Text

Addenda to Rev 3 of NPPD Cooper Nuclear Station ISI Program
	ML20085G189
Person / Time
Site:	Cooper
Issue date:	09/30/1991
From:	; NEBRASKA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20085G183	List: ML20085G189; NLS9100661, Forwards Addenda to Rev 3 to NPPD Cooper Nuclear Station ISI Progam, Incorporating Exam of Selected Components Supports & Attachments of HPCI Sys Suction Line;
References
	NUDOCS 9110240180
	Download: ML20085G189 (172)
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{{#Wiki_filter:- - _ _ _ _ _ _. F Septesber 1991 Addenda Date of Issue: September 30, 1991 l 7'~~g - 6 's_ / NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION INSERVICE INSPECTION PROCRAM FOR ASME CLASS 1, 2 AND 3 COMPONENTS Revision 3 This is an addenda to the losse-loaf version of the Inservice Inspection Program for ASME Cicss 1, 2, and 3 Components, Revision 3, and is issued in the form of replacement or edditional pages, revisions, additions, or deletions, and are incorporated directly into the affected pages, Egmnary of Chances: This is the ninth addenda to be published to the Inservice Inspection Program for ASME class 1, 2, and 3 components, Revision 3, This change aifects both the ASME Section XI and the Augmented ISI sections of the CNS ISI Program. Changes given below are identified in the text by a margin note (September 91) next to the affected item. SECTION PAGE(s) DESCRIPTION Introduction Section All Remove and destroy existing pages and replace with attached pages. Extends second interval, deletes h references to specific augmented y inspections, and corrects typo-graphical errors. Material Specification Page 4 Remove and destroy existing page 4 Abbreviations Section and replace with attached page. Updates listings to include new material descriptions /specifica-tions. NDE Procedures All Add this new section between existing CNS UT Calibration Standards and ASME Class 1 Categories sections to identify NDE procedures referenced in the CNS ISI Program. ASME Class 1 Categories Section Tabs B-D, B-E, Remove and destroy existing compo-B-F, B-G-1, nent listing pages and replace B-G-2, B-H, B-J, with attached pages. Updatea B-K-1, B-L-2, component listings due to CNS B-M-2, B N-1, plant design changes and NDE B-N-2, B-P specficiation listing changes. ASME Class 2 Categories Section Tabs C-A, C-B, Remove and destroy existing compo-C-C, 0-F, C-H nent listing pages and replace with attached pages. Updatas component listings due to CNS s j plant design changes and NDE \\s / specification listing changes. oi4 Page 1 of 2 l h O 00 8 O )

D SECTION PACE (s) DESCRIPTION l [ME Class 3 Categories Section Tabs D A and D C Remove and destroy existing component listing pages and replace with attached pages. Updates component listings due to CNS plant design changes and NDE specification listing changes. ~ ASME Class 3 Categories Section Tab D-B Remove and destroy existing (Continued) component listing pages and replace with attached pages. Adds integral attachments on llPCI suction line per NPPD commitment to U.S. NRC, reference IR 90-15. ASME Section XI llangers and Tabs F-A, F-B, Remove and destroy existing Supports Section FC component support listing pages and replace with attached pages. Adds compenent supports on ASME Class 3 IIPCI suction line from emergency condensate storage tank per NPPD commitment to U.S. NRC, reference IR 9015. Update component listings due to CNS plant design changes. Augmented Inservice Inspections Pages 1 and 2 Remove and destroy existing pages Section 1 and 2 and replace with attached pages. Updated AISI listing for { } Tab 5 and adds new Tab 15 for RPV stud augmented exams. Augmented Inservice Inspections Tab 5 Remove and destroy first page and Section replace with attached page. Add attached pages listing Core Spray and Reactor Recirculation system piping subject to CL 88-01, Add requirements for weld crown condi-tioning for UT (future welds) per General Electric SIL Mo. 117R3. Augmented Inservice Inspections Tab 15-A11 Pa;es idd this new tab and attached Section pages. Added Type 15 AISI, RPV stud inspections per G.E. RICSIL No. 55. ,~ s a e '\\,j Page 2 of 2 o

m. e r l INTRODUCTION SECTION l l: i l-l l I l j __.,7 --_,,_q.5.. .,7 y ,.y-p p. _,m ,,,,e 4 ,_,,,y ,,4.y g .Nr - - - r -7w --w-- --m e-ve'-ww--w-r-v'w-w-r-*-v - - - - ' -e T

e l JEflLVICfdEliEGn0tL PROGRAGESCRTPTIOB EASIS FOR THE INSERVICE INSECTION PROCRAM 1.0 v; -The base document from which che ISI plan, schedule, and program is developed is 10CFR50.55a(g). Burns and Roe, The Architect Engineer for Cooper Nuclear Station during constructiosi, were NPPD's agent in determining the ASME Code classifications of our plant components, NRC _ Regu atory Gut e 1.26 was not yet published; hence, it was not used l d during Burns and Roe's classificacion. NPPD has not, nor are there plans to reclassify all the plant corrponents in accordance with the newer regulations and guides, however, as a guidance for component's inspection applicability, the NRC Regulatory Cuide 1.26, Revision 3 February,1976, l Sept. 91 is being used for determining examination boundaries primarily for Class 2 and 3 systerns. 2.0 IJ1E SECOND TEN YEAR INSPECTION INTERVAL 2.1 The commercial operation date for Cooper Nucicar Station is July 1.1974. The end of the first interval was extended from June 30,1984, to September 1, 1985, due to the 1984/85 pipe replace'eent outage. This was allowed by IWA-2400(c). The end of Sept. 91 the second interval was originally scheduled for June 30, 1994. Due to the change from 12 month to 18 month fuel cycles, the end of the second interval will be extended to December 31, 1994, as allowed ny IWA 2400(c). 2.2 The three inspection periods during the second interval are as O follows: t u O First Period: From September 1, 1985, to June 30, 1988 Second Period: From July 1,1988, to December 31, 1991 Sept. 91 Third Period: From January 1,1992, to December 31, 1994 3.0 6PPLICABLE ASME COE The Inservice Inspection of ASME Class 1, 2, and 3 components are performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition through Winter 1981 Addenda, except where specific written relief from examinations and testings determined to be impractical has been granted by the NRC pursuant to 10CFR Part 50, Section 50.55a(g)(6)(i). 4.0 INSERVICE INSPECTION PROCRAM 4.1 The Inservice Inspection (ISI) Program is a detailed living documer.t which is available to the regulatory authorities for audit. 4.2 The.ISI Program provides: 4.2.1 ASME Section XI examination category and item listin5 4.2.2 Specific examination identification and location. ^ ( 4.2.3 Weld joint configurations. \\' 4.2.4 Nondestructive examination procedures. _- a

q j d l' 4'2.5 Ultrasonic calibration st.andards list. AJ-4.2.6-Applicable list of isometric drawings. 4.2.7 Augmented Inservice Inspections. 4.2.8 Relief Requests. ~ 4. 2. 9 - Status of completed inspections. 4.2.10 Summary of previous Inservice-Inspections. 4.3 The scheduling of Inservice Inspections is in accordance with' ASME -Section XI. The mechanism for specific component scheduling is contained 1in Cooper-Nuclear Station - procedures. This form of scheduling provides for optimization of component inspection during normal refueling and maintenance activities and naintains exposure All.RA. i 4.4 All records and reports are prepared in accordance with ASME Section XI IWA 6000. 5'. 0 EXTENT OF EXAMINATIONS 5.1. All components are examined in accordance with MME Section XI, subsections IWA, IWB, IWC, IMD, and IWF. .j 5.1.1-The extent of examinations for Code Class 1 and 2 pipe welds are determined by the requirements of Table IVB 2500 -and IWB 2600, Category B-J for. Class 1 and paragraph IWC 2411 Category C-F and' C C - for Class 2 in t.he 1974 ? Edition through the Summer 1975 Addenda of ASME Section XI. ~ 5.1.2 -ASME-Section XI (Winter 1981), Subsection IWE, " Require-ments for Class MC Components of Light Water Cooled Power Plants", was added to Section XI since - Summer 1975. Eowever, 10CFR50. 55a presently only incorporates those portions of Section XI-that address the-ISI requirements for Class 1, 2, and 3 components and their supports. The-regulation does not currently address the ISI of contain-9 { ments. ' Hence, ISI of MC components and their supports are not included in t'.te ISI Plan and Program. (NOTE: From the Foderal Register /Vc',ume 48, No. 26; 10CFR Part 50,-dated Monday, February 7,-1983, Supplementary Information.) 5.2 All Class 1, 2,' and 3 piping components have been reviewed against the respective ASME Section XI. IWB 1220, IWC-1220, and IWD-1220 exemption criteria. 5.3 Appropriate Code Class 2 pipe welds in Residual Heat Removal Systems, Emergency Core Cooling System, and Containment Heat Removal Systems shall be examined. 5.4 System pressure tests will be conducted in accordance with the Sept. 91 3Q requirements of ASME Section XI, Subsections IWA, IWS, IWC, IWD, and IWF... -. . -.)

