ML20128E485
| ML20128E485 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/30/1985 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20128E467 | List: |
| References | |
| NUDOCS 8505290397 | |
| Download: ML20128E485 (99) | |
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Revision 1 April 30, 1985 INSERVICE INSPECTION EXAMINATION REPORT YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION MARCH 31, 1984 THROUGH JUNE 9, 1984 8505290397 850522 gDR ADOCK 05000029 PDR 1
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~ PREFACE This summary report covers the inservice inspection of Yankee Nuclear Power Station during the period March 31, 1984 through June 9, 1984.
-Included in this report is the NIS-1 form as required by the provisions of ASME Section II.
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TABLE OF CONTENTS Page PREFACE.......................................................... 11 NIS-1 OWNERS DATA REP 0RT.......................................... 1
1.0 INTRODUCTION
...................................................... 8
' 2.0 NONDESTRUCTIVE EKAMINATION PROCEDURES............................. 9
- 3.0 EVALUATION OF DATA................................................ 11 4.0
SUMMARY
REPORT.................................................... 12 5.0 SAFETY VALVE TESTING.............................................. 26
6.0 CONCLUSION
S....................................................... 27 ill
FORSt NIS 1 OWNERS
- DATA REPORT FOR INSERVICE INSPECTIONS As required by the Prosisjons of the ASAtE Code Rules Yankee Atomic Electric Co., 1671 Worcester Road, Framingham, Ma. 01701
- 1. owner (Name and Address of Owner)
Yankee Nuclear Power Station, Rowe, Massachusetts 01367
- 2. Plant (Name and Address of Plant)
Yankee Rowe 4. Owner Certificate of Authorization (if required) DPR-3
- 3. Plant Unit
- 5. Commercial Service Date 7/1/61 6. National Board Number for Unit _ Reactnr #NR-7106d
- 7. Components inspected Manufacturer Manufacturer or Installer State or National Component or Board No.
Appurtenance or Installer Serial No. Province No.
Reactor Vessel B&W 610-0011 N/A 23964 Steam Generator W N/A N/A 404E(Z)1-Piping S6W N/A N/A N/A Pressurizer B5W 610-0011 N/A 2397 4E 22 l
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Note: Supplemental sheets in form of lists. sketches, or drawings may be used provided (1) sire is 8% in. x 11 in..
(2) information in items I 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 otetained from the Order Dept., ASME. 345 E. 47th St., New Yuk, N.Y. 10017 Page 1 of 28
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FORM NIS-1 (back)
- 8. Examination Dates 3/31/84 ,,6/9/84 9. Inspection Interval from 12/1/74 to L 12/1/R4 10.~ Abstract of Examinations. Include a list of examinations and a statement concerning status of work required for current interval. Page 3 of 28 )
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- 11. Abstra'ct of Conditions Noted Page 6 of 28 Page 7 of 28 1'2. Abs' tract 'of Corrective Measures Recommended and Taken
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' .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. .
1YV W Signed YM 4C A D 4 s C. By _ '
tk% ^
Date '
Owner li DOR-3 Expiration Date 11/4/07 Certificate of Authorization No. (if apUSde ble)C Facility License No.
CERTIFICATE OF INSERVICE INSPECTION I, the. undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel of Inspectors and/or the State or Province of Massachusetts and employed by }{SBI & I Co.
Connecticut have inspected the components described in this Owners' Data Report during the period
. and state that to the best of my knowledge and belief, the Owner 1 3/31/84 to 6/9/R4 has performed examinations and taken corrective measures described in this Owners' Data Report in accordance with the requirements of th'e 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 [- 19 MASS. 1182 .
h Commissions National Board, State Province and No.
Inspector's Signature Revision ~1' Date i;-/>
i 19 2 ( Signad- *Yjf</ily t1TDAss L By Owner h '*
k/Ww Dat $ /(s 19 W Jn J)M/ , Commissions //B 99 % Ms 154
/ " g eef7P s' Signature National Board, State, Province and No.
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4 YANKEE ATOMIC ELECTRIC COMPANY NIS-1 OWNERS DATA REPORT f
[ 10. Abstract of Examinations (Safety Class 1)
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t ASME Code (
Category No. Components Examined Method [
i B-A Closure Head Flange Circulation Weld
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UT I Vessel to Flange Weld UT B-B (2) Welds Steam Generator #3 UT (3) Pressurizer Nozzles UT l
B-D (4) Outlet Nozzles UT.VT l (4) Inlet Nozzles UT.VT (4) Outlet Nozzles to Reducer UT.VT (4) Inlet Noczles to Reducer UT VT B-E (1) Feed / Bleed Heat Exchanger Nozzle VT B-F (1) Pressurizer Safe End Weld PE.UT B-C-1 (52) Rx Head Studs, Nuts, and Washers VT,MT,UT Flange Ligaments UT ,
B-C-2 (2) Safety Valves SV-181/182 Bolting VT (20) Studs / Nuts Main Coolant Check Valve #316 UT.VT (16) Stator Cap Bolts, #2 Main Coolant Pump UT,MT VT (12) Studs / Nuts Pressurizer Primary Manway UT VT (40) Studs, #2 Steam Generator Primary Manway UT NT,VT (38) Studs, #3 Steam Generator Primary Manway UT,NT.VT (40) Studs, #4 Steam Generator Primary Manway UT MT,VT B-H (4) Welded Supports Feed / Bleed Heat Exchangers PE B-J (34) Safety Class 1 Pipe Welds UT.PE.VT B-K-1 (3) Integrally Welded Pump Supports PE F B-K-2 (5) Nonwelded Supports VT B-M-2 (1) Main Coolant Check Valve #316 Internals VT B-N-1 Interior of Reactor Vessel vT I B-N-3 Upper Core Support Barrel VT
. Lower Core Support Barrel 0.D. VT Flcf Baffle Assembly, Spacer Blocks, VT and Source Vanes Shroud Tubes from I.D. of Lower Core VT Support Barrel Bolting Upper Core Suppert Plate to VT Baffle Assembly Shroud Tube Tie-Plate Bolting VT Lower Core Support Barrel 1.D. VT Page 3 of 27 a
YANKEE ATOMZC ELECTRIC COMPANY NIS-1 OWNERS DATA REPORT ASME Code Catezory g2.:. Components Examined Methg Core Barrel Lateral Support Padu VT Thermal Thield Surface, Spacer Pins. VT Seam Clamps, and Suppor t Luss, I
U-0 Control Rod Drive Housings PT B-P System Leakage Test Conducted on all Safety VT Class 1 Systems 1
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YANKEE ATOMIC ELECTRIC COMPANY NIS-1 OWNERS DATA REPORT
- 10. Abstract of Examinatione (Safety Classes 2 and 3)
ASME Code Category No. Components Examined Method C-A (1) Weld Low Pressure Surge Tank Heat Exchanger UT C-B (1) Nozzle Low Pressure Surge Tank Heat Exchanger PE CC-CE (2) Welded Supports #3 Steam Generator PE (1) Meenanical Support #1 Steam Generator VT (3) Charging Pump Supports VT (2) Nonwelded Supports #3 Steam Generator VT (30) Nonwolded Supports Shutdown C0oling System VT (1) Welded Support Shutdown Cooling System PE (15) Nonwelded Supports, Main Steam Line VT (17) Nonwelded Supports, Feedwater Line VT (1) Vapor Container Heating, Welded Support VT (1) Low Pressure Surge Tank SV Discharge Support VT C-F (14) Shutdown Cooling System Piping Welds PE (11) Safety Injection System Piping Walds PE (5) Feedwater System Piping Welds NT.VT (10) Main Steam System Piping Welds. UT.MT,VT C-H (9) Safety Class 2 System Leakage Tests VT (13) Safety Class 2 System Hydrostatic Tests VT D-A (5) Safety Class 3 System Leakage Tests VT (3) Safety Class 3 System Hydrostatic Tests VT Page 5 of 27 h-
YANKEE ATOMIC ELECTRIC COMPANY NIS-1 OWNERS DATA REPORT
- 11. Abstract'of Conditions Noted B-D Ultrasonic inner radius examination of pressurizer nozzles PRZN-2, 5, and 6 revealed I recordable indication on nozzle PRZN-2.
Ultrasonic examination of RPV outlet nozzle FF (I.D.) revealed 1 reportable indication in the weld region.
B-F An unacceptable linear indication was identified by visual testing on the pressurizer drain line weld PRZ-SE-2.
B-G-1 Examination of 52 reactor head studs revealed a degeaded condition of two threads on r : actor stud #4.
B-G-2 Inspection of 16 bolts on #2 main coolant pump stator esp identified a break in the plating on 1 bolt.
B-J Examination of main coolant piping weld MC-1-4 identified an unacceptable linear indication.
B-K-2 Inspection of main coolant pump support CRM-H-12 identified one bolt / nut having lack of thread engagement.
B-N-3 Visual in.pection of the flow baffle assembly (lower core barrel I.D.) revealed a reportable condition.- cracked tack weld.
CC-CE Visual examination identified a total of 15 supports with discontinuities on the Shutdown Cooling System.
Visual examination identified a total of 4 supports with discontinuities on the main steam line.
Visual examination identified a total of 4 supports with discontinuities on the Feedwater System.
C-F Ultrasonic examination of main' steam line welds MS-1-15 and MS-2-ll identified two recordable indications.
Examination of 4 nonnuclear safety main steam line welds identified a nonservice-induced discontinuity (LAP) .
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YANKEE ATOMIC ELECTRIC COMPANY NIS-1 OWNERS DATA REPORT
- 12. Abstract of Corrective Measures Recommended and Taken B-D The ultrasonic indication recorded on PRZN-2 was subsequently evaluated in accordance with IWB-3514.5 and determined to.be a geometric reflector. -
The indication found in RPV outlet nozzle FF was evaluated as defined in IWB-3600. Appendix B to this report contains that evaluation. Subsequent evaluation of the sizing method was performed. Based on the data developed by this sizing study, the indication is acceptable per IWB-3512-1. Appendix C con-tains the sizing study and all pertinent information.
B-F The linear indication identified on weld PRIN-SE-2 was subsequently removed by light buffing and re-examined as satisfactory.
B-C-1 The two degraded threads on reactor stud #4 were subsequently evaluated by engineering and found to be acceptable for continued service.
B-C-2 The bolt on #3 main coolant pump stator cap which had a break in the plating was replaced in kind.
3-J The indication on main coolant piping weld MC-1-4 was subsequently removed by light buffing and re-examined satisfactory.
B-K-2 The lack of thread engagement condition identified on main
-coolant pump support CRM-H-12 was subsequently corrected and re-examined as satisfactory.
B-N-3 The indication identified on the flow baffle assembly was subsequertly evaluated and deemed inconsequential to the s t ructure .
CC-CE Tne 15 supports on the Shutdown Cooling System identified as having discontinuities were subsequently corrected and re-examined as satisfactory.
The 4 supports on the main steam line identified as having discontinuities were subsequently corrected and re-examined as satisfactory.
The 4 supports on the Feedwater System identified as having discontinuities were subsequently corrected and re-examined as satisfactory.
C-F The ultrasonic indications recorded on main steam line welds MS-1-5 and MS-2-Il were subsequently evaluated in accordance with IWB-3514.5 and determined to be geometric refle tors.
The discontinuity on #2 nonnuclear safety main steam line wrld was subsequently determined to be nonst? ice induced and accepted as is.
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1.0 INTROD'UCTION This report covers the inservice inspection of Yankee Nuclear Power Station during the period of March 31, 1984 to June 9, 1984.
The examinations performed are those of the third period of the second interval .
The required ten-year reactor pressure vessel examinations were performed this refueling. A summary of the examinations performed and conditions noted are included as part of this report along with attachments.
With the exception of several hydrostatic pressure tests, the examinations performed this refueling complete the required inspections for the third period of the second interval.
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2.0 NONDESTRUCTIVE EKAMINATION PROCEDURES Nondestructive examinations were performed in accordance with the procedures contained in the Yankee Atumit Electric Company Engineering Guidelines, Book III, " Inservice Inspection NDE Procedures", and Nuclear Energy Services examination procedures.
