Letter Sequence RAI |
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MONTHYEARML0415604642004-05-27027 May 2004 Proposed Technical Specification Changes Implementation of Alternate Source Term Project stage: Other ML0423304632004-08-20020 August 2004 Facsimile Transmission, Draft RAI to Be Discussed in an Upcoming Conference Call Project stage: Draft RAI ML0426602502004-09-22022 September 2004 Facsimile Transmission, Draft Request for Additional Information (RAI) to Be Discussed in an Upcoming Conference Call Project stage: Draft RAI ML0435100392004-12-29029 December 2004 Request for Additional Information Regarding Use of Alternate Source Term Project stage: RAI ML0500402642005-02-0909 February 2005 RAI, Use of Alternate Source Term Project stage: RAI ML0515400932005-04-25025 April 2005 Summary of Meeting with Dominion Nuclear Connecticut, Inc.,Surry & North Anna, Units 1 & 2, & Millstone, Units 2 & 3, Discuss Current Licensing Issues with Dominion. Enclosure 3 - Top Ten Millstone Licensing Tasks Project stage: Meeting ML0530704822005-11-18018 November 2005 RAI Regarding Use of Alternate Source Term Project stage: RAI ML0619901352006-09-15015 September 2006 Issuance of Amendment Alternate Source Term Project stage: Approval 2004-09-22
[Table View] |
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Category:Code Relief or Alternative
MONTHYEARML23188A0432023-07-31031 July 2023 Authorization and Safety Evaluation for Alternative Request No. IR-04-11 ML22201A5082022-07-28028 July 2022 Authorization and Safety Evaluation for Alternative Request No. IR-04-09 ML21284A0062021-10-29029 October 2021 Authorization and Safety Evaluation for Alternative Request No. RR-05-04 and IR-4-02 ML21174A0202021-08-0202 August 2021 Relief Request for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML21167A3552021-07-16016 July 2021 Authorization and Safety Evaluation for Alternative Request No. RR-05-06 ML20312A0022020-12-10010 December 2020 Relief Request for Limited Coverage Examinations Performed in the Third 10-Year Inservice Inspection Interval (EPID L-2020-LLR-0027 Through L-2020-LLR-0032) ML20312A0012020-12-10010 December 2020 Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML20287A4712020-10-20020 October 2020 Proposed Alternative RR-05-05 to the Requirements of the ASME Code Containment Unbonded Post-Tensioning System Inservice Inspection Requirements ML20189A2062020-07-16016 July 2020 Relief Request IR-4-03 Concerning Non-Code Methodology to Demonstrate Structural Integrity of Class 3 Moderate-Energy Piping ML20080K5082020-03-24024 March 2020 Alternative Request RR-05-03 for the Fifth 10-Year Inservice Inspection Interval ML19340A0012019-12-18018 December 2019 Proposed Alternative Request IR-4-01 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML19338G3722019-12-18018 December 2019 Alternative Requests RR-05-01 and RR-05-02 for the Fifth 10-Year Inservice Inspection Interval ML19340A0002019-12-13013 December 2019 Relief Request IR-3-39, Proposed Alternative to ASME Code Weld Preheat Requirements ML19275D2522019-09-24024 September 2019 Alternative Request RR-05-03, Extension of ASME Code Case N-770-2 Volumetric Inspection Frequency for Reactor Coolant Pump Inlet and Outlet Nozzle Dissimilar Metal Butt Welds ML19098A0342019-04-30030 April 2019 Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3, and Quad Cities, Units 1 and 2 - Revision to Approved Alternative to Use BWR Vessel and Internal Proj Guidelins ML18290A6022018-11-13013 November 2018 Alternative Requests Related to the Fifth and Fourth 10-Year Interval Pump, Valve, and Snubber Inservice Testing Programs, Respectively (EPID L-2018-LLR- 0012 Through EPID L-2018-LLR-0022) ML18252A0032018-09-17017 September 2018 Alternative Requests RR-04-27 and IR-3-38 for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography ML18066A5222018-02-28028 February 2018 Proposed Alternative Requests RR-04-27 and IR-3-38 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography in Accordance with 10 CFR 50.55(z)(1) ML17132A1872017-05-25025 May 2017 Alternative Relief Requests RR-04-24 and IR-3-30: Reactor Pressure Vessel Threads in Flange ML17135A2962017-05-25025 May 2017 Alternative Relief Request RR-04-25 Boric Acid Pump P-19B Stuffing Box Cover ML17122A3742017-05-0303 May 2017 Alternative Relief Request RR-04-26 Boric Acid Pump P-19B Stuffing Box Cover ML17125A2522017-04-28028 April 2017 ASME Section XI Relief Request RR-04-26 ML17090A1102017-03-29029 March 2017 ASME Section XI Relief Request RR-04-25 ML16363A0892017-01-23023 