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Category:Letter
MONTHYEARNMP1L3570, Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-02-0101 February 2024 Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation IR 05000220/20230042024-02-0101 February 2024 Integrated Inspection Report 05000220/2023004 and 05000410/2023004 NMP1L3569, CFR 50.46 Annual Report2024-01-26026 January 2024 CFR 50.46 Annual Report ML24004A2122024-01-0808 January 2024 Senior Reactor and Reactor Operator Initial License Examinations ML23354A0012024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0059 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines NMP1L3566, Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station2023-12-14014 December 2023 Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station IR 05000410/20243012023-12-14014 December 2023 Initial Operator Licensing Examination Report 05000410/2024301 ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 NMP1L3564, Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements ML23291A4642023-12-0707 December 2023 Issuance of Amendment No. 251 Regarding the Adoption of Title 10 the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of SSC for Nuclear Power Plants ML23289A0122023-12-0606 December 2023 Issuance of Amendment No. 250 Regarding the Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b NMP1L3563, Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-12-0404 December 2023 Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) IR 05000220/20234022023-11-28028 November 2023 Security Baseline Inspection Report 05000220/2023402 and 05000410/2023402 NMP1L3557, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums IR 05000220/20234202023-11-0101 November 2023 Security Baseline Inspection Report 05000220/2023420 and 05000410/2023420 ML23305A0052023-11-0101 November 2023 Operator Licensing Examination Approval IR 05000220/20230032023-10-25025 October 2023 Integrated Inspection Report 05000220/2023003 and 05000410/2023003 IR 05000220/20235012023-10-17017 October 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000220/2023501 and 05000410/2023501 IR 05000220/20230112023-10-16016 October 2023 Comprehensive Engineering Team Inspection Report 05000220/2023011 and 05000410/2023011 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans NMP1L3554, Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases2023-10-0606 October 2023 Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases C IR 05000220/20233032023-09-20020 September 2023 Retake Operator Licensing Examination Report 05000220/2023303 ML23250A0822023-09-19019 September 2023 Regulatory Audit Summary Regarding LARs to Adopt TSTF-505, Rev. 2, and 10 CFR 50.69 ML23257A1732023-09-14014 September 2023 Requalification Program Inspection IR 05000220/20230052023-08-31031 August 2023 Updated Inspection Plan for Nine Mile Point Nuclear Station, Units 1 and 2 (Report 05000220/2023005 and 05000410/2023005) RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs NMP2L2851, Relief Request Associated with Successive Inspections for Generic Letter 88-01 / BWRVIP-75-A Augmented Examinations2023-08-25025 August 2023 Relief Request Associated with Successive Inspections for Generic Letter 88-01 / BWRVIP-75-A Augmented Examinations ML23151A3472023-08-21021 August 2023 Issuance of Amendments to Adopt TSTF-295-A, Modify Note 2 to Actions of PAM Table to Allow Separate Condition Entry for Each Penetration NMP1L3534, License Amendment Request - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2023-08-18018 August 2023 License Amendment Request - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation ML23220A0262023-08-0808 August 2023 Licensed Operator Positive Fitness-for-Duty Test IR 05000220/20234012023-08-0808 August 2023 Cyber Security Inspection Report 05000220/2023401 and 05000410/2023401 (Cover Letter Only) NMP1L3545, Supplemental Information Letter to Adopt TSTF-505, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems .2023-08-0404 August 2023 Supplemental Information Letter to Adopt TSTF-505, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems . RS-23-087, Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor2023-08-0404 August 2023 Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor IR 05000220/20230022023-08-0101 