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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217N3901999-10-25025 October 1999 Advises That Info Provided in & Affidavit Re Holtec Position Paper WS-115,rev 1,repts HI-87113, Rev 0,HI-87114,rev 0,HI-87102 Rev 0 & HI-87112,rev 0,marked Proprietary,Will Be Withheld from Public Disclosure ML20217L8591999-10-21021 October 1999 Discusses 990921 Request for Approval to Perform Alternative Testing as Part of Vermont Yankee Nuclear Power Station IST Program.Informs That Submittal Reviewed Against ASME Code Section XI Requirements & Forwards Safety Evaluation ML20217M1181999-10-19019 October 1999 Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Requal Pass Dates ML20217D9711999-10-13013 October 1999 Responds to Request That Information Titled Addl Info Re Cycle Specific SLMCPR for Vermont Yankee Cycle 21 Be Withheld from Public Disclosure.Determined Info to Be Proprietary & Will Be Withheld from Public Disclosure ML20217F1261999-10-12012 October 1999 Forwards Update to Previously Submitted RELAP5 Analytical Assumptions for App R,Re RAI of 961104 BVY-99-130, Provides Clarification of Method for Determining MSIV Maximum & Minimum Pathway at Vermont Yankee Nuclear Power Station1999-10-0808 October 1999 Provides Clarification of Method for Determining MSIV Maximum & Minimum Pathway at Vermont Yankee Nuclear Power Station ML20217C1501999-10-0707 October 1999 Forwards Insp Rept 50-271/99-11 on 990809-27.No Violations Noted.Insp Focused on Effectiveness of Engineering Functions in Providing for Safe Operation of Plant BVY-99-128, Submits Listed Addl Info in Support of 990414 Request for Clarification to SER Confirming Adequacy of Space Cooling for HPCI & RCIC Sys,Re Item II.K.3.24 of NUREG-0737.Copy of NEDE-24955,encl1999-10-0606 October 1999 Submits Listed Addl Info in Support of 990414 Request for Clarification to SER Confirming Adequacy of Space Cooling for HPCI & RCIC Sys,Re Item II.K.3.24 of NUREG-0737.Copy of NEDE-24955,encl ML20212J7891999-10-0404 October 1999 Informs That Licensee 980804,0628,29 & 990921 Responses to GL 98-01, Y2K Readiness of Computer Sys at NPPs Acceptable.Nrc Consider Subj GL to Be Closed for Plant ML20212J6501999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of VYNPS on 990913. No New Areas Identified in Which Licensee Performance Warranted Addl Insp Beyond Core Insp Program.Historical Listing of Plant Issues & Insp Plan Through Mar 2000 Encl ML20216J3531999-09-29029 September 1999 Responds to NRC Re Violations Noted in Insp Rept 50-271/99-12 on 990628-0811.Corrective Actions:Based on RFO 20 Maint Rule Outage Performance Review,Task Was Generated to Clarify & Enhance SD Monitoring Process BVY-99-122, Notifies of Intention to Reinstate Original Version of App F in FSAR & Correct Docket Re Assumption That Electrical Power Sys Are Designed IAW Requirements of GDC-171999-09-28028 September 1999 Notifies of Intention to Reinstate Original Version of App F in FSAR & Correct Docket Re Assumption That Electrical Power Sys Are Designed IAW Requirements of GDC-17 BVY-99-114, Provides Notification That Licensee Completed Y2K Remediation Efforts Described in Util 990608 Response to NRC GL 98-01,Suppl 11999-09-21021 September 1999 Provides Notification That Licensee Completed Y2K Remediation Efforts Described in Util 990608 Response to NRC GL 98-01,Suppl 1 BVY-99-113, Requests Approval to Perform Alternative Testing to That Specified by ASME Boiler & Pressure Vessel Code,Section XI & Asme/Ansi OM, Operation & Maint of Nuclear Power Plants. Attachment 1 Provides Justification for Alternative Testing1999-09-21021 September 1999 Requests Approval to Perform Alternative Testing to That Specified by ASME Boiler & Pressure Vessel Code,Section XI & Asme/Ansi OM, Operation & Maint of Nuclear Power Plants. Attachment 1 Provides Justification for Alternative Testing BVY-99-116, Informs of Determination That Wh Schulze,License SOP-10528-1,will No Longer Maintain License at Facility. Termination of License Requested1999-09-21021 September 1999 Informs of Determination That Wh Schulze,License SOP-10528-1,will No Longer Maintain License at Facility. Termination of License Requested BVY-99-121, Requests Extension Until 990929 to Respond to Violations Noted in Insp Rept 50-271/99-12,dtd 990819.Licensee Did Not Receive Rept Until 990830 & Addl Time Is Needed to Prepare & Allow for Adequate Review of Violation Response Submittal1999-09-20020 September 1999 Requests Extension Until 990929 to Respond to Violations Noted in Insp Rept 50-271/99-12,dtd 990819.Licensee Did Not Receive Rept Until 990830 & Addl Time Is Needed to Prepare & Allow for Adequate Review of Violation Response Submittal ML20212C1621999-09-17017 September 1999 Forwards Amend 175 to License DPR-28 & Safety Evaluation. Amend Revises TSs to Enhance Limiting Conditions for Operation & Surveillance Requirements Relating to Standby Liquid Control System BVY-99-118, Responds to RAI Concerning GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions1999-09-16016 September 1999 Responds to RAI Concerning GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions BVY-99-115, Forwards non-proprietary & Proprietary Responses to 990714 RAI Re Civil & Mechanical Engineering Considerations for Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355.Proprietary Encls Withheld1999-09-16016 September 1999 Forwards non-proprietary & Proprietary Responses to 990714 RAI Re Civil & Mechanical Engineering Considerations for Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355.Proprietary Encls Withheld ML20216F3171999-09-13013 September 1999 Forwards Insp Rept 50-271/99-06 on 990621-0801.One Violation Identified & Being Treated as Noncited Violation BVY-99-110, Informs of Util Intent to Replace Commitments Made in Licensee & Subsequently Ack in NRC with Containment Insp Criteria Defined in 10CFR50.55a(b)(2)(vi),per Drywell Coating Insp1999-08-31031 August 1999 Informs of Util Intent to Replace Commitments Made in Licensee & Subsequently Ack in NRC with Containment Insp Criteria Defined in 10CFR50.55a(b)(2)(vi),per Drywell Coating Insp BVY-99-111, Informs That Encl TS Bases Page 91 Has Been Revised to Allow Reactivity Anomaly BOC Steady State Core Reactivity to Be Normalized Between off-line Uncorrected Solution & on-line 3D-Monicore Exposure Corrected Solution1999-08-31031 August 1999 Informs That Encl TS Bases Page 91 Has Been Revised to Allow Reactivity Anomaly BOC Steady State Core Reactivity to Be Normalized Between off-line Uncorrected Solution & on-line 3D-Monicore Exposure Corrected Solution ML20211G4791999-08-27027 August 1999 Forwards Notice of Withdrawal of 990420 Amend Request Re TS on Reloading & Unloading Sequence of Fuel in Reactor Core When All Fuel Removed from Core BVY-99-107, Submits Response to NRC RAI Re Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355 Fuel Assemblies1999-08-26026 August 1999 Submits Response to NRC RAI Re Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355 Fuel Assemblies ML20211E8841999-08-25025 August 1999 Requests That Licensee Provide bldg-specific Justification for Use of Method A.1 at Locations Where Amplification Significantly Exceeds 1.5 Limit Above 8 Hz ML20211E1371999-08-20020 August 1999 Forwards from J Bean to H Miller & FEMA Final Exercise Rept for 990427-29 Plume Exposure & Ingestion Pathway Exercise for Vermont Yankee Nuclear Power Station.No Deficiencies Noted.Areas Requiring C/A Identified ML20211H0851999-08-19019 August 1999 Forwards Insp Rept 50-271/99-12 on 990628-0711 & Nov. Violation Re Failure to Monitor Unavailability of Specific Sys,Structures & Components During Refueling Outage Did Not Allow Adequate Assessment of Maint Effectiveness BVY-99-108, Requests That Gv Bogue,Bj Croke,Vs