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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 ML17241A5001999-10-21021 October 1999 Forwards Rev 3 to Emergency Response Data Sys (ERDS) Data Point Library for St Lucie Unit 1.Rev Provides Replacement Pages & Follows Format Recommended by NUREG 1394, ERDS Implementation, Rev 1,App C ML17309A9981999-10-19019 October 1999 Forwards Revised Epips,Including Rev 3 to EPIP-10 & Rev 25 to HP-202.EPIP-10 Added Onsite Monitoring Points,Made Administrative Changes & Incorporated New Attachments & HP-202 Added Red Team Survey Points 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 ML20217F6171999-10-0808 October 1999 Forwards Insp Repts 50-335/99-11 & 50-389/99-11 on 990827 & 990907-09.No Violations Identified.Matl Encl Contained Safeguards Info as Defined by 10CFR73.21 & Disclosed to Unauthorized Individuals Prohibited by Section 147 of AEA 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 ML17241A4811999-10-0101 October 1999 Reports Number of Tubes Plugged During Unit 1 Refueling Outage SL1-16,per TS 4.4.5.5.a 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 ML20212M1601999-09-28028 September 1999 Refers to 990908 Engineering Meeting Conducted at NRC Region II to Discuss Engineering Issues at Lucie & Turkey Point Facilities.List of Attendees & Copy of Presentation Handout Encl ML17241A4701999-09-25025 September 1999 Forwards Info Requested by NRC Staff During 990916 Telcon to Complete Staff Review of Request for risk-informed Extension of Action Completion/Aot Specified for Inoperable Train of LPSI Sys at Plant ML17241A4721999-09-24024 September 1999 Forwards Rev 1 to Plant Change/Mod (PCM) 99016 to St Lucie Unit 1,Cycle 16 COLR, IAW TS 6.9.1.11.d.Refueling Overhaul Activities Are Currently in Progress & Reactor Operations for Cycle 16 Are Scheduled to Commence in Oct 1999 ML17241A4681999-09-22022 September 1999 Requests Restriction Be Added to Senior Operator License SOP-21093 for TE Bolander.Nrc Forms 369,encl.Encl Withheld Per 10CFR2.790(a)(6) 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-99-113, 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 ML17241A4671999-09-20020 September 1999 Forwards Completed NRC Form 536, Operator Licensing Exam Data, for St Lucie Units 1 & 2,as Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams. 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 ML20212B1681999-09-13013 September 1999 Forwards Insp Repts 50-275/99-12 & 50-323/99-12 on 990711- 08-21.Four Violations Being Treated as Noncited Violations ML17241A4581999-09-13013 September 1999 Forwards Info Requested by NRC Staff During 990630 & 0816 Telcons,To Complete Review of Proposed License Amend for Fuel Reload Process Improvement Program ML17241A4521999-08-31031 August 1999 Withdraws Relief Request 16 & Suppl Relief Request 15 with Info Requested During 990526 Telephone Conference Re ISI Insp Plan,Third 10-yr Interval ML17241A4531999-08-31031 August 1999 Informs That No Candidates from St Lucie Plant Will Be Participating in PWR Gfes Being Administered on 991006 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-99-110, 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 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 ML17241A4501999-08-26026 August 1999 Informs That FPL Has Reviewed Reactor Vessel Integrity Database,Called RVID2,re Closure of GL 92-01,rev 1,suppl 1. Requested Corrections & Marked Up Pages from Rvid 2 Database Summary Repts That Correspond to Comments,Attached 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 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 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-99-103, 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 ML17241A4371999-08-13013 August 1999 Forwards fitness-for-duty Program Performance Data for six- Month Period Ending 990630,per 10CFR26.71(d) ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl ML17241A4461999-08-11011 August 1999 Requests That W Rept Entitled, Evaluation of Turbine Missile Ejection Probability Resulting from Extending Test Interval of Interceptor & Reheat Stop Valves at St Lucie Units 1 & 2, Be Withheld from Public Disclosure ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams 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 L-99-171, Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)1999-07-29029 July 1999 Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3) 1999-09-30
[Table view] Category:OUTGOING CORRESPONDENCE
