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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20211B8931999-08-17017 August 1999 Safety Evaluation Supporting Amend 143 to License NPF-2 ML20210T2161999-08-0606 August 1999 Draft SE Supporting Proposed Conversion of Current TS to ITS for Plant ML20206L4551999-05-10010 May 1999 Safety Evaluation Supporting Amends 142 & 134 to Licenses NPF-2 & NPF-8,respectively ML20206G7471999-05-0404 May 1999 Safety Evaluation Accepting Corrective Actions Taken by SNC to Ensure That Valves Perform Intended Safety Functions & Concluding That SNC Adequately Addressed Requested Actions in GL 95-07 ML20203J0631999-02-19019 February 1999 Safety Evaluation Supporting Amends 107 & 85 to Licenses NPF-2 & NPF-8,respectively ML20199D8611999-01-12012 January 1999 SER Accepting Relief Request for Inservice Insp Program for Plant,Units 1 & 2 ML20198S4091999-01-0606 January 1999 Revised SE Supporting Amends 140 & 132 to Licenses NPF-2 & NPF-8,respectively.Page Contains Vertical Lines to Indicate Areas of Change ML20195E2281998-11-16016 November 1998 Safety Evaluation Authorizing Relief Request for Second 10-year ISI Program Relief Request 56 for Plant,Unit 1 ML20155H4801998-11-0303 November 1998 Safety Evaluation Supporting Amends 139 & 131 to Licenses NPF-2 & NPF-8,respectively ML20155E0271998-10-29029 October 1998 SER Approving & Denying in Part Inservice Testing Program Relief Requests for Plant.Relief Requests Q1P16-RR-V-3 & Q2P16-RR-V Denied Since Requests Do Not Meet Size Requirement of GL 89-04 ML20154B6121998-10-0101 October 1998 Safety Evaluation Granting Second 10-year ISI Requests for Relief RR-13 & RR-49 Through RR-55 for Jm Farley NPP Unit 1 ML20237C5471998-08-20020 August 1998 Suppl to SE Re Amends 137 & 129 to Licenses NPF-2 & NPF-8, Respectively.Se Being Supplemented to Incorporate Clarifications/Changes & Revise Commitment for Insp of SG U-bends in Rows 1 & 2 for Unit 2 Only ML20236U6141998-07-23023 July 1998 Safety Evaluation Authorizing Use of Alternative Alloy 690 Welds (Inco 52 & 152) as Substitute for Other Weld Metal ML20236R8671998-07-0909 July 1998 Safety Evaluation Concluding That Southern Nuclear Operating Co USI A-46 Implementation Program Has Met Purpose & Intent of Criteria in GIP-2 & Staff SSER-2 on GIP-2 for Resolution of USI A-46 ML20236L2451998-07-0707 July 1998 Safety Evaluation Accepting Licensee Request for Exemption from Certain Requirements of 10CFR50.71(e)(4) Re Submittal of Revs to UFSAR for Facility Changes Made Under 10CFR50.59 for Plant,Units 1 & 2 ML20217P0571998-04-29029 April 1998 Safety Evaluation Supporting Amends 137 & 129 to Licenses NPF-02 & NPF-08,respectively ML20217D2591998-04-21021 April 1998 Safety Evaluation Accepting Licensee Proposed Alternative Re Augmented Exam of Reactor Vessel Shell Welds for Plant ML20217H3191998-03-31031 March 1998 Safety Evaluation Accepting Proposed Changes to Plant Matl Surveillance Programs ML20217D4081998-03-24024 March 1998 Safety Evaluation Accepting Proposed Changes to Maintain Calibration Info Required by ANSI N45.2.4-1972 ML20216H6731998-03-17017 March 1998 SER Accepting Quality Assurance Program Description Change for Joseph M Farley Nuclear Plant,Units 1 & 2 ML20203B4731998-02-0505 February 1998 Safety Evaluation Supporting Amends 135 & 127 to Licenses NPF-2 & NPF-8,respectively ML20203B7581998-02-0505 February 1998 Safety Evaluation Supporting Amends 134 & 126 to Licenses NPF-2 & NPF-8,respectively ML20199F8931998-01-23023 January 1998 Safety Evaluation Supporting Amends 133 & 125 to Licenses NPF-02 & NPF-08,respectively ML20199B0371998-01-22022 January 1998 SER Accepting Request for Relief (RR-27) for Plant,Units 1 & 2 from Certain Provisions of Section XI to ASME Boiler & Pressure Vessel Code.Relief Will Remove Insulation on ASME Code Class 1 Sys During Inservice Insp ML20198R5221997-10-29029 October 1997 Safety Evaluation Supporting Amends 132 & 124 to Licenses NPF-02 & NPF-08,respectively ML20212F4791997-10-23023 October 1997 Safety Evaluation Supporting Amend 131 to License NPF-2 ML20212A5711997-10-17017 