The examination procedures were reviewed and approved by personnel certified to Level III per SNT-TC-1A,1975 Edition, and the authorized Nuclear Inservice Inspector.
These procedures conform to the requirements of the ASME Code,Section XI,
" Rules for Inservice Inspection of Nuclear Power Plant Components", 1977 Edition, Summer 1978 Addenda.
The following procedures were used during the 1984 inservice inspections:
- 1. YA-PE-2 Rev. 3 " Liquid Penetrant Examination"
- 2. YA-MP-111 Rev. O " Procedure for Magnetic Particle Examination"
- 3. YA-VT-ll Rev. 1 " Visual Examination Procedure"
- 4. YA-RT-111 Rev. O " Radiographic Examination"
- 5. OP-4200 Rev. 10 " Main Coolant System Leak Inspection"
- 6. YA-UT-1 Rev. 2 " Ultrasonic Examination, General Requirements"
- 7. YA-UT-7 Rev. 2 " Ultrasonic Examie.ation of Bolting"
- 8. YA-UT-8 Rev. 1 " Ultrasonic Examination of Reactor Closure Nuts"
- 9. YA-UT-9 Rev. 2 " Ultrasonic Examination of Piping - Ferritic Welds"
- 10. YA-UT-10 Rev. A " Ultrasonic Examination of Piping - Austenitic Welds"
- 11. YA-UT-13 Rev. 1 " Ultrasonic Examination of Vessel Nozzles - Inner Radius"
- 12. YA-UT-14 Rev. 2 " Ultrasonic Examination of Piping - Base Metal and HAZ"
- 13. YA-UT-22 Rev. O " Ultrasonic Examination of Vessels - Circumferential, Longitudinal, Meridional, and Flange Welds"
- 14. YA-UT-44 Rev. O " Ultrasonic Examination of Vessels - Nozzle to Vessel Welds"
- 15. YA-UT-116 Rev. O Ultrasonic Examination of Full Penetration Welds"
- 16. YA-UT-6 Rev. 2 " Ultrasonic Examination of Flange Ligaments" Pags ~ 27
- 19. 83A0312 Rev. O "UT Examination of RPV Upper Shell to Flange Weld from the Flange Mating Surface"
- 20. 83A0311 Rev. O " Operation of NES Mini-Scanner System"
- 21. 83A0317 Rev. O " Visual Examination Procedure"
- 22. 83A0318 Rev. O " Visual Examination Scan Plan" The following technique sheet was used to perform the subject examination:
Technique Sheet No. Rev. Subject YA-UT-3, TS-1 1 Flange to Vessel Weld From Flange Face 4
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3 3.0 EVALUATION OF DATA All inservice examinations were performed, evaluated, and reviewed by personnel certified to Level II in accordance with SNT-TC-1A, 1975 Edition and ASME Section XI, 197 7 Edition, Summer 19 78 Addenda.
The examination methods, volumes, and evaluation of indications were in accordance with ASME Boiler and Pressure Vessel Code,Section II, 1977 Edition, Summer 1978 Addenda, axcept for Class 1 piping ultrasonic calibration. This was conducted in accordance with Article III-2000 of Appendix III, ASME Section II, Summer 1976 Addenda, as required per Plant Technical Specifications.
Summaries of the examinations that were pe eformed are contained in Section 4.0 of this report. The detailed examination data along with the calibration records, procedures, personnel, and equipment certifications are maintained at the plant site.
Attached is a summary of the examination methods, volumes, and the results and evaluation of test data thereof, including any corrective measures taken.
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.3 4.0
SUMMARY
REPORT The following is a summary of all the examinations pet f orraed. the conditions noted and corrective measures taken during the 1984 inservice inspections, with the exception of the ten-year RPV examination that is included as part of this report.
CODE CATEGORY B-B PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS Ultrasonic examinatien was performed on two welds on steam generator #3. The longitudinal shell to barrel weld SC-3-8 and the longitudinal barrel to transitional piece S0-3-9 were inspected with no recordable indications.
CODE CATEGORY B-D FULL PENETRATION WELDS OF NOZZLES IN VESSELS Ultrasonic examination was performed on 3 pressurizer nozzle. welds. Nozzle welds PRZN-2, 5 and 6 were examined with no recordable indications. A nozzle inner radius examination was also performed on all 3 nozzles. Only one indication was recorded, on nozzle PRZN-2, which was subsequently evaluated and determined to be a geometric reflector in accordance with IWB-3514.5.
Ultrasonic examination was performed on eight (8) RPV outlet and inlet no::le 1;
I.D.s. Outlet nozzle FF revealed one reportable indication. This was subsequently evaluated and determined to be acceptable (see Attachment A, B and C).
CODE CATEGORY B-E PRESSURE RETAINING PARTIAL PENETRATION WELD 9 IN VESSELS 8 No. 1 feed and bleed heat exchanger nor:le FB-1-3 was visually examined during the Main Coolant System leakage inspection (OP-4200). No leasage was noted.
CODE CATEGORY B-F PRESSURE RETAINING DISGINILAR METAL WELDS 8 An ultrasonic and 11guld penetrant examination was' performed on pressurizer safe end weld PR2-SE-1. No indications were recorded. During the examination of PRZ-SE-1, a visual inspection of weld PRZ-SE-2 identified two linear indications.
The areas were lightly buffed and re-examined with liquid penetrant which revealed the indications had been removed. A liquid penetrant examination was also performed on the weld adjacent to PRZ-SE-2. This examination resulted in no unacceptable indications.
CODE CATEGORY B-C-1 PRESSURE RETAINING BOLTING LARGER THAN 2" IN DIAMETER Fifty-two (52) reactor head closure sto , nuts, and washers (Set No. L-43) including 2 reduced diameter studs #1-638-1 and #1-674-1 were inspected. A visual examination of the washers and a visual, magnetic particle, and ultrasonic examination of the nuts resulted in no recordable indications.
Page 12 of 27
A visual and ultrasonic examination was performed on all 52 studs. During the visual examination of stud #4, two threads were found to be degraded. An engineering evaluation was conducted which determined that the stud was still acceptable for usc. The subsequent ultrasonic examination did not reveal any indications.
CODE CATEGORY B-G-2 PRESSURE RETAINING BOLTING 2" AND SMALLER IN DIAMETER A base line visual examination was performed on 16 new bolts and nuts prior to installation on pressurizer safety valves SV-181 and 182. This inspection resulted in no unacceptable indications noted.
A base line visual examination was also conducted on 6 new bolts installed on PR-MOV-191. No unacceptable indications were noted.
The following bolting / studs and nuts were inspected per I&E Bulletin 83-02,
" Bolting in RCPB Closure Connections Greater Than Six Inches":
A visual and ultrasonic examination was performed on 20 studs and nuts from
- 3 main coolant loop check valve #316. No unacceptable indications were noted.
A visual, magnetic particle and ultrasonic examination was performed on 16 bolts removed from #2 main coolant pump stator cap. A linear indication was noted on one bolt which was determined to be a break in the plating. A new bolt was installed in its place, prior to which a visual, magnetic particle, and ultrasonic examination was performed on it, with no unacceptable indications noted.
Twelve (12) pressurizer primary manway studs were visually and ultrasonically inspected with no unacceptable indications noted.
. A visual, ultrasonic, and magnetic particle examination was conducted on 40 studs from steam generator #2 primary manways, 38 studs from steam generator #3 primary manways, and 40 studs from steam generator #4 primary manways. These examinations resulted in no unacceptable indications.
CODE CATEGORY B-H VESSEL SUPPORTS Feed and bleed heat exchanger integrally welded supports BL-H-6, 8, and 10 were liquid penetrant inspected with no unacceptable indications noted.
~ CODE CATEGORY B-J PRESSURE RETAINING WELDS IN PIPING The following Main Coolant System piping welds were inspected as follows:
Weld No.
MC-1-10 Liquid penetrant examination performed - no unacceptable indications noted.
MC-1-13 Ultrasonic and liquid penetrant examination performed - no unacceptable indications noted.
MC-1-16 Liquid penetrant examination performed no unacceptable indications noted.
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Weld No.
MC 17 Liquid penetrant examination performed - no unacceptable indications noted.
MC-1-15 Liquid penetrant examination performed - no unacceptable indications noted.
MC-2-15 Liquid penetrant examination performed - no unacceptable indications noted.
MC-2-10 Liquid penetrant examination performed - no unacceptable indications noted.
MCB-2-5 Liquid penetrant and ultrasonic examination performed - no unacceptable indications noted.
.MCB-4BR-2 Liquid penetrant and ultrasonic examination performed - no unacceptable indications noted.
MC-1-4 Liquid penetrant examination was performed which identified several unacceptable linear indications. The areas were subsequently buffed and re-examined which verified the indications had been removed. An additional weld was examined in accordance with IWB-2430, which identified no unacceptable indications.
MC-2-3 Liquid penetrant examination performed - no unacceptable indications noted.
Safety Valve Ultrasonic and liquid penetrant base line examination 182 Weld performed - no unacceptable indications noted.
PRS-206-23 Liquid penetrant base line examination - no unacceptable indications ,nated.
PRS-206-26 Liquid penetrant base line examination - no unacceptable indications noted.
PRS-206-29 Liquid penetrant examination - no unacceptable indications noted.
PRS-206-39 Visual and liquid penetrant examination - no unacceptable indications noted.
PRS-43 Liquid penetrant examination - no unacceptable indications noted.
PRS-52 Liquid penetrant examination - no unacceptable indicationr noted.
L- PRS-50 Liquid penetrant eAamination - no unacceptable indicatione noted.
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E -- 7
-Weld No.
PRS-206-24A Liquid penetrant base line examination - no unacceptable indications noted.
PRS-206-24B Liquid penetrant base line examination - no unacceptable indications noted.
PRS-206-22A Liquid penetrant base line examination - no unacceptable indications noted.
PRS-206-22B Liquid penetrant base line examination - no unacceptable indications noted.
PRS-206-40 Liquid penetrant examination - no unacceptable indications noted.
Pressurizer safety valve discharge header weld caps #1 and #2 - liquid penetrant examination - no unacceptable indications noted.
SAFETY INJECTION PIPING INSPECTIONS 4
Weld No.
SI-3-7 Ultrasonic and liquid penetrant examination performed - no unacceptable indications noted.
SI-3-7A Ultrasonic and liquid penetrant examination performed - no unacceptable indications noted.
SI-3-8 Ultrasonic and liquid penetrant examination performed - no I unacceptable indications noted.
SI-4-1 Ultrasonic and liquid penetrant examination performed - no unacceptable indications noted.
f SI-147 Liquid penetrant examination perforiaed - no unacceptable indications noted.
SI-153 Ultrasonic examination performed - no unacceptable indications noted.
SI-150 Ultrasonic examination performed - no unacceptable indicatisns noted.
SI-137 Liquid penetrant examination performed - no unacceptable indications noted.
CODE CATEGORY B-K-1 SUPPORT MEMBERS FOR PIFING. PUMPS. AND VALVES A liquid penetrant examination was performed on 3 main coolant pump integrally welded supports , :'RM-H-5, CRM-H-12, and CRM-H-26. No unacceptable indications were identified.
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CODE CATEGORY B-K-2 COMPONENT SUPPORTS FOR PIPING PUMPS AND VALVES A visual examination (VT-3 and VT-4) was performed on the following component rupports:
MCB1-H-1 CRM-H-26 CRM-H-5 SDC-H-1 CRM-H-12 BL-H-1 Only one discontinuity was noted. CRM-H-12 was rejected due to a nonservice-induced condition (lack of thread engagement), which was subsequently corrected and re-examined ss satisfactory.
CODE CATEGORY B-M-2 PUMP CASINGS AND VALVE BODIES Main Coolant System loop #3, check valve 316, was disassembled and a visual inspection was performed on the internals. This resulted in no unacceptable indications.