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5.5( 'Dntermination of ASME Section'XI Code-categories for Class 1,-2

'd and 3 components.

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All Class 1, 2 and 3 components are categorir.ed in accordance with ASME Section XI, Table 1WB-25001, Table IWC-2500 1, ard Table IWD 2500-1, respectively. 6.0 ' EAMINATION BOUNDARIM 6.1 Examination boundaries are these ASME Class 1, 2, and 3 fabricated t and installed components,' their attachments and supports. ~6. 2 -- ASME Class 1, 2 and 3 boundaries are identified on color coded P& ids and are located at Cooper Nuclear-Station. 6.3 The following systems or portions of-systems are included in the ISI examination boundaries: OP6TD SYSTEM 6.3.1 2004 Condensate and Feedwater 6.3.2 2022 - Primary Containment Cooling and Nitrogen -Inerting 16.3.3 2026 Reactor Vessel Instrumentation g -: / ). "(] ' 6.3.4 2027 Reactor Recirculation and Suppression Oct. 85-Chamber Vent 6.3.5 2028 Reactor Building and Drywell Equipment Drain 6.3.6 2030 Fuel-Pool Cooling and Cleanup 6.3.7 2031 Reactor Building Closed Cooling Water 6.3.8 2032-liigh Conductivity Floor Drains 6.3.9 2037 Standby Cas Treatment and off Gas Filters 6.3.10 2038 Reactor Building Floor and Roof Drains 6.3.11 2039 Control Rod Drive liydraulic 6.3.12 '2040 -Residual Heat Removal 6.3.13: 2041 Main Steam - Reactor Building 6.3.14 2042 Reactor Water Cleanup 6.3.15 2043 Reactor Core Isolation Aj 6.3.16 2044 liigh Pressure Coolant injection 6.3.17 2045 Core Spray and Standby Liquid Control u

l l l -,m OP61D SYS1EM 6.3,18 2049 Condensate Supply 6.3.19 2084 Atmospheric Containment Atmospheric Dilution 6.3.20 2077 Diesel Cencrator Building Service Water, Starting Air, Fuel, Oil, Sump, and Roof Drains 6.3.21 6000302 Augmented Off Cas 6.3.22 13095-12-FSK-1-1 post Accident Sampling 7.0 AUCMENTED INSERVICE INSPECTIONS 7.1 Augmented Inservice Inspections (AISI) are not ASME Section X1 Code requirements, but are 1) additional examination areas, or 2) increased inspection frequency or combinations of both which are requested by the Nuclear Regulatory Conunission. 7,2 When examination components fall into the scheduled testing of both ISI and AISI, then credit for both requirements are taken (no double testing). / i 7.3 There are several types of Augmented Inservice It.spection required Sept. 91 at Cooper Nuclear Station; refer to the beginning of the " Augmented Inservice Inspections" section for a complete listing of these inspections. 7.4 The following A3ME Category welds are being climinated from the A'igmented Inspection and are listed in a separate section of the ISI program. 7.4.1 Welds susceptible to Intergranular Stress Corrosion Crack-ing that are required to be inspected per NURFG-0313, 7.4.2 Welds, category B-F and B-J, that are designated as " Pipe Whip". I 8.0 RELIL REOU F,STS Oct. 85 8.1 When an ASME Code Class 1, 2, or 3 component is determined to be impractical to inspect in accordance with ASME Section XI IWA 2000, IWB-2000, IWC-2000, IWD-2000, or IWF-2000, a specific written Relief Request from the ASME Code is submitted to the NRC in accordance with Section 3.0 (above). Each written Relief Request contains the following information as a minimum: 8.1.1 Identification of component (s) for which relief is requested. 8.1.2 ASME Section III Code Class. 8.1.3 The specific ASME Code requirement that has been determined to be impractical...

-1 L ;t g.j-I 8'.1.4 -Cooper Nuclear Station relief justification (s) information pT' '"j : for requesting relief. E 8.1.5 Specific. alternative inspection (s) in lieu of ASME Code Section-XI_ requirement (s). 8.2-The following is a list of Relief Requests: Relief Reauest [if Number DescrIntion 1-ASME __ Category B J. Inaccessible

Welds, Primary Containment-2 See Note 1.

l Hay 87 3 ASME Category B D, RPV Top Head. Nozzle,- j Inner Radii-I 4 ASME Category. C-F, Inaccessible Velds in Floor Penetrations. 5-ASME Category C A, Inaccessible Welds on the - h -EllR lleat Exchanger. ASME Category B A, Inaccessible RPV Welds 6 7_ %. NOTE 1:: Relief Request Number 2, "ASME Category C F, RllR Drywell ( s .. Spray Internal to Drywell", was-not granted by the NRC and 'thus, removed per the May, 1987, addenda. Reference Tab 8-of correspondence section. L i (A - 1

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MATERIAL SPEC.IUCATION/ DESCRIPTION MATERIAL SPECIFICATION KATERIAL SPECIFICATION / DESCRIPTION ABBREVIATION F 25 SA-216 May 87 P-20 316 NG stainless steel in accordance with NPPD May 87 Contract 83 41, Section G, page C 13. RPV-2 A 508 Oct. 89 RPV-3 A-508 Oct. 89 S-1 SA 540 GR B24/RPV Studs Oct. 89 NOTE: Material specification abbreviations correspond with Cooper Nuclear Station's Material Specification Coding Tables used during construc-tion, except for material specifications greater than P-17 and F-22, and RPV-1, RPV-2, RPV-3, and S-1, which have been added. P 21 SA 312 seamless and welded austenitic stainless ' Sept. 91 steel pipe, grade TP316L. F-26 SA 403 wrought austenttic stainlass steel piping Sept. 91 and fittings, grade WP316L. F-27 SA-182 forged or rolled alloy steel pipe flanges, Sept. 91 forged fittings, and valves and parts for high temperature service, grade F316L. __- _____

NDE PROCEDURES SECTION (NEW) 3 L__.__.______.______._.___

n i 1 \\ j NDE PRM fdlVBLS %s The following NDE procedures are used by the CNS ISI Program. The numbers listed in the VT, pts MT, UTO, UT45, and UT60 columns of the component listing pages in the ISI Prograr correspond to the Tab /Index Number of the NDE procedure listed below that may be used for that examination. Where there is more than one procedure referenced for a specific UT examination, the option exists to use only one of the procedures. All NDE procedures listed below are on file at Cooper Nuclear Station. The most current revision of a referenced NDE procedure approved by the CNS Station Operation Review Committee (SORC) will be used. TAB /,INDEX PROCEDURE PROCEDURE TITLE NF BER NUMBER 1 GE-UT 300 Procedure for Manual Examination of Reactor Vessel Assembly Welds 2 GE.UT-308 Procedure for Manual Ultrasonic Examination of the RPV Flange Ligaments 3 GE-UT-303 Procedure for Manual Ui uasonic Examination of Nozzle Inner Radius Greater Than 10" Diameter 4 MIUSK-W812 Ultrasonic Examination of Support Skirt to ('~] Reactor Pressure Vessel Wolds ij \\ 5 MIUB-W812 Procedure for Ultrasonic Inspection of Pressure Retaining Bolting Two Inches or Greater in Diameter 6 GE-UT-106 Procedure for Manual Ultrasonic Examination of Pressure Retaining Welds in Ferritic and Austenitic Piping and Components 7 CE-PT-100 Procedure for Color Contrast Liquid Penetrant Examination 8 IVl-W812 Procedure far VT-1 Visual Examination 9 IV2-W812 Procedure for VT-2 Visual Examination 10 GE-VT-100 Procedure for VT-3 Visual Examination 11 IV4-W812 Procedure for VT-4 Visual Examination 12 MIUB-NXI2 UT of Non-ASME Bolts and Studs 13 GE-UT-305 Procedure for Manual Ultrasonic Examination of Nozzle Inner Radius and Bore Regious on Small Bore Nozzles 14 GE-PT-101 Procedure for Liquid Penetrant Examination of Nozzle Inner Radius and Bore Areas 15 GE-MT-101 Procedure for Wet Fluorescent Magnetic Particle (p Examination y v 16 CE-MT-100 Procedure for Magnetic Particle Examination Page 1 ot 3

