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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
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+ P. O. BOX 60340 l OUISI&AN POWER A LIGHT / 317NEW BARONNESTREET ORLEANS, LOUISIANA 70160 + (504) 595-3100
$?ONEsY$
October 23, 1986 W3P86-3379 A4.05 QA Mr. George W. Knighton, Director PWR Project Directorate No. '7 Division of PWR Licensing-B Office of Nuclear Reactor Regulation Washington, D.C. 20555
SUBJECT:
Waterford SES Unit 3 Docket No. 50-382 Technical Specification Change Requests NPF-38-23, 40, 42 Additional Information
REFERENCES:
(1) W3P86-2163 dated June 24, 1986 (2) W3P86-3323 dated September 25, 1986
Dear Mr. Knighton:
By the Reference (1) letter LP&L requested a Technical Specification change (NPF-38-23) to implement a bypass for the non-safety related high steam generator level trip. In Reference (2), changes to the shutdown margin (NPF-38-40) and special test exception (NPF-38-42) Technical Specifications were requested to support Cycle 2 operation.
In subsequent discussions, your staff requested additional information concerning these changes. Enclosed please find our response:
Attachment 1 - NPF-38-40 '
i Attachment 2 - NPF-38-42 Attachment 3 - NPI-38-23 I Should you require any further information please contact Mike Meisner at j l
(504) 595-2832.
Yours very truly, 8610290059 861023 jA PDR ADOCK 05000382 g'd L Q' /
P PDR K.W. Cook Nuclear Support & Licensing Manager l Enclosures cc: B.W. Churchill, W.M. Stevenson, R.D. Martin, J.H. Wilson, L. Kopp (NRC/NRR) ,
NRC Resident Inspector's Of fice (W3) M "AN EQUAL OPPORTUNITY EMPLOYER" l\g\
l
Attachment 1 to W3P86-3379 Page 1 of 3 Additional Information License Amendment Request NPF-38-40 QUESTION:
In reference to proposed change number 40 to Technical Specifications 3.1.1.1 and 3.1.1.2, Shutdown Margin:
(a) Which of the measured RCS cold leg temperatures is used for T-cold?
Response
The lowest of the eight (8) safety-related cold leg temperatures are generally used to determine the Shutdown Margin requirements. However, the equivalent boron concentrations calculated per Operations procedure OP-902-090, Shutdown Margin, have sufficient conservatism to account for normally observed differences in the RCS T-cold indications. Thus, any safety-related RCS T-cold value may be used for the calculation of required Shutdown Margin.
(b) How does the Shutdown Margin when all full-length CEAs are fully inserted account for the highest worth CEA fully withdrawn (stuck-out)?
Response
By definition, the Shutdown Margin assumes that the CEA of highest worth is always fully withdrawn (stuck-out). When the actual Shutdown Margin is calculated at Waterford-3, one CEA is assumed to be stuck-out to be con-sistent with the definition. However, in the safety analyses supporting this Technical Specification change, credit was taken for all CEAs being fully inserted. If, for example, the stuck CEA was worth 1%, then a Shut-down Margin of 1% would mean that the core is actually subcritical by 2%
(i.e., 1% Shutdown Margin plus the worth of the stuck-out CEA) since all CEAs have been verified to be fully inserted. The safety analyses that support this change (zero power steam line break, boron dilution, etc.)
were initiated therefore, with the core subcritical by 2% since all CEAs were assumed to be inserted.
(c) Why does the 1% Shutdown Margin in Technical Specification 3.1.1.2 (Figure 3.1-0) extend to 400 F rather than 200 F as in Tech Spec 3.1.1.1?
Response
Above approximately 200 F, the Shutdown Margin requirements depicted in Figure 3.1-0 are a direct result of the all rods in configuration and the analysis of the RCS cooldown and resulting reactivity transients associated with Steam Line Break accidents at different (initial) RCS temperatures. As the initial RCS temperature decreases, the potential RCS cooldown and associated reactivity transient are less severe. This less severe RCS cooldown and more favorable reactivity transient result in a 1% Shutdown l
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Attachmint 1 to W3P86-3379 Page 2 of 3 Additional Information License Amendment Request NPF-38-40 Margin requirement at 400 F. This explicit analysis as a function of initial RCS temperature was not done for Cycle 1; hence, the Shutdown Margin determined at hot zero power conditions was conservatively extended to 200 F.
(d) Justify the linear interpolation over 100 F between the 1% and the 4.15%
Shutdown Margin values in Figure 3.1-0.
Response
As stated in the response to question (c), above 200 F the Shutdown Margin requirements are a direct result of reanalyzing the RCS cooldown associated with the Steam Line Break accident at various (initial) RCS temperatures.