CODE CATEGORY B-P ALL PRESSURE RETAINING COMPONENTS A system hydrostatic pressure test was conducted on all repaired / replaced sections of the Main Coolant System prior to startup.
Plant Procedure OP-4200, Rev. 10 " Main Coolant System Leak Inspection or ISI Pressure Test", was perform 9d at 2040 psi at 5140F for two hours. The areas inspected were uninsulated. The visual inspection (VT-2) was acceptable.
In conjunction with the above test, the remaining portions of the Reactor Coolant System was subjected to the required system leakage tests.
No serious degradation was noted during the inspection other than normal packing leaks which were corrected at the time of the inspection.
SAFETY CLASS 2 COMPONENTS CODE CATEGORY C-A PRESSURE RETAINING WELDS IN PRESSURE VESSELS An ultrasonic examination was performed on the low pressure surge tank heat exchanger head circumferential weld (LPST-Rx-H-1). No recordable indications were observed.
CODE CATEGORY C-B PRESSURE RETAINING NOZZLE WELDS IN VESSELS The low pressure surge tank heat exchanger nozzles to shell welds #1 and #2 (LPST-Hx-N1 and 2) were subjected to a liquid penetrant examination. No unacceptable indications were recorded.
An ultrasonic inner radius examination was attempted on #3 steam generator main steam outlet nozzle SG-3-S01. Due to physical limitations of the component, no relevant data could be obtained, therefore the examination was deleted.
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CODE CATECORY CC-CE Due to access limitations, a "best effort" liquid penetrant examination was performed on #3 steam gencrator integrally welded supports (SG- 3-E and SG-3-W). No unacceptable indications were identified.
Mechanical support SG-1-236 was functionally tested (VT-4) after removal from steam generator #1. Functional acceptability was verified as satisfactory.
Component supports for charging pumps P-15-1, 2, and 3 and steam generator #3 supports SG-3-SE and SG-3-SW were subjected to a visual examination (VT-3) and found to be acceptable.
The following Shutdown Cooling System support members were subjected to visual examirations with results as follows:
Support No.
CRT-H-66 Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservice-related discontinuity (lack of thread engagement) which was subsequently corrected and re-examined as satisfactory.
CRT-H-59 Subjected to e VT-3/VT-4 visual examination. VT-3 identified a nonservice-related discontinuity (lack of thread engagement) which was subsequently corrected and re-examined as satisfactory. VT-4 identified an incorrect support setting.
This was evaluated by Engineering and found to be acceptable as is. An additional examination was performed per IWC-2430 and found acceptable.
SDC-H-1 Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-H-43 Subjected to a VT-3/VT-4 visual examination. VT-3 identified an unacceptable condition (loose nuts). This condition was corrected and re-examined as satisfactory. An additional examination was performed per IWC-2430 and found to be acceptable.
PRCL-H-53 Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservice-induced unacceptable condition (lack of thread engagement). This condition was corrected ~and subsequently re-examined and found to be acceptable.
CRT-H-71 Subjected to a VT-3/VT-4 visual examination and found to be acceptable.
CRT-H-62 Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservice-induced unacceptable condition (lack of thread engagement) which was subsequently corrected and re-examined as satisfactory.
Page 17 of 27
Support No.
PRCL-H-54 Subjected to a VT-3/VT-4 visual examination. This was an additional examination as required by IWC-2430(?). Thic examination resulted in an unacceptable VT-3/VT-4 condition.
In accordance with IWC-2430(b) the remaining number of supports within the system were subjected to applicable VT-3/VT-4 examinations with results as follow:
PRCL-H Subjected to VT-3/VT-4 examination. VT-3 identified ar. unacceptable condition (missing bolt). Subsequently corrected and re-examined as satisfactory.
PRCL-H Subjected to VT-3/VT-4 examination. VT-3 identified an unacceptable condition (loose nuts). VT-4 identified incorrect support settings. Both conditions were subsequently corrected and reinspected as satisfactory.
PRCL-H Subjected to a VT-3/VT-4 visual examination. VT-4 identified an incorrect support setting which was subsequently corrected and re-examined as satisfactory.
PRCL-H-61 , Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservice related unacceptable condition (1sek of thread engagement) which was subsequently corrected and re-examined as satisfactory.
i PRCL-S Subjected to a VT-3 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination. VT-3 identified an unacceptable nonservice-induced condition (lack of thread engagement). VT-4 identified an incorrect support setting. Both items were subsequently corrected and reinspected as satisfactory.
PRCL-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-S Subjected to a VT-3 visual examination and found acceptable.
PRCL-S-51B/S Subjected to a VT-3 visual examination which identified an unacceptable condition (loose nuts). Subsequently corrected and re-examine d as satisfactory.
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i Support No. ;
1 CRT-H Subjected to a VT-3/VT-4 visual examination. VT-3 I identified an unacceptable nonservice-induced condition (lack of thread engagement) which was subsequently corrected and re-examined as satisfactory.
CRT-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
CRT-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
CRT-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
CRT-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination. -VT-3 identified an unacceptable condition (loose bracket) which was subsequently corrected and re-examined as satisfactory.
PRCL-H Subjected to a VT-3/VT-4 visual examination. VT-4 identified an incorrect support setting which was subsequently corrected and re-examined as satisfactory.
PRCL-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination. VT-4 identified an unacceptable support setting which could not be adjusted. A new support was replaced in kind. A base line VT-3/VT-4 was conducted and found acceptable.
PRCL-S-51A- Subjected to a VT-3 visual examination and found acceptable.
PRCL-H Subjected to a VT-3/VT-4 visual examination. VT-4 identified an incorrect support setting which was subsequently corrected and re-examined as satisfactory.
A reinspection will be conduct d on selected support s on the Shutdown Coolitm System next refueling to assure no excessive movement has occurred.
Page 19 of 27
-The following main stcam line supports were visually examined as follows:
Support No.
SHP-H-83 Subjected to a VT-3/VT-4 visual examination and found to be acceptable.
SHP-H-81 Subjected to a VT-3/VT-4 visual examination and found to be acceptable.
SHP-H-91 Subjected to a VT-3/VT-4 visual examination. VT-3 identified an unaccep*.able condition (loose bolts) which was subsequently corrected and re-examined as satisfactory. In accordance with ASME IWC-2430 an additional examination was conducted and found to be acceptable.
SHP-H-3 Subjected to a VT-3/VT-4 visual examination. VT-3 identified an unacceptable condition (lack of thread engagement) which was determined not to be inservice-induced. This condition was corrected and subsequently re-examined as satisfactory.
SHP-H-65 Subjected to a VT-3/VT-4 visual examination and found acceptable.
1 SHP-H-74 Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservicq-induced unacceptable condition (lack of thread engagement) which was subsequently corrected and re-examined as satisfactory.
SHP-H-71 Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservice-induced unacceptable condition (loose bolt) which was subsequently corrected and re-examined as satisfactory.
SKP-RH-3 Subjected to a VT-3 visual examination and found acceptable.
SHP-H-66 Subjected to a VT-3/VT-4 visual examination and found acceptable.
SHP-H-92 Subjected to a VT-3/VT-4 visual examination and found acceptable.
SHP-H-91R Subjected to a VT-3 visual examination and found acceptable.
SKP-H-81R Subjected to a VT-3 visual examination and found acceptable.
SMP-H-65R Subjected to a VT-3 visual examination which identified an unacceptable nonservice-induced condition (lack of thread engagement). Subsequently corrected and re-examined as satisfactory.
SHP-H-71R Subjected to a VT-3 visual examination and found acceptable.
SHP-H-75 Subjected to a '.**- 3 visual examination and found acceptable Page 20 of 27
o Th2 following Feedwater System supports were visually examined as follows:
Support No.
WCBD-H-12 Subjected to a VT-3 visual examination and found acceptable.
WCBD-H-139 Subjected to a VT-3/VT-4 visual examination. VT-3 identified an unacceptable condition (loose nuts) which was subsequently corrected and re-examined as satisfactory. In accordance with IWC-2430, an additional examination was conducted and found acceptable.
WCBD-H-130-1 Subjected to a VT-3 visual examination and found acceptable.
WCBD-H-3 Subjected to a VT-3/VT-4 visual examination and found acceptable.
WCBD-H-138 Subjected to a VT-3/VT-4 visual examination. VT-3 identified a nonservice-induced unacceptable condition (improper support installment) which was subsequently corrected and re-examined as satisfactory.
WCBr '131 Subjected to a VT-3 visual examination and found acceptable.
WCoD-hn-122 Subjected to a VT-3 visual examination and found acceptable.
WCBD-RH-119 Subjected to a VT-3 visual examination and found acceptable.
WCBD-N-114 Subjected to a VT-3/VT-4 visual examination and found acceptable.
WCBD-N-6 Subjected to a VT-3/VT-4 visual examination and found acceptable.
WCBD-RH-117 Subjected to a VT-3 visual examination which identified an unacceptable condition (loose nuts) which was subsequently' corrected and re-examined as satisfactory. An additional examination was performed in accordance with IWC-2430 and found to be acceptable.
WCBD-H-8 Subjected to a VT-3/VT-4 visual examination and found acceptable.
WCBD-H-126 Subjected to a VT-3 visual examination. VT-3 examination identified a nonservice induced unacceptable condition (lack of thread engagement) which was subsequently corrected and reinspected satisfactory.
WCBD-H-129 Subjected to a VT-3 visual examination and found acceptable.
WCBD-H-134-1 Subjected to a VT-3 visual examination and found acceptable.
WCBD-H-134-4 Subjected to a VT 3 visual examination and found acceptable.
WCBD-H-120-4 Subjected to a VT-3 visual examination and found acceptabic.
Page 21 of 27
OTHER CC-CE SUPPORTS VC Heating A-1 Subjected to a VT-1 visual examination and found acceptable.
Penetration #4-6" Subjected to a base line VT-3 visual examination BRL (EDCR 83-10) and found acceptable.
CODE CATEGORY C-F PRESSURE RETAINING WELDS IN PIPING The following Shutdown Cooling System piping welds were subject to liquid penetrant examination with no unacceptable indications noted:
SDC-L-15 SDC-4-6 SDC-3-16 SDC-4-21 SDC-3-19 SDC-4-24 SDC-3-22 SDC-4-27 SDC-3-25 SDC-3L-21 SDC-3L-20 SDC-4-39 SDC-3L-22
-The following Safety Injection System piping welds were liquid penetrant examined with no unacceptable indications noted:
SI-003 SI-093 SI-006 SI-Op6 SI-009 SI-117 SI-018 SI-120 SI-019 SI-126 SI-027 Pressurizer piping weld PR-2-26 was liquid penetrant examined with no unacceptable indications noted.
The following Feedwater System piping welds were subjected to magnetic particle examination with no unacceptable indications noted:
FW-4-15 FW-4-18 FW-4-17 94-8D No. 3 feedwater nozzle to pipe weld (MT and VT).
MAIN STEAM SYSTEM PIPING WELDS MS-4-20 Subjected to a ultrasonic, magnetic particle, and visual examination with no unacceptable indications noted.
MS-1-15 Subjected to an ultrasonic, magnetic particle, and visual ex'mination.
a One indication was recorded during the ultrasonic examination which was subsequently evaluated as a geometric reflector.
MS-2-6 Subjected to an ultrasonic, magnetic particle, and visus 1 examination - no indicstions noted.
Page 22 of 27
MS-2-7 Subjected to an ultrasonic, magnetic particle, and visual examination - no indications noted.
MS 11 Subjected to an ultrasonic, magnetic particle, and sisual examination. One indication was recorded during the ultrasonic examination which was subsequently plotted and evaluated as being a geometric reflector in accordance with IWB-3514.5.
MS-2-12 Subjected to an ultrasonic, magnetic particle, and visual examination - no indications noted.
In accordance with the Integrated Safety Assessment Systematic Evaluation Program for Yankee Nuclear Power Station - NUREC 0825 (3-5.6). The first weld downstream of the non-return valves (all four loops) was magnetic particle and visually examined.