p ^_) ( TAB /INDEX PROCEDURE PROCEDURE TITLE NUMBER NUMBER 17 -GE UT-205-Procedure for Automated Ultrasonic Examination of Pressure Retaining Welds in Ferritic and Austenitic Piping Components l 18 GE-UT-200 Procedure for Automated Ultrasonic Examination of Similar and Dissimilar Piping Welds for 1GSCC 19 VT-06 Procedure for Cooper Nuclear Station Reactor Pressure Vessel Internal Invessel Visual Inspection (IVVI) 20 GE-UT-600 Procedure for Ultrasonic Thickness Measurements of Nuclear Components 21 GE UT-601 Procedure for Ultrasonic Thickness Measurements for Erosion / Corrosion 22 GE-UT-301 Procedure for Manual Ultrasonic Examination of Pressure Retaining Vessel Welds Less Than Two Inches in Thickness 23 GE-UT-307 Procedure for Ultrasonic Examination of RPV Closure Studs N 24 GE-UT-309 Planer Flaw Elaing for Nozzle Inner Radius and f j Bore Regions %/ 25 GE UT-310 Verification of feedwater Sparger Secondary Thermal Sleeve Location 26 CE-UT-102 Procedure for Manual Ultrasonic Examinatior,of Similar and Dissimilar Piping Welds for ICSCC 27 CE-UT-104 Procedure for Manual Ultrasonic Planer Flaw Sizing 28 GE-UT-105 Procedure for Manual Ultrasonic Examination of Dissimilar Metal Nozzle To-Safe End Welds 29 GE-UT 202 Procedure for Automated Ultrasonic Exam!. nation of Dissimilar Metal Nozzle-To Safe End Welds 30 GE-UT-400 Procedure for Remote Ultrasonic Examination of RPV Welds 31 GE-UT-402 UT Alternative to NUREG 0619 Nozzle Radius and Bore PT Requirement 32 GE-UT-501 Procedure for Remote, Ultrasonic Examination of Shroud Head Hold-Down Bolting 33 TP-508-1477 Procedure for the Ultrasonic Inspection of Bell Housing-Type Recirculation Pump Shafts j/"'} 34 C-lVCV-01 Cooper Invessel Clearance Verification Procedure 's / 35 246-GP-48 Procedure for Control Rod Blade Inspectioa m Page 2 of 3

TAB /INDEX PROCEDURE PP.OCEDURE TITLE NUMBER NUMBER 36 CE-ADM-1001 Procedure for Performing Linearity Checks on Ultrasonic Instruments 37 GE-ADM-1002 Procedure for Review Process and Analysis of Recorded Indications 38 GE-ADM-1003 Procedure for Operational Guidelines with " Smart UT" System 39 GE-ADM-1005 Procedure for Zero Reference and Data Recording for Non-Destructive Examination 40 GE-ADM-1006 Procedure for Compliance uith U.S. NRC Regulatory Guide 1.150 Page 3 of 3

