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Category:INTERNAL OR EXTERNAL MEMORANDUM
MONTHYEARML20211P4961999-09-0909 September 1999 Revised Notification of 990916 Meeting with Tva.Meeting Will Be Closed to Public Due to Proprietary Info Contained in Westinghouse TR WCAP-15128 to Be Discussed ML20211H7841999-08-31031 August 1999 Notification of 990916 Meeting with Util in Rockville,Md to Discuss TVA Intent to Aply for License Amend to Use Alternate Steam Generator Repair Criteria for Stress Corrosion Cracking Evaluated in W TR WCAP-15128 ML20210F6981999-07-27027 July 1999 Discusses 990430 Request for Waiver of 10CFR170 Fees for Certain Insp Efforts Re NRR Pilot Regulatory Oversight Program.Determined That Thirteen Plants in Pilot Program Will Be Exempted from 10CFR170 Fees for Insp Efforts ML20210G9951999-07-20020 July 1999 Discusses 990628 Meeting Re New NRC Reactor Insp & Oversight Process.List of Attendees & Matl Used in Presentation Enclosed ML20196H8491999-06-28028 June 1999 Forwards June 1999 Performance Indicator (PI) Data Rept for Revised Reactor Oversight Process Pilot Program.Rept Constitutes Second Monthly PI Data Submittal from Pilot Plant Licensees,Which Includes Data Through May 1999 ML20207A4771999-05-18018 May 1999 Notification of 990629 Meeting with Util in Rockville,Md to Discuss long-range Planning with Regard to Dry Storage of Sf at Sequoyah Site ML20206M0061999-05-12012 May 1999 Notification of 990616 Meeting with TVA in Rockville,Md to Discuss long-range Planning with Regard to SG Replacement.Draft Agenda Encl ML20210F7021999-04-30030 April 1999 Requests Waiver of Fees for Thirteen Dockets That Compromise Plant Population for NRR Pilot Regulatory Oversight Program. Purposes of Pilot Program Described in Commission Papers SECY-99-007 & SECY-99-007A ML20207M6711999-03-12012 March 1999 Forwards Operator Licensing Exam Repts 50-327/98-302 & 50-328/98-302 Administered on 980928-1001,with as Given Written Exam Encl ML20204B5541999-03-12012 March 1999 Forwards NRC Approved Operator Licensing Exam (Facility Outline & Initial Exam Submittal & as-given Operating Exam) for Tests Administered on 980928-1001 ML20196H1561998-12-0303 December 1998 Notification of 981215 Meeting with TVA in Rockville,Md to Discuss Current & Future Licensing Activities ML20151X3091998-09-11011 September 1998 Notification of 980930 Meeting with Util in Rockville,Md to Discuss License Amend Request for Extending Sequoyah Emergency Diesel Generator Allowed Outage Time from 3 Days to 7 Days ML20237D4351998-08-21021 August 1998 Notification of 980901 Meeting W/Util in Rockville,Md to Discuss Planned SG Insp During Sequoyah Unit 1 Cycle 7 Refueling Outage in September 1998 ML20236X7041998-06-22022 June 1998 Notification of 980714 Meeting W/Tva in Rockville,Md to Discuss Status of Current Licensing Activities Re Sequoyah Nuclear Plant.Meeting Replaces Meeting W/Tva Scheduled for 980715 ML20249B7451998-06-19019 June 1998 Notification of 980708 Meeting W/Util in Rockville,Md to Discuss NRC Review of Installation of Lead Use Assemblies Containing U-236 Into Plant Core ML20249B7391998-06-19019 June 1998 Notification of 980715 Meeting W/Util in Rockville,Md to Discuss Status of Current Licensing Activities Re Plant ML20216B7171998-05-0707 May 1998 Notification of 980514 Meeting W/Util in Rockville,Md to Discuss NRC RAI Re TS Change Request to Extend Emergency Diesel Generator Allowed Outage Time from 3 Days to 7 Days ML20236N6021998-02-25025 February 1998 Responds to 980203 Memo Requesting Assistance in Responding to Operator License Appeal.Ts Branch Concurred in Interpretation ML20203K3261998-02-19019 February 1998 Revised Notification of 980305 Meeting W/Tvs in Rockville, Md.Meeting Being Held at Request of Utility to Discuss Special Project Being Pursued W/Doe,Use of Downblended Highly Enriched U in Fuel Used at Sequoyah Nuclear Plant ML20202D1571998-02-12012 February 1998 Requests That Encl Documents Be Placed on Docket for Sequoyah,Units 1 & 2 & in