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Category:Report
MONTHYEARML23010A0882023-01-15015 January 2023 Summary of Regulatory Audit in Support of License Amendment Request to Revise Control Rod Technical Specifications PLA-8026, Biennial 10 Crf 50.59 and 72.48 Summary Report and Changes to Regulatory Commitment (PLA-8026)2022-10-19019 October 2022 Biennial 10 Crf 50.59 and 72.48 Summary Report and Changes to Regulatory Commitment (PLA-8026) PLA-7998, Submittal of Unit 1 Cycle 23 Core Operating Limits Report and Reload Safety Analysis Report2022-03-29029 March 2022 Submittal of Unit 1 Cycle 23 Core Operating Limits Report and Reload Safety Analysis Report PLA-7984, Supplement to License Amendment Requesting Adoption of TSTF-505, Revision 22022-03-0808 March 2022 Supplement to License Amendment Requesting Adoption of TSTF-505, Revision 2 PLA-7959, Owner'S Activity Report PLA-79592021-07-15015 July 2021 Owner'S Activity Report PLA-7959 PLA-7910, Submittal of Unit 2 Cycle 21 Fuel Rod Design Report to Support License Amendment Requesting Application of Advanced Framatome Methodologies PLA-79102020-12-10010 December 2020 Submittal of Unit 2 Cycle 21 Fuel Rod Design Report to Support License Amendment Requesting Application of Advanced Framatome Methodologies PLA-7910 PLA-7757, Final Integrated Plan to Comply with June 06, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (NRC-Order EA-13-109), Revision 12018-11-27027 November 2018 Final Integrated Plan to Comply with June 06, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (NRC-Order EA-13-109), Revision 1 ML18024A1142018-01-24024 January 2018 a Reaction to Cracking of Austenitic Stainless Steel Piping in Boiling Water Reactors (Includes Susquehanna Ses Design Modifications) ML17093A6682017-03-27027 March 2017 Technical Requirements Manual Unit 2, Revision 93 with Revision 13 to Table of Contents PLA-7559, Flooding Mitigating Strategies Assessment (MSA) Report2016-12-19019 December 2016 Flooding Mitigating Strategies Assessment (MSA) Report ML16231A5092016-08-25025 August 2016 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Phase 2 of Order EA-13-109 (Severe Accident Capable Hardened Vents)(Cac Nos. MF4364 and MF4365) PLA-7491, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insight from the Dai-Ichi Accident2016-06-30030 June 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insight from the Dai-Ichi Accident ML16054A2432016-02-0909 February 2016 NPDES Permit No. PA 0047325 Renewal Application ML15356A2472016-01-20020 January 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulation Part 50, Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force PLA-7414, Redacted - Susquehanna, Units 1 and 2 - Submittal of 10 CFR 71.95 Report Regarding Energy Solutions 8-120B Cask Certificate of Compliance 91682015-11-19019 November 2015 Redacted - Susquehanna, Units 1 and 2 - Submittal of 10 CFR 71.95 Report Regarding Energy Solutions 8-120B Cask Certificate of Compliance 9168 PLA-7374, Submittal of 10 CFR71.95 Report Regarding Energy Solutions 8-120B Cask Certificate of Compliance 91682015-08-20020 August 2015 Submittal of 10 CFR71.95 Report Regarding Energy Solutions 8-120B Cask Certificate of Compliance 9168 PLA-7311, Negative Blind Specimen Report, PLA-73112015-04-0909 April 2015 Negative Blind Specimen Report, PLA-7311 ML15090A3002015-04-0101 April 2015 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-13-109 (Severe Accident Capable Hardened Vents) (TAC Nos. MF4364 & MF4365) PLA-7287, Flood Hazards Reevaluation Report PLA-72872015-03-0303 March 2015 Flood Hazards Reevaluation Report PLA-7287 ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 PLA-7175, (SSES)- 10 CFR 50.46 - Annual Report2014-06-20020 June 2014 (SSES)- 10 CFR 50.46 - Annual Report ML14156A2342014-06-16016 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (MF0288 and MF0289) ML14056A4492014-05-0606 May 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident (Tacs MF0183, MF0184) ML14113A5552014-04-23023 April 2014 Draft Staff Assessment of Seismic Walkdown Report Near-term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident; PPL Susquehanna, LLC; Susquehanna Steam Electric Station, Unit 1. Docket No. 50-387 ML14085A3982014-03-26026 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(F) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident PLA-7145 ML14085A4262014-03-26026 March 2014 Apendix B IPEEE Adequacy Review ML14010A3712014-01-15015 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Susquehanna Steam Electric