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I ,m 6119ERTfdl_lLSERVICE INSPECT 1QNS ) - J Augmented Inservice Inspections (AISI) are not ASME Section XI Code requirements, but are 1) additional examina.tions areas or 2) increased inspection frequency or combinations of both which are requested by the Nuclear Regulator
~~Commission, recommended in General Electxic Company Service Infocmation Letters or added for other reasons.~~
When examination components fall into the scheduled testing of ISI and are also AISI requirements, then credit for both requirements may be taken (no double testing). The following ate types of Augmented Inservice Inspections required at July 89 Cooper Nuclear Station. The TAB number corresponds to tabbed pages that follow which contain information on the specific type of AISI. TAB TYPE DESCRIPTION REVISION DATE 1 1 All ring girder bolting and ring girder anchor Original bolting is to be volumetrically inspected each ten Release - 3/85 year interval. The anchor bolting adjacent to the inboard MSIV is to be visually inspected each ten year interval. (
~~Reference:~~
NRC DRO Bulletin #74-34 ) 2 2 Ultrasonic examination of the feedwater nozzle safe original /'~'y ends, bores, and insido blend radii, liquid Release - 3/85 / penetrant examination of the feedwater nozzles, and visual inspection of the feedwater spargers as required per Table 2 and Section 4.3,2.4 of NUREG-0619. 3 3 Visual inspection of the Core Spray spargers and Original the Core Spray piping inside the RPV shall be Release - 3/85 conducted each refueling outage. (
~~Reference:~~
IE Bulletin No. 80-13.) 4 4 Ultrasonic examinations, utilizing G.E. Procedure Original TP508.0654, Revision D, or equivalent, are Release 3/85 conducted to assess the integrity of the jet pump hold down beams at the mid-length ligament areas bounding the beam bolt. These examinations shall be performed once during the second ten year interval. These examinations may be deferred to the end of the interval. 5 5 Ultrasonic examinations per Generic Letter (G.L.) Sept. 91 88-01. G.L. 88-01 applies to all BVR piping made of austenitic stainless steel that is four inches or larger in nominal diameter, and contains reactor coolant at above 200*F during poaer operation regardless of code classification. All accessible welds will be examined in accordance with CNS G.L. 88-01 commitments. Add requirements for weld crown ,-s, [ \\ conditioning for UT (future welds) per Generic \\,,) Electric SIL No. 117R3. l l Page 1 of 3 \\ l L i
4 ~ m TAB TYPE DESCRIPTION REVISION DATE J -6 6 Visual inspection of steam dryer channel welds Feb. 89 during refueling outages (Refererce General Electric SIL No. 474.) 7 7 Visual inspection of jet pump 'ozzles and mixer July 89 inlets in conjunction with jet pump inspection. (
~~Reference:~~
General Electric SIL No. 465.) 8 8 Ultrasonic examination of the shroud support access Aug. 91 hole covers once every three years beginning with the Spring 1993 Refueling Outage. Visual ) examination VT 1 of the shroud support access hole covers during the 1991 Refueling Outage. (
~~Reference:~~
~~General Electric SIL No. 462 S2, R1, and G.E. memo RCil9143, dated April 4, 1991.) -9 9 Visual inspection of the Core Spray T-junction box July 89 welds inside the react <.r vessel. (~~
~~Reference:~~
General Electric SIL No. 289, R1, S1.) 10 10 Visual examination of the Reactor Recirculation July 89 (RR) pumps' shafts, pump. covers, Impeller / shaft attachment reLion.(including bolts), and hydrostatic bearings (including baffle plate). "requency of examination shall coincide with the regularly scheduled RR pumps inspection. -O (
~~Reference:~~
General Electric S't No. 459 and \\Q RICSIL No. 038.) 11 11 Visual examination of all accessible areas of the Aug. 91 Intermediate Range Monitor (IRM) and Source Range Monitor (SRM) dry tubes each refueling outage. (
~~Reference:~~
~~General Electric SIL No. 409 R1.) 12 12 Ultrasonic (UT) examination of all remaining old Aug. 91 design creviced Inconel 600 Shroud llead Bolts (SilBs) each refueling outage. (~~
~~Reference:~~
General 1.lectric SIL No. 433.) 13 13 Rationale for updating of General Electric reactor Aug. 91 pressure vessel ultrasonic inspection procedures to incorporate U.S. NRC Regulatory Guide 1.150 requirements. -(
~~Reference:~~
General Electric SIL No. 515.) I 14 14 Augmented Inservice Inspection (AISI) Program for Aug. 91 Service Water (SW) and Reactor Equipment Cooling (REG) pipe supports outside the scope of the CNS ASME Section XI ISI Program. This AISI Program requires only VT-3 and/or VT-4 examinations of selected supports. O Page 2 of 3 ~..
~. ~, i 1
~~,,~-~~
[ 'Y TAB TYPE DESCRIPTION REVISION DATE \\'}. 15 15 Ultrasonic examination of CNS RPV V'ad Studs in . Sept, 91 accordance with General Electric Ri: ill No, 055, RPV studs 1-21 and 26 scheduled for examination based on raw material heat treatod hardness values (ASME Section XI Category B G 1 Code credit will be taken for RPV studs 18-21 and 26). em \\ ( l l g'~N (s_.- l Page 3 of 3 l
~ .. ~ -... O l AUGMENTED INSERVICE INSPECTIONS SECTION O TAB 5 l l l l O
m. e v COOPER NUC1. EAR STATION ISI PROGRAM ~ i x" Augimented inservice Inapection Program In Accordance with U.S. NRC Ceneric Letter 88 01 {
~~References:~~
~~1. NRC position on TCSCC in BWR austenitio stainless steel piping (Generic Lettet 88 01). 2.~~
~~Letter, C.~~
~~A. Trevors (h?rD) to U.S.~~
~~NRC, dated October 9, 1990.~~
~~Subject:~~
~~Generic Letter 88 01, Cooper Nuclear Station. 3. L.tter, P. W. O'Connor to C. R. Horn (!WPD), dated May 14, 1991.~~
~~Subject:~~
Reactor Vater Cleanup (RWCU) Pipe i Vela inspection. As a result of previous commitments to Reference 1. NPPD replaced the snajority l of the Category D Intergranular Stress Corrosion Cracking (ICSCC) susceptible piping and welds with Category A ICSCC resistant piping and welds during the 1990 Refueling Outage. The only Cat 6 gory D piping and volds thac were not replaced conststs of the RWCU return piping from the regenerative heat exchanger outle to the RWCU/RCIC attachment to the feedwater inlet line. In Reference 2 NPPL proposed to revise its previous conantement to replace this piping and instead conduct continued inspection of this piping in accordance with the requirements f of CL 88 01. This request was approved and future inspection requirements for all RWCU piping weli inspections were documented in Reference 3. [ -y f I -The following pages document the CNS Augmented ISI requirementr for all of the applicable RWCU piping velds in accordance with Referenco 3. All inaccessible i Category D welds have been reclassified us Category C welds in accordance with [ CL 88 01. Also, weld RCA BF.1 (CRD nozzle cap weld).has been reclassified as Category D'and included Jn the attached program. It should be noted that thire are additional ICSCC Category A welds in Class 1 Sept.91 portions of the reacter recirculation, cora spray, and RWCU systems. These welds are included in the CNS ASME Section XI ISI Pro 6 ram. These welds are also documented in the following pages in order to provide a complete list of all welds subject to C.L. 88 01, NPPD will conform to the NRC staff position on reporting requirements as stated in CL 88 01.. The NRC will be notified of any flaws identified that do not meet 1WB 3500 criteria from-Section XI of the Code for continued operation without evaluation or a range found in condition of welds previously known to be cracked. The NRC Lal be notified of any flaw evaluation required for continued + operation:and/or v repair plans. ' All Category A welds in austenitic stainless ste '.' piping at CNS have had the Sept.91 ,outside surface' weld crowns condition $d (machined)_to provids optimum surface condition for ultrasonic (UT) examination. This was done in accordance with General Electric Company recommendations which are now documented in C.E. Service Information Letter'(SILS No. 117, Revision 3, _ All future welds in austenitic l stainless steel piping systems subject to Generic Letter 88 01 wi? ? be condi. tiened/ machined for ultrasonic inspection in accordance with the recommendations of SIL No. 117 Revision 3. A copy of this SIL is included at the end of this section for further information, ii I