PDRs for Plant ML20202C5291998-02-11011 February 1998 Requests That Encl Documents Be Placed on Docket & in PDRs for Plant ML20202D1681998-02-11011 February 1998 Discusses MOV Terminal Voltage Issue,Per GL 98-10 Program at Plant.Concludes That Licensee Approach for Determining Operating Voltage for safety-related Movs,Acceptable ML20203A0311998-02-0505 February 1998 Notification of 980305 Meeting W/Tva in Rockville,Md to Discuss Special Projects Being Pursued W/Doe Including Use of Downblended Highly Enriched U in Fuel Used at Plant ML20236N5961997-12-31031 December 1997 Responds to Request for Technical Assistance (TIA 97-019) Re Interpretation of Intent of Sequoyah Nuclear Plant TS 4.10.3.2 ML20203B2651997-12-0303 December 1997 Notification of Meeting W/Tva on 971218 in Rockville,Md to Discuss Plans to Employ Electrosleeving Process to Repair Minor Cracks in Plants SG ML20202H9991997-12-0303 December 1997 Notification of 971210 Meeting W/Util in Rockville,Md to Discuss Current Licensing Activities Re Plant ML20236N4851997-10-16016 October 1997 Responds to 961016 Memo from Region II Requesting Technical Assistance (TIA 96-017) Re Interpretation of Sequoyah TS 3.2.4 on Quadrant Power Tilt Ratio ML20236N4471997-10-0101 October 1997 Responds to 961003,request from Region II for Technical Assistance Re Interpretation of Sequoyah TS 3/4.3.1, Specifically Items 10 & 11 on Table 3.3-1 ML20217D9131997-09-25025 September 1997 Notifies of 971002 Meeting W/Tva in Rockville,Md to Discuss Plans to Employ an Electrosleeving Process to Repair Minor Cracks in Plant SG ML20217P3451997-08-20020 August 1997 Notification of Licensee Meeting W/Util in Rockville,Md to Discuss Test Methods Used & Results of Plant Unit 1 SG Tube Insp Conducted in Apr 1997.Time & Meeting Room Changed ML20210K3301997-08-15015 August 1997 Notification of 970822 Meeting W/Util in Rockville,Md to Discuss Test Methods Used & Results of Plant,Unit 1 SG Tube Insp Conducted in Apr 1997 ML20205M3331997-07-28028 July 1997 Provides Final Agenda for Region II Enforcement Meeting to Be Held on 970730.OI Attendance Requested for Issue Involving Termination of Turkey Point Radwaste Operator Who Failed to Follow Procedure ML20205D9761997-07-0202 July 1997 Ack Receipt of Licensee Response to NOV (SL-IV) on Firewatches (EA 97-092).Licensee Requested That Region II Reconsider Violation as NCV ML20140F6191997-06-11011 June 1997 Forwards Documents Re Snp to Be Placed on Docket & in PDR ML20148F7201997-05-30030 May 1997 Forwards Semiannual Status Rept on Fire Protection Task Action Plan & plant-specific Thermo-Lag Correction Action Programs ML20141K7251997-05-28028 May 1997 Forwards Signed Original of Order Imposing Civil Monetary Penalty (Tva),For Transmittal to Ofc of Fr for Publication. W/O Encl ML20140F6471997-05-0707 May 1997 Forwards P Blanch Comments Re Snp Pressurizer Draindown Event for Review ML20140F6351997-05-0505 May 1997 Forwards P Blanch Comments on TVA Evaluation of 970324 Pressurizer Draindown Event.Disagrees W/Conclusion That TVA Failed to Identify Root Cause ML20138D1481997-04-29029 April 1997 Forwards Documents Re Meetings W/Tva Ig Ofc Concerning Review of TVA Nuclear Employees Concerns Resolution Program. Documents Should Be Forwarded to NRC PDR ML20138B5721997-04-25025 April 1997 Notification of 970502 Meeting W/Util in Rockville,Md to Discuss Methods & Schedule to Be Employed at Plant,Units 1 & 2 to Upgrade Fire Barriers That Contain Thermo-Lag Matl ML20140F4151997-04-25025 April 1997 Notification of 970425 Meeting W/Tva in Rockville,Md to Discuss Methods & Schedule to Be Employed at Plant to Upgrade Fire Barriers That Contain Thermo-Lag Matl.Meeting Cancelled ML20205M1931997-04-14014 April 1997 Informs That Lieberman Offers No Objection to Releasing TVA OI Report 2-96-009 to Licensee & PDR Provided That Report Released in Accordance with Guidance in Section 8.2.4 of Investigative Procedures Manual ML20205E3591997-04-10010 April 1997 Discusses Sequoyah NOV Response Extension.Two