January 2017 Alternative Relief RR-04-23 and IR-3-28 from the Requirements of ASME Code Section XI Regarding Radiographic Examinations ML16277A6782016-10-18018 October 2016 Alternative Request RR-04-22 to Implement Extended Reactor Vessel Inservice Inspection Interval ML16172A1352016-07-13013 July 2016 Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inservice Inspection Interval ML16038A0012016-02-16016 February 2016 Alternative Request IR-3-27 for Implementation of Extended Reactor Vessel Inservice Inspection Interval ML15257A0052015-09-21021 September 2015 Relief from the Requirements of ASME Code Section XI Regarding Radiographic Examinations ML15216A3632015-07-30030 July 2015 ASME Section XI Inservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inspection Interval ML15216A3592015-07-30030 July 2015 ASME Section XI Inservice Inspection Program Relief Requests for Limited Coverage Examinations Performed in the Fourth 10-Year Inspection Interval ML15082A4092015-04-24024 April 2015 Alternative Use of Weld Overlay as Repair and Mitigation Technique ML14217A2032014-09-0404 September 2014 Relief from the Requirements of the ASME Code Section XI, Requirements for Repair/Replacement of Class 3 Service Water Valves (Tac No. MF1314) ML14163A5862014-07-10010 July 2014 Relief from the Inservice Testing Requirements of American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants ML14091A9732014-04-0404 April 2014 Issuance of Relief Request RR-04-16 Regarding Use of Encoded Phased Array Ultrasonic Examination in Lieu of Radiography ML1130001002011-11-0909 November 2011 Issuance of Relief Request RR-04-12 Regarding the Temporary Non-Code Compliant Condition of the Class 3 Service Water System 10 Inch Emergency Diesel Generator Supply Piping Flange ML11234A0772011-08-19019 August 2011 Relief Request RR-04-12 for the Temporary Non-Code Compliant Condition of the Class 3 Service Water System 10 Inch Emergency Diesel Generator Supply Piping Flange ML1118810292011-07-27027 July 2011 Issuance of Relief Request RR-04-04 Regarding Use of Alternative Pressure Testing Requirements ML1118706002011-07-22022 July 2011 Issuance of Relief Request RR-04-05 Regarding Use of Alternative Pressure Testing Requirements ML1106800802011-03-24024 March 2011 Issuance of Relief Request lR-3-14 -- Use of Risk-Informed Inservice Inspection Program Plan ML1014700992010-06-11011 June 2010 Issuance of Relief Request IR-3-05 Regarding Use of American Society of Mechanical Engineering Code, Section XI, Appendix Viii ML1012412192010-05-13013 May 2010 Relief Request IR-3-02 Regarding Use of American Society of Mechanical Engineering Code, Section XI, 2004 Edition ML1010400422010-04-29029 April 2010 Issuance of Relief Request lR-3-11 Regarding Use of American Society of Mechanical Engineering Code, Section XI, 2004 Edition ML0935702372010-04-26026 April 2010 Issuance of Relief Request RR-89-67 Regarding the Repair of Reactor Coolant Pump Seal Cooler Return Tubing and Weld ML1011301872010-04-19019 April 2010 ASME Section XI Inservice Inspection Program Relief Request for Limited Coverage Examinations Performed in the Second 10-Year Inspection Interval ML1008407382010-04-15015 April 2010 Issuance of Relief Requests IR-3-13 Regarding Use of American Society of Mechanical Engineering Code, Section XI, 2004 Edition ML1006801182010-04-0606 April 2010 Issuance of Relief Request IR-3-01 Regarding Use of American Society of Mechanical Engineering Code, Section XI, Appendix Viii ML1009002002010-03-30030 March 2010 Relief Requests RR-04-02, Alternative VT-2 Pressure Testing Requirements for the Lower Portion of the Reactor Pressure Vessel, and RR-04-03, Alternative Evaluation Criteria for Code Case N-513-2, Temporary Acceptance of Flaws In.. ML1006404462010-03-12012 March 2010 Issuance Relief ML1005402202010-02-19019 February 2010 Relief Request IR-3-01 Supplemental Information Re Snubber Inspection and Testing for Third 10-Year Interval ML0923901412009-08-24024 August 2009 Relief Request for Millstone Power Station, Unit 3, Relief Request IR-3-04, Response to Request for Additional Information for Alternative Brazed Joint Assessment Methodology 2023-07-31
[Table view] Category:Letter
