Damping Values for Seismic Design of Nuclear Power Plants
Regulatory Guide 1.61# QUERYae8f90c8857d73a44087c1e50c035440
Regulatory Guide 1.61# QUERY2a6dc70a88de519a702cfaf9c4289687
Regulatory Guide 1.61# QUERY2f424dc36dc53148ebe16fe2b866f7dd

Regulatory Guide 1.61# QUERY63d5ea34e6aa3a2e0bea4e8bf23559a9
Regulatory Guide 1.61# QUERY6e8e6098e14870e6e2c4dfebdb64e5a9
Regulatory Guide 1.61# ERR1dd423a8697e32d6813ecd1a7aae482c
NRC 2017-0017, License Amendment Request 278, Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances
SBK-L-16108, Status of Required Actions for EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design Basis External Events
ONS-2015-072, Updated Final Safety Analysis Report, Revision 24, Chapter 9 - Auxiliary Systems
JAFP-15-0007, Seismic Walkdown Report Update - Entergys Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Re the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accide
ONS-2014-086, Updated Final Safety Analysis Report, Revision 23, Chapter 9, Auxiliary Systems
LR-N13-0172, Enclosure 1 - Salem Nuclear Generating Station Unit 1 Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic
ULNRC-06033, Proposed Revision to Final Safety Analysis Report FSAR-SP Section 3.7(B) & 3.7(N), Seismic Design, (License Amendment Request LDCN 12-0041)
SBK-L-13038, Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)
RC-13-0028, Overall Integrated Plan as Required by March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Event( Order EA-12-049)
L-12-065, Submittal of Shield Building Root Cause Evaluation
LR-N11-0122, Response to Request for Additional Information, License Amendment Request - Operation with Final Feedwater Temperature Reduction and Feedwater Heaters Out-of-Service
DCL-17-038, Diablo Canyon Power Plant, Units 1 & 2, Revised Updated Final Safety Analysis Report, Rev. 23, Chapter 5, Reactor Coolant System
LR-N17-0088, Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 1, Introduction and General Description of Plant
NL-17-0534, Joseph M. Farley Nuclear Plant, Updated Final Safety Analysis Report, Revision 27, Chapter 3, Part 1
LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 3A, Pse&G Positions on Us NRC Regulatory Guides
LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 3.7, Seismic Design
NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Figures 3.7.B.2-27 Through 3D-17
NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant
L-2016-198, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 5, Structures
NRC-16-0034, Fermi, Unit 2, Revision 20 to Updated Final Safety Analysis Report. Appendix a
RC-16-0083, Virgil C. Summer, Unit 1 - Updated Final Safety Analysis Report, Chapter 3, Design of Structures, Components, Equipment and Systems
NL-15-183, Watts Bar Nuclear Plant, Unit 2, Amendment 114 to Final Safety Analysis Report, Section 3.0 - Design of Structures, Components, Equipment, and Systems, Pages 3.8.5-1 - 3.12-6
NL-15-183, Watts Bar Nuclear Plant, Unit 2, Amendment 114 to Final Safety Analysis Report, Section 3.0 - Design of Structures, Components, Equipment, and Systems, Pages 3.6-1 - 3.7-223
RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 1 - Introduction and General Description of Plant
NG-17-0111, Duane Arnold Energy Center, Revision 24 to Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant
RS-13-009, Clinton, Unit 1, Updated Safety Analysis Report, Revision 15, Chapter 3 - Design of Structures, Components, Equipment and Systems
NL-12-2270, SNCV061-RPT-01, Ver 1.0, Vogtle Unit 1 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic. Part 1 of 20
NL-12-2271, Vogtle, Unit 2, SNCV061-RPT-02, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 2 of 14
NL-12-2271, Vogtle, Unit 2, SNCV061-RPT-02, Ver. 1.0, Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, Part 1 of 14
JAFP-12-0134, FitzPatrick - Seismic Walkdown Report, Response to NRC Request for Information Per 10 CFR 50.54 (F) Re the Seismic Aspect of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
JAFP-12-0134, Engineering Report, JAF-RPT-12-00015, Rev. 0, Jaf Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3 Seismic, Book 1 of 3
L-2010-212, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report - Unit 4 Cycle 24 Update, Chapter 5, Structures
DCL-10-102, Response to NRC Request for Additional Information Regarding License Amendment Request 10-03, Damping Values for the Seismic Design and Analysis of the Reactor Vessel Integrated Head Assembly
DCL-10-066, License Amendment Request 10-03, Damping Values for the Seismic Design and Analysis of the Reactor Vessel Integrated Head Assembly
DCL-10-065, Supplement to Response to NRC Request for Additional Information Regarding License Amendment Request 09-06, Critical Damping Value for Structural Dynamic Qualification of the Control Rod Drive Mechanism Pressure Housings
DCL-10-045, Response to NRC Request for Additional Information Re License Amendment Request 09-06, Critical Damping Value for Structural Dynamic Qualification of the Control Rod Drive Mechanism Pressure Housings
DCL-10-039, 10 CFR 50.59 Report of Changes, Tests, and Experiments for the Period January 1, 2008, Through December 31, 2009
DCL-09-086, License Amendment Request 09-06, Critical Damping Value for Structural Dynamic Qualification of the Control Rod Drive Mechanism Pressure Housings
NL-09-138, Updated Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.