The linear interpolation shown in Figure 3.1-0 bounds the Shutdown Margin requirements for initial RCS temperatures between 400 F and 500 F. That is, specific RCS cooldowns were analyzed at intermediate temperatures (between 400 and 500 F) and were found to require less Shutdown Margin than that required by the linear interpolation shown in Figure 3.1-0.
(e) Discuss the results of an inadvertent boron dilution event initiated from an initial subcriticality of 1% allowed by Figure 3.1-0.
Response
As discussed in the response to question (b), when all full-length CEAs are inserted a Shutdown Margin of 1% (as shown in Figure 3.1-0) means the core is actually subcritical by over 2% (1% Shutdown Margin plus the worth of the most reactive CEA). The inadvertent boron dilution with a Shutdown Margin as allowed by Figure 3.1-0 is described in Section 7.4.4 of the Reload Analysis Report. Under these conditions there is at least 15 minutes from the time an alarm makes the operator aware that an unplanned boron dilution is in progress until a loss of Shutdown Margin occurs. This is consistent with Standard Review Plan Section 15.4.6 and shows that sufficient time exists for the operator to identify the event and take the required action to terminate it.
(f) Discuss the results of the CEA Withdrawal event from a subcritical or low power condition assuming the Shutdown Margin values given in Figure 3.1-0.
Response
The Shutdown Margin required to preclude (or mitigate) the consequences of a CEA withdrawal event from a low power or subcritical condition is bounded by the Steam Line Break (at high temperatures). At lower temperatures, protection from the CEA_githdrawl event is provided by automatic removal of the CPC bypass at 10 power. Additional discussion of the CEA with-drawal event from a subcritical or low power condition is provided in Section 7.4.1 of the Reload Analysis Report.
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Attachment 1 to
. W3P86-3379 Page 3 of 3 Additional Information License Amendment Request NPF-38-40 (g) How does proposed Technical Specification 3.3.1.2 affect the definition of Shutdown Margin given in Section 1.0?
Response
There will be no change to the definition of Shutdown Margin given in Section 1.0 of the Technical Specifications. That is, the Shutdown Margin requirements described in proposed Tech Spec 3.1.1.2 do not include the worth of the most reactive CEA.
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Attachm:nt 2 to W3P86-3379 Page 1 of 2 Additional Information License Amendment Request NPF-38-42 QUESTION:
In relation to proposed change number 42 to Technical Specification 3/4.10.3, Special Test Exceptions:
(a) Why must CPC trips be bypassed during physics testing?
Response
In order to measure CEA worths during Cycle 2 physics testing, it is necessary to insert and withdraw CEAs outside of the normally prescribed sequence. If the CPCs were not bypassed, they would generate out of sequence per,alty factors that would result in a reactor trip.
The intent of this proposed Tech Spec change is not to request permission to bypass the CPCs during physics testing since this is already allowed by Technical Specification 3.3.1 (Table 3.3-1, Notes c and d); but rather to credit the log power trip for protection against power transients initiated at low power levels instead of relying on reduced reactor trip setpoints in the linear power channels (which is currently required by Special Test Exception 3.10.3).
In order to bypass all 4 CPC channels without generating a reactor trip, it is necessary to increase the CPC operating bypass permissive bistable toavalue_g%and10bove between 10 the_ ower level of thermal where physics testing is performed power). (usually
, the CPCs would automatically come out of bypass at a thermal If this were notof power dong 10 % and potentially trip the reactor. Since this same bistable also serves as the threshold value for bypassing the log power trip (i.e., it is the minimum thermal power below which the log power trip can not be bypassed), setting it above the log power trip setpoint (0.257% power per Technical Specifi-cation 2.2.1, Table 2.2-1) removes the possibility of bypassing the log power trip. Thus, by bypassing all 4 CPCs, as described in proposed change number 42, a type of " electrical interlock" is created that precludes the log power trip from being bypassed. Therefore, the log power trip will provide the necessary protection for increasing power transients precluding the need to change the high linear power setpoint.
It should also be noted that increasing the CPC operating bypass permissive bistable setpoint above the log power trip setpoint does not preclude the CPCs from performing a protective function. That is, if the thermal power were to exceed the bistable setpoint (without causing a log power trip),
the CPCs would automatically come out of bypass and, if necessary, trip the reactor.
Finally, proposed change number 42 inadvertently omitted the word "either" following LCO 3.10.3a and the word "or" following LC0 3.10.3b. The corrected page is included at the end of this Attachment.
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. Attachment 2 to W3P86-3379 Page 2 of 2 Additional Information License Amendment Request NPF-38-42 (b) Discuss any possible transients, such as CEA withdrawal events, which may arise due to the removal of the CEA withdrawal prohibit when bypassing the CPCs.