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 ML20217F6171999-10-0808 October 1999 Forwards Insp Repts 50-335/99-11 & 50-389/99-11 on 990827 & 990907-09.No Violations Identified.Matl Encl Contained Safeguards Info as Defined by 10CFR73.21 & Disclosed to Unauthorized Individuals Prohibited by Section 147 of AEA 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 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 ML20212M1601999-09-28028 September 1999 Refers to 990908 Engineering Meeting Conducted at NRC Region II to Discuss Engineering Issues at Lucie & Turkey Point Facilities.List of Attendees & Copy of Presentation Handout Encl 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 ML20212B1681999-09-13013 September 1999 Forwards Insp Repts 50-275/99-12 & 50-323/99-12 on 990711- 08-21.Four Violations Being Treated as Noncited Violations ML20216F3171999-09-13013 September 1999 Forwards Insp Rept 50-271/99-06 on 990621-0801.One Violation Identified & Being Treated as Noncited Violation 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 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 ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams ML20210H6181999-07-27027 July 1999 Forwards Insp Repts 50-275/99-07 & 50-323/99-07 on 990503- 0714.Apparent Violations Being Considered for Escalated Enforcement Action ML20216D7321999-07-26026 July 1999 Forwards Insp Rept 50-271/99-05 on 990510-0620.Two Viiolations Being Treated as Noncited Violations ML20209G6931999-07-14014 July 1999 Forwards Request for Addl Info Re Spent Fuel Storage Capacity Expansion 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 ML20209F1541999-07-0606 July 1999 Informs That NRC in Process of Conducting Operational Safeguards Response Evaluations at Nuclear Power Reactors. Plant Chosen for Such Review Scheduled for Wk of 990823-26 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 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 ML20195J7221999-06-14014 June 1999 Forwards Insp Rept 50-271/99-03 on 990329-0509.Two Severity Level IV Violations Identified & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy ML20195F3871999-06-11011 June 1999 Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5) IA-99-247, Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5)1999-06-11011 June 1999 Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5) ML20195E1441999-06-10010 June 1999 Ack Receipt of Correspondence to NRC Commissioners Re Vermont Yankee Nuclear Power Station.Correspondence Forwarded to Staff for Appropriate Action ML20196J3001999-06-0404 June 1999 Informs That NRR Reorganized,Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Was Created. Reorganization Chart Encl ML20207G0921999-06-0404 June 1999 Forwards Insp Rept 50-271/99-04 on 990426-28.No Violations Noted.Insp Evaluated Performance of Emergency Response Organization During 990427,Vermont Yankee Nuclear Power Station full-participation Exercise ML20207E6001999-05-28028 May 1999 Forwards Operator Licensing Exam Rept 50-271/99-302 on 990510-11.Exam Addressed Areas Important to Public Health & Safety.Exam Developed & Administered Using Guidelines of NUREG-1021,Interim Rev 8.Both Applicants Passed Exam ML20207B8201999-05-25025 May 1999 Informs Licensee of Individual Exam Results for Applicants on Initial & Retake Exams Conducted on 990510-11 at Licensee Facility.Without Encls ML20207C7531999-05-17017 May 1999 Discusses Issue Identified by FPL in Feb 1998 Involving Potential for Fire to Cause Breach of Rc Sys High/Low Pressure Interface Boundary & NRC Decision for Exercise of Enforcement Discretion ML20206K1481999-05-0606 May 1999 Forwards Insp Rept 50-271/99-02 on 990215-0328.Three Severity Level IV Violations of NRC Requirements Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy ML20206J9951999-05-0505 May 1999 Informs That Util Authorized to Administer Initial Written Exams to Applicants Listed on 990510.Region I Operator Licensing Staff Will Administer Operating Test ML20206G6391999-05-0404 May 1999 Informs That Version of Holtec Intl Rept HI-981932 Marked as Proprietary Submitted by Util Will Be Withheld from Pubic Disclosure Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954,as Amended ML20206E8641999-04-29029 April 1999 Forwards SER Concluding That Flaw Evaluation Meets Rules of ASME Code Concerning Util 990329 Request to Extend Reinspection Period for Jet Pump Riser Circumferential Weld Flaws Discovered During 1998 Refueling