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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20205L1831986-03-13013 March 1986 Comments on Proposed Rule 10CFR9.Rule Opposed Due to Opinion That Broad New Class of Secret Sessions or Meetings W/O Transcripts Will Be Created ML20151R5611986-01-30030 January 1986 Summarizes 851216 Mgt Meeting at Oak Ridge Operations Ofc to Review & Inspect Redress & Reclamation of Crbr Site.Drawings Reviewed,Discussions Held & Tour of Site Performed.Site Restoration Concluded to Be as Described in Plans & Specs ML20138Q8301985-12-0303 December 1985 Further Response to FOIA Request for Records Re Voluntary or Required Redress of Sites Where Const Was Terminated, Including Crbr & Legal Analysis.Forwards App E Documents.App D & E Documents Available in Pdr.Photographs Also Available ML20128B2801985-06-27027 June 1985 Updates DOE .Bids for Redress of Crbr Site Opened on 850604.Contract Awarded to Beaver Excavating Co,Canton, Oh,On 850613.Contractor Scheduled to Complete Site Redress on or Before 851216 ML20133C7371985-06-14014 June 1985 Further Response to FOIA Request for Documents Re Site Redress Where Plant Const Begun,Including Clinch River Facility & NRC Legal Analysis Re Redress.Forwards App B Documents.App C Document Withheld (Ref FOIA Exemption 5) ML20128R0001985-06-0606 June 1985 Partial Response to FOIA Request for Records Re Desirability of Voluntary or Required Redress of Nuclear Plant Sites Where Work Undertaken But Const No Longer Contemplated. Forwards Documents Listed in App a ML20133C6881985-05-0808 May 1985 FOIA Request for Documents Re Redress of Sites Where Nuclear Plant Const Begun & NRC Legal Analysis of Need for Site Redress W/ or W/O Current CP & LWA ML20107M8141984-11-0707 November 1984 Clarifies & Reaffirms Util Commitments Re Redress of Site in Accordance W/Util 840305 Final Site Redress Plan & NRC 840606 Approval of Plan ML20107H8031984-11-0606 November 1984 Reaffirms Commitments to Redress Site in Accordance W/ 840305 Final Site Redress Plan & NRC 840606 Approval Ltr. Related Correspondence ML20140C6121984-06-18018 June 1984 Advises That Time for Commission to Review ALAB-761 Expired. Commission Declined Review.Decision Became Final Agency Action on 840611.Served on 840618 ML20087B4361984-03-0808 March 1984 Confirms That ASLB Intends to Vacate Notice of 840314 Conference & Reschedule Later Date Due to Delay in Funding for Review of Redress Plan & Possible Lack of Availability of One ASLB Member.Certificate of Svc Encl ML20086T4631984-03-0505 March 1984 Forwards Site Redress Plan. Minor Clarifications & Corrections to Draft Plan Submitted 840227 Made in Response to NRC Comments ML20087A4141984-03-0202 March 1984 Forwards Page 15 Inadvertently Omitted from 840227 Transmittal of Draft Crbr Plant Redress Plan.Related Correspondence ML20080T9331984-03-0101 March 1984 Forwards Page 15 of Crbr Program Redress Plan,Inadvertently Omitted from 840227 Transmittal.Certificate of Svc Encl ML20128R0161984-02-29029 February 1984 Expresses Thanks for 840222 Review of Site Redress Planning. Concurs W/Conceptual Approach & 1-yr Period for Development of Final Site Redress Plan & Investigation of Potential Use of Site.Ml Lacy Encl ML20080S6661984-02-27027 February 1984 Forwards Draft Site Redress Plan,In Response to N Grace 831208 Request.Plan Will Be Finalized for Submission on 840302,following Receipt of Comments ML20079F9411984-01-13013 January 1984 Informs of Receipt & Storage,Through S&W Engineering,Of Spent Fuel Transfer Port Assembly Large Shield Plug.Due to Failure of Congress to Appropriate Addl Funding,Doe No Longer Seeking CP & Is Closing All Licensing Activities ML20083G3551984-01-10010 January 1984 Advises That Svc of DOE & Project Mgt Corp 831227 Notification Re Project Termination Affected Again on All Parties on Attached Svc List ML20083H2331983-11-15015 November 1983 Summarizes Current Status of SER Open Items Re Structural Response During Faulted Conditions & Beyond Dbas.Program Lacks Planned Analytical Support.Models to Support Experimental Efforts Should Be Developed ML20081B9721983-10-24024 October 1983 Summarizes 831004 Meeting W/Nrc,Acrs & Lnr Assoc Re Mgt of Crbr PRA Program.Viewgraphs & List of Meeting Attendees Encl ML20078A7571983-09-0707 September 1983 Forwards Evaluation Repts of Faults 1,2 & 3 Discovered on Site During Foundation Excavation.Faults Not Capable within Meaning of App a to 10CFR100.W/seven Photographs ML20076A8171983-08-17017 August 1983 Confirms Redirection for Shipping Applicant Voluminous Exhibits.Certificate of Svc Encl ML20076A7761983-08-17017 August 1983 Advises That Author Will Present Oral Argument on Behalf of Applicants & Forwards Motion Requesting Argument Be Rescheduled for 830928 ML20077J0781983-08-11011 August 1983 Forwards Pages from Transcript of Crbr CP 830810 Hearings Per ASLB Direction.Certificate of Svc Encl ML20081A5601983-08-11011 August 1983 Summarizes 830808 Informal Meeting on Contract Re PRA Review of Crbr (Task 4) & NRC Concerns Associated W/Technology for Energy Corp Deliverables Schedule ML20076H8811983-08-0909 August 1983 Expresses Appreciation for NRC Presentation on 10CFR21 & 10CFR50.55(e) Requirements.Info Should Be Most Useful to Personnel Involved in Project in Following Requirements ML20024E0391983-08-0505 August 1983 Forwards Errata Sheets for Applicant Prepared Testimony. Certificate of Svc Encl.Related Correspondence ML20077D1321983-07-25025 July 1983 Advises of Omission in Applicant 