Letter Sequence Request |
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MONTHYEARNOC-AE-15003315, Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies for Cycle 202015-12-0303 December 2015
Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies for Cycle 20
Project stage: Request
NOC-AE-15003318, Response to Request for Additional Information and Supplement to Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies for Cycle 202015-12-0909 December 2015
Response to Request for Additional Information and Supplement to Emergency License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies for Cycle 20
Project stage: Supplement
NOC-AE-16003351, License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies2016-04-0707 April 2016
License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: Request
ML16127A4522016-05-12012 May 2016
Supplemental Information Needed for Acceptance of Requested Licensing Action, Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: Acceptance Review
NOC-AE-16000338, Supplement to License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies2016-05-25025 May 2016
Supplement to License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: Supplement
ML16158A0622016-06-0606 June 2016
Acceptance of Requested Licensing Action Following Supplement, License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: Acceptance Review
ML16188A3682016-07-0606 July 2016
Audit Presentation - Westinghouse Methodology for South Texas Unit 1 LAR: Operation with 56 Control Rods
Project stage: Request
ML16214A2912016-08-26026 August 2016
Summary of June 28-30, 2016, Regulatory Audit at Westinghouse in Rockville, MD, License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: Other
ML16246A0952016-09-15015 September 2016
Correction to 8/26/16 Request for Additional Information Enclosure, License Amendment Request to Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: RAI
ML16319A0102016-12-21021 December 2016
Issuance of Amendment No. 211, Revise Technical Specification 5.3.2 to Allow Operation with 56 Full-Length Control Rod Assemblies
Project stage: Approval
2016-05-12
[Table View] |
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Category:Slides and Viewgraphs
MONTHYEARML22243A1312022-08-31031 August 2022
STP Nuclear Operating Company August 31, 2022 Presentation - Pre-Submittal Meeting for Updated Main Steam Line Break and Locked Rotor Dose Consequence Analysis to Address Extended Cooldown Timelines
ML22214A0972022-08-0202 August 2022
Presentation by South Texas Project
ML22153A2942022-06-17017 June 2022
richardsd-hv-th28
ML21203A2702021-07-26026 July 2021
Presentation - 7-26-21 Meeting with STPNOC Proposed South Texas Project Moderator Temperature Coefficient TS Change
ML20248H5322020-09-0404 September 2020
STP Nuclear Operating Company September 14, 2020, Presentation - Pre-Submittal Meeting for Proposed South Texas Project Moderator Temperature Coefficient Surveillance Requirement Technical Specification Change
ML20125A3362020-05-0404 May 2020
STP Presubmittal Meeting Exigent Accumulator LAR 2020-05-05 Presentation L-2020-LRM-0039
ML20034E8192020-02-0606 February 2020
Pre-Application Public Meeting for Proposed License Amendment Request to Revise the STPEGS Emergency Plan
ML19095A6562019-04-0404 April 2019
NRR E-mail Capture - (External_Sender) Handout for 4/10/19 Meeting to Discuss a Proposed Request for South Texas Project, Units 1 and 2, Regarding Repair of Piping Using Carbon Fiber Repair Methods
ML17004A0292017-01-0404 January 2017
Slides for January 4, 2017, Public Meeting on South Texas Project's Resposne to Aluminum Bronze Aging Management Request for Additional Information
ML16356A0612016-12-21021 December 2016
12/21/2016 South Texas Project Aluminum Bronze Aging Management Request for Additional Information
ML16350A3272016-11-17017 November 2016
Transcript of Advisory Committee on Reactor Safeguards Plant License Renewal Subcommittee Meeting - November 17, 2016
ML16264A3632016-09-20020 September 2016
Pre-Application Public Meeting for Proposed License Amendment Request to Revise the STPEGS Emergency Plan Staff Augmentation Times
ML16172A0192016-07-25025 July 2016
STP Aluminum Bronze Selective Leaching Public Meeting - June 21, 2016
ML16188A3682016-07-0606 July 2016
Audit Presentation - Westinghouse Methodology for South Texas Unit 1 LAR: Operation with 56 Control Rods
ML16062A3042016-03-0202 March 2016
TS 5.3.2 LAR Pre-Submittal Meeting Presentation - March 2, 2016
ML16033A0042016-01-19019 January 2016
Applicant's Slides for STP Meeting on Aluminum Bronze Aging Management
ML15334A3972015-11-30030 November 2015
TS 5 3 2 Emergency LAR Pre-Submittal Meeting Presentation
ML15274A5992015-10-0101 October 2015
10-01-15 Public Phone Call
ML15034A1512015-02-0505 February 2015
Integrated Leak Rate Test (ILRT) from 10 Years to 15 Years. NRC Public Meeting Slides - ILRT License Amendment
