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MONTHYEARULNRC-05521, Request for Extension of Corrective Actions Completion Date for NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.2008-06-24024 June 2008 Request for Extension of Corrective Actions Completion Date for NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors. Project stage: Request ML0818503312008-06-27027 June 2008 Email Response from Scott Maglio to Mohan Thadani, Jack Donahew, Leon Whitney Generic Letter - 2004 - 02 Project stage: Other ML0818503552008-06-27027 June 2008 Attachment, Summary of Telephone Conference Regarding Request for Extension of Completion of Corrective Actions for NRC Generic Letter 2004-02 Project stage: Request ML0818406062008-07-0303 July 2008 Approval of Second Extension Request for Corrective Actions GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors. Project stage: Other ML0920101192008-12-18018 December 2008 Email Re Request for Additional Information, Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: RAI ML0920101222008-12-18018 December 2008 Request for Additional Information, Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: RAI ML0920306202009-07-31031 July 2009 Notice of Forthcoming Conference Call with Union Electric Company and Wolf Creek Nuclear Operating Company Re Draft Requests for Additional Information on Generic Letter 2004-02 for Callaway Plant and Wolf Creek Creek Generating Station Project stage: Draft RAI ML0922205722009-08-27027 August 2009 Request for Additional Information Related to Generic Letter 2004-02, Potential Impact of Debris Blockage of Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Project stage: RAI ML0924607142009-09-28028 September 2009 Summary of Meeting with Union Electric Co., and Wolf Creek, to Discuss Draft Requests for Additional Information on GL 2004-02 for Callaway, Unit 1, and Wolf Creek Project stage: Draft RAI ML0928003162009-10-16016 October 2009 Notice of Meeting with Union Electric Company and Wolf Creek Nuclear Operating Company to Discuss NRC Requests for Additional Information on Generic Letter 2004-02 Sent 7/31/0209 and 8/27/2009, Respectively Project stage: RAI ML0929300152009-10-20020 October 2009 Revised Notice of Meeting with Union Electric Company and Wolf Creek Nuclear Operating Company to Discuss NRC Requests for Additional Information on Generic Letter 2004-02 Project stage: RAI ML0933801482009-11-20020 November 2009 Licensee Handouts from November 20, 2009, Public Meeting with Union Electric Company and Wolf Creek Nuclear Operating Company to Discuss GL 2004-02 Response Rai'S Project stage: Request ULNRC-05668, Request for Extension of Response Date to NRC Request for Additional Information Re Responses to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at PWRs2009-11-24024 November 2009 Request for Extension of Response Date to NRC Request for Additional Information Re Responses to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at PWRs Project stage: Request ML0933805662009-12-17017 December 2009 Summary of Meeting with Union Electric Company and Wolf Creek Nuclear Operating Company to Discuss NRC Requests for Additional Information on Generic Letter 2004-02 Sent 7/31/2009 and 8/27/2009, Respectively Project stage: RAI ULNRC-05989, Path Forward for Resolution of GSI-1912013-05-16016 May 2013 Path Forward for Resolution of GSI-191 Project stage: Request ML13263A1682013-09-23023 September 2013 Request for Additional Information, Closure of Option 2 to Address In-Vessel Mitigative Measures for Potential In-Vessel Blockage; Related to Generic Safety Issue 191 Project stage: RAI ULNRC-06045, Response to Request for Additional Information Closure of Option 2 to Address In-Vessel Mitigative Measures for Potential In-Vessel Blockage2013-10-31031 October 2013 Response to Request for Additional Information Closure of Option 2 to Address In-Vessel Mitigative Measures for Potential In-Vessel Blockage Project stage: Response to RAI 2009-11-24
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Category:Letter
MONTHYEARULNRC-06853, Submittal of 2023 Fitness for Duty Performance Data Per Per 10 CFR 26.7172024-01-29029 January 2024 Submittal of 2023 Fitness for Duty Performance Data Per Per 10 CFR 26.717 IR 05000483/20230042024-01-19019 January 2024 Integrated Inspection Report 05000483/2023004 ML24008A0552024-01-19019 January 2024 Acceptance of Requested Licensing Action - Proposed Alternative to the Requirements of the ASME Code (EPID L-2023-LLR- 0061) ML23353A1712024-01-18018 January 2024 Issuance of Amendment No. 237 to Clarify Support System Requirements for the Residual Heat Removal System and Control Room Air Conditioning System Under Technical Specifications 3.4.8, 3.7.11, and 3.9.6 ML23317A0012024-01-12012 January 2024 Audit Summary Regarding LAR to Clarify Support System Requirements for the Residual Heat Removal and Control Room Air Conditioning System Under Technical Specifications 3.4.8, 3.7.11, and 3.9.6 ML23347A1212024-01-11011 January 2024 Issuance of Amendment No. 236 to