Similar Documents at Surry |
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Category:Drawing
MONTHYEARML20034F0532020-01-31031 January 2020 Proposed License Amendment Request: Allowed Outage Time Extension for Replacement of Reserve Station Service Transformer C 5KV Cables to Transfer Bus F Response to Request for Clarification and Additional Information ML19105B0552019-04-15015 April 2019 Superseded Pages Per Revision No. 01 to Environmental Sampling Program ML0501805082004-07-19019 July 2004 Associated System Drawings in Support of North Anna Power Station Units 1 & 2, and Surry Power Station Units 1 & 2, ASME Section XI Inservice Inspection Program End of Intervals System Pressure Testing ML0413302072004-05-0505 May 2004 Drawing 11548-FM-1E, Rev 10, Machinery Location Reactor Continment Sections A-A, X-X & F-F Surry Power Station - Unit 2. Sheet 1 of 1 ML0413302082004-05-0505 May 2004 Drawing 11548-FM-1F, Rev 9, Mach. Loc. Reactor Cont. Sections B-B. E-E. Y-Y & Z-Z Surry Power Station - Unit 2. Shet 1 of 1 ML0413302092004-05-0505 May 2004 Drawing 11548-FM-10, Rev 9, Mach Loc Reactor Cont Sections C-C & D-D Surry Power Station - Unit 2. Sheet 1 of 1 ML0414804032004-01-30030 January 2004 Drawing 11448-FE-7B, Rev 21, Wiring Diagram Annunciator (ANN-VSP) Surry Power Station - Unit 1. Sheet 1 of 1 ML0414804142002-05-29029 May 2002 Drawing 11448-FB-27A, Rev 16, Plumbing and Fire Protection Service Building Surry Power Station - Unit 1. Sheet 1 of 1 ML0414804152002-01-31031 January 2002 Drawing 11448-FB-25E, Rev 22, Ventilation & Air Conditioning Service Building - El. 9' - 6 Surry Power Station - Unit 1. Sheet 1 of 1 ML19095A2821978-02-22022 February 1978 Response to Request for Information on LOCA Pipe Break Inside Reactor Vessel Cavity ML19095A2841978-02-15015 February 1978 Presenting, Unit 2 Steam Generator Surveillance Program Outage Schedule & Information of Program ML19093B0471977-12-23023 December 1977 Request for NRC Review of Test Performance Data on Defense Apparel Air Hood & for Permission to Continue Using of Hood Indicated by VEPCO Letter of 12/6/1977 with Serial No. 558 ML19093B0821977-12-0101 December 1977 Response to a Letter of 09/14/1977 Regarding the Potential for a Boron Dilution Accident at Surry Unit Nos. 1 & 2 Due to Injection of Contents of Naoh Tank Into Reactor Coolant System ML19093A9671977-09-0101 September 1977 Letter Serial # 374 Dated 08/29/1977 Provided Information Regarding NPSH Available for Lead Safety Injection Pumps at Surry. Determined That Procedures Should Be Implemented to Throttle Valves in Lhsi Pump Discharge ML19098B4441977-06-28028 June 1977 Provide Requested Information Relative to Site Settlement to Close Previously Unresolved Item Identified as 76-15/1, Site Settlement Surveillance ML19095A5541977-01-11011 January 1977 Submit Licensee Event Report No. AO-S1-76-08 Re Broken Stud from 2A Steam Generator Girth Strap with Unit 2 ML19105A1811975-12-31031 December 1975 Submit Supplement to Initial Report Dated 12/4/1975, Which Responded to IE Bulletin 75-04B, as Requested ML19098B2971975-10-28028 October 1975 10/28/1975 Letter Supplemental Information in Technical Specification Change No. 33 2020-01-31
[Table view] Category:Letter