830722 Response Re CP Evidentiary Hearings.Hearings Did Not Commence on 830718 But Were Postponed Per 830713 Order.Order of 830719 Rescheduled Hearings for 830808-12.Certificate of Svc Encl ML20024D0291983-07-22022 July 1983 Forwards Clinch River Breeder Reactor Plant Sys Design Description - Nuclear Island HVAC Sys, as Example of Procedure Outlines Available for Performance of PRA ML20080A8311983-07-20020 July 1983 Opposes Plant Const Since Little Prior Experience Exists W/Breeder Reactor Design ML20024D5121983-07-19019 July 1983 Requests Specs for Electrical Power Cable Insulation to Be Used at Facility ML20077A5331983-07-19019 July 1983 Advises of Incorrectly Cited Ref on Page 1 of Attachment B & on Page 2 of Attachment C to .Certificate of Svc Encl ML20077H1911983-07-19019 July 1983 Responds to NRC Re Violations Noted in IE Insp Rept 50-537/83-05.Corrective actions:Westinghouse-Oak Ridge Audit Program Revised to Be Computerized Sys.Implementation Throughout Yr Will Be Measured by Planned Surveillances ML20072P1101983-07-15015 July 1983 Forwards Applicant Proposed Exhibit List for CP Hearings,For Review.Stipulation as to Authenticity & Admissibility Requested.Certificate of Svc Encl ML20024C1621983-07-0808 July 1983 Informs That Auxiliary Feedwater Sys Evaluation,Per PSAR App C,Section C.6.4 & App H,Section II.E.1.1,scheduled for Completion by mid-1985 ML20085A7291983-07-0606 July 1983 Forwards Rev 6 to Vol 2 to CRBRP-3, Assessment of Thermal Margin Beyond Design Base (Tmbdb) ML20105B9551983-07-0606 July 1983 Forwards Addl Info Per Request at 830610 Meeting on Programmatic Objectives Re Fuel Burnup.Fftf Operates W/Peak Burnup of Over 61,000 Megawatt Day/Mt.Burnup Occurred Under Temp & Power Conditions Similar to Crbr Conditions ML20079R7401983-06-23023 June 1983 Summarizes 830606 Meeting W/Crbr Project Personnel Re Schedule for Resolution of Confirmatory Items.All Identified Items & Preliminary Schedule Info Discussed.List of Attendees Encl ML20079R2661983-06-21021 June 1983 Lists Typographical & Transcription Errors in 830512 Deposition.Certificate of Svc Encl ML20024A6781983-06-20020 June 1983 Informs of Planned Optional Use of Mechanical Couplers for Reinforcing Bar Splice Sys in Nuclear Island Mat.Qa Program Will Be Established.Exception to ASME Code,Section III & Reg Guide 1.136 Requirements Encl ML20076J0951983-06-17017 June 1983 Summarizes Programmatic Objectives 830610 Meeting Re Fallbacks Identified in Chapter 4 of SER & Impact on Crbr Project.Viewgraphs & Supporting Documentation Encl ML20076J0511983-06-16016 June 1983 Advises That DOE Addressees Include Tj Garrish,L Silverstrom & Wd Luck.Rt Johnson & WE Bergholz Should Be Deleted from Svc List.Certificate of Svc Encl ML20023D9611983-05-27027 May 1983 Submits Agreements Reached at 830524 Meeting W/Crbr Project Re Pra.Description of Addl Tasks Needed to Integrate Plan I & II Efforts Encl ML20072B3431983-05-27027 May 1983 Forwards Crbr Erosion & Sediment Control Plan Rept, Providing Implementation Status of Control Plan Measures Currently Utilized ML20023D4031983-05-20020 May 1983 Forwards Amend 77 to PSAR ML20076D3151983-05-19019 May 1983 Forwards Rev 1 to Crbr Project Heat Transport Sys In-Containment Piping Reserve Seismic Margins & Rept Re Consequences of Leaks from Small Diameter Primary Heat Transport Sys Piping ML20076D2281983-05-17017 May 1983 Forwards Rev 5 to Vol 2 to Thermal Margin Beyond Design Base. Rev Incorporates Isotopic Inventory for Heterogeneous Core,Current Meteorology,Addl Organ Doses & More Realistic Pu Sparging Calculations ML20023C5821983-05-16016 May 1983 Submits Supplemental Info to 830401 Ltr Re Cable Separation by Confirming That Approx 75 Ft of DHR Svc & Steam Generator Auxiliary Heat Removal Sys Cable Will Be Run in Separate Conduits or Encl Raceways ML20079Q2881983-05-10010 May 1983 Forwards Corrected 830509 Response to NRDC & Sierra Club First Set of Interrogatories & Request to Produce Directed to Applicant.Original Document Not Identified as Response ML20024D9551983-05-0909 May 1983 Submits Estimate of LMFBR Safety & Licensing Review Needs Over Next Several Yrs.Preparation of Portions of FSAR Will Begin in 1984 1986-03-13
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20205L1831986-03-13013 March 1986 Comments on Proposed Rule 10CFR9.Rule Opposed Due to Opinion That Broad New Class of Secret Sessions or Meetings W/O Transcripts Will Be Created ML20128B2801985-06-27027 June 1985 Updates DOE .Bids for Redress of Crbr Site Opened on 850604.Contract Awarded to Beaver Excavating Co,Canton, Oh,On 850613.Contractor Scheduled to Complete Site Redress on or Before 851216 ML20133C6881985-05-0808 May 1985 FOIA Request for Documents Re Redress of Sites Where Nuclear Plant Const Begun & NRC Legal Analysis of Need for Site Redress W/ or W/O Current CP & LWA ML20107M8141984-11-0707 November 1984 Clarifies & Reaffirms Util Commitments Re Redress of Site in Accordance W/Util 840305 Final Site Redress Plan & NRC 840606 Approval of Plan ML20107H8031984-11-0606 November 1984 Reaffirms Commitments to Redress Site in Accordance W/ 840305 Final Site Redress Plan & NRC 840606 Approval Ltr. Related Correspondence ML20087B4361984-03-0808 March 1984 Confirms That ASLB Intends to Vacate Notice of 840314 Conference & Reschedule Later Date Due to Delay in Funding for Review of Redress Plan & Possible Lack of Availability of One ASLB Member.Certificate of Svc Encl ML20086T4631984-03-0505 March 1984 Forwards Site Redress Plan. Minor