MONTHYEARIR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 IR 05000336/20234022024-01-30030 January 2024 Security Baseline Inspection Report 05000336/2023402 and 05000423/2023402 (Cover Letter Only) ML24030A7522024-01-30030 January 2024 Technical Specification Bases Pages ML23341A0172024-01-12012 January 2024 Issuance of Amendment No. 288 Revision to Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures IR 05000336/20234402024-01-11011 January 2024 Special Inspection Report 05000336/2023440 and 05000423/2023440 (Cover Letter Only) ML24004A1052024-01-0404 January 2024 Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000336/2024010 & 05000423/2024010 ML23361A0942023-12-21021 December 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies and Core Operating Limits Report . ML23283A3052023-12-20020 December 2023 Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary) ML23361A0312023-12-20020 December 2023 Intent to Pursue Subsequent License Renewal ML23352A0202023-12-18018 December 2023 Senior Reactor and Reactor Operator Initial License Examinations ML23334A2242023-11-30030 November 2023 Request for Exemption from Enhanced Weapons Firearms Background Checks, and Security Event Notifications Implementation ML23324A4302023-11-20020 November 2023 Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment 15 ML23324A4222023-11-20020 November 2023 Reactor Vessel Internals Inspections Aging Management Program Submittal Related to License Renewal Commitment 13 ML23317A2702023-11-13013 November 2023 Core Operating Limits Report, Cycle 23 IR 05000336/20230032023-11-0606 November 2023 Integrated Inspection Report 05000336/2023003 and 05000423/2023003 ML23298A1652023-10-26026 October 2023 Requalification Program Inspection IR 05000336/20234202023-10-0404 October 2023 Security Inspection Report 05000336/2023420 and 05000423/2023420 ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23226A0052023-09-26026 September 2023 Issuance of Amendment No. 287 Supplement to Spent Fuel Pool Criticality Safety Analysis IR 05000245/20230012023-09-19019 September 2023 Safstor Inspection Report 05000245/2023001 IR 05000336/20230102023-09-0808 September 2023 Commercial Grade Dedication Report 05000336/2023010 and 05000423/2023010 IR 05000336/20230052023-08-31031 August 2023 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Report 05000336/2023005 and 05000423/2023005) ML23248A2132023-08-30030 August 2023 Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature. ML23242A0142023-08-30030 August 2023 Operator Licensing Examination Approval ML23223A0552023-08-18018 August 2023 Request for Withholding Information from Public Disclosure for License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies, and COLR Related to Framatome Gaia Fuel ML23223A0482023-08-18018 August 2023 Request for Withholding Information from Public Disclosure for License Amendment Request to Use Framatome Small Break and Realistic Large Break LOCA Evaluation Methodologies for Establishing COLR Limits IR 05000336/20230022023-08-0909 August 2023 Integrated Inspection Report 05000336/2023002 and 05000423/2023002 ML23188A0432023-07-31031 July 2023 Authorization and Safety Evaluation for Alternative Request No. IR-04-11 ML23208A0922023-07-26026 July 2023 Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of Framatome ORFEO-GAIA and OORFE-NMGRID CHF Correlations in the Dominion Energy Vipre-D Computer Code Response ML23188A0202023-07-26026 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML23207A1102023-07-26026 July 2023 NRC Regulatory Issues Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations IR 05000336/20234012023-07-17017 July 2023 Material Control and Accounting Program Inspection Report 05000336/2023401 and 05000423/2023401 - (Cover Letter Only) ML23175A0052023-07-12012 July 2023 Alternative Request P-07 for Pump Periodic Verification Testing Program for Containment Recirculation Spray System Pumps ML23193A8562023-06-28028 June 2023 Submittal of Updates to the Final Safety Analysis Reports ML23178A1682023-06-26026 June 2023 2022 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML23153A1732023-06-16016 June 2023 Correction to Amendment Nos. 346 & 286 Millstone, 294 & 277 North Anna, 311 & 311 Surry, and 225 Summer to Revise Technical Specifications to Adopt TSTF-554,Rev Reactor Coolant Leakage Requirement ML23151A0742023-06-12012 June 2023 Review of the Spring 2022 Steam Generator Tube Inspection Report ML23159A2202023-06-0808 June 2023 Associated Independent Spent Fuel Storage Installation Revision to Emergence Plan - Report of Changes 2024-08-13