ML15034A1142015-02-0404 February 2015
STP Risk- Informed Approach to GSI - 191 Roverd 2015 Meeting Blue 1 29 15
ML14192A9842014-08-0404 August 2014
Summary of Pre-Licensing Meeting with STP Nuclear Operating Company to Discuss Future License Amendment Request to Revise TS 6.9.1.6.b.9 to the Leading Edge Flow Meter Technology for Feedwater Flow Measurement
ML14149A0892014-05-21021 May 2014
NRR E-mail Capture - STP-GSI-191 Presentation
ML14120A0112014-04-29029 April 2014
4/29/2014 - South Texas Project Regulatory Approach for Risk-Informed Pilot Submittal Presentation Relating to 10 CFR 50.46c
ML13352A1242013-12-16016 December 2013
NRC Staff Slides South Texas Project Risk-Informed Generic Safety Issue - 191, December 16, 2013
ML13352A1422013-12-16016 December 2013
GSI-191 Licensing Submittal Comparison of Changes Between Rev 1 and Rev 2
ML13316B9052013-11-13013 November 2013
Albrz Testing Update for NRR Final
ML13150A2162013-05-23023 May 2013
Licensee Slides for 5/23/13 Public Meeting Regarding GSI-191
ML13140A2562013-05-20020 May 2013
Licensee Slides for 5/23/13 Meeting - STP Pilot Submittal for Risk Informed Approach to Resolving GSI-191
ML13023A3342013-02-25025 February 2013
1/15/2013 Summary of Public Meetings Conducted to Discuss Draft Supplemental Environmental Impact Statement Related to Review of South Texas Project, Units 1 & 2, License Renewal Application
ML13051A8552013-02-20020 February 2013
Licensee Slides for 3/5/13 Pre-Application Meeting for Proposed Licensing Action to Revise the Fire Protection Program at STP
ML13029A4972013-01-15015 January 2013
Slides - Preliminary Site-Specific Results of the Environmental Review for South Texas Project License Renewal
ML12297A3312012-11-27027 November 2012
Summary of Prelicensing Meeting with STP Nuclear Operating Company to Discuss Proposed Amendment for Approval of Revised Fire Protection Program Related to Alternate Shutdown Capability for South Texas, Units 1 and 2
ML12264A3202012-10-11011 October 2012
Meeting Handout for 10/11/12 Pre-licensing Meeting License Amendment Request to Revise the Fire Protection Program
ML1209004042012-03-30030 March 2012
3-27-2012 - STP Public Meeting Handout from Nuclear Innovation North America, LLC on Fukushima Lessons Learned for South Texas Project Units 3 and 4
ML1204400652012-02-0909 February 2012
Licensee Slides for 2/9/12 Meeting Regarding GSI-191
ML1133505632011-12-0101 December 2011
Licensee Presentation from 12/1/11 Meeting Via Conference Call to Discuss Risk-Informed GSI-191,Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance
ML1130506322011-10-26026 October 2011
Licensee Handouts from November 1, 2011 Meeting with STP Nuclear Operating Company on GSI-191
ML1127701762011-10-0303 October 2011
Licensee Handouts from 10/3/11 Meeting STP LOCA Frequency
ML1127702032011-10-0303 October 2011
Licensee Slides from 10/3/11 Meeting LOCA Frequencies, Final Results
ML1123508832011-08-22022 August 2011
LOCA Initiating Event Frequencies and Uncertainties Status Report
ML1123508732011-08-22022 August 2011
Models and Methods Used for Casa Grande
ML1120707292011-07-26026 July 2011
Meeting Notice with South Texas Project, Units 1 and 2 - Licensee Slides Computational Fluid Dynamics Validation Plan
ML1120707072011-07-26026 July 2011
Meeting Notice with South Texas Project, Units 1 and 2 - Licensee Slide Models and Methods Used for Casa Grande
ML1118903802011-07-0707 July 2011
Licensee Slides, LOCA Initiating Event Frequencies and Uncertainties(Draft)
ML1118904202011-07-0606 July 2011
Licensee Slide from 7/7/11 Meeting with STP Nuclear
ML11157A0102011-06-0202 June 2011
STP Nuclear Operating Company, Licensee Handouts, 6/2/2011 Meeting Risk-Informed GSI-191 Closure Plan for GSI-191 Resolution
ML1114501952011-05-24024 May 2011
Summary of Meeting with South Texas Project Electric Generating Station 2010 Performance
ML1105503952011-02-22022 February 2011
Licensee Slides from 2/22/11 Public Meeting with STP Nuclear Operating Company
ML1103206432011-01-27027 January 2011
1/27/11 Presentation on Risk-Informed GSI-191 Project Overview
ML1021803392010-08-0505 August 2010
Ltr to E. Halpin Re Participation in Commission Meeting on the Resolution of Generic Safety Issue (GSI) - 191, Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Performance on 09/29/10
2022-08-31
[Table view] |
Text
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Westinghouse Methodology Impacts from the Removal of the RCCA from Core Location D-6 Brian Guthrie, Transient Analysis Danielle Schmitt, Nuclear Design 1
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Overview of Reload Method (WCAP-9272-P-A)
- The STP engineering design change process is used to implement the removal of RCCA D-6
- Impacts of the change on UFSAR Chapter 6 and Chapter 15 analyses have been evaluated through a detailed review of the AORs and the reload design process (WCAP-9272-P-A).