Adopt TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Volume Values to Licensee Control EPID L-2023-LLA-0046) ML24011A1492024-01-11011 January 2024 Notification of Post-Approval Site Inspection for License Renewal and Request for Information (05000483/2024011) ULNRC-06847, Supplement to Relief Request from Requirements of ASME BPV Code, Section XI, Subsection Iwl Regarding Examination and Testing of the Unbonded Post-Tensioning System (Relief Request C3R-01)2023-12-21021 December 2023 Supplement to Relief Request from Requirements of ASME BPV Code, Section XI, Subsection Iwl Regarding Examination and Testing of the Unbonded Post-Tensioning System (Relief Request C3R-01) ULNRC-06849, License Renewal Resolution for Commitments 34 and 35 Perform Evaluation of Crack Initiation and Propagation in Steam Generator Divider Plate and Tube-To-Tubesheet Welds2023-12-20020 December 2023 License Renewal Resolution for Commitments 34 and 35 Perform Evaluation of Crack Initiation and Propagation in Steam Generator Divider Plate and Tube-To-Tubesheet Welds ML23346A0392023-12-14014 December 2023 Supplemental Information Needed for Acceptance of Requested Licensing Action Request for Relief from Requirements of ASME Code, Section Xl, Examination and Testing Unbonded Post-Tensioning System ULNRC-06844, Request for Exemption from Specific Requirements in 2023 Security Rule, Enhanced Weapons, Firearms Background Checks, and Security Event Notification2023-12-0707 December 2023 Request for Exemption from Specific Requirements in 2023 Security Rule, Enhanced Weapons, Firearms Background Checks, and Security Event Notification 05000483/LER-2023-001, Submittal of LER 2023-001-00 for Callaway, Unit 1, Inoperable Instrument Tunnel Sump Level Indication Resulted in Condition Prohibited by Technical Specifications2023-11-29029 November 2023 Submittal of LER 2023-001-00 for Callaway, Unit 1, Inoperable Instrument Tunnel Sump Level Indication Resulted in Condition Prohibited by Technical Specifications ULNRC-06827, Supplement to License Amendment Request Regarding Support System Requirements for Residual Heat Removal and Control Room Air Conditioning Systems Under Technical Specifications 3.4.8, 3.7.11, and 3.9.6 (LDCN 22-0029)2023-11-20020 November 2023 Supplement to License Amendment Request Regarding Support System Requirements for Residual Heat Removal and Control Room Air Conditioning Systems Under Technical Specifications 3.4.8, 3.7.11, and 3.9.6 (LDCN 22-0029) IR 05000483/20230102023-11-15015 November 2023 NRC License Renewal Phase 1 Inspection Report 05000483 2023010 IR 05000483/20233012023-11-0909 November 2023 NRC Examination Report 05000483-2023301 ML23311A2082023-11-0909 November 2023 Reassignment of U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch IV ML23297A2502023-11-0606 November 2023 Individual Notice of Consideration of Issuance of Amendment to Renewed Facility Operating License, Proposed No Significant Hazards Consideration Determination and Opportunity for a Hearing (EPID L-2022-LLA-0176) - Letter IR 05000483/20240122023-10-24024 October 2023 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection (05000483/2024012) ML23293A2652023-10-24024 October 2023 3rd Quarter 2023 Integrated Inspection Report ML23240A3692023-10-0505 October 2023 Issuance of Amendment No. 235 to Revise Technical Specifications to Use of Framatome Gaia Fuel (EPID L-2022-LLA-0150) (Non-Proprietary) ML23234A1522023-10-0505 October 2023 Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2022-LLE-0030) (Letter) ML23305A0942023-10-0202 October 2023 10-CW-2023-09 Post-Exam Submittal ML23270B9662023-09-27027 September 2023 10 CFR 50.55a(z)(I) Request for Relief from ASME OM Code Pump and Valve Testing Requirements for Fifth 120-Month Inservice Testing Interval ML23228A0252023-09-25025 September 2023 Issuance of Amendment No. 234 to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Test Frequencies ML23261C3852023-09-25025 September 2023 Safety Evaluation for Operating Quality Assurance Manual Revision 36A ML23166B0882023-09-20020 September 2023 Issuance of Amendment No. 233 for Adoption of Alternative Source Term and Revision of Technical Specifications ML23206A1992023-09-15015 September 2023 Regulatory Audit Summary Regarding License Amendment and Regulatory Exemptions Request for Fuel Transition to Framatome Gaia Fuel (Epids L-2022-LLA-0150 and L-2022-LLE-0030) IR 05000483/20234012023-09-13013 September 2023 NRC Security Baseline Inspection Report 05000483/2023401 ML23240A7572023-08-31031 August 2023 NRC Initial Operator Licensing Examination Approval 05000483/2023301 IR 05000483/20230052023-08-23023 August 2023 Updated Inspection Plan for Callaway Nuclear Power Plant, Unit 1 (Report 05000483/2023005) - Mid Cycle Letter 2023 ULNRC-06824, Response to Request for Additional Information Regarding Operating Quality Assurance Manual (Oqam) Revision 36A2023-08-17017 August 2023 Response to Request for Additional Information Regarding Operating Quality Assurance Manual (Oqam) Revision 36A ML23219A1392023-08-15015 