WM 09-0001, Response to NRC Request for Additional Information Regarding Main Steam and Feedwater Isolation System (Msfis) Controls Modification
RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Appendix a Through Appendix N
RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Chapter 3.0 - Design of Structures, Components, Equipment, and Systems
ULNRC-05481, Response to Request for Additional Information Regarding Response for NRC Generic Letter 2004-002, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.
NL-08-025, Supplemental Response to GL-04-002, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors
HNP-08-015, Supplemental Response to NRC GL-04-002, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-water Reactors
CP-200800263, Supplement to License Amendment Request (LAR) 07-004 Revision to the Operating License and Technical Specification 1.0, Use and Application to Revise Rated Thermal Power from 3458 Mwt to 3612 Mwt
RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 3.0, Design of Structures, Components, Equipment and Systems
NLS2007012, Non-Proprietary Version of Response to Request for Additional Information Regarding License Amendment Request for Onsite Spent Fuel Storage Expansion
RAIO-1018-62158, LLC Supplemental Response to NRC Request for Additional Information No. 200 (Erai No. 9021) on the NuScale Design Certification Application
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 Information
ET 08-0053, Corporation Response to Request for Additional Information Response to Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors
ET 08-0003, Operating Corporation Response to Request for Additional Information Response to Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors
PLA-1097, Forwards Addl Info in Response to NRC Concerns Re Chugging Load Definition.Info Will Be Incorporated Into Design Assessment Rept Scheduled for Submittal on 820608
1CAN040302, License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration TS 3. 7.14, the Loading Restrictions in the SFP in TS 3,7,15m and to Modify the Fuel Storage Design Features in TS 4.3
2CAN010304, Arkansas, Unit 2, License Amendment Request to Change Spent Fuel Pool Loading Restrictions
IA-86-669, Partial Response to FOIA Request for Info Re Generic Safety Issues & Gimcs.Forwards App C Document.App B & C Documents Available in PDR
JPN-99-033, Provides Response to Questions Contained in 990712,facsimile from NRC Re Ja FitzPatrick USI A-46 Program.Questions Were Also Discussed Between Members of Util & NRC Staff During Telcon
A05975, Requests NRC Confirmation of Util Interpretation of Reg Guide 1.84 Applicability to plant-specific Use of ASME Code Case N-4111, Alternate Damping Values for Seismic Analysis of Class 1,2 & 3 Piping Sections,Section Iii,Div I
B12480, Const Design Info Submittal:New Switchgear Bldg Const. W/38 Oversize Encl
2CAN038604, Requests NRC Approval to Utilize ASME Code Case N-411, Alternative Damping Values for Seismic Analysis of Piping, for Response Spectrum Analysis of Selected Piping.Planned Use of Code Case Listed.Fee Paid
B12042, Requests Authorization by 860508 for Use of ASME Code Case N-411, Alternate Damping Values for Seismic Analysis.... More Realistic Damping Values Will Reduce Need for Numerous Snubbers & Improve Plant Reliability
B12844, Rev 0 to Spent Fuel Pool Capacity Expansion Project Description & Sar
B15017, Responds to Providing NRC W/Calculations That Documented Technical Justification of Approach for Developing Realistic,Median Centered in-structure Response Spectra for Plant Auxiliary Bldg
B11743, Provides Pvrc Damping Curve Info to Clarify Proposed Changes to Criteria Document for safety-related Piping Analyses Under SEP Topic III-6, Seismic Design Considerations
NRC-96-4862, Forwards Responses to NRC Dser Section 3.12 Open Items