Outage ML20206A6871999-04-22022 April 1999 Informs of Completion of Review of Re Nepco in Capacity as Minority Shareholder in Vermont Yankee Nuclear Power Corp,Yaec,Myap & Connecticut Yankee Atomic Power Co ML20205M0431999-04-13013 April 1999 Eighth Partial Response to FOIA Request for Records.App Q & R Records Encl & Being Made Available in PDR ML20205P1551999-04-0909 April 1999 Discusses 990225 PPR & Forwards Plant Issues Matrix & Insp Plan.Insp Program Subject to Rev ML20205K7461999-04-0808 April 1999 Advises That Info Contained in Holtec Intl Affidavit, Will Be Withheld from Public Disclosure,Per 10CFR2.790 ML20205J3381999-04-0808 April 1999 Informs That Time Provided by NRC Regulation within Which Commission May Act to Review Director'S Decision Expired. Commission Declined Any Review & Became Final Agency Action on 990406.With Certificate of Svc.Served on 990409 ML20205K7531999-04-0707 April 1999 Discusses Alternative Proposal for Reexamination of Circumferential Welds in Plant Rpv.Nrc Has Determined That Alternative Proposal Meets Conditions in BWRVIP-05 Rept. Forwards Safety Evaluation ML20205J0331999-03-31031 March 1999 Informs That on 980423 NRC Oi,Region I Field Ofc,Initiated an Investigation to Determine Whether Williams Power Corp Employee,Working at Vynp,Had Been Threatened & Eventually Fired in April 1998 ML20205C3701999-03-29029 March 1999 Final Response to FOIA Request for Documents.App a Documents Encl & Available in Pdr.App B Records Being Completely Withheld (Ref FOIA Exemption 5) ML20204J4321999-03-19019 March 1999 Forwards from Ve Quinn to Cl Miller Forwarding IEAL-R/85-11, Vermont Yankee Nuclear Power Station Site- Specific Offsite Radiological Emergency Preparedness Alert & Notification Sys QA Verification, for Info ML20204D7481999-03-16016 March 1999 Forwards Insp Rept 50-271/99-01 on 990104-0214.No Violations Noted.Security Program Insp Found That Licensee Implementing Security Program That Effectively Protects Against Acts of Radiological Sabotage DD-99-04, Informs That Time Provided by NRC Regulation within Which Commission May Act to Review Director'S Decision (DD-99-04) Has Expired.Commission Declined Review.Decision Became Final Action on 990308.With Certificate of Svc.Served on 9903121999-03-11011 March 1999 Informs That Time Provided by NRC Regulation within Which Commission May Act to Review Director'S Decision (DD-99-04) Has Expired.Commission Declined Review.Decision Became Final Action on 990308.With Certificate of Svc.Served on 990312 IR 05000271/19980141999-03-0101 March 1999 Forwards Request for Addl Info Based on Review of Util 961115 & 970313 Responses to GL 96-05 & Insp Rept 50-271/98-14 of GL 96-05 Program at Vermont Yankee Nuclear Power Station 1999-09-30
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. Soe8640M k **/ April 15, 1997
! Mr. Paul M. Blanch 135 Hyde Road West Hartford, CT 06117
Dear Mr. Blanch:
i I as writing to respond to the technical concerns described in your letters
- dated January 3 and January 13, 1997, regarding a loss of spent fuel pool i (SFP) water inventory. I have forwarded your concerns about statements by Mr. Wayne Lanning and other NRC officials to the NRC Inspector General for
, disposition.
Regarding your request for additional analyses of postulated SFP events i specific to the Millstone site, I believe that the staff's evaluations of SFP
- safety have been focused, methodical, and well documented. Consequently, I i
believe that a careful reading of the staff's reviews will highlight that the i staff has thoroughly considered the issues you raised. The following i paragraphs describe my basis for this statement.
I As you observed in your January 3, 1997 letter, a purpose of the NUREG-1353 study (NUREG-1353, " Regulatory Analysis for the Resolution of Generic Issue 82 l [GI-82), 'Beyond Design Basis Accidents in Spent Fuel Pools'") was to evaluate the level of safety associated .with high density storage of irradiated fuel and determine if the cost of any proposed enhancements to fuel storage facilities are commensurate with the increase in the level of safety provided l by the modification. Many of the concerns you have expressed involve assumptions made during the performance of the regulatory analysis and how the
! results of the analysis fit into the NRC's regulatory process.
I Each SFP and related system at licensed reactors has an associated design i
basis that the NRC staff reviewed and approved during licensing proceedings.