August 2023 Request for Withholding Information from Public Disclosure ULNRC-06830, Transmittal of Updated Technical Specification Markup and Clean Pages for License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak2023-08-15015 August 2023 Transmittal of Updated Technical Specification Markup and Clean Pages for License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak ML23215A1972023-08-0303 August 2023 Supplement to License Amendment and Exemption Request Regarding Use of Framatome Gaia Fuel (LDCN 22-0002) ULNRC-06223, Minor Correction to License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Test Frequencies (LDCN 2020-0004)2023-07-25025 July 2023 Minor Correction to License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Test Frequencies (LDCN 2020-0004) ULNRC-06799, Submittal of Licensee Event Report 2022-003-01, Class 1E Electrical Air Conditioning System Thermal Expansion Valve Failure Resulted in Condition Prohibited by Technical Specifications2023-07-13013 July 2023 Submittal of Licensee Event Report 2022-003-01, Class 1E Electrical Air Conditioning System Thermal Expansion Valve Failure Resulted in Condition Prohibited by Technical Specifications IR 05000483/20230022023-07-10010 July 2023 Integrated Inspection Report 05000483/2023002 ML23174A1272023-06-23023 June 2023 Cw FFD Document Request List 2023 ML23171A9942023-06-22022 June 2023 Acceptance of Request for Approval of Operating Quality Assurance Manual Revision 36a ULNRC-06821, Post-Audit Follow-Up Information in Support of Callaway'S License Amendment Request and Proposed Exemption to Allow Use of Framatome Gaia Fuel (LDCN 22-0002) (EPID L-2022-LLA-0150 and EPID L-2022-LLE-00301)2023-06-21021 June 2023 Post-Audit Follow-Up Information in Support of Callaway'S License Amendment Request and Proposed Exemption to Allow Use of Framatome Gaia Fuel (LDCN 22-0002) (EPID L-2022-LLA-0150 and EPID L-2022-LLE-00301) IR 05000483/20230112023-06-15015 June 2023 Comprehensive Engineering Team Inspection (CETI) Inspection Report 05000483/2023011 ULNRC-06822, Additional Information Regarding Request for NRC Approval of Operating Quality Assurance Manual (Oqam) Revision 36a2023-06-14014 June 2023 Additional Information Regarding Request for NRC Approval of Operating Quality Assurance Manual (Oqam) Revision 36a ML23158A1462023-06-13013 June 2023 Notification of Post-Approval Site Inspection for License Renewal and Request for Information Inspection (05000483/2023010) ULNRC-06815, Request for NRC Approval of Operating Quality Assurance Manual, Revision 36a2023-06-0505 June 2023 Request for NRC Approval of Operating Quality Assurance Manual, Revision 36a ULNRC-06818, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate .2023-06-0505 June 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate . ML23093A0952023-05-10010 May 2023 Issuance of Amendment No. 232 Regarding Technical Specification Changes for Spent Fuel Storage ULNRC-06816, Withdrawal of Previously Submitted Enclosures Regarding License Amendment Request for Adoption of Alternative Source Term and Revision of Technical Specifications (LDCN 21-0015)2023-05-0909 May 2023 Withdrawal of Previously Submitted Enclosures Regarding License Amendment Request for Adoption of Alternative Source Term and Revision of Technical Specifications (LDCN 21-0015) ML23129A7942023-05-0909 May 2023 Post-Audit Supplement to License Amendment Request and Exemption to Allow Use of Framatome Gaia Fuel (LDCN 22-0002) (Iepid L-2022-LLA-0150 and EPID L-2022-LLE-0030) ML23122A3172023-05-0808 May 2023 Review of the Spring 2022 Steam Generator Tube Inservice Inspections ML23118A3492023-05-0808 May 2023 Request for Withholding Information from Public Disclosure 2024-01-29
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML23200A2982023-07-19019 July 2023 NRR E-mail Capture - Callaway Plant, Unit 1 - Final Request for Additional Information (RAI) - Request for Approval of Oqam, Revision 36a - EPID L-2023-LLQ-0000 ML23174A1272023-06-23023 June 2023 Cw FFD Document Request List 2023 ML23158A1462023-06-13013 June 2023 Notification of Post-Approval Site Inspection for License Renewal and Request for Information Inspection (05000483/2023010) ML23163A1572023-06-0606 June 2023 In-service Inspection Request for Information ML23096A0072023-04-0505 April 2023 NRR E-mail Capture - Final Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request to Revise TS 5.5.16 for Permanent Extension of Integrated Leak Rate Testing - EPID L-2022-LLA-0165 ML23080A1382023-03-21021 March 2023 Notification of Inspection (NRC Inspection Report 05000483/2023003) and Request for Information ML23073A0262023-03-13013 March 2023 NRR E-mail Capture - Final - Request for Additional Information - Callaway Plant, Unit 1 - SG Inspection Report Review - EPID L-2022-LRO-0143 ML23037A7092023-02-0606 February 2023 April 2023 Emergency Preparedness Exercise Inspection - Request for Information ML23026A0212023-01-24024 January 2023 NRR E-mail Capture - Final - Request for Additional Information - Callaway