Response
The CEA withdrawal event, or any increasing power transient, would result in a reactor trip when the core thermal power reached the high log power trip setpoint (0.257%) or, when the core thermal power reached the value at which the CPCs would automatically come out of bypass and, if warranted, trip the reactor.
(c) What low power transients rely on the CPC trips?
Response
The transients that are initiated from a low initial core power level (e.g., the CEA withdrawal event) credit the CPC variable overpower trip for protection. Since the CPCs automatically come out of bypass, they will be available to provide protection against any transients that are initiated from low power.
(d) Does the high log power level trip require a faster rate of power increase to trip as compared to the high LPD trip? If so, what is the effect of the fact that it may not catch rapid power increases as quickly as the CPC trip?
Response
The high log power level trip does not require any rate of power increase to trip the reactor. The high log power level trip setpoint is set at an absolute value of 0.257% thermal power per Technical Specification 2.2.1, Table 2.2-1.
(e) operator to reset the CPC operating bypass permissive What reminds setpoint to 10 thg% after testing?
Response
Once the low power physics test program has been completed the CPC operatjngbypasspermissivesetpointisreturnedtoitsnominalvalue of 10 % power per station procedure NE-2-003.
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> . .Y SPECIAL TEST EXCEPTIONS 3/4.10.3 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.3 The noted requirements of Tables 2.2-1 and 3.3-1 may be suspended during the performance of startup and PHYSICS TESTS, provided:
a.
The THERMAL POWER does not exceed SX of RATED THERMAL POWER, and Bill)ef I b.
The reactor trip setpoints of the OPERA 8LE power level channels are set at less than or equal to 20% of RATED THERMAL POWER, or l f APPLI ILITY: During startup and PHYSICS TESTS.
ACTION:
With trip the the THERMAL reactor. POWER greater than 5% of RATED THERMAL POWER, immediately SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%
j of RATED THERMAL POWER at least once per hour during startup and PHYSICS TESTS.
4.10.3.2 Each wide range logarithmic and power level neutron flux monitoring channel shall be subjected to a CHAMEL FUNCTIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating startup and PHYSICS TESTS.
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Attachm:nt 3 to W3P86-3379 Page 1 of 4 Additional Information License Amendment Request NPF-38-23 QUESTION:
(a) Technical Specification proposed change number 23 requests providing the plant operators with the capability of bypassing the high steam generator level reactor trip. It appears that credit is indirectly taken for this >
trip in events such as the increase in feedwater flow (FSAR Sec. 15.1.1.2) by using it to set the maximum water level used as an initial condition.
Please discuss. The statement is made that even should the main steam line piping be postulated to rupture due to the water loading, the resulting event is bounded by the main stea,n line break event analyzed in the FSAR.
Justify this by providing the results of a main steam line break analysis occurring at the highest possible steam generator water level and comparing these results (e.g., minimum DNBR, return to power, containment pressure and temperature) with the results of the FSAR steam line break analysis. Also, confirm that the resulting containment pressure and temperature is less than the LOCA containment pressure and temperature. Discuss the effect of possible higher temperatures than previously considered in the vicinity of the break on equipment qualification.
Response
Proposed Technical Specification Change Number 23 was requested to assist operators in reducing reactor trips during low power operation when steam generator level is being manually controlled. Although not noted in the change request submittal, use of the high steam generator water level bypass will be administratively controlled such that the bypass cannot be enabled above 20% power. These controls will ensure that the high steam generator level trip is available to maintain steam generator inventory below the trip setpoint for events occurring above 20% power.
In this context, a distinction can be made between high and low power scenarios and the means available to limit steam generator water level.
At power levels greater than 20% the proposed Technical Specification change has no effect on plant operations -- i.e. the high water level trip remains.
available as in Cycle 1. For power levels below 20% the high level trip may be bypassed. However, for this case additional mechanisms exist to place an upper limit on water level. First, the main feedwater isolation valves (MFIVs) receive a safety-related close signal (independent of the feedwater control system) when the steam generator level reaches 89.5%
narrow range (compared to the high steam generator level trip setpoint of 87.7% narrow range which may be in bypass). Second, when the proposed bypass is in use the operator will have manual control of the feedwater system and be directed by procedure to maintain steam generator level at approximately 68% narrow range. The combination of these two factors ensures that steam generator level at low power levels will not increase to the point of filling the main steam piping.
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Attachm nt 3 to W3P86-3379 Page 2 of 4 Additional Information License Amendment Request NPF-38-23 Response (Cont'd):
The NRC's question consists of three elements:
- 1. The results of a postulated rupture of the main steam line due to water loading,
- 2. A discussion of the events which may indirectly credit a maximum steam generator water level, and
- 3. Confirmation that peak containment temperature and pressure due to Item 1 remain below that for LOCA.