MONTHYEARIR 05000280/20244012024-02-0606 February 2024 Security Baseline Inspection Report 05000280/2024401, 05000281/2024401 and 07200002/2024401 IR 05000280/20230102024-01-18018 January 2024 Age Related Degradation Inspection Report 05000280-2023010 and 05000281-2023010 ML23312A1922024-01-18018 January 2024 Issuance of Amendment Nos. 316 & 316 Regarding a Risk Informed Approach for Tornado Classification of the Fuel Handling Trolley Support Structure ML24003A9022024-01-0303 January 2024 Stations, Units 1 and 2, Emergency Plan Revision - Relocation of the Technical Support Center (TSC) - Editorial Correction IR 05000280/20233012023-12-28028 December 2023 NRC Operator License Examination Report 05000280/2023301 and 05000281/2023301 ML23283A3052023-12-20020 December 2023 Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary) ML23352A3692023-12-18018 December 2023 Associated Independent Spent Fuel Storage Installation, Revision to Emergency Plan Report of Change ML23334A2342023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23242A2292023-11-0707 November 2023 Issuance of Amendment Nos. 315 and 315, Regarding Emergency Procedure Relocation of the Technical Support Center ML23200A2622023-11-0202 November 2023 Issuance of Amendment Nos. 314 and 314, Regarding Technical Specification and Spent Fuel Pool Criticality Change IR 05000280/20230032023-10-25025 October 2023 Integrated Inspection Report 05000280/2023003 and 05000281/2023003 ML23286A0372023-10-13013 October 2023 Annual Submittal of Technical Specifications Bases Changes Pursuant to Technical Specification 6.4.J ML23285A0922023-10-12012 October 2023 Inservice Testing Program for Pumps and Valves Sixth Interval Update and Associated Relief and Alternative Requests ML23277A1172023-10-0303 October 2023 Operator Licensing Examination Approval 05000280/2023301 and 05000281/2023301 ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23256A2672023-09-21021 September 2023 Review of the Fall 2022 Steam Generator Tube Inspection Report ML23244A2352023-09-18018 September 2023 Request to Use a Subsequent Edition of the ASME Code, Section XI ML23230A0012023-09-15015 September 2023 Request to Use a Provision of a Previous Edition of the ASME Boiler and Pressure Vessel Code, Section XI, for Non-Destructive Examinations IR 05000280/20220042023-09-0101 September 2023 Reissue - Surry Power Station - Integrated Inspection Report 05000280/2022004 and 05000281/2022004 ML23244A0142023-08-30030 August 2023 Inservice Inspection Owner'S Activity Report IR 05000280/20230052023-08-23023 August 2023 Updated Inspection Plan for Surry Power Station, Units 1 and 2 (Report 05000280/2023005 and 05000281/2023005) ML23226A1862023-08-10010 August 2023 Proposed License Amendment Request - Reclassification of Regulatory Guide 1.97 Variable for Low Head Safety Injection Flow Instrumentation ML23220A1492023-08-0707 August 2023 Independent Spent Fuel Storage Installation - Submittal of Cask Registrations for Spent Fuel Storage ML23220A1482023-08-0707 August 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Response to Request for Additional Information IR 05000280/20230022023-08-0404 August 2023 Integrated Inspection Report 05000280/2023002 and 05000281/2023002 ML23208A0922023-07-26026 July 2023 Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of Framatome ORFEO-GAIA and OORFE-NMGRID CHF Correlations in the Dominion Energy Vipre-D Computer Code Response ML23207A1102023-07-26026 July 2023 NRC Regulatory Issues Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations ML23198A0322023-07-18018 July 2023 Regulatory Audit Report Regarding Review of License Amendment to Modify Emergency Plan Staff Augmentation Times ML23193A9382023-07-18018 