Weeks Extension Has Been Granted ML20137F7941997-03-28028 March 1997 Requests Technical Assistance Re Design of AFW Control Sys at Plant ML20137C1841997-03-20020 March 1997 Notification of Upcoming Meeting W/Tva on 970321 in Rockville,Md to Discuss Status of Currently Ongoing Applications for Amend of TS for Plant ML20137A6441997-03-19019 March 1997 Forwards Signed Original of Order Imposing Civil Monetary Penalty (Tva),For Transmittal to Ofc of Fr for Publication ML20135A6471997-02-25025 February 1997 Notification of 970227 Meeting W/Util in Rockville,Md to Discuss Response to NRC Request for Info Re Plant Amend Request 96-01 ML20203C0591997-02-12012 February 1997 Discusses OI Rept 2-96-009 Re Alleged Falsification of Firewatch Journals.Both Individuals Were Subsequently Terminated ML20205D9591997-02-11011 February 1997 Forwards Copy of OI Rept 2-96-009 Issued on 970124 Re Alleged Falsification of Firewatch Journals at Sequoyah Nuclear Plant.Concludes That Both Firewatch Personnel Failed to Patrol All Assigned Areas & Falsified Firewatch Journals ML20133E8681997-01-0909 January 1997 Notification of 970116 Meeting W/Framatome Cogema Fuels in Rockville,Md to Discuss Use of Bwu Critical Heat Flux Correlation at Sequoyah 1999-09-09
[Table view] Category:MEMORANDUMS-CORRESPONDENCE
MONTHYEARML20211P4961999-09-0909 September 1999 Revised Notification of 990916 Meeting with Tva.Meeting Will Be Closed to Public Due to Proprietary Info Contained in Westinghouse TR WCAP-15128 to Be Discussed ML20211H7841999-08-31031 August 1999 Notification of 990916 Meeting with Util in Rockville,Md to Discuss TVA Intent to Aply for License Amend to Use Alternate Steam Generator Repair Criteria for Stress Corrosion Cracking Evaluated in W TR WCAP-15128 ML20210F6981999-07-27027 July 1999 Discusses 990430 Request for Waiver of 10CFR170 Fees for Certain Insp Efforts Re NRR Pilot Regulatory Oversight Program.Determined That Thirteen Plants in Pilot Program Will Be Exempted from 10CFR170 Fees for Insp Efforts ML20210G9951999-07-20020 July 1999 Discusses 990628 Meeting Re New NRC Reactor Insp & Oversight Process.List of Attendees & Matl Used in Presentation Enclosed ML20196H8491999-06-28028 June 1999 Forwards June 1999 Performance Indicator (PI) Data Rept for Revised Reactor Oversight Process Pilot Program.Rept Constitutes Second Monthly PI Data Submittal from Pilot Plant Licensees,Which Includes Data Through May 1999 ML20207A4771999-05-18018 May 1999 Notification of 990629 Meeting with Util in Rockville,Md to Discuss long-range Planning with Regard to Dry Storage of Sf at Sequoyah Site ML20206M0061999-05-12012 May 1999 Notification of 990616 Meeting with TVA in Rockville,Md to Discuss long-range Planning with Regard to SG Replacement.Draft Agenda Encl ML20210F7021999-04-30030 April 1999 Requests Waiver of Fees for Thirteen Dockets That Compromise Plant Population for NRR Pilot Regulatory Oversight Program. Purposes of Pilot Program Described in Commission Papers SECY-99-007 & SECY-99-007A ML20204B5541999-03-12012 March 1999 Forwards NRC Approved Operator Licensing Exam (Facility Outline & Initial Exam Submittal & as-given Operating Exam) for Tests Administered on 980928-1001 ML20207M6711999-03-12012 March 1999 Forwards Operator Licensing Exam Repts 50-327/98-302 & 50-328/98-302 Administered on 980928-1001,with as Given Written Exam Encl ML20196H1561998-12-0303 December 1998 Notification of 981215 Meeting with TVA in Rockville,Md to Discuss Current & Future Licensing Activities ML20151X3091998-09-11011 September 1998 Notification of 980930 Meeting with Util in Rockville,Md to Discuss License Amend Request for Extending Sequoyah Emergency Diesel Generator Allowed Outage Time from 3 Days to 7 Days ML20237D4351998-08-21021 August 1998 Notification of 980901 Meeting W/Util in Rockville,Md to Discuss Planned SG Insp During Sequoyah Unit 1 Cycle 7 Refueling Outage in September 1998 ML20236X7041998-06-22022 June 1998 Notification of 980714 Meeting W/Tva in Rockville,Md to Discuss