Ferrizzi,Me French, Bk Mcnutt,Jf Meyer & DM Navarro Take BWR Gfes of OL Exam Administered on 991006.DA Daigler & ST Brown Will Have Access to Exams Before Tests Administered1999-08-19019 August 1999 Requests That Gv Bogue,Bj Croke,Vs Ferrizzi,Me French, Bk Mcnutt,Jf Meyer & DM Navarro Take BWR Gfes of OL Exam Administered on 991006.DA Daigler & ST Brown Will Have Access to Exams Before Tests Administered BVY-99-103, Informs That Util Expects to Submit Approx Twenty Licensing Actions in FY00 & FY01,in Response to Administrative Ltr 99-021999-08-18018 August 1999 Informs That Util Expects to Submit Approx Twenty Licensing Actions in FY00 & FY01,in Response to Administrative Ltr 99-02 BVY-99-100, Forwards Revised Floor Response Spectra Diagrams,Originally Sent as Attachment 1 to Licensee to Nrc.Revised Diagrams Have More Legible Scale Markings1999-08-0202 August 1999 Forwards Revised Floor Response Spectra Diagrams,Originally Sent as Attachment 1 to Licensee to Nrc.Revised Diagrams Have More Legible Scale Markings ML20210M5791999-07-30030 July 1999 Responds to NRC 990726 Telcon Re Status of Resolution for USI A-46 Outliers.Written Summary,By Equipment Category, Listed ML20211E1701999-07-28028 July 1999 Forwards Copy of Final Exercise Rept for 990427-29,full- Participation Plume Exposure & Ingestion Pathway Exercise of Offsite Radiological Emergency Response Plans site-specific to VYNPS ML20210G5041999-07-27027 July 1999 Responds to NRC 990301 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions. Licensee Will Submit Info Re Proposed Sys Mod by 990916 ML20210J3031999-07-27027 July 1999 Submits Proposed Changes to Eals.Attachment 1 Provides Listing of Changes to EALs Along with Ref to Bases Documents Supporting Change ML20210G4271999-07-27027 July 1999 Forwards Testing Data & Associated Results for Fitness for Duty Program at Plant for 990101-0630 ML20216D7321999-07-26026 July 1999 Forwards Insp Rept 50-271/99-05 on 990510-0620.Two Viiolations Being Treated as Noncited Violations ML20209G2721999-07-14014 July 1999 Discusses Licensee Response to RAI Re GL 92-01,Rev 1,Suppl Suppl 1, Rv Structural Integrity, for Vermont Yankee Nuclear Power Station ML20209J0601999-07-14014 July 1999 Forwards Rev 11 to Vols 1-10 of State of Nh Radiological Emergency Response Plan & Vols 11-50 to Town Radiological Emergency Response Plans,In Support of Vermont Yankee & Seabrook Station.Vols 17-19 of Were Not Included ML20209G6931999-07-14014 July 1999 Forwards Request for Addl Info Re Spent Fuel Storage Capacity Expansion ML20209G1531999-07-12012 July 1999 Discusses Util Setpoint Control Program Implementation Schedule,As Committed to in Licensee 990514 Response to Notice of Violation,Insp Rept 50-271/97-10 ML20196J2321999-06-30030 June 1999 Submits Input from Util Technical Staff Re Soil Disposal on-site Under 10CFR20.2002 & Expresses Interest in Pursuing Approval to Use Same Methodology (Implemented Through Util ODCM & Reported as Noted) If Possible ML20196J7421999-06-29029 June 1999 Informs NRC That Vygs Has Implemented Severe Accident Management,As Committed to in Licensee to NRC ML20209B6111999-06-29029 June 1999 Resubmits Summary of Vynp Commitments Page to Replace Original Page Submitted with Responding to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20196J2431999-06-29029 June 1999 Informs That Author Received Call from NRR on Dirt Spreading Ltr & Questions Re Cover Ltr Statement Where Util Asks to Be Allowed to Dispose of Future Soil in Same Manner Provided Same Acceptance Criteria Met ML20209C3751999-06-28028 June 1999 Forwards non-proprietary Rev 16 to EPIP OP 3524, Emergency Actions to Ensure Initial Accountability & Security Response & Proprietary Rev 12 to EPIP OP 3531, Emergency Call-In Method. Proprietary Encl Withheld ML20209B5861999-06-28028 June 1999 Provides Alternative Y2K Readiness Status Described in Supplement 1 to GL 98-01, Y2K Readiness of Computer Sys at Npps. Y2K Readiness Disclosure Rept Encl ML20196G5241999-06-22022 June 1999 Responds to Re Changes to Vermont Yankee Guard Training & Qualification Plan,Rev 8,Errata A.No NRC Approval Is Required.Encl Will Be Withheld from Public Disclosure Per 10CFR73.21 BVY-99-084, Forwards Proprietary Application & Medical Certificate for Mod of Listed SRO License,For Gj Leclair.Gj Leclair Will Be Trained & Evaluated in Accordance with Util Lsro Training Description.Proprietary Info Withheld,Per 10CFR2.7901999-06-18018 June 1999 Forwards Proprietary Application & Medical Certificate for Mod of Listed SRO License,For Gj Leclair.Gj Leclair Will Be Trained & Evaluated in Accordance with Util Lsro Training Description.Proprietary Info Withheld,Per 10CFR2.790 ML20212J0541999-06-17017 June 1999 Responds to Requesting That NRC Staff ...Allow BWR Plants Identified to Defer Weld Overlay Exams Until March 2001 or Until Completion of NRC Staff Review & Approval of Proposed Generic Rept,Whichever Comes First ML20195H1741999-06-15015 June 1999 Forwards Original & Copy of Request for Approval of Certain Indirect & Direct Transfer of License & Ownership Interests of Montaup Electric Co (Montaup) with Respect to Nuclear Facilities Described as Listed 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217F1261999-10-12012 October 1999 Forwards Update to Previously Submitted RELAP5 Analytical Assumptions for App R,Re RAI of 961104 BVY-99-130, Provides Clarification of Method for Determining MSIV Maximum & Minimum Pathway at Vermont Yankee Nuclear Power Station1999-10-0808 October 1999 Provides Clarification of Method for Determining MSIV Maximum & Minimum Pathway at Vermont Yankee Nuclear Power Station BVY-99-128, Submits Listed Addl Info in Support of 990414 Request for Clarification to SER Confirming Adequacy of Space Cooling for HPCI & RCIC Sys,Re Item II.K.3.24 of NUREG-0737.Copy of NEDE-24955,encl1999-10-0606 October 1999 Submits Listed Addl Info in Support of 990414 Request for Clarification to SER Confirming Adequacy of Space Cooling for HPCI & RCIC Sys,Re Item II.K.3.24 of NUREG-0737.Copy of NEDE-24955,encl ML20216J3531999-09-29029 September 1999 Responds to NRC Re Violations Noted in Insp Rept 50-271/99-12 on 990628-0811.Corrective Actions:Based on RFO 20 Maint Rule Outage Performance Review,Task Was Generated to Clarify & Enhance SD Monitoring Process BVY-99-122, Notifies of Intention to Reinstate Original Version of App F in FSAR & Correct Docket Re Assumption That Electrical Power Sys Are Designed IAW Requirements of GDC-171999-09-28028 September 1999 Notifies of Intention to Reinstate Original Version of App F in FSAR & Correct Docket Re Assumption That Electrical Power Sys Are Designed IAW Requirements of GDC-17 BVY-99-113, Requests Approval to Perform Alternative Testing to That Specified by ASME Boiler & Pressure Vessel Code,Section XI & Asme/Ansi OM, Operation & Maint of Nuclear Power Plants. Attachment 1 Provides Justification for Alternative Testing1999-09-21021 September 1999 Requests Approval to Perform Alternative Testing to That Specified by ASME Boiler & Pressure Vessel Code,Section XI & Asme/Ansi OM, Operation & Maint of Nuclear Power Plants. Attachment 1 Provides Justification for Alternative Testing BVY-99-114, Provides Notification That Licensee Completed Y2K Remediation Efforts Described in Util 990608 Response to NRC GL 98-01,Suppl 11999-09-21021 September 1999 Provides Notification That Licensee Completed Y2K Remediation Efforts Described in Util 990608 Response to NRC GL 98-01,Suppl 1 BVY-99-116, Informs of Determination That Wh Schulze,License SOP-10528-1,will No Longer Maintain License at Facility. Termination of License Requested1999-09-21021 September 1999 Informs of Determination That Wh Schulze,License SOP-10528-1,will No Longer Maintain License at Facility. Termination of License