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML23318A0952023-11-14014 November 2023 Request for Additional Information (E-mail Dated 11/14/2023) LAR to Revise TSs Related to Framatome Gaia Fuel ML23199A2832023-07-18018 July 2023 Request for Additional Information (E-mail Dated 7/18/2023) LAR to Revise the Pressure-Temperature Limits ML23171B0522023-06-20020 June 2023 NRR E-mail Capture - RAI for Review of Appendix F to DOM-NAF-2-P (Redacted) ML23089A1882023-03-29029 March 2023 Request for Additional Information (E-mail Dated 3/29/2023) Spring 2022 Steam Generator Tube Inspection Report ML23073A1802023-03-14014 March 2023 Notification of Commercial Grade Dedication Inspection (05000336/2023010 and 05000423/2023010) and Request for Information ML22251A1292022-09-0909 September 2022 Request for Additional Information Related to Response to Generic Letter 2004-04 ML22242A0452022-08-23023 August 2022 NRR E-mail Capture - Dominion GL 04-02 Response Draft RAIs (L-2017-LRC-0000) ML21225A5022021-08-13013 August 2021 Request for Additional Information for North Anna Power Station, Units 1 and 2, Surry Power Station, Units 1 and 2, and Millstone Power Station, Units 2 and 3, Appendix E of Fleet Report Com-Naf-2 (EPID:L02021-LLT-0000) Public Version ML21195A2442021-07-13013 July 2021 NRR E-mail Capture - Draft RAI Appendix E to Topical Report DOM-NAF-2 Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code ML21182A2902021-07-0101 July 2021 7/1/2021 E-mail - Request for Additional Information Alternative Frequency to Supplemental Valve Position Verification Testing Requirements ML21112A3082021-04-22022 April 2021 Request for Additional Information (E-mail Dated 4/22/2021) License Amendment Request for Measurement Uncertainty Recapture Power Uprate ML21105A1152021-04-15015 April 2021 Request for Additional Information (e-mail Dated 4/15/2021) Proposed LAR for Addition of Analytical Methodology to the Core Operating Limits Report for a LBLOCA ML21078A0332021-03-18018 March 2021 NRR E-mail Capture - Millstone Power Station, Unit 2 - Request for Additional Information - Proposed LAR to Revise TSs for Steam Generator Inspection Frequency ML21076A4622021-03-17017 March 2021 NRR E-mail Capture - Draft RAI for Versatile Internals and Component Program for Reactors Vipre Code Review (L-2021-LLT-0000) ML21077A1652021-02-26026 February 2021 (02/26/2021 E-mail) Draft Request for Additional Information LAR to Revise Technical Specifications for Steam Generator Inspection Frequency ML21034A5762021-02-0303 February 2021 Request for Additional Information (e-mail Dated 1/3/2021) Alternative Request RR-05-06 Inservice Inspection Interval Extension for Steam Generator Weld Inspections ML20246G5612020-09-0202 September 2020 Request for Additional Information (E-mail Dated 9/2/2020) License Amendment Request for a One-Time Deferral of the Steam Generator Inspections ML20244A2902020-08-31031 August 2020 Request for Additional Information (E-mail Dated 8/31/2020) License Amendment Request to Revise Battery Surveillance Requirements ML20231A7172020-08-18018 August 2020 Request for Additional Information (e-mail Dated 8/18/2020) License Amendment Request to Revise TS Table 3.3-11 Accident Monitoring Instrumentation ML20184A0052020-07-0202 July 2020 Request for Additional Information (E-mail Dated 7/2/2020) Relief Request IR-3-33 for Limited Coverage Examiniations ML20075A0322020-03-15015 March 2020 Request for Additional Information (E-mail Dated 3/15/2020) Alternative Request IR-4-03 for Use of an Alternative Non-Code Methodology to Demonstrate Structural Integrity ML20050F9212020-02-19019 February 2020 Request for Additional Information (E-mail Dated 2/19/2020) Proposed LAR to Revise TS 3.8.1.1, AC Sources Operating ML20023D0372020-01-23023 January 2020 Request for Additional Information (E-mail Dated 1/23/2020) Proposed LAR to Revise TS 3.8.1.1, AC Sources Operating ML19361A0382019-12-23023 December 2019 12/23/19 E-mail from R.Guzman to S. Sinha - Draft Request for Additional Information Planned Audit of LAR to Revise TS 3.8.1.1 ML19277B9082019-10-0707 October 2019 Supplemental Information Needed for Acceptance of Requested Licensing Action Extension of Technical Specification 3.8.1.1, A.C. Sources - Operating, Allowed Outage Time ML19248D7552019-09-0505 September 2019 Request for Additional Information Via E-mail from R.Guzman to S. Sinha Proposed License Amendment Request to Adopt 10 CFR 50.69 ML19169A0962019-06-17017 June 2019 NRR E-mail Capture - Request for Additional Information Emergency Action Level Scheme Change for Millstone Power Station - Units 1, 2, and 3, North Anna Power Station - Units 1 and 2, and Surry Power Station - Units 1 and 2 ML19162A0052019-06-10010 June 2019 NRR E-mail Capture - Draft RAIs for Dominion EAL Scheme Change ML19042A1822019-02-11011 February 2019 Request for