1 [d N-v V CENERIC LETTER 88-01 AUCMDTTED ISI EXAMINATIONS CATECORY A REACTOR RECIRCUILATION AND CORE SPRAY PIPE WEIDS Sept-91 VELD ID CONF. SIZE ISO. MAT. UT PROC INTERVAL REMARKS D AGE CSB-BJ-2 P-SE 10 CNS-CS-3 P20 49~ 6 2 1 ICSCC Cat. A per G.L. 88-01. Completed F86. CSB-BJ-3 P-P 10 CNS-CS-3 P20 t49 6 2 1 ICCC Cat. A per C.L. 88-01. Completed F86.. i CSB-SJ-4 E-P 10 CNS-CS-3 P20 49 6 2 1 IGSCC Cat. A per C.L. 88-01. Completed F86. CSA-BJ-2 P-SE 10 CNS-CS-4 P20.$ 6 ICSCC Cat A per G.L. 88-01. CSA-BJ-3 P-P 10 CNS-CS-4 P20 6 IGSCC Cat. A per G.L. 88-01. CSA-BJ-4 E-P 10 CSN-CS-4 P20 6 ICSCC Cat. A per G.L. 88-01. RAD *BJ-1 PU-P 20 CNS-RR-37 P20 56 6 2 1 IGSCC Cat. A per G.L. S8-01. Completed F86. RAD-BJ-2 P-VA 28 CNS-RR-37 P20 '6 6 2 1 IGSCC Cat. A per G.L. 38-01. Completed F86. RAD-BJ-3 VA-P 28 CNS-RA-37 P20 56 6 2 1 IGSCC Cat. A per C.L. 88-01. Completed F86. RAD-BJ-4 P-T 28 CNS-LR-37 P2O 56 6 2 1 ICSCC Cat. A per G.L. 88-01 Completed F86. RAD-BJ-5 T-4V 30 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RAD-EJ-6 P-T 24 CNS-RR-37 P20 6 Cat. A per C.L. 83-01. RAH-BJ-1 4W-P 22 CNS-RR-37 P20 6 Is; SCC Cat. A per G.L. 88-01. RAH-BJ-2. 4V-P 22 CNS-RR-37 P20 6 ICSCC Cat. A per C.L. 88-01. Page 1 of 5 ,.,y-ww g J-c-4# r-p ,.-p. 3 m y-y w
' (G N (~~N /"N \\ p 'Qf GENERIC LETTER 83-01 AUGMENTED ISI EXAMINATIONS CATECORY A REACTOR RECIRCUIIATION AND CORE SPRAY PIPE VEIDS Sept. 91 WELD ID CONF. SIZE ISO. MAT. UT FROC INTERVAL REMARKS O. O GE RAS-BJ-10 W-E 9 CIS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RAS-BJ-11 E-E 20 CNS-RR-37 P20 53 6 2 1 ICSCC Cat. A per G.L. 85-01. Completed F86. RAS-BJ-13 P-V 20 CNS-RR-37 P20 53 6 2 1 IGSCC Cat. A per G.L. 88-01. Completed F86. RAS-BJ-2 SE-P 23 CNS-RR-37 P20 56 6 2 1 IGSCC Cat. A per G.L. 88-01. Completed F86. RAS-BJ-3 P-T 2S CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RAS-BJ-4 T-P 28 CNS-RR-37 P20 6 IGSCC Cat. A per G.L. 88-01. RAS-BJ-5 P-VA4 28 CNS-RR-37 P20 '6 I ICSCC Cat. A per G.L. 88-01. RAS-BJ-6 VA-P 28 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RAS-BJ-6A P-VE 28 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RAS-BJ-6B VE-F 4 CNS-RK-37 P20 6 IGSCC Cat. A per G.L. 88-01. RAS-BJ-7 P-E 28 CNS-RR-37 P20 6 ICSCC Cat. A pyr G.L. 88-01. RAS-BJ-8 E-PU 28 CNS-RR-37 P20 6 1GSCC Cat. A per G.L. 88-01. RAS-BJ-9 T-E 20 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. i RRF-BJ-2 P-SE 12 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RRF-BJ-3 P-P 12 C'5-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RRF-BJ-4 P-P 12 CNS-2R-37 P20 6 IGSCC Cat. A per G.L. 88-01. RRF-BJ-5 R-P 12 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. Page 2 of 5 e.
k / y / i ( i Q Q,) G' CENERIC LETTER 88-01 AUGMENTED 151 EXAMINATIONS CATECORY A REACTOR RECIRCUII/LTION AND CORE SPRAT PIPE UEIIS Sept. 91 WELD ID CONF. SIZE ISO. MAT. ^ ~~ . JO. O' AE RRG-BJ-2 P-SE 12 CNS-RR-37 P20 6 ICSCC Cat. A per C.L. 88-01. ERG-M-3 T-P 12 CNS-RR-37 P20 6 ,ICCCC Cat. A per C.L. 88-01. RRri-BJ -2 P-SE 12 CNS-RR-37 P20 50 6 2 IGSCC Cat. A per C.L. 88-01. Completed F86. RRH-BJ-3 R-P 12 CNS-RR-37 P20 50 6 2 1 ICSCC Cat. A per C.L. 88-01. Completed F86. l RRJ-BJ-2 P-S E 12 CNS-RR-37 P20 6 ICSCC O t. A per C.L. 88-01. l l RRJ-BJ-3 T-P 12 CNS-RR-37 P20 6 ICSCC Cat. A per C.L. 88-01. l RRK-EJ-2 P-SE 12 CNS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01. RRK-EJ-3 P-P 12 CNS-RR-37 P20 6 ICSCC Cat. A per C.L. 88-01. RRK-3J-4 P-P 12 CNS-RR-37 P20 6 ICS C Cat. A per C.L. 68-01. RRK-BJ-5 P-P 12 CIS-RR-37 P20 6 ICSCC Cat. A per G.L. 88-01 RBD-Ba-1 PU-P 28 CiS-RR-38 F20 6 ICSCC Cat. A per C.L. 88-01. RBD-BJ-2 P-VA 28 CNS-RR-36 P20 6 ICSCC Cat. A per C.L. 88-01. RBD-BJ-3 VA-P 28 Ci3-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. 6 IGSCC Cat. A per C.L. 88-01. J RBD-BJ-4 P-T 28 CIS-ER-3 S P20 RB9-BJ-5 T-47 30 CJS-RR-38 P20 6 IGSCC Cat. A per G.L. 88-01. RBD-BJ-6 P-T 24 _ lCNS-RR-38 P20 6 ICSCC Cat. A per C.L. 88-01. RBH-BJ-1 47-P 22 CNS-RR-38 P20 6 ICSCC Cat. A per C.L. 31-01. Page 3 of 5
/"%. rz (N (r,); ) / Q,l GENERIC IETTER 88-01 AUGMENTFD 151 EXAMINATIONS CATECOP.Y A REACTOR RECIRCUIIATI0h AFD CORE SPRAY PIPE WELDS Sept. 91 l' UELD ID CONF. SIZE -150. MAT. UT PROC INTERVAL REMARKS O. REH-BJ-2 4W-P 22 CPC-RR-38 P20 6 IGSCC Cat. A per G.L. 88-01. RBS-BJ-2 SE-P' 28 CNS-RR-38 P20 E IGSCC Cat. A per C.L. 88-01. RBS-BJ-3 P-P 28 CNS-RR-38 P20 6 ICSCC Cat. A per C.L. 88-01. RBS-BJ-4 P-P 25 CNS-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. RBS-BJ-5 P-VA 28 CNS-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. RBS-EJ-6 VA-P 28 CNS-RR-38 P20 6 ICSCC Cat. A per C.L. 88-01. RBS-BJ-6A P-VE 28 CNS-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. RBS-BJ-6B UE-F 4 CNS-RR-38 P26 6 IGSCC Cat. A per G.L. 88-01. RBS-BJ-7 P-E 28 CNS-RR-38 P2G 6 ICSCC Cat. A per G.L. 88-01. RBS-BJ-8 E-FU' 2S CNS-RR-38' P20 6 ICSCC Cat. A per G.L. 38-01. RRA-BJ-2 P-SE; 12 CNS-RR-38 P20 6 IGSCC Cat. A per G.L. 88-01. I l RRA-BJ-3 P-P 12 CNS-ER-38 P20 6 IGSCC Cat. A per G.L. 88-01. 1 REA-M-4 P-E' 12-CNS-RR-38 P20 6 ICSCC Cat. A per C.L. 88-01. l RRA-M -5 R-P 12 CNS-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. l RRB-BJ-2 P-SE 12 CNS-RR-38 P20 6 IGSCC Cat. A per G.L. 88-01. RRB-BJ-3 T-P 12 CNS-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. RRC-EJ-2 P-SE 12 lCNS 7R-38 P20 6 ICSCC Cat. A per G.L. 88-01. g RRC-BJ-3 R-P 12 CNS-RR-38 P20 6 ICSCC Cat. A per G.L. 88-01. Page 4 of S w
~ -~
~~Q Q y,~~
-fN ] b"i 7 o o (f r s. -k CENERIC LETTER git-01 AUGMENTED ISI EXAMINATIONS CAIM,vnx A REACTOR RECIRCUIIATION AND CORE SPRAT FIPE WEIDS 'l 4 -l S( t_ d ~ WELD ID CONF. SIZE 150. MAT. UT PROC Ih*TERVAL REMARES A 2 I ~ RRD-BJ-2 P-SE 12 CNS-ER-38 P20. 6 IGSCC Cet. A per G.L. 88-01. RRD-BJ-3 T-P 12 CNS-RR-38 P20 6 IGSCC Cat. A per G.L. 88-01. RRE-EJ-2 P-SE 12 CNS-RR-33 P20 50 6 2 1 ICSCC Cat. A per G.L. 88-01. Completed F86. RRE-EJ-3 P-P 12 CNS-RR-38 F20-50 6 2 1 ICSCC Cat. A per G.L. 88-01.I j RRE-1 -4 P-E 12 CNS-RR-38 P20 50 6 2 1 ICSCC Cat. I. M r G.L. 88-01. Completed F86. .I RRE-BJ-5 R-P 12 CNS-RR-38 P20 50 6 2 1 ICSCC Cat. A per C.L. 88-01. Completed F86. i i i 6 Page 5 of 5 m.w . J
/]
r CENERIC LETTER 88-01 AUGMENTED IS1 EXAMINATIONS' CATACORY A REACTOR WATER CLEANUP UUDS ( WELD ID CONF. SIZE ISG; MAT. ITT PROC INTERVAL RDiARKS A ' CWA-M-1 P-P. 6: CNS-RUCU-3 .: P20 - 6 ICSCC Cat. A per C.L. 88-01. CWA-BJ-10 E-P. 6 CNS-RWCU-3
~~P20 48 6'~~
'2 l' ICSCC Cat. A per C.L. 88-01. Completed F86. CWA-BJ-12 .P-VA' 6 CNS-RUCU-3 'P20 48 6' 2 1 ICSCC Cat. A per C.L. 88-01. Completed F86. CUA-BJ P-P 6.- CNS-RWCU-3 P20 48 6-2. 1 ICSCC Cat. A per C.L. 88-01. 4 Completed F86. J CWA-BJ-2 P-P. 6 CNS-RWCU-3 P20 48 6-2 1 ICSCC Cat. A per C.L. 88-01. y Completed F86. l CWA-BJ P-P 6 CNS-RVCU-3 P20' 6 ICSCC Cat. A per C.L. 88-01. l-CWA-BJ-4' P-VA 6-CNS-RWCU-3 P20 6 ICSCC Cat. A per C.L. 88-01. CWA-BJ-3 VA-P 6 CNS-RWCU-3 P20 6 IGSCC Cat A per G.L. 88-01. CUA-BJ-6 P-P - 6 CNS-RUCU-3 P20 6 ICSCC Cat. A per C.L. 88-01. CWA-BJ-7 P-VA 6 CNS-RWCU-3 P20 6 2 1, 2 ICSCC Cat. A per C.L. 88-01. Completed F86, S88. CWA-BJ-8 VA-P 6 CNS-RWCU-3 P20 48 6 2
~~1, 2 IGSCC Cat. A per G.L. 88-01.~~
Completed F86, S88. CWA-BJ-9 P-E 6-CNS-RUCU-3 P20 6 ICSCC Cat. A per C.L. 88-01. 1 i' NOTE: These welds are also listed in the CNS ASME Section II Class 1 ISI Program. Page 1 of 1 _. ~. - -. _..., _. -. -
] ~ GC NucIcar Enetyy [' ..,,... m m h. s m om,4 4 ~ w.e.: s. m Services Information 1. citer September 27,1990 SIL No.117 Prvision 3 Category 1 DETECTION AND EVALUATION OF IGSCC IN WELDED AUSTENITIC PIPING
~~. Background~~
