|
---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARML17347B5881990-03-0101 March 1990 Responds to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. Info Covers Time Spent by Key Power Plant Managers in Responding to Operational Insps & Audits ML18094B3221990-02-28028 February 1990 Forwards Executed Amend 14 to Indemnity Agreement B-74 ML15217A1031990-02-28028 February 1990 Forwards Semiannual Radioactive Effluent Release Rept for Jul-Dec 1989 for McGuire Nuclear Station Units 1 & 2 & Revised Process Control Programs & Offsite Dose Calculation Manuals ML20011F3821990-02-26026 February 1990 Confirms Amount Electronically Transferred to Us Dept of Treasury,Nrc on 900223 for Payment of NRC Review Fees of 10CFR50 Applications & 10CFR55 Svcs Per 10CFR170,for Period of 890101-0617 for Listed Invoices ML20055C3921990-02-26026 February 1990 Approves Util 900214 Request for Use of B&W Steam Generator Plugs W/Alloy 690 as Alternative to Alloy 600.Alternate Matl Is nickel-base Alloy (ASME Designation SB-166) ML20006G0621990-02-22022 February 1990 Forwards Revised Proprietary Pages to DPC-NE-2004, Core Thermal Hydraulic Methodology Using VIPRE-01, Reflecting Minor Methodology Changes Made During Review & Approval Process.Pages Withheld ML20006E5881990-02-20020 February 1990 Forwards Proprietary Response to NRC 890725 Questions Re Vipre Core Thermal Hydraulic Section of Topical Rept DPC-NE-3000 & Rev 2 to Pages 3-69,3-70,3-78 & 3-79 of Rept. Encls Withheld (Ref 10CFR2.790) ML20006E1441990-02-16016 February 1990 Forwards Suppl to Rev 1 to Updated FSAR for Braidwood Station,Units 1 & 2 & Byron Station,Units 1 & 2,per 881214 & 891214 Submittals ML20006E9071990-02-16016 February 1990 Discusses Plants Design Control Program.Util Adopted Concept of Design Change Implementation Package (Dcip).Dcip Will Contain or Ref Design Change Notice Prepared Per Approved Procedures ML20006E4201990-02-14014 February 1990 Requests NRC Approval for Use of Alloy 690 Steam Generator Tube Plugs for Facility,Prior to 900301,pending Final ASME Approval of Code Case for Alloy 690 ML18094B3291990-02-14014 February 1990 Forwards Printouts Containing RW-859 Nuclear Fuel Data for Period Ending 891231 & Diskettes ML20011E6151990-02-12012 February 1990 Forwards Revs 1 to Security Plan & Security Training & Qualification Plan & Rev 2 to Security Contingency Plan. Salem Switchyard Project Delayed.Revs Withheld (Ref 10CFR73.21) ML20011E5571990-02-0808 February 1990 Forwards Us Bankruptcy Court for Eastern District of Tennessee Orders & Memorandum on Debtors Motion to Alter or Amend Order & Opinion Re Status of Sales Agreement Between DOE & Alchemie.Doe Believes Agreement Expired on 890821 ML20011E4991990-02-0606 February 1990 Discusses Liability & Funding Requirements Re NRC Decommissioning Funding Rules & Verifies Understanding of Rules.Ltr from NRC Explaining Liability & Requirements of Rule Requested ML20011E5981990-02-0505 February 1990 Requests That Listed Individuals Be Deleted from Svc List for Facilities.Documents Already Sent to Dept of Environ Protection of State of Nj ML20006D6911990-02-0202 February 1990 Provides Alternative Design Solution to Dcrdr Implementation at Facilities.Simpler Design Devised,Using Eyelet Screw Inserted in Switch Nameplate Which Is Identical to Providing Caution Cards in Close Proximity to Switch Handle ML20006C5661990-01-31031 January 1990 Provides Certification Re Implementation of Fitness for Duty Program Per 10CFR26 at Plants ML20006D6611990-01-29029 January 1990 Advises That 900117 License Amend Request to Remove Certain cycle-specific Parameter Limits from Tech Specs Inadvertently Utilized Outdated Tech Specs Pages.Requests That Tech Specs Changes Made Via Amends 101/83 Be Deleted ML20011E2521990-01-29029 January 1990 Forwards Proprietary Safety Analysis Physics Parameters & Multidimensional Reactor Transients Methodology. Three Repts Describing EPRI Computer Code Also Encl.Proprietary Rept Withheld (Ref 10CFR2.790) ML20006C6711990-01-29029 January 1990 Responds to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Plants Have Established Preventive Maint Program for Intake Structure & Routine Treatment of Svc Water Sys W/Biocide to Control Biofouling ML20006B7961990-01-29029 January 1990 Forwards Summaries of Latest ECCS Evaluation Model Changes ML18153C0951990-01-29029 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Belief in Appropriateness to Address Generic Ltr 89-13 Concerns within Context of Established Programmatic Improvements Noted ML20006D2431990-01-26026 January 1990 Provides Info Re Emergency Response Organization Exercises for Plants.Exercises & Callouts Would Necessitate Activation of Combined Emergency Operations Facility Approx Eight Times Per Yr,W/Some Being Performed off-hours & Unannounced ML18153C0871990-01-26026 January 1990 Responds to NRC Bulletin 89-003, Potential Loss of Required Shutdown Margin During Refueling Operations. Refueling Procedures to Be Revised & Familiarization Sessions Will Be Conducted Prior to Each Refueling Outage ML18094B2861990-01-26026 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Problems Affecting Safety-Related Equipment. Aggressive Program of Monitoring,Insp & Matl Replacement Initiated in Advance of Generic Ltr 89-13 Issuance ML19354E4191990-01-25025 January 1990 Comments Re Issuance of OL Amends & Proposed NSHC Determination Re Transfer of Operational Mgt Control of Plants & Views on anti-trust Issues Re Application for Amend for Plants ML19354E6711990-01-24024 January 1990 Requests Approval to Use Alloy 690 Plugs as Alternative to Requirements of 10CFR55(a),codes & Stds for Plants Prior to 900226 ML17347B5451990-01-24024 January 1990 Informs of Plans to Apply ASME Code Case N-356 at Plants to Allow Certification Period to Be Extended to 5 Yrs.Rev to Inservice Insp Programs Will Include Use of Code Case ML19354E4461990-01-22022 January 1990 Forwards Proprietary Rev 1 to DPC-NE-2001, Fuel Mechanical Reload Analysis Methodology for MARK-BW Fuel, Adding Section Re ECCS Analysis Interface Criteria & Making Associated Administrative Changes.Rev Withheld ML19354E4451990-01-22022 