 12/16/78 9. Inspection Interval from 12/1/74 to 12/1/84

10. Abstract of Examinations. Include a list of examinations and a statement concerning status of work required for current interval. See pages 2 and 3

11. Abstract of Conditions Noted See pagea 3 and 4

12. Abstract of Corrective Measures Recommended and Taken see page 4 We certify that the statements msde in this report are correct and the examinations and corrective mc.-

sures taken conform to the rules of the ASME C de, Sectio h h_

Date /ftbMk l(

l9h Signed

.r_i__v7_s m

_ By Owner Certificate of Authorization No. (if ppplicable)

DPR-3 Expiration Date 11/4/97

( U.S.NRC Facility License No.)

CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and/or the State or Province of VasL and employed by H.S.E.I.&I. Co.

of ected the components described in this Owners' Data Report during the period have insp/16/78

__10/21/78 to 12

. 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.

Var. 14 19 79 Date C

Nat'l Ed.

7933

- ' ' Inspector's Signature { ANII[ Commissions National Board, State. Province and No.

iv

Job Order List The following repairs, were performed in accordance with Section XI.

Documentation is available at the plant site.

Job Order #

1.

Post LOCA Recirculation System 78-62 2.

Pressurizer Spray Line Repair 78-206 3

1. 4 Main Coolant Check Valve Repair 78-37 4.

1. 3 MC Pump Vent Valve Repair 78-188 5.

Steam Generator Tube Repair (Explosive Plugging)78-237 6.

CC-V-676 Component Cooling Valve, Repair 18-250 7.

CC-V-654 Component Cooling Valve, Repair 78-256

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Abstract of Examinations Component Category Examination Method Reactor B-C Flange to vessel weld, 25%

Mut B-C Flange to head weld, 25%

Mut B-G-1 2 Reduced dia. studs Mut, Vis.

B-G-1 50 Closure studs Mut B-G-1 Ligaments between stud holes, 25%

Mut B-I-l Closure head cladding Vis.

Pressurizer B-D Safety valve nozzle to vessel weld Mut B-H 1 Vessel support Mut, Vis.

B-E-1 Heater nozzle to shell weld Mut, Vis.

B-G-2 Pressure retaining bolting <2" diameter Vis.

B-G-1 Pressure (manway bolting) retaining bolting >,2" Mut G-2 Safety valve flange studs <2" Mut Steam Generator #2 B-F Nozzle to safe end weld, inlet Vis., Mut, PT C-C Integrally welded support lug PT B-F Nozzle to safe end weld, outlet Vis., Mut, PT C-D Sec. side manway bolting, 20 Vis. - 8 MP Vis, MP Steam Generator #4 G-2 Primary manway b.olting Vis.

C-D Sec. side manway bolting, 20 Vis. - 8 MP Vis., MP Steam Generator #1 C-D Sec. side manway bclting, 20 Vis. - 8 MP Vis., MP Steam Generator #3 C-D Sec. side manway bolting, 20 Vis. - 8 MP Vis., MP Feed & Bleed Rx B-B Circumferential welds (two lower)

PT Pumps G-1 Pressure retaining bolting, M.C.P.#3 Mut, Vis., MP K-2 Hangers, M.C.P. #2 Vis.

G-1 Pressure retaining bolting, >2" Vis.

B-L-1 Pump casing, MCP#3 Vis.

Valves B-M-2 2 Valve bodies Vis.

B-G-1 Valve bolting Mut B-G-2 13 Valves, bolting Vis.

Piping B-J 4 Main coolant pipe welds Mut, Vis.

B-F 2 Main coolant safe-end welds Mut, Vis., PT B-K-2 13 Piping support components Vis.

C-D 2 Valves, bolting Vis.

C-E-2 31 Piping support components Vis.

C-C 3 Integrally welded support Vis.

C-B LPST Ex nozzle to vessel weld MP B-J Loop #4 charging line PT Piping B-J 30 circumferential welds Mut (Base line-ECCS System) 11.

Abstract of Conditions Noted

1) Visual examination of two (2) hangers PRZ5-H1 and PRZ5-H2 on the pressurizer spray line showed loose bolting.

2). Visual examination on the loop four (4) check valve showed excess wear on the pivot blocks.

3) Liquid penetrant examination on the north lug of steam generator

I no. 2 revealed numerous rounded indications.

12.

Abstract of Corrective Measures Recommended and Taken

1) The hanger bolting on the 2 pressurizet spray line hangers were tightened and adjusted.

2) The pivot blocks on loop four (4) check valve were replaced by plant personnel. One main loop check valve is disassembled each outage in rotation. Each will be visually examined in turn.