Previous Revisions and Supplements to SIL No.117 have tracked the significant occurrences regarding intergranular stress corrosion cracking (IGSCC) in the. primary stainless steel piping systems in GE HWRs. The purpose of these com - i munications has been to alert GE UWR owners to IGSCC problems nr they oc- . curred and to make specific inspection recommendations. The most recent. modification of SIL No.117 was Supplement 1 to Revision 2," Specific Elements -r for UT Procedures for Detection and Evaluation of IGSCC in Welded j' Austenitic Piping,' issued in December 1983. Its purpose was to provide specific procedural elements to GE HWR owners to assist thern in meeting the then new r requirements for training and qualification set forth in USNRC Generic Letter 83-02. Since 1983, industry advances in training non destructive examination personnel and qualifying related equipment and procedures, along with the issuance of subsequent USNRC inspection guidelines have contributed to improved detec. ' tion and sizing of IGSCC indications in piping welds during in service inspec-tions.
~~lhe purpose of this Revision 3 to SIL No.117 is to. furnish to GE BWR owners~~
. OE Nuclear Energy's current recommendations regarding the detection and . evaluation of IGSCC indications in welded austenitic piping systemsc 1his Revi-sion 3 to SIL No.117 volds the originst SIL No.117 and all of its previously issued Revisions and Supplements. These include SIL Nos. 117,117 R1, 117RIS1,117RIS2,117RIS3,117R2 and 117R2S1 ("R" means Revision; *S" means Supplement.) Recommended GE Nuc! car Energy recommends that owners of GE BWRs incorporate the fol-Action lowing guidelines into their in service inspection programs regarding IGSCC in ~ austenitic piping system welds. ~ 1.~ Adhere to the USNRC's recommendations provided in NUREG 0313 Re-vision 2,
~~Technical Report on Material Selection and Process Guidelines e~~
~~for HWR Pressure Boimdary Piping~~
~~and with the most recent USNRC ge-~~
'( neric letters and bulletins related to this subject. \\
r i j ,") 2. Qualification requirements for detecting and sizing IGSCC indications in ( \\ j piping we'ds are set forth in the NRC EPRI BWROG Coordination Plan (Reference NUREO 0313 Revision 2, Paragraph 5.2.1). Apply crdy those ?rocedures and equipment which are qualified to those standards and spec-fications and assign only those personnel who are qualified to them in de-tecting and sizing 10 SCC Indications in piping welds. 3. Where possible, use automated ultrasonle testing (UT data acquisition sys-tems that record A scan waveforms, which increase) inspection reliabilit and record data for comparisons with data from previous and subs (quent inspections. 4. Condition the outside surlace weld crowns of all piping welds to provido a good surface for UT. 'Ihe ideal weld surface is a flush surface, which csn require complete weld crown removal. liowever, because weld shrinkage and minimum wall thickness limitations make this impossible, OE Nuclear Energy recommends the weld crown conditioning shown in Figure 1 and described below. Smooth, rounded translito[ns -\\ j 0.030
~~ruam 3~~
s A t 1t7 N n h ) ~ l ( (, j T / y-- sj g Weld Crown Conditionlut for Nondestructive Examination Recommended weld crown surface preparation: Grind the weld flush with the adjoining base metal, if practical, without o violating the minimum thickness T. o if unobstructed access of at least 2T + 2 inches is available on both weld edges, the height of the weld reinforcement beyond the pipe outside diameter must not exceed 0.030 inch. The weld crown must blend smoothly and gradually with the base metal. 'the transition from base metal to weld crown should not exceed 3:1. If steeper transitions exist, the weld crown must be ground flush with the base metal. . (xSil, No !!7 ( Zenision 3 v Category 1 Page 2
'O if uncbstructed access of at least 2T + 2 inches is available only on one o \\] weld ed;e, grind the accessible weld edge ilush with the adjoinmg base metal. To acceive additionalinformation on this subject or for assistance in imple-menting a recommendation, please conuct yourlocal GE Nuclear Energy Ser-vice Representr*e. Technical J. P. Clai k Source Notice his SIL pertains cnly to GE DWRs. GE Nuclear Energy prepareJ this SIL ex. clusively as a service for owners of GE DWRs. GE Nuclear Lnergy has not con. sidered or evaluated the applicability, if any, of information contained in this SIL to any plant or iacility other than GE BWRs. Determination of applicability of information contained in this SIL in a specific BWR and implementation of recommended action are the responsibilities of the owner of that BWR. No warranty or iepresentation expressed or implied is made with respect to the accuracy, completeness or usefulness of this Information. General Electric Company assumes no responsibility for liability er damage which may result from the use of this information. Issued by k 09 6 / EEKiocre Cusmmer Service Communications Manager Product A71. Plant Recommendations Reference SIL No.117 Revision 3 C Category 1 ( Page 3
AUGMENTED INSERVICE INSPECTIONS SECTION TAB 15 (NEW) l
3 CENERAL ELECTRIC COMPANY RICHIL NO.- 05b RPV llEAD STUD CRACKING General Electric Company RICSIL No. 055 oveuments cracking observed in RPV head studs a t a dome s t ic IWR. The cracking occurred in the first two fully engaged threads below the RPV top flange. The cracking mechanir,m was identified as t ransgranular ot.ress corrosion cracking (SCC). A review of CNS RPV stud f abrication records was conductad, and studs 1 21 and 26 were f at.ricated from innt erial most susceptible to SCC, 1ased on RICSIL No. 055 guidelinen. CNS will perform an ultrasonic (UT) examination of RPV studs 1-21 ano 26 during the 1991 Refueling Outage. This examination will be a straight beam (0 degree L wave) examination performed frein the top end of the stud. This examination will be conducted in accordance with the requirements of the ASME Boiler and Pressure Vessel Code Section XI, and RICSIL No. 055 recommendations for examinat. ion sensitivity will be incorporated, Any indication observed will be evaluated in accordance with ASME Code Section XI requirements. A shear wave UT examination from the RPV stud center drilled hole will be considered f or uso in sizing any observed indications, If evaluation of observed UT indications requires removal of an RPV stud, the stud will be magnotic particle (MT) inspected to confirm the prescuce of cracking. The res ults of all RPV stud examinat ions will be documented in the NDE services report for the 1991 Tall Refueling Outage, Since RPV studs 1 17 were previously examined (1988) for AS!!E Section XI Code credit, only UT examination of RPV studo 18 21 and 26 will be applied for ASME Section XI code credit. A copy of RICSit No. 055 and the Ct1S response to the RICSIL are attached for additional information.
- !?!: I gy \\ GE NucInr En:rgy %7.LWs% untn n,I T, J t l'etwary 1,1991 GE Nuclear Energy has issued the RICSIL Idendfled below, a copy of which is enclosed wint this letter for yourinformation. -.........,...,.,..,,,ww.;,,.,.g,.,.,,,~g, ,,y w ', f9,5b;J; 'ln'hll:$h}.$-$- la. / II a!.kWWW & S....: .tj If you have questions about this RICS!L or if you wish to change the address to which GII distributes SILs and RICSILs to you, please contact your local GII Nuclear Servlees Manager. I /'Jwj t J. G. Moore i Customer Service Cammunications Manager V ce;. SIL Distribution. t t t 9 6 l _ .\\s
( ( l 6: RgM informitton Communicitlere Servicesinforttilhvtl.sitet - ~ l !( Petsuory 1,1991 HiCSII, No. 055 RPVHead Stud Cracking During routine in-service inspection (ISI) in the w.n ; 1 -L . - --,im s on[' p" ' y i! /d -"' y- - ' 4 'd cleventh refueling outage at a GE llWR located in the. ; United States, suspected <nck indleadons were old ~ ~ + setved by uhrusonic testing (lfl) in twontor 'the studs in which crackIndications were observed. pressure vessel (RPV) head studs below the thst - were fatxicated from material specified as SA 193 - engaged diread in the RPV flange, lhe Indications Class 3 with Code Case 1335 1. 'Ihe outside diame. west detected with a Section X1 Code UT technique ter of the stud is approximately 6 inches and the inside in whicts a straight beam (0 degree L wave) intro-dianneteris 1 inch. During refueling, the stud,includ-duced from the upper end traverses the length of die Ing thelower portion which is duraded, is exposed to . stud. De calitation reflector in Utis case was a 3/8 the stagnant, oxygenated water environment follow-Inch diameter flat bottomed hole.