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML23173A0602023-07-11011 July 2023 Audit Summary for License Amendment Request Revision of the Emergency Plan to Support Relocation of the Technical Support Center ML23136B1392023-06-29029 June 2023 Issuance of Amendment No. 313 Administrative Changes to Subsequent Renewed Operating Licenses and Technical Specifications ML23178A1682023-06-26026 June 2023 2022 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML23152A0432023-06-22022 June 2023 Audit Plan License Amendment Requests to Change Emergency Plan Staff Augmentation Times ML23153A1732023-06-16016 June 2023 Correction to Amendment Nos. 346 & 286 Millstone, 294 & 277 North Anna, 311 & 311 Surry, and 225 Summer to Revise Technical Specifications to Adopt TSTF-554,Rev Reactor Coolant Leakage Requirement ML23163A2352023-06-12012 June 2023 Request to Use a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI, for the SPS Sixth Inservice Inspection (Isi) Intervals and VCSNS Fifth ISI Interval ML23151A6612023-05-31031 May 2023 ES-2.1-1, Surry 2023-301 Corporate Notification Letter (Revised) ML23167B0072023-05-31031 May 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Supplemental Information - LOCA Dose Calculation Summary ML23117A0952023-05-10010 May 2023 Audit Plan License Amendment Request to Support Relocation of the Technical Support Center ML23130A3742023-05-10010 May 2023 Request to Delay NRC Intial Licensing Examination 2024-02-06
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000281/LER-2020-001, Cancellation of LER 2020-001-00 for Surry Power Station Unit 2 Re Loss of Containment Cooling Affecting Containment Partial Pressure Indication2020-11-0606 November 2020 Cancellation of LER 2020-001-00 for Surry Power Station Unit 2 Re Loss of Containment Cooling Affecting Containment Partial Pressure Indication 05000280/LER-2019-0022020-02-0404 February 2020 LER 2019-002-00 for Surry Power Station Unit 1, Items Non-Conforming to Design for Tornado Missile Protection 05000280/LER-2017-0012017-10-0606 October 2017 1 OF 3, LER 17-001-00 for Surry, Unit 1, Regarding Shutdown due to an Unisolable Leak in Reactor Coolant Pressure Boundary 05000281/LER-2016-0012016-12-0202 December 2016 Unit 2 Reactor Trip due to Generator Differential Lockout, LER 16-001-00 for Surry Power Station, Unit 2, Regarding Reactor Trip Due to Generator Differential Lockout 05000280/LER-2016-0012016-07-11011 July 2016 1 OF 4, LER 16-001-00 for Surry Power Station, Unit 1 and 2, Regarding Emergency Service Water Pump Inoperable Due to Corrosion of Valve Support ML1015505662009-03-0404 March 2009 Event Notification for Surry on Relief Valve Failure Results in Tritium and Cesium Spill ML0702501902007-01-17017 January 2007 Special Report on Qpt Not Less than Two Percent for Twenty Four Hours ML19105B1281979-08-21021 August 1979 LER 1979-011-03X for Surry Units 1 & 2 Loss of Power with Both Units at Cold Shutdown, Observed Radiation Monitors Without Power ML19093B4141978-09-26026 September 1978 Reporting Condenser Cooling Water Outlet Temperature to James River Decreased by 3.5oF as Measured at Station Discharge Structure ML19105A3001978-09-0808 September 1978 LER 78-013-03X-1, LER 78-029-03L-0, & LER 78-031-03L-0 for Surry 2 Re Critical Below Minimum Insertion Limit on D Control Bank ML19105A3011978-09-0101 September 1978 LER 1978-028-03L-0 for Surry 2 Re Containment Pressure Exceeded Allowable Limit ML19095A4971978-08-25025 August 1978 Submit Licensee Event Report