- As detailed in Table 7 of NOC-AE-16003351, the existing key safety parameters continue to apply (and are confirmed on a cycle specific basis) and no other aspects of the AORs are impacted.
WCAP-9272 methodology is still applicable when control rod pattern is changed 2
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Overview of Reload Method (WCAP-9272-P-A) 3
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Overview of Reload Method (WCAP-9272-P-A)
- The AORs are established using conservative inputs that are expected to bound future cycles.
- The key safety parameters that could be impacted by a reload become the Reload Safety Analysis Checklist (RSAC) which is issued every cycle by the Safety Analysis groups to Nuclear and Thermal Hydraulic Design for confirmation
- Any RSAC limit that is not met for a given cycle is provided to the impacted Safety Analysis group for evaluation
- Evaluation or re-analysis of the AOR is performed to address the RSAC exception and the results are reported in the cycle specific Reload Evaluation.
WCAP-9272 methodology is still applicable when control rod pattern is changed 4
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Safety Analysis Accidents
- Example - HZP Steamline Break
- RETRAN-02 uses a point kinetics neutronics model and does not directly model a control rod pattern
- Limiting case is generated using assumed physics characteristics.
- Nuclear Design implicitly confirms acceptability of the physics characteristics by comparing ANC calculated parameters to RETRAN-02 calculated parameters.
- An unacceptable mismatch requires evaluation or reanalysis of the event using adjusted physics parameters.
- This approach is consistent with the analysis methodology for the HZP steamline break core response described in WCAP-9226-P-A, Revision 1; the reload methodology for this event is described in WCAP-9272-P-A; and the qualification of RETRAN-02 for use in analyzing HZP steamline break core response is described in WCAP-14882-P-A.
No UFSAR safety analysis calculations explicitly model control rod patterns.
5
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Cycle-Specific Core Design Analysis
- Each cycle the reload core is modeled by Nuclear Design in detail (fuel pattern, control rod pattern, burnable absorbers, etc.)
- This detailed model is used to calculate cycle-specific values for comparison to the RSAC key safety parameters
- Key safety parameters assumed in the AOR include values that are impacted by RCCAs such as shutdown margin, total rod worth, and trip reactivity
- Key safety parameters assumed in the AOR do not include the number of RCCAs, RCCA configuration, or a symmetric RCCA pattern Nuclear Design calculations explicitly model control rod pattern.
6
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Cycle-Specific Core Design Analysis
- Nuclear Design performs the cycle-specific analysis using:
- ANC [WCAP-10965-P-A]
- APOLLO [WCAP-13524-P-A]
- PARAGON/NEXUS [WCAP-16045-P-A, Addendum 1-A]
- These codes were rigorously benchmarked and qualified for a variety of reactor types, fuel types, and burnable poisons
- These codes are capable of modeling an asymmetric core:
- Highest worth rod stuck out of the core (N-1) is a conservatism assumption for many RSAC calculations
- Asymmetric core temperature distribution with an N-1 condition is modeled for the Steamline Break accident
- N-2, N-3, etc are calculated for use in plant Operations Nuclear Design codes are capable of modeling the removal of RCCA D-6.
7
Westinghouse Non-Proprietary Class 3 © 2016 Westinghouse Electric Company LLC. All Rights Reserved.
Conclusions
- Plant changes are handled by STP engineering design change process
- As a part of the engineering design change process, Westinghouse and the STP Nuclear Fuel and Analysis group performed an impact review of the respective AORs including the continued applicability of WCAP-9272-P-A
- Nuclear Design codes are capable of modeling the removal of RCCA D-6 8