August 2023 Integrated Inspection Report 05000220/2023002 and 05000410/2023002 ML23207A0762023-07-14014 July 2023 EN 56557 - Update to Part 21 Report Re Potential Defect with Trane External Auto/Stop Emergency Stop Relay Card Pn: XI2650728-06 NMP1L3544, Fifth Inservice Inspection Interval, First Inservice Inspection Period 2023 Owner'S Activity Report for RFO-27 Inservice Examinations2023-07-14014 July 2023 Fifth Inservice Inspection Interval, First Inservice Inspection Period 2023 Owner'S Activity Report for RFO-27 Inservice Examinations ML23186A1642023-07-0606 July 2023 Operator Licensing Retake Examination Approval NMP2L2846, Nine Mire Point Nuclear Station, Units 1 and 2, General License 30-day Cask Registration Notifications2023-07-0505 July 2023 Nine Mire Point Nuclear Station, Units 1 and 2, General License 30-day Cask Registration Notifications ML23192A0622023-06-30030 June 2023 Engine Systems, Inc., 10CFR21 Reporting of Defects and Non-Compliance, Report No. 10CFR21-0136, Rev. 0 IR 05000220/20230102023-06-29029 June 2023 Biennial Problem Identification and Resolution Inspection Report 05000220/2023010 and 05000410/2023010 ML23131A4242023-06-23023 June 2023 Issuance of Amendment No. 249 Regarding the Revision to Technical Specification 3.3.1 to Adopt Technical Specifications Task Force Traveler TSTF-568 RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations NMP1L3539, Day Commitment Response - Relief Request I5R-11 Concerning the Installation of a Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-06-0909 June 2023 Day Commitment Response - Relief Request I5R-11 Concerning the Installation of a Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) ML23159A0052023-06-0505 June 2023 56557-EN 56557 - Paragon - Redlined RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling IR 05000410/20233022023-05-15015 May 2023 Initial Operator Licensing Examination Report 05000410/2023302 2024-02-01
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML23264A7992023-09-21021 September 2023 NRR E-mail Capture - Final RAI - Constellation Energy Generation, LLC Fleet Request License Amendment Request to Adopt TSTF-580, Revision 1 ML23205A2432023-07-19019 July 2023 NRC Staff Follow-up Question on Audit Question 18 TSTF-505 and 50.69 Regulatory Audit (E-mail Dated 7/19/2023) (EPIDs L-2022-LLA-0185 and L-2022-LLA-0186) ML23087A2912023-03-28028 March 2023 Request for Additional Information (3/28/2023 E-mail) - Proposed Emergent I5R-11 Alternative Associated with a Weld Overlay on RPV Recirculation Nozzle N2E DM Weld ML23061A0522023-03-0202 March 2023 Request for Additional Information (3/2/2023 E-mail) - Proposed Alternative Associated with a Weld Overlay Repair to the Torus ML23012A2002023-01-13013 January 2023 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection 05000220/2023401 and 05000410/2023401 ML22207A2162022-07-26026 July 2022 Information Request to Support Triennial Baseline Design-Basis Capability of Power-Operated Valves Inspection; Inspection Report 05000220/2022010 and 05000410/2022010 ML22194A9412022-07-13013 July 2022 Request for Additional Information Relief Request CS-PR-02 (7/13/2022 e-mail) ML22041B5362022-02-10010 February 2022 NRR E-mail Capture - Constellation Energy Generation, LLC - Request for Additional Information Regarding Fleet License Amendment Request to Adopt TSTF-541 ML22020A0642022-01-13013 January 2022 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative for Repair of Water Level Instrumentation Partial Penetration Nozzles ML21320A3472021-11-16016 November 2021 Request for Additional Information LAR to Revise TSs to Adopt TSTF-582, Revision 0 ML21306A3312021-11-0202 November 2021 Request for Additional Information Alternative Request GV-RR-10 (11/2/2021 e-mail) ML21256A1902021-09-10010 September 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding License Transfer Application ML21144A2132021-05-24024 May 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding License Transfer Application ML21117A0342021-05-0505 May 2021 Request for Additional Information Regarding Proposed Alternative to Use ASME Code Case N-893 ML21125A1282021-05-0404 May 2021 OPC Document Request - Feb 2021 ML21110A5112021-04-20020 April 2021 Request for Additional Information Review of License Amendment Request to