3) The rounded indications noted on the north lug of steam generator no. 2 were evaluated and determined to be from original manufacture.

Yankee Rowe will examine one lug per generator per outage.

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1.0 INTRODUCTION

This report describes the inservice inspection performed during the 1978 refueling outage at the Yankee Nuclear Power Station, Rowe, Massachusetts. The non-destructive p ocedures used for Inservice Inspection were in accordance with the ASME Boiler and Pressure Vessel Code,Section XI,

" Rules for Inservice Inspection of Nuclear Reactor Coolant System", S75 edition except that B-F & B-J category welds were examined per the S-76 addenda Appendix III as referenced by the 4nt technical specifications.

Pressure boundary welds and adjacent 5 e metal were inspected to the extent that the system design and non-destructive testing technology permitted.

This report summarizes the areas examined, the type of examinations, the results of the test data, evaluations and repairs. Manual ultrasonic testing techniques were employed in conjunction with visual, liquid penetrant, and magnetic particle examinations.

1.1 Examination Methods All non-destructive examinations were performed in accordance with the procedures contained in the Yankee Atomic Electric Company, Engineering Guidelines, Book III, " Inservice Inspection NDE Procedures". The examination procedures were reviewed and approved by personnel qualified to SNT-TC-1A Level III..These procedures conform to the requirements of ASME Section XI (S'76) and the referenced parts of ASME Section V (S'76), except where these editions are in conflict with the technical specification requirements.

The inservice examinations were performed and evaluated by technicians qualified to the 1975 Edition of SNT-TC-1A.

The procedures used for these examinations are as follows:

Procedure Number Rev.

Title YA-ISI-l 2

Inservice Inspection Program Requirements YA '.T-1 2

Visual Examination Procedure YA-PE-2 2

Liquid Penetrant Examination YA-MP-1 1

Magnetic Particle Examination YA-UT-1 1

Ultrasonic Examination - Genera' Requirements YA-UT-2 1

UT of Vessels-Circumferential, Longitudinal, Meridional and Flange Welds YA-UT-3 1

UT of Vessels-Fla..ge to Shell Weld from Flange Face YA-UT-4 1

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

UT of Vessels - Integral Support At tac hment YA-UT-6 0

UT of Flange Ligaments YA-UT-7 1

UT Exam of Bolting YA-UT-9 1

UT Exam of Piping-Ferritic Welds YA-UT-10 1

UT Exam of Piping - Austenitic Welds UA-UT-ll 1

UT Exam of Piping - Dissimilar Metal Welds YA-UT-16 0

UT Exam of Full Penetration Welds Per Section V Article V OP-4200 Main Coolant System Leak Inspection or Pressure Test r-

The Technique Sheets used for these examinations are as follows:

Technique Sheet No.

Rev.

Subject YA-UT-3,TS-1 2

Flange to vessel weld from flange face-YR-ll 1.2 Evaluation of Data The examination results were reviewed at the site by personnel qualified to SNT-TC-1A Levels II and III.

Indications were evaluated to the acceptance standards as defined in the Yankee Nuclear Power Station Technical Specifications.

1.3 Exanination Results Summaries of all the examinations that were performed are contained in Section 2.0 of this report. The detailed examination data along with the calibration record, procedures, equipment certifications and personnel qualifications are maintained at the plant site.

There am ao reportable indications found with ultrasonic examination method.

1.3 Examination Results There were no reportable indications found with liquid penetrant examination methods.

Visual examination established that there was no change to the cladding defects in the I.D. of the RPV head (see Section 2.1).

Magnetic particle examination methods revealed no reportable indications.

There were no reportable indications found with altrasonic examination methods.

2.0

SUMMARY

OF EXAMINATION This section summarizes the inservice inspection data for the 1978 refueling outage at the Yankee Nuclear Power Station, Rowe, Massachusetts.

The results are summarized by ASME Code Categories of Section XI.

2.1 Reactor Vessel Data Category C - Head to Flange and Vessel to Flange Circumferential Welds The head to flange and vessel to flange welds were examined ultrasonically adjacent to the area of stud holes 15 through 38.

No reportable indications were found.

Category C Closure Studs The "L" set of closure studs including two (2) reduced diameter studs were examined ultrasor.ically. There were no recordable indications. The "B" set was replaced in the system. This set had prior examination.

Catenory G Ligaments Eetween Threaded Stud Holes The flange ligaments between stud holes 15 through 38 vere ultrasonically examined. No reportable indic ations were found.