Additional Information LAR for TS Changes to Spent Fuel Storage and New Fuel Storage, 02/11/2019 E-mail from R. Guzman to S. Sinha ML18291B3262018-10-18018 October 2018 Request for Additional Information LAR Regarding Proposed Technical Specification Changes for Spent Fuel Pool Storage and New Fuel Storage, E-mail from R.Guzman to S.Sinha ML18277A1762018-10-0404 October 2018 Request for Additional Information LAR to Revise TS Action Statement for Loss of Control Building Inlet Ventilation Rad Monitor Instrumentation Channels (e-mail from R.Guzman to M.Whitlock, EPID: L-2018-LLA-0099) ML18213A1152018-08-0101 August 2018 Request for Additional Information Alternative Request P-06 'C' Charging Pump (08/01/2018 E-mail from R. Guzman to M. Whitlock) ML18171A1532018-06-20020 June 2018 Request for Additional Information Via E-mail from R.Guzman to W.Craft Alternative Requests for Inservice Testing Program (E-mail from R.Guzman to W.Craft) (EPID: L-2018-LLR-0012-LLR-0022) ML18137A3132018-05-17017 May 2018 Request for Additional Information Via E-mail from R.Guzman to W.Craft Proposed License Amendment Request to Revise ILRT Type a and Type C Test Intervals ML17297A8472017-10-27027 October 2017 Request for Supplemental Information Regarding Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools (CAC No. MF9430; EPID L-2016-LRC-0001) ML17033B6142017-02-0202 February 2017 Request for Additional Information, 2017/02/02 E-mail from R.Guzman to W.Craft Proposed Alternative RR-04-24 and IR-3-30 for the Elimination of Reactor Pressure Vessel Threads in Flange Examination (CAC Nos. MF8468/MF8469) ML17005A3282017-01-0505 January 2017 MPS2 EOC-23 and MPS3 EOC-17 Steam Generator Tube Inspection Reports - Requests for Additional Information ML16260A0062016-09-16016 September 2016 2016/09/16 E-mail from R. Guzman to W. Craft Request for Additional Information - License Amendment Request Realistic Large Break Loss of Coolant Accident Analysis (MF7761) ML16267A0022016-09-12012 September 2016 2016/09/12 E-mail from R.Guzman to W.Craft Request for Additional Information - LAR to Revise ECCS TS and FSAR Ch. 14 to Remove Charging (MF7297) ML16133A3562016-05-12012 May 2016 E-mail from R. Guzman to C. Sly Request for Additional Information Supporting License Amendment Request to Remove Charging Credit ML16068A4142016-03-0808 March 2016 Request for Additional Information - Relief Request RR-04-19 Examination Cat R-A, Risk-Informed Piping Examinations (MF6569) ML16055A5302016-02-24024 February 2016 2016/02/24 E-mail from R.Guzman to W.Craft Request for Additional Information - LAR to Adopt Dominion Core Design and Safety Analysis Methods (MF6251) ML16047A1712016-02-16016 February 2016 2016/02/16 E-mail from R.Guzman to M. Whitlock Request for Additional Information - Small Break Loss of Coolant Accident Re-Analysis (MF6700) ML16043A5232016-02-12012 February 2016 Request for Additional Information 2016/02/12 E-mail from R. Guzman to M.Whitlock Relief Requests IR-3-22 and IR-3-23 (CAC Nos. MF6573 and MF6574) ML16013A4772016-01-13013 January 2016 NRR E-mail Capture - Millstone Power Station, Units 2 and 3 - Request for Additional Information License Amendment Request for Removal of Severe Line Outage Detection (Slod) from the Offsite Power System - (MF6430, MF6431) ML16011A5172016-01-11011 January 2016 NRR E-mail Capture - Millstone Power Station, Unit 3, Reactor Vessel Surveillance Capsule Withdrawal Schedule - Request for Additional Information ML15349A8312016-01-0808 January 2016 Request for Additional Information License Amendment Request to Adopt Dominion Core Design and Safety Analysis Methods ML15337A1512015-12-0303 December 2015 NRR Email Capture - Millstone Unit 2 Request for Additional Information ML15279A1092015-10-0606 October 2015 NRR e-mail Capture - Millstone Power Station, Unit 3, NRC Staff Request for Additional Information - Relief Request Nos. IR-3-19,3-20,3-21 ML15275A1432015-10-0202 October 2015 2015/10/02 E-mail from R.Guzman to W.Craft - Follow-up Request for Additional Information TSTF-523, Revision 2, GL 2008-01, Managing Gas Accumulation 2023-07-18
[Table view] |
Text
February 9, 2005 Mr. David A. Christian Senior Vice President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 3 - REQUEST FOR ADDITIONAL INFORMATION REGARDING USE OF ALTERNATE SOURCE TERM (TAC NO. MC3333)
Dear Mr. Christian:
By letter dated May 27, 2004, you requested an amendment to the Millstone Power Station, Unit No. 3 (MP3) Technical Specifications. The amendment would allow for use of an alternate source term for MP3.