January 1990 Submits Update on Status of RHR Sys Iconic Display at Facilities,Per Generic Ltr 88-17 Re Loss of Dhr.Computer Graphics Display Data in Real Time & Reflect Status of Refueling Water Level & RHR Pump Parameters ML20005G7161990-01-20020 January 1990 Forwards Rev 1 to Updated FSAR for Braidwood & Byron Units 1 & 2.Changes in Rev 1 Include Facility & Procedures Which Were in Effect as of 890610.W/o Encl ML20006A8001990-01-19019 January 1990 Forwards Response to NRC 891220 Ltr Re Violations Noted in Plant Insps.Response Withheld (Ref 10CFR73.21) ML16152A9091990-01-18018 January 1990 Forwards Public Version of Rev 33 to Crisis Mgt Implementing Procedure CMIP-1, Recovery Manager & Immediate Staff & Rev 24 to CMIP-2, News Group Plan. W/900131 Release Memo ML18153C0771990-01-17017 January 1990 Forwards North Anna Power Station Emergency Plan Table 5.1, 'Min Staffing Requirements for Emergencies' & Surry... Table 5.1, 'Min Staffing Requirements...', for Approval,Per 10CFR50.54(q),NUREG-0654 & NUREG-0737,Suppl 1 ML20006A6241990-01-16016 January 1990 Forwards Draft Qualified Master Trust Agreement for Decommissioning of Nuclear Plants,For Review.Licensee Will Make Contributions to Qualified & Nonqualified Trust as Appropriate ML20006A2011990-01-16016 January 1990 Responds to NRC Bulletin 89-002 Re Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel in Anchor Darling Swing Check Valves.Eight Subj Valves Identified in Peach Bottom Units 1 & 2 & Will Be Returned to Mfg ML18153C0731990-01-15015 January 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or or Valves.... Util Replaced Studs in twenty-five Valves ML20006A8201990-01-10010 January 1990 Forwards Errata to Rev 3 to BAW-1543,Tables 3-20 & E-1 of Master Integrated Reactor Vessel Surveillance Program Reflecting Changes in Insertion Schedule for A5 Capsule for Davis-Besse & Crystal River ML20006B8821990-01-10010 January 1990 Reissued Ltr Correcting Date of Util Ltr to NRC Which Forwarded Updated FSAR for Byron/Braidwood Plants from 881214 to 891214.W/o Updated FSARs ML20005G6431990-01-10010 January 1990 Responds to Generic Ltr 89-21 Re Implementation of USI Requirements,Consisting of Revised Page to 891128 Response, Moving SER Ref from USI A-10 to A-12 for Braidwood ML20005G7601990-01-0404 January 1990 Forwards Public Version of Rev 33 to Crisis Mgt Plan. Privacy Info Should Be Deleted Prior to Placement in Pdr.W/ D Grimsley 900118 Release Memo ML18153C0491990-01-0303 January 1990 Advises of Implementation of fitness-for-duty Program Which Complies w/10CFR26.Util Support Objective of Providing Assurances That Nuclear Power Plant Personnel Will Perform Tasks in Reliable & Trustworthy Manner ML20005F4641990-01-0303 January 1990 Advises That Licensee Implemented 10CFR26 Rule Re fitness-for-duty Program W/One Exception.Util Has Not Completed Background Check for Some of Program Administrators.Checks Expected to Be Completed by 900105 ML18094B2331990-01-0303 January 1990 Certifies Util Implementation of fitness-for-duty Program, Per 10CFR26.Training Element Required by Rule Completed on 891215.Chemical Testing for Required Substances Performed at Min Prescribed cut-off Levels,Except for Marijuana ML17347B5051990-01-0202 January 1990 Certifies That Util Has fitness-for-duty Program Which Meets Requirements of 10CFR26.Util Adopted cut-off Levels Indicated in Encl ML20042D3731990-01-0202 January 1990 Forwards Revised Crisis Mgt Implementing Procedures, Including Rev 32 to CMIP-1,Rev 29 to CMIP-4,Rev 33 to CMIP-5,Rev 38 to CMIP-6,Rev 37 to CMIP-7,Rev 32 to CMIP-9, Rev 1 to CMIP-14 & Rev 30 to CMIP-21 ML17347B4961989-12-28028 December 1989 Responds to Generic Ltr 89-10, Safety-Related Motor- Operated Valve Testing & Surveillance. Util Considering Expansion of Plants to Include Addl safety-related & Position Changeable Valves W/ Emphasis on Maint & Testing ML20042D3381989-12-28028 December 1989 Forwards Response to Generic Ltr 89-10, Safety-Related Motor-Operated Valve Testing & Surveillance. Util Will Comply W/Ltr Recommendations W/Noted Exceptions.Response to Be Completed When Ltr Uncertainties Cleared ML18094B2201989-12-27027 December 1989 Advises of Intent to Provide follow-up Response to Generic Ltr 89-10 by 900831 to Describe Status of Program, Recommendation Exceptions & Any Schedule Adjustments ML18094B2291989-12-27027 December 1989 Requests to Apply ASME Section XI Code Case N-460 to Facilities Re Reduction in Exam Coverage on Class 1 & 2 Welds.Fee Paid 1990-03-01
[Table view] Category:ENGINEERING/CONSTRUCTION/CONSULTING FIRM TO NRC
MONTHYEARML19327A8691989-09-0707 September 1989 Submits Info Re Alchemie & Anderson County Bank Financing Transaction ML19332F2171989-07-10010 July 1989 FOIA Request for Documents Re Communications Between Ofcs of Edo,Deputy Edo,Ofc of Director,Regional Administrators & Commissioners Ofcs Re Plants During period,890301-0615 ML20246F1311989-06-26026 June 1989 FOIA Request for Minutes of Meeting Ref in 820210 Memo from NRR Re Design & Const Assurance for Upcoming OL Cases ML20245D9851989-06-22022 June 1989 Forwards 21 Insp Rept Executive Summaries,Per NRC Contract NRC-03-87-029,Task Order 037.Individual Quality Evaluations of Insp Repts Also Prepared ML20247P9411989-05-17017 May 1989 FOIA Request for Final Open Item Transmittal Ltrs Per NRC Insp Procedure 94300B for Listed Plants ML20245C1421989-04-0303 April 1989 Forwards Endorsements 75,108,108,96 & 110 to Maelu Policies MF-56,MF-26,MF-58,MF-39 & MF-52,respectively & Endorsements 93,129,127,109 & 122 to Nelia Policies NF-186,NF-76,NF-188, NF-151 & NF-173,respectively ML20247N1551989-03-31031 March 1989 Forwards Revised Proprietary Conformance of HPCS Div to NUMARC 87-00 Alternate AC Criteria, for Review as Result of Comments from 890216 Meeting.Rept Withheld ML20246M7331989-03-15015 March 1989 Responds to NRC Info Notice 88-082, Torus Shells W/ Corrosion & Degraded Coatings in BWR Containments. Summary of Relevant Projects for Various Utils Successfully Employing Underwater Alternative to Draining Vessel Encl ML20246N1281989-02-27027 February 1989 FOIA Request for Jl Smith to NRC Re Spent