October 1997 Safety Evaluation Supporting Amends 130 & 123 to Licenses NPF-2 & NPF-8,respectively ML20217E8501997-10-0101 October 1997 Safety Evaluation Supporting Amends 129 & 122 to Licenses NPF-2 & NPF-8,respectively ML20216G9521997-09-0404 September 1997 Safety Evaluation Authorizing Request for Relief for IEEE 279-1971,Section 4.7.3 Requirements Concerning Steam Generator Water Level Control ML20236N3331997-08-21021 August 1997 SER Re Request for Interpretation of EDG TS 4.8.1.1.2.e for Farley Nuclear Plant,Units 1 & 2 ML20140B8571997-06-0202 June 1997 Corrected Page 7 to Safety Evaluation Supporting Amend 124 to License NPF-2 ML20140A8471997-05-29029 May 1997 Errata to Safety Evaluation Supporting Amends 123 & 118 to Licenses NPF-2 & NPF-8,respectively.Corrects Pp 2 of Subj Safety Evaluation ML20137D6961997-03-24024 March 1997 Safety Evaluation Supporting Amend 124 to License NPF-2 ML20137E2951997-03-24024 March 1997 Safety Evaluation Supporting Amends 125 & 119 to Licenses NPF-2 & NPF-8,respectively ML20137B4371997-03-20020 March 1997 SER Accepting Request for Relief for 120-month Update of Facility Inservice Insp & Inservice Testing Programs & Code Addition & Addenda of Asme/Ansi Parts 6 & 10 ML20135E4811997-03-0404 March 1997 Safety Evaluation Accepting Implementation of 10CFR50.55a Requirements Related to Repair & Replacement Activities for Containment at Plant ML20134F3721997-02-0303 February 1997 Safety Evaluation Supporting Amends 123 & 118 to Licenses NPF-2 & NPF-8,respectively ML20129K3781996-11-20020 November 1996 Safety Evaluation Supporting Amend 117 to License NPF-8 ML20129A9361996-10-11011 October 1996 Safety Evaluation Supporting Amend 115 to License NPF-8 ML20117H4861996-09-0303 September 1996 Safety Evaluation Supporting Amends 121 & 113 to Licenses NPF-2 & NPF-8,respectively ML20115J9481996-07-19019 July 1996 Safety Evaluation Supporting Amends 120 & 112 to Licenses NPF-2 & NPF-8,respectively ML20112B0741996-05-20020 May 1996 Safety Evaluation Supporting Amend 110 to License NPF-08 ML20095J2561995-12-0808 December 1995 Safety Evaluation Supporting Amends 118 & 109 to Licenses NPF-2 & NPF-8,respectively ML20092N1841995-09-28028 September 1995 Safety Evaluation Supporting Amends 116 & 108 to Licenses NPF-2 & NPF-8,respectively ML20084M0401995-05-22022 May 1995 Safety Evaluation Supporting Amends 115 & 107 to Licenses NPF-2 & NPF-8, Respectively ML20082G4311995-04-0707 April 1995 Safety Evaluation Supporting Amends 114 & 105 to Licenses NPF-2 & NPF-8,respectively ML20081H5101995-03-20020 March 1995 Safety Evaluation Supporting Amends 112 & 103 to Licenses NPF-2 & NPF-8,respectively ML20081J6481995-03-20020 March 1995 Safety Evaluation Supporting Amends 113 & 104 to Licenses NPF-2 & NPF-8,respectively ML20080N3461995-03-0101 March 1995 Safety Evaluation Supporting Amends 112 & 103 to Licenses NPF-2 & NPF-8,respectively ML20080D7111994-12-28028 December 1994 Safety Evaluation Supporting Amends 111 & 102 to Licenses NPF-2 & NPF-8,respectively 1999-08-06
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217P0761999-10-0606 October 1999 Non-proprietary, Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs ML20217G0361999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20212E7451999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Hcgs,Unit 1.With Summary of Changes,Tests & Experiments Implemented During Aug 1999.With ML20216E4941999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Jmfnp.With ML20211B8931999-08-17017 August 1999 Safety Evaluation Supporting Amend 143 to License NPF-2 ML20210T2161999-08-0606 August 1999 Draft SE Supporting Proposed Conversion of Current TS to ITS for Plant ML20211B2011999-08-0404 August 1999 Informs Commission About Results of NRC Staff Review of Kaowool Fire Barriers at Farley Nuclear Plant,Units 1 & 2 & Staff Plans to Address Technical Issues with Kaowool & FP-60 Barriers ML20210R6031999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With 05000364/LER-1999-001-05, :on 990621,plant Was Manually Tripped Due to Decreasing Vacuum in Condenser.Caused by Broken Steam Dump Drain Line.Broken Section of Line Was Repaired & Licensee Will Implement Addl Design Change1999-07-0202 July 1999