ST-HL-AE-2154, Forwards Addl Info Re Seismic Model Used in Qualification of High Head Safety Injection Pump & Verification of Computer Code Used for Analysis,Per Util Agreement During Meeting W/ NRC
ST-HL-AE-2136, Forwards Revised Pages to Util FSAR Which Reflect Change in Commitment to Reg Guide 8.27 Re Radiation Protection Aspects of General Employee Training Courses
ST-HL-AE-2101, Forwards Rev to FSAR Table 3.12-1, Reg Guide Matrix, Which Revises Commitment to Reg Guide 1.133 to Show Testing & Operability Requirements for Loose Parts Monitoring Sys Contained in FSAR Rather than Tech Specs
ST-HL-AE-1674, Forwards Comments & marked-up SER (NUREG-0781) Re Operation of South Texas Projects 1 & 2
ST-HL-AE-1761, Forwards Minutes from Sqrt/Pump & Valve Operability Review Team 860916 pre-audit Meeting in Bethesda,Md.Related Correspondence
AECM-86-0127, Requests Authorization to Use Damping Recommendations of ASME Code Case N-411, Alternative Damping Values for Seismic Analysis ... as Alternative to Reg Guide 1.61 Damping Values.Fsar Changes Will Be Made in Future Update
B15912, Forwards Response to RAI Re Realistic Median Centered Spectra Generated for Resolution of USI A-46
RBG-20-998, Forwards Info in Response to SER Sections 3.10.1 & 3.10.2 & Open Item 5.FSAR Tables 3.10A-1 & 3.10B-1 Revised Providing Summary of Seismic Qualification Results
B11541, Forwards Response to Request for Addl Info Re Design Basis of Facility Spent Fuel Racks (SER Confirmatory Item 10)
B11587, Forwards Rev 3 to Safety-Related Piping Seismic Qualification Program, Per SEP Topic III.6.Changes to Damping Values,Peak Shifting & Independent Support Motion Described
ET-NRC-93-3795, Forwards Responses to NRC RAIs on AP600 from Ltrs 920923, 1001 & 09.Remainder of RAIs Contained in NRC Ltrs of 920923 & 1001 Will Be Provided Prior to 930123
A04590, Responds to NRC 850111 Request for Addl Info Re NUREG-0737, TMI Item II.D.1 Concerning Performance Testing of Relief & Safety Valves,Selection of Transients & Valve Inlet Conditions,Valve Operability & thermal-hydraulic Analysis
NSD-NRC-96-4745, Provides Info Required for Closure Re Status of Dser Open Items Re Piping Analysis
NSD-NRC-96-4759, Responds to NRC Position on Damping Values to Be Used in Piping Analyses When Sys Model Includes Flexible Equipment. Requests NRC Reaffirm Position in Subsection 3.12.5.19 of Dser That Use of 5% Damping,Acceptable for Piping Sys
RBG-19294, Forwards Responses to All Deficiencies,Unresolved Items & Details A2.8-1 & A3.3-1,per Integrated Design Insp Rept 50-458/84-18 on 840409-0601.Limited Design Review to Be Conducted by Util,Per NRC
B11215, Forwards Addl Info Requested to Close Draft SER & Structural Audit Open & Confirmatory Items 3,8,9,19,27,29,32,33,36,39 & 42.All Remaining Structural Confirmatory Items Expected to Be Resolved at Early June 1984 Meeting
B11252, Forwards Responses to Structural Audit Items 17 & 38 Re Vertical Flexibility of Floor Slabs & Peak Broadening Procedure,Respectively.Items Closed.List of Attendees from 840615 Meeting Also Encl
B11088, Summarizes Structural & Geotechnical Engineering Branch 840227-0302 Structural Design Audit of Facility Per Draft Ser.Audit Agenda,List of Attendees & Status of Items Developed During Audit Encl
NRC-94-4259, Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600 & Provide Electronic Files in Wordperfect 5.1 File for RAI 230.63 Rev 2
NRC-94-4237, Forwards Responses to NRC RAIs on AP600 from Ltrs of 940428, 29,0502,05,16 & 26.Revs of Responses Previously Submitted Also Provided
NRC-94-4208, Provides Advance Reactor Corp Technical Core Group Rept on Piping Initiatives,App G,Piping Design/Analyst Users Guide
B14892, Forwards Rev 0 to Calculation 0024-00099-MOT-01, Selection of Motion for A-46 Evaluation, Per GL 87-02,Suppl 1, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-46