Die reflections ing vessel flooding, 'Ihc stud Oncads near the vessel received from theindications had amplitudes of10% flenge surface may remain wet until plant start up, at to 25% of the calitution reflector and, therefore, which time the flange reaches a temperature high were not sequired by Code to be reconfed, llowever, enough to evaporate the water.
(O, these indicadons were evaluated as crackIndications. ') For uds reason, studs with indications wert itmoved 1he metallurgical examinadon of the stud with the for addidonal testing. Two availabic sparc studs were 0.7 inch deep crack showed transgranular craclJng installed in place of die two with indications.' All ' and brancidng which is typical of SCC in utis alloy. other studs were TTI' inspected wlDi no observation of The probable cause of the SCC is the exposure of the additional crackindicadons, studs in the preloaded condition to oxygenated water - during outages. Extensive oxidation and pttting were Magneuc particle examinadon was performed on observed on the studs. De metallurgical evaluation both studs to verify the presence of cracks. One stod Indicates that cruck!ng originated on the outer edge of showed ends in seven threads extending to 40% of the stud at pitted k>cadons near the Ducad icots. The - thecircumfc.vnce.1hcotherstudcontainedcracksin cracked studs had been in service for about 18 years two ducads extending 50% of (Se circumference, which included 11 refueling outages. Depth measurements were made with a 70. degree shear wave probe from the center tore hole. Maxi. The specification for the studs sets a minimum tensile mum dep01s were estimated to be 0.88 inch in one - strength of 145 kal. At the time the GE IlWR in stud and 2.09 inches in the other. Subsequently, a quesdon was designed, no specificadons on maxi-metallurgical examination was performed on nie stud mum stud tensile or yield strength were defined. . with the lesser crack depth and it was deterrnined that USNRC Regulatory Guide 1.65, issued In October 'the actual depdi was 0,7 inch -1he metallurgical 1973, vquires a maximum tensile strength of 170 ksi evaluadon eported the cause of Ole cracidng io be on RPV studs, because stud material with grcater than r, tress corrosion cracking (SCC) inidaung at pits. 170 ksi(with the corresponding hardness of R,38) had shown susceptibility to SCC, 'Ihe metallurgical - 1he purpose of this RICSIL is to descrite this situ-cvaluation reported a tensile strengut value of IS0 ksi y p ' adon and provide laterim recommeixiations for owres and a hardness of R,38 for the matedalin the outside t I ducaded area ofone crack'x1 stud. %e metallurg;1 cal ' U. of other GE IlWRr, while continuing investigations - a' undeway, pago'f
evaluation also rcport' d an impact toutimessi f 21 it. percent of f ull screen height. Normally, uds e Ibs at 410 degrees F, widle the Certifled hiates.al Test will require about 20 dll above reference g-(N Report (ChilR) for diat heat had reported 36 to S2 it-level assuming calibration on a 3/81neh \\ t, . Ibs at 410 degrees F. 'Ihc reasons for the apparrnt diameter flat bottomed hole. difference in toughness are being investigated. o Record and evaluate any indications in ac-A preliminary review by GE Nuclear Energy indi-cordance with Code. cates that there is sigrdficant structural margin in die deslgts of the RPV head closure such that the RPV o In addition, any indication int isjudged to could still meet AShiE Code rcquirements with sotne be suspect, regardless of a aptitude, should cracked head studs at the observed toughness level. be recorded and evaluated. An ultrasonic ex. In fact, for the plant at which the stud cracking was aminadon from the bore shoul:1 be consid-observed, the required Code margins could be main-cred for sizing the enck indications if tr-tained even if 10 percent of the studs were fully quired. cracked. Nevertheless, interim recommendations are fumished in this R1CSIL so that ownen of GB DWRs o Studs with suspect indicadons should be re-can assure themselves that these margins are main-moved wMre possible and examined by . tained as their plants age. magnetic particle testing for confirmadon of the UT results. Mib It ecpg.puhw$$$NM,b Nk egre < %~
3. As an alternative, the complete UTexamination While im esugations are undctway, GE Nuclear Energy can be performed from the center drilled hole cf recommends that ownen of GE IlWRs take the the studs in accordance with the AShiE Code following acdons.
Case number N 3071. L Review the hardness and tensile strength of the 4. If ntplacement head studs are required, fabrica-( [Q \\ vessel studs to determine potential suscepdbility tion specifications should be in accordance wida to cracking, if tenslie or hardness data are not Regulatory Guide 1.65. ,, readily 1,vailable, field hardness tesu ng should be - considered. GE Nuclear Energy will furnish addidonal informa-don on uds subject to owners of GE IlWRs as il
~~2. In addidon to or as part of the normal AShtE becomes available.~~
Code Section Xi stud lSI program. conduct an ul- -TeclinicalSh0$c PQ J trasonic examination of at least five RPV head i 4 studs either during die next refueling outage or at (y ov.h 2 f. % the next available opportunity. 'Ihe studs chosen T. A. Caine for inspection should be those most suscepuble to SCC of all the studs in the head based on the , Product Rc/crenco? % > Idghest hardness either as measured or from the ChtTR or tensile strength values. *lhe following Il11 - Reactor Pressure Vessel recommendadons apply when a Code examina-JNotico % 1; tion is performed with a straight beam (0 degree L wave) from the end of the stud. This RICSIL pertains only to GH HWRs. GE Nuclear o Use a 3/4 inch to 1 inch diameter transducer Energy is distribudng uds RICSIL to all owners of with a frequency not less than 2,25 biliz and GE IlWRs. At the time of this distnbution, applica, not greater than 5.0 biliz. bility of the information contained in this RICSl! to specific GE IlWRs hs J not been defined. if, in the o increase lhe sensidvity of the examination to opinion of GE, subsequent acdon related to die O obtain a background noise level of about 5 information contained in this RICSIL is necessary, i V pago 2 RICSIt. No. OSS t
( Gl! willinfmu owners of Gl! IIWits. C 1 No warranty or representation expressed or implied is made with nespect to the axuracy, completeness or l' N usefulness of this informailon. General lilecuic Issued by J. G. Moore Company assumes no responsibility for liability or Custorner Service Communications Manager damage which may result fmm the use of this infor. Gli Nucicer Lacrgy mation. 175 Curtner Avenue, San. lose, CA 95125 M Pago 3 RICSil No. 055 l
dRF 4 \\ - MSS CENERAL E12CTRIC RICSIL No. 055 lipV lIl%D STUD CRACKING p ( RICSIL No. 055 documents the discovery of two cracked RPV studs at a domestic v IWR, and the subsequent notallurgical evaluation. To date, this is the only BWR that has reported RPV head stud cracking. This RICSIL has been reviewed and determined to be applicable to CNS. MCKWOUND The tactallurgical evaluation performed on one of the cracked RPV stude determined the cause of cracking to be transgranular stress corrosion cracking (SCC), initiating at corrosion pita in the thread roots below the first fully engaged thread in the RPV flange. The most likely cause for SCC is the combined effects of exposure of the studs to highly oxygenated water during refueling outages (the RpV studs are submerged during refueling, wetting the lower threaded portion of the studs) and proloading or tensioning of the stude during installat. ion of the RpV closure hood. The studs typically remain wet while under tension until plant startup when RPV temperatures increase and t.he trapped water is evaporaced. The cracked studs were made frots AIS1 4340 low alloy steel. In general. AISI 4000 series alloy steels (alloys 4130, 4140, 4340, 4340M) have increased susceptibility to SCC when heat treated to tensile strengths of 170 RSI or greater. As part of the metallurgical examination of