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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217L0421999-10-21021 October 1999 Forwards Insp Rept 50-382/99-20 on 990815-0925 & Notice of Violation.Two Severity Level IV Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20217N2111999-10-19019 October 1999 Forwards Insp Rept 50-382/99-14 on 990913-17 & 1004-08.No Violations Noted.Licensed Operator Requalification Program, Effective,Utilized Systems Approach to Training & Showed Continued Improvements Over Previous Insp Findings ML20217L0101999-10-18018 October 1999 Provides Update of Waterford 3 Effort for Review of Ufsar. Info Listed Includes Background Mgt Expectations,Review Status & Results,Clarifications Re Review & Conclusions ML20217L0141999-10-18018 October 1999 Submits Update to NRC Staff Re Circumstances & Plans for Submitting Certification Rept on Waterford 3 Plant Specific Simulator ML20217G7051999-10-14014 October 1999 Forwards Comments on Four of NRC RO Examination Questions for Exam Administered During Week of 991004 05000382/LER-1999-014, Forwards LER 99-014-00,providing Details of Reactor Shutdown Due to Loss of RCP Controlled bleed-off Flow.Attached Commitment Identification/Voluntary Enhancement Form Identifies All Commitments Contained in Submittal1999-10-12012 October 1999 Forwards LER 99-014-00,providing Details of Reactor Shutdown Due to Loss of RCP Controlled bleed-off Flow.Attached Commitment Identification/Voluntary Enhancement Form Identifies All Commitments Contained in Submittal ML20217D5151999-10-0707 October 1999 Forwards Application for Renewal of SRO License for C Fugate License SOP-43039-3,IAW 10CFR55.57.Without Encls ML20217C6251999-10-0505 October 1999 Informs That NRC Reviewed Util Ltr & Encl Exercise Scenario Package for Waterford 3 Emergency Plan Exercise Scheduled for 991013.Based on Review,Nrc Determined That Exercise Appropriate to Meet Objectives ML20212J6921999-09-29029 September 1999 Forwards Insp Rept 50-382/99-18 on 990830-0902.One Noncited Violation Identified Re Failure to Follow Procedural Instructions to Ensure That Members on Fire Brigade Shift Were Qualified ML20216G2441999-09-27027 September 1999 Forwards Insp Rept 50-382/99-19 on 990830-0903.No Violations Noted 05000382/LER-1999-013, Forwards LER 99-013-00,providing Details of Exceeding TS Limits for RCS Cooldown Rates.All Commitments Contained in Submittal Are Identified on Encl Commitment Identification/ Voluntary Enhancement Form1999-09-23023 September 1999 Forwards LER 99-013-00,providing Details of Exceeding TS Limits for RCS Cooldown Rates.All Commitments Contained in Submittal Are Identified on Encl Commitment Identification/ Voluntary Enhancement Form IR 05000382/19993011999-09-21021 September 1999 Informs That NRC License Exam Previously Associated with NRC Insp Rept 50-382/99-301 Will Be Incorporated Into NRC Insp Rept 50-382/99-14 ML20212D8761999-09-16016 September 1999 Informs That on 990818,NRC Staff Completed Midcycle PPR of Waterford 3.During Assessment Period,Number of Personnel Errors Occurred,Which Demonstrated Lack of Attention to Detail by Plant Personnel.Historical Listing of Issues,Encl ML20212C2471999-09-16016 September 1999 Forwards Five Final Applications for RO Licenses for G Esquival,Jm Hearn,Md Lawson,Re Simpson & PI Wood.Written Exam & Operating Test to Be Administered,Is Requested. Encls Withheld ML20212C2391999-09-16016 September 1999 Requests Cancellation of SRO Licenses for Bn Coble,License SOP-43835,due to Job Assignment Location & CA Rodgers, License SOP-43537-1,due to Resignation from Company, Effective 990901 ML20212C5881999-09-14014 September 1999 Forwards Insp Rept 50-382/99-15 on 990719-23 with Continuing in Ofc Insp Until 0819.No Violations Noted ML20211Q4421999-09-0909 September 1999 Forwards Insp Rept 50-382/99-07 on 990601-11.Three Violations Being Treated as Noncited Violations ML20211P4121999-09-0707 September 1999 Requests NRC Staff Review & Approval of Integrated Nuclear Security Plan (Insp) & Integrated Security Training & Qualification Plan (Ist&Q), for Use by All Entergy Operations,Inc.Encl Withheld,Per 10CFR2.790(d) ML20211M8391999-09-0303 September 1999 Forwards Revised Epips,Including Rev 25 to EP-001-020,rev 24 to EP-001-030,rev 25 to EP-001-040,rev 30 to EP-002-100,rev 22 to EP-001-010,rev 27 to EP-002-010,rev 26 to EP-002-102 & Rev 16 to EP-002-190.Listed Proprietary Revs to Epips,Encl ML20211L3681999-09-0202 September 1999 Forwards Five Preliminary Applications for Reactor Operator Licenses for Individuals Listed,Iaw 10CFR55.31.Encls Withheld ML20211K9741999-09-0101 September 1999 Forwards Insp Rept 50-382/99-16 on 990704-0814.Two Severity Level IV Violations Identified & Being Treated as Noncited Violations,Consistent with App C of Enforcement Policy 05000382/LER-1999-011, Forwards LER 99-011-00,providing Details of Reactor Shutdown Due to Loss of Controlled bleed-off Flow.All Commitments Contained in Submittal Identified on Attached Commitment Identification/Voluntary Enhancement Form1999-08-31031 August 1999 Forwards LER 99-011-00,providing Details of Reactor Shutdown Due to Loss of Controlled bleed-off Flow.All Commitments Contained in Submittal Identified on Attached Commitment Identification/Voluntary Enhancement Form ML20211M3641999-08-30030 August 1999 Forwards Written Examination,Operating Tests & Supporting Ref Matl Identified in Attachment 2 of ES-210,in Response to NRC .Encl Withheld ML20211G5751999-08-27027 August 1999 Forwards RAI Re IPEEE Submittal.Please Provide RAI within 60 Days of Receipt of Ltr,Per Util Response to GL 88-20,suppl 4 ML20211E3281999-08-26026 August 1999 Forwards fitness-for-duty Performance Data for Period of 990101-0630,IAW 10CFR26.71(d).Ltr Does Not Contain Commitments 05000382/LER-1999-009, Forwards LER 99-009-00 Re Discovery of Condition of Noncompliance with App R Involving Inadequate Separation of Essential Cables Routed in Fire Area RAB-30 in Rab. Compensatory Measures Were Established Immediately1999-08-26026 August 1999 Forwards LER 99-009-00 Re Discovery of Condition of Noncompliance with App R Involving Inadequate Separation of Essential Cables Routed in Fire Area RAB-30 in Rab. Compensatory Measures Were Established Immediately 05000382/LER-1999-010, Forwards LER 99-010-00,providing Details of Inadequate Pumping Capacity in Dry Cooling Tower Area.All Commitments Contained in Submittal Are Identified on Attached Commitment Identification Voluntary Enhancement Form1999-08-26026 August 1999 Forwards LER 99-010-00,providing Details of Inadequate Pumping Capacity in Dry Cooling Tower Area.All Commitments Contained in Submittal Are Identified on Attached Commitment Identification Voluntary Enhancement Form ML20211F5421999-08-24024 August 1999 Forwards Proposed marked-up TS Page Xviii, Index Administrative Controls, Correcting Page Number Re TS Change Request NPF-38-220.Editorial Changes for TS Change NPF-38-221 Discussed ML20211F3561999-08-24024 August 1999 Forwards CTS Pages & TS Proposed marked-up Pages for Insertion Into TS Change Request NPF-38-207 Re Efas, Originally Submitted on 980702.Original NSHC Determination Continues to Be Applicable ML20211F4611999-08-24024 