No. LER 78-024/03L-0 Re No Alarm During a Planned Release of Liquid Radioactive Waste ML19095A4981978-08-22022 August 1978 Submit Licensee Event Report Nos. LER 1978-019/03L-0, LER 1978-021/03L-0, LER 1978-022/03L-0, LER 1978-023/03L-0, LER 1978-025/03L-0, LER 1978-026/03L-0 for Surry Unit No. 1 ML19105A0771978-08-16016 August 1978 Submit LER 78-027/03L-0 Re Seven Snubbers Not Meeting Inspection Criteria ML19105A0781978-08-16016 August 1978 Submit LER 78-027/03L-0 Re Seven Snubbers Not Meeting Inspection Criteria ML19105A0791978-08-11011 August 1978 LER 78-026/03L-0. Follow-up Inspection for Valve Cleanliness, Discovered That Containment Spray Isolation Valves (MOV-CS-201C and D) Contained No More That Five Milliliters. Technical Specification TS 6.6.2.b.(3) ML19095A4991978-08-0404 August 1978 Sutmit Licensee Event Report No. LER 78-018/03L-0 Re Two Main Steam Safety Valves Were Not within Plus or Minus 3% of Setpoint as Required ML19105A0801978-07-28028 July 1978 LER 1978-022-03L-0 Re Axial Power Distribution Monitoring Was Exceeded,& Three LERs 1978-021-03L-0, LER 1978-023-03L-0, & LER 1978-024-03L-0 Re Inlet Circulating Water Temperature Exceeded ML19095A5001978-07-28028 July 1978 Submit Licensee Event Report Nos. LER 78-017/03L-0 Re Not Documenting Four Electric Maintenance Procedures & Two Periodic Tests, & 78-020/03L-0 Re Suspended Air Bubbles in Reservoir Fluid ML19095A2281978-07-27027 July 1978 Reporting & Providing Additional Information on Incidence of Missing Two Steam Generator Tube Plugs Found During Inspection ML19095A2291978-07-26026 July 1978 Reporting of Cooling Water Discharge Temperature Change Rate on 7/9/1978 & Found No Evidence of Any Detrimental Effects in River or Its Inhabitants ML19095A5011978-07-21021 July 1978 Submit Licensee Event Report No. LER 78-016/03L-0 Re Valve MOV-RH-100 Was Not Opening ML19105A0821978-07-17017 July 1978 LER 1978-025-03L for Surry Unit 2 Excessive Gap Around Battery Room Ventilation Duct ML19105A0811978-07-17017 July 1978 LER 1978-025-03L for Surry Unit 2 Excessive Gap Around Battery Room Ventilation Duct ML19095A5021978-07-0707 July 1978 Submit Licensee Event Report Nos. LER 78-013/03L-0 Re Six Snubbers Not Meeting Test Criteria, & 78-014/03L-0 Re Mount Plate Was Warped, & 3 of 8 Bolts Were Insufficiently Embedded ML19095A5031978-06-27027 June 1978 Submit Licensee Event Report No. LER-78-015/03L-0 Re Valve Cleanliness Inspection Found Foreign Material on Valve Seating Surfaces ML19095A5041978-06-27027 June 1978 Submit Licensee Event Report No. LER-78-015/03L-0 Re Valve Cleanliness Inspection Found Foreign Material on Valve Seating Surfaces ML19105A0831978-06-23023 June 1978 LER 1978-017-03L, LER 1978-018-03L, LER 1978-019-03L, & LER 1978-020-03L for Surry Unit 2 Related to Snubbers ML19095A2481978-06-19019 June 1978 06/19/1978 Letter Reporting Increase of Condenser Cooling Water Discharge ML19095A5051978-06-0707 June 1978 Submit Licensee Event Report No. LER-78-012/03L-0 Re Inside Recirculation Spray Pump 1-RS-P-1A, Failure to Rotate by Hand ML19105A0851978-05-23023 May 1978 LER 1978-015-03L for Surry Unit 2 Finding of Air Leak on Header of Control Room Emergency Air Supply Bank ML19095A5061978-05-23023 May 1978 Submit Licensee Event Report Nos. LER-78-008/03L-0, LER-78-009/03L-0, LER-78-010/03L-0, LER-78-011/03L-0 for Surry Unit No. 1 ML19105A0861978-05-12012 May 1978 LER 1978-013-03L for Surry Unit 2 Reactor Brought Critical Below Minimum Insertion Limit on D Control Bank