- on 990621,plant Was Manually Tripped Due to Decreasing Vacuum in Condenser.Caused by Broken Steam Dump Drain Line.Broken Section of Line Was Repaired & Licensee Will Implement Addl Design Change
ML20209G0661999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20196J3791999-06-30030 June 1999 Safety Evaluation of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs. Rept Acceptable L-99-267, Monthly Operating Repts for June 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With 05000348/LER-1999-002-02, :on 990527,Unit 1 Reactor Trip Occurred Following Loss of 1A SG Feedwater Pump.Caused by Personnel Error.Unit Was Stabilized in Hot Standby.With1999-06-25025 June 1999
- on 990527,Unit 1 Reactor Trip Occurred Following Loss of 1A SG Feedwater Pump.Caused by Personnel Error.Unit Was Stabilized in Hot Standby.With
L-99-023, Monthly Operating Repts for May 1999 for Jfnp Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Jfnp Units 1 & 2. with ML20206L4551999-05-10010 May 1999 Safety Evaluation Supporting Amends 142 & 134 to Licenses NPF-2 & NPF-8,respectively ML20206G7471999-05-0404 May 1999 Safety Evaluation Accepting Corrective Actions Taken by SNC to Ensure That Valves Perform Intended Safety Functions & Concluding That SNC Adequately Addressed Requested Actions in GL 95-07 ML20206C9461999-04-30030 April 1999 1:Final Cycle 16 Freespan ODSCC Operational Assessment ML20204D4391999-03-31031 March 1999 Unit-1 1999 Voltage-Based Repair Criteria 90-Day Rept L-99-161, Monthly Operating Repts for Mar 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20204D7271999-03-15015 March 1999 ISI Refueling 15,Interval 2,Period 3,Outage 3 for Jm Farley Nuclear Generating Plant,Unit 1 ML20207M6421999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20203J0631999-02-19019 February 1999 Safety Evaluation Supporting Amends 107 & 85 to Licenses NPF-2 & NPF-8,respectively ML20203A2651999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Jm Farley Nuclear Plant,Units 1 & 2.With 05000348/LER-1998-008-01, :on 981223,reactor Vessel Support Concrete Design Bases Temperature Exceeded Due to Closed Cooling Damper.Caused by Personnel Error.Damper Opened & Secured in Position.With1999-01-18018 January 1999
- on 981223,reactor Vessel Support Concrete Design Bases Temperature Exceeded Due to Closed Cooling Damper.Caused by Personnel Error.Damper Opened & Secured in Position.With
ML20199D8611999-01-12012 January 1999 SER Accepting Relief Request for Inservice Insp Program for Plant,Units 1 & 2 ML20198S4091999-01-0606 January 1999 Revised SE Supporting Amends 140 & 132 to Licenses NPF-2 & NPF-8,respectively.Page Contains Vertical Lines to Indicate Areas of Change ML20199E6591998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20206C8081998-12-31031 December 1998 Alabama Power 1998 Annual Rept 05000348/LER-1998-007-02, :on 981106,found Several TSP Circumferential Indications & Several TSP Axial Indications Extending Just Beyond Edge of Tsp.Caused by Tube Defects.Sg Tubes Have Been Plugged or Repaired as Required1998-12-22022 December 1998
- on 981106,found Several TSP Circumferential Indications & Several TSP Axial Indications Extending Just Beyond Edge of Tsp.Caused by Tube Defects.Sg Tubes Have Been Plugged or Repaired as Required
05000348/LER-1998-006-03, :on 981124,PRF Sys Suction Damper Was Outside Design & Licensing Basis.Caused by 1976 Personnel Failure to Identify Noted Inconsistency.Scheduled Design Change in 1999 to Modify Dampers to Ensure Licensing Basis Are Met1998-12-18018 December 1998
- on 981124,PRF Sys Suction Damper Was Outside Design & Licensing Basis.Caused by 1976 Personnel Failure to Identify Noted Inconsistency.Scheduled Design Change in 1999 to Modify Dampers to Ensure Licensing Basis Are Met
ML20198K4091998-12-18018 December 1998 COLR for Jm Farley,Unit 1 Cycle 16 05000364/LER-1998-007-01, :on 981116,ESF Actuation Occurred During DG 1000 Kw Load Rejection Test.Caused by Poor Jumper Electrical Connection.Improved Jumpers Will Be Used on Appropriate Terminals in Load Rejection Test Procedures.With1998-12-11011 December 1998