The Nuclear Regulatory Commission staff is reviewing your application and has determined that additional information is required. The enclosed request for additional information (RAI) was forwarded electronically to Mr. Paul Willoughby of your staff on December 14, 2004. I understand that you intend to respond to this RAI by February 28, 2005. If you have any questions, I can be reached at (301) 415-1494.
Sincerely,
/RA/
George F. Wunder, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosure:
As stated cc w/encl: See next page
Millstone Power Station, Unit No. 3 cc:
Lillilan M. Cuoco, Esquire Mr. William D. Meinert Senior Counsel Nuclear Engineer Dominion Resources Services, Inc. Massachusetts Municipal Wholesale Building 475, 5th Floor Electric Company Rope Ferry Road Moody Street Waterford, CT 06385 P.O. Box 426 Ludlow, MA 01056 Edward L. Wilds, Jr., Ph.D.
Director, Division of Radiation Mr. J. Alan Price Department of Environmental Protection Site Vice President 79 Elm Street Dominion Nuclear Connecticut, Inc.
Hartford, CT 06106-5127 Building 475, 5th Floor Rope Ferry Road Regional Administrator, Region I Waterford, CT 06385 U.S. Nuclear Regulatory Commission 475 Allendale Road Mr. Chris Funderburk King of Prussia, PA 19406 Director, Nuclear Licensing and Operations Support First Selectmen Dominion Resources Services, Inc.
Town of Waterford 5000 Dominion Boulevard 15 Rope Ferry Road Glen Allen, VA 23060-6711 Waterford, CT 06385 Mr. David W. Dodson Mr. John Markowicz Licensing Supervisor Co-Chair Dominion Nuclear Connecticut, Inc.
Nuclear Energy Advisory Council Building 475, 5th Floor 9 Susan Terrace Rope Ferry Road Waterford, CT 06385 Waterford, CT 06385 Mr. Evan W. Woollacott Co-Chair Nuclear Energy Advisory Council 128 Terry's Plain Road Simsbury, CT 06070 Senior Resident Inspector Millstone Power Station c/o U.S. Nuclear Regulatory Commission P. O. Box 513 Niantic, CT 06357 Ms. Nancy Burton 147 Cross Highway Redding Ridge, CT 00870
ML050040264 *See previous concurrence OFFICE PDI-2/PM PDI-2/LA SPSB/SC PDI-2/SC NAME GWunder CRaynor RDennig DRoberts DATE 2/9/05 2/9/05 1/19/05 2/9/05
REQUEST FOR ADDITIONAL INFORMATION MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423 USE OF ALTERNATE SOURCE TERM (AST)
General Questions
- 1. What is the basis for assuming that the control room ventilation system isolation timing is the same for the fuel handling accident as for the entire spectrum of accidents at Millstone Power Station, Unit No. 3 (MP3)?
- 2. The control room analyses need to consider the doses to the Millstone Power Station, Unit No. 2 (MP2) control room operators. Provide those consequences and the operating mode and conditions of the control room ventilation systems in response to the MP3 accident. Provide the manner in which the MP2 control room ventilation systems are initiated for the MP3 accident and the timing associated with that initiation.
- 3. What is the impact of MP2 accidents on the doses to the MP3 control room operators as a result of these changes to MP3?
- 4. What is the basis for two-spray operation during the recirculation phase of the loss-of-coolant accident (LOCA)?
- 5. Describe what the neutral operating conditions are for the MP3 control room. What is the basis for assuming 350 cfm of unfiltered inleakage? What were the American Society for Testing Materials E741 test results when the control room was tested in this condition?