The design basis consists of the information that identifies 'the function of 1
! the structure or system and the ranges of values for important parameters for j
- which the structure or system is designed to perform its function. As used in -
! the title to NUREG-1353, the term "beyond design basis accidents" refers to
, classes or magnitudes of postulated events that were not used in the design
, process to define the function of structures or systems nor to establish .the
- ranges of values for important parameters. Exclusionofeventsinacertjin class of or a certain magnitude from the design process is not necessarily- '
, indicative of their likelihood relative to other events, such as large break l loss-of-coolant accidents (LOCAs), that were considered during plant design.
j However, the Commission must ensure that the integrated capability that 2
evolves from the events used in the design process provides adequate
- protection for public health and safety.
- With regard to the statement about the reporting requirements of 10 CFR 50.72
- in your letter dated January 3,1997, the reporting requirements mentioned
- (i.e., 10 CFR 50.72(b)(ii)(A) and (B)) apply following the actua7 occurrence l of an event or condition that places the plant in an unanalyzed condition or a s l
9704250019 970415 PDR ADOCK 05000245 NRC fnE CEnlTER 0
~
G PM
(
P. Blanch condition outside the design basis of the plant. The postulated occurrence of events or conditions involving functional capabilities greater or different than those considered in the design process or assumed parameter values outside the range of values used in the design process do not constitute reportable events or conditions pursuant to 10 CFR 50.72. However, Northeast Nuclear Energy, the licensee for the Millstone reactors, has submitted several reports pursuant to 10 CFR 50.72 and 50.73 about actual events or conditions related to the SFP that the licensee believed placed the plant in an unanalyzed condition or a condition outside the design basis of the plant (e.g., piping having the function of retaining pressure following exposure to the stress values imposed by a seismic event had not been analyzed for that capability, and irradiated fuel was transferred to the SFP with a decay time that was outside of the range of values used in the design evaluation of the SFP cooling system).
Nevertheless, analyses of postulated event sequences serve an important purpose in the regulatory process. Through these analyses, event sequences can be evaluated for their level of risk. Where the risk is significant, the NRC employs the backfit process of 10 CFR 50.109 to modify the design or operation of a nuclear power plant. The evaluation documented in NUREG-1353 determined that the level of risk from beyond design basis SFP accidents was low, but the evaluation also examined the potential of specific enhancements to improve the level of safety.
A number of SFP event analyses were documented in NUREG-1353 to support resolution of GI-82. Events evaluated included seismic events of a magnitude greater than that used in the design process for the SFP structure, pneumatic seal failures, inadvertent drain down due to pipe breaks or system misalignments, and extended loss of cooling and makeup events. The analyses documented in NUREG-1353 estimated the probable radiological consequences of these events and the frequency with which the event would progress to the state producing the estimated radiological consequence.
In response to your concerns about adequate consideration of full core off-loads, you should note that the staff did consider the size and timing of the fuel off-load for events where the frequency of occurrence was strongly influenced by these factors. For example, the evaluation explicitly considered the impact of a full core off-load occurring 5 days after reactor shutdown for extended lo:s of pool cooling and makeup sequences because the size and timing of the off-load have an effect on the probability of recovery.
Conversely, the size and timing of fuel off-load has negligible influence on the progression of events initiated by large-magnitude seismic and other low-probability external events. For example, SFP structural failures due to such events as missile strikes and aircraft accidents were assumed to progress to rapid cladding oxidation regardless of the :in and timing of the most recent off-load because the event was assumed to preclude coolant addition to the SFP. Because of the independence of the progression of externally initiated events from the size and timing of the m st recent off-load, the staff believes that the use of a more probable radioneclide inventory (e.g., a l
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P. Blanch !
one-third core discharge at 90 days decay) rather than an extreme inventory (e.g., a full core discharge with a short decay time) provides a more accurate
! assessment of risk.
With regard to the consequences of the various events analyzed, the staff calculated best-estimate and worst-case consequences for events progressing to rapid cladding oxidation. The best-estimate case assumed that one third of a -
core ignited as a result of the particular event sequence. The best-estimate case calculated radioisotope ' inventory based on the one-third core being placed in the SFP 90 days prior to the event occurring. Finally, as you observed, the best-estimate case assumed a typical population density of 340 people per square mile.
In contrast, the worst-case consequence estimate assumed that (1) the event
- progression involved the igniticn of the entire inventory of stored spent fuel (a full fuel pool consisting of multiple cores stored over the hfe of the 4
plant), (2) the radioisotope inventery included the inventory of he niost
- recent one-third core discharge with 30 days decay plus all previou s discharges, and (3) the assumed population density was 860 people per square mile, which was based on the Zion facility - a high population der sity site.