Clarifications & Corrections to Draft Plan Submitted 840227 Made in Response to NRC Comments ML20087A4141984-03-0202 March 1984 Forwards Page 15 Inadvertently Omitted from 840227 Transmittal of Draft Crbr Plant Redress Plan.Related Correspondence ML20080T9331984-03-0101 March 1984 Forwards Page 15 of Crbr Program Redress Plan,Inadvertently Omitted from 840227 Transmittal.Certificate of Svc Encl ML20080S6661984-02-27027 February 1984 Forwards Draft Site Redress Plan,In Response to N Grace 831208 Request.Plan Will Be Finalized for Submission on 840302,following Receipt of Comments ML20079F9411984-01-13013 January 1984 Informs of Receipt & Storage,Through S&W Engineering,Of Spent Fuel Transfer Port Assembly Large Shield Plug.Due to Failure of Congress to Appropriate Addl Funding,Doe No Longer Seeking CP & Is Closing All Licensing Activities ML20083G3551984-01-10010 January 1984 Advises That Svc of DOE & Project Mgt Corp 831227 Notification Re Project Termination Affected Again on All Parties on Attached Svc List ML20083H2331983-11-15015 November 1983 Summarizes Current Status of SER Open Items Re Structural Response During Faulted Conditions & Beyond Dbas.Program Lacks Planned Analytical Support.Models to Support Experimental Efforts Should Be Developed ML20081B9721983-10-24024 October 1983 Summarizes 831004 Meeting W/Nrc,Acrs & Lnr Assoc Re Mgt of Crbr PRA Program.Viewgraphs & List of Meeting Attendees Encl ML20078A7571983-09-0707 September 1983 Forwards Evaluation Repts of Faults 1,2 & 3 Discovered on Site During Foundation Excavation.Faults Not Capable within Meaning of App a to 10CFR100.W/seven Photographs ML20076A8171983-08-17017 August 1983 Confirms Redirection for Shipping Applicant Voluminous Exhibits.Certificate of Svc Encl ML20076A7761983-08-17017 August 1983 Advises That Author Will Present Oral Argument on Behalf of Applicants & Forwards Motion Requesting Argument Be Rescheduled for 830928 ML20081A5601983-08-11011 August 1983 Summarizes 830808 Informal Meeting on Contract Re PRA Review of Crbr (Task 4) & NRC Concerns Associated W/Technology for Energy Corp Deliverables Schedule ML20077J0781983-08-11011 August 1983 Forwards Pages from Transcript of Crbr CP 830810 Hearings Per ASLB Direction.Certificate of Svc Encl ML20076H8811983-08-0909 August 1983 Expresses Appreciation for NRC Presentation on 10CFR21 & 10CFR50.55(e) Requirements.Info Should Be Most Useful to Personnel Involved in Project in Following Requirements ML20024E0391983-08-0505 August 1983 Forwards Errata Sheets for Applicant Prepared Testimony. Certificate of Svc Encl.Related Correspondence ML20077D1321983-07-25025 July 1983 Advises of Omission in Applicant 830722 Response Re CP Evidentiary Hearings.Hearings Did Not Commence on 830718 But Were Postponed Per 830713 Order.Order of 830719 Rescheduled Hearings for 830808-12.Certificate of Svc Encl ML20024D0291983-07-22022 July 1983 Forwards Clinch River Breeder Reactor Plant Sys Design Description - Nuclear Island HVAC Sys, as Example of Procedure Outlines Available for Performance of PRA ML20080A8311983-07-20020 July 1983 Opposes Plant Const Since Little Prior Experience Exists W/Breeder Reactor Design ML20024D5121983-07-19019 July 1983 Requests Specs for Electrical Power Cable Insulation to Be Used at Facility ML20077H1911983-07-19019 July 1983 Responds to NRC Re Violations Noted in IE Insp Rept 50-537/83-05.Corrective actions:Westinghouse-Oak Ridge Audit Program Revised to Be Computerized Sys.Implementation Throughout Yr Will Be Measured by Planned Surveillances ML20077A5331983-07-19019 July 1983 Advises of Incorrectly Cited Ref on Page 1 of Attachment B & on Page 2 of Attachment C to .Certificate of Svc Encl ML20072P1101983-07-15015 July 1983 Forwards Applicant Proposed Exhibit List for CP Hearings,For Review.Stipulation as to Authenticity & Admissibility Requested.Certificate of Svc Encl ML20024C1621983-07-0808 July 1983 Informs That Auxiliary Feedwater Sys Evaluation,Per PSAR App C,Section C.6.4 & App H,Section II.E.1.1,scheduled for Completion by mid-1985 ML20085A7291983-07-0606 July 1983 Forwards Rev 6 to Vol 2 to CRBRP-3, Assessment of Thermal Margin Beyond Design Base (Tmbdb) ML20105B9551983-07-0606 July 1983 Forwards Addl Info Per Request at 830610 Meeting on Programmatic Objectives Re Fuel Burnup.Fftf Operates W/Peak Burnup of Over 61,000 Megawatt Day/Mt.Burnup Occurred Under Temp & Power Conditions Similar to Crbr Conditions ML20079R7401983-06-23023 June 1983 Summarizes 830606 Meeting W/Crbr Project Personnel Re Schedule for Resolution of Confirmatory Items.All Identified Items & Preliminary Schedule Info Discussed.List of Attendees Encl ML20079R2661983-06-21021 June 1983 Lists Typographical & Transcription Errors in 830512 Deposition.Certificate of Svc Encl ML20024A6781983-06-20020 June 1983 Informs of Planned Optional Use of Mechanical Couplers for Reinforcing Bar Splice Sys in Nuclear Island Mat.Qa Program Will Be Established.Exception to ASME Code,Section III & Reg Guide 1.136 Requirements Encl ML20076J0951983-06-17017 June 1983 Summarizes Programmatic Objectives 830610 Meeting Re Fallbacks Identified in Chapter 4 of SER & Impact on Crbr Project.Viewgraphs & Supporting Documentation Encl ML20076J0511983-06-16016 June 1983 Advises That DOE Addressees Include Tj Garrish,L Silverstrom & Wd Luck.Rt Johnson & WE Bergholz Should Be Deleted from Svc List.Certificate of Svc Encl ML20023D9611983-05-27027 May 1983 Submits Agreements Reached at 830524 Meeting W/Crbr Project Re Pra.Description