Station, Units 1 and 2, TAC Nos.: MF0888 and MF0889 PLA-6938, Response to Request for Information Pursuant to 10 CFR 50.54 (F) Regarding Results of the SSES Flooding Walkdown (References 1 and 2) PLA-69382012-11-21021 November 2012 Response to Request for Information Pursuant to 10 CFR 50.54 (F) Regarding Results of the SSES Flooding Walkdown (References 1 and 2) PLA-6938 PLA-6809, Proposed Amendment No. 309 to License NPF-14 & Proposed Amendment No. 280 to License NPF-22: Change to Technical Specification Surveillance Requirement (SR) 3.8.1.19 to Increase Diesel Generator E Minimum Steady.2012-09-18018 September 2012 Proposed Amendment No. 309 to License NPF-14 & Proposed Amendment No. 280 to License NPF-22: Change to Technical Specification Surveillance Requirement (SR) 3.8.1.19 to Increase Diesel Generator E Minimum Steady. ML13168A4282012-03-29029 March 2012 NRC 2012 Susquehanna Steam Electric Station Unit 2 - TN1179 ML13168A4212012-03-29029 March 2012 NRC 2012 Susquehanna Steam Electric Station Unit 1 - TN1178 ML12068A1952012-02-23023 February 2012 Technical Requirements Manual Unit 2 ML11348A1102011-12-0101 December 2011 TRM1, Technical Requirements Manual, Unit 1 ML13046A1372011-11-30030 November 2011 Enclosure 8 ML1127000692011-09-26026 September 2011 Enclosure 2, Mfn 10-245 R4, Description of the Evaluation and Surveillance Recommendations for BWR/2-5 Plants ML1108712502011-03-24024 March 2011 BWR Vessel and Internals Inspection Summaries for Spring 2010 Outages ML1102504712011-01-14014 January 2011 Technical Requirements Manual PLA-6657, 10 CFR 50.59 Summary Report and Changes to Regulatory Commitments2010-10-0707 October 2010 10 CFR 50.59 Summary Report and Changes to Regulatory Commitments ML1020401442010-05-12012 May 2010 Replacement Steam Dryer Report, Unit 1, Start-Up, 107% Power Test Plateau 05/11/10 ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants ML1004801852009-12-31031 December 2009 BWRVIP-117NP-A: BWR Vessel and Internals Project - RAMA Fluence Methodology Plant Application-Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5, Final Report PLA-6568, Submittal of 10CFR26.719(c)(1) Report for Drug and Alcohol Testing Errors PLA-65682009-10-0505 October 2009 Submittal of 10CFR26.719(c)(1) Report for Drug and Alcohol Testing Errors PLA-6568 ML0921703332009-07-31031 July 2009 Enclosure 2 to PLA-6542 - SSES Replacement Steam Dryer and Flow Induced Vibration Report Unit 2 Start-Up 107.0% Power Test Plateau. (Non-Proprietary) PLA-6523, Fourteen Refueling Outage Owner'S Activity Report PLA-65232009-07-31031 July 2009 Fourteen Refueling Outage Owner'S Activity Report PLA-6523 ML0918004602009-06-23023 June 2009 Submittal of Work Environment Improvement Plan, PLA-6528 PLA-6510, Engineering Report 0000-0101-0766-NP-R0, Main Steam Line Limit Curve Adjustment During Power Ascension.2009-04-30030 April 2009 Engineering Report 0000-0101-0766-NP-R0, Main Steam Line Limit Curve Adjustment During Power Ascension. ML0907704472009-02-28028 February 2009 Enclosure 5 to PLA-6484, Susquehanna Replacement Steam Dryer Updated Stress Analysis at Extended Power Uprate Conditions, Non-Proprietary Version Engineering Report 0000-0095-2113-NP-R0 ML0907704482009-02-27027 February 2009 Enclosures 6, 7, 8 and 9 to PLA-6484, Revised Susquehanna Replacement Steam Dryer Limit Curves - Main Steam Line Mounted Instrumentation, Non-Proprietary Version Engineering Report 0000-0096-5766-NP-R1 PLA-6438, Proposed Amendment No. 274 to Unit 2: MCPR Safety Limits, Including Enclosure 2 to PLA-6438, PPL Evaluation of the Proposed Changes Unit 2 Minimum Critical Power Ratio Safety Limits, Non-Proprietary Version2008-10-30030 October 2008 Proposed Amendment No. 274 to Unit 2: MCPR Safety Limits, Including Enclosure 2 to PLA-6438, PPL Evaluation of the Proposed Changes Unit 2 Minimum Critical Power Ratio Safety Limits, Non-Proprietary Version ML0828804202008-10-14014 October 2008 River Drainage Yoy Smallmouth Bass Disease Investigations 2005 and 2007 2023-01-15
[Table view] Category:Technical