The staff does not believe further analysis of the offsite consequences considering the radioisotope inventory of a recently off-loaded full core is warranted given the highly conservative nature of the existing worst-case estimate.
i It is also important to understand that the worst-case consequence estimates i were carried over into the value impact analysis. The value impact portion of ;
NUREG-1353 quantified cost and benefits for two alternatives that were judged l likely to mitigate the sequences that resulted in spect "uel fires (i.e., the l 1 use of low density racks or increased use of dry storage). Both the best- l estimate and the worst-case consequence estimates were factored into these cost-benefit evaluations. The staff also quantified costs and benefits for improvements to SFP cooling systems. Those cooling system modifications l considered were judged likely to have only limited impact on averted offsite dose because the modeling of loss of pool cooling events did not support the assumption of rapid cladding oxidation. Instead, the analyses of loss of pool i cooling events assumed cladding failure, which does not result in significant offsite consequences even when the decay time is as short as a few days.
i The bases for your assertions that "the probabilities and consequences of this accident have increased significantly" and that "the risk is now 100 times i
that previously assumed" are not clear. However, it appears that they are attempts to reconcile the results of recent staff evaluations of SFP safety with the results and conclusions in NUREG-l'13.
Note that the postulated loss of spent fuel water inventory events quantified i in NUREG-1353 were total loss of water events that resulted in ignition of the stored spent fuel. NUREG-1353 specifically stated that "... spent fuel pools are designed to preclude significant (a few inches) fuel mcovery due to the i
l P. B1anch leakage..." as a result of seal failures, pipe breaks, and other leakage events considered during plant design. The analytical models used to support the analyses in NUREG-1353 indicated that near total fuel uncovery (i.e.,
water level more than 12 feet below the top of the fuel) would be necessary for the significant fuel damage and ignition that formed the basis of the severe consequences quantified in NUREG-1353.
As part of the recent Task Action Plan for Spent Fuel Storage Pool Safety (Action Plan), the staff examined the features incorporated in the design of every operating reactor SFP to prevent coolant loss. The staff determined that all SFP designs had been reviewed and accepted by the NRC staff and that the designs provided adequate protection against loss of coolant events.
However, certain plants had features inconsistent with current SFP design guidance that could be enhanced to further reduce the probability of significant coolant inventory loss. As described in the staff's report to the Commission on the Action Plan, which was provided to you on September 11, 1996, the staff is performing regulatory analyses for those plants in order to determine if a substhntial increase in the level of safety at a justifiable cost can be achieved by modifying the SFP design or operation at those facilities. No enhancements to SFP design features were identified at the Millstone units with regard to reducing the probability of a significant coolant inventory loss, but Millstone Unit I was identified for potential enhancement of SFP temperature instrumentation if it is justified on a cost-benefit basis.
The staff study, AE0D/S96-02, which you cited in your letter, further quantified the probability of SFP leakage events. AE00/S96-02 did state that events involving a loss of SFP inventory greater than one foot have occurred at a rate of about one per 100 reactor years. The report notes that, as a result of human interaction, all actual events were terminated with approximately 20 feet of water remaining over the top of the fuel. As you correctly point out, an estimate for the frequency of loss of pool level of greater than one foot is in no way equivalent to an estimate of the frequency of a postulated loss of pool level down to the top of the stored fuel much less a total loss of inventory. Because NUREG-1353 accounts for human intervention to mitigate loss of inventory events due to seal failures or inadvertent drainage, there is no reason to believe that the estimates for these events in NUREG-1353 are inadequate or are inconsistent with the findings of AE00/S96-02. Thus, the requantification of the events in NUREG-1353 based on the findings of AE00/S96-02 is not warranted for Millstone or any other site.
The staff believes that the SFP safety studies conducted over the past few years, including NUREG-1353, the Task Action Plan for Spent Fuel Storage Pool Safety, and AE0D/S96-02, have provided significant insight into the relative risks posed by the storage of irradiated fuel in storage pools at the nations's power reactor facilities. These studies have identified the issues and facilities where specific design and operational improvements may be justified by reviewing previous analyses, such as NUREG-1353, and new information in a methodical manner. At the same time, these studies provide
I P. Blanch ,
s l
part of the justification for the staff's statement in Partial Director's l Decision DD-96-23 that the safety significance of certain full are off-load practices at Millstone 1 was low. The staff's November 9, 1995, safety
- evaluation supporting the license amendment related to the practice of full core off-loads was also used to develop DD-96-23.