Fuel Shipment from Brunswick Nuclear Power Station to Harris Plant ML17285A2351989-02-0606 February 1989 Forwards Proprietary Draft Conformance of HPCS Div to NUMARC 87-00 App B Aac Criteria, for 890214 Meeting ML17285A2341989-01-0606 January 1989 Discusses Issues Highlighted at BWR/6 Alternate Ac Task Force Meeting on 881115,including Need for Capability of Div III Sys to Maintain Plant in Safe Shutdown Condition (Hot Shutdown) for Min of 4 H ML20206H0511988-11-14014 November 1988 Urges Relicensing of Pilgrim & Expedited Operation of Seabrook.Newspaper Clipping Encl ML20150D5721988-03-0808 March 1988 Provides Summary of Utils Test Results & Calculations on Emergency Diesel Generators,Including Review of Design of Static Exciter & Voltage Regulator for Emergency Diesel Generators ML20196C1591988-02-0303 February 1988 Forwards Monthly Progress Rept P-C6177-5, Independent Analysis & Assessment, for Period Ending 880131 ML20147G0741988-01-18018 January 1988 FOIA Request for All Documents Re NRC Investigation of Wg Dick Allegations About S&W & Lilco Re Performing NRC Instructions to Bring Facility Up to Fuel Load Stds ML20235A1251987-12-16016 December 1987 Forwards Info Re Resource Technical Svcs,Inc,Including Summary of NRC Contract Work,Nrc Form 26 for Three Existing Contracts,Audit Info,Work History & Lists of Expertise Available for Special Insps & of Current Resource Svcs ML20237B8051987-11-25025 November 1987 FOIA Request That Encls to Listed Documents,Including NRC Forwarding Amend 1 to License NPF-73,be Placed in PDR ML20236S4291987-10-20020 October 1987 FOIA Request for Listed Documents,Including Encls from NRC Requesting Addl Info on Gpu Topical Repts TR-033 & TR-040 & Encl to NRC Meeting Summary Re SPDS ML20236U5221987-10-19019 October 1987 FOIA Request for LERs for Listed Plants,Including All Attachments & Encls from Original Documents ML20235V1321987-08-28028 August 1987 Forwards EGG-NTA-7471, Technical Evaluation Rept,Reactor Trip Sys Reliability Conformance to Item 4.5.2 of Generic Ltr 83-28.... Based on Licensee Responses,Plants Reviewed Conform W/Exceptions Listed in Section 14 ML17342A7741987-07-13013 July 1987 Forwards Technical Evaluation of Rept, Retran Code: Transient Analysis Model Qualification, Dtd Jul 1985. Criteria for Use of Single & Two Loop Plant Models Listed. NRC Audit of Util QA Procedure Recommended ML20235K8731987-07-0909 July 1987 Informs That Tayloe Assoc Cannot Produce Mag or nine-track Tapes of Hearing Transcripts Until NRC Finalizes Arrangements W/Others to Provide Lexis Format,Including Library & File Numbers & Segmentation Info ML20237J2141987-07-0202 July 1987 FOIA Request for Listed Documents Ref in NUREG-1150 & Related Contractor Repts ML20238E3011987-06-29029 June 1987 FOIA Request for All Documents Described in App,Including Listed LERs & Revs for Plants,W/Original Attachments & Encls ML18052B1911987-06-17017 June 1987 Forwards EGG-NTA-7720, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface:Calvert Cliffs-1 & -2,Millstone-2 & Palisades, Final Informal Rept. Plants Conform to Generic Ltr Item ML20234B6211987-05-12012 May 1987 Requests That Listed Plants Be Added to Encl 870508 FOIA Request Re 94300 Region Input on Plant Readiness ML20234B6571987-05-0808 May 1987 FOIA Request for Placement,In Pdr,Region Input to NRC Headquarters,Nrr Re Status of Listed Plants in Terms of Plant Readiness for OL IE Manual,Chapter 94300 ML20214R4051987-05-0808 May 1987 FOIA Request for Region Input to NRR Re Status of Listed Plants Readiness for Ol,Per IE Manual Chapter 94300 ML18150A1861987-05-0101 May 1987 Forwards EGG-NTA-7612, Conformance to Generic Ltr 83-28, Item 2.2.2 - Vendor Interface Programs for All Other Safety- Related Components,North Anna Units 1 & 2 & Surry Units 1 & 2, Final Informal Rept ML20214Q8801987-04-17017 April 1987 Forwards EGG-NTA-7591, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28 Reactor Trip Sys Vendor Interface,Hatch-1 & 2,Millstone-1, Final Rept.Plants Conform to Item ML20214R0621987-04-17017 April 1987 Forwards EGG-NTA-7613, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface,Arnold, Brunswick-1 & 2, Final Rept.Plants Conform to Item ML18150A1171987-04-14014 April 1987 Forwards Final rept,EGG-NTA-7625, Conformance to Item 2.1 (Part 2) Generic Ltr 83-28,Reactor Trip Sys Vendor Interface,North Anna 1 & 2 & Surry 1 & 2. ML20214R8611987-03-27027 March 1987 Forwards EGG-NTA-7614, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface:Cook-1 & -2,Haddam Neck, Final Informal Rept.Facilities Conform to Generic Ltr ML20214R1361987-03-26026 March 1987 Forwards Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface:Maine Yankee, St Lucie 1 &-2 & Waterford 3, Final Rept.Plants Conform to Generic Ltr ML20214R1861987-03-26026 March 1987 Forwards Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components,Haddam Neck & Millstone 1,2 & 3, Final Rept ML20211D6631987-02-12012 February 1987 Notifies of 830204 Meeting W/Util,Idvp,Nrc & BNL in San Francisco,Ca to Discuss Status of Containment Annulus Steelwork & Status of Auxiliary Bldg ML20211D7031987-02-12012 February 1987 Notifies of 830209 Meeting in San Francisco,Ca to Discuss Shake Table Tests of Electrical Equipment ML20211D7441987-02-12012 February 1987 Notifies of 830517 Meeting in San Francisco,Ca to Discuss Development of Piping Stress Intensification Factor ML20209A8551987-01-16016 January 1987 FOIA Request for Documents to Be Placed in Pdr,Including NRC Re Calibr of Test Equipment allegation,1986 Inservice Insp Repts for McGuire 1 & Surry 1 & NRC 830307 SALP on Nine Mile Point 2 ML20207K0151986-12-19019 December 1986 FOIA Request That Encls to Insp Rept 50-247/86-26,Byron Semiannual Radioactive Effluent Rept & Millstone 1 & 2 SALP Rept Be Placed in PDR ML20211P2051986-11-24024 November 1986 FOIA Request for La Crosse & Big Rock Point Semiannual Effluent Repts & Turkey Point & St Lucie SALP Repts ML20214R8831986-11-0505 November 1986 FOIA Request for Encls to 860821 SALP Repts ML20213F8841986-10-30030 October 1986 FOIA Request for Encls to NRC 860724 & 31 Requests for Addl Info Re Vermont Yankee Spent Fuel Pool