Requested BVY-99-121, Requests Extension Until 990929 to Respond to Violations Noted in Insp Rept 50-271/99-12,dtd 990819.Licensee Did Not Receive Rept Until 990830 & Addl Time Is Needed to Prepare & Allow for Adequate Review of Violation Response Submittal1999-09-20020 September 1999 Requests Extension Until 990929 to Respond to Violations Noted in Insp Rept 50-271/99-12,dtd 990819.Licensee Did Not Receive Rept Until 990830 & Addl Time Is Needed to Prepare & Allow for Adequate Review of Violation Response Submittal BVY-99-115, Forwards non-proprietary & Proprietary Responses to 990714 RAI Re Civil & Mechanical Engineering Considerations for Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355.Proprietary Encls Withheld1999-09-16016 September 1999 Forwards non-proprietary & Proprietary Responses to 990714 RAI Re Civil & Mechanical Engineering Considerations for Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355.Proprietary Encls Withheld BVY-99-118, Responds to RAI Concerning GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions1999-09-16016 September 1999 Responds to RAI Concerning GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions BVY-99-110, Informs of Util Intent to Replace Commitments Made in Licensee & Subsequently Ack in NRC with Containment Insp Criteria Defined in 10CFR50.55a(b)(2)(vi),per Drywell Coating Insp1999-08-31031 August 1999 Informs of Util Intent to Replace Commitments Made in Licensee & Subsequently Ack in NRC with Containment Insp Criteria Defined in 10CFR50.55a(b)(2)(vi),per Drywell Coating Insp BVY-99-111, Informs That Encl TS Bases Page 91 Has Been Revised to Allow Reactivity Anomaly BOC Steady State Core Reactivity to Be Normalized Between off-line Uncorrected Solution & on-line 3D-Monicore Exposure Corrected Solution1999-08-31031 August 1999 Informs That Encl TS Bases Page 91 Has Been Revised to Allow Reactivity Anomaly BOC Steady State Core Reactivity to Be Normalized Between off-line Uncorrected Solution & on-line 3D-Monicore Exposure Corrected Solution BVY-99-107, Submits Response to NRC RAI Re Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355 Fuel Assemblies1999-08-26026 August 1999 Submits Response to NRC RAI Re Proposed Change to TS to Increase Spent Fuel Storage Capacity from 2,870 to 3,355 Fuel Assemblies BVY-99-108, Requests That Gv Bogue,Bj Croke,Vs Ferrizzi,Me French, Bk Mcnutt,Jf Meyer & DM Navarro Take BWR Gfes of OL Exam Administered on 991006.DA Daigler & ST Brown Will Have Access to Exams Before Tests Administered1999-08-19019 August 1999 Requests That Gv Bogue,Bj Croke,Vs Ferrizzi,Me French, Bk Mcnutt,Jf Meyer & DM Navarro Take BWR Gfes of OL Exam Administered on 991006.DA Daigler & ST Brown Will Have Access to Exams Before Tests Administered BVY-99-103, Informs That Util Expects to Submit Approx Twenty Licensing Actions in FY00 & FY01,in Response to Administrative Ltr 99-021999-08-18018 August 1999 Informs That Util Expects to Submit Approx Twenty Licensing Actions in FY00 & FY01,in Response to Administrative Ltr 99-02 BVY-99-100, Forwards Revised Floor Response Spectra Diagrams,Originally Sent as Attachment 1 to Licensee to Nrc.Revised Diagrams Have More Legible Scale Markings1999-08-0202 August 1999 Forwards Revised Floor Response Spectra Diagrams,Originally Sent as Attachment 1 to Licensee to Nrc.Revised Diagrams Have More Legible Scale Markings ML20210M5791999-07-30030 July 1999 Responds to NRC 990726 Telcon Re Status of Resolution for USI A-46 Outliers.Written Summary,By Equipment Category, Listed ML20211E1701999-07-28028 July 1999 Forwards Copy of Final Exercise Rept for 990427-29,full- Participation Plume Exposure & Ingestion Pathway Exercise of Offsite Radiological Emergency Response Plans site-specific to VYNPS ML20210G5041999-07-27027 July 1999 Responds to NRC 990301 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions. Licensee Will Submit Info Re Proposed Sys Mod by 990916 ML20210G4271999-07-27027 July 1999 Forwards Testing Data & Associated Results for Fitness for Duty Program at Plant for 990101-0630 ML20210J3031999-07-27027 July 1999 Submits Proposed Changes to Eals.Attachment 1 Provides Listing of Changes to EALs Along with Ref to Bases Documents Supporting Change ML20209J0601999-07-14014 July 1999 Forwards Rev 11 to Vols 1-10 of State of Nh Radiological Emergency Response Plan & Vols 11-50 to Town Radiological Emergency Response Plans,In Support of Vermont Yankee & Seabrook Station.Vols 17-19 of Were Not Included ML20209G1531999-07-12012 July 1999 Discusses Util Setpoint Control Program Implementation Schedule,As Committed to in Licensee 990514 Response to Notice of Violation,Insp Rept 50-271/97-10 ML20196J2321999-06-30030 June 1999 Submits Input from Util Technical Staff Re Soil Disposal on-site Under 10CFR20.2002 & Expresses Interest in Pursuing Approval to Use Same Methodology (Implemented Through Util ODCM & Reported as Noted) If Possible ML20209B6111999-06-29029 June 1999 Resubmits Summary of Vynp Commitments Page to Replace Original Page Submitted with Responding to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20196J7421999-06-29029 June 1999 Informs NRC That Vygs Has Implemented Severe Accident Management,As Committed to in Licensee to NRC ML20209C3751999-06-28028 June 1999 Forwards non-proprietary Rev 16 to EPIP OP 3524, Emergency Actions to Ensure Initial Accountability & Security Response & Proprietary Rev 12 to EPIP OP 3531, Emergency Call-In Method. Proprietary Encl Withheld ML20209B5861999-06-28028 June 1999 Provides Alternative Y2K Readiness Status Described in Supplement 1 to GL 98-01, Y2K Readiness of Computer Sys at Npps. Y2K Readiness Disclosure Rept Encl BVY-99-084, Forwards Proprietary Application & Medical Certificate for Mod of Listed SRO License,For Gj Leclair.Gj Leclair Will Be Trained & Evaluated in Accordance with Util Lsro Training Description.Proprietary Info Withheld,Per 10CFR2.7901999-06-18018 June 1999 Forwards Proprietary Application & Medical Certificate for Mod of Listed SRO License,For Gj Leclair.Gj Leclair Will Be Trained & Evaluated in Accordance with Util Lsro Training Description.Proprietary Info Withheld,Per 10CFR2.790 ML20195H1741999-06-15015 June 1999 Forwards Original & Copy of Request for Approval of Certain Indirect & Direct Transfer of License & Ownership Interests of Montaup Electric Co (Montaup) with Respect to Nuclear Facilities Described as Listed ML20195C5891999-05-27027 May 1999 Forwards Response to NRC 990301 RAI Re GL 96-05 Program at Vermont Yankee Nuclear Power Station ML20195D5341999-05-27027 May 1999 Forwards Description of Vermont Yankees Plans for Insp of & Mods to Certain Reactor Vessel Internals BVY-99-074, Forwards Application & Medical Certificate Required for Renewal of Jd Livingston,License OP-10049,RO License.Medical Certificate Withheld1999-05-26026 May 1999 Forwards Application & Medical Certificate Required for Renewal of Jd Livingston,License OP-10049,RO License.Medical Certificate Withheld ML20195B4081999-05-24024 May 1999 Withdraws Licensee Commitment,Contained in ,To Reinitiate ITS Project Following Completion of FSAR Accuracy Verification Project.Util Will Continue to Modify Current TS with Number of Improvements BVY-99-067, Informs That Bw Metcalf,License SOP-1761-9,has Retired from VYNPS & Will No Longer Require License.Nrc Is Requested to Terminate License1999-05-21021 May 1999 Informs That Bw Metcalf,License SOP-1761-9,has Retired from VYNPS & Will No Longer Require License.Nrc Is Requested to Terminate License ML20196L1801999-05-18018 May 1999 Withdraws Licensee & Attachment,Containing Rev 2 to Vermont Yankee Operational QA Manual, from Further Consideration by Nrc.Summary of Commitments Encl ML20206K3201999-05-0707 May 1999 Forwards Response to RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML20206J2801999-04-30030 April 1999 Forwards 1998 Annual Financial Repts for CT Light & Power Co,Western Ma Electric Co,Public Svc Co of Nh,North Atlantic Energy Corp,Northeast Nuclear Energy Co & North Atlantic Energy Svc Corp,License Holders ML20206D3731999-04-27027 April 1999 Informs NRC of Changes in Recipients of NRC Docketed Correspondence ML20206B1401999-04-23023 April 1999 Forwards Replacement of Section 3(a) of NSHC Determination Provided by Re TS Proposed Change 208,suppl Section 6 ML20205S3381999-04-16016 April 1999 Submits Revised Schedule for Response to NRC 990226 RAI Re 980630 Submittal of IPEEE Rept.Info Will Be Submitted by 991231 ML20205S3891999-04-16016 April 1999 Forwards non-proprietary & Proprietary Revised Page to Holtec Rept HI-981932,supplementing TS Proposed Changed 207 Re Spent Fuel Pool Storage Capacity Expansion ML20205S3031999-04-15015 April 1999 Forwards Revised TS Bases Pages 90,227,164 & 221a,accounting for Change in Reload Analysis from Yaec to GE Methodology, Reflecting Change in Condensation Stability Design Criteria & Accounting for More Conservative Calculation ML20205P9291999-04-14014 April 1999 Requests That Rev to NRC 821029 SER for NUREG-0737,Item II.K.3.24,be Issued to Clarify Util Installed RCIC & HPCI HVAC Configuration,As Discovered During Preparation of DBDs for Sys ML20205P8191999-04-13013 April 1999 Forwards Rev 2 to COLR for Vermont Yankee Cycle 20, Dtd Feb 1999,IAW TS Section 6.7.A.4 ML20205M3191999-04-0707 April 1999 Forwards 1998 Annual Rept of Results of Individual Monitoring, Per 10CFR20.2206(b).Licensee Is Submitting Matl to Only Addressee Specified in 10CFR20.2206(c).Without Encl ML20205K0351999-03-31031 March 1999 Informs That Certain Addl Corrections Warranted for 990121 SER for Amend 163 to License DPR-28 Re Suppression Pool Water Temp.Suggested Corrections Listed ML20205K1821999-03-31031 March 1999 Informs of Modifications That Util Made to CO(2) Fire Suppression Sys,Due to Sen 188 Which Occurred at Ineel on 980728.Compensatory Actions Will Remain in Place Until Modifications Are Complete & Systems Are Returned to Svc ML20206A6951999-03-29029 March 1999 Request Confirmation That No NRC Action or Approval,Required Relative to Proposed Change in Upstream Economic Ownership of New England Power Co,Minority Shareholder in Vermont Yankee Nuclear Power Corp,Yaec,Myap & Connecticut Yankee 1999-09-29
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059J2831990-09-10010 September 1990 Forwards Updated Operator Licensing Exam Schedule for FY91, FY92,FY93 & FY94,per Generic Ltrs 90-07 & 89-12 ML20059D9831990-08-28028 August 1990 Forwards fitness-for-duty Program Performance Data for 900103-0630,per 10CFR26.71.NRC Review of Data Will Provide Realization That Positive Testing Rate Extremely Low & Limited to pre-access Testing Population BVY-90-087, Forwards Addl Info on Use of RELAP5YA Program for LOCA Analyses.Proprietary Encl Withheld1990-08-28028 August 1990 Forwards Addl Info on Use of RELAP5YA Program for LOCA Analyses.Proprietary Encl Withheld BVY-90-086, Responds to NRC Re Violations Noted in Insp Rept 50-271/90-06.Corrective Actions:Incident Rept Initiated & All Required Locking Devices in Place by 9007061990-08-24024 August 1990 Responds to NRC Re Violations Noted in Insp Rept 50-271/90-06.Corrective Actions:Incident Rept Initiated & All Required Locking Devices in Place by 900706 ML20059F6681990-08-22022 August 1990 Comments on Review of Amend 115 to License DPR-28,including Safety Evaluation.Requests Explanation of Statement in NRC Re How NRC Considers Comments & What Resolution Could Be for Each Util Comment in BVY-90-085, Informs That Sys Testing & Operator Training Successfully Completed & SPDS Declared Operable on 900813.Util Intends to Operate SPDS in Parallel W/Original Honeywell Gepac Plant Computer Until mid-Nov 19901990-08-15015 August 1990 Informs That Sys Testing & Operator Training Successfully Completed & SPDS Declared Operable on 900813.Util Intends to Operate SPDS in Parallel W/Original Honeywell Gepac Plant Computer Until mid-Nov 1990 BVY-90-084, Notifies NRC of Intentions to Install Test Fuel Assemblies & Test Control Blades During Cycle 15 Refueling Outage in Sept 19901990-07-24024 July 1990 Notifies NRC of Intentions to Install Test Fuel Assemblies & Test Control Blades During Cycle 15 Refueling Outage in Sept 1990 BVY-90-082, Informs That Effective 900723 Facility Implemented Rev 4 of Procedure Generating Package & Corresponding Revs to Eops. Revs Developed Per Rev 4 of BWR Owners Group Emergency Procedure Guidelines1990-07-24024 July 1990 Informs That Effective 900723 Facility Implemented Rev 4 of Procedure Generating Package & Corresponding Revs to Eops. Revs Developed Per Rev 4 of BWR Owners Group Emergency Procedure Guidelines BVY-90-071, Forwards Rev 2 to Training & Qualification Plan.Rev Withheld (Ref 10CFR73.21)1990-07-20020 July 1990 Forwards Rev 2 to Training & Qualification Plan.Rev Withheld (Ref 10CFR73.21) BVY-90-078, Forwards List of Refs for Proposed Change 161 to Facility OL & Tech Specs1990-07-17017 July 1990 Forwards List of Refs for Proposed Change 161 to Facility OL & Tech Specs BVY-90-072, Forwards Supplemental Effluent & Waste Disposal Semiannual Rept for Third & Fourth Quarters 1989,Including Annual Radiological Impact on Man for 19891990-06-27027 June 1990 Forwards Supplemental Effluent & Waste Disposal Semiannual Rept for Third & Fourth Quarters 1989,Including Annual Radiological Impact on Man for 1989 ML20043G4351990-06-15015 June 1990 Requests Temporary Waiver of Compliance from Tech Spec Requirements for Limiting Conditions for Operation for Certain post-accident Monitoring Instrumentation Listed in Tech Spec Table 3.2.6.Parameters Listed ML20043E4011990-06-0808 June 1990 Responds to Second Request for Addl Info on Use of RELAP5YA. Explanation Re Why More Accurate View Factor Calculation Not Included in Huxy Code Addressed ML20043C6131990-06-0101 June 1990 Forwards YAEC-1659-A, Simulate-3 Validation & Verification. ML20043C5991990-06-0101 June 1990 Forwards Accepted Version of YAEC-1683-A, MICBURN-3/ CASMO-3/TABLES-3/SIMULATE-3 Benchmarking of Vermont Yankee Cycles 9 Through 13. ML20043C4821990-05-30030 May 1990 Informs of Three Organizational Changes That Will Become Effective on 900601.WP Murphy,Jp Pelletier & DA Reid Will Be Senior Vice President of Operations,Newly Created Vice President of Engineering & Plant Manager,Respectively ML20043B7561990-05-23023 May 1990 Informs That Util Intends to Utilize Relationship Between Frosstey & FROSSTEY-2 to Support Cycle 15 Calculations.Nrc Approval of FROSSTEY-2 Needed by Aug 1990 for LOCA Analysis Program ML20043B6481990-05-17017 May 1990 Forwards Rev 19 to Physical Security Plan.Rev Withheld (Ref 10CFR73.21) BVY-90-058, Forwards Public Version of Vermont Yankee Nuclear Power Station Emergency Response Preparedness Exercise 1990. Exercise Scenario Package Includes All Info Pertinent to Performance of Exercise Scheduled for 9007181990-05-17017 May 1990 Forwards Public Version of Vermont Yankee Nuclear Power Station Emergency Response Preparedness Exercise 1990. Exercise Scenario Package Includes All Info Pertinent to Performance of Exercise Scheduled for 900718 ML20042G9061990-05-10010 May 1990 Forwards Proprietary Supplemental Info to 900419 Response to NRC 900309 Ltr Re FROSSTEY-2 Fuel Performance Code.Info Withheld ML20042F6471990-05-0404 May 1990 Ack That NRC Will Issue Supplementary Info to NRC 900307 Request for Installation of Neutron Flux Monitoring Instrumentation That Conforms to Requirements of Reg Guide 1.97 & 10CFR50.49 at Plant ML20042E7291990-04-23023 April 1990 Forwards Pages Omitted from 900314 Revs 16-18 to Physical Security Plan.Revs Withheld ML20012F3511990-03-30030 