Plant, Unit 1 - LAR for Proposed Changes to TS for SFP - ML22287A0952022-10-14014 October 2022 NRR E-mail Capture - Final - Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request for Adoption of Alternate Source Term and Revision of Technical Specifications - EPID L-2021-LLA-0177 ML22269A4312022-09-26026 September 2022 November 2022 Emergency Preparedness Program Inspection - Request for Information ML22173A0562022-06-22022 June 2022 Information Request, Security IR 2022402 ML22167A0252022-06-15015 June 2022 NRR E-mail Capture - Final - Request for Additional Information - Callaway Plant, Unit 1 - LAR for Proposed Revision to Radiological Emergency Response Plan Regarding Response & Notification Goals - EPID L-2022-LLA-0024 ML22157A0572022-06-0606 June 2022 Notification of NRC Design Bases Assurance Inspection (Programs) (05000483/2022013) and Request for Information ML22154A0122022-06-0202 June 2022 NRR E-mail Capture - Final - Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request for Adoption of Alternate Source Term and Revision of Technical Specifications - EPID L-2021-LLA-0177 ML22151A0512022-05-27027 May 2022 NRR E-mail Capture - Final - Request for Additional Information - Columbia Generating Station - LAR to Change TS 3.4.11 - Reactor Coolant System Pressure and Temperature Limits - EPID L-2021-LLA-0191 ML22137A0292022-05-16016 May 2022 NRR E-mail Capture - Draft - Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request for Adoption of Alternate Source Term and Revision of Technical Specifications - EPID L-2021-LLA-0177 ML22096A0232022-04-0505 April 2022 NRR E-mail Capture - Callaway Plant - Final RAIs - License Amendment and Regulatory Exemptions for a Risk-Informed Approach to Address Generic Safety Issue - 191 and Respond to GL 2004-02 (EPIDs L-2021-LLA-0059 and L-2021-LLE-0021) ML21336A6392021-12-0202 December 2021 .05 Sec Doc Request ML21319A0062021-11-30030 November 2021 Supplemental Information Needed for Acceptance of Requested Licensing Actions License Amendment Request for Adoption of Alternate Source Term and Revision of Technical Specifications ML21258A0382021-09-14014 September 2021 NRR E-mail Capture - Final - Request for Additional Information - Callaway, Unit 1 - LAR to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors- EPID L-2020-LLA-023 ML21159A2352021-06-17017 June 2021 Notification of NRC Triennial Heat Exchanger/Heat Sink Performance Inspection (05000483/2021003) and Request for Information ML21130A5882021-05-11011 May 2021 Supplemental Information Needed for Acceptance of Requested Licensing Actions to Adopt a Risk-Informed Approach to Address GSI-191 and Respond to Generic Letter 2004 02 ML21088A3872021-03-30030 March 2021 Notification of Evaluations of Changes, Tests, and Experiments Inspection (Inspection Report 05000483/2021002) and Request for Information ML21007A1622021-01-0606 January 2021 NRR E-mail Capture - Final - Request for Additional Information - (COVID-19) Callaway Plant, Unit 1 - Additional Request for Exemption from Specific Requirements of 10 CFR Part 26, Fitness for Duty Programs - EPID L-2021-LLE-0242 ML20203M3682020-07-21021 July 2020 NRR E-mail Capture - Draft Request for Additional Information - Callaway Plant, Unit 1 - License Amendment Request - Revision to Technical Specification (TS) 5.3.1 and Deletion of TS 5.3.1.1 and 5.3.1.2 - EPID L-2020-LLA-0046 ML20162A1882020-06-10010 June 2020 Request for Supporting Information for the Callaway SPRA Audit Review - Draft Supplement ML20280A5442020-03-25025 March 2020 Cwy 2020 PIR Request for Information ML20064B6582020-02-27027 February 2020 Second, Third, and Fourth Request for Information for Callaway Dba Teams Inspection 2020011 ML19317E6332019-11-13013 November 2019 Request for Supporting Information for the Callaway SPRA Audit Review ML19107A5152019-04-11011 April 2019 Cwy 2019410 RFI Cyber Security Gap ML19078A2932019-03-18018 March 2019 Notification of an NRC Triennial Fire Protection Baseline Inspection (NRC Inspection Report 05000483/2019010) and Request for Information ML19023A2032019-01-22022 January 2019 Notification of NRC Triennial Heat Sink Performance Inspection (05000483/2019001) and Request for Information ML19009A3442019-01-0909 January 2019 NRR E-mail Capture - Formal Release of RAIs Ref: Callaway Plant Class 1E LAR, L-2018-LLA-0062 ML18355A4882018-12-20020 December 2018 NRR E-mail Capture - Formal Release of RAI Ref: Callaway Plant EAL Changes, L-2018-LLA-0239 ML18331A2052018-11-27027 November 2018 NRR E-mail Capture - Formal Release of RAIs Ref: Callaway Relief Request EPID L-2018-LLR-0051 ML18025B4672018-01-24024 January 2018 NRR E-mail Capture - Request for Extension of Due Date for RAI Response ML17304B1912017-10-31031 October 2017 NRR E-mail Capture - Requests for Additional Information Concerning Callaway License Amendment - Thermal Overload Protection ML17142A1352017-05-19019 May 2017 Notification of NRC Design Bases Assurance Inspection (05000483/2017007) and Initial Request for Information ML17115A0622017-04-25025 April 2017 NRR