of Addl Tasks Needed to Integrate Plan I & II Efforts Encl ML20072B3431983-05-27027 May 1983 Forwards Crbr Erosion & Sediment Control Plan Rept, Providing Implementation Status of Control Plan Measures Currently Utilized ML20023D4031983-05-20020 May 1983 Forwards Amend 77 to PSAR ML20076D3151983-05-19019 May 1983 Forwards Rev 1 to Crbr Project Heat Transport Sys In-Containment Piping Reserve Seismic Margins & Rept Re Consequences of Leaks from Small Diameter Primary Heat Transport Sys Piping ML20076D2281983-05-17017 May 1983 Forwards Rev 5 to Vol 2 to Thermal Margin Beyond Design Base. Rev Incorporates Isotopic Inventory for Heterogeneous Core,Current Meteorology,Addl Organ Doses & More Realistic Pu Sparging Calculations ML20023C5821983-05-16016 May 1983 Submits Supplemental Info to 830401 Ltr Re Cable Separation by Confirming That Approx 75 Ft of DHR Svc & Steam Generator Auxiliary Heat Removal Sys Cable Will Be Run in Separate Conduits or Encl Raceways ML20079Q2881983-05-10010 May 1983 Forwards Corrected 830509 Response to NRDC & Sierra Club First Set of Interrogatories & Request to Produce Directed to Applicant.Original Document Not Identified as Response ML20024D9551983-05-0909 May 1983 Submits Estimate of LMFBR Safety & Licensing Review Needs Over Next Several Yrs.Preparation of Portions of FSAR Will Begin in 1984 ML20073S2761983-05-0505 May 1983 Forwards Revised Responses to SER Item 6 Re Qa,Including Info to Complete Identification of safety-related Structures,Sys & Components Controlled by Crbr QA Program for PSAR.Marked-up Tech Specs Encl ML20073Q3491983-04-28028 April 1983 Forwards Revised Response to SER Item 6, Qa. Response Provides Addl Info Re Identification of safety-related Structures,Sys & Components Controlled by QA Program ML20073R2221983-04-27027 April 1983 Requests to Make Limited Appearance Statement at 830718 CP Hearings Re Regional Socioeconomic Impacts ML20069L1801983-04-27027 April 1983 Informs That Evaluation of Seismic Adequacy of Primary Heat Transport Sys Branch Line & Consequences of Line Failure Being Conducted,In Response to ACRS 830419 Request.Results Will Be Forwarded by 830517 ML20073K3441983-04-18018 April 1983 Forwards Static Tests of 1/20-Scale Models of Crbr Head in Support on LMFBR Safety Program ML20071G2761983-04-18018 April 1983 Recommends That Commission Retain Technical Cadre of Experts to Review Crbr & Overall DOE Breeder Program 1986-03-13
[Table view] Category:OTHER U.S. GOVERNMENT AGENCY/DEPARTMENT TO NRC
MONTHYEARML20128B2801985-06-27027 June 1985 Updates DOE .Bids for Redress of Crbr Site Opened on 850604.Contract Awarded to Beaver Excavating Co,Canton, Oh,On 850613.Contractor Scheduled to Complete Site Redress on or Before 851216 ML20107H8031984-11-0606 November 1984 Reaffirms Commitments to Redress Site in Accordance W/ 840305 Final Site Redress Plan & NRC 840606 Approval Ltr. Related Correspondence ML20086T4631984-03-0505 March 1984 Forwards Site Redress Plan. Minor Clarifications & Corrections to Draft Plan Submitted 840227 Made in Response to NRC Comments ML20087A4141984-03-0202 March 1984 Forwards Page 15 Inadvertently Omitted from 840227 Transmittal of Draft Crbr Plant Redress Plan.Related Correspondence ML20080S6661984-02-27027 February 1984 Forwards Draft Site Redress Plan,In Response to N Grace 831208 Request.Plan Will Be Finalized for Submission on 840302,following Receipt of Comments ML20079F9411984-01-13013 January 1984 Informs of Receipt & Storage,Through S&W Engineering,Of Spent Fuel Transfer Port Assembly Large Shield Plug.Due to Failure of Congress to Appropriate Addl Funding,Doe No Longer Seeking CP & Is Closing All Licensing Activities ML20081B9721983-10-24024 October 1983 Summarizes 831004 Meeting W/Nrc,Acrs & Lnr Assoc Re Mgt of Crbr PRA Program.Viewgraphs & List of Meeting Attendees Encl ML20078A7571983-09-0707 September 1983 Forwards Evaluation Repts of Faults 1,2 & 3 Discovered on Site During Foundation Excavation.Faults Not Capable within Meaning of App a to 10CFR100.W/seven Photographs ML20076H8811983-08-0909 August 1983 Expresses Appreciation for NRC Presentation on 10CFR21 & 10CFR50.55(e) Requirements.Info Should Be Most Useful to Personnel Involved in Project in Following Requirements ML20024D0291983-07-22022 July 1983 Forwards Clinch River Breeder Reactor Plant Sys Design Description - Nuclear Island HVAC Sys, as Example of Procedure Outlines Available for Performance of PRA ML20077H1911983-07-19019 July 1983 Responds to NRC Re Violations Noted in IE Insp Rept 50-537/83-05.Corrective actions:Westinghouse-Oak Ridge Audit Program Revised to Be Computerized Sys.Implementation Throughout Yr Will Be Measured by Planned Surveillances ML20024C1621983-07-0808 July 1983 Informs That Auxiliary Feedwater Sys Evaluation,Per PSAR App C,Section C.6.4 & App H,Section II.E.1.1,scheduled for Completion by mid-1985 ML20105B9551983-07-0606 July 1983 Forwards Addl Info Per Request at 830610 Meeting on Programmatic Objectives Re Fuel Burnup.Fftf Operates W/Peak Burnup of Over 61,000 Megawatt Day/Mt.Burnup Occurred Under Temp & Power Conditions Similar to Crbr Conditions ML20085A7291983-07-0606 July 1983 Forwards Rev 6 to Vol 2 to CRBRP-3, Assessment of Thermal Margin Beyond Design Base (Tmbdb) ML20079R7401983-06-23023 June 1983 Summarizes 830606 Meeting W/Crbr Project Personnel Re Schedule for Resolution of Confirmatory Items.All Identified Items & Preliminary Schedule Info Discussed.List of Attendees Encl ML20024A6781983-06-20020 June 1983 