March 1990 Provides Supplemental Response to Station Blackout Rule (10CFR50.63).Util Will Use Alternate Ac Power Source Available within 10 Minutes of Onset of Station Blackout to Meet Requirements of Station Blackout Rule ML20012D0301990-03-19019 March 1990 Forwards Response to Generic Ltr 89-19 Re Resolution of USI A-47.Feedwater Sys Trip Relays,Interfacing W/Feedwater Pump Control Circuitry,Powered from Supplies Originating from safety-related Dc Sources ML20012D0241990-03-16016 March 1990 Forwards Supplemental Info Re Feedwater Check Valve V28B Flaws Evaluation,Per NRC Request.Util Remains Committed to Replacement of Subj Valve During Upcoming 1990 Refueling Outage ML20012C6381990-03-15015 March 1990 Forwards Vermont Yankee Nuclear Power Corp Financial Statements 891231,1988 & 1987. ML20012C6071990-03-15015 March 1990 Forwards Method for Generation of One-Dimensional Kinetics Data for RETRAN-02, Per NUREG-0393 & 891211 Request ML20012B8311990-03-0909 March 1990 Forwards Proprietary Vermont Yankee Evaluation Model Sample Problem 0.7 Ft(2) Break in Recirculation Discharge Loop, in Response to 900208 Telcon.Rept Withheld (Ref 10CFR2.790) ML20012B6131990-03-0909 March 1990 Informs of Schedular Changes Made W/Regard to Plant Licensed Operator Requalification Training Program ML20006E8871990-02-15015 February 1990 Provides NRC W/Results of Licensee Review of Design Bases & Operability Status of torus-to-reactor Bldg Vacuum Breakers ML20011E6791990-02-0505 February 1990 Responds to Weaknesses Noted in SALP Rept 50-271/88-99 for Jul 1988 to Sept 1989.Implementation of Emergency Response Facility Info Sys Nearing Completion & Remaining Safety Class Vendor Manuals Will Be Completed During 1990 ML20006D1571990-02-0202 February 1990 Responds to 891226 Request for Addl Info Re YAEC-1683 on MICBURN-3/CASMO-3/TABLES-3/SIMULATE-3 Benchmarking.Hot Eigenvalue Std Deviation on Table 5.7 of YAEC-1683 Reduced to 0.00098 w/SIMULATE-3 ML20006B1351990-01-22022 January 1990 Forwards Responses to Generic Ltr 89-13 Re Svc Water Sys Problems Affecting safety-related Equipment.Establishment of Program Revs Prior to Startup from Next Refueling Outage, Scheduled for Fall 1990,planned ML20006A4441990-01-16016 January 1990 Forwards Revised Page 127 of Tech Specs to Clarify Proposed Change 134, Rev of Pressure Suppression - Reactor Bldg Vacuum Breaker Sys Operability Requirements. Change Involves Adoption of Language Consistent W/Bwr STS ML19354E8001990-01-16016 January 1990 Forwards Addl Info Re Testing of Cable Vault C02 Suppression Sys During 891031-1102,per NRC 890518 & 0821 Requests.Encl Final Test Rept Demonstrates That Carbon Dioxide Sys Will Satisfy Design Bases for Greater than 10 Minutes in Room ML20005G0841990-01-10010 January 1990 Responds to NRC Bulletin 89-002 Re Stress Corrosion Cracking of high-hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Design ML20005E8201990-01-0202 January 1990 Forwards Minutes of NRC 890907 Meeting W/Util in Rockville,Md Re Util LOCA Analysis Program.List of Attendees Also Encl ML20005F0551990-01-0202 January 1990 Informs That Util Has Implemented Fitness for Duty Program, in Compliance w/10CFR26 ML20005E3531989-12-28028 December 1989 Forwards Response to Generic Ltr 89-10 Re safety-related motor-operated Valve Testing & Surveillance.Util Intends to Extend Existing IE Bulletin 85-003 Program to Cover motor- Operated Valves within Scope of Ltr ML20005E3191989-12-28028 December 1989 Responds to Violations Noted in Insp Rept 50-271/89-17 on 890906-1016.Corrective Actions:Plant Procedures Revised & Addl Meetings Between Plant Manager,Dept Supervisors & Personnel to Take Place ML19332G1791989-12-12012 December 1989 Forwards Rev 0 to Vermont Yankee Nuclear Power Station Cycle 14 Core Operating Limits Rept. ML19332F2781989-11-30030 November 1989 Forwards Rev 1 to YAEC-1693, Application of One-Dimensional Kenetics to BWR Transient Analysis Methods, Per 891106 Ltr.Rept Presents Methodology,Verification & Justification for Application of RETRAN-02 One Dimensional Option ML19332E3511989-11-29029 November 1989 Forwards Annual Cashflow Statements for 1989 as Evidence of Util Maint of Approved Guarantee,Per Requirements of 10CFR140.21 Re Licensee Guarantees of Payment of Deferred Premiums ML19332E5281989-11-28028 November 1989 Requests Removal of Change B to Proposed Change 148 Re Rev to Pages 5b & 6a Correcting Administrative Error in Tech Spec 2.1 ML19332D3801989-11-22022 November 1989 Responds to NRC Generic Ltr 89-21 Re Request for Info Re Status of Implementation of USI Requirements.Encl Table Details Implementation Status for USIs for Which Final Technical Resolution Achieved ML19324C1501989-11-10010 November 1989 Responds to NRC Bulletin 88-010,Suppl 1 Re Molded Case Circuit Breakers.Program Initiated to Ensure That Breakers Can Perform Safety Functions ML19324C2201989-11-0606 November 1989 Requests Change in Review & Approval Basis from Facility Specific to Generic Because Methods Described in YAEC-1693 & YAEC-1694 Applicable to All BWRs ML19325F0261989-11-0606 November 1989 Responds to Generic Ltr 89-07, Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs. Util Has Evaluated Listed Considerations,Including Safe Standoff Distances for Vital Equipment ML19324B7431989-10-30030 October 1989 Responds to Generic Ltr 89-16 Re Installation of Hardened Wetwell Vent.Util Expects to Establish Specific Design Criteria to Install Enhanced Containment Overpressure Protection Capability by End of 1992 Refueling Outage ML19324B8481989-10-30030 October 1989 Provides NRC W/Test Acceptance Criteria for Alternate Test of CO2 Suppression Sys,Per 891025 Meeting.Ability to Contain CO2 at Appropriate Concentration for Required Duration,As Well as Ability to Withstand Dynamics of Discharge,Verified 1990-09-10
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VERMONT YANKEE NUCLEAR POWER CORPORATION
- FVY 88-47 RD 5. Box 169 Ferry Road, Brattleboro, VT 05301 ,my ,o ENGINEERING OFFICE Sqr 1671 WORCESTER ROAD FR AMINGH AM, MASSACHUSETTS 01701
. T E L E PHON E 617472-8100 June 7, 1988 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk
References:
a) License No. DPR-28 (Docket No. 50-271) b) Letter, VYNPC to USNRC, FVY 86-34, "Proposed Technical Specification Change for New and Spent Fuel Storage", dated 4/24/86 c) Letter, VYNPC to USNRC, FVY 88-17, "Vermont Yankee Proposed Change No. 133 - Spent Fuel Pool Expansion,"
dated 3/2/88 d) Letter, USNRC to VYNPC, NVY 88-093, "Spent Fuel Poo*
Expansion Reracking - Amendment No. 104," dated 5/20/88
Dear Sir:
Subject:
Vermont Yankee Proposed Technical Specification Change for New and Spent Fuel Storage By letter dated April 24, 1986 (Reference b)], Vermont Yankee sub-mitted a proposed license amendment request to revise Section 5.5, "Spent and New Fuel Storage" of the Vermont Yankee Technical Specifications to increase the number of spent fuel assemblies allowed to be stored in the spent fuel pool. By letter dated March 2, 1988 (Reference c)], Vermont Yankee commited to design, install, test, and make operational, a redundant seismically designed Spent Fuel Pool Cooling System prior to the time Vermont Yankee exceeds the existing Technical Specification limit of 2,000 spent fuel assembly storage limit in the Vermont Yankee spent fuel pool.
Subsequently, by letter dated May 20, 1988 (Reference d)], Amendment No.
104 to Vermont Yankee's license was issued allowing the installation of racks of a new design in the spent fuel pool sufficient to accommodate 2,870 fuel assemblies, and the storage of fuel assemblies in the new racks up to the present Technical Specification limit of 2,000 assemblies in the pool. Use of the remaining 870 storage positions for the storage of fuel assemblies was not authorized by the license amendment.