E-mail Capture - Requests for Additional Information -- Callaway Plant, Unit 1, Technical Specification 5.6.5, Core Operating Limits Report CAC MF8463 ML17038A2292017-02-0707 February 2017 NRR E-mail Capture - RAI Formal Release for Callaway SG Tube Inspection Report, MF8474 ML16111B3222016-04-20020 April 2016 Notification of Evaluations of Changes, Tests, and Experiments, and Permanent Plant Modifications Inspection (05000483/2016007) and Request for Information ML15316A1532015-11-12012 November 2015 Request for Additional Information Email, Relief Request 13R-11 (Pressurizer Welds) from Code Case N-460 Requirements, Third 10-Year Inservice Inspection Interval ML15096A0942015-04-0606 April 2015 Notification of Inspection (NRC Inspection Report 05000483/2015003) and Request for Information ML14353A1172014-12-22022 December 2014 Request for Additional Information, Round 3, License Amendment Request to Revise Final Safety Analysis Report Standard Plant Section 3.6 for High Density Polyethylene (Hdpe) Crack Exclusion ML14294A7752014-10-28028 October 2014 Request for Additional Information, Round 2, License Amendment Request to Revise Final Safety Analysis Report Standard Plant Section 3.6 for High Density Polyethylene (Hdpe) Crack Exclusion ML14203A0632014-07-25025 July 2014 Request for Additional Information, Relief Request I3R-17, Proposed Alternative to ASME Code, Section XI Requirements, Which Extends Rv ISI Frequency from 10 to 20 Years, Third 10-Year ISI Interval ML14178A8232014-07-0101 July 2014 Request for Additional Information, License Amendment Request to Revise Final Safety Analysis Report- Standard Plant Section 3.6 for High Density Polyethylene (Hdpe) Crack Exclusion ULNRC-06117, Callaway Plant, Unit 1, License Revewal Application, Request for Additional Information (RAI) Set 31 Responses2014-04-24024 April 2014 Callaway Plant, Unit 1, License Revewal Application, Request for Additional Information (RAI) Set 31 Responses ML14114A1102014-04-24024 April 2014 License Revewal Application, Request for Additional Information (RAI) Set 31 Responses 2023-07-19
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 27, 2009 Mr. Adam C. Heflin Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 65251 SUB..IECT: CALLAWAY PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO GENERIC LEDER 2004-02, "POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED-WATER REACTORS" (TAC NO. MC4671)
Dear Mr. Heflin:
By letter dated February 29,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080810491), Union Electric Company (the licensee) submitted a supplemental response to Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors," for Callaway Plant, Unit 1.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal.
The process involved detailed review by a team of approximately 10 subject matter experts, with a focus on the review areas described in the NRC's "Content Guide for Generic Letter 2004-02 Supplemental Responses" (ADAMS Accession No. ML073110389). Based on these reviews, the staff has determined that additional information is needed in order to conclude that there is a reasonable assurance that GL 2004-02 has been satisfactorily addressed for Callaway Plant, Unit 1. A draft request for additional information (RAI) was sent to the licensee bye-mail on December 18, 2008. This letter documents the RAI e-mail.
The NRC requests that the licensee respond to the enclosed RAls within 90 days of the date of this letter. However, the NRC would like to receive only one response letter for all RAls with exceptions stated below. If the licensee concludes that more than 90 days are required to respond to the RAls, the licensee should request additional time, including a basis for why the extension is needed.
The NRC staff considers in-vessel downstream effects to not be fully addressed at Callaway Plant, Unit 1, as well as at other pressurized-water reactors. The licensee's submittal refers to draft WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid." At this time, the NRC staff has not issued a final safety evaluation (SE) for WCAP-16793.
The licensee may demonstrate that in-vessel downstream effects issues are resolved for Callaway, by showing that the licensee's plant conditions are bounded by the final WCAP-16793 and the corresponding final NRC staff SE, and by addressing the conditions and limitations in
A. Heflin -2 the final SE. The licensee may also resolve RAI 17 by demonstrating, without reference to WCAP-16793 or the NRC staff SE, that in-vessel downstream effects have been addressed at Callaway. The specific issues raised in RAI 17 should be addressed regardless of the approach the licensee chooses to take.
The licensee should report how it has addressed the in-vessel downstream effects issue and the associated RAI referenced above within 90 days of issuance of the final NRC staff SE on WCAP-16793.
Bye-mail dated July 16, 2009, your staff provided a comparison between the enclosed RAls and those recently issued to Wolf Creek Generating Station (ADAMS Accession No.