Only one indication was noted. During the magnetic particle examination of loop #2, a lap-like indication was revealed. Subsequently, an area of the indication was blended out to verify in fact it was a processing-related discontinuity. This indication was evaluated and determined to be nonservice-induced and accepted as is.
CODE CATEGORY C-H ALL PRESSURE RETAINING COMPONENTS The following systems were subjected to system leakage tests:
- 1. Safety Injection System
- 2. Low Pressure Surge Tank Cooling
- 4. Service Water System
- 5. Vapor Container Heating System
- 6. Main Coolant Drain System
- 7. Charging and Volume control System
- 8. Purification System
- 9. Chemical Shutdown System No serious degradation or leakage was noted other than packing leaks which were subsequently corrected, s The following system hydrostatic tests were conducted after modification or repair to safety class systems. Testing was performed in accordance with IWA/IWC-5000; no serious degradation or leakage was identified:
)
l l
Page 23 of 27
, - - b
m -.
Test System Pressure Duration Procedure Main Coolant Vent System 2200 psi 10 min OP-2000.136 Low Pressure Safety Injection Header 787 psi 10 min OP-2000.130 Charging and Volume Lantrol 52 psi 10 min OP-2000.130 Vapor Container Drain Header 190 psi 10 min OP-2000.112.4 Pressure Control and Relief System 300 psi 10 min OP-2000.123 Feedwater System 1214 psi 10 min OP-2000.126 Bleed Line 3000 psi 10 min OP-2000.127 Post Accident Hydrogen Vent System 102 psi 10 min OP-2000.128/129 Valve Stem Leakoff 375 psi 10 min OP-2000.132 Low Pressure Surge Tank 94 psi 10 min OP-2000.134 Main Steam System 1035 psi 10 min OP-2000,135 EBFP Steam Inlet , 1170 psi 10 min OP-2000.137 An Appendix "J" test was conducted on modifications to the low pressure surge tank safety valve discharge header. The test was conducted at 32 psi.
Reference plant Procedure OP-2000.133. The test was conducted and found acceptable.
SAFETY CLASS 3 CODE CATECORY D-A PRESSURE RETAINING C0KPONENTS Portions of the followin6 systems were subjected to either inservice or system leakage tests and found acceptable:
- 1. Emergency Boiler Feedwater System
- 2. Main Coolant Vent System
- 3. Spent Fuel Pit Cooling
- 4. Primary Pump Seal Water
- 5. Service Water System hydrostatic tests were. conducted on the following systems after repair or modification and found acceptable:
Page 24 of 27 1
l
Test System Pressure Duration Procedure
- 1. Emergency Boller Feed Pump Steam
-Inlet 188 psi 10 min OP-2000.137
- 2. Primary Pump Seal Vater 345/414 psi 15 min OP-2000.124
- 3. Domineralized Water / Service Water Crosstie 165 psi 10 min OP-2000.125 ADDITIONAL EKAMINATIONS In addition to the code-required examinations, the following inspections were conducted:
A visual examination was performed on the reactor head underside to monitor the known flaws. This inspection revealed that no change has occurred.
A visual inspection was also performed on the cladding of #3 steam generator hot leg and cold leg. The visual inspection resulted in no unacceptable indications noted.
The pressurizer internals were subjected to a visual inspection. This inspection concluded that no change has occurred since the last visual inspection.
l Page 25 of 71
5.0 SAFETY VALVE TESTING The following safety vslves were subjected to testing and found to be acceptable:
SV-215A Low Pressure Surge Tank Safety SV-409A Main Steam Safety SV-409B Main Steam Safety SV-409C Main Ste=m Safety SV-409D Main Steam Safety SV-409E Main Steam Safety SV-409F Main Steam Safety SV-409C Main Steam Safety SV-409H Main Steam Safety SV-409I Hain Steam Safety SV-409L Main Steam Safety PRSV-181 Pressure Code Safety Relief Valve e
l PRSV-182 Pressure Code Safety Relief Valve replaced this refueling.
During plant operation, valve setpoint was found to be out of tolerance. See LER 84-11.
Page 26 of 27
6.0 CONCLUSION
S With the exception of several hydrostatic pressure tests (which will be completed within this second interval), the examinations performed during this outage complete the inservice inspection requirements of the Yankee Nuclear
' Station Technical Specifications for the third period of the.second interval.
Page ?? of ??
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ATTACINENT A i
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ATTACHMENT A Procedure No. ,7 4/! a </c'/
Subject:
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EVALUATI0t! SHEET FOR !!ICORDABLE INDICATI0t!S Supplement C -
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A. Zone number /2)!N Evaluation Proc. No. k'dA8$3 I s /
B. Raster No. [E /[O7 / /9 Weld number [d OV -(( ' /
C. Indication number R 2/3
/
D. Applicable ASME Code Standards used for evaluation:
Sectionf Sjir7/71[t? /97V Article lll'0* ?(W)
E. Size'of indication:
Length SE6 N7a/ Depth /, y' t 7~'; C/.LSA I /3/,t [
~
Width 568 /3Fl49 Plate thickness (c) 7 75" W 1
, F. Characterization of Flaw Indication per Fara. / 41/i-332r)
- 1. Type of Flaw: SM<,, f/'M [M/',88
- 2. Sketch of indication:
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- b. y 2 t. c 5~ Ms0 : /.?,,50
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- 3. Flaw characteristic calculations:
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f* 4 * ' ll y - If.3l^ s
-( ; ).d.M ,Y 3.9 "
G. Comparison of pertinenhs' valuation standard (Para. /h/. W/> / )
to actual flaw size. g g ,g , ,.y Qi f f
- s- /
Acci.ptable Reportable Sf6 2/ /Fk/ O 5#4T'
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i ATTACEMENT B
c WESTINGHOUSL PROPRIETARY CLASS 2 ABSTRACT Finite element stress analyses for the inlet and outlet nozzles of the Yankee Pressurized Water Reactor (PWR) vessel subjected to postulated Level A, Level
B, and most limiting Level C and D Transients were carried out at Westinghouse for the Yankee Atomic Electric Company. Complete data and calculation results are stored on microfiche.
Two-dimensional finite element models were constructed for the inlet and outlet nozzles, and the stresses due to the pressure and temperature loadings were computed for all transients. This report presents detailed stress results for critical regions in the two nozzles for selected transients.
During the inservice inspection of April-May 1984, an indication was discovered in the outlet nozzle of loop 1 which required further evaluation under the provisions of the ASME Code,Section XI. The stress analysis results which were calculated for the outlet nozzle were then used to perform a fracture evaluation to demonstrate the acceptability of the indication without repair.
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WES11NGH0VSE PROPRIETARY CLASS 2 '
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FRACluRE EVALUATION OF DE1EC1ED INDICATION 7-1 ININ000C110N A fracture analysis per Section X1 of the ASML Code has been performed to investigate the acceptability of the indication found in the outlet nozzle of l loop 1 during the recent inservice inspection.
I The indication is shown in Figure 7-1, and was found at the boundary between the nozzle forging material and the weld by which it is attached to the l
vessel. The indication was characterized as an embedded ellipse oriented circumferentially with its plane being directed nearly radially f rom the centerline of the nozzle, as shown in the figure. The indication was centered at an angle of 232' from top dead center with respect to the centerline of the nozzle.[8]. To evaluate the acceptability of the indication, it must be I assumed to be a flaw.
The evaluation consists of three parts:
1 (1) Crack Growth Analysis - the final crack width of the initial flaw is to be l
determined as the basis for the calculation of maximum stress intensities K for normal / upset / test conditions.
g (2) Evaluation of the critical depth f or level A and B conditions.
(3) Evaluation of the critical flaw depth for level C and D conditions.
The acceptability is to be evaluated by the criteria of applied stress intensity factor of the Code Section XI and its appendices, as described l in Section 7-2.
i l
7.2 CODE ACCEPTANCE CRITERIA There are two sets of flaw acceptance criteria for continued service without j Namely, j .epair in paragraph IWB-3600 of ASML Code Section XI.
7-1 >
5927Q:10/061884
WESilNGHOUSL PROPRILIARY CLASS 2
- 1. Arceptar.ce Criteria Based on F law Size (IWB-3611)
- 2. Acceptance Criteria Based on Stress Intensity Factor (IWB-3612)
Both criteria are comparable in accuracy for surface flaws in thick sections.
The acceptance criteria (2) have been assessed by past experience to be less restrictive for thin sections, as well as for evaluation of embedded flaws in either thick or thin sections. For completeness, however, both criteria will be given.
7.2.1 CRITLRIA BASED ON FLAW SIZE The code acceptance criteria stated in IWB-3611 of Section XI are:
a For Level A and B Conditions f < .1 a and e
a f < .5 ag For level C and D Conditions where l
l ag - The maximum size to which the detected flaw is calculated to grow at the en( of design life, or until the next inspection time.
a The minimum critical flaw size under level A and B conditions (upset and test condi'.lons inclusive) ag - The minimum critical flaw size for initiation of nonarresting growth under level C and 0 conditions To determine whether a flaw is acceptable for continued service without repair, both criteria must be met simultaneously.
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59270:10/061884 7-2
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'9F- f WESilNGHOU5L PROPRillARY CLASS 2 7.2.2 CRiiLRIA BASLO ON SlRLSS INIEN511Y FAC10R h
As mentioned in the proceeding paragraphs, the criteria used fflaws or t ein evaluation of embedded flaws in thin or thick sections, and surface Namely, thin sections are f rom IWB-3612 Section XI.
K K < 18 For level A and B conditions I
.[UE K
K < 1E For level C and 0 conditions 1 jy*
where
=
The maximum applied stress intensity f actor f or the flaw size K
g a
to which a detected flaw will grow, during the conditions f
under consideration, to the next inspection.
'l
=
Fracture toughness based on crack arrest for the corresponding K
g crack tip temperature.
=
Fracture toughness based on f racture initiation f or the K
g corresponding track tip temperature.
7.3 FATIGUE CRACK GROWlH ANALY$15 7.3.1 EMBEDDED FLAW VS. SURFACL FLAW f
Since the indication discovered by inspection is near the inside surf ace o the nozzle wall, prior to crack growth analysis being perf ormed, the first item that has to be determined is whether the flaw should be consi embedded as it apparently is, or, be considered as a surface flaw.
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WE511NGHOUSE PROPRillARY CLASS 2 i h r? .imensions and the location of the f law are shown in F igure 7 -1
- Tha rule for treating the flaw is found in Section XI, and is:
< 1 - (2) ( )
4 Where the nomenclature is defined in Figure 7-1.
Dimensions as measured:
t = 8.06" S = 1.48" e = 2.13 a = .42" t = 3.9" (refer to Figure 7-1 for definitions) then, f=0.528 whereas, 1 - (2) ( ) = 0.792
)
, , f < 1 - (2) (
therefore, the flaw can be considered as embedded.
7.3.2 OUILL) N077LE TRANSIENTS AND STRESSES The transients used in the f atigue crack growth evaluation are the Level A and B transients developed in Section 2 of this report. The number of cycles f or each of the transients has been provided earlier in Table 2-1 for one year of service, and these occurrances were used directly in the analysis.
The stresses used in the analysis were obtained from the detailed finite element results for the outlet nozzle as presented in Section 6 of this report. The stress values used as input to the analysis are provided in Table 7-1. '
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WLSilNGHOUst PROPRIEIARY CLASS 2 7.3.3 FATIGUE CRACK GROWTH ANALYSIS in applying code acceptance criteria as introduced in Section 1.2, the final flaw size af used in criteria (2) is defined as the flaw size to which the detected flaw is calculated to grow at the end of 40 years design life, or until the next inspection time. In this work, periods in increments of ten year s were used f or reporting crack growth results.