Status of Current Licensing Activities Re Sequoyah Nuclear Plant.Meeting Replaces Meeting W/Tva Scheduled for 980715 ML20249B7451998-06-19019 June 1998 Notification of 980708 Meeting W/Util in Rockville,Md to Discuss NRC Review of Installation of Lead Use Assemblies Containing U-236 Into Plant Core ML20249B7391998-06-19019 June 1998 Notification of 980715 Meeting W/Util in Rockville,Md to Discuss Status of Current Licensing Activities Re Plant ML20216B7171998-05-0707 May 1998 Notification of 980514 Meeting W/Util in Rockville,Md to Discuss NRC RAI Re TS Change Request to Extend Emergency Diesel Generator Allowed Outage Time from 3 Days to 7 Days ML20236N6021998-02-25025 February 1998 Responds to 980203 Memo Requesting Assistance in Responding to Operator License Appeal.Ts Branch Concurred in Interpretation ML20203K3261998-02-19019 February 1998 Revised Notification of 980305 Meeting W/Tvs in Rockville, Md.Meeting Being Held at Request of Utility to Discuss Special Project Being Pursued W/Doe,Use of Downblended Highly Enriched U in Fuel Used at Sequoyah Nuclear Plant ML20202D1571998-02-12012 February 1998 Requests That Encl Documents Be Placed on Docket for Sequoyah,Units 1 & 2 & in PDRs for Plant ML20202C5291998-02-11011 February 1998 Requests That Encl Documents Be Placed on Docket & in PDRs for Plant ML20202D1681998-02-11011 February 1998 Discusses MOV Terminal Voltage Issue,Per GL 98-10 Program at Plant.Concludes That Licensee Approach for Determining Operating Voltage for safety-related Movs,Acceptable ML20203A0311998-02-0505 February 1998 Notification of 980305 Meeting W/Tva in Rockville,Md to Discuss Special Projects Being Pursued W/Doe Including Use of Downblended Highly Enriched U in Fuel Used at Plant ML20236N5961997-12-31031 December 1997 Responds to Request for Technical Assistance (TIA 97-019) Re Interpretation of Intent of Sequoyah Nuclear Plant TS 4.10.3.2 ML20202H9991997-12-0303 December 1997 Notification of 971210 Meeting W/Util in Rockville,Md to Discuss Current Licensing Activities Re Plant ML20203B2651997-12-0303 December 1997 Notification of Meeting W/Tva on 971218 in Rockville,Md to Discuss Plans to Employ Electrosleeving Process to Repair Minor Cracks in Plants SG ML20236N4851997-10-16016 October 1997 Responds to 961016 Memo from Region II Requesting Technical Assistance (TIA 96-017) Re Interpretation of Sequoyah TS 3.2.4 on Quadrant Power Tilt Ratio ML20236N4471997-10-0101 October 1997 Responds to 961003,request from Region II for Technical Assistance Re Interpretation of Sequoyah TS 3/4.3.1, Specifically Items 10 & 11 on Table 3.3-1 ML20217D9131997-09-25025 September 1997 Notifies of 971002 Meeting W/Tva in Rockville,Md to Discuss Plans to Employ an Electrosleeving Process to Repair Minor Cracks in Plant SG ML20217P3451997-08-20020 August 1997 Notification of Licensee Meeting W/Util in Rockville,Md to Discuss Test Methods Used & Results of Plant Unit 1 SG Tube Insp Conducted in Apr 1997.Time & Meeting Room Changed ML20210K3301997-08-15015 August 1997 Notification of 970822 Meeting W/Util in Rockville,Md to Discuss Test Methods Used & Results of Plant,Unit 1 SG Tube Insp Conducted in Apr 1997 ML20205M3331997-07-28028 July 1997 Provides Final Agenda for Region II Enforcement Meeting to Be Held on 970730.OI Attendance Requested for Issue Involving Termination of Turkey Point Radwaste Operator Who Failed to Follow Procedure ML20205D9761997-07-0202 July 1997 Ack Receipt of Licensee Response to NOV (SL-IV) on Firewatches (EA 97-092).Licensee Requested That Region II Reconsider Violation as NCV ML20140F6191997-06-11011 June 1997 Forwards Documents Re Snp to Be Placed on Docket & in PDR ML20148F7201997-05-30030 May 1997 Forwards Semiannual Status Rept on Fire Protection Task Action Plan & plant-specific Thermo-Lag Correction Action Programs ML20141K7251997-05-28028 May 1997 Forwards Signed Original of Order Imposing Civil Monetary Penalty (Tva),For Transmittal to Ofc of Fr for Publication. W/O Encl ML20140F6471997-05-0707 