ML092030628). The Wolf Creek RAls, though issued before the enclosed Callaway RAls, were developed after the enclosed RAls and reflect additional NRC staff review of certain issues pertinent to both plants. Based on a phone conversation with Mr. Thomas Elwood of your staff on July 16, 2009, we understand that you plan to address all the RAls issued to Wolf Creek because they apply to the two plants, which are very similar. We also understand that both licensees wish to interact with the NRC staff together on the issues raised by the RAls. Your written response should address all the enclosed RAls as well as those for Wolf Creek. Where a given item is substantially identical in the two sets, a cross-reference from one set of responses to the other is all that is required.
As part of the written response to the additional RAls, we request that you include a safety case.
This safety case should describe, in an overall or holistic manner, how the measures credited in the Callaway licensing basis demonstrate compliance with the applicable NRC regulations as discussed in GL 2004-02 and should describe your approach to responding to the RAls. As appropriate, the safety case may describe how the licensee reached compliance even in the presence of remaining uncertainties. The NRC staff views the safety case as informing, not replacing, responses to the RAls.
If you have any questions, please contact me at 301-415-1476 or via e-mail at mohan.thadani@nrc.gov.
Sincerely, Mohan C. Thadani, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosure:
As stated cc w/encl: Distribution via Listserv
CALLAWAY PLANT, UNIT 1 REQUEST FOR ADDITIONAL INFORMATION SUPPLEMENTAL RESPONSE DATED FEBRUARY 29,2008, TO GENERIC LETTER 2004-02, "POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED-WATER REACTORS"
- 1. Please provide information that verifies that the break selection process was completed considering the reduced zones of influence (ZOls) based on WCAP-16710-P, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) of Min-K and NUKON Insulation, for Wolf Creek and Callaway Nuclear Operating Plants," dated October 2007, or that the originally selected breaks remain bounding from a debris generation perspective after reducing the ZOls for Min-K and jacketed Nukon.
- 2. Please explain whether secondary breaks (main steam or feedwater) could require recirculation to supply containment spray. If one or more secondary breaks require recirculation for containment spray, provide information that shows whether the analysis for any loss-of-coolant accident (LOCA) bounds the secondary break(s). If secondary breaks are not bounded by LOCA analyses, please address the impact of such breaks on emergency core cooling system (ECCS) strainer performance, including the method used to determine the limiting main steam line break (MSLB) location.
- 3. Considering that the Callaway Plant, Unit 1 (Callaway) debris generation analysis diverged from the approved guidance in Nuclear Energy Institute (NEI) 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," please provide details on the testing conducted that justified the ZOI reductions for encapsulated Min-K and the jacketed Nukon. The information should include the jacket materials used in the testing, geometries and sizes of the targets and jet nozzle, and materials used for jackets installed in the plant. Provide information that compares the mechanical configuration and sizes of the test targets and jets, and the potential targets and two-phase jets in the plant. Evaluate how any differences in jeUtarget sizing and jet impingement angle affect the ability of potentially impacted insulation to resist damage from jet impingement. State whether the testing in WCAP-1671 O-P was bounding for the Callaway insulation systems. If not, prOVide information that compares the Callaway encapsulation and jacketing systems structure with the system that was used in the testing, showing that the testing conservatively or prototypically bounded potential damage to the insulation materials.
- 4. In the February 29, 2008, supplemental response (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML080810491), Union Electric Company (the licensee) showed that the Callaway debris generation/ZOI analysis contained three size categories of fibrous insulation debris: small fines, large pieces, and intact. However, as stated in the NRC staff's safety evaluation (SE) on NEI 04-07, Enclosure
-2 dated December 6, 2004 (ADAMS Accession No. ML043280641), in order to conduct adequate transport analysis and head loss testing, the small fines should be further broken down into fines (suspended fibers) and small pieces (less than 4 inches on a side). Using this categorization system (or justifying a different categorization), please provide additional information on the amounts of fibrous debris predicted to be generated from various breaks. Note that reduced lOis generally result in increased percentages of small and fine debris.
- 5. The licensee's February 29, 2008, supplemental response (page 12 of 81) stated that the Min-K at Callaway is located near the reactor vessel. This raises the question as to whether spherical resizing was done and whether it is appropriate for this location. The NRC staff's audit report for San Onofre Nuclear Generating Station discusses a potentially similar issue (Open Item 1 in Section 3.2, ADAMS Accession No.
ML071240024) regarding Microtherm insulation that was located on the reactor vessel, for which spherical resizing was considered inappropriate by the staff due to the constraints imposed by the biological shield wall and reactor vessel. Please state whether a spherical lOI was assumed in this region for which substantial physical obstructions could result in a significantly non-spherical destruction zone, and, if applicable, provide a technical basis for the use of a spherical lOI.
- 6. The WCAP-16710-P lOI reduction for jacketed l'Jukon insulation was also taken for Thermal Wrap at Callaway. Please provide information on the jacketing, banding and/or latching, and cloth cover for the Thermal Wrap insulation to provide confidence that it is comparable to the jacketing system for the Nukon insulation system that was tested.