Informs of Planned Optional Use of Mechanical Couplers for Reinforcing Bar Splice Sys in Nuclear Island Mat.Qa Program Will Be Established.Exception to ASME Code,Section III & Reg Guide 1.136 Requirements Encl ML20076J0951983-06-17017 June 1983 Summarizes Programmatic Objectives 830610 Meeting Re Fallbacks Identified in Chapter 4 of SER & Impact on Crbr Project.Viewgraphs & Supporting Documentation Encl ML20076J0511983-06-16016 June 1983 Advises That DOE Addressees Include Tj Garrish,L Silverstrom & Wd Luck.Rt Johnson & WE Bergholz Should Be Deleted from Svc List.Certificate of Svc Encl ML20023D9611983-05-27027 May 1983 Submits Agreements Reached at 830524 Meeting W/Crbr Project Re Pra.Description of Addl Tasks Needed to Integrate Plan I & II Efforts Encl ML20072B3431983-05-27027 May 1983 Forwards Crbr Erosion & Sediment Control Plan Rept, Providing Implementation Status of Control Plan Measures Currently Utilized ML20023D4031983-05-20020 May 1983 Forwards Amend 77 to PSAR ML20076D3151983-05-19019 May 1983 Forwards Rev 1 to Crbr Project Heat Transport Sys In-Containment Piping Reserve Seismic Margins & Rept Re Consequences of Leaks from Small Diameter Primary Heat Transport Sys Piping ML20076D2281983-05-17017 May 1983 Forwards Rev 5 to Vol 2 to Thermal Margin Beyond Design Base. Rev Incorporates Isotopic Inventory for Heterogeneous Core,Current Meteorology,Addl Organ Doses & More Realistic Pu Sparging Calculations ML20023C5821983-05-16016 May 1983 Submits Supplemental Info to 830401 Ltr Re Cable Separation by Confirming That Approx 75 Ft of DHR Svc & Steam Generator Auxiliary Heat Removal Sys Cable Will Be Run in Separate Conduits or Encl Raceways ML20024D9551983-05-0909 May 1983 Submits Estimate of LMFBR Safety & Licensing Review Needs Over Next Several Yrs.Preparation of Portions of FSAR Will Begin in 1984 ML20073S2761983-05-0505 May 1983 Forwards Revised Responses to SER Item 6 Re Qa,Including Info to Complete Identification of safety-related Structures,Sys & Components Controlled by Crbr QA Program for PSAR.Marked-up Tech Specs Encl ML20073Q3491983-04-28028 April 1983 Forwards Revised Response to SER Item 6, Qa. Response Provides Addl Info Re Identification of safety-related Structures,Sys & Components Controlled by QA Program ML20069L1801983-04-27027 April 1983 Informs That Evaluation of Seismic Adequacy of Primary Heat Transport Sys Branch Line & Consequences of Line Failure Being Conducted,In Response to ACRS 830419 Request.Results Will Be Forwarded by 830517 ML20073K3441983-04-18018 April 1983 Forwards Static Tests of 1/20-Scale Models of Crbr Head in Support on LMFBR Safety Program ML20073D7311983-04-12012 April 1983 Forwards Addl Info to Resolve SER Open Item 4 Re Solid State Programmable Logic Sys ML20073A1201983-04-0808 April 1983 Forwards Responses to Questions Re Identification of safety- Related Structures,Sys & Components by Crbr QA Program.Info Should Enable NRC to Complete Review of QA List & Close Out SER Open Item 6 ML20073A4591983-04-0808 April 1983 Provides Addl Clarification of Info Contained in PSAR Section 13.3, Emergency Planning (SER Open Item 5). Responses to Questions Identified in Crbr 830321 Meeting in Bethesda,Md & Annotated PSAR Pages Encl ML20073A0941983-04-0808 April 1983 Forwards Response to SER Open Item 3 Re Plant Protection Sys Monitor Mods ML20073A8031983-04-0707 April 1983 Commits to Listed Actions to Close Out SER Open Item 1, Review of Rdt Stds F9-4T & F9-5T. Appropriate Info on Verification of Analysis Methods & Computer Programs Used in High Temp Design Will Be Provided at OL Review Stage ML20069H5331983-04-0404 April 1983 Forwards Analysis of Wall Liner at Square Penetrations for Crbr Cell Liner Design Verification Program,Per Commitment in DOE ML20072P9561983-04-0101 April 1983 Commits to Provide Min Physical Separation of 5 Ft Between DHR Sys & Steam Generator Auxiliary Heat Removal Sys Cabling in Same Div.Commitment Resolves SER Open Item 2 ML20069G0761983-03-23023 March 1983 Forwards New Info Re Cell Liner Design Validation Program. Analysis of Wall Liner at Circular Penetrations Encl ML20071F1931983-03-14014 March 1983 Forwards Adequacy of Crbr Program Reactor Vessel NDE Insps. Rept Supports Conclusion That Exam Performed on Reactor Vessel Adequate.Rept Also Applicable to ex-vessel Storage Tank & Heat Transport Sys Vessels ML20071D2591983-03-0808 March 1983 Comments on Design Change to fission-driven Compaction. Compaction of Core Could Lead to Energetics During Hypothetical Core Disruptive Accident.Design Change Feasible ML20071D2511983-03-0707 March 1983 Forwards Corrected PSAR Table 5.6-12 Covering DHR Operating Cases & Sensitivity Evaluations to Be Included in Future PSAR Amend ML20071C7171983-03-0404 March 1983 Addresses ACRS Questions Re Capability of Facility to Continue Shutdown Heat Removal on Natural Circulation W/Loss of Bulk Ac Power for Greater than 2 H.Info Previously Presented to NRC & Documented in PSAR Section 5.7.5 ML20072B6181983-03-0404 March 1983 Forwards Crbr Program Heat Transport Sys Incontainment Piping Reserve Seismic Margins. Rept Addresses Questions Raised During ACRS 830211 Committee Meeting ML20071C6891983-03-0404 March 1983 Forwards Description of Cell Liner Design Validation Program & Fallback Plan If Base Criteria Not Validated,Per NRC Request ML20071C7021983-03-0404 March 1983 Forwards Response to Questions Re Thermal Margin Beyond Design Base,Design & Analysis.Overly Conservative Assumptions Made by NRC Consultants Re Treatment