The NRC letter of May 20, 1988 transmitting Amendment No. 104 to the Vermont Yankee license stated that the staff would complete its review of the thermal-hydraulic aspects of Vermont Yankne's proposed change and con-sider a decision to increase the Technical Specification limit to 2,870 foO\
\
8806150435 880607 PDR ADOCK 05000271 {
P n r.n
VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission June 7, 1988 Page 2 assemblies after learning more about Vermont Yankee's plans for enhancing the Spent Fuel Pool Cooling System. Accordingly, Vermont Yankee submits as Attachment A to this letter a description of the enhanced Spent Fuel Pool Cooling System in the format of a revised Final Safety Analysis Report (FSAR). The design, installation and testing of the enhanced system will be in accordance with 10 CFR 50.59. On the basis of the information pro-vided in Attchment A, Vermont Yankee requests issuance of the subject license amendment allowing Vermont Yankee use of the full 2,870 storage positions for storage of fuel assemblies in the spent fuel storage pool.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION tS(4 / m Warren P. Nur y Vice Pre ident a Manager of Ope ns
/dm cc Mr. V. Rooney, USNRC USNRC Regional Administrator, Region I USNRC Resident Inspector, VYNPC ASLB Service List
__ __ _ __________- - _____j
m.
VYNPS FUEL P00L' COOLING AND DEMINERALIZER SYSTEM TABLE OF CONTENTS Section Title Page.
A.1 Power Generation 0bjective............................ A-1 oA . 2 Safety 0bjective...................................... A-1 A.3 ' Power Generation Design
- Bases......................... A-1 A.4 Sa f e t y De s i gn Ba s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2 A.5 Description........................................... A-2 A.6 Safety Evaluation..................................... A-10 A.7 Inspection and Testing................................ A-12 i
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A-i 7
VYNPS
- i. _, v00L COOLING AND DEMINERALIZER SYSTEM LIST OF IIGURES Figure No. Title A-1 Fuel Pool Cooling System A-2 Fuel Pool Filter Demineralizer System A-il
z
. . O) .*.,
VYNPS FUEL POOL COOLING AND DEMINERALIZER SYSTEM LIST OF TABLES 4
i i ~ Table No. Title
- A.1 Fuel Pool Cooling and Demineralizer System ' System Specifications.
A.2 Fuel Decay Heat - After Normal Refueling or Full Core Discharged to Pool - Estimated Using SRP 9.1.3 1.
J 4
9 A-iii
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I VYNPS A FUEL POOL COOLING AND DEMINERALIZER SYSTEM ,
A.1 Power Generation Objective The objective of the Fuel Pool Cooling and Demineralizer System is to remove-the decay heat released from the spent fuel elements. The system maintains a
'specified fuel pool water temperature, purity, water clarity, and water level.
A.2 Safety Objective The safety objective of the Fuel Pool Cooling and Demineralizer System f r *.o remove decay heat from the stored fuel and maintain fuel pool watei ,
temperature at a level which will help maintain the Reactor Building environment within the bounding limits of the environmental qualification of electrical equipment.
t A.3 Power Generation Design Bases
- 1. The Fuel Pool Cooling and Demineralizer System shall minimize corrosion product buildup within the spent fuel pool and shall maintain proper water clarity, so that the fuel assemblies can be efficiently handled underwater.
- 2. The Fuel Pool Cooling and Demineralizer System shall minimize fission product concentration in the spent fuel pool water, thereby minimizing the radioactivity which could be released from the pool to the Reactor Building environment.
- 3. The Fuel Pool Cooling and Demineralizer System shall monitor fuel pool I
water level and maintain a water level above the fuel sufficient to provide shielding for normal building occupancy.
i l 4. The Fuel Pool Cooling System shall be capable of maintaining the spent fuel pool temperature below 150 F.
i I
! A-1 1
_ _ _ _ _ _ _ _ . _ _ _ _/
>+. .
'_y.
e
. VYNPS A.4 Safety Deslan Basis The Fuel Pool Cooling and Demineralizer System shall be designed to remove the decay heat from the fuel assemblies and maintain fuel pool-water temperature at a level which will help maintain the Reactor Building environment within the bounding limits of the environmental qualification of electrical equipment.
A.5 Description General The Fuel Pool Cooling and Demineralizer System (FPCDS) consists of four heat exchangers, four pumps, two decineralizers, piping and sufficient valves for control of the design functicas and required isolation capability. The Fuel Pool Cooling and Demineralizer System pumps and heat exchangers are located in the Reactor Building below the bottom elevation of the fuel pool.
The fuel pool concrete structure, metal liner, spent fuel storage racks, and the Emergency Standby Subsystem of the FPCDS are designed to withstand Seismic Class I earthquake loads.
The FPCDS equipment is arranged in such a way as to provide a system with two independent means of cooling the spent fuel pool.
Normal spent fuel pool cooling and cleanup is provided by using the Normal Fuel Pool Cooling Subsystem. This subsystem consist of Pumps P-9-1A and IB and Heat Exchangers E-19-1A and IB which are arranged in two parallel trains with one train normally lined up and operating during plant operation. This subsystem of the FPCDS is used to provide pool water filtration and demineralization to maintain proper pool water clarity and cleanliness for refueling operations. The Normal Fuel Pool Cooling Subsystem also provides sufficient pool cooling to maintain pool temperatures within specified limits during normal refuelings (nominal one-third core discharge) and plant operations.
l l A-2 l .
VYNPS
.s However, should an unusually high spent fuel decay heat load be placed in the pool, or a seismic event occur, the Emergency Standby Subsystem can be utilized to maintain pool temperatures within specified limits. The Emergency Standby Subsystem of the FPCDS consists of Pumps P-19-2A and 2B and Heat Exchangers E-19-2A and 2B which are normally lined up as two parallel trains in a standby mode to the Normal Fuel Pool Cooling Subsystem. Each train of the Emergency Standby Subsystam can be placed in service remotely.
Calculations of expected decay heat loads from normal refuelings and from a foil core discharge both with previous cycles of spent fuel in the reeks were performed in accordance with the guidance provided in NRC Standard Review Plan 9.1.3, Revision 1, dated July 1981. The normal discharges were assumed discharged to the pool at six days and ten days following shutdown from normal operation. The full core discharge was assumed discharged to the pool ten days following shutdown from normal operation for refueling. Six days following shutdown for a normal refueling is derived from the guidance provided in NRC Standard Review Plan 9.1.3. Ten days following shutdown for a normal refueling or a full core discharge is the earliest time at which the refueling cavity gates could be replaced isolating the reactor vessel from the spent fuel pool. The transfer of the spent fuel assemblies from the reactor vessel to the spent fuel pool is assumed to occur instantly at the six-day or ,
ten-day time period providing a conservative fuel decay heat load in the spent fuel pool. Data from these analyses are provided in Table A.2. Examination of this data shows that while the Normal Fuel Pool Cooling Subsystem heat exchanger capacity may be exceeded for relatively short spent fuel decay times, the backup capability of the Emergency Standby Subsystem of the FPCDS is more than sufficient and can be placed in service until the fuel decay heat load is reduced.
The operating temperature of the fuel pool is permitted to rise up to ?"[F above the administrative temperature limit (125 F) when the circulation flow is temporarily interrupted or when larger than normal batches of spent fuel are placed in the pool.
A-3 l
.. .. I F VYNPS Emergency Standby Subsystem The' Emergency Standby Subsystem (ESS) of the FPCDS is shown in Figure A-1.
l .
E The Emergency Standby Subsystem of the FPCDS is a two train, Seismic Class I, Safety Class.3 System designed to prevent a single active failure from disabling both trains. It is designed as a standby system that can remotely be placed in operation from the Control Room. This portion of the system cools the fuel storage pool by transferring the spent fuel decay heat (see Table A.2) to the Service Water System. The pumps circulate the pool water in a closed loop, taking suction from the spent fuel storage pool through the heat exchangers and discharging it back into the fuel pool.
The emergency standby heat exchangers are of the shell and tube design, with all parts in contact with the pool water being corrosion resistant raaterial.