August 1999 Informs That NRC Reviewed Ltr & Encl Objectives for Waterford 3 Emergency Plan Exercise Scheduled for 991013.Exercise Objectives Appropriate to Meet Emergency Plan Requirements ML20211G1731999-08-23023 August 1999 Informs That Info Submitted in ,B&W Rept 51-1234900-00,will Be Withheld from Public Disclosure,Per 10CFR2.790 ML20211C5101999-08-19019 August 1999 Forwards Certified Copies of Liability Insurance Policy Endorsements Issued in First Half of 1999 for Each Entergy Operations,Inc Nuclear Unit,Per 10CFR140.15 ML20210T9791999-08-18018 August 1999 Discusses Which Responded to Reconsideration of Violation Denial (EA 98-022) Enforcement Action Detailed in .Concludes That Violation Occurred as Stated 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 ML20210S0561999-08-12012 August 1999 Submits Voluntary Response to NRC AL 99-02, Operating Reactor Licensing Action Estimates, for NRC Fys 2000 & 2001 for Waterford 3 ML20210Q6161999-08-12012 August 1999 Forwards Corrected Copy of Monthly Operating Rept for July 1999 for Waterford 3.Original Rept,Submitted with ,Contained Typos ML20217F2661999-08-12012 August 1999 Forwards Copy of 1999 Waterford 3 Biennial Exercise Package to Be Performed Using Waterford 3 CR Simulator ML20210R9231999-08-11011 August 1999 Forwards Insp Rept 50-382/99-10 on 990719-23.Violations Noted.Nrc Has Determined That One Severity Level IV Violation of NRC Requirements Occurred 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 05000382/LER-1999-008, Forwards LER 99-008-00,re Failure to Perform Testing of ESF Filtration Units Per TS Srs.Commitments Made by Util Also Encl1999-07-29029 July 1999 Forwards LER 99-008-00,re Failure to Perform Testing of ESF Filtration Units Per TS Srs.Commitments Made by Util Also Encl ML20210H4291999-07-29029 July 1999 Forwards Response to NRC Rai,Associated with TS Change Request NPF-38-208,proposing to Replace Ref to Supplement 1 with Ref to Supplement 2 of Calculative Methods for CE Small Break LOCA Evaluation Model, in ACs Section of TSs ML20210F9451999-07-27027 July 1999 Forwards Proprietary & non-proprietary Version of Rev 29 to EPIP EP-002-100, Technical Support Ctr Activation,Operation & Deactivation. Proprietary Info Withheld,Per 10CFR2.790 ML20210D3171999-07-23023 July 1999 Submits Proposal for Final Resolution of Reracking Spent Fuel Pool at Plant,Per License Amend 144,issued by NRC in .No New Commitments Are Contained in Ltr 05000382/LER-1999-007, Forwards LER 99-007-00,providing Details of Operation Outside Tornado Missile Protection Licensing Basis for turbine-driven Emergency Feedwater Pump Exhaust Stack & Steam Supply Piping.All Commitments Identified on Attached1999-07-23023 July 1999 Forwards LER 99-007-00,providing Details of Operation Outside Tornado Missile Protection Licensing Basis for turbine-driven Emergency Feedwater Pump Exhaust Stack & Steam Supply Piping.All Commitments Identified on Attached ML20210D8701999-07-23023 July 1999 Forwards Safety Evaluation Re First 10-yr Interval Inservice Insp Plan Requests for Relief ISI-018 Through ISI-020 for Entergy Operations,Inc,Unit 3 ML20210B1521999-07-15015 July 1999 Forwards Insp Rept 50-382/99-13 on 990523-0703.Three Violations Being Treated as Noncited Violations ML20209G9771999-07-13013 July 1999 Forwards Objectives & Guidelines for Waterford 3 Emergency Preparedness Exercise Scheduled for 991013.List of Objectives cross-referenced Where Applicable to Relevant Sections of NUREG-0654 IR 05000382/19990081999-07-12012 July 1999 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-382/99-08 Issued on 990503 ML20209E5231999-07-0909 July 1999 Informs That as Result of NRC Review of Util Responses to GL-92-01,rev 1 & Suppl 1,staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2.This Closes Staff Efforts Re TAC MA0583 ML20209D4051999-07-0707 July 1999 Forwards Revised TS Pages to Replace Attachment C,Entirely in Original TS Change Request NPF-38-207,per 990519 Discussion with C Patel of Nrc.Changes to Action 20 Delete Word Requirement & Revise Word Modes to Mode 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217L0101999-10-18018 October 1999 Provides Update of Waterford 3 Effort for Review of Ufsar. Info Listed Includes Background Mgt Expectations,Review Status & Results,Clarifications Re Review & Conclusions ML20217L0141999-10-18018 October 1999 Submits Update to NRC Staff Re Circumstances & Plans for Submitting Certification Rept on Waterford 3 Plant Specific Simulator ML20217G7051999-10-14014 October 1999 Forwards Comments on Four of NRC RO Examination Questions for Exam Administered During Week of 991004 05000382/LER-1999-014, Forwards LER 99-014-00,providing Details of Reactor Shutdown Due to Loss of RCP Controlled bleed-off Flow.Attached Commitment Identification/Voluntary Enhancement Form Identifies All Commitments Contained in Submittal1999-10-12012 October 1999 Forwards LER 99-014-00,providing Details of Reactor Shutdown Due to Loss of RCP Controlled bleed-off Flow.Attached Commitment Identification/Voluntary Enhancement Form Identifies All Commitments Contained in Submittal ML20217D5151999-10-0707 October 1999 Forwards Application for Renewal of SRO License for C Fugate License SOP-43039-3,IAW 10CFR55.57.Without Encls 05000382/LER-1999-013, Forwards LER 99-013-00,providing Details of Exceeding TS Limits for RCS Cooldown Rates.All Commitments Contained in Submittal Are Identified on Encl Commitment Identification/ Voluntary Enhancement Form1999-09-23023 September 1999 Forwards LER 99-013-00,providing Details of Exceeding TS Limits for RCS Cooldown Rates.All Commitments Contained in Submittal Are Identified on Encl Commitment Identification/ Voluntary Enhancement Form ML20212C2391999-09-16016 September 1999 Requests Cancellation of SRO Licenses for Bn Coble,License SOP-43835,due to Job Assignment Location & CA Rodgers, License SOP-43537-1,due to Resignation from Company, Effective 990901 ML20212C2471999-09-16016 September 1999 Forwards Five Final Applications for RO Licenses for G Esquival,Jm Hearn,Md Lawson,Re Simpson & PI Wood.Written Exam & Operating Test to Be Administered,Is Requested. Encls Withheld ML20211P4121999-09-0707 September 1999 Requests NRC Staff Review & Approval of Integrated Nuclear Security Plan (Insp) & Integrated Security Training & Qualification Plan (Ist&Q), for Use by All Entergy Operations,Inc.Encl Withheld,Per 10CFR2.790(d) ML20211M8391999-09-0303 September 1999 Forwards Revised Epips,Including Rev 25 to EP-001-020,rev 24 to EP-001-030,rev 25 to EP-001-040,rev 30 to EP-002-100,rev 22 to EP-001-010,rev 27 to EP-002-010,rev 26 to EP-002-102 & Rev 16 to EP-002-190.Listed Proprietary Revs to Epips,Encl ML20211L3681999-09-0202 September 1999 Forwards Five Preliminary Applications for Reactor Operator Licenses for Individuals Listed,Iaw 10CFR55.31.Encls Withheld 05000382/LER-1999-011, Forwards LER 99-011-00,providing Details of Reactor Shutdown Due to Loss of Controlled bleed-off Flow.All Commitments Contained in Submittal Identified on Attached Commitment Identification/Voluntary Enhancement Form1999-08-31031 August 1999 Forwards LER 99-011-00,providing Details of Reactor Shutdown Due to