MONTHYEARPLA-8026, Biennial 10 Crf 50.59 and 72.48 Summary Report and Changes to Regulatory Commitment (PLA-8026)2022-10-19019 October 2022 Biennial 10 Crf 50.59 and 72.48 Summary Report and Changes to Regulatory Commitment (PLA-8026) PLA-7998, Submittal of Unit 1 Cycle 23 Core Operating Limits Report and Reload Safety Analysis Report2022-03-29029 March 2022 Submittal of Unit 1 Cycle 23 Core Operating Limits Report and Reload Safety Analysis Report PLA-7910, Submittal of Unit 2 Cycle 21 Fuel Rod Design Report to Support License Amendment Requesting Application of Advanced Framatome Methodologies PLA-79102020-12-10010 December 2020 Submittal of Unit 2 Cycle 21 Fuel Rod Design Report to Support License Amendment Requesting Application of Advanced Framatome Methodologies PLA-7910 PLA-7757, Final Integrated Plan to Comply with June 06, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (NRC-Order EA-13-109), Revision 12018-11-27027 November 2018 Final Integrated Plan to Comply with June 06, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (NRC-Order EA-13-109), Revision 1 ML18024A1142018-01-24024 January 2018 a Reaction to Cracking of Austenitic Stainless Steel Piping in Boiling Water Reactors (Includes Susquehanna Ses Design Modifications) ML17093A6682017-03-27027 March 2017 Technical Requirements Manual Unit 2, Revision 93 with Revision 13 to Table of Contents PLA-7491, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insight from the Dai-Ichi Accident2016-06-30030 June 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insight from the Dai-Ichi Accident ML16054A2432016-02-0909 February 2016 NPDES Permit No. PA 0047325 Renewal Application PLA-7287, Flood Hazards Reevaluation Report PLA-72872015-03-0303 March 2015 Flood Hazards Reevaluation Report PLA-7287 ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 ML14010A3712014-01-15015 January 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Susquehanna Steam Electric Station, Units 1 and 2, TAC Nos.: MF0888 and MF0889 PLA-6938, Response to Request for Information Pursuant to 10 CFR 50.54 (F) Regarding Results of the SSES Flooding Walkdown (References 1 and 2) PLA-69382012-11-21021 November 2012 Response to Request for Information Pursuant to 10 CFR 50.54 (F) Regarding Results of the SSES Flooding Walkdown (References 1 and 2) PLA-6938 ML13168A4282012-03-29029 March 2012 NRC 2012 Susquehanna Steam Electric Station Unit 2 - TN1179 ML13168A4212012-03-29029 March 2012 NRC 2012 Susquehanna Steam Electric Station Unit 1 - TN1178 ML11348A1102011-12-0101 December 2011 TRM1, Technical Requirements Manual, Unit 1 ML1102504712011-01-14014 January 2011 Technical Requirements Manual ML1004801852009-12-31031 December 2009 BWRVIP-117NP-A: BWR Vessel and Internals Project - RAMA Fluence Methodology Plant Application-Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5, Final Report ML0921703332009-07-31031 July 2009 Enclosure 2 to PLA-6542 - SSES Replacement Steam Dryer and Flow Induced Vibration Report Unit 2 Start-Up 107.0% Power Test Plateau. (Non-Proprietary) PLA-6510, Engineering Report 0000-0101-0766-NP-R0, Main Steam Line Limit Curve Adjustment During Power Ascension.2009-04-30030 April 2009 Engineering Report 0000-0101-0766-NP-R0, Main Steam Line Limit Curve Adjustment During Power Ascension. ML0907704472009-02-28028 February 2009 Enclosure 5 to PLA-6484, Susquehanna Replacement Steam Dryer Updated Stress Analysis at Extended Power Uprate Conditions, Non-Proprietary Version Engineering Report 0000-0095-2113-NP-R0 ML0907704482009-02-27027 February 2009 Enclosures 6, 7, 8 and 9 to PLA-6484, Revised Susquehanna Replacement Steam Dryer Limit Curves - Main Steam Line Mounted Instrumentation, Non-Proprietary Version Engineering Report 0000-0096-5766-NP-R1 PLA-6438, Proposed Amendment No. 274 to Unit 2: MCPR Safety Limits, Including Enclosure 2 to PLA-6438, PPL Evaluation of the Proposed Changes Unit 2 Minimum Critical Power Ratio Safety Limits, Non-Proprietary Version2008-10-30030 October 2008 Proposed Amendment No. 274 to Unit 2: MCPR Safety Limits, Including Enclosure 2 to PLA-6438, PPL Evaluation of the Proposed Changes Unit 2 Minimum Critical Power Ratio Safety Limits, Non-Proprietary Version ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 ML0828300742008-07-31031 July 2008 GE-NE 0000-0085-2413 NP-R0, Susquehanna Unit 1, Replacement Steam Dryer Vibration Instrumentation Program NRC Summary Test Report ML0810105462008-04-0101 April 2008 Attachment 2 to PLA-6346, Non-Proprietary Version of Unit 1 Main Steam Line Limit Curves ML0806602552008-02-22022 February 2008 GE-NE-000-0080-2994-NP-R2, Revision 2, Susquehanna Replacement Steam Dryer Instrumentation Acceptance Criteria - Dryer Mounted Instrumentation. ML0803800892008-01-31031 January 2008 Engineering Report, GE-NE-0000-0079-2250-NP-R0, Susquehanna Replacement Steam Dryer Stress Analysis at Extended Power Uprate Conditions ML0728501222007-10-10010 October 2007 BWRVIP-145NP, BWR Vessel and Internals Project, Evaluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology ML0720103642007-06-30030 