May 1997 Forwards P Blanch Comments Re Snp Pressurizer Draindown Event for Review ML20140F6351997-05-0505 May 1997 Forwards P Blanch Comments on TVA Evaluation of 970324 Pressurizer Draindown Event.Disagrees W/Conclusion That TVA Failed to Identify Root Cause ML20138D1481997-04-29029 April 1997 Forwards Documents Re Meetings W/Tva Ig Ofc Concerning Review of TVA Nuclear Employees Concerns Resolution Program. Documents Should Be Forwarded to NRC PDR ML20140F4151997-04-25025 April 1997 Notification of 970425 Meeting W/Tva in Rockville,Md to Discuss Methods & Schedule to Be Employed at Plant to Upgrade Fire Barriers That Contain Thermo-Lag Matl.Meeting Cancelled ML20138B5721997-04-25025 April 1997 Notification of 970502 Meeting W/Util in Rockville,Md to Discuss Methods & Schedule to Be Employed at Plant,Units 1 & 2 to Upgrade Fire Barriers That Contain Thermo-Lag Matl ML20205M1931997-04-14014 April 1997 Informs That Lieberman Offers No Objection to Releasing TVA OI Report 2-96-009 to Licensee & PDR Provided That Report Released in Accordance with Guidance in Section 8.2.4 of Investigative Procedures Manual ML20205E3591997-04-10010 April 1997 Discusses Sequoyah NOV Response Extension.Two Weeks Extension Has Been Granted ML20137F7941997-03-28028 March 1997 Requests Technical Assistance Re Design of AFW Control Sys at Plant ML20137C1841997-03-20020 March 1997 Notification of Upcoming Meeting W/Tva on 970321 in Rockville,Md to Discuss Status of Currently Ongoing Applications for Amend of TS for Plant ML20137A6441997-03-19019 March 1997 Forwards Signed Original of Order Imposing Civil Monetary Penalty (Tva),For Transmittal to Ofc of Fr for Publication ML20135A6471997-02-25025 February 1997 Notification of 970227 Meeting W/Util in Rockville,Md to Discuss Response to NRC Request for Info Re Plant Amend Request 96-01 ML20203C0591997-02-12012 February 1997 Discusses OI Rept 2-96-009 Re Alleged Falsification of Firewatch Journals.Both Individuals Were Subsequently Terminated ML20205D9591997-02-11011 February 1997 Forwards Copy of OI Rept 2-96-009 Issued on 970124 Re Alleged Falsification of Firewatch Journals at Sequoyah Nuclear Plant.Concludes That Both Firewatch Personnel Failed to Patrol All Assigned Areas & Falsified Firewatch Journals ML20133E8681997-01-0909 January 1997 Notification of 970116 Meeting W/Framatome Cogema Fuels in Rockville,Md to Discuss Use of Bwu Critical Heat Flux Correlation at Sequoyah 1999-09-09
[Table view] |
Text
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so; UNITED STATES NUCLEAR REGULATOAY coMM:ssiCN
[l y' i A ASMING TON O C 00155 5 , %;nd !
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.....< JUL 2 41980 MEMORANDUM FOR: Richard H. Vollmer, Director .
Division of Engineering, NRR FROM: L. C. Shao, Assistant Director for General Reactor Safety Research, RES
SUBJECT:
PEER REVIEW 0F A DIFFERING PROFESSIONAL OPINION, SEQUOYAH UNIT ONE PRESSURIZER RELIEF LINE WELD REPAIR
Reference:
Memo to Shao from Vclimer, " Peer Review of a Differing Professional Opinion," dated June 30, 1980 In May 1979 Sequoyah Unit One pressurizer relief line was deformed as a result of a support malfunction during hot functional testing of the reactor coolant system. In order to restore the pipe to its original configuration, TVA used a weld draw-bead technique, employing two deeply grooved welds that i.ere ground out and rewelded, in the 6-inch stainless steel pipe. The NRC staff accepted this repair, but Mr. J. Halspatz expressed concerns about the repair. Subsequently, a 304 stainiess steel mockup was evaluated and in situ metallography on the field repair was performed.
Based on tneir review of the mockup and field repair, the NRC staff iccepted the TVA repair procecure. Nevertheless, Mr. J. Halapatz of tne NRC staff suomitted a differing professional opinion regarding tne acceptance of the weld recair. On June 30, 1980, you (by referenced memo) requested that a peer review of Mr. Halapatz's differing professional opinion be conducted.