Expansion & Browns Ferry Seismic Reevaluation Program,Respectively ML20209D1691986-10-29029 October 1986 Forwards Rev 3 to EGG-EA-7035, Conformance to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Braidwood Units 1 & 2,Bryon Station Units 1 & 2,Callaway Plant Unit 1,Indian Point.... Licensees Conform to All Items W/Exception of Trojan ML20214J9761986-10-15015 October 1986 FOIA Request for Containment Event Trees for Listed Facilities,Technical Repts & Memoranda Re Interpretation & Quantification & Identification of FIN Numbers,Contractors & Investigators Involved in Creation/Analysis of Event Trees ML20214K0231986-10-15015 October 1986 FOIA Request for All Documentation Re Accident Sequence Evaluation Program Repts Re Listed Facilities in Preparation for NUREG-1150 ML20245A4451986-09-25025 September 1986 Forwards Revised Draft EGG-NTA-7188, Conformance to Generic Ltr 83-28 Item 2.1 (Part 1) Equipment Classification Hope Creek,Peach Bottom 2 & 3,Perry 1 & 2 & Pilgrim 1 ML20215M9941986-05-21021 May 1986 FOIA Request for Documents Re Eg Case 771201 Memo to Tj Mctiernan Concerning Chronology of Events Associated W/ Facility Fault Assessments ML20210T4441986-03-27027 March 1986 FOIA Request for Initial (Cycle 1) Startup Test Repts & Suppls for Seven Plants & Original Monthly Operating Rept for June 1983 for Virgil C Summer Plant ML20195C5951986-03-14014 March 1986 FOIA Request for Structural Integrity Test Repts for Shoreham,Limerick-1,Nine Mile Point-2 & Susquehanna 1989-09-07
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17347B5881990-03-0101 March 1990 Responds to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. Info Covers Time Spent by Key Power Plant Managers in Responding to Operational Insps & Audits ML18094B3221990-02-28028 February 1990 Forwards Executed Amend 14 to Indemnity Agreement B-74 ML15217A1031990-02-28028 February 1990 Forwards Semiannual Radioactive Effluent Release Rept for Jul-Dec 1989 for McGuire Nuclear Station Units 1 & 2 & Revised Process Control Programs & Offsite Dose Calculation Manuals ML20011F3821990-02-26026 February 1990 Confirms Amount Electronically Transferred to Us Dept of Treasury,Nrc on 900223 for Payment of NRC Review Fees of 10CFR50 Applications & 10CFR55 Svcs Per 10CFR170,for Period of 890101-0617 for Listed Invoices ML20006G0621990-02-22022 February 1990 Forwards Revised Proprietary Pages to DPC-NE-2004, Core Thermal Hydraulic Methodology Using VIPRE-01, Reflecting Minor Methodology Changes Made During Review & Approval Process.Pages Withheld ML20006E5881990-02-20020 February 1990 Forwards Proprietary Response to NRC 890725 Questions Re Vipre Core Thermal Hydraulic Section of Topical Rept DPC-NE-3000 & Rev 2 to Pages 3-69,3-70,3-78 & 3-79 of Rept. Encls Withheld (Ref 10CFR2.790) ML20006E1441990-02-16016 February 1990 Forwards Suppl to Rev 1 to Updated FSAR for Braidwood Station,Units 1 & 2 & Byron Station,Units 1 & 2,per 881214 & 891214 Submittals ML20006E9071990-02-16016 February 1990 Discusses Plants Design Control Program.Util Adopted Concept of Design Change Implementation Package (Dcip).Dcip Will Contain or Ref Design Change Notice Prepared Per Approved Procedures ML20006E4201990-02-14014 February 1990 Requests NRC Approval for Use of Alloy 690 Steam Generator Tube Plugs for Facility,Prior to 900301,pending Final ASME Approval of Code Case for Alloy 690 ML20011E6151990-02-12012 February 1990 Forwards Revs 1 to Security Plan & Security Training & Qualification Plan & Rev 2 to Security Contingency Plan. Salem Switchyard Project Delayed.Revs Withheld (Ref 10CFR73.21) ML20011E5571990-02-0808 February 1990 Forwards Us Bankruptcy Court for Eastern District of Tennessee Orders & Memorandum on Debtors Motion to Alter or Amend Order & Opinion Re Status of Sales Agreement Between DOE & Alchemie.Doe Believes Agreement Expired on 890821 ML20011E4991990-02-0606 February 1990 Discusses Liability & Funding Requirements Re NRC Decommissioning Funding Rules & Verifies Understanding of Rules.Ltr from NRC Explaining Liability & Requirements of Rule Requested ML20011E5981990-02-0505 February 1990 Requests That Listed Individuals Be Deleted from Svc List for Facilities.Documents Already Sent to Dept of Environ Protection of State of Nj ML20006D6911990-02-0202 February 1990 Provides Alternative Design Solution to Dcrdr Implementation at Facilities.Simpler Design Devised,Using Eyelet Screw Inserted in Switch Nameplate Which Is Identical to Providing Caution Cards in Close Proximity to Switch Handle ML20006C5661990-01-31031 January 1990 Provides Certification Re Implementation of Fitness for Duty Program Per 10CFR26 at Plants ML20006B7961990-01-29029 January 1990 Forwards Summaries of Latest ECCS Evaluation Model Changes ML20006C6711990-01-29029 January 1990 Responds to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Plants Have Established Preventive Maint Program for Intake Structure & Routine Treatment of Svc Water Sys W/Biocide to Control Biofouling ML20006D6611990-01-29029 January 1990 Advises That 900117 License Amend Request to Remove Certain cycle-specific Parameter Limits from Tech Specs Inadvertently Utilized Outdated Tech Specs Pages.Requests That Tech Specs Changes Made Via Amends 101/83 Be Deleted ML20011E2521990-01-29029 January 1990 Forwards Proprietary Safety Analysis Physics Parameters & Multidimensional Reactor Transients Methodology. Three Repts Describing EPRI Computer Code Also Encl.Proprietary Rept Withheld (Ref 10CFR2.790) ML18153C0951990-01-29029 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Belief in Appropriateness to Address Generic Ltr 89-13 Concerns within Context of Established Programmatic Improvements Noted ML18094B2861990-01-26026 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Problems Affecting Safety-Related Equipment. Aggressive Program of Monitoring,Insp & Matl Replacement Initiated in Advance of Generic Ltr 89-13 Issuance ML18153C0871990-01-26026 January 1990 Responds to NRC Bulletin 89-003, Potential Loss of Required Shutdown Margin During Refueling Operations. Refueling Procedures to Be Revised & Familiarization Sessions Will Be Conducted