BECO-94-036, Requests Approval to Use Alternative Reactor Building Response Spectra,Reg Guide 1.61 & ASME Code Case N411 Modal Damping as Alternate to Existing Design Response Spectra
NRC-94-4087, Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600 from 940126.Listing of NRC Requests for Addl Info Responded to Ltr Contained in Attachment a
LIC-90-0564, Forwards Updated, Position Paper Re Seismic Capability of Welded Steel Piping. Documentation Demonstrates No Generic Thermal or Seismic Safety Issue Exists for Welded Steel Small Diameter Piping at Facility
TXX-6571, Forwards Description of Amend 64 to Fsar,Transmitted by .Amend Provides Updates,Additions & Editorial Changes as Well as Revised Effective Page Listing.Topics Included in Revs Listed.W/O Encl
TXX-5006, Forwards Revised FSAR Pages Scheduled for Future FSAR Amend. Changes Resulted from Requalification of Class 2 & 3 Piping & Class 1,2 & 3 Pipe Supports
W3P86-1914, Requests Approval to Use ASME Code Case N-411-1 Re Alternate Damping Values for Response Spectra Analysis to Optimize Number of Seismic Restraints in Various Piping Sys.Code Case Criteria Only Used for Response Spectra Type Analyses
BECO-LTR-96-007, Requests NRC Review & Approval of Engineering Rept Used as Closure for Long Term Program (LTP) Item Number 599,in semi-annual LTP Rept Submittal of 940228
LD-92-091, Forwards Responses to 920409 Request for Addl Info Re Sys 80+ Severe Accident Design Features
PY-CEI-NRR-1499, Suppls Re Snubber Optimization of Feedwater Sys. Results of post-mechanical Stress Improvement Program Revealed No Significant Change in Size of Existing Indications
LD-92-063, Forwards Info to Supplement 920325 Responses to NRC Request for Addl Info Re Analysis of Ruptured Reactor Coolant Pump Seal Cooler Tube & Single Failure Assumptions for Analysis of moderate-energy-line Break
LD-92-043, Forwards Response to Remaining Outstanding Request for Addl Info (RAI) 230.1 & Revs to Responses for RAIs 220.6 & 220.7, Providing Figure 3.7-2 & Expanding Description of How Average Power Spectral Density Calculated,Respectively
TXX-4316, Forwards Response to Structural & Geotechnical Engineering Branch Request for Addl Info Re Questions 130.37,130.38 & 130.39 for FSAR Amend 53
LIC-93-0248, Documents Change in Commitment Made in 930211 Response to Generic Ltr 87-02 Re Usar,App F in-structure Response Spectra & Alternate Seismic Criteria & Methodology,W/Regard to Resolution of USI A-46
NRC Inspection Manual 0326, Operability Determinations & Functionality Assessments for Conditions Adverse to Quality of Safety
05000293/LER-1998-003, :on 980220,discovered That RBCCW & Tbccw Heat Exchanger Supports Outside Design Basis Seismic Requirements.Caused by Calculation Error.Reviewed Similar Calculations & Counseled Individuals Associated W/Errors
05000254/LER-1986-024-03, :on 860811,RHR Svc Water Supports Exceeded Allowable Stress During Safe Shutdown Earthquake Loading. Caused by Inadequate Design Control.Rhr Svc Water Supports Modified
ST-HL-AE-1975, Forwards Response Spectrum Piping Stress Analysis Using Multiple Support Input Response Spectra W/Asme Code Case N-411 Damping, Summary Rept,Per 870122 Commitment.Util Intends to Maintain Current Calculations Based on Study
LD-92-007, Forwards Response to Matls & Chemical Engineering Branch 910808,Structural & Geosciences Branch 910926 & Plant Sys Branch 911010 Requests for Addl Info Re SSAR - Design Certification
05000333/LER-1983-029-03, /03L-0:on 830829,hydraulic Snubber H12-94 (12-1-S-20) Failed to Lockup in Compression When Tested. Caused by Poppet Spring Lodged Between Compression Poppet & Seat.Modified Valve Assembly Installed
NSD-NRC-97-5041, Forwards Responses for Open Items Re Markup of Proposed Changes to SSAR Section 2.5 & 3.7 & Appendix 3f
AECM-82-122, Forwards Comparison of Finite Element & Lumped Parameter Soil Structure Interaction for Grand Gulf Power Block Structure, Addressing Confirmatory Item 1.10(4) in SER, NUREG-0831