Revise Technical Specifications to Adopt TSTF-582 ML21088A2682021-03-30030 March 2021 Notification of Conduct of a Fire Protection Team Inspection ML21062A0652021-03-0101 March 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative to Documentation Requirements for Pressure Retaining Bolting ML21049A2572021-02-18018 February 2021 Request for Additional Information Byron/Dresden Proposed Changes to Site Emergency Plans to Support Post-Shutdown and Permanently Defueled Conditions (EPID-2020-LLA-0240 & EPID-2020-LLA-0237) ML20365A0092020-12-30030 December 2020 Request for Additional Information Concerning Review of License Amendment Request and Relief Request to Change Excess Flow Check Valve Testing Frequency (EPIDs L-2020-LLA-0188 and L-2020-LLR-0114) ML20358A2602020-12-28028 December 2020 Changes to Draft Request for Additional Information Regarding License Amendment Request and Relief Request to Change Excess Flow Check Valve Testing Frequency ML20272A2802020-09-28028 September 2020 Withdrawal and Replacement of Request for Additional Information to Support Review of License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times ML20248H5192020-09-28028 September 2020 Changes to Draft Request for Additional Information Regarding Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times ML20246G6362020-09-0202 September 2020 Request for Additonal Information to Support Review of License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times NMP2L2739, Request for Additional Information for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 22020-08-28028 August 2020 Request for Additional Information for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 ML20239A7982020-08-25025 August 2020 NRR E-mail Capture - Exelon Generation Company, LLC - Fleet License Amendment Request to Adopt TSTF-568, Revision 2 ML20212L8702020-07-30030 July 2020 Request for Additonal Information Review of License Amendment Requests Regarding Riskinformed Categorization and Treatment of Structures, Systems and Components (L-2019-LLA-0290) ML20213A9352020-07-30030 July 2020 Request for Additional Information Review of License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times ML20153A7042020-06-0101 June 2020 NRR E-mail Capture - Preliminary RAI for Fleet Request to Use Alternative OMN-26 ML20135H1972020-05-14014 May 2020 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Request to Extend Safety Relief Valve Test Interval ML20045E3582020-02-14014 February 2020 Draft Request for Additional Information Regarding License Amendment Request to Increase Allowable MSIV Leakage Rates ML19296A1862019-10-23023 October 2019 Request for Additional Information Regarding License Amendment Request to Increase Allowable MSIV Leakage Rates (L-2019-LLA-0115) ML19275H1362019-10-0202 October 2019 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Request to Use ASME Code Case N-879 ML19179A0612019-07-19019 July 2019 Three Mile Point 1 - Supplemental Information Needed to Proposed Alternative to Use ASME Code Case N-879 ML19151A8132019-05-31031 May 2019 Licensed Operator Positive Fitness-For-Duty Test ML19025A1572019-01-25025 January 2019 Request for Additional Information Regarding Primary Containment Oxygen Concentration License Amendment Request ML19025A1202019-01-24024 January 2019 NRR E-mail Capture - Calvert Cliffs, Fitzpatrick, and Nine Mile Point - Request for Additional Information Regarding License Amendment Request to Revise Emergency Response Organization Staffing ML18341A2212018-12-0707 December 2018 Request for Additional Information Regarding Emergency Tech Spec Change Re HPCS Completion Time (EPID -L-2018-LLA-0491) ML18228A6932018-08-15015 August 2018 Request for Additional Information Regarding Reactor Pressure Vessel Water Inventory Control License Amendment Request (L-2017-LLA-0426) ML18205A3922018-07-24024 July 2018 Request for Additional Information Regarding Reactor Pressure Vessel Water Inventory Control License Amendment Request (L-2017-LLA-0426) ML18184A2882018-07-0303 July 2018 Request for Additional Information Regarding Removal of Boraflex Credit from Spent Fuel Pool License Amendment Request(L-2018-LLA-0039) ML18102A2372018-04-12012 April 2018 NRR E-mail Capture - Calvert Cliffs, Ginna, and Nine Mile Point - Request for Additional