of Sodium Aerosols Inappropriate ML20072B5191983-03-0202 March 1983 Advises That Reliability Assurance Program Will Be Modified to Comply W/Criteria Presented in NRC 830225 Submittal. Project Will Provide Reliability Assurance Program Plan & Schedule Approx 6 Months After Issuance of CP ML20071C2201983-02-28028 February 1983 Forwards Revised PSAR Pages Re Inservice Insp & Addl Info Re NDE Procedure,Per 830214 Discussions.Pages Clarify Design & Provide Info Relative to Piping Insulation & NDE Requirements for Cell Liners & Fire Suppression Decks ML20071C1331983-02-28028 February 1983 Forwards Updated Page to PSAR Chapter 4.4,design Bases Section.Info Concerns Primary Sodium Gas Entrainment & Assembly Flow Blockage Criteria.Page Will Be Submitted W/ Next PSAR Amend ML20071C1391983-02-25025 February 1983 Forwards Addl Info on Psar,Section 13.3 Re Coordination W/State of Nc & Local Officials for Emergency Planning.Info Will Be Incorporated in Next PSAR Amend ML20071B0621983-02-24024 February 1983 Forwards Revised Responses to Mechanical Engineering Branch Items 26 & 70 Re Low Temp Components.Info to Be Included in Next Amend to PSAR ML20071B0301983-02-23023 February 1983 Forwards Amended PSAR Page 10.4-7 Re Consequences of Flooding Due to Failures in Circulating Water Sys.Pages Respond to NRC Questions & Will Be Included in Next PSAR Amend 1985-06-27
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:209 Dr. J. Nelson Grace Director FEB 11 m3 CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ,
Washington, D.C. 20555
Dear Dr. Grace:
ADDITIONAL INFORMATION CONCERNING GAS ENTRAINMENT IN THE PRIMARY HEAT TRANSPORT SYSTEM (PHTS)
Enclosed are amended Preliminary Safety Analysis Report (PSAR) pages to Clinch River Breeder Reactor Plant PSAR Chapters 4 and 5 that provide additional information concerning the potential for gas entrainment in the PHTS. These pages are in response to questions raised by the Nuclear Regulatory Commission staff reviewer and will be included in the next amendment to the PSAR.
Questions regarding the enclosure may be addressed to Mr. D. Hornstra (FTS 626-6110) or Mr. D. Robinson (FTS 626-6098) of the Project Office Oak Ridge staff.
Sincerely, l &*
John R. Longenecker Acting Director. Office of Breeder Demonstration Projects Office of Nuclear Energy Encicsure cc: Service List Standard Distribution Licensing Distribution 9#
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ENCLOSURE the Doppler feedback in CRBRP. Reducing th'e Reference 9 uncertainty for the already conservative neglect of global temperature-importance .
weighting results in an extrapolated Doppler uncertainty of less than ,
110% (la) for CRBRP. !
' Further confirmation of the accuracy cf.LMFBR Doppler constant '
--- predictions is provided by the small-sample measurements in zero power l critical mockups. Small, heated-sample Doppler constant measurements '
have been performed in the. Zero power plutonium Reactor (ZPPR) critical assemblies simulating the CRBRP core configuration. The analysis of the .
Doppler constant experiments in IPPR-2, 3, and 5 are reported in i
l References 10, ll, and 12, respectively. Although these small-sample '
measurements do not represent a direct experimental detennination of the '
- total core Doppler constant .the good agreement between calculated and measured values .does provide substantial confidence that the U-238 ,
resonance parameters in the core spectrum are accurately predicted.
l From the total of 52 small-sample Doppler measurements throughout the core under a variety of reactor conditions (flooded, voided.. control '
rods inserted and withdrawn, etc.), the mean calculation-to-experiment ratio is 0.98 (slightly conservative underprediction of Doppler) with
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an uncertainty of 15.4% (la). . .' ,
- The CRBRP Doppler constant uncertainty, based principally on !
the aforementioned SEFOR evaluation, and supported by the small-sample measurements in ZPPR, is + 10% (la). This value is intended for use in
{ operationel'and duty cycle.
design transient evaluations within the reacto
- and the extf emely unlikely class of events are based on Doppler constants with -2o (80% of nominal Doppler) and -3o (70% of nominal Doppler) uncertainties, respectively.
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4.3.2.3.2 Sodium Void Worth 4 g4g The sodium void worth relates the change in neutron Small, multipli-distribu-cation to the presence of voids in the sodium coolant.
ted voids such as gases entrained in the coolant 3are adequately treated by use of the sodiur, density coefficient (Section 4.3.2.
- l unlikely situation. In the following discussion, the reactivity associated with this latter type of voiding is developed for use in accident analysis in Chapter 15.
Figure 4.3-28 is a flow chartCross showing the are sections method for calculating processed both the sodium voiding reactivity worth.
~with sodium and without sodium to properly account for resonance self-shielding and the change in spectrum 1 when m :.sodium
. ~ is removed.
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The above discussion refers to demonstrating the ability of the rods to meet their design requirements when subjected to the various l
- categories of design transients. Such topics as transient effects due to rod itilures and continued operations '
with fa'iled rods- are presented
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iri S~ection 15.4.