ML19105A0871978-05-0505 May 1978 LER 1978-014-03L for Surry Unit 2 Charging Pump Service Water Pump a Tripped on High Current ML19095A5071978-05-0505 May 1978 Submit Licensee Event Report No. LER-78-007/03L-0 Re Leak on Body to Bonnet Joint Diaphragm Valve 1-CH-98 ML19105A0881978-04-19019 April 1978 LER 1978-006-01T for Surry Unit 2 Review of Fire Protection Yard Piping Revealed That Two Pipe Extensions May Not Have Been Seismically Qualified ML19095A5081978-04-19019 April 1978 Submit Licensee Event Report No. LER-78-006/01T-0 Re Seismically Disqualified Two Pipe Extensions on Fire Protection Yard Piping ML19095A5091978-04-0707 April 1978 Forwarding Licensee Event Report No. LER-78-005/01T-0 Re Westinghouse'S Error in LOCA-ECCS Evaluation Model, Which Will Result in Higher Calculated Peak Clad Temperature ML19105A0901978-04-0606 April 1978 LER 1978-012-01T for Surry Unit 2 Failure to Meet Containment Integrity ML19105A0921978-04-0404 April 1978 LER 1978-011-03L for Surry Unit 1 Loss of Service Water Flow Through Charging Pump Intermediate Seal Cooler ML19095A5101978-04-0404 April 1978 Submit Licensee Event Report No. LER-78-004/03L-0 Re Loss of Service Water Flow Through Charging Pump Intermediate Seal Cooler ML19105A0931978-03-30030 March 1978 LER 1978-008-03L Re Snubber Body in Contact with Structure; LER 1978-009-03L Re Trip of Boric Acid Transfer Pump; & LER 1978-010-03L Re Failure of Pressurizer Channel ML19095A5111978-03-28028 March 1978 Submit Licensee Event Report No. LER-78-003/03L-0 Re Failed Heat Tracing, Panel 8 Circuit 24 ML19095A5121978-03-21021 March 1978 Submit Licensee Event Report No. LER-78-002/01T-0 Re Additional Radial Peaking Factor Surveillance ML19105A0941978-02-27027 February 1978 LER 1978-006-03L Re Failure of Sola Transformer, & LER 1978-007-03L Re Trip of Boric Acid Transfer Pump on Thermal Overload ML19095A5131978-02-24024 February 1978 Submit Licensee Event Report No. LER-78-001/03L-0 Re Hydraulic Shock Suppressor Found Inverted with Empty Reservoir ML19105A0951978-02-16016 February 1978 LER 1978-003-03L Re Erroneous Readings of Loop a Hot Leg Temperature Protection Channel; LER 1978-004-03L Re Unqualified Welding of Two Diaphragm Welds on Heat Exchanger; & LER 1978-005-03L Re Low Current Condition from Heat Tracing ML19105A0971978-02-0202 February 1978 LER 1978-001-03L for Surry Unit 2 Inoperability of Condenser Circulating Water Outlet Valve ML19095A5141978-01-0909 January 1978 Submit Licensee Event Report Nos. LER-77-021/03L-0 & LER-77-022/03L-0 ML19105A0991978-01-0909 January 1978 LER 1977-021-03L Re Routine Sampling of Bast a Revealed Tank Was at Concentration of 11.5%, & LER 1977-022-03L Re Failure of Comparator to Trip During Routine Test of a Steam Generator Level Protection Channel II 2020-02-04
[Table view] |
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CONTROL BLO~K: I l. I I I I [PLEASE PAINT ALL REQUIRED INFORMATION]
1 6 LICENSEE LICENSE EVENT NAME LICENSE NUMBER TYPE TYPE
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lo I I I lIJ 57 58 59 w 60 I61 ols lo I-lo l2lalol 68 11 lo . I ol z I z I G74I LQ..l 69 75 1:Lo I 6 J1 lz80I EVENT DESCRIPTION lo'21 I With Unit No. 2 at cold shutdown, a broke*n stud from the 2A steam generator girth I 1 7 8 9 . 80 loj3j I strap was discovered while completinq a modification to the 2A steam generator. I 7 89 , 80 4
7 1°1 81 9I Further investigation* to the. llnit Na
mare cracked studs . 80I joj5!8 9!This event. is reportable .per Technical 7 .