These heat exchangers are each sized to maintain the fuel pool water temperature below 150 F after a normal refueling. Considering one train (one heat exchanger and one pump), this heat removal capability encompasses the normal maximum hest load from completely filling the pool with 2,870 spent fuel asseu2blies from the last normal discharge. The combined heat removal capability considering both trains (two heat exchangers and two pumps) operating encompasses a full core discharge heat load completely filling the pool with 2,870 spent fuel assemblies. This provides sufficient heat removal capacity to preclude any impact on plant operation'due to insufficient spent fuel pool cooling.
i' The heat exchangers are cooled by the seismically qualified safety-related l
l Service Water System (SWS). The design of the system places the heat 1
exchangers on the cuction side of the pumps. In order to protect against fuel pool water leakage into the Service Water System, a positive differential pressure is maintained. The fuel pool water side of the heat exchangers has a 1,
l maximum operating pressure equivalent to the stai.ic pressure head f rom the pool surface to the heat exchanger. The Service Water System side of the heat exchangers has a minimum operating pressure which is greater than the maximum pressure on the fuel pool side of the heat exchangers. By providing a l-
! A-4
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' operation, leakage of fuel pool water to the environment is prevented. The differential pressure across each heat exchanger is monitored by a differential presaure indicator in the Control Room.
The Emergency Standby Subsystem of the FPCDS includes two centrifugal pumps each with a design flow of 700 gpm. All the parts of the pump in contact with
^
the pool water are corrosion-resistant material. The pumps are Seismic Class I and environmentally qualified to ensure operability af ter exposure to 4
a harsh environment. The pumps are located within the FPCDS cubicle in such a manner to prevent common mode failure from fire, flooding, or missiles. A low discharge pressure alarm indicates in the Control Room, plus, the pumps are automatically tripped on a low suction pressure condition. One pump alone is designed to provide sufficient flow for the maximum normal heat load from a normal refueling discharge. For an abnormal heat load, such as full core discharge, two pumps can be running concurrently (one in each train)
(reference Table A-1).
Four Motor-Operated Valves (MOVs) provide isolation from the nonseismic Normal Fuel Pool Cooling Subsystem and isolation and throttling of the service water through the heat exchangers. Each heat exchanger service water outlet MOV is powered by the same electrical source as its respective Emergency Standby Subsystem pump. These two MOVs V-19-J and K are throttling-type valves providing service water flow control through its respective heat exchanger, and thereby controlling both pool temperature and service water to fuel pool cooling differential pressure.
The two Normal Fuel Pool Cooling Subsystem Isolation Valves V-19-H and I are nonthrottling MOVs, each powered by a different safety-related electrical power supply. These isolation valves receive a signal to close on low pool level, providing automatic pool isolation from the Normal Fuel Pool Cooling Subsystem in case of a line break in this nonseismic portion of the FPCDS. In conjunction with the two Normal Fuel ?ool Cooling Subsystem isolation MOVs in the supply line, there are two discharge line check valves. These Check Valves V-19-18 and V-19-G provide isola. ion of the nonseismic Normal Fuel Pool Cooling Subsystem from the A-5
VYNPS Emergency Standby seismic portion of the system. Thus, isolation of the nonseismic portions of the Normal Fuel Pool Cooling Subsystem is assured.
Piping associated with the Service _ Water supply and discharge to the heat exchangers and the fuel pool water piping will be cf corrosion resistant material. The piping is designed and constructed in accordance with the requirements of ANSI B31.1-77. Valves in the fuel pool water piping are chosen considering their propensity not to collect cor' .on products, pressure tight sealing capability, and ease of maintenance.
Indication is provided in the Control Room and/or locally near the equipment.
Control Room indication for each train includes direct pool temperature, fuel pool water temperature out of the heat exchangers, pump run lights, pump discharge pressures, service water flow, SWS to ESS heat exchanger DP and valve position lights. Local indication inclunes fuel pool water temperature into the heat exchangers, pump discharge pressures, and heat exchanger DP.
Pool temperature is provided by redundant thermocouples located within the pool. All other transmitters and sensors are located in or near the Fuel Pool Cooling System cubicle.
Control for the two pumps and four MOVs is provided in the Control Room.
Control Room controls include pump on/off switches, service water throttle valves control switches, and Normal Fuel Pool Cooling Subsystem isolation valves control switches. These remote controls and instrumentation are provided to detect and control pump operation, pool temperature, and system flow, thereby ensuring operability of the Emergency Standby Subsystem of the FPCSD.
Normal Fuel Pool Cooling Subsystem The Normal Fuel Pool Cooling Subsystem (NFPCS) is shown in Figure A-1. The system cools the fuel storage pool by transferring the spent fuel decay heat (see Table A.2) through heat exchanger (s) to the Reactor Building Closed Cooling Water System. Water purity and clarity in the storage pool, reactor well, and dryer-separator storage pit are inaintained by filtering and A-6
'..^. .. .
, VYNPS demineralfzing the pool water through filter-demineralizer(s), which is shown in Figure A-2.
The system consists of two circulating pumps connected in parallel, two heat
'exchangers, two filter-demineralizers, and the required piping, valves and instrumentation. Each pump has a design capacity equal to a filter-demineralizer design flow rate'(450 gpm) and is capable of simultaneous operation. Two filter-demineralizers are provided. The pumps circulate the pool water in a closed loop, taking steticn from the spent fuel storage pool, circulating the water through the heat exchanger (s) and filter demineralizer(s), and returning it to the fuel pool and reactor well.
The fuel pool. filter demineralizers are located in the Radwaste Building.
The pools (dryer-separator storage pit, reactor well, and fuel storage pool) are filled from the Condensate Transfer System. Make-up to the pools is supplied by the Condensate Transfer System or the Demineralized Water System.
Water is removed from the pools via the fuel pool pumps through the fuel pool filter-demineralizer units to the condensate storage tank.
Fuel pool water is continuously recirculated except during the temporary periods when the reactor well and dryer-separator pit are being drained. The Normal Fuel Pool Cooling-Subsystem is capable of removing the decay heat load of the normal discharge batch of spent fuel with sufficient decay heat reduction. The Emergency Standby Subsystem can be used in lieu of the Normal Fuel Pool Cooling Subsystem to increase pool cooling in the event that a larger than normal amount of fuel is discharged into the pool or the normal fuel pool cooling heat transfer capacity is exceeded. During refueling, when the reactor well is flooded and the gates between the well and the pool are j removed, the RHR System is also available to cool the fuel pool in concert with reactor vessel core cooling. The RHR System has more than enough capacity to cool the reactor vessel core plus the entire inventory in the spent fuel pool.
I A-7 l
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. ,* 3 VYNPS
-Two small skimmer pumps are provided which take suction from the top of the pool to remove surface debris, These pumps pump this water through cartridge filters then back to the pool through the service boxes located around the pools.
Pool water clarity and purity are maintained by a combination of filtering and ion exchange processes. The filter-demineralizer maintains total heavy element content (Cu, Ni, Fe, Hg, etc.) at 0.1 ppm or less, with a pH range of 5.8 to 8.0 for compatibility with the fuel racks and other equipment.
Particulate material is removed from the circulated water by the pressure precoat filter-demineralizer unit in which a finely divided disposable filter medium is supported on permanent filter elements. The filter medium is replaced when the pressure drop is excessive or the ion exchange resin is depleted. Backwashing and precoating operations are manually controlled from the Radwaste Building. The spent filter medium is flushed from the elements and transferred to the condensate phase separator tanks by backwashing with air and condensate. The new filter medium is mixed in a precoat tank and transferred as a slurry by a precoat pump to the filter where the solids d.eposit on the filter elements. The holding pump maintains circulation through the filter in the interval between the precoating operation and the return to normal system operation to hold the precoat on the elements. The pump starts automatically on loss of system flow to maintain sufficient flow through the filter media to retain it on the filter elements.
A post-strainer is provided in the effluent stream of the filter-demineralizer to limit the migration of the filter material. The filter holding element is capable of withstandir.g a differential pressure greater than the developed pump head for the system. The maximum pressure drop across the filter and f_
associated process valves ano piping should not exceed 25 psid at the time of
( filter media replacement. The Backwash System is used to completely remove l
L resins and accumulated sludge from the filter demineralizers with a minimum i
volume of water. Backwash slurry drains to a phase separator. The Precoat l-System is designed to rapidly apply a uniform precoat of filter medir 'o the.
holding elements of a filter demineralizer. One centrifugal precoat pump and associated piping and valves are provided to precoat either A-8 1
l l
VYNPS filter-demineralizer and recirculate the water to the precoat tank or suction side of the precoat pump. The filter-demineralizer units are located separately in shielded rooms. Each room contains only the filter-demineralizer ar.d piping. All inlet, outlet, recycle, vent, drain, and other valves are located on the outside of one shielding wall of the room,
.together with necessary piping and headers, instrument elements, and coatrols. Penetrations through shielding walls are located so as not to
, compromise radiation shielding requirements.