- 4.4.3.7 Potentially Damaging Temperature Effects During Transients i
Ivi general, a single anticipated event is not damaging to the reactor structu'res; it is the total sum of all the occurrences (in this -
event classification) over the particular lifetime that may cause the -
- structure to approach its design limit.- As mentioned in Section 4.4.2.9, the current cladding requirement for fuel pins is that considering all ..*
- normal'and anticipated events, the cumulative cladding damage must not l preclude the capability to survive at least one of the wprst unlikely -
events without loss of cladding integrity. .
Maximum 3c fuel and blanket hot rod temperatures during v'rious a limiting core design events of the plant duty cycle which are described i in Appendix B have been analyzed. Detailed temperatures for various '
axial and radial positions along the rods have been evaluated in deter-mining the core design adequacy as described in Section 4.2.1. Con-servative assumptions used in calculating these temperatures are described in Section 15.1.4. These assumptions include: full power operation, hot rod analyzed in highest power and temperature fuel assembly of all core
- conditions, worst case Doppler coefficient with uncertainties included, a 200 millisecond delay between trip signal and the start of control rod insertion, 3a hot channel factors, the single most reactive control rod assumed to be stuck in the withdrawn position for both sets of control i
rods, highest. core pressure drop, and.the most rapid flow coastdown of the primary pumps following pump trip. In addition. Section 15.1.4 shows
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that of the safety related events, both of the overpower and undercooling l
58} type, the Safe Shutdown Earthquake (SSE) event (60c step reactivity insertion occurring under SSE conditions) results in the highest core temperatures.
The conclusion of the Chapter 15.0 safety evaluations and the Section 4.2.1 design evaluations is that the design changes incurred in going from the homo-geneous to the heterogeneous scheme are not expected to significantly change the design or safety capability of the core.
D@ N.4.3.8 Thermal Description of the Direct Heat Removal Service (DHRS)
The thermal description of the Direct Heat Removal Service will be found in Section 5.6.2. .
55l 4.4.4 Testing and Verification At the present state-of-the-art in reactor design, scale model 51 tests of reactor flow systems provide the most useful tool for studying reactor hydraulics. pressure drop through complicated flow paths, thermal (i C 4.4-75 Amend. 58 Nov. 1980
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An evaluation has also been made to determina if expected levels of gas entrained in the sodium could impair heat transfer in the reactor.
It was found that at the limiting value of one volume
- - - percent established for reactor neutronics effects (section The
- 4.3.2.3.2) there is no significant effect on heat transfer.
expected entrainment. level (Section 4.4.4.1) is much lower than this limiting value.
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biet plenum as a function of the plenum height, Inlet nozzle discharge angle, and the elevation of the nozzles relative to the Core Support Structure. The test served a scoping purpose for use in designing the 1/4-scale IPFM test previously described.
The gross flow patterns were predictable and produced no unusual or unexpected -
results. - ,
.~' Intaa/al'Reac.ter Flow Model. Phase T Testtr.u - ou+Ie+ Planum Featura -
,fjow and Vibration Test -
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This test: 1) measured the velocity pattern in the outlet plenum and in the vicinity of the major outlet plenum structures; 2) determined the pressure -
drop characteristics cf the outlet plenum and major outlet plenum structures;
- 3) determined the mixing characteristics and transport times in the outlet plenum and at locations of probable hot / cold interf aces; 4) evaluated flow induced vibration characteristics of selected outlet plenum structures; and 5) . .
- evaluated the gas entrainment characteristics of the suppressor plate.
&SMWIntearal Reactor Flow Model. Phase Il Testina *
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The purpose of this te~st which is ln progress, is to verify the final design for hy.draulic and vibration performance using the Integral reactor flow model. ,
j Those components used in the Phase I testing, whose design has changed to the extent that hydraulic, and vibration performance is influenced, will be modified in a subsequent Phase 11.
Outlet Plenum Flow Stratification Test ,
. The Outlet Plenum Flow Stratification Test was performed in a model of 0.55 scale simulating a 120 sector of the CRBRP reactor vessel outlet plenum, containing a portion of the UIS, and an outlet main coolant pipe.
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Sane important conclusions and observations derived from the evaluation of the test data (Ref. 60) are: ,
- 1) The transient temperature response at the outlet nozzle is less severe than that predicted by the plant simulation model in the DEM5 code, thus demonstrating that this UlS design goal is satisfied.
- 2) The standard height Uls chimneys resulted in less severe transient -
temperature ranp rates at the outlet nozzle than shorter chimneys.
- 3) The nominal prototypic gep beneath the UlS skirt of 1.0 inch resulted in less severe outlet nozzle transient temperatures than either larger cr smaller gaps.
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A.4-77 Amend. 58
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INSERT FOR FAGE 4.4-77 The gas entrainment tests were performed to determine the amount of cover gas that might be entrained at the free surface inside Using conservative Weber Number the CRDRP reactor vessel.
podeling of the entrainment process, the average reactor vessel outlet void fraction when at normal liquid level and scaled to I 115 percent of CRBRP average full flow is 1.25 a 10-4 The paximum void fraction measured at any reactor vessel outlet nozzle for this same case and extrapolated to 115 percent flow l was 3.0 x 10-4 In addition to determining primary entrainment pheno.sena within the reactor vessel, tests were run to determine the o9uilibrium PHTS void fraction that results for a given gas entratament rate in the reactor vessel. These tests were based on the conservative assumptionThis thatassumption all gas removal occurs results in in highest the the out1ct plenum of the reactor.
equilibrium void fraction in the PHTS, Based on the entrainment rate that gave the void fraction of 1.25 x 10-4 and Weber Number modeling of the entrainment proceta, the equilibrium void fraction Gasin the PHTs removal at 115 by the percent IHX vent and oftherated primaty flow is 2.0 x 10-3 pump standpipe bubbler will reduce this invel.
The test results, as well as a complete description of the subject testing, are presented in Reference 76.
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- 73. F. C. Engel'. R. A. Markley and B. 'Minushkin. " Buoyancy Effects on Sodium Coolant Temperature Profiles Measured in an Electrically . . .
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. Heated Mock-up of a 61-rod Breeder Reactor Blanket Assembly.",, ~'
'ASME 78-WA/HT-25.' .