The analysis procedure involves postulating an initial flaw at the location of interest and predicting the growth of that flaw due to an imposed series of loading transients. The input required for a f atigue crack growth analysis is which basically the information necessary to calculate the parameter AK depends on crack and structure geometry and the range of applied stresses in the area where the crack exists. Once AKg is calculated, the growth due to that particular stress cycle can be calculated by equations given in figure 7-2. This increment of growth is then added to the original crack size, and The procedure is continued in the analysis proceeds to the next transient.
this manner until all the transients known to occur in the period of evaluation have been analyzed.
t The transients considered in the analysis are all the design. transients
! These contained in the vessel equipment specification as shown in Section 2.
l Faulted -
transients are spread equally over the design lifetime of tNe vessel.
conditions are not considered because their frequency of occurrence is too low to affect fatigue crack growth.
Crack growth calculations were carried out f or a range of flaw sizes, beginning with the detected indication size. For all cases the flaw was assumed to maintain a constant shape as it grew.
The stress intensity factor expression provided in Appendix A of Section XI The flaw shape was used directly, which requires linearizing the stresses.
was set as described in Figure 7-1 with the eccentricity set at 2.13, as described earlier. ., . , ,, . , .-... 3s
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59270:10/061884 I-5
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For section 3) f or the main steam line break, and are shown in F igure 7 -4.
1 l ref erence, the stress distribution f or the small break LOCA is shown in Figure t
t i
7-5.
i i
! 7.4.2 STRESS INTENSITY FACTOR CALCULATIONS One of the key elements of the critical flaw size calculations is the i
determination of the driving force or stress intensity f actor. This was done f using expressions available from the literature. In all cases the stress j
intensity factor for the critical flaw size calculations utilized a This f
representation of the actual stress profile rather than a linearization. I was necessary to provide the most accurate determination possible of the i
critical flaw size, and is particularly important f or consideration of Level C and D conditions, where the stress profile is generally nonlinear and often f The stress profile was represented by a cubic polynomial: ,
' very steep.
i i
2 3 a(x) = A0 + ^1 + ^2 ( } * ^3 ( }
l where x is the coordinate distance into the wall j
t = wall thickness j o stress perpendicular to the plane of the, crack !
f l
The stress intensity f actor calculation f or an embedded flaw was taken f rom i work by Shaw and Kobayashi (10) which is applicable to an embedded flaw in an infinite medium, subjected to an arbitrary stress profile. This expression l l
has been shown to be applicable to embedded flaws in a thick-walled pressure i
vessel in a recent paper by Lee and Bamf ord (11].
7.4.3 FRACTURE 10VGHNES$ ,
The other key element in the determination of critical flaw sizes is the fracture toughness of the material. The fracture toughness has been taken directly from the reference curves of Appendix A,Section XI. In the transition temperature region, these curves can be represented by the following equations: '
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59270:10/070384 L . s . . .' . J . i .
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WESilNGHOUSL PROPRIETARY CLASS 2 The track growth rate curves used in the analysis were taLEn directly from Appendix A of Section XI of the ASML Code. The air environment curve was used f or the embedded flaws analyzed, and is shown in F igure ? -2.
= (0.0267 x 10-3) g 3.726 where, = Crack growth rate, micro-inches / cycle K; = stress intensity factor range, ksi b iman Imin I 7.3.4 FAi!GUE CRACA GROWlH RESUL15 The analysis of all the embedded flaws showed negligible crack growth, as shown in Table 7-2. This result is not unexpected, since the crack growth rate for air environments is relatively low.
7.4 CRITICAL FLAW Sill CALCULA110NS 7.4.1 SELECTION OF KEY TRANSl[NTS The key parameters used in the evaluation of any indications discovered during-inservice inspection are the critical flaw parameters. These are as follows; e (a ) for the governing Level A and B conditions e (a g) for the governing Level C and 0 conditions.
The selection of the governing transient for Level A and 8 conditions can be done easily based on the results of the f atigue crack growth analyses, where all the transients were considered, and stress intensity f actors calculated for each one. The governing transient was found to be the reactor trip (with loss of MCP) and the stress distribution in the section of interest is provided in figure 73 for the Level C and 0 :oncitions the two most limiting transients are the small break LOCA snd tne main steam line break for the outlet nottle region, the highest str esses were f ound (in c ro s
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WLSilNGHOUSL PROPRillARY CLASS 2 K 33.2 + 2.806 exp. (0.02 (1-RI
- g NDI Kg , - 26.8 + 1.233 esp. [0.0145 (T-RI ND1 + 60*F))
~
where K g and K g areinksifin.
The upper shalf temperature regime requires utilization of a shelf toughness which is not specified in the ASME Code. A value of 200 ksi in has been used here. This value is consistent with general practice in such evaluations, as shown for example in reference {12) which provides the background and technical basis of Appendix A of Section XI.
lhetoughnessusedforboththegoverningconditionswas200ksi[in,because the temperature throughout the wall thickness remains above 400*F at all times '
for both. The temperature distributions at the governing time step for each transient are shown in Figures 7-6 and 7-7, f or the Reactor Trip and Main Steam Line Break respectively. The RI 'UI ' 8" U"' " " ##'"
ND1 was estimated at 60*F, but does not play a role, because of the high metal temperature.
7-5 RESULTS OF FRACTURL LVALUA110N The f racture and f atigue evaluation of the indication f ound in the Yankee Rowe Outlet Nozzle of Loop 1 showed that the indication meets the acceptance criteria of IWB 3600 of the ASHL Code Section Al, therefore, no repair will be necessary.
The criteria are summarized below, along with the results of the fracture evalaation, the first step in the evaluation is the calculation of f atigue crack growth according to the methods suggested in Section XI. 1he ref erenc e crack growth law for air environments was used, and the resulting crack growth was negligible f or the entire remaining lif e of the reactor vessel, as shown in Table 7-2. Therefore, the indication as detected was used in the evaluation. The configuration of the indication is shown in Figure 7-1.
59210: 10/061804 1H
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WL511NGHOUSL PROPRILIARY CLAS$ 2 i
For level A and B conditions the allowable stress intensity f attor is:
l i l I
K Ia 200 63.2 hst (In f
K < !
I
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J As seen in Figure I-8, the max ito'Jm appliedg K for all the Level A and B transients is 5.2 ksi b found in the reactor trip transients, which is far 4 below the allowable.
I f
i For level C and D conditions the allowable stress intensity f actor is:
I i
K It 200 141 ksi 5 K
I I<@ h The maximum applied stress intensity factor for the governing Level C and 0 l
condition (the main steam line break), is shown in Figure 7-9, and is found to be 4.8 kst fin, which is well below the allowable as well.
Therefore the indication easily meets both the acceptance standards of Section ,
XI, IWB 3600, for acceptance without repair, for the entire remaining life of j the plant. f t
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59270: 10/061084 1-9 i
. . ._ . - _ . . - _ . - _ - - . . . . _ ~ _ _ - - - - _ _ - - . - _ - - - . . - - . __
.l IAut t 7-1 l
l STRESS RLSULIS U5Lii rdR FATIGUL CRACL GROWTH ANALY$lh i I i o i LOAD S1EP TRANSIENT ,
(Inside Stress) (Outside Str e max / min l
- 1. Reactor Trip with Loss of MCP max 19.544 7.249 l min 0.0 0.0 f f 3.742 0.6 l 2. Heatup and Cooldown 50'F/Hr max !
min 0.0 2.320 l
- 3. Heatup and Cooldown 100f /Hr max 6.435 0.0 }'
min 0.0 2. 3 M
- 4. Unit Loading and Unloading (0-50 ) max 0.0 2.845 l min 0.0 2 625 Unit Loading and Unloading (50-1001.) max 0.0 J.376 (
5.
min 0.0 2.845 ;
t
- 6. Step Load Decrease max 0.0 3.514 l min 0.0 3. R4 t 0.0 3.347 i
- 7. Steady State Fluctuations Temp. max Period Used min 0.0 3.28C l
Steady State Fluctuations Pressure nax 0.0 3.345
- 8. l min 0.0 3.279 ;
Period Used
- 9. Feedwater Cycling max 0.0 2.Eg[ ,
min 0.0 2.2ci l max 0.0 3.999
- 10. Loss of Load 3.02c min 0.0 ,
max 0.0 4.137
- 11. Loss of Power 2.715 min 0.0
- 13. MCP Startup & Shutdown Pump Westart max 0.0 2.269 i min 0.0 2.160 l Conditions
- 14. MCP Startup & Shutdown Hot Plant max 0.0 2.J47 ;
min 0.0 J.:74 i Conditions
- 15. Reactor Trip f rom Full Power (Nomal) max 1.343 3.70 !
min 0.0 3.477 l max 4.968 3. 51 :-
- 16. Reactor Trip 3.rP min 0.0 max 2.496 4.59 l
- 17. Inadvertant Steamline 0.0 1.;10 NRV Closure (Affected Lnor) min l man 3.293 4.91 ? (
- 18. Inadvertent Steamline *Wd !
NRV Closure (Other i oop'.) min 0. (1 .
l 9.776 1./49 l
- 19. Excessive feedwater flow (Variation 1) max 0.0 1 ~m min l 11.039 t 1 !
- 20. Execssive feedwater ' low (variati m 2) max ,,, i n o.n i..,. 1
'"d " UD '
- 21. Primary
- Side
- I eab * 'a'p . .
0.0 ' , '
.. r - .
, , min l r . ./ . .
7-10
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i' 1
T AHl.1 7.P l RESULTS Of FAT!GUE CRACK GROWTH A.*JAL YSIS -
i 1 (note crack length "a" is half of embedded crack width i l I
INITIAL CRACA CRACK LENGTH AFTER YEAR l LENGTH 10 20 30 40 420* 42008 .42016 .42024 42032 l
l l .500 .50011 .50023 .50034 .50045 1.000 1.00051 1.00103 1.00:54 1.00205 ,
- Actual detected size I
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STDESS INf tNSITY FAC708 #ANGC, Au (uSI T I g
Figure 7.2 udder Bound Fatigue Crack Growth Data for Reactor Vessel Steels l
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Figure 7.5 Outlet Nozzle Stresses at Section 3 at 50 minutes into Small LOCA 20 -
10 m
N Y
' 0 l I I i
= l '
$ 1 2 3 4 0; .
7 j 9 g am o
. . }
- ~
10 -
l DISTANCE THROUGH THE WALL: X (INCHES)
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l DISTANCE (INCHL5) rigurc 1,6 Teriperature Olstrit>ution at section 3 at 3.33 Hinutes r into Reactor Trip with loss of MCP - Outlet Notile i
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FIGUPC 7.7 Tenter a tur e Ot t,t' It.u tion a t 'we t ion 3 at
, t) . l fi7 ni nu t e ?. i n ta Mi t n $ teami t ne lis i..th
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.. ANGULAR POSITION ON CRACK O. DEGREES }
i x-Figure 7-8 fpplied Stress :ntensity Factor around the Periphery of the Indication '#
- _ - Governing tesel A and B Condition (Reactor Trip) c.. I s !
s e
e 6
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Wi Si l NGHOU5l l'Rf D'R il. l AR Y LIASS 2
- I i ;
+ SEC110N 8 l
l
SUMMARY
AND CONCItlSIONS 2
I Applicable transients were developed for the thermal and stress evaluation of l 3
i the inlet and outlet nozzles of the Yankee Pressurized Water Reactor Vessel.
Using .ill the pre',sure .ind temperature data f rom these developed transient s and finite element modeis developed for each nozzie, the stresses at is4 time steps were calculated for each finite element run for Levels A and B transients. Thirty two time steps were calculated for level C and 0 f conditions. .
i These stress analysis results were used to perform fracture and fatigue i
analyses to demonstrate the acceptability of the indication discovered in the !
outlet nozzle of loop 1 during the inservice inspection of April 1984. l l
I The maximum applied stress intensity factors for all transients analyzed are l well below the allowable stress intensity factors. The indication easily (
meets both the requirement s of sect ion XI, lWB 3600, f or acceptance without l e ep.i i e , Ior ihe crit i t .. r enu in inq IiIe is t I he p l.ir i I
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- t. . . .. : .