June 2007 GE-Hitachi Nuclear Energy Americas LLC Report GE-NE-0000-0061-0595-NP-R1, Susquehanna Replacement Steam Dryer Fatigue Analysis. ML0700403832006-12-31031 December 2006 GE-NE-OOOO-0061-0595-NP-RO, Engineering Report: Susquehanna Replacement Steam Dryer Fatigue Analysis. PLA-6136, Proposed License Amendment, Arts/Mella Implementation Response to Request for Additional Information PLA-61362006-11-29029 November 2006 Proposed License Amendment, Arts/Mella Implementation Response to Request for Additional Information PLA-6136 PLA-5995, Proposed Amendment No. 269 to Unit 1 & Amendment No. 236 to Unit 2: DC Electrical Power Systems Technical Specifications Rewrite - Response to Request for Additional Information (RAI)2005-12-15015 December 2005 Proposed Amendment No. 269 to Unit 1 & Amendment No. 236 to Unit 2: DC Electrical Power Systems Technical Specifications Rewrite - Response to Request for Additional Information (RAI) PLA-5983, Proposed License Amendment Numbers 272 for Unit 1 Operating License No. NPF-014 and 241 for Unit 2 Operating License No. NPF-022, Power Range Neutron Monitor Digital Upgrade Supplemental Information2005-12-0101 December 2005 Proposed License Amendment Numbers 272 for Unit 1 Operating License No. NPF-014 and 241 for Unit 2 Operating License No. NPF-022, Power Range Neutron Monitor Digital Upgrade Supplemental Information PLA-6130, Rev. 0 to 0000-0039-3825 Susq A-M T506-RBM-Calc-2005, Instrument Limits Calculation, PPL Susquehanna, LLC, Susquehanna Steam Electric Station Units 1 & 2, Rod Block Monitor (Numac ARTS-MELLA).2005-10-31031 October 2005 Rev. 0 to 0000-0039-3825 Susq A-M T506-RBM-Calc-2005, Instrument Limits Calculation, PPL Susquehanna, LLC, Susquehanna Steam Electric Station Units 1 & 2, Rod Block Monitor (Numac ARTS-MELLA). ML0601203532005-10-13013 October 2005 Proposed Amendment No. 281 and Proposed Amendment No. 251: Application for License Amendment and Related Technical Specification Changes to Implement Full-Scope Alternative Source Term in Accordance with 10 CFR 50.67 ML0507504332005-03-11011 March 2005 PROJ0704 - BWRVIP-1 17NP: BWR Vessel and Internals Project, RAMA Fluence Methodology Plant Application - Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 ML0710400422003-08-0101 August 2003 Environmental Studies in the Vicinity of the Susquehanna Steam Electric Station 2002 Water Quality and Fishes. PLA-5616, Annual Personnel Exposure Report PLA-56162003-04-28028 April 2003 Annual Personnel Exposure Report PLA-5616 ML18024A0561978-09-15015 September 1978 Letter Supplementing PP&L Letters Dated April 17, 1978 and July 14, 1978 and Attaching a Final Report of a Deficiency in Seismic Qualification of Medium Voltage, Metal Clad, Safeguards Switchgear ML18026A2181978-04-14014 April 1978 SSES Design Assessment Report, Revision 1 ML18025A6661978-02-28028 February 1978 Structural Integrity Test Report Containment Structure Unit 1 ML18025A6671977-12-30030 December 1977 Final Report on Shear Studs ML18025A1901977-06-22022 June 1977 Letter Enclosing an Interim Report of Reportable Deficiency Regarding Unsatisfied Field Welding and Inspection of Shear Studs ML18026A1191977-03-22022 March 1977 Results of Non-Nuclear Hot Test with the Relief System in the Philippsburg Nuclear Power Plant ML18025A0911976-05-31031 May 1976 Ecological Studies of the North Branch Susquehanna River in the Vicinity of the Susquehanna Steam Electric Station - Progress Report for the Period January-December 1974 ML18023A8811975-08-29029 August 1975 Letter Transmitting Report on Deficiency in Box Column Supports for the Susquehanna Nuclear Plant. ML18025A2011975-08-29029 August 1975 Definitive Report of Deficiencies in Fabricated Structural Steel Members (Box Column Supports) ML18023A8841975-07-25025 July 1975 Letter Regarding an Interim Report of Reportable Deficiency ML18023A8871975-04-0404 April 1975 Letter Attaching an Interim Report on a Potential Deficiency Which Was Discussed on March 6, 1975 ML18026A1181972-12-31031 December 1972 Formation and Oscillations of a Spherical Gas Bubble Under Water 2022-03-29
[Table view] |
Text
SUSQUEHANNA STEAM ELECTRIC STATION DESIGN ASSESSMENT REPORT (DAR)
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Revision 1
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PR EPACE This Report contains data, descriptions and anaylsis relative to the adequacy of the Susquehanna Steam Electric Station design to accommodate loads resultinq from safety relief valve (SRV) discharqe and/or loss-of-coolant accident (LOCA) conditions.