R. A. Markley and B. Minushkin. " Heat Transfer T'est 74.~ F.. C. Engel .
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' Data of a 61-rod Electrically' Heated LMFBR Blanket Assembly Mor.kup' '
and their use for.Subchannel Code Calibration" in Fluid Flow and * ~ ~
.- . Heat' Transfer Over' Rod or Tube Bundles symposium of ASME 1979 Winter Annua.1 Meeting. pp. 223-229 December 1979. .
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- 75. F. C. Engel. 8 Minubkin R'.' J.' Atkins and R. A. Markley. "Characteriz-
- ation of Heat Transfer and Temperature Distributions in kn Electrically -
Heated Model of an LMFBR Assembly, to be published in a special issue 58 . of Nuclear Engineering Design. -
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Gas entrainment within (11 ft/sec) at design flow frem the reactor vessel.
' the pump is minfalzed by designing and testing, to ensure that alI pump parts >
which need to be submerged during operation at pony motor speed or restart to pony motor speed are located below the alntmum sodium level.
jsed -PDe max! mum oxygen content in the primaryThis
- system level ofsodlum sodiumisimpurity specifled to be 42 -
ppm at 800*F or above and 35 ppm below 800 F. -- ,
will not ,. affect the pump operating characteristics.
The biolog! cal shielding for the PHTS sodium pumps as well as the pump .
assembiles is designed to withstand the loadings associated with the Safe . -
. Shutdown Earthquake (SSE) and the transient overpressures. for ex unlikely plant conditions. -
the PHTS pump as a Class I component in accordance with the rules of the ASME )
Code Section 111 and modifying RDT Standards. . '
I The biological shlalding for the PHTS sodium pumps is p'rovided by (1) an f annular shield tank which surrounds the pump shaft, (2) the pump shaf t itself l' which is designed to provide an Integral part of the shield requirements, (3) the pump support structure which is part of the operating floor, and (4) l special precautions to preclude streaming along the Instrumentati penetrations. flange of the pump pressure boundary containment vessel which accordance with the ASME Code for Class I nuclear components.
supporting assembly and the annular shield structura The design of this joint provides for the operating floor pump motor w' ell. i 1 ds and provides a seal at
' .the dual function of resisting static and dynam c .oa the boundary.between the pump atmosphere and the ,
rings.'
The SSE For the PHTS pump SSE. seismic analysis, a 2$ damping value Following was used.
loadings were considered to occur in conjunction with a plant trip.
the SSE, the Intermediate Heat Transport System, Steam Generator Syste Steam Generator Auxillary Heat RemovalComputer stored and decay heat.
System programs, must provi flow without loss of structural Integrity after the SSE.such Descriptions of these computer programs can be found in primary pumps.
Appendix A. .
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5.3-43 . . . . . <<.
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INSERT FOR PAGE 5.3-43 The pump standpipe bubbler also provides a vent for gas de-entrained fro:n the PETS. The maximum gas flow that the '
. . standpipe can accommodate without flow instability corresponds to an entrained gas volume fractionef 3 x 10-4 This volume fraction has been shown to have no impact on pump performance or heat transfer in the IBX. Outlet plenum tests which give expected gas entrainment levels are discussed in Section 4.4.4.1.
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i 5.3.3.4 Valve- Cha racterisitics The essential characteristics and features for the check valve i
design to satisfy the system operating requirements are included in Table 5.3-13.
The CRBRP Cold Leg Check Valve (CLCV) is hydraulically similar to the FFTF 16" check valve. As noted in Reference 62, extensive tests have been performed on the FFTF valve to confirm its hydraulic performance. In operation at FFTF, no problems due to gas accumulation in the CLCV have been identified. Secondly, tests performed on the FFTF hot leg isolation valve verified the self-venting characteristics (that any accumulated gas will be swept out) of the main valve body of both valves. Since the valve bodiee of both CRBRP and FFTF CLCV's and the FFTF hot leg isolation valve are geometrically equivalent, the results of these tests are applicable to the CRBRP CLCV. Air deliberately introduced into the valve was quickly entrained and removed by the turbulent fluid at pipe flow velocities greater than 7 fps, which correspond to a 2 fps (average) velocity in the valve. The experimenters noted that, "This is the approximate limiting value for air removal from pipelines as established by previous research" (Reference 63). Under full flow conditions the flow velocity in the CLCV inlet pipe is about 25 fps and about 6 fps (average) in the valve body. At 40 percent flow the velocity in the valve body is still above the 2 fps level, and at this reduced flow rate there is little likelihood for any gas entrainment at the principal gae entrainment site (the reactor vessel outlet plenum). Thus, no gas accumulation is expected in the main body of the CLCV.
Although no mechanism has been identified which would allow the de-entrainment of gas in the valve, an evaluation has been performed with the following conservative assumptions. The dome on the top of the check valve is initially ful- gas at the conditions consistent with the thermal hydraulic os_ign condition, the pumps are tripped, and the pressure in the valve has been minimized by assuming that the sodium level in the reactor is at the minimum safe level. The evaluation showed no break of siphon. Therefore, gas accumulation in the check valve would not be a problem for CRBRP.
5.3-43b
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References:
- 54. P. Soo, "Seiection of Coolant-Boundary Materials for the Clinch River -
Breeder Reactor Plant", WARD-0-0010, August 1974. .
- 55. Deleted
- 56. Deleted
- 57. P. Soo (Compiler), " Analysis of Structural Materials for LWBR Coolant-Boundary Components - Materials Property Evaluations", WARD-3045T3-5, November 1972.
- 58. Deleted .
- 59. R. A. Leasure, "Ef feet of Carbon and Nitrogen and Sodium Environment on the Mechanical Properties of Austenitic Stainless Steels",
WARD-NA-94000-5, De'2 ember 1980.
- 60. ES-LPD-82-007, CRBRP Engineering Study Report, "CERP Transition Joints", April 1982.
- 61. ES-LPD-82-008, CERP Engineering Study Report, "CERP Materials Data Base", May 1982.
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i 5.3-75c Amend. 71 Sept. 1982
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