In completing tnis review, I have drawn on information and evaluations developeo by my staff, principally J. Muscara, C. Z. Serpan, Jr., and J. E. Richardson. We have reviewed the information included in your memo, have conoucted interviews with NRC personnel from NRR, IE Headquarters and SD, and have held two telephone interviews with personnel from TVA. Both J. Muscara and C. Z. Serpan have met with TVA at their headquarters offices and at a nuclear plant construction site to review and evaluate previously available and aeditional inforr.ation. In completing the conclusions and recommendations, I have considered all of the metallurgical information provided by my staff and have also considered the overall safety implications of this repair weld against the history and experience of other welds and repairs in nuclear service.
S O Ogg10 2 8 4
s .
R. H. Vollmer 2 M 2 '- 2 Two main points of the differing professional opinion of Mr. J. Halacat:
are that (1) the draw-bead welds might have penetrated to the pipe I.D.
thereby requiring a hydrotest, and (2) the draw-bead welding might have caused sufficient sensitization of the piping to render it susceptible to intergranular stress corrosion cracking (IGSCC). Some of Mr. Halapatz's contentions regarding item (2) have merit.
The following paragraphs describe the results of this peer review with respect to weld penetration and sensitization, with conclusions and recomendations following; a discussion of technical issues and comparison of the weld repair to BWR nonconforming service-sensitive lines is enclosed.
First, with respect to weld penetration, adequate information has been presented and reviewed by NRC IE Region II to indicate that the weld did not penetrate tne I.D. Further, a hydrotest in this case would not provide any additional useful information on weld integrity. We believe that the NDE performed. i.e., radiography and UT, have shown the soundness of the repair.
With respect to sensiti:ation, it is concluded that the pressuri:er relief pipe has been sensitized under the repair welds cown to the inside surface and most likely has been made susceptible to IGSCC by the repair process.
This opinion derives from a review of the available information attached to your referenced memo, a review of research results of the past 5 years, and from results of the second mockuo prepared by TVA, as disclosed to C. Z. Serpan, Jr. ano J. Muscara during a visit to TVA on July 17 and 18,1980.
Those results show a region of sensitization below the bottom of the weld groove that is sufficiently large as to encroach on the pipe I.D. on the field recair. This sensiti:ation, apparent to one degree or another all the way frem the weld root to the wall I.D., when coupled with the relatively hign 02 content in tne steam phase in the line (reported by TVA to be a maximum of 0.2 ppm), plus high residual tensile stresses, could result in initiation of IGSCC.
Conclusions and Recommendations We believe that the Sequoyah Unit One pressurizer relief line draw-beac repair (1) did not oenetrate the pipe wall so the hydrostatic testing is not necessary, and (2) did become sensiti:ed and, thus, could be susceptible to IGSCC in service. As stated above, in addition to metallurgical reasons, the conclusions and recommendations are based on an evaluation of the overall situation which includes field expvience, operating stresses, successful experiences in detecting small leaks, and behavior under accident loads. Therefore, although this stainless steel line most likely is sensiti:ed to some degree and will operate in an environment of 02 concentration
R. H. Vollmer 3 JUL 2 41980 somewhat similar to that of BWRs under full power operating conditions wnich have experienced cracking in service, this specific repair weld appears to be neither worse nor more serious than many other welds that are currently in service, especially those welds in BWR nonconfoming service-sensitive lines (see page 6 of the enclosure). We, therefore, reconrnend resolution of this matter to be the same as for service-sensitive nonconforming lines as stated in NUREG-0313, Revision 1. -
Altnough we believe the above reconsnendations will satisfactorily resolve this issue, if one wishes to better understand the status of the repair weld, or to minimize the potential crack growth of the repair weld, other tests could be performed on the second mockup or on the pressurizer relief line itself. For the mockup: (1) Electrochemical Potentiokinetic Reactivation (EPR) tests could be perfomed to evaluate the degree of sensitization on the irside surface of the pipe, and (2) Slow Extension Rate Testing (SERT) of inside surface specimens could be perfomed in a simulated service environment (saturated steam at 650'F and 2485 PSI containing 0.2 ppm 07 )
to determine the susceptibility to IGSCC. For the repaired pipe, technTques exist for conducting in situ EPR tests on the inside surface or near-inside surface to determine the degree of sensitization. In the absence of conducting any further tests, acoustic emissior, can be used for limited area monitoring of the repaired section of the pipe to detect, on-line, the possible growth cf IGSCC and thereby avoid any leakage in service.
It is pointed out that the whole issue of weld repair is not well addressed oy the ASME Boiler and Pressure Vessel Code, nor by the NRC in Regulatory Guices or other technical positions. It would be beneficial for this section of the Code to be strengthened or other NRC positions to be developed to better address the sensitization problem in weld repairs.
m f-L -- '.,..