Loss of Controlled bleed-off Flow.All Commitments Contained in Submittal Identified on Attached Commitment Identification/Voluntary Enhancement Form ML20211M3641999-08-30030 August 1999 Forwards Written Examination,Operating Tests & Supporting Ref Matl Identified in Attachment 2 of ES-210,in Response to NRC .Encl Withheld ML20211E3281999-08-26026 August 1999 Forwards fitness-for-duty Performance Data for Period of 990101-0630,IAW 10CFR26.71(d).Ltr Does Not Contain Commitments 05000382/LER-1999-010, Forwards LER 99-010-00,providing Details of Inadequate Pumping Capacity in Dry Cooling Tower Area.All Commitments Contained in Submittal Are Identified on Attached Commitment Identification Voluntary Enhancement Form1999-08-26026 August 1999 Forwards LER 99-010-00,providing Details of Inadequate Pumping Capacity in Dry Cooling Tower Area.All Commitments Contained in Submittal Are Identified on Attached Commitment Identification Voluntary Enhancement Form 05000382/LER-1999-009, Forwards LER 99-009-00 Re Discovery of Condition of Noncompliance with App R Involving Inadequate Separation of Essential Cables Routed in Fire Area RAB-30 in Rab. Compensatory Measures Were Established Immediately1999-08-26026 August 1999 Forwards LER 99-009-00 Re Discovery of Condition of Noncompliance with App R Involving Inadequate Separation of Essential Cables Routed in Fire Area RAB-30 in Rab. Compensatory Measures Were Established Immediately ML20211F3561999-08-24024 August 1999 Forwards CTS Pages & TS Proposed marked-up Pages for Insertion Into TS Change Request NPF-38-207 Re Efas, Originally Submitted on 980702.Original NSHC Determination Continues to Be Applicable ML20211F5421999-08-24024 August 1999 Forwards Proposed marked-up TS Page Xviii, Index Administrative Controls, Correcting Page Number Re TS Change Request NPF-38-220.Editorial Changes for TS Change NPF-38-221 Discussed ML20211C5101999-08-19019 August 1999 Forwards Certified Copies of Liability Insurance Policy Endorsements Issued in First Half of 1999 for Each Entergy Operations,Inc Nuclear Unit,Per 10CFR140.15 ML20210Q6161999-08-12012 August 1999 Forwards Corrected Copy of Monthly Operating Rept for July 1999 for Waterford 3.Original Rept,Submitted with ,Contained Typos ML20210S0561999-08-12012 August 1999 Submits Voluntary Response to NRC AL 99-02, Operating Reactor Licensing Action Estimates, for NRC Fys 2000 & 2001 for Waterford 3 ML20217F2661999-08-12012 August 1999 Forwards Copy of 1999 Waterford 3 Biennial Exercise Package to Be Performed Using Waterford 3 CR Simulator 05000382/LER-1999-008, Forwards LER 99-008-00,re Failure to Perform Testing of ESF Filtration Units Per TS Srs.Commitments Made by Util Also Encl1999-07-29029 July 1999 Forwards LER 99-008-00,re Failure to Perform Testing of ESF Filtration Units Per TS Srs.Commitments Made by Util Also Encl ML20210H4291999-07-29029 July 1999 Forwards Response to NRC Rai,Associated with TS Change Request NPF-38-208,proposing to Replace Ref to Supplement 1 with Ref to Supplement 2 of Calculative Methods for CE Small Break LOCA Evaluation Model, in ACs Section of TSs ML20210F9451999-07-27027 July 1999 Forwards Proprietary & non-proprietary Version of Rev 29 to EPIP EP-002-100, Technical Support Ctr Activation,Operation & Deactivation. Proprietary Info Withheld,Per 10CFR2.790 ML20210D3171999-07-23023 July 1999 Submits Proposal for Final Resolution of Reracking Spent Fuel Pool at Plant,Per License Amend 144,issued by NRC in .No New Commitments Are Contained in Ltr 05000382/LER-1999-007, Forwards LER 99-007-00,providing Details of Operation Outside Tornado Missile Protection Licensing Basis for turbine-driven Emergency Feedwater Pump Exhaust Stack & Steam Supply Piping.All Commitments Identified on Attached1999-07-23023 July 1999 Forwards LER 99-007-00,providing Details of Operation Outside Tornado Missile Protection Licensing Basis for turbine-driven Emergency Feedwater Pump Exhaust Stack & Steam Supply Piping.All Commitments Identified on Attached ML20209G9771999-07-13013 July 1999 Forwards Objectives & Guidelines for Waterford 3 Emergency Preparedness Exercise Scheduled for 991013.List of Objectives cross-referenced Where Applicable to Relevant Sections of NUREG-0654 ML20209D4051999-07-0707 July 1999 Forwards Revised TS Pages to Replace Attachment C,Entirely in Original TS Change Request NPF-38-207,per 990519 Discussion with C Patel of Nrc.Changes to Action 20 Delete Word Requirement & Revise Word Modes to Mode ML20209B6081999-06-30030 June 1999 Submits Response to NRC GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants. Disclosure Encl 05000382/LER-1999-005, Forwards LER 99-005-00,providing Details of Discovery of Untested Electrical Contacts in safety-related Logic Circuits1999-06-24024 June 1999 Forwards LER 99-005-00,providing Details of Discovery of Untested Electrical Contacts in safety-related Logic Circuits ML20196G5731999-06-24024 June 1999 Forwards Operator Licensing Exam Outlines Associated with Exam Scheduled for Wk of 991004.Exam Development Is Being Performed in Accordance with NUREG-1021,Rev 8 ML20212J4121999-06-23023 June 1999 Responds to NRC Re Reconsideration of EA 98-022. Details Provided on Actions Util Has Taken or Plans to Take to Address NRC Concerns with Ability to Demonstrate Adequate Flow Availability to Meet Design Requirements ML20196E9371999-06-22022 June 1999 Forwards Revs Made to EP Training Procedures.Procedures NTC-217 & NTC-217 Have Been Deleted.Procedure NTP-203 Was Revised to Combine Requirement Previously Included in Procedures NRC-216 & NTC-217 ML20196A1021999-06-17017 June 1999 Provides Supplemental Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, Per 990513 Request of NRC Project Manager ML20195F3671999-06-0909 June 1999 Forwards Rev 21,Change 0 to EP-001-010, Unusual Event. Rev Reviewed in Accordance with 10CFR50.54(q) Requirements & Determined Not to Decrease Effectiveness of Emergency Plan ML20195C7801999-06-0303 June 1999 Submits Response to Violations Noted in Insp Rept 50-382/99-08.Corrective Actions:All Licensee Access Authorization Personnel Were Retrained Prior to Completion of Insp ML20195C2951999-05-28028 May 1999 Forwards Annual Evaluation of Changes & Errors Identified in Abb CE ECCS Performance Evaluation Models Used for LOCA Analyses.Results of Annual Evaluation for CY98 Detailed in Attached Rept,Based Upon Suppl 10 to Abb CE Rept ML20195C0241999-05-28028 May 1999 Notifies NRC of Operator Medical Condition for Waterford 3 Opertaor Sp Wolfe,License SOP-43723.Attached NRC Form & Memo Contain Info Concerning Condition.Without Encls ML20196L3281999-05-24024 May 1999 Informs That Entergy Is Withdrawing TS Change Request NPF-38-205 Re TS 3.3.3.7.1, Chlorine Detection Sys & TS 3.3.3.7.3, Broad Range Gas Detection Submitted on 980629 ML20206S4691999-05-17017 May 1999 Requests Waiver of Exam for SRO Licenses for an Vest & Hj Lewis,Iaw 10CFR55.47.Both Individuals Have Held Licenses at Plant within Past Two Year Period,But Licenses Expired Upon Leaving Util Employment.Encl Withheld 05000382/LER-1999-004, Forwards LER 99-004-00 Re Discovery That Response Time Testing Had Not Been Performed for ESFAS Containment Cooling Function,As