Prior to Each Refueling Outage ML20006D2431990-01-26026 January 1990 Provides Info Re Emergency Response Organization Exercises for Plants.Exercises & Callouts Would Necessitate Activation of Combined Emergency Operations Facility Approx Eight Times Per Yr,W/Some Being Performed off-hours & Unannounced ML19354E4191990-01-25025 January 1990 Comments Re Issuance of OL Amends & Proposed NSHC Determination Re Transfer of Operational Mgt Control of Plants & Views on anti-trust Issues Re Application for Amend for Plants ML19354E6711990-01-24024 January 1990 Requests Approval to Use Alloy 690 Plugs as Alternative to Requirements of 10CFR55(a),codes & Stds for Plants Prior to 900226 ML17347B5451990-01-24024 January 1990 Informs of Plans to Apply ASME Code Case N-356 at Plants to Allow Certification Period to Be Extended to 5 Yrs.Rev to Inservice Insp Programs Will Include Use of Code Case ML19354E4451990-01-22022 January 1990 Submits Update on Status of RHR Sys Iconic Display at Facilities,Per Generic Ltr 88-17 Re Loss of Dhr.Computer Graphics Display Data in Real Time & Reflect Status of Refueling Water Level & RHR Pump Parameters ML19354E4461990-01-22022 January 1990 Forwards Proprietary Rev 1 to DPC-NE-2001, Fuel Mechanical Reload Analysis Methodology for MARK-BW Fuel, Adding Section Re ECCS Analysis Interface Criteria & Making Associated Administrative Changes.Rev Withheld ML20005G7161990-01-20020 January 1990 Forwards Rev 1 to Updated FSAR for Braidwood & Byron Units 1 & 2.Changes in Rev 1 Include Facility & Procedures Which Were in Effect as of 890610.W/o Encl ML20006A8001990-01-19019 January 1990 Forwards Response to NRC 891220 Ltr Re Violations Noted in Plant Insps.Response Withheld (Ref 10CFR73.21) ML16152A9091990-01-18018 January 1990 Forwards Public Version of Rev 33 to Crisis Mgt Implementing Procedure CMIP-1, Recovery Manager & Immediate Staff & Rev 24 to CMIP-2, News Group Plan. W/900131 Release Memo ML18153C0771990-01-17017 January 1990 Forwards North Anna Power Station Emergency Plan Table 5.1, 'Min Staffing Requirements for Emergencies' & Surry... Table 5.1, 'Min Staffing Requirements...', for Approval,Per 10CFR50.54(q),NUREG-0654 & NUREG-0737,Suppl 1 ML20006A2011990-01-16016 January 1990 Responds to NRC Bulletin 89-002 Re Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel in Anchor Darling Swing Check Valves.Eight Subj Valves Identified in Peach Bottom Units 1 & 2 & Will Be Returned to Mfg ML20006A6241990-01-16016 January 1990 Forwards Draft Qualified Master Trust Agreement for Decommissioning of Nuclear Plants,For Review.Licensee Will Make Contributions to Qualified & Nonqualified Trust as Appropriate ML18153C0731990-01-15015 January 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or or Valves.... Util Replaced Studs in twenty-five Valves ML20005G6431990-01-10010 January 1990 Responds to Generic Ltr 89-21 Re Implementation of USI Requirements,Consisting of Revised Page to 891128 Response, Moving SER Ref from USI A-10 to A-12 for Braidwood ML20006A8201990-01-10010 January 1990 Forwards Errata to Rev 3 to BAW-1543,Tables 3-20 & E-1 of Master Integrated Reactor Vessel Surveillance Program Reflecting Changes in Insertion Schedule for A5 Capsule for Davis-Besse & Crystal River ML20006B8821990-01-10010 January 1990 Reissued Ltr Correcting Date of Util Ltr to NRC Which Forwarded Updated FSAR for Byron/Braidwood Plants from 881214 to 891214.W/o Updated FSARs ML20005G7601990-01-0404 January 1990 Forwards Public Version of Rev 33 to Crisis Mgt Plan. Privacy Info Should Be Deleted Prior to Placement in Pdr.W/ D Grimsley 900118 Release Memo ML18094B2331990-01-0303 January 1990 Certifies Util Implementation of fitness-for-duty Program, Per 10CFR26.Training Element Required by Rule Completed on 891215.Chemical Testing for Required Substances Performed at Min Prescribed cut-off Levels,Except for Marijuana ML18153C0491990-01-0303 January 1990 Advises of Implementation of fitness-for-duty Program Which Complies w/10CFR26.Util Support Objective of Providing Assurances That Nuclear Power Plant Personnel Will Perform Tasks in Reliable & Trustworthy Manner ML20005F4641990-01-0303 January 1990 Advises That Licensee Implemented 10CFR26 Rule Re fitness-for-duty Program W/One Exception.Util Has Not Completed Background Check for Some of Program Administrators.Checks Expected to Be Completed by 900105 ML20042D3731990-01-0202 January 1990 Forwards Revised Crisis Mgt Implementing Procedures, Including Rev 32 to CMIP-1,Rev 29 to CMIP-4,Rev 33 to CMIP-5,Rev 38 to CMIP-6,Rev 37 to CMIP-7,Rev 32 to CMIP-9, Rev 1 to CMIP-14 & Rev 30 to CMIP-21 ML17347B5051990-01-0202 January 1990 Certifies That Util Has fitness-for-duty Program Which Meets Requirements of 10CFR26.Util Adopted cut-off Levels Indicated in Encl ML17347B4961989-12-28028 December 1989 Responds to Generic Ltr 89-10, Safety-Related Motor- Operated Valve Testing & Surveillance. Util Considering Expansion of Plants to Include Addl safety-related & Position Changeable Valves W/ Emphasis on Maint & Testing ML20042D3381989-12-28028 December 1989 Forwards Response to Generic Ltr 89-10, Safety-Related Motor-Operated Valve Testing & Surveillance. Util Will Comply W/Ltr Recommendations W/Noted Exceptions.Response to Be Completed When Ltr Uncertainties Cleared ML18094B2201989-12-27027 December 1989 Advises of Intent to Provide follow-up Response to Generic Ltr 89-10 by 900831 to Describe Status of Program, Recommendation Exceptions & Any Schedule Adjustments ML18094B2291989-12-27027 December 1989 Requests to Apply ASME Section XI Code Case N-460 to Facilities Re Reduction in Exam Coverage on Class 1 & 2 Welds.Fee Paid ML20005E1911989-12-26026 December 1989 Forwards Revised Page 2 Correcting Plant Implementation Date for USI A-24 Requirements in Response to Generic Ltr 89-21 ML18153C0261989-12-26026 December 1989 Responds to Generic Ltr 89-10 Re safety-related motor-operated Valve Testing & surveillance.Motor-operated Valve Program Structured to Allow Similar Approach to motor-operated Dampers.Results Will Be Submitted in 30 Days 1990-03-01
[Table view] |
Text
_-- -_-_ . -
./ -
$1..-w
.m C
)p%@ap$?