Information Regarding License Amendment Request to Revise Emergency Action Level Schemes (EPID-L-2017-LLA-0237) ML17331B1342017-12-12012 December 2017 2, and R.E. Ginna Nuclear Power Plant - Request for Additional Information - Regarding ML17285B1962017-10-27027 October 2017 Request for Additional Information Regarding Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools. ML17272A0112017-10-10010 October 2017 Request for Additional Information Regarding License Amendment Concerning Reactor Pressure Vessel Water Inventory Control ML17234A3592017-08-30030 August 2017 Request for Additional Information Regarding Relief Request NMP-RR-001 to Utilize Code Case N-702 ML17241A2752017-08-29029 August 2017 Request for Additional Information Response MSA (MF7946,7) and FE (MG0087,8) E-mail Attachment ML17240A3102017-08-15015 August 2017 NRR E-mail Capture - Nine Mile Point MSA and FE RAI ML17172A0842017-06-27027 June 2017 Request for Additional Information Regarding Relief Request to Utilize ASME Code Case N-702 ML17065A1622017-03-14014 March 2017 Request for Additional Information Regarding Proposed Alternative to Perform Pressure Isolation Valve Leakage Testing at Frequencies Consistent with 10 CFR Part 50, Appendix J 2023-09-21
[Table view] |
Text
MAR 08 f978 Docket No 2,50-43.0 Niagara Mohawk Power Corporation ATXM: Ikr. Gerlad. K. Diode Vice President - Mrgineering 300 Erie Boulevard Nest Syracuse, New, York 13202 Gentlemen:
SUBJECT:
NBC STAFF POSXTICM OH THE USE OF AUSTEMXTIC STAINLESS STEEL IN BOILIHG PPZER REACTOR FACXLITXESWIHE BILE HHbJT NU~+
SZATIC8, UNIT 2 During the past several years, the MC and its predecessor agency, the ABC, have conducted an extensive investigation to evaluate the cracking of austenitic stainless steel piping. This effort was initiated following the detection in late 1974 and early 1975 of a series of cracks in the piping of boiling water reactor facilities. As a result of this investigation, we have concluded that the types of austenitic stainless steel currently used in boiling water reactor piping are susceptible to stress corrosion cracking.
The staff believes the probability is extremely low that such stress corrosion cracks vill propagate far enough to create a significant safety hazard to the public. However, we have also concluded that steps should be taken to elimi-nate this condition. To this end, we have developed a position to set forth acceptable reethods to reduce the susceptibility of boiling water reactor piping to stress corrosion cracking. This position is contained in HUREG-0313, dated July 1977, a copy of which is enclosed. Pe have also incorporated the position contained in NUREG-0313 as Branch Technical Position NTEB 5-7 and issued it as. a revision to the Standard Review Plan. ,
You should note that the implen~tation schedule set forth in the position provides for varying degrees of conformance, depending upon the status of the application. 1'equire that you provide a schedule for your response to this position within 14 days of receipt of this letter. Your response should address each of the subsections in Section II and XIX of the position.
Forty (40) c'opies of your response are needed for use by the stMf.
OKKICK~
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NRC FORM 318 (9-76) NRCM 0240 4 UI S, OOVKRNMKNTPRINTINO,OI'PICK> 1979 429 924
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Niagara Mohawk Power Corporation .
MAR 0 8 1978 lf you require any clarification of this request, staff's assigned Licensing Project 2~1anager.
please contact the
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This request for generic information was approved by GEO under a blanket clearance Ho. R0071. This clearance expires September 30,"1978.
Sincerely,
'lso'P ~~" 407/
Or<
Steven A. Varga, Chief Light Rater Reactors Branch No. 4 Division of Project t4anagement
Enclosure:
NUREG-0313, dated July 1977 cc vr/enclosure:
See page 3 DFFICS~
SURNAME~ W cm SAIItj a DATd~ / oz, /78 0)/f /78 NRC FORM 318 (9-76) NRCM 0240 4 U, S, OOVSRNMSNT FRINTINO OFFICSs ISTS d2d d24
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Niagara Mohawk Power Corporation JAR 0 8 l9/9 ccs.
Eugene B. Thomas, Jr.
LeBoeuf, Lamb, Leiby & MacRae 1757 N Street, N. W.
Washington, D. C. 20036 Anthony Z. Roisman, Esq.