Category I-I - Closure Head Cladding Reactor engineering visually examined the closure head cladding over the entire inside surface of the reactor vessel head. Particular attention was paid to the known clad defect (See 77 ISI Report) near stud hole #13.

The defect was measured and photographed. It was determined no change has occurred.

2.2 Pressurizer Data Category D - Two Safety Valve Nozzle to Vesse? Welds The safety valve nozzle to vessel welds were examined using ultrasonic examination techniques. No recordable indications were found.

Category E Heater Nozzle to Shell Weld The northwest and northeast Heater Bundles were visually examined. No reportable indications were found. Ultrasonic examination of the heater nozzles was also performed. No reportable indications were found.

Category C Pressure Re'_nining Bolting Greater than 2 Inches Twelve (12) manway studs were examined ultrasonically.

No recordable indications were found.

Category G Pressure Retaining Bolting Less Than 2 Inches Twelve (12) studs on the northwest heater were visually examined. No recordable indications were found.

Category H - Integrally Welded Vessel Supports The north lug was examined visually and ultrasonically. The ultrasonic examination was for laminar tearing under the lug in the vessel plate.

No recordable indications were found.

2.3 Steam Generators Category F - Nozzle to Safe-End Welds The inlet and outlet nozzle safe ends of steam generator number 2 were examined using ultrasonic, penetrant, and visual techniques.

The ultrasonic examination was performed from both the austenitic and ferritic sides, using calibration blocks of comparable material. No recordable indications were found by any of the examination techniques.

Category G Pressure Retaining Boltira Less Than 2 Inches Diameter All the primary side manway bolting on steam generator number 4, (24) was visually examined. No recordable indications were found.

Category C Pressure Retaining Bolting Less Than 2 Inches Diameter All the secondary side manway bolting (80) on steam generators number 1,2,3,4 was visually examined. No recordable indications were found. Magnetic

. particle examination was performed on 10% of the bolting, per code requirement. No recordable indications were found.

Category C-C - Integrally Welded Vessel Support The north lug was examined through liquid penetrant methods. Ultrasonic examination was not possible due to the configuration. Liquid penetrant was possible only in the lower 60% of the vertical sides and 100% on the lower side. No linear indications were found. Numerous rounded indications were detected and determined to be from original manufacture. Yankee Rowe will examine one lug per generator per outage.

No recordable indications were found.

2.4 Main Coolant Pumps Category G Pressure Retaining Bolting >2" Diameter All bonnet bolting on main coolant pump number 3 was examined throu; h ultrasonic and visual means. Bolts were additionally examined in a best effort using magnetic particle techniques. The pump was disassembled for maintenance. No recordable indications were found.

Category G Pressure Retaining Bolting >2" Diameter All bonnet bolting on main coolant pump numb ~er 2 was visually examined in place. No recordable indications were found.

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Category L Pump Casings A visual examination was performed on number 3 main coolant pump casing.

The pump was disassembled for maintenance.

No recordable indications were found.

Category K Supports and Hangers The pump supports and hangers on main coolant pump number 2 were visually examined. No recordable indications were found.

Deficiencies in Pumps and Valves Program Due to procedural oversight, four power operated valves in the nitrogen supply system for ths ECCS accumulator were full stroke cycled per Section XI requirements, but the travel times were not recorded. These valves are:

Valve Procedure SI-TV-604 OP-4636 SI-TV-605 OP-4636 SI-TV-606 OP-4636 SI-TV-608 OP-4615 The procedures are being revised to include travel time.

2.5 Feed and Bleed Heat Exchangers Category B-B - Pressure Retaining Welds in Vessels Circumferential welds on the lower shell of two (2) feed and bleed heat exchangers was examined through liquid penetrant methods. The cast austenitic stainless cor.struction of the heat exchangers prevented ultrasonic examination. No recordable indications were found.

2.6 Piping Data Category C Pressure Retaining Bolting Less thau 2 Inches Diameter All the bolting on the flanges of pressurizer safety valve SV-181 and SV-182, was examined ultrasonically. No recordable indications were found.

Category J - Piping Pressure Boundary Welds Ultrasonic and visual examination was performed on four (4) welds MC-2-13, MC-2-14, MC-2-18 and MC-2-19 of the Loop 2 main coolant piping. No recordable indications were found.

Pressurizer Piping Liquid penetrant examination was performed on four (4) spray line welds 206-5, 206-7, 206-8 and 206-9.

No recordable indications were found.

Charging Line Piping Liquid penetrant examination was performed on three (3) pipe socket welds and one (1) butt weld in the charging system, between CH-MOV-524 and CH-611A. No recordable indications were four,d.