- on 981116,ESF Actuation Occurred During DG 1000 Kw Load Rejection Test.Caused by Poor Jumper Electrical Connection.Improved Jumpers Will Be Used on Appropriate Terminals in Load Rejection Test Procedures.With
ML20198B2561998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20195E2281998-11-16016 November 1998 Safety Evaluation Authorizing Relief Request for Second 10-year ISI Program Relief Request 56 for Plant,Unit 1 05000348/LER-1998-005-03, :on 981021,automatic Start of B Train Penetration Room Filtration Occurred Due to Filling Sf Tc. Caused by Inadequate Procedure.Changed Procedure to Provide Specific Guidance for Filling SFP Tc.With1998-11-12012 November 1998
- on 981021,automatic Start of B Train Penetration Room Filtration Occurred Due to Filling Sf Tc. Caused by Inadequate Procedure.Changed Procedure to Provide Specific Guidance for Filling SFP Tc.With
ML20155H4801998-11-0303 November 1998 Safety Evaluation Supporting Amends 139 & 131 to Licenses NPF-2 & NPF-8,respectively ML20195C9681998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20155E0271998-10-29029 October 1998 SER Approving & Denying in Part Inservice Testing Program Relief Requests for Plant.Relief Requests Q1P16-RR-V-3 & Q2P16-RR-V Denied Since Requests Do Not Meet Size Requirement of GL 89-04 ML20154B6121998-10-0101 October 1998 Safety Evaluation Granting Second 10-year ISI Requests for Relief RR-13 & RR-49 Through RR-55 for Jm Farley NPP Unit 1 ML20154H6001998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20154H0121998-09-30030 September 1998 Submittal-Only Screening Review of Farley Nuclear Plant IPEEE (Seismic Portion) ML20151V8341998-09-30030 September 1998 Non-proprietary Rev 2 to NSA-SSO-96-525, Jm Farley Nuclear Plant Safety Analysis IR Neutron Flux Reactor Trip Setpoint Change 05000348/LER-1998-004-03, :on 980909,turbine Trip & Consequent Reactor Trip Occurred.Caused by Reactor Protection Sys Card Failure. Failed Card Was Replaced & Unit 1 Returned to Power on 980910.With1998-09-28028 September 1998
- on 980909,turbine Trip & Consequent Reactor Trip Occurred.Caused by Reactor Protection Sys Card Failure. Failed Card Was Replaced & Unit 1 Returned to Power on 980910.With
05000348/LER-1998-003-04, :on 980816,determined That Wgdt Hydrogen & Oxygen Had Exceeded Concentration Limits,Per TS 3.11.2.5. Caused by Undetected Leak.Leak Was Identified & Isolated & Waste Gas Sys Was Returned to Svc on 980818.With1998-09-11011 September 1998
- on 980816,determined That Wgdt Hydrogen & Oxygen Had Exceeded Concentration Limits,Per TS 3.11.2.5. Caused by Undetected Leak.Leak Was Identified & Isolated & Waste Gas Sys Was Returned to Svc on 980818.With
05000348/LER-1997-003, :on 970315,determined That TS SR 4.5.3.2 Had Not Been Performed,Per Operating Procedure.Caused by Personnel Error.Verified That RHR Discharge to Charging Pump Suction MOVs 8706A & 8706B Were Closed.With1998-09-0808 September 1998
- on 970315,determined That TS SR 4.5.3.2 Had Not Been Performed,Per Operating Procedure.Caused by Personnel Error.Verified That RHR Discharge to Charging Pump Suction MOVs 8706A & 8706B Were Closed.With
05000348/LER-1998-005, :on 980315,failure to Perform Nuclear Instrumentation Surveillance Requirements Prior to Mode 2 & 3 Entry,Was Discovered.Caused by Personnel Error.Revised Applicable Procedures1998-09-0303 September 1998
- on 980315,failure to Perform Nuclear Instrumentation Surveillance Requirements Prior to Mode 2 & 3 Entry,Was Discovered.Caused by Personnel Error.Revised Applicable Procedures
05000348/LER-1998-002-05, :on 980816,SG Tube Leakage Investigation,Repair & Evaluation,Occurred.Caused by ODSCC in Two Locations on Same Tube.Operational Leak Rate Limit Requiring Plant Shutdown Has Been Administratively Reduced1998-09-0303 September 1998
- on 980816,SG Tube Leakage Investigation,Repair & Evaluation,Occurred.Caused by ODSCC in Two Locations on Same Tube.Operational Leak Rate Limit Requiring Plant Shutdown Has Been Administratively Reduced