Required by TS SR 4.3.2.31999-05-14014 May 1999 Forwards LER 99-004-00 Re Discovery That Response Time Testing Had Not Been Performed for ESFAS Containment Cooling Function,As Required by TS SR 4.3.2.3 ML20206N1921999-05-10010 May 1999 Provides Revised Attachment 2 for Alternative Request IWE-02,originally Submitted 990429 Re Bolt Torque or Tension Testing of Class Mc pressure-retaining Bolting as Specified in Item 8.20 of Article IWE-2500,Table IWE-2500-1 ML20206J1471999-05-0606 May 1999 Requests That Implementation Date for TS Change Request NPF-38-211 Be within 90 Days of Approval to Allow for Installation of New Monitoring Sys for Broad Range Gas Detection Sys ML20206J1721999-05-0606 May 1999 Notifies That Proposed Schedule for Plant 1999 Annual Exercise Is Wk of 991013.Exercise Objective Meeting Scheduled for 990513 at St John Baptist Parish Emergency Operations Ctr ML20206G8021999-05-0404 May 1999 Provides Revised Response to NRC Re Violations Noted in Insp Rept 50-382/99-01.Licensee Denies Violation as Stated.Change Made Is Denoted by Rev Bar & Does Not Materially Impact Original Ltr ML20206E7811999-04-29029 April 1999 Proposes Alternatives to Requirements of ASME B&PV Code Section XI,1992 Edition,1992 Addenda,As Listed.Approval of Alternative Request on or Before 990915,requested ML20205T2531999-04-22022 April 1999 Forwards LER 99-S02-00,describing Occurrence of Contract Employee Inappropriately Being Granted Unescorted Access to Plant Protected Area ML20205R2611999-04-20020 April 1999 Forwards Rev 19 to Physical Security Plan,Submitted in Accordance with 10CFR50.54(p).Plan Rev Was Approved & Implemented on 990407.Rev Withheld,Per 10CFR73.21 ML20205Q3241999-04-16016 April 1999 Submits Addl Info Re TS Change Request NPF-38-215 for Administrative Controls TS Changes.Appropriate Pages from New Entergy Common QA Program Manual Provided as Attachment to Ltr 1999-09-07
[Table view] Category:UTILITY TO NRC
MONTHYEARW3P90-1505, Forwards Proposed Operator Licensing Exam Schedule & Proposed Requalification Exam Schedule,Per Generic Ltr 90-071990-09-17017 September 1990 Forwards Proposed Operator Licensing Exam Schedule & Proposed Requalification Exam Schedule,Per Generic Ltr 90-07 W3P90-1163, Forwards Relief Requests Associated w/10-yr Inservice Insp Program Per Section 50.55a(g)(6)(i) of 10CFR501990-09-0606 September 1990 Forwards Relief Requests Associated w/10-yr Inservice Insp Program Per Section 50.55a(g)(6)(i) of 10CFR50 W3P90-1191, Responds to Violations Noted in Insp Rept 50-382/90-15. Corrective Actions:Tech Spec Surveillance Procedure PE-005-004 Will Be Revised to Ensure That Normally Closed Valves Opened & Verified to Close for Toxic Gas Signal1990-08-31031 August 1990 Responds to Violations Noted in Insp Rept 50-382/90-15. Corrective Actions:Tech Spec Surveillance Procedure PE-005-004 Will Be Revised to Ensure That Normally Closed Valves Opened & Verified to Close for Toxic Gas Signal W3P90-1194, Submits Fitness for Duty Performance Data for 6-month Period from Jan-June 19901990-08-29029 August 1990 Submits Fitness for Duty Performance Data for 6-month Period from Jan-June 1990 W3P90-1184, Responds to Violations Noted in Insp Rept 50-382/90-14. Corrective Actions:Local Leak Rate Test Activities Shall Be Administratively Controlled to Require Use of Test Method Other than Pressure Decay1990-08-20020 August 1990 Responds to Violations Noted in Insp Rept 50-382/90-14. Corrective Actions:Local Leak Rate Test Activities Shall Be Administratively Controlled to Require Use of Test Method Other than Pressure Decay W3P90-1187, Forwards Booklet Entitled, Safety Info - Plans to Help You During Emergencies, Recently Distributed to General Public1990-08-17017 August 1990 Forwards Booklet Entitled, Safety Info - Plans to Help You During Emergencies, Recently Distributed to General Public W3P90-1189, Forwards Waterford 3 Steam Electric Station Emergency Preparedness Exercise for 901024. Annual Exercise Will Be Performed Using Control Room Simulator1990-08-17017 August 1990 Forwards Waterford 3 Steam Electric Station Emergency Preparedness Exercise for 901024. Annual Exercise Will Be Performed Using Control Room Simulator W3P90-1162, Forwards Rev 4 to 10-Yr Inservice Insp Program First Interval 1985-19951990-08-16016 August 1990 Forwards Rev 4 to 10-Yr Inservice Insp Program First Interval 1985-1995 W3P90-1174, Forwards Rev to Emergency Plan & QA Program,Consisting of Chart Indicating Changes to Util Organization1990-08-0707 August 1990 Forwards Rev to Emergency Plan & QA Program,Consisting of Chart Indicating Changes to Util Organization W3P90-1177, Forwards Revised Objectives for Emergency Preparedness Exercise Scheduled for 9010241990-08-0303 August 1990 Forwards Revised Objectives for Emergency Preparedness Exercise Scheduled for 901024 W3P90-1164, Forwards Waterford Steam Electric Station Unit 3 Basemat Monitoring Program Special Rept 3. Rept Documents Continued Integrity of Basemat as Verified by Program from Time of Inception of Monitoring in 1985 Through Mar 19901990-08-0303 August 1990 Forwards Waterford Steam Electric Station Unit 3 Basemat Monitoring Program Special Rept 3. Rept Documents Continued Integrity of Basemat as Verified by Program from Time of Inception of Monitoring in 1985 Through Mar 1990 W3P90-1167, Forwards Rev 12 to Emergency Plan Implementing Instruction EP-001-001, Recognition & Classification of Emergency Conditions, Reflecting Name Change of State Agency to Louisiana Radiation Protection Div1990-07-19019 July 1990 Forwards Rev 12 to Emergency Plan Implementing Instruction EP-001-001, Recognition & Classification of Emergency Conditions, Reflecting Name Change of State Agency to Louisiana Radiation Protection Div W3P90-1148, Responds to NRC 900503 Submittal Concerning Review of Util Rev 6,Change 1 to Inservice Testing Program for Pumps & Valves1990-07-17017 July 1990 Responds to NRC 900503 Submittal Concerning Review of Util Rev 6,Change 1 to Inservice Testing Program for Pumps & Valves W3P90-1143, Advises That 900404 Request for Addl Info Re Tech Spec Change Request NPF-38-103 Will Be Provided by 900803.Change Will Extend Test Frequency of Channel Functional Tests for ESF Actuation Sys & Reactor Protection Sys Instrumentation1990-07-0606 July 1990 Advises That 900404 Request for Addl Info Re Tech Spec Change Request NPF-38-103 Will Be Provided by 900803.Change Will Extend Test Frequency of Channel Functional Tests for ESF Actuation Sys & Reactor Protection Sys Instrumentation W3P90-1379, Provides Notification That Util Has Consolidated Operation of All Nuclear Facilities,Effective 9006061990-07-0202 July 1990 Provides Notification That Util Has Consolidated Operation of All Nuclear Facilities,Effective 900606 ML20044A5541990-06-26026 June 1990 Forwards Response to Generic Ltr 90-04 Requesting Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions ML20044A5551990-06-22022 June 1990 Describes Changes Required to Emergency Plan as Result of Transfer of Operations to Entergy Operations,Inc. Administrative Changes to Plan Necessary to Distinguish Support Functions to Be Retained by Louisiana Power & Light W3P90-1365, Provides Notification of Change