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N'%.J,.;, 3 - '
I 5927Q: 10/061884 H1 ,
l 7t [ i'/ " U COMMITMENT-TRACKING SYSTEM' PROCTR Applicability Y N Commitment Responsibility:
NSD Plant O sotn Outgoing Letters:
Responsibility:
Due Date:
NSD Service Request Required Y hN Distribution for 84-96 J. E. Tribble D. E. Vandenburgh L. H. Heider D. Hunter J. D. Haseltine J. A. Kay R. L. Berry J. K. Thayer W. G. Jones A. H. Shepard - 2 N. St. Laurent - YR John Ritsher, Esq. - Ropes & Gray J. G. Robinson R. L. Boutwell John T. Taylor /W. K. Peterson Resident Inspector (Rowe)
W. L. Whipple ,
J. Lance T. Rennell N. Fetherston (Rowe)
Licensing Copy File
SUPPLEMEh"T I ./
IMS DOCUMENT INPUT FORM - L //
REQUIRED INFORMATION
- 2. PLANT iR 3. CI.ASSIFICATION TYPE L
- 4. RECORD TYPE NO. G ot ( 6,2 M/
- 5. IMS SURJECT NO. O)d 3 dI (f f
- 6. DATE WRITTEN [M - /Y
/
- 7. DOCUMENT FORM
- 8. DOCUMENT LOCATION SUPPLEMENTAL INFORMATION
- 9. PRIMARY DOCUMENT NO. F1R P4- %
- 11. TITLE 9////6/ [AM//N/ b M 1/ / d //f ) M
- 12. KEYWORDS [b If P J/
- 13. ORICINATOR
- 14. RECEIVER
- 17. REFERENCE DOCUMENT I.
I i
- 21. WORK ORDER NO.
ACTION
- ADD / CHANCE / DELETE (CIRCLE ONE)
- 1. ACCESSION NUMBER /7/ /,d 2 1
'
- If ADD, provide all document information and circle ADD.
If CHANCE, provide document Accession Number, corrected document characteristic inf ormation and circle CHANCE. If deleting all document l
information for a particular characteristic, write DELETE in appropriate i entry.
If DELETE, provide document Accession Number and circle DELETE.
I
. -l; ' * " ' l t. . ,,
, " *, ,,! .' i . . ..],4 , . . , .9 ,, ; g, , ,;l,p ;y - .; 'j,y ;p . . , , .
ATTACIB!ENT C NOTE: This section added-Revision 1.
EVALUATION OF INDICATION REPORTED IN YANKEE ROWE j N0ZZLE NUMBER RPV-FF-1 t
Prepared by:
//
George Martens Technical Consultant i
l
~- , , . . , .
' l. SCOPE 1.1 This document summarizes a study to determine the size of the flaw that produced reportable indication #R2/CH.3 in nozzle number RPV-FF-1 at the Yankee Rowe Nuclear Power Station.
1.2 The determination of flaw size is based on comparison of data obtained from the nozzle with data from a special test block.
- 2. BACKGROUND 2.1 At examination time, NES advised Yankee Atomic that the indication was sized as reportable pursuant to ASME Code / Reg. Guide 1.150 criteria, but that the actual size of the flaw was probably within acceptable limits. NES reasoned that the conservative indication sizing criteria applied to a longitudinal wave examination more than doubles the actual flaw size in the specific location and orientation of detection.
2.2 NES subsequently demonstrated the oversizing parameter on an NES test block and estimated the actual flaw size to have a 2a dimension of 0.31",'resulting in an allowable a/t of 1.59%.
2.3 Yankee Atomic had SWRI fabricate a special test block to the original Yankee Rowe specification to assure accurate data. The SWRI test block certifications are in Addendum B. This block was fabricated with 0.200" wide and 0.400" wide slots to bracket the NES estimate of flaw size. This report covers the analysis of the flaw size based on data from this block.
- 3. EXAMINATION OF THE SPECIAL TEST BLOCK 3.1 A dimensional inspection of the special test block YR-D-00-056 (SWRI Drawing No. D-8328-600) was performed. All important dimensions are within specified limits except some from the clad surface. This is realistic and totally acceptable. The data is in Addendum C.
3.2 The equipment used for this examination is the sane os that used at Yankee Rowe except for the coaxial cables and examination head (search unit) which are in hot storage. Identical cable type and lengths were used. The examination head used is the identical spare that was taken to Yankee Rowe but was not put in service.
3.3 The calibration was in total conformance to the procedure used at Yankee Rowe (NES Document No. 83A0314, Rev. 1), with two exceptions:
-the sensitivity was increased 2 dB to adjust for the measured sensitivity loss between curved and flat entry
-the sweep was adjusted to expand the time of interest for more accurate measurement Examination documentation is in Addendum D. ,
- =%
3.4 The block was scanned in the same orientation and with the same index (.2") as the nozzles at Yankee Rowe. Examination data is in Addendum E.
- 4. DATA ANALYSIS 4.1 The composite data sheet is attached. The plots depict apparent depth in the material from left to right. This represents Code /
Reg. Guide indication sizing of the 2a dimension.
4.2 The flaw was initially considered to be at 20' because it was located at a 20* fusion line, the amplitude was into saturation, and it was not detected with any other angles. The test block data substantiates this, however, it shows that the flaw is not located as deep as measurement indicated. The data clearly shows that the angle of the reflector strongly influences the apparent depth by changing the effective detection angle. This effect is greater with the longitudinal beam because of its wider beam divergence. The peak amplitude points are at 2.02" for the 0.2" 30* reflector, 2.52" for the 0.2" 20' reflector, and 2.82" for the 0.4" 20* reflector. It is understandable that the larger reflector at the same angle has a stronger peak effect because the reflected energy has less divergence.
4.3 No attempt is made to establish correction for the l dimension because of the beam divergence difference between flat and curved entry in this orientation. NES estimates a possible oversize in this dimension of up to 0.75", however, this would not substantially effect the aspect ratio.
- 5. FLAW SIZE 5.1 The 20* slots, 0.2" and 0.4" wide, bracket the data from the flaw.
The 0.2" slot was oversized by 0.58", and the 0.4" slot was oversized by 0.54". The average oversize dimension of 0.56" cubtracted from the 0.84" flaw data results in a 2a dimension of 0.28". This results in an a/t of 1.4%. The decrease in depth discussed in paragraph 4.2 changes the center of the flaw depth from 1.9" to 1.35". The estimated decrease in oversize dimension due to decreased effective beam width is 15%. This results in a 2a dimension of 0.36" and an a/t of 1.8%. This is the estimated flaw size. The tolerances in measurement of ! .1" result in 1.3% to 2.4% outside limits.
- 6. CONCLUSION 6.1 All data clearly indicates that the flaw is within the allowable limit of 2.4% (IWB 3512.1). The flaw had to be at least 20* to produce the peak amplitude in excess of 150% DAC. If the flaw is over 20*, the "oversizing" would have been even greater.
_ _ _. . - - -. _ . .-. _ . _ . . _ _ . _ .__ . _ _- _ me.. .. _ . - .
4 g %
1 I The possibility of'this. flaw being " service induced"'.is highly.
. 6.2
- remote. The' shape, orientation, location, and size indicate that
- the-flaw is lack of. fusion to the parent material at part of one weld. pass.
Y Y.
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en-#-~-~~~e- 't-n<' V=~c*- 4
COMPOSITE DATA PERPENDICULAR PLOT KNOWN AMOUNT g TARGET OF APPARENT 2a DIMENSION 2a OVERSIZED A '
I a .
f Cos. 20' O.77 Special Test Block 10 0 - .
x -0.19 .
20*, .2" Wide 0.2 = q 50 - o.77-
= 0.58 7 20 - .-
,o Depth to the center Cos. 20' O.92 f all slots is 1.9" soo . __ . P _._
20*, .4" Wide x -0.38 0.4 =
yo . - -
o,9f .. ___
= . 0.54
.38 20 . . ..-
s o o ___ .P -.- -
Cos. 30* 1.02 30*, .2" Wide x -0.17 0.2 -
50 ',F/h 0 2. - - . - - - = 0.85
.17 20 - y - - - l i_
Cos. 10* (50%
' x w uld have 10*, .2" Wide 20
- 0.2 been sized
- 'l l ,
= differently g
.20 Apparent
- f.6 1.0 I.y f,( y,2 , n
- Depth in
- Estimated Inch 88 e.+'
s.s f 2.1 2,6 3.0 7.t .
, , 2a ,a/1 a/t Allowable I
[
Indication in 8y 7, {, e,,, N.A. 20* avg. 0.84 .18/ .18/ with an Nozzle i,
~ ' re g. - x .85 -0.48 3.9 9.75 aspect o
T12' . , l 50
, depth = = = ratio of
' I j correction 0.36 .046 1.8% 0.05 h I g [ 0.48 2.4%
.i ADDENDA A. YANKEE R0WE 1SDICATION DATA B. SPECIAL TEST 3 LOCK DATA FROM SWRI C. SPECIAL TEST BLOCK DIMENSIONAL TEST DATA D. DOCUMENTATION ON EXAMINATION OF SPECIAL TEST BLOCK i.
E. DATA FROM SPECIAL TEST BLOCK t
a i
F.
I I
b -- - u ,.d_ ..--.e a __
A ADDENDUM A YANKEE ROWE INDICATION DATA
a O Y A I T ,1 I
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Procedure No. , $3f.41/4/
Subject:
, /l[/ f,./73/gj Page /,4 of_ n.s3 Plant /Unic Y/4.E"/f4 47/8 Cal. Data Pkg. F R/4 /- /
EVALUATION SHEET FOR RECORDABLE INDICATIONS Sucolement C A. Zone number [_I/h Evaluation Proc. No. b A ST3 I
/
B. RasterNo.kE/[O7/A Weld number /[OV-8[-/
C. Indication number R 2./3 D. Applicable ASME Code Standards used for evaluation:
' Section[ $_'/r'/T/J. /f/'/i' Article /[(.'b' S/ W )
E. Size of indication: ,
Length i f$ /L'>'r.'^~ Depth [. L/ P G Cl.'3.RI /'*/6 I l
) Width 6E6 # Ele Place thickness (t) 9 7s~"/#
F. Characterization of Flaw Indication per Para. / l/ > e ? -33 'd/'
- 1. Type of Flaw: $ >i s <. zrsd"[ P/_de_8/?
t
- 2. Sketch of indication:
Cto2c Di4T' cf CE x ri y GOS/ZnL' MQ' /41.S( L M & 2 7 Coif 05lV7+ Li Cf X 0X G?LS!Z?? M illt A*2Y7.5/ ~~ l.50 '
- 6. v 3 /s c .
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.s c Acceptable Reportable SCE, /.17d/df/) 5447'
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Approval / /M I
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NUCLEAR ENEMGY SEMwCES. WC. -
FORM 028-120/1 8/83
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I ADDENDUM B SPECIAL TEST BLOCK DATA FROM SWRI
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Yankee Rows Nuclear Power Station Special Test Block Certifications ,
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Purchase Order No. 105016 Dated February 12, 1985 ,'
Identification No. YR-D-00-056 1 (RE-a SOUTHWEST RESEARCH INSTITUTE :
SAN ANTONIO '+ . 1 HOUSTON
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> SOUTHWEST RESEARCH INSTITUTE P--NUMBER CLASSIFICATION FOR CALIBRA TION BLOCKS Y~~ ~
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CALISRATION SLOCX 3 GROUP 3 IN ACCORDANCE WITH CLA S SIFIED AS P- hum 8ER SECTION II 1980 EclfloN OF THE ASME SolLER AND PRES 3URE VESSEL CCOE. THE P - NUM SER CL AS SIFICATION FOR THIS CALIBR ATIO N SLOCX 15 $USSTANTI ATED- WITH THE ATTACHED CH EMICAL AN ALYSIS REPORT ' FOR SA508, CL2, HT Q2Q 106 NQT IN ACCoROANCE wiTH THE M ATERI ALS SP.ECIFICATION SECTION :: OF THE ASME 80lLER ANO PRESSURE CODE.