Roisman, Kessler & Cashdan 1025 15th Street, NW Washington, D. C. 20036 Mr. Richard Goldsmith Syracuse University College of Law E. I. White Hall Campus Syracuse,"Sew"York '13210 T. K. DeBoer, Director Technological Development Programs New York State Energy Office Swan Street Building Core 1 2nd Floor Etrpire State Plaza Albany, New York 12223
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NUR EG-0313 TECHNICAL REPORT ON MATERIALSELECTION AND PROCESSING GUIDELINES FOR BWR COOLANT PRESSURE BOUNDARY PIPING Manuscript Completed: July 1977 Date Published: July 1977 Division of Operating Reactors Division of Systems Safety Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
TABLE OF CONTENTS I ." INTRODUCTION ~ ~ 1 II.
SUMMARY
OF ACCEPTABLE METHODS TO MINIMIZE CRACK SUSCEPTIBILITY ~ \ 3 I I I. INSERVICE INSPECTION AND LEAK DETECTION REQUIREMENTS FOR BWRs WITH VARYING CONFORMANCE TO MATERIAL SELECTION AND PROCESSING GUIDELINES 4 IV. IMPLEMENTATION OF MATERIAL SELECTION AND PROCESSING GUIDELINES ~ 9 V. GENERAL RECOMMENDATIONS .10
I. INTRODUCTION Small, hairline cracks in austenitic stainless steel piping in boiling water reactor (BWR) facilities were observed as early as 1965. In each case, it was believed that the situation had been corrected or substan-tially reduced by better control of welding, contaminants and/or design modifications. In September, 1974, when the first of a series of cracks in the piping of the more modern BWRs was found at Dresden Unit No. 2.,
the then Atomic Energy Commission (AEC) initiated an intensive investiga-tion to evaluate the cause, extent, and safety .implications of the observed cracking. In January 1975, a special Pipe Cracking Study Group was formed to coordinate and accelerate the staff's continuing invest'iga-tions of the occurrences of pipe cracking. This group included represen-tatives of the Nuclear Regulatory Commission (NRC) and their consultants.
In October, 1975, the Study Group issued a report, NUREG-75/067 "Tech-nical Report, Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants." During the same general time span, the General Electric Company (GE) conducted an indepen-dent evaluation of the cracking occurrences and submitted their findings and recommendations to the NRC. This paper sets forth the NRC technical position based on the information available at this time.
Plant operating history indicates that Type 304 and 316 austenitic stainless steel piping in the reactor coolant pressure boundary of boiling water reactors are susceptible to stress corrosion .cracking.
Studies have shown that such cracking is caused by a combination of
.the presence of significant amounts of oxygen in the coolant, high stresses, and some sensitization of metal adjacent to welds. Such cracks have occurred in the heat affected zones adjacent to welds but are not expected to occur outside these areas, provided that the pipe material is properly annealed.
Pipe runs containing stagnant or low velocity fluids have been observed to be more susceptible to stress corrosion cracking than pipes containing a continuously flowing fluid during plant operation. Historically, these cracks have been identified either by volumetric examination, by leak detection systems, or by visual inspection. Because of the inherent high material toughness of austenitic stainless steel piping, stress corrosion cracking is unlikely to cause a rapidly propagating failure resulting in a loss-of-coolant accident.
Although the probability is extremely low that these stress corrosion cracks will propagate far enough to create a significant safety hazard t
to the public, the presence of such cracks is undesirable. Steps should therefore be taken to minimize stress corrosion cracking in BWR piping systems to eliminate this condition and to improve overall plant reliability.
It is the purpose of this position to set forth acceptable methods to reduce the stress corrosion cracking susceptibility of BWR piping and thereby also provide an increased level of reactor coolant pressure boundary integrity. Recognizing that the most straightforward and
desirable approach or methods may not be practicable, or even possible, for all plants, the bases for varying degrees of conformance to our guidelines are provided. Augmented inservice inspection and leak detec-tion requirements are established for plants that have not fully implemented the provisions .contained in Part II of this document.
I I.
SUMMARY
OF ACCEPTABLE METHODS TO MINIMIZE CRACK SUSCEPTIBILITY The material selection and processing guidelines listed below identify alternative acceptable methods to minimize susceptibility to stress corrosion in BWR pressure boundary piping. It is expected that adoption of these practices will result in a high degree of protection against stress corrosion cracking.