These elements are covered in the following discussions.
- 1. Main Steam Rupture Due to Water Loading As discussed above, for situations when the high steam generator level trip may be in bypass (i.e..less than 20% power) sufficient automatic and manual controls exist to ensure that the main steam piping does not become water filled.
- 2. Indirect Credit For Maximum Steam Generator Water Level At power levels above 20% the high steam generator water level trip will not be bypassed and, therefore, will not affect FSAR events analyzed at high power levels. Low power events are discussed below.
Section 15.1.1.2 of the Waterford 3 FSAR states, in a discussion of the Increased Feedwater Flow Event, that protection against high steam generator water level is provided by a high steam generator level trip. Although at low power levels this trip could potentially be bypassed, allowing the steam generator to fill beyond the high steam generator level trip setpoint, the results of this event will not be as severe as the Increased Main Steam Flow which is discussed in FSAR Sections 15.1.1.3.
For Cycle 1, the Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (IOSGADV) Event was initiated at zero power and an initial steam gen-erator water level just below the high steam generator water level trip set-i point in order to maximize secondary side water inventory and the radiologi-i cal consequences. If the high level trip was bypassed, the potential exists for the initial steam generator water level to be as high as 89.5% (narrow range), at which point the Main Feedwater Isolation Valves (MFIVs) receive a close signal. LP&L has performed an analysis to determine the potential impact of the increased steam generator inventory on the radiological consequences of the IOSGADV. The analysis conservatively assumed a 2%
process error on the closure signal to the MFIVs and took no credit for steam generator internals (e.g., steam separators) displacing some of the water. This resulted in a maximum increase in the steam generator inventory of approximately 21,000 lbm.
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Attachm:nt 3 to W3P86-3379 Page 3 of 4 Additional Information License Amendment Request NPF-38-23 Response (Cont'd):
Assuming the additional steam generator inventory was at the Technical Specification limit for iodine activity (0.1 uCuries/ gram) and all of the inventory is released directly to the environment during the first two hours of the event, the resulting increase in the thyroid dose at the site boundary is approximately 0.31 rem. This compares with the calculated site boundary doses shown in the FSAR of 5.5 rem (for the IOSGADV with no loss of off-site power) and 6.0 rem (for the IOSGADV with a concurrent loss of off-site power). This increase is within the calculative uncertainties of the analysis and the total dose is well within the guidelines established by 10CFR100.
The steam line break events presented in Section 7.1.5 of the Reload Anal-ysis Report, Steam System Piping Failures, are bounding for the following reasons:
The minimum DNBR during a steam line break accident occurs for full power initial conditions when a loss of off-site power is assumed to occur coin-
~
cident with the reactor trip. For this case (described in section 7.1.5a of the Reload Analysis Report), the minimum DNBR occurs well before the affec-ted steam generator empties and, therefore, the initial steam generator water level has no impact on the results.
The minimum post-trip DNBR (and maximum post-trip power excursion) occurs assuming full power initial conditions and a loss of off-site power at the time of reactor trip (presented in Section 7.1.5b of the Reload Analysis Report). At full power the high steam generator water level alarm and reactor trip is available to limit the inventory. This full power case is more severe than a zero power case with a higher steam generator water level due to a greater number of delayed neutrons which contribute to the post-trip power increase.
The maximum radiological consequences of a steam line break occur during an outside containment break from full power initial conditions and a loss of off-site power at the time of the reactor trip (presented in Section 7.1.5a of the Reload Analysis Report). The full power case is again more limiting than the case with the maximum steam generator inventory (zero power) because it results in more postulated fuel failures.
- 3. Peak Containment Temperature and Pressures Operation with the high steam generator level trip in bypass may occur only below 20% power.
The peak containment temperature and pressure result from a 7.4765 ft 2 steam line break at 75% power as shown in Figures 6.2-7a and 6.2-7b of the FSAR. Therefore, a steam line break at 20% power or less will result in lower peak containment temperatures or pressures.
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Attachment 3 to a
W3P86-3379 Page 4 of 4 Additional Information License Amendment Request NPF-38-23 QUESTION:
(b) Where in the FSAR is there a discussion of the Steam Generator Overfill event? Specifically where does it evaluate filling of the Main Steam Lines up to the Main Steam Isolation Valves? If the evaluation is not in the FSAR what is the source document containing the evaluation?
Response
The FSAR does not address a steam generator overfill event. As discussed in response to the previous question, sufficient controls exist to ensure that the main steam piping does not become water filled.
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