DESIGN CRITE RI A" This special test block was designed by Yankee Atomic Electric Company. This is a verification that the block was fabricated in accordance with Yankee Atomic Electric Company, Purchase Order No. 105016 requirements and to the requirements specified on Southwest Research Institute Drawing No. D-8328-600.
- Due to the availability of the ultrasonic procedure Southwest Research Institute was unable to perform the final ultrasonic calibration on the machined reflectors. Final ultrasonic acceptance to be performed by Yankee Atomic Electric Company.
ATTACHMENTS MILL TEST REPORT / CHEMICAL ANALYSIS REPORT 79 ELIMINt 9Y UT QATA SHEETS ( Sw Ril CIMENSION AL QATA SHEETS ( M ACHIN E SHOP =QC)
FitfAL UT ACCEPTANCE DATA SHEETS (Sw all
- N/A QRAWING (Sw Ril D-8328-600 Welding Electrode Certifications Postweld Heat Treat Certification Project Manaaer March 5. 1985 Robert3 9." 9 0ards
Lenape Forge %mi m. ism con w ra u.aut.crunas como.ny T*";5,'g3pr e o sox sse t .co r irisi nwm en 2s4 west catsren. PENNsytVANIA 19340 C1351-03 OATE 8/23/03 MATERIAL TEST REPORT S.O.NO I DISTRIBUTOR PURCHASER 11200 OfSTRIBUTOR'S CROER NO.
PURCHASEKS ORDER NO.
SPEC HEAT NO. ' M/O No. BHN I M OTY, PRODUCT Rgh. Frg. Blk to Fin. 8 X 14 ' 16 A508-74A Q20106NOT 133J k1 2 CL. 2 Rgh. Frg. Blk to Fin. 8 X 14 X 36 020106NOT 134J
}2 1 CHEMICAL ANALYSIS ANO MECHANICAL PROPERTIES HEAT NO. C JAN ,
P S $1 ,, CR ,NI ,
MO 7 _/,( W HEAT TREATMENT V V V' V V / /
V / / NORM, CUENCH & TEMPER O20106NOT .23 .82 .008 .007 .27 .33 .72 .59 .04 LADLE StrRI P. 0, /// 7 I P. R. I79N9 L::o 1RT 7
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YtELD RON A . SHN 40'F % SHEAR HEAT NO. TENSILE vg- LAT FXP 020106NQT Long'l O' 85,000 63,500 26 68 144-128-110 .085 .094 .086 70-70-60 180* 93,500 72,000 23 68 150-1.39-198 .091 .090 .103 70-60-100 i
we moreoy certify the neove results to De correct
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i t MATERIAL 1EST REPORY w.iding Producta Divi..on "vPTEI . .... EDYNE McKAY
. . .. . ~ . _ .. i,..,
vw c.a., No 1936 u.,6.a ro, . CSS-309 11/79 WCLDERS SUPPLY C0 our o,o., No. G7680 SAN ANTONIO TX 78238 o i. sn.po.o 04/27/83 u.i. ..i o..c,.,u . s,.citic.n.n. Coamg Ser. in Ci.ss t at in E we.gne so E309-15 DC Lime C 60 5/32 E309-16 AC-0C E309-16 DCT 3 b 6 So.celec. teen i Hea No LotNo AWS AS.4-78; ASME SFAS.4 i
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- n. ,a inns. . os. v..io. as. ' s tiong Charpy V-Notch Imoact, ft.lb.
statement: i ( The 5/32" DCT electrode is not ( 2 88,000 67,000 37 recommended for use in the ( 3 vertical and overhead positions, , i s 6 we n.<.or c.nier mai in. nove p.oouci n.. e n po, noi.,,,,,,on - ei.. i a.n.cco, nc. ..in. .o c.c.i..
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g CSS-308 g MATERIAL TEST REP @R Welding Products Division NTELEDYNE McKAY PO sea 9609 4 0Gancie, A.ae. ve a. Pe 17406 CSS-308 11/79 Toid Of0w No gg7g ue w v WCLDERS SUPPLY CO SAN ANTONIO TX. %, o,,,, n, sn.opea 09/28/83 u...n. o. n.e.a .ao sp. cmc.a.a.
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Welding Research & Development Quality Assurance Systems and Engineering Division , Southwest Research Institute pA0aEcT NO. /7 f317-2% onTE 2.//7/FE
,PART NO./ID NO. d d.l. O Io c k [a.t Ice. /hfe mtd.
DRAWING NO. p- P3 U- 6 do i STRESS-RELIEF TEMPERATURE *F TIAEATTEMPERATURE*
//56*F 2 hours
- *F/PER HOUR SPrerAL'INSTRUCn0NS ,
~
R.T_U,- T. 1 Al e r fr/A - 7J Pdd*F- f 7,5"//u_ co0L-00wM RATE , REMARIG M#M. > m PERSON PERFORMING STRESS RELIEF:
/V!! L G A SIGNATURE 2/ir/rr ORTE e
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Date [NA8C/1 23# SvRI QA Signoff W Project Nucher/7dPJ1.7 .10 C Utility Aff eg/t. ,/A w 8e, e Plant QUALITY ASSURANCE REVIEW - PREOPERATIONAL ACTIVITY e PRIOR TO SHIPMENT OF TEST EQUIPMENT AND
# DEPARTURE OF PERSONNEL ! List Ultrasonic Calibration Blocks:
- Mill Insp.
Serial Number pf Test Rept. Dwg. Remarks YA-D-oo-os'C / / v V u 7 Da r^
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Yoke Calibration valid for 6 months
;l POA-1-C 12/79 Page [ of ]
e L-
r ADDENDUM C SPECIAL TEST BLOCK DIMENSIONAL TEST DATA
CALIBRATION BLOCK DDd.DiSIONAL INSPECTION p l g Project Name and ifo.: - Y%ket 6e 5 6to -ioo Celibration Block No.: Specut Tes t %ic.ek 90-t-oo-oS4 Au s Drawing No. and Rev.: "" W'*} t - R 12 A-- 6eo Dimensional Specification No. and Rev.: 'J /d 3 - 2, .e s-Inspected by: O P4 tevel It Dace RC, viewed by: Level Date Spool / Blank ( ) Prelim. Machising ( ) Tabricated Block ( / ) No. Drawing Dimension Limits Act. Dim. Remarks Ak l% CIA ok 12 %t.A Lemykk 6 Bl.c.k W;alk ih 6.03o ow 3.oM - 1. cia 3 Blode Thukw.s s If2 uk
- o. sis go 4 200 W;dLL .9 m.ht vi@ s\As view A.A I.coS 3, g , , ,, g ,
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CAI,IBRATICN BI.OCK DIMENSICNAI INSPECTION Ptg3 of 1 INSPECTICt1 RECORD 31Cck No: YR.s-ne -osc I[ DRAWING DIMENSION I.IMITS ACTUAI. DIM. REMARKS i....u.- .Nuu t- u..w exy uog Ik b tente,- (Nshk sl Leh Ude eS A A) 08 Ok 22 * ~t oo W.ALL e(m.bb 1 its.ct of g-a 1 O'F c isr p C . io r a. go e. k 50o WR ok hstek left St Ae eE 3. n. 1*C50 0.'493 ok 24 2.co o wth 6 eta sJ.u .C m.uwt t.Ctse t. . c ie i.99a e% I L'astk eE wstek -(. le{k OAe sQ s-a I 'IT O 91 *l en gg 3 Loc.b.n .T w.ttk ba =4*s*A. > bL 4a **.Aev- (a.b L c.t t eh v Ac ei s e) , Il g
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Cle.114 Tk;ckmess, N c ** . Ok 32 "33 34 35 MOTCl S ECTec4 A- A t% bnAwM 37 eNt04afe744 c N T! 4 t h b f1 A Q 8 4 O__. 38 _ 39' 40 4I 42 43 44 45
k ADDENDUM D DOCUMENTATION ON EXAMINATION OF SPECIAL TEST BLOCK
4 Raport No. ULTRASONIC SEARCH UNIT QUALIFICATION
! SEARCH UNIT DATA REAL TIME WAVEFORM i
Manufacturer: NES .5 uS /Div. .5 Volts /Div.
" 1 Serial No.: C- 4 gg., g . .
- . . ............1
. .....m.
Mfgr. Designation: 80D8120-1 spare , .. .c r.
~
- 1. . .
- .,4 UU c % c.:i.,, . . ' , :: L,;,Q[if.
l Nomini Angle: 45 * (Check One) -l l . -, .[ b'[ [ .' 3..[. f. h . ,, - [3 Long. m Shear N. -.
. .: , . . y. . ;. e , ['.
.<;, ' - ^
.' %. . 1.
Nominal Frequency: 2.25 MHz ,s.I . L . . ...cM i;. ; f: . - f r [ [ I . .'., l [. - m 7;,- Element Dimensions: 1x1 . ) h;(; [ '."1 ) 4 Element Material: LTZ-2 r. C< ~9 : - . ' ., .., -/ .(. 1, ,,:, .. -. ,.S _ j i h,.__., ~ l,
. w . ..
Tuning: (Check One) J7[%* E 'C l'. y' Y ~. ' 1" VY'. i*M l [3 Tuned O Untuned .[ y ] g [ ~ [ f Q ?l [ [ ! Intended Use: (Check One) i l OC'ontact GImmersion O Othet 1 f Connector: BNC attached SPECI' RUM ANALYSIS Peak Distance:
.. MAR rI.R;h 55 '00' .0MM- ).[.
~
b aEF -10.0.'h-i ..N (Chsek One) 4 2 3sa n 'e . M/o :aem . . O m?s, sa W y '-
~
=.;,.+3 4 Wh.f r.
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Flat Yn 96 '.,. ,\ < ' ; ". 4li . l, NA Peak (for.duall NA Lins F s NA Point Foett[ ~ NA ,
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' '5 '*4 x. l TEST DATA
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Test Block S/N: 4608 . ....e.,',.
.. a ,'
_~ Blo'ck Material: Carbon Steel . , .
. % / D .7 i .7 + '.
]N.) s b[' :Md[' I O 4* N'N k Block Dimension Used: 4" radius I .. .
Cable Type: Mini noise coax '
.I ,. Y s n. . . .
4.:. [ .. I , f . ?l f . - , Cable Length: 75' l ,,, c. -A !
.nshp aw, ; .M^ 3a g. svors oo0 c 2 000 0 wri s
? sr..it' '
- w si stc- ,. u ; l Pulser / Receiver l
Cains 40 Energy: 4 _ _ . . __. .. __ j i Receiver Attenus cion: 6 db pm:s , 50 ! Receiver Damping : . Ohms 3 db Limits: Lc<er 1 96 eH: Op >er 2.35 gg2 Center Freq.: 2.155 ha: , , 4 db Limits ver 1.88 MHz CERTIFICATION PERFORMED BL
~,
.I f RC
, s Ut?er 2 44 MHz DATE: 4/14/04 ACCEPTED REJ'ECTED sured Peak Freq.: 2.155 332 37: jfM/b UT LE7EL III idth: .56 MHz RECERNF7BEFORE: 4/14/85 j
!aa ed angla: 45
- in ligo EE "uc'8^^ **'ane sexvers. ec. -
n
- 020-137/2 S/83 REF: 60A5536 l 1
l
RF WAVEFORM RECORD SHEET /# j f bf / i I CHECK ONE: PRE-EXAM FIELD WAVEFOR!!, SITE: ,MM l _T/l/Fd POST-EXAM FIELD WAVEFORM, SITE: I/6;f ,ky_ /d/ u-G t l l 3 HlulW r Iledh, 4 i TRANSDUCER PULSER RECEIVER ULTRASONIC INSTRUIENT
)
I S/N: d-/ S/N: //fo7['9 S/N: /Oc2f/V~9 I MFGR.: 4)[d liODEL NO.: 81f MFGR.: AT MFGR. STYLE: _f" <f d A//9 DESIGNATION: PULSER RECEIVER SETTINGS 90-DAY LINEARIU ELHENT ATTEN. db: 2/ ,P DUE DATE: J- / 7 ~8f REF. (Fine Gain): A//9Y REFERENCE BLOCK I TY E h /7Y v S/N: ,Pd b N Ved ~/ CABLE - , MODE: #b TYPE: Oa /4/rr/ f/r'M LENGTH: 2/70 / 7 J ~ FREQ.: J.d[ 1fATERIAL: be vfm Me' d [ Q-* V REFLECTOR FIXTURE OR HEAD PULSE LENGTH: //74X USED: JACK (T or R): 7 S/N: f0 /d [' ~ ! COMMENTS: TEST (Thru or [ ANGLE: 4[ L No m U: Yd l [W /$D YM "] .