Rev. 2, 5/8p
DAR TABLE OF CONTENTS Chapter 1 GENERAL. INFORMATION 1.1 Purpose of Report 1.2 History of Problem
- 1. 3 SSES Containment Program 1.4 Plant Description 1.5 Figures 1 6 Tables Chapter 2
SUMMARY
2.1 Load Definition Summary 2.2 Design Assessment Summary I
Chapter 3 SRV DISCHARGE AND LOCA TRANSIENT DESCRIPTION
- 3. 1 Description of Safety Relief Valve (SRV) Discharge 3.2 Description of Loss-of-Coolant Accident (LOCA)
Chapter 4 LOAD DEFINITION
- 4. 1 Loads from Safety Relief Valve Discharge 4.2 Loads from Loss-of-Coolant Accident 4.3 Annulus Pressurization 4.4 Figures 4.5 Tables Chapter 5 LOAD COMBINATIONS FOR STRUCTURES~ PIPING~ A'ND EQUIPMENT
- 5. 1 Concrete Containment and Reactor Building Load Combinations 5.2 Structural Steel Load Combinations 5.3 Liner Plate Load Combinations 5.4 Downcomer Load Combinations 5.5 Piping, Quencher, and Quencher Support Load Combinations 5.6 NSSS Load Combinations 5.7 Equipment Load Combinations 5.8 Electrical Raceway System Load Combinations
- 5. 9 HVAC Duct System Load Combinations 5.10 Figures 5.11 Tables Chapter 6 DESIGN CAPABILITY ASSESSMENT
- 6. 1 Concrete Containment and Reactor Building Capability Assessment Criteria 6.2 Structural Steel Capability Assessment Criteria 6.3 Liner Plate Capability Assessment Criteria'ev.
2, 5/80
TABLE OF CONTENTS fContinued)-
6.4 Downcomer Capability Assessment Criteria 6.5 Piping, Quencher, and Quencher Support Capability Assessment Criteria 6.6 NSSS Capability Assessment Criteria 6 7 Equipment Capability Assessment Criteria Chapter 7 DESIGN ASSESSMENT
- 7. 1 Assessment Methodology 7.2 Design Capability Margins 7.3 Fiqures Chapter 8 SSE~S- U~ECHER VERIFICATION TEST
- 8. 1 Introduction 8 2 Test Facility and Instrumentation 8.3 Test Parameters and Matrix 8.4 Test Results 8.5 Data Analysis 'and Verification of Load Specification 8.6 Figures 8.7 Tables Chapter 9 GKM IIM STEAM BLOWDOWN TESTS 9.1 Introduction 9.2 Test Facility and Matrix 9.3 Test Parameters and Matrix 9.4 Test Results 9.5 Data Analysis 9.6 Verification of the Design Specification 9.7 Figures Chapter 10 RESPONSES TO NRC~UESTIONS 10 1 Identification of Questions Unique to SSES 10 2 Questions Unique to SSES and Responses Thereto
- 10. 3 Questions Pertaining to the NRC s Reviev of the
~
DAR and Response Thereto
- 10. 4 Figures Chapte'r 11 REFERENCES Appendix A CONTAINMENT DESIGN ASSESSMENT A. 1 Containment Structural Design Assessment A.2 Submerged Structures Design Assessment Appendix B CONTAINMENT RESPONSE SPECTRA DUE TO SRV AND LOCA LOADS 1 Containment Mode Shapes B.2 Containment Response Spectra Appendix C REACTOR BUILDING RESPONSE SPECTRA DUE TO SRV AND XOCA LOADS Rev. 2, 5/80
Appendix D pRgGRgg VgRIFICQTIQN D.1 Poolsvell Model Verification D.2 Velpot Proqram Verification D.3 Fiqures D. 4 Tables A ppendix F. R EACTOQ BUILDING STRUCTURAL DESIGN A SS L'S SM EP T Appendix F BOP- AND NSSS GRIPING DESIGN ASSESSMFNT Appendix G VASSS QESIGN ASSESSMENT QDELETFOQ Appendix H EQUIPMENT DESIGN ASSESSMENT /DELETED}
Appendix I SUPPRESSION POOL TFMPERATJJRE RFSPON SE TO SRV DISCHARGE-Appendix J VFRIFICATION OF SRV SUBMERGED STRUCTURE DRAG LOAD (PRO'PRIETARY)
Appendix K DBYMELL FLOOR VACCUM BREAKER JVB} CYCLING DURING C HUGGING Appendix L .SUppLEMEQTAL DESIGN. ASSFSSMENT L. 1 Assessment Methodoloqy L. 2 Assessment Results J..3 Fiqures L.4 Tables Rev. 8, 2/83
I CHAPTER "
'1 GENERAL INFORMATTION TABLE OP CONTENTS 1.1 PURPOSE OF REPORT 1.2 HISTORY OF PROBLEM 1-3 SSES CONTAI NMENT PROGRAM' 4 PLANT DESCRIPTION 1.4 1 Primary'ontainment'.
- 1. 4 1. Penetrations 1.4. 1.2 1