The fuel pool filter-demineralizers are also used to process liquid radioactive wastes. See Chapter 9 of the Vermont Yankee FSAR for details.
The system instrumentation is provided for both automatic and remote manual operations. Instrumentation and controls are provided to detect, control and record pump operation, pool temperature, and system flow. A pool Leak Detection System has been provided to monitor leakage and thus indicate pool
. integrity.
The pumps can be controlled locally or at Panel 20-22 in the Radwaste Control Room. Pump low suction pressure automatically trips the pumps. A pump low discharge pressure alarm indicates in the Radwaste Control Room and a common trouble alarm in the Main Control Room.
The flow rate through each of the filter-demineralizers is indicated by a flow indicator on the Pump Room panel and in the Radwaste Control Room. The flow indicators can be seen by the operators from the vicinity of the Fuel Pool Cooling System control valves.
A high rate of leakage through the refueling bellows assembly, drywell to l
reactor seal, or the fuel pool gates is indicated by lights on the operating l- floor instrument racks and is alarmed in the Main Control Room.
The filter-demineralizers are controlled from a local panel in the Radwaste Building. Differential pressure and conductivity instrumentation are provided j for each filter-demineralizer unit to indicate when backwash is required.
A-9
r=
u . .
VYNPS Suitable alarms, differential pressure indicators, and flow indicators are provided to monitor.the condition of the filter-demineralizers.
A.6 Safety Evaluation Maximum normal heat load in the pool will be the sum of the heat from all previous batches plus that just discharged from the current refueling. The Normal Fuel Pool Cooling Subsystem of the Fuel Pool Cooling and Demineralizer System is used normally to maintain the pool water temperature below administrative limits during refuelings and plant operation. The Emergency
-Standby Subsystem is available to provide additional cooling, if needed, to ensure that the pool temperature does not exceed 150 F.
Maximum possible heat load would be the sum of the heat from all previous batches plus the heat from a full core discharge. If such a situation arose, the Emergency Standby Subsystem would be used to provide the cooling capacity needed under these conditions, or other high heat load conditions, to maintain the pool water temperature less than 150 F. Also, as an additional means of cooling the spent fuel pool during refueling operations, when the fuel pool and the refueling cavity are connected and filled with water, the Residual Heat Removal (RHR) System can be utilized to provide concurrent cooling to the core and spent fuel pool by circulating the water from the core to the pool and back to the core. In this mode, the RHR System will be in operation providing cooling to the core and can be shif ted to provide concurrent reactor core and spent fuel pool cooling. The RHR System has more than enough capacity to cool both the reactor core and the entire inventory of stored spent fuel in the spent fuel pool.
The Emergency Standby Subsystem is designed to provide pool cooling under all licensed plant conditions. This portion of the system is designed as Seismic Class I using the Seismic Class I Service Water System to remove spent fuel decay heat to the ultimate heat sink (Connecticut River). Essential electrical components in this portion of the system are also environmentally qualified to ensure operability under design basis accident conditions. In addition, the equipment is located in such a manner as to prevent common mode A-10
4 VYNPS failure from fire, flooding, or missiles. Providing sufficient pool cooling and environmental qualification, assures that the spent fuel will be cooled and boiling will not occur in the spent fuel pool. Therefore, the Reactor Building environment will not be subject to the consequences of a boiling spent fuel pool.
Leakage of potentially radioactive water from the Emergency Standby Subsystem through the heat exchanger into the Service Water System is prevented by providing a higher service water pressure than the Emergency Standby Subsystem pressure. This differential pressure ensures that leakage, if any, will go into the pool. Indication of this differential pressure is provided in the Control Room along with the controls for initiating the emergency standby portion of the system.
Leakage of the potentially radioactive water from the Normal Fuel Pool Cooling Subsystem to the Service Water System is prevented by using an intermediate closed loop cooling system, Reactor Building Closed Cooling Water (RBCCW),
which transfers the heat from the Normal Fuel Pool Cooling Subsystem to the Service Water System. This Closed Loop System arrangement ensures that fuel pool water leakage, if any, is contained within the RBCCW System and not released into the Service Water System.
(
l The normal fuel pool cooling flow rate is designed to be larger than that
[
required of two complete water changes per day of the fuel pool, or one change
( per day of the fuel pool, reactor well, and dryer-separator pit. The Emergency Standby Subsystem flow rate (700 gpm) is approximately 50% greater l
l than the normal fuel pool cooling flow rate (450 gpm). The maximum Normal Fuel Pool Cooling Subsystem flow rate is twice the flow rate needed to maintain the specified water quality.
An analysis has been made to determine the consequences of dropping a fully l
l loaded spent fuel shipping cask into the fuel storage pool. The results of that analysis showed that the bottom of the pool would lose its water-tight integrity, thereby making it difficult to maintain adequate shielding and cooling of the stored spent fuel. To prevent any load-drop occurrence, the l
A-11 c -
VYNPS Reactor Building crane is designed to be single-failureproof. (See Section 12.2.2.2. of the Vermont Yankee FSAR)
A.7 Inspection and Testing No special tests are required of the Normal Fuel Pool Cooling Subsystem because at least one pump, heat exchanger, and filter-demineralizer are normally in operation while fuel is stored in the pool. Redundant units are operated periodically to handle abnormal heat loads or to replace a unit for servicing. Routine visual inspection of the system components, pumps, heat exchangers, instrumentation, and trouble alarms are adequate to verify system operability.
The redundant units of the Emergency Standby Subsystem are periodically operated to ensure that the active components of the subsystem can isolate and provide pool cooling by remote manual initiation. Routine visual inspections of th'e system components, pumps, heat exchangers, instrumentation, and alarms are adequate to verify system operability.
l A-12
VYNPS TABLE A.1 FUEL POOL COOLING AND DEMINERALIZER SYSTEM -
SYSTEM SPECIFICATIONS System Function System Specification Normal Fuel Pool Cooling Subsystem Total pool, well, and pit volume 81,500 ft3 Fuel storage pool volume 41,600 f t3 System design flow 450 gpm Maximum flow 900 gpm Pump characteristics 450 gpm, 225 feet TDH, 25 feet NPSH Heat exchanger - Capacity each 2.23 x 106 Btu / hour, FPC temperature 1250F, RBCCW temperature 1000F, RBCCW flow 350 gpm Filter-demineralizer 267 square feet, 450 gpm, 25 psi maximum differential pressure (dirty)
Emergency Standby Subsystem System design flow 700 gpm Maximum flow 1400 gpm g Pump characteristics 700 gpm, 80 feet TDH, 24 feet NPSH i
Heat exchanger - Capacity each 11.0.x 106 Btu / hour, FPC temperature 1500F, SW temperature 900F, SW flow 700 gpm A-13 t
. .9 -
VYNPS' TABLE A.2
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FUEL DECAY HEAT (ESTIMATED), ArfER 0PERATION OF 18 MONTHS
- NORMAL REFUELING, 136 ASSEMBLIES DISCHARGED
- FULL CORE DISCHARGE, 368 ASSEMBLIES DISCHARGED-Degay Heat (10 Btu /hr)
Normal Refueling Discharge Full Core Discharge Number of 6 Days 10 Days Number of 10. Days Cycle Bundles After After Bundles After Discharged In Pool Shutdoen Shutdown In Pool Shutdown 13 (1) 1,586 8.75 7.59 1,818 16.84 14 1,722 9.00 7.79 1,954 17.18 15 1,858 9.18 7.96 2,090 17.37 16 ' 1,994 9.35 8.12 2,226 17.53 17 2,130 9.50 8.28 2,362 17.69 18 2,266 9.65 8.42 2,498 17.84 19 2,402 9.80 8.57 2,634 17.99 20 (2) 2,538 9.94 8.71 2,770 18.13 21 2,674 10.07 8.84 2,906 (3) 18.26 22 2,810 10.20 8.97 N/A 23 2,946 (3) 10.33 9.10 N/A NOTE: The decay heat from the previous cycle discharges is included in the above-estimated heat loads.
- 1. Vermont Yankee is currently in Cycle 13; estimated shutdown is 2/1989.
- 2. Loss of full core reserve discharge capability.
- 3. Excee 's capacity of reracked fuel pool.
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VERMONT YANKEE NUCLEAR POWER STATION l FINAL SAFETY ANALYSIS REPORT
, M Fuel Pool Filter Domineralizer System Figure A -2
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