- 1. Corrosion Resistant Materials All pipe and fitting material including weld metal should be of a type and grade that has been shown to be highly resistant to oxygen-assisted stress corrosion in the as-installed condition.
Unstabilized wrought austenitic stainless steel with >0.035/
carbon does not meet this requirement unless all such piping including welds is in the solution annealed condition. The acceptability of alternative materials, processes, or other methods f
to provide an adequate degree of corrosion resistance will be made on a case-by-case basis.
- 2. Corrosion Resistant "Safe Ends" All unstabilized wrought austenitic stainless steel piping with carbon contents >0.0355 should be in the solution annealed condition.
If welds joining these materials are not solution annealed, they should be made between case (or weld overlaid) austenitic stainless steel surfaces (5/ minimum ferrite) or other materials having high resistance to oxygen-assisted stress corrosion. The joint design must be such that any unstabilized wrought austenitic stainless steel containing >0.0351. carbon, which may become sensitized as a result of the welding process, is not exposed to the reactor coolant.
- 3. Other proposed methods to provide protection against stress corrosion cracking will be reviewed on a case by case basis.
Regulatory Guide 1.44 "Control of the Use of Sensitized Stainless Steel",
dated May, 1973 will be revised to provide additional guidance on acceptable practices.
III. INSERVICE INSPECTION AND LEAK DETECTION RE UIREMENTS FOR BWRs WITH VARYING CONFORMANCE TO MATERIAL SELECTION AND PROCESSING GUIDELINES 1.. For plants where all ASME Code Class I reactor coolant pressure boundary piping subject to inservice inspections under Section XI meets the guidelines stated in Part II, no augmented inservice inspection or leak detection requirements are necessary.
- 2. Piping in all other plants is subject to additional inservice inspection and leak detection requirements, as described below.
The degree of inspection of such piping depends on whether the specific piping runs are conforming or non-conforming, and on whether the specific piping runs are classified as "Service
Sensitive". "Service Sensitive" lines are defined as those that have experienced cracking in service, or that are considered to be particularly susceptible to cracking because of high stress, or because they contain relatively stagnant, intermittent, or low flow coolant.
Examples of piping runs considered to be service sensitive include, (but are not limited to): core spray lines, recirculating by-pass lines (or "stub tubes" on plants that have removed the by-pass lines)
CRD hydraulic return lines, isolation condenser lines, and shut down heat exchanger lines.
A. For non-conforming lines that are not service sensitive:
(1) Inservice inspection of the non-conforming lines should be conducted in accordance with the schedule specified in ASME Code,Section XI - Subsection IWB, as required by the applicable examination Categories B-F and B-J, with the exception that the required examination should be completed in no more than 80 months (two thirds of the time perscribed in the schedule in the ASNE Boiler and Pressure Vessel Code Section XI). If examinations conducted during the first 80 month period reveal no incidence of stress corrosion cracking, the examination schedule thereafter can revert to the schedule perscribed in Section XI of the ASME Boiler and Pressure Vessel Code.
The piping areas subject to examination, the method of examina-tion, the allowable indication standards and examination pro-cedures should comply with the requirements of the Edition and Addenda of the ASME Code,Section XI identified as applicable by 10 CFR Part 50, Section 50.55a, Paragraph (g), "Codes and Standards."
(2) The reactor coolant leakage detection system should be operated under the following Technical Specification requirements in order to enhance the discovery of unidentified leakage that may include through-wall cracks developed in austenitic stainless steel piping:
- a. The source of reactor coolant leakage should be identifiable to the extent practical, using leakage detection and collec-tion systems that meet the position described in Section C, Regulatory Position of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," or an acceptable equivalent system.
- b. Plant shutdown should be initiated for inspection and corrective action when the leakage system indicates, within a period of four hours or less, an increase in the rate of unidentified leakage in excess of two gallons per minute, or when the total unidentified leakage attains a r ate of five gallons per minute, whichever occurs first.
- c. Unidentified leakage should include all leakage other than:
- 1. Leakage into closed systems, such as pump seal or valve packing leakage that is captured, metered, and conducted to a sump or collecting tank,
- 2. Leakage into the containment atmosphere from sources that are specifically located and known either not to interfere with the operation of the unidentified leakage detection system, nor not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.