- 7. The NRC staff has concerns that the size of the nozzle being used for the NUKON destruction testing at Wyle Laboratories may have resulted in non-conservatively exposing only a limited area of the target material to the peak jet pressure, particularly for the tests conducted at the smaller lOI radii. Since a LOCA jet could be much larger than 3 inches in diameter, the testing may not be representative of an actual LOCA at close ranges where the pressures of the smaller-diameter jet used for the testing would decay significantly more rapidly in the radial direction. This potential non-prototypicality from the debris generation testing affects not only the determination of lOI size, but also the determination of the size distribution of the debris formed within that ZOL Appendix" to the NRC's SE for NEI 04-07 indicates that essentially all low-density fiberglass within 7 pipe diameters (70) of a pipe rupture would become small fines.
However, based on the potentially non-conservative NUKON destruction testing performed at Wyle Laboratories discussed above, for Callaway only 60 percent small fines were assumed to be generated within 5 pipe diameters (50) of a LOCA jet, and 100 percent intact pieces were assumed to be generated between 50 and 70 of a LOCA jet. Please provide additional information to justify why the quantity of small fines debris assumed for Callaway is conservative or prototypical.
- 8. The NRC staff's SE for NEI 04-07 stated that a maximum of 15 percent holdup of debris should be assumed in inactive holdup regions during pool fill up. For the case of single train sump operation for Callaway, a two-sump plant, the sump that is not operating essentially becomes an inactive holdup region. From this point of view, the staff observed that Callaway appeared to credit a 15 percent inactive holdup volume in the
-3 containment pool, plus 14 percent holdup in the inactive recirculation sump for single train cases, for a total of 29 percent of debris held up in inactive volumes for these single-train cases (e.g., the Loop D cross-over break). The staff considers this credit a deviation from the approved guidance in the SE, which stated that the limit for inactive hold up should be 15 percent unless a computational fluid dynamics (CFD) analysis was performed that considered the time-dependent containment pool flows during pool fill-up.
Please provide additional basis for the assumed total inactive holdup fraction of 29 percent or revise this value to within the accepted SE range.
- 9. The licensee's February 29,2008, supplemental response discusses Stokes' Law, but does not specifically quantify the credit taken for application of this methodology. Please state the quantities of fine debris assumed to settle onto the containment floor by applying the Stokes' Law methodology. If credit is taken for such settling, technical justification is needed regarding the following points: (1) (lack of) experimental benchmarking of analytically derived turbulent kinetic energy (TKE) metrics; (2) uncertainties in the predictive capabilities of TKE models in CFD codes, particularly at the low TKE levels necessary to suspend individual fibers and 10-micron particulate; (3) the basis for analytical prediction of settling velocities in quiescent water due to the specification of shape factors and drag coefficients for irregularly shaped debris; and (4) the basis for theoretical correlation of the terminal settling velocity to turbulent kinetic energy that underlies the Alion Science & Technology methodology for fine debris settling. Please address these points to demonstrate that the credit taken for fine debris settling is technically justified.
- 10. Please identify the source of the erosion testing used to justify 10 percent erosion of fiberglass in the containment pool for Callaway and specify the velocity, turbulence, and chemical conditions for which the testing is applicable, and the velocity, turbulence, and chemical conditions present in the Callaway containment pool.
- 11. The licensee's February 29, 2008, submittal indicated that its analyses and/or testing were substantially incomplete in the head loss and vortexing area. The NRC staff will review the remaining information when the licensee submits it and, as a result of such review, the staff could request additional information in this subject area. Among items that should be addressed are:
- a. At the beginning of recirculation for a small-break LOCA, the strainer stacks are not submerged by about 6 inches. This condition should be evaluated for vortexing, air ingestion, and failure of the strainer to pass adequate flow.
- b. The air ingestion evaluation should include an analysis of the potential for de-aeration of the sump fluid as it passes through the debris bed and strainer. If de-aeration can occur resulting in entrained air reaching the pumps' suctions, a correction to the affected pumps' NPSH R should be calculated as described in Appendix A to NRC Regulatory Guide 1.82, Revision 3, "Water Sources for Long Term Recirculation Cooling Following a Loss-of-Coolant Accident," dated November 2003.
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- c. The Callaway strainer testing was witnessed by the NRC staff. The staff observed that significant agglomeration of debris occurred during testing. The staff also noted that the amount of fine debris predicted to reach the strainer was extremely low compared to other plant evaluations that used test methods the staff has found to be generally acceptable. Because the testing was designed to credit near-field settling, these issues could have significantly affected the results of the testing in a non-conservative manner. It was noted in the debris characteristics section of the February 29, 2008, supplemental response that the small debris contained about 30 percent fines. However, if the fines were not separated from the smalls prior to addition, it is likely that they would become entangled or agglomerated with the larger debris. This would reduce fine debris transport and the ability of the fibrous debris to create a thin bed. In fact, in PCI testing witnessed by the staff after the Callaway testing, high head losses occurred with the addition of only particulate and fine fibrous debris.