Section 1.3 Analysis Assumptions and Key Parameter Values
- 1. There is a discussion in Section 1.3.1 about the radiological consequences of the waste gas system failure and the radioactive liquid waste system leak or failure (atmospheric release). It is stated that these analyses will be retained in Chapter 11 of the MP3 final safety analysis report and that the whole-body and thyroid doses will be converted to total effective dose equivalent (TEDE). These two analyses originate out of Standard Review Plans (SRPs) 15.7.1 and 15.7.2 of NUREG-75/087. The acceptance criteria for these two events was 500 mrem whole-body which was the Title 10 of the Code of Federal Regulations (10 CFR) Part 20 limit at the time. Additional guidance on the consequences of a waste gas decay tank leak is provided in Branch Technical Position 11-5 of SRP 11.3. If there is to be a conversion to TEDE, the acceptance criterion for the conversion should be based upon the criterion for the present 10 CFR Part 20, 100 mrem TEDE rather than the Part 100 criterion. In addition, any change to a TEDE dose criterion should not be limited to only the consequences from noble gases and iodides but also include the other isotopes associated with Regulatory Guide (RG) 1.183 and the AST.
- 2. Provide the basis for increasing the acceptance criterion for methyl iodine penetration of the control room emergency air filtration system in surveillance requirements (SRs) 3/4.7.7.c.2 and 3/4.7.7.d from 2.5% to 5%.
- 3. What is the basis for the high efficiency particulate air filter and charcoal adsorber efficiencies for the control room emergency ventilation system and why are they inconsistent with the values of RG 1.52?
Section 3.1 Design Basis LOCA Reanalysis
- 1. Section 3.1.4 provides a discussion of the manner in which containment sprays are utilized during the course of a LOCA and the manner in which their use was modeled in the determination of the radiological consequences. The manner in which the information in this Section and in Table 3.1-4 were incorporated in the calculations, however, is unclear. Please provide a table which notes the time post-LOCA when the sprays were initiated and stopped, the spray removal coefficient for elemental and particulate forms of iodine utilized during these periods, when the sprays were terminated, and the decontamination factor (DF) at the time of termination.
- 2. Was the Powers model used for aerosols in the sprayed region after the quench spray was secured and after it was assumed that there was no removal of iodine by the sprays?
- 3. How is the tripping of the non-nuclear safety grade exhaust fan accounted for in the assessment of the control room operators dose?
- 4. In Table 3.1-4, what is the rationale for concluding that the containments leak rate will be reduced after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the control room consequences but only after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the offsite consequences?
- 5. Provide your calculations demonstrating the time at which the particulate and elemental iodine DF are achieved. If these calculations do not describe how the DF is defined, provide the definition.
- 6. At what time does the spray removal coefficients for particulate become 1.27 and 1.61?
What regions of containment do these spray removal coefficients apply?
- 7. Table 3.1-5 provides information on contaminated inflow to the rad waste storage tank (RWST). Is the time information in the Table the duration of the inflow from a given source or is it the time the leak begins post-accident?
- 8. How full are each of the lines to the RWST when the backflow begins? What is the volume of each line?
- 9. Table 3.1-6 provides information on contaminated inflow to the RWST. Similar to Table 3.1-5, it provides a summary of times. However, as with Table 3.1-6, it is unclear whether the time information in the Table is the duration of the inflow from a given source or is the time the leak begins post-accident.
It is also unclear what SIS R is and why the volume for the SIS R is shown to be zero.
Clarify this table.
- 10. Provide the calculation which demonstrates how the RWST airflow rate of 8.7 cfm was determined.
- 11. What iodine isotopes contributed to the value of 10,000 grams of iodine in the core?
Provide the calculation for this number.
- 12. The discussion of backleakage to the RWST in Section 3.1.5.3 of the application describes the amount of leakage and the resulting iodine concentration in the tank. The discussion does not provide a value for the pH of the solution in the RWST or the amount of iodine in the air space. Since the iodine concentration and the pH influence the potential release of radioactive iodine, please discuss the pH in the RWST and the amount of iodine released to the air space in the tank during the 30-day LOCA period. If pH calculations were performed, please provide the results, inputs, and explanation of the inputs. If pH calculations were not performed, please explain why this was considered unnecessary for assuring that elemental iodine would not be evolved from the tank.
Section 3.2 Fuel Handling Accident (FHA)
- 1. What is the control building isolation signal based upon?
- 2. It is proposed that no credit is taken for the control room envelope (CRE) pressurization system. Therefore, the system can be deleted from the technical specification (TS). In fact, credit is taken for the pressurization system. The analyses which support this amendment request base the inleakage characteristics of the CRE for the first hour of the FHA and other accidents upon the pressurization system working. If the pressurization system is not operating, the inleakage characteristics of the CRE will be different. What is the inleakage characteristics of the CRE with the normal control room ventilation system operating, with no system operating, or with the pressurization system operating? Which results in the worst case condition? Is the proposed deletion of the pressurization system from the TSs still appropriate?
- 3. What is the limiting CRE inleakage characteristics during this accident? Did the determination of the limiting inleakage characteristics consider the FHA scenario with and without the loss of offsite power? This would seem important since the inleakage characteristics of the CRE will be affected by the ventilation systems operating internal and external to the CRE.