g j ? W~
)
HOLMES 8 N A RV E R . I N C.Lg yne COPY.
E N C I N E E R 5 C O N S T R U C T O R.5
- ere souTw riovcnoA sTntrT LoS ANGELES 17 waoisom um LOS ANGELES HONOLOLO I ..
1 l
. July 16, 1963 to 4
Dr. Robert H. Bryan, Chief '
i l
Research and Power Reactor Safety Branch ~ C l Division of Licensing and Regulation [2- S -8 U. S. Atomic Energy Commission 1 Washington 25, D. C. N g
Dear Sir:
t _
6 The following material relating to Bodega Bay Atomic Park Unit No. 1 is transmitted herewith for your information. It relates to questions which have been raised in meetings at Bethesda and 3 others stemming from the Argonne meeting of July 2. !
Item 1 attempts to shed some light on the question of the costs involved in providing seismic protection, in response to the l l query which arose in the meetings of June 27 and 28 at Bethesda.
I y , Item 2 comprises an initial attempt at some conclusions for the j . final report as requested in the Bethesda meetings of June 27 and
- 28.
Item 3 considers the advantages and disadvantages of using working stresses as a design basis as compared to using yield point stresses.
Item 4 comments on the spectra considered at the Argonne meeting of July 1.
Item 5 discusses items which the writer feels should be included in.
the revised criteria, which it is anticipated, will be submitted by the Company.
....7-g 3& '
r.m.,.
8709220215 851217 ES -665 PDR h 5106
)
R.H. Bryan 7/16/63 Comments relating to the problem of defining the critical items of the reactor and the special requirements needed to insure their l integrity were submitted to you at Chicago on July 1,1963, In a preliminary draft dated June 29, 1963, titled Special Requirements for Earthquake Resistant Design.
It is believed that the data herein and that previously submitted consider the currently important factors to the extent feasible l under existing circumstances. Accordingly, I will await further direction from you before proceeding further.
l Sincerely yours, HOLMES & NARVER, INC.
R. A. Williamson enc 1.
l l
6 l 1
1 I
l l
l i
- -- -.w. . p. . . .e.... . , - . . _ , , . . - , . .
a u _-___ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _
[' dated' July' 16,l 1963 d.
C. T h . ?
- 1. Cost of Seismic Protection p g ylle Co;7 ( ,4-c 7 ps/
During our discussions on June 27 in Bethesda, Mr. F.N.*, Watson.
indicated an interest in knowing what the cost penalty might be"in de-signing according to the seismic criteria of the PHSR as com, pared to the criteria conventionally used. This is a particularly difficult l question to answer which can be resolved accurately only on the basis of comparative preliminary designs. The following material, based 1
on updated cost information for concrete and reinforcing steel', amends
]
some previous tentative' conclusions.
To provid'e a very approximate answer, calculations were made for the walls of the portion of the reactor building above elevation + 62.
It is in this portion of the structure that seismic effects appear to have a maximum influence on cost. It was assumed that changes in seismic input would be met by variations in the amount of reinforcing steel, without changing wall thickness, and that concrete with approximately 200 lbs. of reinforcing steel per yd. would cost $65/yd . Reinforcing steel cost was taken as $0.12/lb. It was further estimated that under conventional assumptions the steel percentages in each face would be i
! 0. 25, whereas under the seismic input proposed, on the basis of rough f ., calculations, the percentages of vertical steel in each face would be-c ome 1. 0.
1 The calculations led to the conclusion that the more severe seismic forces being considered would increase cost of the reinforced concrete in these walls by about 40% as compared to conventional construction in seismic areas. This is obviously an upper bound which is far too high to serve as an accurate index for the whole structure.
t A major portion of the concrete is located in the' substructure, l i
HoLMis a N ARVER INC j l
i t_ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -. _ __ _ _ _ . . _ _ _ _ .__ _ _ _ _ _ _
. )
where steel requirements and wall thicknesses are dictated by other considerations aside from ground motion. Here the cost penalty will be quite small. A similar condition, to a lesser degree, applies to the turbine generator substructure.
Accounting for these factors, it might be reasonable to Expect that the average increase in the cost of the reinforced concrete features of the facility would be less than 15%. If about two thirds of the cost of the structures and improvements and one third of the overhead con-struction costs, as given in the Company's Amendment No. 3 are chargeable to. reinforced concrete, it appears that the cost increase i 1
in these items might be in the range of $1,000,000 or about 7% of their combined total. l The percentage increase in engineering costs could be expected to ;
1 be relatively greater due to the design refinements needed to replace the more casual practices usually used. As a completely unsupported estimate, this increase would probably be less than $500,000 (about 18%
of the $2,830,000 allotted to engineering, superintendence, and account-ing in the Company estimate). The combined total of $1,500,000 would be about 10% of the $14,500,000 allotted to structures and improvements
. and overhead construction costs. i
? -
P Again, it is emphasized that these numbers are not to be considered as anything more than an extremely rough estimate of low probable ac-curacy. The only fairly firm conclusion to be drawn is that the cost of the additional seismic protection is very likely to be in the range of 5 to 15% of the cost allotted to structures, improvements, and con-struction overhead.
a m_____-------_---------- - - . - - - - - - _ - - . - - _ - -
)
l I
- 2. Conclusions for Report i
In the course of the meetings in Bethesda on June 27 and 28, it was requested that the writer submit conclusions for incorporatiori in the final report.