L. C. Shao, Assistant Director for General Reactor Safety Research Division of Reactor Safety Research
Enclosure:
Discussion of Tecnnical Issues cc w/ encl:
R. J. Budnitz, RES H. R. Denton, NRR E. G. Case, NRR T. E. Murley, RES C. Z. Serpan, RES J. E. Richardson, RES l J. Muscara, RES )
J. Halapatz, NRR '
l
DISCUSSION OF ISSUES RELATED TO DRAW-BEAD WELD REPAIR OF SE000YAH UNIT ONE PRESSURIZER RELIEF PIPE TVA prepared a mockup to qualify the procedure for draw-bead field repair.
Documentation and quality control of the draw-bead. repair appear to be w tain normal requirements, but are not adequate for a careful review of what was actually done or to evaluate the metallurgical consequences of the repair.
Altncugh the draw-bead repair is stated to be Code acceptable, and is practiced in fisid repairs, the Code does not address the relevant topics of residual welding stresses, sensitization and susceptibility to stress corrosion cracking of austenitic stainless steels all of which can result from repair welding and can be responsible for field failures. Further, the draw-bead weld repair technicue is not specifically discussed and accepted by the Code, but upon a cuery to ASME by TVA, ASME found this draw-bead method as being implied in the rules for weld repair.
First Mockuo - 30a SS - The original mockup prepared and studied by TVA and j reviewed by NRC staff and consultant was not adequate to address the issues i of sensiti:ation and cracking with respect to the field repair. The mockup was of 304 stainless steel while the field repair was of 316 stainless steel.
Althougn the welding heat inout was claimed to te much higner in the mockup ,
- wnen TVA was interviewed, they agreed that the travel sceed both in the mockup and in the 'ield were not measured. TVA stated that the mockup constitutes a conservative representation of field repairs because of the higner heat input ano hign carbon 304 stainless steel used in the mockuo. Further, TVA claims tnat the field repair was acceptable because the specimers from nis mock-up
- :assed the A262E intergranular corrosion test. It is agreed that mcckup
! soecimens passed the A262E test, even :nougn metallograony did snow a J
A r.
2 considerable level of sensitization; however, the A262E tests over the last 5 years nave been reported to be insensitive to levels of sensitization in welded stainless steel that are nevertheless high enough to crack stainless steels in operating plants.
Therefore, the mockuc was unrepresentative and also it cannot unquestionably be considered conservative. Further, the corrosion tests performed do not accurately predict the susceptibility to IGSCC, and tnus, such tests cannot be used to establish the acceptability of the repair.
The review shows that the caroon level in tne 304 stainless steel mockup was not cocumentec out was saic to be abcut 0.065 w/o. Questions still exist witn resoect to accurate information on weld neat input in neckup versus field altnough the mockup heat inout was reported to be about 179 Kj/in. versus about 2u Kj/in, fcr-the field.
In situ Meta 11ocrachy - At a meeting between TVA and NRC staff on March 13, 1980, NRC requested, and TVA subsequently conducted, in situ metallography on the outside surface of the draw-bend weld repaired pressuri:er relief pipe. The in situ metallography was witnessed by IE Region II personnel. TVA and IE Region II concluded from an evaluation of this in situ metallography that the pice HAZs were not sensitized, tnat sensiti:ation cid not encroach on the pipe insioe surface, and that, therefore, the weld repair was acceptable and the pipe would not be susceptible to IGSCC. NRR also accepted these conclusions.
o .-
3 It is pointed out nat outside surface metallography can be totally misleading with respect to the state of sensitization on the inside surface of the pice.
Recent research results have snown that the outside surface of a pipe may not sensitize appreciably due to welding even though the inside of the pipe does. Very often the inside surface of the pipe is sensitized enough to render it susceptible to cracking hile the outside has very little or no sensitiration.
If the outside surface of piping is found to be sensit.ized then :ne inside surface will be even more sensitizec; tnis situation can arise wnen the heat of material is especially sensitive or is already sensiti:ed in :ne as-received condition before welcing.