Section 3.3 Steam Generator Tube Rupture (SGTR) Accident
- 1. Why are there no steam releases from 0.8183 hours0.0947 days <br />2.273 hours <br />0.0135 weeks <br />0.00311 months <br /> until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />? What precipitates the steaming of the affected steam generator at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and not before?
- 2. What isolates the control room on a SGTR?
- 3. Is the CRE pressurization system assumed to operate in the event of a SGTR?
- 4. Was the iodine spike assumed to occur during the entire duration of the accident?
- 5. Why was it assumed that there was no flashing of the break flow from 0.927 hours0.0107 days <br />0.258 hours <br />0.00153 weeks <br />3.527235e-4 months <br /> until 1.554 hours0.00641 days <br />0.154 hours <br />9.160053e-4 weeks <br />2.10797e-4 months <br />?
Section 3.4 Main Steam Line Break Analysis What was the duration of the iodine spike for this accident?
Section 3.5 Locked Rotor Accident
- 1. Has the closure of the stuck-open atmosphere dump valve within 20 minutes been demonstrated to be a reasonable action for the control room operators, or has it been maintained because it is part of the existing licensing basis?
- 2. Table 3.5-1 provides, as a function of time, the total steam flows to atmosphere and the mass flow rates from the three intact steam generators. The total steam flow to atmosphere seems to reflect the release from two steam generators and not three. How are these numbers calculated?
Section 3.6 Control Rod Ejection Accident
- 1. What is the basis for assuming that the primary system pressure is less than the secondary side after 20 minutes?
- 2. Why are there no steam releases from 20 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />?
Section 7.0 TS and Bases Change
- 1. The use of the thyroid dose conversion factors is acceptable; however, the staff believes that the proposed definition of dose equivalent I-131 should be changed. The licensee should consider whether they agree to the following definition change (note the words in bold):
DOSE EQUIVALENT I-131 shall be that concentration of I-131 (micro curie per gram) which alone would produce the same TEDE dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under Inhalation in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."
A change to SRs c.2 and d of TS 3/4.7.7, Control Room Emergency Air Filtration System, was proposed which would increase the acceptance criteria for methyl iodide penetration from 2.5% to 5%. The basis for this change appears to be revised AST analysis which decreased the adsorber efficiency for the elemental and organic forms of iodine to 90% and 70%, respectively. This proposed change in adsorber efficiency and associated change to the SRs are unacceptable. RG 1.52 details the adsorber
efficiencies which should be assigned based upon the various depths of charcoal. The adsorber efficiencies selected for the elemental and organic forms of iodide are inconsistent with this guidance. Therefore, this decrease in adsorber efficiency cannot be approved nor can the acceptance criteria for SRs c.2 and d, unless justification is provided which satisfactorily explains the basis for the change.
It has been proposed that TS 3/4.7.8, Control Room Envelope Pressurization System, be deleted since it is no longer credited in the accident analyses for AST. All of the AST accident analyses have assumed operation of the CRE pressurization system. It appears that the analyses of the control room operators dose assumed that the control room would be in a neutral pressure condition even though the CRE should be pressurized by the CRE pressurization system. The AST amendment request did not elaborate on what control room ventilation systems would be operating if the CRE pressurization system did not operate. The assessment of the control room operators dose must account for the manner of operation during the course of the action. Since it is proposed to delete the CRE pressurization system as an engineered safeguards features (ESF) system intended to protect the control room operators in the event of a radiological accident, an assessment needs to be performed which identifies the various configuration of control room ventilation systems that may be functioning following a radiological event. For example, it has been proposed that the CRE pressurization system, while no longer ESF grade, will be functioning during the first hour post-accident. What is the CREs inleakage characteristics when it is? What if the CRE pressurization system does not operate? Now what control room ventilation systems are operating and what is the CREs inleakage characteristics when they are operating?
What is the effect upon CRE cooling under these scenarios?
The following are comments on the BASES Section.
- 1. In 3/4.6.1.1 General Design Criteria (GDC) 19 should not be replaced with RG 1.183. GDC 19 does apply.
- 2. In 3/4.7.7 it specifies that the control room emergency ventilation system may be operated in either the isolation and recirculating mode of operation or in the pressurization mode of operation. However, the AST analyses reflects only a pressurization and recirculating mode of operation. The BASES should reflect the actual intended modes of operation for the system. If these other modes of operation are possible then the dose consequences associated with these other modes need to be provided as well as the inleakage characteristics of the CRE during these modes of operation. In addition, in various sections there is reference to modes of operation which are inconsistent with the mode of operation described in the AST analyses.
Examples include the background and SR 4.7.7.e.2.