Specification 6.6.2a(5). A similar problem I 80 lo!6j !was identified on Unit No. 1 (AO-S1~76-08). I 7 8 9 . 80 PRIME SYSTEM CAUSE COMPONENT COMPONENT CODE CODE COMPONENT CODE SUPPUER MANUFACTURER VIOLATION 7
@El 8 9 Is10I H I []J 11 12 IH I A IN .I G IE !R 17 I ~ IX l 9 l 9 ! 9 47l Ll!J 43 44 48 CAUSE DESCRIPTION
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7 8 9 IStress corrosion cracking caused by degradation of the coating on the stud in I 80 9
1°1 1 I 7 8 9 conjunction with
- high stresses developed d11ring the roao11factJJring of the st11ds 80 I
II@! !See attached report for additional information. I
- 7. 8 9 80 FACILITY METHOD OF STATUS % POWER OTHER STATUS DISCOVERY DISCOVERY DESCRIPTION 7
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N_,_/...,_A 80 PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION .
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13 . 60 PERSONNEL INJURIES NUMBER DESCRIPTION EE IO I OI 01 ~_:.aN/_A___________________________-,--_ _~
7 8 9 11 12 80 OFFSITE CONSEQUENCES
[lfil. N/A 7 8 9 80 LOSS OR DAMAGE TO FACILITY
. . TYPE . OESCRIPl"ION
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- 7. 8 9 10 80 PUBLICITY (ill} NA 7 8 9 ,80
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7 8 9 80 NAME: _ _ _ T.._y--nd_a__l;..;l;......;;;L'""'.'.....;;;.Ba;;::.;u:::.;c=o=m'----------- PHONE: (804) 357-3184
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e LICENSEE EVENT REPORT REPORT NO. AO-Sl-76-08 STRESS CORROSION'CRACKING'OF STUDS FOR STEAM.GENERATOR.RESTRAINT SYSTEM DECEMBER 3, 1976 DOCKET NOS. 50-280
... 50-281 LICENSE NOS. DPR-32 DPR-37 SURRY POr/ER STATION VIRGINIA ELECTRIC AND POWER COMPANY
I. INTRODUCTION In accordance with the requirements of Technical Specification 6.6.
2a(5) for Surry Power Station Operating License Numbers DPR-32 and DPR-37 this report describes a Reportable Occurrence which occurred on October 7, 1976. ,
The Directorate of Regulatory Operations, Region II, was notified October 21, 1976s, The occurrence described herein involved the discovery of cracked Vascomax studs on the upper restraint anchor system for Unit No. 2 Steam Generator. Because a similar problem was identified on Unit No. 1, both occurrences will be discussed in this report.
An occurrence involving the failure of Vascomax material at Surry Power Station, used in the steam generator lower ring support swivel end coupling_s, was discussed in AO-S2-74-09, dated December 12, 1974.
II.
SUMMARY
OF OCCURRENCE All supports in the reactor coolant system including the steam generator supports, are designed to withstand the design basis earthquake acting simultaneously with an instantaneously applied pipe break.
(The part numbers referenced below a're shown on station drawings 11448-FM-511 Rev. 6 and listed in the "Bill of Material". Refer to Figure 1).