I do not believe that performing a site-specific analysis of beyond design ,
basis SFP accidents, which would be necessary to answer many of the detailed l questions in your February 28, 1995, and January 3, 1997, letters, would yield i information that could be used by the staff to improve SFP safety. Rather, I believe that staff's efforts to complete the actions identified for resolution of the Action Plan represent the most effective means of ensuring accomplishment of the NRC's mission of protecting public health and safety with regard to SFPs. Additionally, I believe these actions represent a highly 1 responsive and responsible approach to the concerns you have raised.
On an administrative note, in your January 13, 1997, letter, you stated that your request was being made under the provisions of the Freedom of Information Act (F0IA). The existing staff analyses for spent fuel storage pool safety are documented in NUREG-1353, the July 26, 1996, report to the Commission on the Action Plan and in AEOD/S96-02. NUREG-1353 was provided to you during a drop-in meeting with me in March 1995. The Action Plan report was forwarded l to you on September 11, 1996. From your letters, it appears that you already have access to AE00/S96-02. As noted above, the staff's November 9, 1995, safety evaluation supporting the licensee's full core off-load amendment was also used to develop DD-96-23. This document is publicly available. There are no other " safety analyses" that substantiate the staff's conclusions, therefore, there is nothing to provide to you under F0IA that you do not already have or have access to.
If you have any additional questions on this matter, please do not hesitate to contact me.
Sincerely,
/S/
John A. Zwolinski, Deputy Director Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
- Previous Concurrence n [h ,
OFFICE PdhhA P bf-2/PM PDI-2/D NRR/SP0
- DRP NAME Mdhib Wh JStolz PMcKee GHolahan JZdok1hski DATE Nfhf97 Qho/97 Y/h/97 3/27/97 4/01/97 h/ Mf97 0FFICIAL' RECORD COPY FILENAME: G:\SHEA\ BLANCH.197
P. Blanch '
part of the justification for the staff's statement in Partial Director's Decision DD-96-23 that the safety significance of certain full core off-load practices at Hillstone I was low. The staff's November 9, 1995, safety evaluation supporting the license amendment related to the practice of full
, core off-loads was also used to develop DD-96-23.
I do not believe that performing a site-specific analysis of beyond design basis SFP accidents, which would be necessary to answer many of the detailed questions in your February 28, 1995, and January 3, 1997, letters, would yield information that could be used by the staff to improve SFP safety. Rather, I believe that staff's efforts to complete the actions identified for resolution of the Action Plan represent the most effective means of ensuring accomplishment of the NRC's mission of protecting public health and safety with regard to SFPs. Additionally, I believe these actions represent a highly responsive and responsible approach to the concerns you have raised.
On an administrative note, in your January 13, 1997, letter, you stated that your request was being made under the provisions of the Freedom of Information Act (F0IA). The existing staff analyses for spent fuel storage pool safety are documented in NUREG-1353, the July 26, 1996, report to the Commission on the Action Plan and in AE00/S96-02. NUREG-1353 was provided to you during a drop-in meeting with me in March 1995. The Action Plan report was forwarded to you on September 11, 1996. From your letters, it appears that you already have access to AE0D/S96-02. As noted above, the staff's November 9. 1995, safety evaluation supporting the licensee's full core off-load amendment was also used to develop DD-96-23. This document is publicly available. There are no other " safety analyses" that substantiate the staff's conclusions, therefore, there is nothing to provide to you under F0IA that you do not already have or have access to.
If you have any additional questions on this matter, please do not hesitate to contact me.
g Sincerely, John . Zwolinski, Deputy Director Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
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letter to P. Blanch from J. Zwolinski, dated April 15, 1997.
.DISTRIBUTIQN :
' Docket File (50-245) (w/ original incoming)
PUBLIC (w/ incoming)
PDI-2 Reading (w/ incoming) -
RZimmerman SVarga JZwolinski JStolz.
JShea (w/ incoming)
MThadani M8ayle (e-mail only MLB4)
M0'Brien NRR Mail Room (YT#970005 w/ incoming) (012/G/18)
N01 son CNorsworthy WLanning, RGN-I GHolahan' OGC ACRS WTravers PMcKee RHoefling JLee - '
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