in Operator Status Per 10CFR50.74 Due to Entergy Corp Consolidating Operation of All Nuclear Generating Facilities,Including Plant Under Util1990-06-19019 June 1990 Provides Notification of Change in Operator Status Per 10CFR50.74 Due to Entergy Corp Consolidating Operation of All Nuclear Generating Facilities,Including Plant Under Util ML20043G3431990-06-14014 June 1990 Requests That All NRC Correspondence Re Plant Be Addressed to RP Barkhurst at Address Indicated in 900523 Ltr ML20043F5121990-06-0808 June 1990 Forwards List of Directors & Officers of Entergy Operations, Inc.Operation of All Plants Transferred to Entergy on 900606 ML20043F2621990-06-0606 June 1990 Requests Withdrawal of 900504 Request to Extend Implementation Date of Amend 60 Re Transfer of Operations to Entergy,Inc.All Necessary Regulatory Approvals Obtained & License Conditions Implemented ML20043C1861990-05-29029 May 1990 Submits Response to 900426 Comments Re Investigation Case 4-88-020.Util Issued P.O. Rev Downgrading Order of Circuit Breakers & Eliminating Nuclear Requirements ML20043E5441990-05-24024 May 1990 Forwards Public Version of Change 1 to Rev 2 to EPIP EP-002-015, Emergency Responder Activation. Release Memo Encl ML20043B3501990-05-23023 May 1990 Forwards Response to Concerns Noted in Insp Rept 50-382/90-02.Response Withheld ML20043B3781990-05-23023 May 1990 Requests Change in NRC Correspondence Distribution List, Deleting Rt Lally & Adding DC Hintz,Gw Muench & RB Mcgehee. All Ref to Util Changed to Entergy Operations,Inc.Proposed NRC Correspondence Distribution List Encl W3P90-1314, Requests NRC Concurrence That Design/Controls/Testing to Minimize Potential for Common Header Blockage Acceptable Per 900510 Meeting.Tap Alternatives for Shutdown Cooling Level Indication Sys Discussed1990-05-21021 May 1990 Requests NRC Concurrence That Design/Controls/Testing to Minimize Potential for Common Header Blockage Acceptable Per 900510 Meeting.Tap Alternatives for Shutdown Cooling Level Indication Sys Discussed ML20043B3271990-05-21021 May 1990 Forwards Justification for Continued Operation Re Taped Splice for Use in Instrument Circuits,Per 900517 Request ML20042F5251990-05-0404 May 1990 Requests Extension of 90 Days to Implement Amend 60 to License NPF-38 in Order to Provide Securities & Exchange Commission Time to Review Transfer of Licensed Activities to Entergy Operations,Inc ML20042E5501990-04-17017 April 1990 Responds to Request for Addl Info Re Feedwater Isolation Valve Bases Change Request Dtd 891006 ML20012F4551990-04-10010 April 1990 Forwards Rev 10,Change 4 to Physical Security Plan.Encl Withheld ML20012F5491990-04-0606 April 1990 Advises That Util Installed Two Addl Benchmarks for Use as Part of Basemat Surveillance Program to Increase Efficiency of Survey Readings.New Benchmarks Will Be Shown on FSAR Figure 1.2.1 as Part of Next FSAR Rev ML20012F3181990-04-0606 April 1990 Forwards Util,New Orleans Public Svc,Inc & Entergy Corp 1989 Annual Repts ML20012E8971990-03-30030 March 1990 Submits Results of Evaluation of Util 900414 Response to Station Blackout Rule (10CFR50.63).Station Mod May Be Required to Change Starting Air Sys to Accomodate Compressed Bottled Air ML20012E2551990-03-27027 March 1990 Responds to Violation Noted in Insp Rept 50-382/90-01. Corrective Actions:Qa Review of Licensed Operator Medical Exam Records Conducted & Sys Implemented to Track Types & Due Dates of Medical Exams Required for Operators ML20012E0511990-03-27027 March 1990 Forwards Rev 10,Change 3 to Physical Security Plan.Rev Withheld ML20012D5461990-03-22022 March 1990 Forwards Documentation from Nuclear Mutual Ltd,Nelia & Nuclear Electric Insurance Ltd Certifying Present Onsite Property Damage Insurance ML20012D4911990-03-21021 March 1990 Responds to NRC 900208 Ltr Re Violations Noted in Investigation Rept 4-89-002.Corrective Action:Proper Sequence of Insp Hold Point Placed in Procedure Under Change Implemented on 880425 ML20012C0691990-03-14014 March 1990 Advises That Util Intends to Address Steam Generator Overfill Concerns (USI A-47) Utilizing Individual Plant Exam Process,Per Generic Ltr 89-14 ML20012C0421990-03-12012 March 1990 Forwards Questionnaire in Response to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. Results Not Reflective of Particular Calendar Yr ML20012B6731990-03-0707 March 1990 Responds to NRC Bulletin 88-011,Action 1.a Re Insp of Surge Line to Determine Discernible Distress or Structural Damage & Advises That Neither Surge Line Nor Affiliated Hardware Suffered Any Discernible Distress or Structural Damage ML20006F5321990-02-22022 February 1990 Forwards Payment for Order Imposing Civil Monetary Penalty in Response to Enforcement Action EA-89-069 ML20011F1401990-02-21021 February 1990 Responds to Violations Noted in Insp Rept 50-382/89-41. Corrective Action:Review of Independent Verification Requirements Re Maint Activities Performed ML20006F1731990-02-19019 February 1990 Forwards Corrected Pages 9.2-21 & 9.2-22 of Rev 3 to FSAR, Per 891214 Ltr ML20006E5781990-02-13013 February 1990 Forwards Third Refueling Inservice Insp Summary Rept for Waterford Steam Electric Station Unit 3. ML20006D0571990-02-0202 February 1990 Responds to SALP Rept for Aug 1988 - Oct 1989.Contrary to Info Contained in SALP Rept,Civil Penalty Not Assessed by State of Nv for Radioactive Matl Transport Violations.Issue Resolved W/State of Nv W/O Issuance of Civil Penalty ML20006C1631990-01-30030 January 1990 Requests Extension of Commitment Dates in Response to Violations Noted in Insp Repts 50-382/89-17 & 50-382/89-22 to 900222 & 19,respectively.Violations Covered Use of Duplex Strainers & Missing Seismic Support for Cabinet ML20006C1581990-01-29029 January 1990 Forwards Response to Generic Ltr 89-13 Re safety-related Open Svc Water Sys.Instruments in Place on Component Cooling Water Sys/Auxiliary Component Cooling Water Sys HXs Which Connect to Plant Monitor Computer ML20006C1611990-01-29029 January 1990 Responds to NRC Bulletin 89-003 Re Potential Loss of Required Shutdown Margin During Refueling Operations. Instructions for Determining Acceptable Refueling Boron Concentration Provided in Procedure RF-005-001 ML20006B4121990-01-26026 January 1990 Informs That Photographic Surveys Discontinued,Per Basemat Monitoring Program.Monitoring Program Implementing Procedure Will Be Revised to Reflect Change ML20006A7091990-01-22022 January 1990 Forwards List of Individuals That No Longer Require Reactor Operator Licenses at Plant 1990-09-06
[Table view] |
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LOUISIANA 24a ocanoNot smm P O W E R & L 1 G H T! P o nox 6008 . NEW ORLEANS LOUIS!ANA
. (504) 36670174 2345
$E0YsSSY[U May 29, 1984 W3P84-1492 3-A1.01.04 Director of Nucient Reactor Regulation Attention: Mr. G.W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUBJECT Waterford SES Unit 3 Docket No. 50-382 Response to RSB Questions on Technical Specifications
REFERENCE:
Letter dated May 18, 1984 from Knighton (NRC) to I.eddick (LP&L)
Dear Sir:
In your referenced letter you requested additional information on the Waterford 3 Technical Specifications resulting from a reevaluation by the Reactor Systems Branch.