The function of the subject studs (Item No. 7) is to couple together the steam generator .gir.th straps (Items 1 and 2) thereby forming a contin-uous ring. This coupling is accompUshed through two joint flanges (Item
- 3) on ~ach side of the steam generator. Two vessel girth straps make-up the upper restraint anchor system for each steam generator. Eight studs are required for each joint flange or a total of 32 studs are required for
each anchor system.
- I These studs are pretensioned across the joint flange spacer block (Item No. 9). 'I'he spacer block serves to reduce bending stresses in the studs.
A total of 9 machined shoe openings (Item No. 6) are welded to each vess*el girth strap (Items 1 and. 2 ). These shoe openings accommodate 9 keys (It.em Nos. 4 and 5) which themselves are fastened by dowels (Item No. 8)
- to the large upper restraint: casting which is shown on station drawing 11448-FM-51G as Item No. 7 (Figure 2). These key and shoe openings func-tion to allow vertical thermal expansion of the.steam generator within the I
upper restraint casting, but will restrict lateral movement resulting from forces generated following a main steam line break and/or seismic event.
The upper restraint casting is anchored to the containment floor through hydraulic snubbers.
In the event of a main steam pipe break and/or seismic event the shoe openings in the vessel girth straps act against the keys which result in a tangentic:1-l load on the.girth straps. The subject studs are designed to accommodate the maximum t~ngential load resulting from this accident con-dition *. Existing space restrictions and restraint design required a limita-tion on,stud size and quantity which necessitates the use *of .an ultra-high strength bolting material. Vascomax 350 CVM maraged alloy steel is the only material available with the strength requirements of the stud material.
On October 7, 1976, with Unit No. 2 at cold shutdown conditions, a broken Vascomax stud for 2A steam generator girth strap was discovered. The insulation*. had been removed from around the upper restraint in order to loosen the girth strap to complete a modification *to the 2A ste*am
- generator when the broken stud was found.
Furth~r*investigation of all Vascomax studs on the upper.restraints of the steam generators revealed.more failed studs.
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e* eI III. ANALYSIS OF THE OCCURRENCE Three studs from the Unit No. 2 steam gener~tor upper restraint system were sent to an independent testing laboratory to determine the cause of the stud failure. One of the studs was in two pieces with the nut attached to
. the smaller piece an4 the. other two were intac't.
Visual examination of the threaded areas showed cracks in the threads of the intact studs which were confirmed by flourescerit dye~penetrant exami-nation. The small piece* from the failed stud also had a crack.two to six threads up towards the slot. Examination of the fracture surface showed an apparent "initiator" crack, delineated with a ring of blue oxide, with a crack propagating from this initiator crack. The fracture surface was descaled, *and examined by a Scanning Electronic Microscope (SEM) and the morphology was found to be mixed intergranular and transgranular.
The threaded end from an intact stud was metallographically*examined.
The crack indicated by the visual and PT examination extended from the fourth thread up from the base of the slot, at an angle of approximately 30 degrees from tl:ie horizontal.* The crack had propagated* to within 1/32 inch of the
- slot. Metallography showed cracks from the main crack; the: morphology was
- mixed tr~nsgranula; and intergranular~ The threaded end from the othe~ intact stud which showed visual and PT -cracks was immersed in liquid nitrogen and deliberately fractured. The fracture surface showed that the crack had almost propagated completely*through the cross-section.
- ... These results, together with the k'nownproperties of the 350 *grade niarag-irig steels. and. the en:~i;omnent i~ which these studs operate; indicate that
~ . . . .
1 the ~rack propagates bra stress corr~sion.mecha~ism.. Although, the maraging steels have higher fracture toughness and greater resistance to-stress corro-sion cracking than other steel heat treated to comparable strengths, the 350-
.. e grade maraging steel is ve~y susceptible to stress corrosion cracking. The presence of the branching cracks, together with a lack of striat~ons vis"".
ible in the SEM, indicate that fatigue is not the propagation mechanism.