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il l-1, Safety Iniection I

A baseline examination was performed on all new construction velds in safety Ultrasonic injection class one systems modified during this outage.

techniques were used to perform these examinations. No recordable

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indications were found.

e The following is a breakdown of the butt welds examined:

i

1) 26 pipe welds - 10" seh 40

- stainless 4 pipe welds - 6" sch 40

- stainless Category C-G - Pressure Retaining Welds Two (2) circumferential welds MS-1-1 and MS-1-2 on the 14" main steam line were examined ultrasonically. No recordable indications were found.

Low Pressure Surge Tank Cooling Line and four (4) longitudinal Eight (8) circumferential velds LPSTC-3,4,5,6,7,8,15,16, welds LPSTC-L1, L2, L3, L6 were ultrasonically examined. No recordable indications were found.

Categorv C-B - Low Pressure Surge Tank Cooling Heat Exchanger Nozzle to Vessel Weld No Two (2) saddle welds were examined through magnetic particle methods.

recordable indications were found.

2.7 Hangers and supports

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Visual examination was performed on 45 hangers. These hangers were on Class I and 11 piping systems. They are defined below:

Class I Class II Loop #2 Main Coolant Pipe

-2 Main Steam Lines

- 12 Loop #3 Main Coolant Pipe

-2 Feedwater Lines

- 13 Loop #1 Bypass Line

-1 LPST Cooling Line

-2 Loop #2 Bypass Line

-2 Condensor Dump Line

-4 Shutdown Cooling

-2 VC Heating Header Support

-1 Safety Injection

-2 Pressurizer Spray Line

-2 Loose bolting on 2 PRZ spray line hangers PRZS-H-1 and PRZS-H-2 was reported to plant maintenance. The condition was corrected. No reportable conditions exist.

Low Pressure Surne Tank Cooler - Integrally Welded Support Number one support was visually examined. No recordable indications were found.

2.8 Valves Category C Pressure Retaining Bolting Pressure retaining bolting was visually and ultrasonically examined on seventeen (17) Class I and II valves in the following systems:

1) Main Coclant Lines

.3

2) Safety Injection

-5

3) Main Coolant Bypass

-3

4) Charging Lines

-2

5) Main Steam Throttle Valve

-1

6) Aux. Steam Dump Line Valve (PCV-402)

-1

7) Main Coolant Check Valves-Loop 2 and 3 - 2 Ca tego ry M-2

- Valve Bodies Visual examination was performed on the internal surfaces of three (3) main coolant check valves in Loop 2, 3 and 4.

The main coolant check valve on Loop 4 showed evidence of pivot block wear.

• ihe blocks were replaced. Because of the same inservice time, loop 2 check valve was also opened for inspection, without any reportable conditions noted. To assure themselves, maintenance opened loop 3 check valve, noting no reportable conditions.

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SAFETY VALVE TESTS The following safety valves were subjected to test in accordance with ASME PTC 25.2 - 1966 per Section XI Subsection IWV 3510.

Valve System PRSV-181 Pressurizer Code Safety PRSV-182 Pressurizer Code Safety SV-209 Charging Line Relief SV-201A M.C. Loop Bypass Relief SV-201B M.C. Loop Bypass Relief SV-201C M.C. Loop Bypass Relief SV-201D M.C. Loop Bypass Relief SV-4 09A Main Steam Safety Valve SV-409B Main Steam Safety Valve SV-409C Main Steam Safety Valve SV-4 09D Main Steam Safety Valve SV-409E Main Steam Safety Valve SV-409F Main Steam Safety Valve SV-409G thin Steam Safety Valve SV-40911 Main Steam Safety Valve SV-409I Main Steam Safety Valve SV-409J Hain Steam Safety Valve SV-409K Main Steam Safety Valve SV-409L Main Steam Safety Valve SV-215A Low Pressure Surge Tank Safety SV-215B Low Pressure Surge Tank Safety

3.0 SYSTEM PRESSURE LEAK TESTS Sevente-n (17) hydrostatic tests were performed af ter modification or repair to safety class systems. Cf these only 2 were held for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to comply with Section XI requirements. These are:

1) Portions of auxiliary steam system (Procedure OP 2000.59)

2) Portions of spent fuel pit cooling system (Procedure OF 2000.58)

The reactor coolant system was subj ected to the required system leakage test prior to startup. All test results were acceptable.

40 CONCLUSIONS The examinations performed during this inspection outage completes all the inservice inspection requirements of the Yankee Nuclear Station Technical Specifications for the first outage of the second period of the second interval.

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