B. For non-conforming lines that are service sensitive:
(1) The leakage detection requirements described in III.A above, should be implemented.
(2) The welds and adjoining areas of bypass piping of the discharge valves in the main recirculation loops, and of the austenitic stainless steel reactor core spray piping up to and including the second isolation valve should be examined at each reactor refueling outage or at other scheduled or unscheduled plant shutdowns. Successive examinations need not be=closer'han six months, if shutdowns occur more frequently than six months. This requirement applies to all bypass lines whether the 4-inch valve is kept open or closed during operation.
In the event these examinations find the piping free .of unacceptable indications for three successive inspections, the examination may be extended to each 36 month period (plus or minus by as much as 12 months) coinciding with a refueling outage. In these cases, the successive examination may be limited to one bypass pipe run, and one reactor core spray piping run ~
(3) The welds and adjoining areas of other service. sensitive piping should be examined on a sampling basis. For example, if a system consists of several branch runs with essentially symmetric piping configurations that perform similar system functions, an acceptable inspection program should include at least one, but not less than 25K, of the similar branch runs. The frequency of such examinations should be as described in 2 above. If unacceptable flaw indications are detected in any branch run, the remaining branch runs among the group should be examined.
in the event the examinations reveal no unacceptable indica-tions within three successive inspections, the examination schedule may revert to the ASME Boiler and Pressure Vessel Code, Section XI, "Inservice Inspection of Nuclear Power Plant
, Components" with the exception that. the required examination should be completed during each 80 month period (two-thirds the time perscribed in the schedule in the ASNE Code Section XI).
(4) The method of examination, the allowable indication standards and examination procedures should comply with the requirements of the Edition and Addenda of the ASME Code,Section XI identified as applicable by 10,CFR Part 50, Section 50.55a, Paragraph (g), "Codes and Standards."
IV. IMPLEMENTATION OF MATERIAL SELECTION AND PROCESSING GUIDELINES
- 1. For plants that apply for a construction permit after the issue date of this document, all ASME Code Class I reactor coolant pressure boundary lines should conform to the guidelines stated in Part II.*
- 2. For plants unde'r review, but for which a construction permit has not yet been issued, all service sensitive lines should conform to the guidelines stated in Part II. Other ASME Code Class I reactor, coolant pressure boundary lines should conform to Part I
II 4
to,the extent practicable.
- 3. For plants that have been issued a construction permit, ASME Code Class I reactor coolant pressure boundary lines should conform to the guidelines stated in Part II to the extent practicable.
- After revision, Regulatory Guide 1 ~ 44 may be used as guidance for acceptable materials, process, or other methods.
- 4. For plants that have been issued an operating license, service sensitive lines should be modified to conform to the guidelines stated in Part II, to the extent practicable. Lines in which cracking is experienced should be replaced with piping that conforms to the guidelines stated in Part II.
V.," GENERAL RECOMMENDATIONS The measures outlines in Part II of this document provide for positive actions that are consistent with the current technology. The implemen-tation of these actions should markedly reduce the susceptibility to stress corrosion cracking in BWRs. It is recognized that additional techniques are available to limit the corrosion potential of BWR coolant pressure boundary materials and improve the overall system integrity.
These include plant design and operational considerations to reduce system exposure to potentially aggressive environment, improve material fabrication and welding techniques and provisions for volumetric inspec-tion capability in the design of weld joints. Specifically, considera-tion should be given to:
- 1. Minimizing the total extent of the coolant pressure boundary with special emphasis on stagnant or low flow lines.
- 2. Reducing the oxygen content of the primary coolant.
0
DISTRIBUTION W/ENCLOSURE:
Docket Fi le PDR Local PDR LWR 84 File M. Service Project Manager W F'ane S. A. Varga DISTRIBUTION W/0 ENCLOSURE:
R'. Boyd R. DeYoung D. Vassallo F. Williams H. 'Sm'ith R. Mattson J. Knight S. Pawlicki H. Conrad I 8 E (3)
ELD L. Crocker K. Goller bcc:
TIC ACRS (15)
NSIC
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