Stress corrosion cracks were induced by the degradation of the protective coating (Heresite) combined with the stress intensification of the thread root developed during manufacture. The degradation of the (Heresite) coating on these studs was attributed to the high temperature in the con-.
fined space between the insulation and the steam generator.
IV *. CORRECTIVE ACTION TO PREVENT RECURRENCE
- 1. The replacement studs are H; in.-12UNF-2A Vascomax 18% Ni, maraging grade 350, 326,000 psi minimum yield strength, with nickel cadmium coating.
Stud t4reads were rolled P!ior to maraging. The design of the replace-ment studs and nuts has been modified to minimize stress concentration during manufacture and a coating was added to protect the studs from the environment (Refer to Figure 3). The* following modifications to the stud have been made:
- a. The stud neck diameter hasbeen reduced to the diameter of the thread root to allow a greater proportion of the strain to be taken up in the neck of the stud. This change relieves the strain in .the threaded parts of the stud.
- b. The thread crests and roots have been rounded and the stud has been polished to .a fine finish, (a 32 RMS finish or better) to minimize points at which stresses can concentrate *
. .' c. The stud threads, were rolled rather than cut to lessen. residual stress concentration from fabrication.
- d. The studs were given a protective coating. The coating is a diffused
_4;..
.. - eI nickel cadmium plating performed to Aerospace Material Specification AMS-2416E. The first coat consists of nickel, electrodeposited from a solution of nickel sulfamate solution to a thickness of 0.0002" - 0.0004" followed by a coat of cadmium, electrodeposited fr0m a solution of cadmium cyanide to a thickness of 0.0002" - 0.0004".
The total coating thickness is approximately .0005" to allow proper thread fit.
- e. After plating, the two coatings were diffused by a heating process to 360 degrees F for at least 30 minutes.
- 2. The replacement nuts are 1\ in.-12UNF-3B Vascomax 18% Ni, maraging grade 250, 150,000 psi minimum yield strength~ with Helicoil inserts. Th~
following modifications to the nuts were made: (Refer to Figure 4)
- a. The_ nuts are 50% deeper than standard castellated nuts to provide greater thread surface.
- b. The nuts have *a groove machined in the mating surface concentric with the axis of the nut. Th~s groove makes the region of the nut more flexible and acts to more equally apportion loads to the t,hreads.
This eliminates.a high load concentration on the first few threads of*
.*nut engagement.
- c. The nuts have Helicoil inserts. The inserts also provide a better distribution of thread loading.
- 3. The studs were installed with a moderate preload stress level below the critical crack size for stress corrosion crack porpagation for this .material.
- 4. All studs (192) and .nuts (384) were replaced on'*the*support rings of**the.
generators on both units.
V. ANALYSIS AND EVALUATION OF SAFETY*IMPLICATIONS OF THE OCCURRENCE The upper restraint anchor system discussed herein does not carry any
operational loads and .do not support the steam generator. They serve only to restrain the steam generator in the improbable event of a pipe break of a feedwater, steam or reactor coolant system pipe simultaneously with a design basis seismic event as defined in the Final Safety Analysis Report.
The con'ditions described above did not exist and there was no requirement to utilize the upper restraint anchor system. Therefore, there were no safety implications associated with this occurrence.
VI. CONCLUSIONS
- 1. From the preliminary data received from an independent testing laboratory, it has been concluded that the degradation of the studs was caused by stress corrosion cracking.
- 2. The stress corrosion cracks were induced by the degradation of the pro-tective coating (heresite) combined with the stress intensification at the thread root developed during manufacture and the excessive preload stresses applied during initial installa.tion.
- 3. The design of, the replacement studs and nuts has been modified to minimize stress concentration during manu.facture. A nickel and cadmium coating ha.s been specified* to protect the studs from the .environment.
- 4. The studs were installed with a moderate preload stress that is below the critical crack size for the Vascomax 350.
- 5. The occurrence described herein did not affect the health or safety of the general public.
- 6. The occurren~e described herein did not affect the .safe operation of the station ...
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