ML20197C8991998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Jm Farley Nuclear Plant,Units 1 & 2.With ML20237C5471998-08-20020 August 1998 Suppl to SE Re Amends 137 & 129 to Licenses NPF-2 & NPF-8, Respectively.Se Being Supplemented to Incorporate Clarifications/Changes & Revise Commitment for Insp of SG U-bends in Rows 1 & 2 for Unit 2 Only ML20237B1891998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Jm Farley Nuclear Plant,Units 1 & 2 1999-09-30
[Table view] |
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST TO USE ALTERNATIVE MATERIALS IN THE FABRICATION AND INSTALLATION OF STEAM GENERATORS SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.
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i JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated May 22,1998, the Southern Nuclear Operating Company, Inc. (SNC) requested approval under the provisions of Title 10 of the.C. ode of Federal Regulations (10 CFR)
Section 50.55a(a)(3) to use the American Society of Mechanical Engineers (ASME)Section IX
]
Code Cases 2142-1 and 2143-1 during the impending fabrication and installation of replacement steam generators (SGs) for the Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2.
These two Code Cases introduce and classify new nickel base weld metals that closely match and are intended for welding Alloy 690. Code Case 2142-1 establishes welding classifications and other requirements for a bare wire filler metal. Code Case 2143-1 establishes welding classifications and other requirements for a coated electrode.
The subject Code Cases were approved by the ASME on June 5,1995, and published in the 1995 Edition of the ASME Boiler and Pressure Vessel Code (ASME Code), Code Cases Supplement No.1. However, since the 1995 Edition of the ASME Code has not been incorporated by reference into the regulations, these Code Cases cannot be used without prior NRC review and approval.
SNC intends to use Alloy 690 tubing and components in the fabrication and installation of i
replacement SGs for FNP, Units 1 and 2. SNC believes that use of the new weld metals will enhance the service life of the replacement SGs. Industry studies indicate that these new weld metals are less susceptible to intergranular stress corrosion cracking (IGSCC) than the other nickel based weld metals currently applied.
I Use of Code Cases 2142-1 and 2143-1 is advantageous to SNC because it eliminates the I
burden of requiring qualification of separate welding procedures for each weld metal, as is the case for non-Code welding materials.
l 9807300346 980723 PDR ADOCK 05000349 P
PDR
l l Thus, this relief request incorporates two issues:
- 1. Use of Alloy 690 type weld metals in Code Class 1 construction, and,
- 2. The use of two ASME Code Cases that group the new weld metals in the same welding categories as other commonly employed nickel based weld metals. This allows the use of appropriate existing welding procedures and performance qualifications with the new weld metals.
Section 50.55a(a) requires that systems and components of boiling and pressurized water cooled nuclear power reactors must meet the requirements of the ASME Code as specified in 10 CFR 50.55a(b) through (g) and that components be designed and fabricated to qualify standards commensurate with their safety function. Section 50.55a(3) states that alternatives to the requirements of 50.55a(c) through (h) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 DISCUSSION Due to the extensive history of IGSCC problems in Alloy 600, the industry has sought an attemative alloy. Currently, Alloy 690 is the industry material of choice. This choice is the result of numerous laboratory studies that show that Alloy 690 has little or no susceptibility to IGSCC in environments that simulate pressurized water reactor (PWR) and boiling water reactor (BWR) plant conditions. The NRC staff's review of the laboratory test results has resulted in the staff position that, based upon the available technical evidence, there is no technical reason to disallow the use of Alloy 690 base materialin nuclear plant construction.