Internal Structures",'~ ',,'"
l I' 5" F ZGUR ES I, w 1.6 TABLES II I I II>>
Rev. 2,.5/80 1-1
CHAPTER 1 FIGURES Number Title 1-1 Cross Section of Containment Suppression Chamber, Partial Plan 1-3 Suppression Chamber, Section Vie@
1-0 Quencher Distribution Rev. 2, 5/80 1-2
~O'X TA CHAPTER 1 TABLES
~umber Title SSES Licensing Basis SSES Containment Dimensions 1-3 SSES Containment Design Parameters Comparison of the SSES Program for SRV and LOCA, loadings with the NUREG 0487 Acceptance Criteria, Lead Plant Program and Generic Long Term Program Rev. 2, 5/80 1-3
1 0 G NRRAL INFORMATION 1.1 PU'RPOSE AND ORGANIZATION OF REPORT The purpose of this report is to present evidence that the Susquehanna Steam Electric Station (SSES) design margins are adequate should the plant be subjected to the recently defined thermohydrodynamic loads which result from safety relief valve (SRV) operations and/or discharges during a loss-of-coolant accident (LOCA) in a GE hoilinq water reactor (BWR) .
Rev; 2, 5/80 1-4
1 2 HISTORy'F PROBLEM In April 1972 at the German AEG-Kraftwerk Union Murgassen Nuclear Plant, a boiling water, reactor (BMR) safety relief valve (SRV) was opened during startup testing and failed to close. The reactor remained at full pressure,'and the valve discharged reactor steam into the containment suppression chamber until the suppression pool water heated from gust above ambient to almost 170~C {in approximately 30 minutes). Pulsating condensation developed and large impulsive forces with substantial underpressure amplitudes acted upon the containment, eventually causinq leakaqe from the bottom liner plate. Therefore, expressed that the structural integrity of other BMR pressure concern'as containment systems could be sensitive to SRV induced dynamic loads.
The Nuclear Regulatory Commission (NRC) issued Bulletin 74-14 to all BWR owners on November 14, 1974 to alert them to the potential problems of condensation instability (Murgassen to operation The NRC requested verification that BMR effect)'ue SRV suppression pools had been designed to withstand loads similar to those which were being experienced. In January 1975 the General Electric Nuclear Energy Program Division (GE-NEPD) identified the follcwing dynamic loading conditions which had not been fully considered in the design criteria of Mark II BMR containments:
- a. Hain steam SRV discharge thermo-hydrodynamic phenomena.
- b. Desiqn basis accident (DBA): loss-of-coolant accident (LOCA) hydrodynamic phenomena Following the GE announcement, the containment construction sequence for the SSES was altered to enable the Pennsylvania Power and Light Company (PPSL) and its architect-engineer, Bechtel Power Corporation, to ascertain the effect of these phenomena on the existing SSES design. A task force was formed in Narch 1975 with representatives from Bechtel-San Francisco, GE-NEPD, PPGL, and Philadelphia Electric Company to evaluate existinq design criteria with respect to the newly defined SRV and DBA-LOCA loadinqs. In Nay 1975 Bechtel completed a preliminary study incorporatinq the effects of the new phenomena in the design criteria for the SSES suppression chamber structures and safety related equipment As a result of this investigation, it was decided that the following civil-structural modifications were to be incorporated immediately in the containment design to aid in load transfer and add additional conservatism to the existing design:
- a. The number of reinforcing bars in the suppression chamber vertical walls was increased.