As you know, LP&L has rnet with RSB personnel to discuss these additional questions. Enclosed please find our response as requested. We trust this is sufficient information to close out your reevaluation.
Should you require further information in this matter please contact Mike Meisner at (504) 363-8938.
Yours very truly.
[
K.W. Cook Nuc1 car Support & Licensing Manager KWC/MJM/pco Enclosure cct E.L. Blake. W.H. Stevenson, J.T. Collins, D.M. C.atchficid, J. Wilson, G.L. Constabic, L.B. Marsh, C.Y. Liang $
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QUESTION:
- 1. Reactor Protective' Instrumentation Setpoints Table 2.2.-1 (Section 2.2 page 2-3)
In reviewing the Reactor Protective Instrumentation Setpoint Table, which is used to determine the relationship between the Reactor Protection Instrumentation Trip Setpoints, the allowable values and the values of these parameters which are used in the safety analyses, the following discrepancies were observed:
See Attachment 1-1 Please resolve the above listed discrepancies.
RESPONSE
Local Power Density-High The case for the single part length CEA drop presented in the FSAR (15.4.1.3) did not explicitly use an analysis setpoint based on local power density. The case was tripped at the latest possible time that still resulted in a peak centerline temperature below that corresponding to centerline melt. The actual CPC trip time would have occurred sooner.
1 The CPCs calculate local power density during a PLCEA drop using core power, )
radial and axial peaks and a CEAC penalty factor for the misaligned CEA.
Conservatisms are included in the CPCs to assure that the plant will trip in time to prevent fuel centerline melt. These conservatisms include the effects of modelling and measurement uncertainties that affect the local power density calculation and are applied to the CPC addressable constants (BERR3 and BERR4).
Thus, a technical specification limit of 6 21kw/ft is acceptable, since the uncertainties are included elsewhere in the CPCs.
e DNBR-Low The change in DNBR limit from 1.19 to 1.20 was the result of an NRC imposed penalty due to fuel spacer grid configuration. Both the Core Operating Limit Supervisory System (COLSS) and Core Protection Calculators (CPCs) were adjusted to conservatively incorporate this penalty. The adjustments to COLSS and CPCs preserve the conclusions of Chapter 15 analyses which are:
(i) the calculated fuel failures based on DNB are conservative, and (ii) the DNBR SAFDL is not violated for Anticipated Operational Occurrences.
If a Chapter 15 event is reported to have a minimum DNBR greater than or equal to 1.19, the event will actually have a minimum DNBR greater than or equal to 1.20 due to the extra penalty applied to COLSS and CPC.
Steam Generator Level-High An increase in feedwater flow is hypothesized to be caused by:
a) a steam generator level instrument failing low; b) failure of the feedwater control system (FWCS) causing a further opening of the feedwater control valve (s) or an increase in feedwater pump speed; '
c) loss of instrument air to the feedwater control valve (s); or d) startup of one or more emergency feedwater (EFW) pumps.
These events, are assumed to be mitigated as follows:
a) Upon turbine trip, each FWCS is assumed to automatically reduce the feedwater flowrate by closing the main feedwater control valves and opening the 5%
bypass valves (FSAR Section 7.7.1.3). The operator then has sufficient time to terminate the event, b) The main feedwater pumps are tripped automatically on steam generator high ,
level signal (FSAR Section 10.4.7.5). '
c) The steam control valves to the main feedwater pump turbines are assumed to fail closed on loss of instrument air, terminating main feedwater flow (FSAR Figure 10.4-2),
i d) The inadvertent actuation of emergency feedwater is terminated by operator action ten minutes after receiving a high steam generator level alarm.
Asguming all three EFW pumps feed one steam generator produces a flow of 210 ;
ft / min. As there is approximately 6,000 ft of steam space in each steam '
generator, there is in excess of 25 minutes before the level would reach the steam piping.
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I QUESTION: .
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- 2. Reactor Coolant System Process Variable LCOS
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Are the values used for process variable LCOs indicated values from the
. instrumentation or the actual values in the systems? If they are actual l values, please explain how instrument uncertainty is accounted for when determining if an LCO is met or exceeded.
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- RESPONSE
j LP&L's practice is to put indicated values for process parameters in the i technical specifications. This avoids the need for the operator to provide a !
correction factor. The indicated values are obtained by applying the appropriate instrument error to the range of initial conditions used in the ,
accident analysis. .
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QUESTION:
- 3. Moderator Temperature Coefficient (Section 3.1.1.3, page 3/4 1-4)
Both the loss of condenser vacuum and the feedwater line break events were analgzed at full power with a moderator temperature coefficient of 0.0 x10 Ak/k/*F. The technical specifications (3.1.1.3) permit plant operagionat70%powerwithamoderatortemperaturecoefficientof+0.2 x 10 Ak/k/*F. Are the e 'nts analyzed at full power with a moderator coefficient of 0.0 more limiting g an operating at 70% power with a moderator coefficient of +0.2 x 10 Ak/k/*F7
RESPONSE
In order to determine if a loss of condenser vacuum (LOCV) or a feedwater line break (FWLB) transient is more limiting operating at full power with a moderator temperaturg coefficient (MTC) of 0.0, or operating at 70% power with a MTC of
+0.2 x 10 Ak/k/*F a comparison of the core power at the time of trip must be made. The scenario which results in the greater reactor power at the time of trip will result in the greater RCS pressurization, because of the largor amount of energy that has to be removed from the RCS following the reactor trip.