B10328, Forwards Response to 810521 Meeting Questions Re SEP Topic III-06 on Seismic Design Considerations.Info Should Resolve Seismic re-evaluation Program Concerns
AECM-81-345, Forwards Response to NRC Request for Addl Info Re Soil Structure Interaction,Ultimate Heat Sink,Shared Sys & NUREG-0737,TMI Action Items.Encl Supersedes 810828 Submittals & Proposes Changes to FSAR
NRC-95-4478, Forwards Draft Rev 4 of Section 3.7 of AP600 SSAR for Discussion at Meeting on Seismic & Structural Issues During Wk of 950612-16
RC-93-0068, Forwards Description of Methodologies Being Used for Reanalysis of Major Piping Connected to Delta 75 SG (Rem 6000-6),including Reactor Coolant Loop,Main Steam,Fw, EFW & SG Blowdown Piping
05000334/LER-1980-008-01, /01T-1:on 800225,during in-plant Insp,Potential Overstress Condition of Cable Spreading Room Floor Indicated Due to Postulated Seismic Loading.Caused by Utilizing Underside of Floor to Higher Degree than Original Design
RAIO-0918-61804, LLC Supplemental Response to NRC Request for Additional Information No. 151 (Erai No. 8974) on the NuScale Design Certification Application
RAIO-0818-61199, LLC Response to NRC Request for Additional Information No. 202 (Erai No. 8911) on the NuScale Design Certification Application
RAIO-0818-61204, Transmittal of Response to NRC Request for Additional Information No. 200 on Design Certification Application
L-90-140, Forwards Response to 900316 Request for Addl Info Re Emergency Power Sys Enhancement project,specifically,1986 Edition of ANSI B31.1 Code Design Basis Code for Diesel Generator Piping
L-89-107, Forwards Supplemental Responses to NRC 890106 Request for Addl Info Re Emergency Power Sys Enhancement Project
L-86-249, Requests Permission to Apply Code Case N-411 Re Spectral Damping Values Used in Piping Stress Analysis Instead of Section Iii,Div I,Table N-1230-1 Values.Fee Paid
RAIO-1117-57010, LLC Supplemental Response to NRC Request for Additional Information No. 27 (Erai No. 8788) on the NuScale Design Certification Application
GO2-87-026, Requests Approval to Apply ASME Code Case N411 on Facility, to Enhance Sys Reliability & Personnel Safety Through Optimization of Snubber Installations.Fee Paid
GO2-83-530, Forwards Responses to NRC Telcon Questions Re B&R Technical Rept, Effect of Hydrodynamic Loads on Safety-Related Equipment & Piping Outside Containment, Transmitted by
L-82-428, Forwards Response to IE Bulletin 80-11, Masonry Wall Design, Supplementing Response to NRC 820805 Request for Addl Info
SBK-L-15025, NextEra Energy Seabrook, LLCs Fourth Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order No. EA-12
NOC-AE-130030, Supplement to Seismic Walkdown Summary Report of Items Classified as Inaccessible
RAIO-1217-57599, LLC Response to NRC Request for Additional Information No. 253 (Erai No. 9184) on the NuScale Design Certification Application
CLI-09-07, Staff Assessment of the Oyster Creek 3-D Finite Element Analysis of the Drywell Shell
NLS2006028, Report HI-2043224, Licensing Report on the Wet Fuel Storage Capacity Expansion at Cooper Nuclear Station
W3F1-2021-0050, Response to U. S. Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt
ST-HL-AE-1509, Forwards Response to Draft SER or FSAR Item C 3.7-1 Re Seismic Design Confirming Frequencies of Floor Subsystems in Each Bldg Higher than Corresponding Limit.Encl marked-up FSAR Pages Will Be Incorporated in Future Amend
B11160, Forwards Responses to Structural & Geotechnical Engineering Branch Draft SER Open Items,Per Bj Youngblood & Items Identified During 840323 Structural Audit.Fsar Amends Will Be Included in Future Amend
DCL-85-306, Applicant Exhibit A-2,consisting of Forwarding Reracking of Spent Fuel Pools,Diablo Canyon Units 1 & 2, Dtd Sept 1985.Unit 1 Refueling Outage Could Occur as Early as Jul 1986