, ci+ " .2
; . l ;
VOLTS /DIV. / 97) #76/ g . f M - f - - - % --f - l : I SEC./DIV. r// ) :-~'!~~~! ~, i p: ; ; : i l .
- p. - - - { 4 - { - - - ,
l e l ! b - d l l I I I _ i i PERFORIED BYJ_ 4 eu ,,LEVET- // DATE:?/91(
\ ! REVIEWED BY:
U' , LEVEL: ' DATE: i
~
l FOR21028-122/ 7 11/84 REF: 60A5532
Y d? ? ULTRASONIC INSTRUMEhT LINEARITY RECORD ULTRASONIC INSTRUMENT CALIBRATION BLOCK MODEL NO. &KO SERIAL NO. /(17I/I/ ~9 TYPE /)// <1/ 4e nSERIAL NO. b bb C b TRANSDUCER BRAND M 64 ke r FREQUENCY M c Y SIZE / STRAIGHT BEAM ( BRAND FREQUENCY SIZE ANGLE BEAM ( ) 000'/3 9 VERTICAL LINEARITY SIGNAL AMPLITUDES IN % FSH HORIZONTAL LINEARITY
^ ^
(CALCULATE) BACK GRID ACCEPT REFLECTOR LOC. LIMITS SIGNAL 1/2 0F ACCEPT. SIGNAL NO. HIGHER LIMITS
- 1 1 1 1 //f/) ( fD) ( % ')-( 6 ) fD 2 d'C 1.90-2.10 2 I !b) IM ~IM 3 3D 2.85-3.15 3 h d- (3/) (dk)-(36) } /'
4 y. b 3.80-4.20 4 El.) I QI) (2 )-I3 b) F 5 T.D 4.75-5.25 5 Yh ( 20) I /b-I ) O 6 (, . D 5.70-6.30 6 2O ( /[) (/ 0 -(M) 1
/
7 '/. O 6.65-7.35 7 JV ( / .2) ( 7 )-(/ 7 ) /A 8 [,h 7.60-8.40 8 .9 O (/0 ) ( f)-(/F) /d 9 9.D 8.55-9.45 9 k# (M) (3 )-(/3) l 10 10 10 10 /.1 ( b) ( ! )-( //) I b
- ACCEPTANCE LIMITS ARE 1/2 OF THE HIGHER SIGNAL i 5% FSH l AMPLITUDE CONTROL LINEARITY
! INITIAL db AMPLITUDE CHANGE RESULT LIMIT 80% FSH D0tJN 6 127 - 4RY 80% FSH DOWN 12 0 16% - 24r 40% FSH UP 6 [d 64% - 96% i 20% FSH UP 12 hD 64% - 96% THIS INSTRUMENT IS CONSIDERED: ( ACCEPTABLE i ( ) NOT ACCEPTABLE SIGNEQ_ - m 1.EVEL ./2 DATE [/9- 7T NUCLEAR ENERGY SERVICES. INC.
,,L.
E s s % % % c. w N S t 8 4 6 6 N l o Y i 4 2 9 9 i. s O T m - - - - a P I i 2 6 4 4 c S h R L 3 1 6 6 i f E q A o R e i l, E N t m s Y d l_ I l T u L u l I R i t 5 6 7 8 L s e m e e n A E l i t O R R o e u g n N m w 0 T N t a a A N 2 2 . a I o 4 b 6 o D O 6 m P h L l 1 1 l C C d - - + + e N r d e R h . o l n T q 0 L . M a i l S i P t / i e e m e N i 8 M i 0 0 0 0 q r t m a v I l A n 8 8 4 2 f e a i x e 1 2 3 4 I M S D T E L F w o 6 6 2 O t s % 8 4 6 6 3 3 t 0 l h / / / Y T i i m 4 2 2 6 f 9 t 9 4 0 P L 3 1 6 6 A S h 0 0 Y 0 I R 8 - f E R l q
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bb E IucI 0#J NIM (Obl M/f[T CALIBRATION DATA SHEET M Cal. Blk. No. b N -l Page of Surface Temp. A/ A F Cal. Data Pkg. ///4 Block Temp. AN F Procedure No83OI'/Rev. RPP 1939 CFl *] 6 JWaSDMA So!L CdC-Instrument Hole Ident Depth In. Sweep Div. Amp % (FSH) Mfg.: /7 / Model: $ 80 f-/ ,$, ,) 7 /,I f-[) Serial No.:/Og////-9 Channel: [ f-Q 4/,j ') .7, '/ (/[2, Nom. Angle: f[) Mode: M jf/ f f.3 , ),
) 3,/ )[
Transducer O-Y /S,Md 4,3 /[ Mfg : ff[ $ Model:/y/f (,-( /I/, $l h.0 Y Serial No.: /[/jf Size: /X/ Calibration Date Time By Freq.: p, #) $' MHz Measured Angler [yd'/, Initial fj/(/[f jg h;{J3 Cable T" pew 6/7t//f, Length:3 00'+ 73 Intermediate h/9/f[/6 3.E hT Head No.:Q) f) P/f70./ Couplant: /[g g7 Intermediate 'y/[f/p[ fj f.1 blQ Main Sweep Range: '
,)d in. Intermediate [
Delay / Setup Range: Mf" sec. Intermediate [ , Main Sweep Calib.: 8, t/g Intermediate [ l Delay: C, () Syne Q/,# / Final fj/j/[( f[pC $M l Rep Rate Range /f M KH: Var:dgz, ggggt gcgg Display / Pulser Sync: /)LT loog Fitr: 7 Mode: [*h.) gog Q: // Freq.: h 80% 4 Jack: ~/~~ 70t Ref.: Min.: (ffC) 60% Pulse Length: Min.: M Primarv Reference Sensitivitv83/t.2 6(Kb 40% Equalized Response: [C (tf C 3O db db
\
30 % Exam Sensitivity: M ~7 c{ h . 3 - J t/ A/g db 20% ( ~ Screen Gated: /U /7 sd to sd Front Interface Signal Position: /l/// sd 10% 1 Maior Screen Division: A/ // in 0 1 '2 3 4' 5 6' 7 8 9 10 TUC Card No_ Pulses: ner
+ - Freq.: B C J nyjn Leye d Date !Nl Pulse Period: C F 2[ M/t Leve d Date Pulse Delay: F 3 Level Date Attenua C F Reviewer Date ance Added:
NUCLEAR ENERGY SERVICES. NC. - FORM 028-102/1 11/82
NOTE:- After calibration documented on calibration data sheet,-sweep was adjusted from long. side of cal ~ block no. 80D8842-1 on holes 0-3 and 0-4, for.more accurate readings, as follows: DEPTH IN MATERIAL - INCHES , 1
~0-3 Hole 0-4 Hole
~
2 2.2 2.4 2.6 2.8 3.0- 3.2 3.4 3.6 0 1 2 3 4 5 6 7 8 . 9 10 HORIZONTAL SCREEN DIVISIONS Sensitivity was increased 2dB to a setting , of 22dB to compensate for curved vs. flat entry, as measured by the difference in amplitude between Hole A and Hole 0-3 (from i the long side of block 80D8842-1). F (pgp f&'fd U Y v b g.-
ULTRASONIC INSTRUMEhT LINEARITY RECORD ULTRASONIC INSTRUMENT CALIBRATION BLOCK MODEL NO. 5~8 O SERIAL NO. / O244 -9 TYPE [mM SERIAL NO. ESC 4S3 TRANSDUCER BRAND h M - G or,,, . //A FREQUENCY.O.25 /Y//:t SIZE / M STRAICHT BEAM ( / ) BRAND N/A FREQUENCY w/a SIZE % ANGLE BEAM (
)
VERTICAL LINEARITY SIGNAL AMPLITUDES IN % FSH HORIZONTAL LINEARITY
^ ^
(CALCULATE) MR HIGHER BACK GRID ACCEPT SIGNAL 1/2 0F ACCEPT. SIGNAL REFLECTOR LOC. LIMITS HIGHER LIMITS
- NO.
l 1 1 1 1 100 (50) (4.5)-( s s) 49 2 .2 .0 1.90-2.10 7 % ( 5) H O )-Wo ) 47 3 80 (40) (M -MS) 40 3 7.0 2.85-3.15 4 4.o 3.80-4.20
' 70 (3F) (3 )-(4 ) 30 5 f.o 4.75'-5.25 5 6O (3o) (25)-(as ) 3o 6 (, . O 5.70-6.30 6 50 (25) ( zo)-(2o 1 26 7 7. 0 6.65-7.35 7 40 (20) ( /S )-(2.5 ) 20 30 Ib 8 f. O 7.60-8.40 8 (/5) (/0)-(20)
I 9 9. 0 8.55-9.45 9 20 ( /O ) ( S )-(/ S ) IO ( O 1-(/O) b 10 10 10 10 (S) l l
- ACCEPTANCE LIMITS ARE 1/2 0F THE HIGHER SIGNAL i 5% FSH AMPLITUDE CONTROL LINEARITY INITIAL db AMPLITUDE CHANGE RESULT LIMIT 80% FSH DOW 6 39 177 - 4RY 80% FSH DOW 12 2O 16T - 24%
40% FSH UP 6 N 64% - 96% 20% FSH UP 12 6O 64% - 96% THIS INSTRUMENT IS CONSIDERED: (!) ACCEPTABLE ( ) NOT ACCEPTABLE SIGNED d m D.b u k LEVEL I DATE / / Aewt 8.S _ _ NUCLEAR ENERGY SERVICES. INC. t
I i RF WAVEFORM RECORD SHEET l CHECK ONE: O PRE-EXAM FIELD WAVEFORM, SITE: l I l b POST-EXAM MEtft-WAVEFORM, SITE: NA t 3? I7 Aw es c i I TRANSDUCER PULSER RECEIVER REMOTE PULSER PREAMP: I S/N: C,- 4 S/N: / / S2 & - 9 l S/N: 183% C_h i f MFCR.: NES MODEL NO. : P E J.S PE/THRU: T%co l MFGR. DESIGNATION: Mone. PULSER RECEIVER SETTINGS ULTRASONIC INSTRUMENT ELEMENT *
- ATTEN. db: 46 S/N: 10244 CI DIMENSIONS: /8/ e/voj REF. (Fine Gain): Dox MFGR.: A 'r PE /2 Cr / 74/U FILTER: 2 STYLE: .f E O CABLE MODE: FW 90-DAY LINEARITY LENGTH:"too' # 75 FREG.: J . 25 r1H a M DATE: 4MM O' N RErERENCE BLOCK l FIXTURE OR HEAD PULSE LENGT11: No.u S/N: nomg4z - 1 '
l S/N: OcDel2a - 1 JACK (T or R): T TYPE: Cedsbetd7 e b ANGLE: /,0
- L. TEST (Thru or MATERIAL: CochnSheel
- NORMAL)
- "Th r-O REFLECTOR 't
_ USED: A" I l 1 lCC:V b00Y . l VOLTS /DIV. lCC NV j SEC./DIV. 5 o 0 of , l a o a l I i ! W i b y' Y w d
- I e r i
l I I . i \ 1 l PERFORMED BY:2m D .M LEVEL:JC DATE:/7 Aft FS i , REVIEWED BY: LEVEL:' DATE: 1 FORM 028-122/81J/84 REF: 00A5532 NR ENERM SEREES. INC. = I
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m ADDENDUM E DATA FROM SPECIAL TEST BLOCK I
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