Therefore, the transient with the larger core power at the time of reactor trip will be more limiting. The core power at the time -o trip in greater for the 100%0MTCcasesthanforthe70% power,+0.2x10{4k/k/'Fcases. An analysis supporting this conclusion was performed for another C-E designed NSSS (St.
Lucie Unit 2). For this plant, the LOCV analysis wgs performed with the initial power equal to 102% and an MTC equal to +0.13 x 10 Ak/k/*F. The core power at the time of trip was 103.5% of full power. The maximum RCS pressure of this run was 2738 psia. 4Another case was run with the initial power equal to 70% and an MTC of +0.5 x Ak/k/*F. The core power at the time of trip was 74.4% of full power. The maximum RCS pressure was 2629 psia or 109 psi Iower than the full power case. The increase in the average moderator temperature prior to trip is greater for the LOCV case than for the FWLB caso due to the more rapid increase in steam generator pressure. As a result, the core power at the time of trip would be lower for the FWLB than for the LOCV causing the reluction in maximum RCS pressure between the 100 and 70% power cases for the FWLB to be of the same order of magnitude as that for the LOCV. Therefore,itcanbeconclugedthata FWLB or LOCV case analyzed at 100% power with an MTC equal to 0.0 x Ak/k/'F wouldbemorelimitingghanaFWLBorLOCVcaseanalyzedat70%powerwithan MTC equal to +0.2 x 10, 4k/k/10*F.
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QUESTION:
- 4. Baron Dilution (Section 3.1.2.9. page 3/4 1-15, 16, 17)
The Chapter 15 analysis for a boron dilution event relics en operator actions and safety-related alarms; however, there are no technical specifications for the alarm availability, setpoint, or surveillance.
Absent this technical specification, describe what assurance exists that this equipment will always be available and will be properly maintained to meet the Chapter 15 accident analysis assumptions. Also, provide bases for the monitoring frequencies for boron dilution detection listed in tabic 3.1-1.
RESPONSE
The boron dilution alarm setpoint, and periodic resetting of the alarm have been added to the Technical Specification. The alarm shall be set tosk2x the existing count rate at intervals dependent on the time after reactor shutdown (starting with a 5-hour interval, extending to weekly). The monitoring frequen-cies were established to ensure that the time interval between determination of boron concentration was less than the time to loss of shutdown margin depending on the number of charging pumps running. For MODES 3. 4 and 5 the time inter-vals are at least 15 minutes ahorter and for MODE 6 at least 30 minutes shorter to allow time for operator action in accordance with Standard Review plan 15.4.6. The times to loss of shutdown margin are shown in the responses to questions 211.95 and 211.49.
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QUESTION:
- 5. RPS/ESF response timen (Table 3.3-2, page 3/4 3-8 and Table 3.3-5. page 3/4 3-23) t Provide the bases for RPS/ESF response times listed in these Tables or i refer to the assumptions made in Chapter 15 of FSAR. [
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RESPONSE
RPS/ESF response times have been included in the Chapter 15 Safety Analyses. !
The values presented fer the RPS response times (Table 3.3-2 of the Technical ;
Specifications) and for the ESF response times (Table 3.3-5 of the Technical l Specifications) have been reviewed against Chapter 15 of the FSAR and have been [
found to be acceptable.
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QUESTION:
6 Steam Generator Water level (Section 3/4.4)
Explain why there is no LCO on the steam generator water levels. What assurance is there tha* the steam generator water Icyc1 will not exceed the values assumed in the safety analyses?
RESPONSE
1 An LCO on steam generator water level is not necessary since the safety analysis considers the range of steam generator water levels from the low steam generator level trip setpoint to the high steam generator water level trip setpoint. For events in which the value of this parameter would have a utgnificant impact on i
the event consequences the value of this parameter la selected to maximize the consequences. For events in which the consequences have a negifgible sensitivity to this parameter the analysis may assume an arbitrary initial water level within the specified initial condition space.
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. QUESTION:
- 7. Operability of the Steam concrators (Section 4.4.1.2.3 and 4.4.1.3.2, page 3/4 4-2 and 3/4 4-4)
Thene surveillance requirements state that the required steam generator (n) shall be determined operable by verifying the necendary side water level to be 10% of wide range indication at least once por 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Provide bases for the 10% steam generator water level as an adequate water level.
RESPONSE
The minimum water level of the Steam Cenerator secondary side in being revised
! to 50% wide range level. Evaluation of the high level requirement on plant operation revented no impact. The 50% level in sufficient to conservatively account for the followings a) Provides sufficient heat transfer area to remove maximum decay heat and Reactor Coolant Pemp heat without raising the RCS j temperature.
b) Provides sufficient inventory to remove decay heat for at least 30 minutes, one hour after reactor shutdown.
c) Account for instrument inaccuracies in a) and b) above.
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I QL'ESTION: l
- 8. Preneurtrer (Section 3.4.3 page 3/4 4-9)
The technical specification for pressuriser levej during steady-state reactor operation is set between 350 and 900 ft. TgeChapter15 L transient and accident events assumed 370 and-800 ft . please justify . ;
how your safety analysis assumptions for pressuriner level bound the '
levels allowed by your preposed technical specifications.
RESPONSE
TheChagter15analysesgenerallyassumedpressuriserwatervolumesassmallas 370 ft. and as large as 800 f (see NRC caseswatervolumesof900ft.j.and975ft.guantion211.33)althoughinafew were assumed. All events which were analyzed using initial preneuriser water levels within the technical specification range were evaluated as to the impact on the censequences of these events of using the Technical Specification ilmits on pressuriser water volume.
There was found to be negligible impact on the consequences of those events. i l
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QUESTION:
- 9. Auxiliary Pressurizer Spray System (Section 3/4.4)
The current Waterford technical specifications do not include a section to address limiting conditions for operation and surveillance requirements on the auxiliary pressurizer spray system (APSS). It is the staff's understanding that the APSS is required for RCS depressurization during plant shutdown per the requirement of the BTP RSB 5-1 (i.e plant cooldown using only safety-related equipment) and during post-SGTR operation. The issue of whether the APSS is required for mitigation of the SGTR or for RSB BTP 5-1 is a license condition for Waterford 3. Does the applicant intend to develop appropriate technical specifications for the APSS if the resolution of this issue shows that this system is necessary?
RESPONSE
In W3P84-1009, dated April 12, 1984, LP&L committed to a resolution of the Staff concern with respect to a potential single failure vulnerability of the Auxiliary Pressurizer Spray (APS) System within six months of receipt of the low power operating license. Resolution will center on the need for APS to satisfy BTP RSB 5-1 and/or SGTR criteria. Should the APS prove necessary in this regard, LP&L agrees that issue resolution shall include a commitment to develop a technical specification to address limiting conditions for operation and surveillance requirements on the APS.
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QUESTION:
- 10. Overpressure Protection System (Section 3.4.8.3, page 3/4 4-34)
The technical specification for overpressure protection systems (Section 3.4.8.3), page 3/4 4-34) references the suction line relief valves as SI-406A and SI-406B. Section 9.3.6.2.2, page 9.3-49. refers to these valves as SI-486 and SI-487. We understand these are LP&L numbers.
What set of valve numbers is correct? How have you assured yourselves that there is no duplication of valve numbers as a result of the different valve numbering systems?
RESPONSE
Valves.SI-486 and SI-487 are the same as SI-406A and SI-406B. The former numbers, used in the FSAR, are Combustion Engineering assigned numbers whereas the latter, used in the Technical Specifications are LP&L unique identification
, numbers. At present, the FSAR typically uses EBASCO valve tag numbers.