= i Since the report is a joint effort of Dr. Newmark and the writer,.
the conclusions should be a joint effort. However, the writer has for-mulated the following tentative conclusions, which are purposely brief l l
and as non-technical as possible, but could be expanded as necessary.
I
- "The following general conclusions' imply the absence of gross differentia 1' foundation movements such as that due to fault slippage g beneath the site.
- a. It is feasible to design all features of the reactor which are essential to the prevention of uncontrolled off-site fission product release, to resist the effects of the maximum credible earthquake of the intensity defined herein.
I
[ b. Adequate design will require approaches utilizing the best earthquake engineering information currently avail-able, including methods which evaluate forces and motions
~
based on a suitable estimate of the properties of the earthquake and those of the structure being considered. l f Attainment of the required degree of seismic resistance I l
will also require careful attention to the details of all .
systems needed for containment of fission products."
- 3. Working Stress Basis vs Yield Stress Basis i
The following considerations relate to the use of conventional work-ing stresses in combination with spectra oflow intensity vs the use of !
l d & m & w 4 M y l _ _ _ . - _ _ - - _ I
)
yield point stresses in combination with spectra of correspondingly high intensity.
The working stress basis is more conservative than the yield stress basis if the component is stressed significantly by other effects.
f in addition to ground motion, because the ground motion effects theri 2 have relatively less influence. As a simple example consider two components, one of which is unstressed except when seismic forces act. The other carries a stress from other causes of 10,000 psi.
With an allowable working stress of 20,000 psi under seismic condi-tions, the components would be designed as follows:
( No. I _No.2 Non-seismic stress 0 10,000 Seismic stress 20,000 10,000 Total 20,000 20,000 If the yield stress in each component is 40,000 psi, a doubling of the 1
l ground motion spectral values would give the following stresses:
l l
No. 1 No.2 1
. Non-seismic stress 0 10,000 l Seismic stress 40,000 20,000 Total 40,000 30,000 Component No. 2 is less highly stressed than component No.1 and will not be subjected to yield stresses in this particular case until the ground motion spectral values are tripled.
On the other hand, in certain other cases, it is necessary to con-sider conditions at yield in order to be sure of satisfactory performance
1
)
{
l
' i
]
under ground motion input represented by spectra having ordinates .
i double those of the so-called maximum probable earthquake. i One such case involves structural steel components, whe e the ratio of yield stress to working stress is typically more like .1. 6 I
Instead of the value of 2 typical for reinforcing steel. Pr esumably, a similar condition might apply to piping and equipment components;-
however, in these instances the use of higher spectral values mentioned in Item 4 might overcome this problem.
A second case involves factors of safety against overturning. For 1
example, c'onsider a case wherein the seismic overturning moment )
1 is 100,000 ft. kips under working stress conditions, and twice this' '
L value under yield conditions, with an available righting moment (due l
to dead load for instance) of 50,000 ft. kips. It is further assumed i that the customary procedure of providing an overturning resistance of 1.5 times the net overturning moment is adhered to. Then parallel calculations for a design based on working. stress conditions would lead to the following results:
j At Working At Yield I r Stress Level Stress Level r j.
l l Seismic overturning moment 100,000 ft k 200,000 ft k I Righting moment 70,000 70,000 Net overturning moment 30,000 130,000 Resisting moment provided 45,000 90,000 Reserve overturning moment 15,000 - 40,000 In this case the resisting moment provided is' insufficient (structure is unstable) if the seismic overturning moment is doubled.
1 4
e [,
s
)
e A related case applies to spread footings bearing directly on soils.
It is not uncommon to find that under the working stress conditions the ,
influence of the compressive stress in the soil due to verticalloads l ey.ceeds the overturning effect associated with seismic forces, leaving a net compression under the entire footing. However, if a doubling, of the seismic forces causes uplifting over a portion of the base, with the soil being incapable of transmitting tension, the soil stresses can be more than doubled.
Finally, there is the problem of seismically induced waves in fluid container s ;("aloshing"). Here, doubling the seismic input will more l- than double the computed wave height in the fundamental " sloshing"
{ mode. If it is important to prevent spillage of the contents under these conditions, as it might be where the fluid is radioactive, then freeboard 1 i
l j requirements should be based on the doubled seismic input.
l 1
7 , The numbers cited in the examples are not necessarily typical but '
I are chosen merely to illustrate the points invoived. However, it seems 4 1
l apparent, in such instances as those cited and certainly in others, that I the investigation of conditions at yielding may be necessary to eliminate weak links in the system and provide a more nearly uniform safety 1
factor, regardless of whether yielding is supposed to occur at double l
the seismic input, or at some other value. Hence, it is urged that the l AEC take the position that maximum credible earthquake should not cause non-linear behavior in the usual structural sense, nor unacceptable I amounts of sloshing, regardless of whether the Company chooses to design at a working stress level or on some other basis. In the writer's opinion, the advantages of designing on a yield point basir outweigh the disadvantages, in this installation except, possibly, in the case of equipment items.
- 4. Spectra '
From the results of the Argonne meeting it seems possible that two 6-
?
M E E__.____..____.__.___.____.._._____ - _ _ _ _ ~ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ _ . J
)
l 2 spectra will be required in the design; one to be used for buildings and t 1 building components, and the other to be used for. equipment. There l are distinct practical disadvantages to this approach: .
- a. There is a hiigh probability of confusion and misinterpretation.
This has already been demonstrated by the variance between . 3 the criteria presented in the PHSR and that which was -
{ intended by Dr. Housner.
1
- b. Further confusion arises in attempting to define the inter-
. faces between equipment and basic structure. l L
- c. The use of two spectra will certainly not ea'se the difficulties .
to be anticipated in the public hearing.
It is strongly suggested that a single spectra be used and that re-duced stresses be applied in the case of those components which are
- considered to require a higher margin of safety. For example, if l spectra corresponding 'to 0.5g are considered appropriate for certain structures, whereas others would require 0. 67g spectra, the allow-able stresses for the latter could be reduced in the ratio 0.50/0.67 .
or 0.75. This would accomplish a result at least as conservative as I that obtained by using two spectra.
l As a result of the Argonne meeting it was agreed that a design equivalent to using 50%g spectra with yield point stresses would be l
considered to be appropriate for the basic structure, this value being l considerably lower than that previously considered to be appropriate.