There are several important reasons wny the outside surface does not sensitize nearly as mucn as tne inside. Firstly, the oatside surface coes not experience as much excesure to heat (total time at sensitization temperatures) as the inside surface. Secondly, two parameters have been found most important in accelerating weld sensiti:ation; these are the thermal cycles and the strain cycles experienced by the underlying (or adjoining) base material. The themal cycles and strain cycles increase the kinetics of sensiti:ation. Since the inside surface of a welded pipe experiences more cumulative thermal and strain cycles tnan the outside surface, more sensiti:ation is found on the inside '
surface. With respect to the draw-bead weld repair of the pressuri:er relief pipe, the 316 stainless steel material used has slower sensiti:ation kinetics than 3Ca stainless steel, so that under nomal comparable welding conditions, 31e would sensiti:e less. The weld recair was acconolished by filling tne groove once then grinding out tne weld metal and refilling :ne groove. This
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procedure, then, has imposed twice the thermal / strain cycles on the base material underlying the root bead leading to the insioe surface than on a nor nal penetration welc. The outside surface did not experience very many ther nal/ strain cycles and can be expected to be nonsensitized or only lign:ly sensitized. The inside surface, however, because of tne accelerating effect of thermal / strain cycle: and the large number of cycles received car, be exoected to be sensiti:ed. Therefore, one cannot infer from the outside surface metallography results, that the inside surface of the draw-bead repaired pipe would not be sensitized.
Secone Mockue - 316 SS - On July 17 and 18, 1980, C. Z. Serpan and J. Muscara reviewed results of a second mockup prepared by TVA in February 1980. This mockup more closely represented the actual pressurizer relief pipe material, anc furtnemore, the welder was instructed to use tne same welcing procecure as for the' field repair. The mockup material used was 316 stainless steel of .051 percent C (pressuri:er pipe is 316 SS of .052 percent C). The mockup contained one penetration weld and two draw-bead repair welds. The two draw-bead welds penetrated approximately 2/3 and 1/2 the oipe wall thickness, respectively. It was evident from the metallography tnat (a) the full penetration weld (wnicn was not reground and rewelded) showed no sensitization on either the outsice or inside surface; (b) the draw-bead welds (wnicn were reground anc reweloed) showed no sensitization on the outsice surface but showed serlsitization along the fusion lines extending up from the root pass to within about 1/4 T of the wali 0.D., as well as a considerable :ene of sensitization 1
in the base material uncerneath the root pass leading towards and even reaching l l
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the inside surface; (c) the sensitized base metal extended about 0.2 inches '
below the root pass towards the inside surface, although the degree of sensiti:ation decreased from the root bead towards the inside surface.
By relating the results of this mockup to the field draw-bead repair, the absence of sensitiration on the outside surface of the field pipe, as shown by in situ metallograohy, is confirmed. Also confirmed is our evaluation that sensiti:ation to some degree extends to the inside surface of the repaired pressuri:er relief pipe since there is less than 0.2 inches of material left on the pipe while the mockup shows a sensitized :ene extending about 0.2 inches from the root bead tcward the inside surface.
Corrosion tests were also performed on this second mockup using ASTM A262 practice E tests. These tests snowed the absence of cracking. As described earlier, A262 practice E tests are not very sensitive for detecting tne susceptibility to IGSCC in LWR environments for materials with sensitization levels wnich are nevertheless high enough'to be susceptible. As a good example of this test's poor discrimination to cracking susceptibility, we highlight the results of the TVA first mockup whien used a high carbon (said by TVA to be .065 percent) 304 stainless steel and a very high heat inout resulting in a condition which would most certainly have failed in LWR environments containing 0.2 pom oxygen: test samples from this 304 stainless steel mockup passed the AS*M A262 practice E test.
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Com:arison of tne Weld Repair to BWR Ncnconformine Service-Sensitive Lines -
The weld recair to the Secuoyan pressuri:er relief line is at least as good as many welds in SWR nonconfoming service-sensitive lines investicated by the Dipe Crack Study Group (NUREG-0313). This conclusion is derived from the following:
- 1. The average car 0cn content in the 304 stainless steel BWR pipe material found to be cracking in service, aoout 0.06 cercent, is about the same as the pipe material in the Sequoyah line (316 stainless steel - 0.052 percent).
- 2. Under full operating conditions, oxygen content of the BUR coolant is slightly higher than that of a DWR pressuri:er relief line (0.2-0.4 con vs 0.005-0.2 pom). Tne im::ortance of this is tnat oxygen is a major centributor to intergranular stress corrosion of sensiti:ed stainless steels.
- 3. Over the cast several years, there have been numerous cases of cracks in BWR stainless steel lines while there have been no reports of cracks in the pressurizer relief lines of PWRs.
J. It is likely that some BWR lines, because of their geometry or location. l may be stressec nigner than tne PWR oressuri:er relief line esoecially wnen large eartncuakes are considered.
- 5. Tne safety significance of one 6-incn PNR relief line weld is much less than the safety significance of a larger amount of welds in SWR nonconforming service-sensitive lines.