However, some FSAR sections written in early Amendments still cite CE valve number, whereas the Technical Specifications use LP&L numbers and often show the corresponding EBASCO numbers. At the plant, procedures use LP&L numbers, and valves are tagged with LP&L numbers. CE numbers are not used at the plant.
Therefore, should duplication within the CE and LP&L numbering systems exist, it will not affect plant operation. Ebasco numbers, which are also shown on valve tags are a different format and cannot be confused with LP&L or CE numbers.
Plant. procedures (not the FSAR) are used to manipulate valves and controls and thus the different numbering systems should not affect operation of the plant.
Flow diagrams retain all three numbering systems, distinguishing each from the others. Retention of the CE numbers in this case is necessary for traceability to engineering documents and QA records.
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- 11. Overpressure Protection Systems (Section 3.4.8.3, page 3/4 4-35)
Section 4.4.8.3.1 states that each shutdown cooling system suction line relief valve shall be demonstrated operable by verifying that each valve in'the suction path between the RCS and the shutdown cooling relief valve is key-locked open in the control room at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Could the auto closure interlock override the key-locked open isolation valves so that the SDC system isolation valves could be closed when RCS pressure exceeding'700 psig? Otherwise explain how the system design preclude a possible event V.
. RESPONSE:
The term " key-locked open" refers to the type control switch used on these
- valves. To operate the switch a key must be inserted; then the valve may be opened using the key. When the valve is shut (normal position during operation) the key is removed to prevent inadvertent operation. Operation of the key switch does not affect the automatic features such as isolation on high RCS pressure.
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QUESTION:
- 12. Reactor Coolant System Vents (Section 3.4.10, page 3/4 4-37)
The current Waterford technical specifications do not ensure the availability of the RCS vents during plant operation.- It is the staff's understanding that the applicant intends to take credit for RCS vents for RCF. depressurization during safe shutdown per BTP RSB 5-1. Does the applicant intend to modify the existing Technical Specification for the RCS vents if'the ongoing assessment shows that this system is necessary for meeting the RSB BTP 5-1 positions?
RESPONSE
The Reactor Coolant System Vents are tied to the resolution of Waterford compliance with.BTP RSB 5-1. In W3P84-0505 dated February 29, 1984 LP&L submitted CEN-259 which included an evaluation of conformance with BTP RSB 5-1.
In that document it was noted that should the APS be unavailable, the primary depressurization could be accomplished via the safety grade RCS Vents. There presently exists.a technical specification (3/4.4.10) on this system which ensures operability of the RCS vents during plant operation. This technical specification will be changed as necessary depending on the results of the NRC's review of BTP RSB 5-1.
I QUESTION:
- 13. Safety Injection "anks (Section 3.5.1, page 3/4 5-1)
-Section 3/4 5.1 describes the modes of operation for the safety injection tanks. The basis for this item implies that the values in the Technical Specification were chosen for compliance with the accident analyses. Address why there are no specifications for the coolant temperature in SIT. Otherwise, justify why the SIT coolant temperatare assumed in the ECCS analyses bound the maximum temperature the SIT could attain.
RESPONSE
The LOCA analysis assumes a temperature for the Safety Injection Tanks (SIT) of 120*F, because for these analyses a higher temperature is more adverse. The temperature of 120*F is assumed to be the maximum, since this is the limit on containment air temperature specified in Technical Specification 3/4.6.1.5. The steam line analyses were performed assuming a SIT temperature of 120*F.
Calculations have shown that by assuming a SIT temperature of 40*F the average RCS temperature will be lowered by less than l'F at the time of minimum DNBR.
This change in temperature has a negligible effect on the return to power and consequently a negligible effect on the minimum DNBR. Thus, for the steam line break analyses the SIT temperature has a negligible impact on the event consequences when a range of 40*F to 120*F is considered.
QUESTION:
- 14. Atmospheric Steam Dump Valves (Section 3/4.7)
The current Waterford technical specification does not include a section to address limiting conditions for operation and surveillance requirements on the atmospheric steam dump valves (ADVs).
Since the ADVs are required during initial phase of plant shutdown per the requirements of the BTP RSE 5-1 (i.e., plant cooldown using only safety-related equipment), and we understand your FSAR Chapter 15 steam generator tube rupture analysis takes credit for these components, explain what assurances exist in the plant that these components will always be operable in accordance with the assumptions made in the safety analyses.
Similarly, the Staff and Commission concluded it was acceptable to defer a decision on the need to install PORVs in your plant based, in part, on the CE PRA study performed for your plant. This PRA placed high reliability on the availability of the ADVs to effect decay heat removal. In responding to the above question, please address how the assurances you are providing are consistent with the reliability assumptions made in your PRA.
RESPONSE
LP&L does not disagree with the intent of developing a technical specification to ensure availability for the Atmospheric Dump Valves. However, the issue is generic. Our position is that development of generic technical specifications requires careful consideration and review on the part of both the NRC and the industry. Unilateral. action does not allow for wide review and, in the long run, could be detrimental to other utilities. We suggest, therefore, that this and other generic technical specification development be remanded to the normal Staff process for Standard Technical Specifications including CRGR review.
In the interim, without explicit Technical Specifications governing the ADVs, LP&L still considers these valves to be governed by the Technical Specifications. Section 4.0.5 invokes the requirements of ASME Section XI valve testing in which the ADVs are covered due to their importance to safe shutdown.
Note that 4.0.5 d states that Section XI requirements are in addition to other specified surveillance requirements. In addition, the definition of OPERABLE-0PERABILITY requires that a system, subsystem, train, component or device must have all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment also capable of performing their related support function (s). In this manner, the ADVs are incorporated into the Specifications for Reactor Coolant Loops having the basis for decay heat removal via bleeding steam from the steam generator.
As to the PRA work on the PORV issue, it should be noted that many systems were considered in the risk assessment presented in CEN 239 which could impact the calculated core damage frequencies. Some of these systems (e.g. main feedwater system) are not safety grade nor are they included in the Technical Specifica-tions for most plants. The availability values assumed for these systems
(including the ADVs) were based upon historical operational data. The issue of whether or not to Tech Spec these systems to meet the assumed availability values is therefore irrelevant.
At the present time, ADVs are also included in the Containment Isolation Valve Technical Specification 3.6.3, requiring that they be OPERABLE. LP8L does not concur with Containment Systems Branch's inclusion of the ADVs as well as other essential valves in Table 3.6-2, and documented this position in W3P84-0577 dated March 16, 1984. Subjecting ADVs to the Action Requirements of this Specification is felt to be non-conservative with respect to their primary function and is an example of problems that result when sufficient inter-branch review on a generic basis is not done.
Y l QUESTION:
- 15. Special Test Exceptions, Reactor Coolant Loops (Section 3/4.10.3 page 3/4 10-3).
This technical specification permits plant operation without any reactor coolant pumps operating up to 5% thermal power on fission heat for ~
startup or physics tests. What safety analyses have been conducted that demonstrate that transients or accidents initiated from this operating condition would be acceptable for Waterford 3? Both the steady state and transient reactor coolant system temperature profiles, margin to saturation, core DNBR, and other related thermal-hydraulic stability, should be assessed. The acceptability of the reactor protective system setpoints during various transients and accidents initiated from this condition must also be justified.
RESPONSE
The Special Test Exception in 3/4.10.3 concerning specification 3.4.1.1 is no longer required and will be removed. This was developed to support the low power Natural Circulation testing. However the test methodology was revised and this exception is no longer necessary.