Alloy 600 type weld metals (such as inco 82 and 182) were widely used during the construction of nuclear power plants. Operating experience showed that inco 182 was also susceptible to IGSCC, although primarily in BWR environments. Weld metals matching Alloy 690 have also been tested in simulated PWR and BWR environments. Commercial development of these weld metals lagged that of the Alloy 690 base metal. No matching weld metal has been commercially available until now. Thus, staff evaluation of these weld metals has not been previously conducted.
Corrosion studies examing the susceptibility of weld metals to IGSCC in SG environments are scant compared to the voluminous base metal studies. This is because the base metal performance is a strong indicator of the expected performance of a matching weld metal.
Results of the principle study, which included weld metals, are found in the Electric Power Research Institute (EPRI) report NP-5882M, titled, " Stress Corrosion Cracking Resistance of Alloys 600 and 690 and Compatible Weld Metals in BWRs." Two experimental Alloy 690 weld metals were tested and found to be immune to IGSCC in pure water environments. However, since these were laboratory simulations of a BWR environment, the results are only an indicator, and not a guarantee, of the weld metals' performance in a PWR environment.
4 in the EPRI report, the designations R-127 and R-135 were used for the experimental weld
. metals. These were the inco designations for the developmental weld metals that became i
inco 52 and 152, respectively. Inco 52 is the commercial filler metal (tig wire) described in i
ASME Code Case 2142-1. Inco 152 is the coated electrode described in Code Case 2143-1.
)
Another paper, *lnconel 690: A New High Nickel Alloy for Corrosive Environments at Elevated Temperature," by A. J. Sedriks, et al., of the inco Research and Development Center, included tests of a matching filler metalin a wide variety of environments. The two most interesting tests were conducted in simulated SG environments: desersted ammoniated and borated water at 316' C. The test results showed the welds and weld metal were highly resistant to gw.sral corrosion.
SCC susceptibility was tested by exposing welds to a variety of chloride environments. The controls used in these tests were Alloy 800 (not 600) and type 304 stainless steel Both of these alloys are known to crack in elevated temperature chloride environments. in all cases, Alloy 690 was tested for periods significantly longer than the time to crack Alloy 800 (the more resistant of the two control alloys), in no case did the Alloy 690 welds crack despite test i
durations 8 times longer than that of the control alloys.
Additional testing for IGSCC susceptibility in pure water environments was conducted. Another group of Alloy 690 welds plus control alloys were exposed to aerated water at elevated temperatures in the presence of a crevice.t Cracking was readily initiated within the controls.
.None of the Alloy 690 welds cracked despite testing durations 24 times longer than for Alloy 600 and 12 times for Alloy 800 and type 304 stainless steel. Thus, use of the Alloy 690 welds provides reasonable assurance of structuralintegrity.
Code Case 2142-1 (inco 52) lists the American Welding Society (AWS) specification AWS A5.14 and Unified Numbering System (UNS) designation UNS N06052 for a filler metal similar to inco 52. The Code Case establishes the F No. of this weld metal as F-No. 43 for both procedure and performance qualification purposes. Code Case 2143-1 (inco 152) lists appropriate AWS and UNS specifications for a coated electrode matching inco 152 and establishes F-No. 43 for this material for welding purposes. By this wt of specifications and F No. assignments, these materials are completely described for welding purposes as similar to other presently used nickel-based weld metals. Thus, existing welding procedures may be i
used with the inco S2 or 152 type weld metals without requalification.
3.0 CONCLUSION
The staff concludes that, based upon the available technical evidence, use of Alloy 690 welds (Inco 52 and 152) as a substitute for other weld metals as requested by SNC in its May 22, 1998, letter will provide reasonable assurance of structuralintegrity. Furthermore, the staff finds that Code Cases 2142-1 and 2143-1 appropriately identify and classify these two weld metals for welding purposes, and there is no need to follow the special procedure and perform 6nce qualifications for non-Code materials.
I 4
Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed attemative to employ the aNemative welding i
materials and the classification categories of Code Cases 2142-1 and 2143-1 is authorized.
Principal Contributor: J. Zimmerman Date:
July 23,1998
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