- b. The number of embedments in the suppression chamber walls for downcomer/piping restraints was increased to accommodate future requirements.
1-5
c Anchor .bolts were placed on the underside of the diaphragm slab to accommodate additional supports SRV discharge piping for horizontal runs should they foz'he be needed.
- d. Additional anchor bolts were placed within the drywell wall to allow installation of additional snubbers and pipe restraints, if required
- e. The diaphragm slab shear reinforcement was changed from a 45~ to a 90~ orientation {with respect to the horizontal plane) to accommodate the most conservative pool swell uplift loadings yet predicted.
It became evident that a complex, technical issue existed for all Nark II plants, and PPGl. sought to create a unif ied utility group to address the matter. A Mark II BMR containment owners group was formed in June 1975 to define precisely the suppression pool dynamic loads and explore ways to assess their impact,. As the direct result of action taken by the Mark II containment owners organization, a generic Dynamic Forcinq Function Information Report, NEDE-21061P Rev. 1, which was also known as the DFFIR, was issued jointly by GE-NEPD and Sargent and Lundy for the Mark II owners in September 1975 Based on the analytical techniques included in the DFFIR, a preliminary SSES unique containment design assessment was submitted by PPGI. to the Nuclear Regulatory Commission {NRC) on Narch 15, 1976 As the body of the useful supportive data increased, Revision 2 of the DFFZH was issued jointly by GE-NEPD and Sargent and t.undy for the Mark II containment owners group on September 1, 1976, as NEDO/NEDE 21061, Rev. 2. It was at this time renamed. the DFFR.
The licensing documentation considered for the SSES is summarized in Table 1-1.
1-6
1 ' SSES- CONTAINMENT PROGRAM.
PPGL is a member of the Hark II owners-group that vas formed in June, 1975 to define and investigate the dynamic loads due to SRV discharge and LOCA. The Mark II ovners group containment program concentrated initially on the tasks required fo'r the licensing of the lead plants (Zimmer, LaSalle, and Shoreham) . This phase of vork, called the short'erm proqram, is complete and a longer term program is underway. The final goal of the Hark II program is to evolve a complete DFFR which vill support the plant-unique DARs submitted by e'ach plant for its license to operate.
After gaining some understanding of the containment .loads through the initial Mark IX work, PPGL decided to find a qualified, consultant to supplement in-house technical resources and assist in the determination of a realistic course of action for Susquehanna In November, 1976, Stanford Research Institute, now called SRI International (SRI), was selected, and an information exchange between SRZ.and PPGL ensued to determine what caused the greatest loads on the containment structure. After conducting a complete review of known data from the Mark II program and other knowledgeable persons and organizations, PPGL and SRI decided that the loads from main steam sa fety relief valve (SRV) d'ischarge vere the key loads to be controlled. A study of possible methods of controlling the load and a review of vhat activities vere occurrinq in Europe led PPGL and SRI to the conclusion that an SRV discharge mitigating device (quencher) should be employed to reduce .this loadinq on the Susquehanna containment. Although the Hark II owners group had quencher-related tasks in their program, these tasks vere not sufficiently timely to satisfy SSES-construction schedule needs.
Prom reviewing the vork done in Europe by such firms as ASEATOM MARVIKEN, and Kraftverk Union, PPGL discovered that all known quencher'designs were based on data from Kraftverk Union (KWU).
Thus, in March, 1977, SRI, Bechtel (the SSES Architect/Engineer) and PPGL visited KWU for discussion and tour of quencher-related facilities. Xn late July, 1977, PPGL employed the services of KWU to design a SSES-unique quencher device.
Kraftwork Union provided PPGL a package of significant design and test reports pertaining to the quencher development to demonstrate design adequacy and quality of their device (refer to Table l-l). These documents were submitted to the NRC in January, 1978. The quencher load specification vas submitted to the NRC in April, 1978. To verify KWU~s design approach, a full-scale SSES unique unit cell test, as described in Chapter 8, was performed by KWU for PPGL. The documentation of this test series and verification of the design specification vas submitted in March, 1979. Subsequently the quencher design by KWU f or use on SSES has been adopted as the SRV discharge used by six of the seven other Hark II owners and the SSES program has become the generic Mark II program.