Dr. Housner's analysis entitled " Acceleration Adjacent to a Fault Produced by the Slip Displacement" appears to substantiate the 50%g figure fairly convincingly. The primary factors in this analysis which might change the picture appear to be variation in the values of gamma i
l
..>,>. . ~
5 and c. If upper bound values larger than those used are a real pos-sibility, then such increases could cause. ground velocities gre.ater than the 2 feet per second given in the analysis, with a corres'p.onding-ly greater damage potential. The writer cannot comment as to the
.. likelihood of this possibility. __
- 5. Revised Criteria Assuming that revised criteria will be submitted by the Company, such criteria should preferably include information relating to items .
such as those discussed in the following paragraphs.
- a. The Company appears to regard the stack as a Class 1 7
structure. If this is considered to be a valid assumption, 4
i ,
it would be well to ask the Company to specify criteria to be used in design of the stack, particularly with regard to such items as the extent to which higher mode responses are to be considered, construction material, whether the stack is to be on ground or carried on a supporting frame, types of lining being contemplate'd, and percentages of ,
j critical damping to be used.
It is not clear whether there will be any elevated tanks in
- f. the Class I category, but if this turns out to be the case, t then certain simple but important precautions are needed .
in design to maximize the available seismic resistance to the fullest possible extent. In the case of an elevated tank of the conventional rod-braced type, it is important to be certain that yielding of the rods in critical panels can occur without yielding or collapse of other features. This situation is not necessarily achieved in earthquake resistant tanks designed on a conventional working stress basis and Fenerally u., - ,
I~ _. . . . _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _
. 1 requires considering conditions at impending failure of the weakest element. The effect on the riser of the large j lateral displacements associated with yielding of the' rods !
in this type of tank.also needs consideration.
- b. The Company should state the criteria to be used fo'f con-sidering the vertical components of ground motion. While i the vertical accelerations are less than the horizontal ac-celebrations, they should not be overlooked.
- c. The Company should indicate the type of analysis (computer l or manual) to be used for piping systems in thermal stress analysis and also for seismic analysis. The amount of
- damping to be assumed should also be stated. l
- d. It is the writer's understanding that the spectra for periods below about 0.2 see or less will be based on an extrapolation.
Since many of the reactor structures and components have periods in this' range and since the spectrum ordinates change
, rapidly with period in this range, this extrapolation may be
- subject to considerable error. It is suggested that the
}
- Company investigate the feasibility of exter. ding these curves t
) ,
on the basis of additional numerical data auch as that obtain-able from the response spectrum analyzer at Caltech.
l e. It would be desirable to request information from the Company relative to the criteria they propose for evaluating seismic effects due to surges in fluid containers, including the amount l of damping to be considered in connection with the fluid os cillations.
1 i
4
_9 9 m .un.ie e. 49 .
L,_,________________-__ _ _ _ _ - - - -
.' c ~; .
q} )
- f. If it is felt that documentation of the seismic resistance of cach critical component should be submitted, then the ,
Company should be appraised of this fact. Perhaps this documentation requirement could be handled as an inter- ,
}
pretation of the statements made on page V-8 of the PHSR, relating to design factors for Class 1 structures.
Ii l
l, l
l I
l 1
i 8!
i
)
3 .-
i 4
l l
l l
l
, July 16,1963
. . _ . - _ . , 4
- - - _ _ . ~ - _ _ _ _ _ _ _
1 t> ~- e n r 4
4*
. . qw . - . .:. -~w r , -w- ~
_.. . , _ ~ . . .m
...+ df;0% M w." j.-. w -ss#g - . . L .-2:wipp6M'%. '
.g,h.-wr.t$h:2.m.
. 9mn$ - 4 '.' * *
. L 9 .. . .
g .
_ .__._ _ .n...e.%. . ~ '....-,....;..... . . . . . . .. . _ . . _ . _ _ . _ ,
oArt or pecumeurs oarc ncesiveD No.: ,i f'>nowz_-
i M
e
~ '
bN . c o, .),oR,, hn$,(
x (& enel)
R. A. hi liwapa .
TOz OksG.: CCs _
OT H Eks .
Eryan x -
ACTION NECES$ARY Q CONCURRENCE O DATE ANSWERED:
e Y, NO ACTION NECESSARY Q COMMENT O
-~
CLAGSIF s POsi OFFICE FILE CODE:
U 50-205 (swp1 only)
REG. Nor DL &CNIP flON: (Must Be Unclassefaed) nEFERRED TO DATE RECElvED BY DAtt Ltr trar.s the fu1 Lowing natarial relab- ~
izy to Ocdega imys g, , g t/ sus. file sy.
ENCLOSUhlBs .
Docunsat, relating te gecations rainsd h sesetine at. batnerda and othen stern frca tre Arscnna su ti ns
[ '
of ? -b3-oost of seismic pvte m en, ..
etc. . .
"'"^"' Lv %(tj l g
M {* '
~
& (t) eA < x _. D ) ~
' Sk,% v.r$"...&,,, 6 dCU -r .
- .....m......,,... . - . =
c.s. ATOMC .:htRGY COMESSION MAIL CONTROL FORM romu uc.sses (s.co)
=,-c.. 2.cc-- ---- ; . x .
~~
7 i 4 7./.W .s
. [c..: w.;c :. g'.
. . _ . . , , . . ~ . . _ . .
c y * ; t..%;.> ,,p::.3 %;;4 ng.s.-9w".p&gvc.;;am . _ m ,- _ . _
(s-_so) ' '-l.
. ; m- v. . - .. .. ; ._ .g.u.~ . ,.gH .. ;...
"2.
g,-_.a.cg . .yg yg.3.,.j.grg.g- =
n .
m ..
~
' . . . .*[ ' ? .M
%Y T' e *'
,R , X*"a.ne.a.ss"$&&WM A ? *****'******'* ^- 4e'y-[***F..
. y =N , ^ * * .4 '~7 i1 s .' . ( ,, ,. j
~.00, ,,. $_ .fh, ',,,,O~. ., '.c',' ' ". ~AR ^'
? Q.h 4 y. ~ . * '
,K- *k .$5 l ....
. .y. , n2 g w ,; ....,..g, w
.y.g-;. .}.gc_% e
~ Q ; . i('; a :_ _ - - .
.qY: *
~
- - .L i ^ y = , .~.y,, -( -h
, ---. y - y ..f yn .- . ,q., .,--z.--;.. m. 4 . , .. _ 7
_ ,. . . . . . :. .~,. , .
..,.. .- - a -. 2";*.* ..
.,.'u. ::,. 3 _ .. - . -;
.n.
. ,. s -
t . .
_ .gp, 'M .-
4 g
.g .
1 1
1
. j I
l !
nh -
.I..; .
OIp ph.8 m_
Y 4
1 I
l J
g
- _.____ .________ _.._ _._ _ _ _ _ _