- d. The February 29, 2008, supplemental response states that no containment accident pressure is credited with regard to head loss, vortexing, air ingestion, or void fraction determination. Considering the small strainer submergence for a large-break LOCA (relative to the head loss across the strainer screen) and lack of submergence for a small-break LOCA, it is not clear to the staff what pressure prevents flashing across the debris bed and strainer.
- 12. The licensee's February 29, 2008, submittal indicated that its analyses and/or testing were substantially incomplete in the net positive suction head (NPSH) area. The NRC staff will review the remaining information when the licensee submits it and, as a result of such review, the staff could request additional information in this subject area. Among items that should be addressed are:
- a. the completed NPSH analyses with the quantitative results for the NPSH margins,
- b. both cold-leg and hot-leg recirculation scenario NPSH margins for all pumps taking suction from the recirculation sump,
- c. the NPSH margin values for the small- and large-break LOCAs,
- d. the pump vendor's criteria for determining the NPSH required (NPSH R) data for the pumps taking suction from the recirculation sump,
- e. the specific methodology used for computing friction head loss in suction piping, and
- f. a summary of the single-failure analysis for the NPSH calculation (single-failure scenarios considered should be identified, and NPSH margin results should be presented).
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- 13. For degraded qualified coatings, the Keeler and Long Report report, "Design Basis Accident Testing of Coating Samples from Unit 1 Containment, TXU Comanche Peak SES," dated April 13, 2006 (ADAMS Accession No. ML070230390), and industry testing are cited as justification of epoxy chip sizes. The NRC's "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses," dated November 21, 2007 (ADAMS Accession No. ML073110389), has accepted use of the Keeler and Long Report, which results in smaller chip sizes than those described in table 3h-2. Please provide justification for using chips larger than those determined in the Keeler and Long report.
In addition, please summarize methods and results of the industry testing reference used to determine the size distribution of degraded qualified coatings.
- 14. Please describe how the quantity of curled chips is determined. In addition, please justify the simplification of the size distribution of the curled chips to a 1.5 inch chip size.
- 15. Please clarify the weight distribution of coating debris surrogates used in head loss testing. Please explain whether it is consistent with table 3h-2 in the submittal. If so, please explain the basis for the distribution in table 3h-2.
- 16. Please provide the quantities of each type of coatings surrogate material used in head loss testing.
- 17. The licensee's February 29, 2008, submittal indicated that its analyses and/or testing were substantially incomplete in the downstream effects, components and systems, fuel and vessel area. The NRC staff will review the remaining information when the licensee submits it and, as a result of such review, the staff could request additional information in this subject area. When submitted, please provide the information requested under item (n) in the NRC's Revised Content Guide. The NRC staff considers in-vessel downstream effects to be not fully addressed at Callaway as well as at other PWRs.
The licensee's submittal refers to draft WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid." The NRC staff has not issued a final SE for WCAP-16793-NP. The licensee may demonstrate that in-vessel downstream effects issues are resolved for Callaway by showing that the licensee's plant conditions are bounded by the final WCAP-16793-NP and the corresponding final NRC staff SE, and by addressing the conditions and limitations in the final SE. The licensee may also resolve this item by demonstrating without reference to WCAP-16793 or the staff SE that in-vessel downstream effects have been addressed at Callaway.
- 18. Please provide the basis that demonstrates that chemicals leaching from insulations and other containment materials that are sprayed but not submerged (Le., located above the flood plane following a LOCA) are not significant to chemical precipitate formation.
- 19. Please identify and justify the assumptions related to phosphate inhibition of aluminum corrosion. For example:
- a. What is the threshold concentration of phosphate assumed to passivate aluminum?
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- b. What time is assumed to reach that phosphate concentration in the pool?
- c. If phosphate inhibition is credited for aluminum in the spray zone, what amount of containment spray time is assumed (after the pool reaches an inhibition threshold of phosphate) before the aluminum is passivated?
ML092030628). The Wolf Creek RAls, though issued before the enclosed Callaway RAls, were developed after the enclosed RAls and reflect additional NRC staff review of certain issues pertinent to both plants. Based on a phone conversation with Mr. Thomas Elwood of your staff on July 16, 2009, we understand that you plan to address all the RAls issued to Wolf Creek because they apply to the two plants, which are very similar. We also understand that both licensees wish to interact with the NRC staff together on the issues raised by the RAls. Your written response should address all the enclosed RAls as well as those for Wolf Creek. Where a given item is substantially identical in the two sets, a cross-reference from one set of responses to the other is all that is required.
As part of the written response to the additional RAls, we request that you include a safety case.
This safety case should describe, in an overall or holistic manner, how the measures credited in the Callaway licensing basis demonstrate compliance with the applicable NRC regulations as discussed in GL 2004-02 and should describe your approach to responding to the RAls. As appropriate, the safety case may describe how the licensee reached compliance even in the presence of remaining uncertainties. The NRC staff views the safety case as informing, not replacing, responses to the RAls.
If you have any questions, please contact me at 301-415-1476 or via e-mail at mohan.thadani@nrc.gov.
Sincerely, IRA!
Mohan C. Thadani, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosure:
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