Rev. 2, 5/80 1-7
The definition of LOCA loads (Section 4.2) is in basic accordance with the Mark II program. In addition though,'PSL has decided to conduct a series of transient steam blowdown tests in a modified GKM II test tank in Nannheim, Germany (refer to Chapter
- 9) . These tests will provide data to resolve NRC concerns on the differences in vent configuration between the original GE 4T facility and a prototypical Mark II containment and, to verify the condensation oscillation load specification used on the SSES design.
Table SSES l-l provides a summary of the documentation supporting the licensing.
In addition, Table 1-4 provides a comparison of the SSES program for SRV. and LOCA loading with the NUREG 0487 acceptance criteria, Lead Plant Program and Generic Long Term Program. I'n accordance with the directions of the NRC staff at the October 19, 1978 meeting with the Mark II Owners Group these positions assume that the use of the SRSS method of load combination will be accepted for use on the Mark II containments.
Rev. 2, 5/80 0-8
- 1. 4 - Pl.M1T DESCRIPTION The SSES, Units 1 and 2, is being built in Salem Township, Luzerne County about 5 miles northeast of the Borough of Berwick. Two generating units of approximately 1,100 megawatts each are scheduled for operation: Unit 1 for November 1, 1980, and Unit 2 for May 1, 1982. General Electric is supplying the nuclear steam supply systems; Bechtel Power Corporation is the architect-engineer and constructor.
The reactor building contains the major nuclear systems and equipment The nuclear reactors for Units 1 and 2 are boiling water, direct cycle types with a rated heat output of 11.2 x 10~
Btu/hr. Each reactor supplies 13.4 x 10~ lb/hr of steam to the tandem compound, double flow turbines.
- 1. 4.1 Primarv Containment The containment is a reinforced concrete structure consisting of a cylindrical suppression chamber beneath a truncated conical drywell. Figure 1-1 shows the geometry of the containment and internal structures. The conical portion of the primary containment (drywell) encloses the reactor vessel, reactor coolant recirculation loops, and associated components of the reactor coolant system. The drywell i.s separated from the wetwell, ie, the pressure suppression chamber and pool, by the drywell floor, also named the diaphragm slab. Major systems and components in the containment include the vent pipe system (downcomers) connecting the drywell and wetwell, isolation valves, vacuum relief system, containment cooling systems, and other service equipment. The cone and cylinder form a structurally integrated reinforced concrete vessel, lined with steel plate and closed at the top of the drywell with a steel domed head. The carbon steel liner plate is anchored to the concrete by structural steel members embedded in the concrete and welded to the plate.
The entire containment is structurally separated from the surrounding reactor building except at the base foundation slab (a reinforced concrete mat, top lined with a carbon steel liner plate) where a cold joint between the two adjoining foundation slabs is provided. The containment structure dimensions and parameters are listed in Tables 1-2 and 1-3. A detailed plant description can be found in the SSES FSAR, Section 3.8.
- 1. 4. 1 1 'Penetrations Services and communication between the inside and outside of the
.containment are made possible by penetrations through the containment wall. The basic types of penetrations are the drywell head, access hatches (eguipment hatches, personnel lock, suppression chamber access hatches, CRD removal hatch),
electrical penetrations, and pipe penetrations. The piping Rev. 2, 5/80 1-9
penetrations consist basically of a pipe with plate flange welded to it. The plate flange is embedded in the concrete wall and provides an anchorage for the penetration to resist normal operatinq and accident pipe reaction loads.
1.4.1. 2 Internal Structures The internal structures consist of reinforced concrete and structural steel and have the major functions of supporting and shielding the reactor vessel, supporting the piping and equipment, and forming the pressure suppression boundary. These structures include the drywell floor (diaphragm slab), the reactor pedestal (a concentric cylindrical reinforced concrete shell resting on the containment base foundation slab and supporting the reactor vessel), the reactor shield wall, the suppression chamber columns (hollow steel pipe columns supporting the diaphragm slab), the drywell platforms, the seismic trusses, the quencher supports, and the reactor steam supply system supports. See Figures 1-1 thr'ough 1-4 and Tables 1-2 and 1-3.
Rev. 2, 5/80 1-10
IL OF PRIMARY CONTAINMENT I
SYM. ABT. (L MORYWELLHEAD 48' OV" 0 D. CONC.
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30'41" R REACTOR BUILDING I.D. CONC. BASE MAT COLD JOINT SUSQUEHANNA STEAM ELECTRIC STATION tJNITS 1 AND 2 DESIGN ASSESSMENT REPORT CROSS SECTION OF CONTAINMENT FIGURE 1-1
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NOTE: BRACING IS NOT SHOWN Rev. 2, 5/80 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT SUPPRESSION CHAMBER PARTIAL PLAN FIGURE 1-2