one failed stud, a tans 11e test specimen and Charpy V notch specimen were removed from an outside threaded area adjacent to the crach. The tensile strength and Charpy V notch test values reported there were 180 RSI and 21 f t lba, respectively. Additional resourch of this issue included review of U.S. NRC Regulatory Guide A 1.65 entitled, "Ma t.c ri als and Ir. spec tions for Reactor Veasel Closure Studs" (d ) (released October 1973). This document recom>nends a maxitou n tens!!c strength of 170 RSI and a minimum Charpy V nouh energy of 45 f t lbs for RpV studs (10CPR50 ' Appendix C has the same minimum lirait for Charpy V notch innpact. onergy). These valuas are based, in part, upon susceptibility of low alloy steals to SCC. It rhould be noted, however, that the USAR for CNS acknowledges the ussi of RPV head studs with ininirnum Charpy V notch impact energ.ies of 30 st lbs. RICS1kNO. 055 INTERItLEQ,njtiERMHQJJji PecomatDdRLi.i!R No. 1: " Review the hardness and tensile strength ot t.he vessel studs to deterinino potentini susceptibility to cracking. If tensile or hardness data are not readily availabic, fic1J hardness testing should be considered." EN1 Enimnrail The certified material test reports (CMTRs) for the 52 CNS RPV studs and four spare studs in the warehouse have been obtained. The CNS RPV studs were made f roin 4340M alloy steel per specification ASME SA $40, grade B24. Tenstic strength, Charpy V notch impact energy, and hardness data for t.he bar stock used to taako these studs was reviewed. It should be noted that the fabrication records.do not identify which piece of bar stock was used to make which RPV stud. This also applies to t.ho fabrication records for the four spare RPV studs. CNS RPV stud nuinbers 1 21 and 26 are fabricated from inaterial most susceptible to stress corrosion cracking, based on their higher tensile strength and hardness values and lower Charpy V notch impact energies. l l page 1 of 4 t I i f
f l dRp _ ct t-410to 6 Tha uso of a p:rtchle bstdn2es tester to obtcin ep2cific hardness data for each CNS RPV stud has been carefully reviewed and is not recommended. The tirou required to /^'- obtain adequate hardness test data for all CNS RIT 6tuds ( ) would very likely exceed the time required to ultrason-N/ ically in pect these studs. Furthermore, factors such as surface coatings (or oxide layera), hardness test locations with respect to material thickness, and conversion of obtained portable hardness numbers to more meaningiul Brinnell or Rockwell scales make portable hardness testing of questionabic benefit. Reautred Actioni None. B.geoamenslation NA.2: "In addition to or as part of the normal ASME Codo Section XI stud ISI program, conduct an ultrasonic examination - of at least five RpV head studs either during the next refueling outage or at the next available opportunity. The studs chosen for inspection should be those most susceptible to SCC of all the studs in the head based on the highest hardness eitber as measured or from the CMTR or tensile strength values. The 'o11owing recommendations apply when a Code examination is perforined with a straight beam (0 degree L wave) fre-the end of the stud. 3/4 inch to 1 inch diameter transducer with Uso a a h equency not less than 2.25 Mllz and not greater than 5.0 Mlle. / \\ Increase the sensitivity of the examination to ( ) obtain a background noise level of about 5% of full screen height. Normally, thir. will require abont 20 dB above reference level, assuming calibti ; ion on a 3/S inch diameter flat bottomed hole. Record and evaluate any (ndications in accordance with the Code. In addition, any indication that is judged to be
~~suspect, regardless of amplitude, should be recorded and evaluated.~~
An ultrasonic examina-tion from the bore should be considered for sizing the crack indications if required. Studo with suspect indications shoulc be removed where possible and examined by inagnetic particle tess.ing for confirmation of the UT resultn." [ CNS Resonnse! As stated above, hardness values for each individual CNS RIT stud are not available. The mechanica14roperty data indicates that CNS RPV studs 1-21 and 26cande froin heat number 37385 would be the most susceptible to SCC, Therefore, these RIT studs will be UT examined during the 1991 Fall Refueling Outaga. Furthermore, to provide additional assurances of RPV stud integrity, the Code- ,-s required IST UT examination of RpV stud numbers 22 25 / j\\ , ( and 27-35 will also be performed during the 1991 Fall l page 2 of 4 L
( dW 9\\-/lOle 6 Refueling Outage. It is enttianted that the UT examina- [( ], tion of t !'e se studs will require approxiteately four j hours. This is actual refuel floor time with the RPV v head reinoved and vessel water Icvel just below the top of the vessel flange. This will affect the outage critical path schedule; however, this issue has been coordinated with CNS Outage and Medifications and steps will be taken to ininiinite the tirpact on outage critical path. UT exeninations of the CN$ RPV scuds will bo conducted in accordance with the requirements cf Section X1 of the ASME Code and the recommendations of RICS1h No. 0$$. All indications connidered to be suspect will be recorded and evaluated. UT exainination from the RPV ntud center bore hole inay be performed to determine the stee and depth of any UT indication evaluated as a possible crack. If eveltation of UT indication requires retcoval of an RPV stud, the stud will be magnetic particle tested (HT) to confirin the presence of
~~cracking, Beaufred AttirPJ Perforan UT exarnination of CNS RPV studs 1-35 during the 1991 Refueling Outage.~~
These examinations have been ac du1ed. g g ygg ; G,qi)l @, q, Recommendelion No. ?: "/.s an altarnative, the complete UT examination can be performed frorn the center-drilled hole of the studs in s accordance with the ASME Code Care No. N 307 1." CNS Retoonsej, As noted above, UT exarnination froin the center bore hole in accordance with ASME Code Case N 307 1 will ha considered as an alternative to the straight besta (0 degree L wayo) met. hod for sizing indications found by the straight heam method. This technique would require additional teachining of the CNS RPV stud UT calibration block and a special UT transducer. General Electric has been tequested to review the CMS RpV stud configuration to determino trensducer requirements and CNS RPV stud calibration block modifications. A decision will be ina.de after C.E. completes their review aa to whether capability to perform this supplemental exare will be needed. E. admired _As1Leni None. This is a supplemental exato not required by ASME Code. PecotmendationJo. 4: "If replacement head studs are required, fabrication specifications slould be in accordance with Regulatory cuide 1.65." CNS Response: CNS has four sparo RpV studa in stock. The certified material test reports (CMTRs) for these studs have been reviewed and although these studs do not meet the requirements of Regulatory cuido 1.65 for tensile [Q strength and Charpy V notch impact energy, they would be acceptable for use in accordance with the CNS USAR. l Page 3 of 4
l dtW-9\\-flOto$ l I l [ Since SCC appears to tr2ke a long Line to develop under favorable conditions (inat.erial/ sustained tensile at ress/ envirotunent) and these condit tons only exist s isuulta - neously for relatively short periods of tinie during outage conditions, theme studs could be used for at least one operating cycle. Tour additional sprte itPV studs have been ordered from AllB.Cotabus t ion 1:nsince ring. The purchase contract for these studs specifies that the tensile strength and charpy impact c.wrgy tocet the requircutent s of 14eculatory Guide 1.65. These studs are scheduled for delivery prior to the beginining of the 1991 Itefueling Outage. PaqttlttiLAtiliitti. None. 9 So 82 'E l h%l)_ c m~- & C1- (tj
~~t.. E 'llickel, J r.~~
~~M. 'J. Sponcce CNS ISI Engineer Englucering Programa Supervisor fly' - k.fI l S.> R f/ Ma 4 page 4 of 4 _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _}}~~

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