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Category:CORRESPONDENCE-LETTERS
MONTHYEARLIC-99-0098, Forwards Revs to Fort Calhoun Station EPIPs & Emergency Planning Forms Manual.Document Update Instructions & Summary of Changes Are Included on Confirmation of Transmittal Form Attached1999-10-15015 October 1999 Forwards Revs to Fort Calhoun Station EPIPs & Emergency Planning Forms Manual.Document Update Instructions & Summary of Changes Are Included on Confirmation of Transmittal Form Attached ML20217G9261999-10-15015 October 1999 Forwards Insp Rept 50-285/99-10 on 990913-17.One NCV Noted ML20217G2331999-10-14014 October 1999 Forwards Insp Rept 50-285/99-09 on 990913-17.No Violations Noted LIC-99-0091, Submits Suppl Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, in Response to NRC RAI During 990817 Meeting with Util1999-10-0808 October 1999 Submits Suppl Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, in Response to NRC RAI During 990817 Meeting with Util LIC-99-0097, Forwards Final RO & SRO License Applications,As Requested in ,Certifying That Training Completed for Six RO Candidates,Three Instant SRO Candidates & One Upgrade SRO Candidate.Without Encls1999-10-0707 October 1999 Forwards Final RO & SRO License Applications,As Requested in ,Certifying That Training Completed for Six RO Candidates,Three Instant SRO Candidates & One Upgrade SRO Candidate.Without Encls ML20217B5111999-10-0606 October 1999 Forwards Amend 193 to License DPR-40 & Safety Evaluation. Amend Revises TS Sections 2.10.4,3.1 & Table 3-3 to Increase Min Required RCS Flow Rate & Change Surveillance require- Ments for RCS Flow Rate ML20216H7231999-09-29029 September 1999 Informs That Util 980325 Response to GL 97-06, Degradation of SG Internals, Provides Reasonable Assurance That Condition of Util SG Internals in Compliance with Current Licensing Bases for Fort Calhoun Station,Unit 1 ML20212L9571999-09-27027 September 1999 Informs That Follow Up Insp for Drill/Exercise Performance Indicator Will Be Conducted During Wk of 991001 LIC-99-0089, Forwards Preliminary License Exam Applications for Six Ros, Three Instant SROs & One Upgrade SRO Candidate,Per Preparation for Ro/Sro Licensing Exams to Be Administered on 991025-29 at Fcs.Without Encl1999-09-24024 September 1999 Forwards Preliminary License Exam Applications for Six Ros, Three Instant SROs & One Upgrade SRO Candidate,Per Preparation for Ro/Sro Licensing Exams to Be Administered on 991025-29 at Fcs.Without Encl LIC-99-0092, Forwards Revised Epips,Including non-proprietary Rev 8 to EPIP-RR-66,rev 3 to EPIP-EOF-24,rev 14 to RERP-Section F,Rev 11 to RERP-Section I,Rev 12 to RERP-Appendix C & RERP & EPIP Indexes.Proprietary Rev 3 to EPIP-EOF-24,pages 1-3,encl1999-09-21021 September 1999 Forwards Revised Epips,Including non-proprietary Rev 8 to EPIP-RR-66,rev 3 to EPIP-EOF-24,rev 14 to RERP-Section F,Rev 11 to RERP-Section I,Rev 12 to RERP-Appendix C & RERP & EPIP Indexes.Proprietary Rev 3 to EPIP-EOF-24,pages 1-3,encl LIC-99-0087, Forwards Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams1999-09-17017 September 1999 Forwards Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams LIC-99-0088, Forwards Documentation Associated with SRO License Renewal Application for Cd Darrow.Applicant Has Satisfactorily Completed Applicable FCS Requalification Program.Without Encls1999-09-16016 September 1999 Forwards Documentation Associated with SRO License Renewal Application for Cd Darrow.Applicant Has Satisfactorily Completed Applicable FCS Requalification Program.Without Encls LIC-99-0080, Requests Approval of Attached Rev to QAP Contained in App a of Usar,Revising Commitment Re Qualifications of Personnel Performing Quality Control Insps of Activities Affecting Fire Protection1999-09-16016 September 1999 Requests Approval of Attached Rev to QAP Contained in App a of Usar,Revising Commitment Re Qualifications of Personnel Performing Quality Control Insps of Activities Affecting Fire Protection LIC-99-0085, Forwards Revs 18 & 19 to TDB-VI, COLR for FCS Unit 1,IAW TS 5.9.5.Rev 18 Resulted from Reevaluation of Cycle 18 Setpoints & Rev 19 Resulted from Changes in Cycle 19 Refueling Boron Concentrations & Addition of TS Parameters1999-09-0808 September 1999 Forwards Revs 18 & 19 to TDB-VI, COLR for FCS Unit 1,IAW TS 5.9.5.Rev 18 Resulted from Reevaluation of Cycle 18 Setpoints & Rev 19 Resulted from Changes in Cycle 19 Refueling Boron Concentrations & Addition of TS Parameters IR 05000285/19990081999-09-0707 September 1999 Forwards Insp Rept 50-285/99-08 on 990907-0811.No Violations Noted.Insp Consisted of Exam of Activities Conducted Under License as Relate to Safety & to Compliance with Commission Rules & Regulations & with Conditions of License ML20211P2581999-09-0303 September 1999 Forwards Insp Rept 50-285/99-06 on 990809-12.No Violations Noted.Insp Consisted of Exam of Activities Under License as Related to EP & to Compliance with Commission Rules & Regulations & with Conditions of License LIC-99-0083, Requests Participation in internet-based Y2K Early Warning Sys (Yews),As Noted in NRC Info Notice 99-25.Plant Point of Contact Will Be JW Chase1999-09-0202 September 1999 Requests Participation in internet-based Y2K Early Warning Sys (Yews),As Noted in NRC Info Notice 99-25.Plant Point of Contact Will Be JW Chase ML20211J9231999-09-0202 September 1999 Forwards SE Accepting Licensee 990301 Requests for Relief Re Third 10-year ISI Interval for Plant,Unit 1.Relief Proposed Alternatives to Requirements in ASME B&PV Section XI,1989 Edition,Paragraphs IWA-5242(a) & IWA-5250(a)(2) ML20211L3241999-09-0202 September 1999 Informs That Due to Administrative Error,Page 2-50 of TSs Contained Error on 2.10.2(2) of Amend 192 to License DPR-40 Issued on 990727.Corrected Page Encl LIC-99-0077, Forwards non-proprietary & Proprietary Revised EP Forms, Including FC-EPF Index,Pp 1 Dtd 990805,pp 2 Dtd 990812,rev 11 to FC-EPF-9,rev 12 to FC-EPF-10,rev 9 to FC-EPF-11,rev 2 to FC-EPF-12 & Rev 3 to FC-EPF-38.Proprietary EP Withheld1999-08-27027 August 1999 Forwards non-proprietary & Proprietary Revised EP Forms, Including FC-EPF Index,Pp 1 Dtd 990805,pp 2 Dtd 990812,rev 11 to FC-EPF-9,rev 12 to FC-EPF-10,rev 9 to FC-EPF-11,rev 2 to FC-EPF-12 & Rev 3 to FC-EPF-38.Proprietary EP Withheld ML20211F4491999-08-25025 August 1999 Forwards Ltr from Ja Miller,Dtd 990819,which Notified State of Iowa of Deficiency Identified During 990810,Fort Calhoun Station Biennial Exercise.One Deficiency Identified for Not Adequately Demonstrating Organizational Capability LIC-99-0076, Forwards fitness-for-duty Program Performance Data for Six Month Period from Jan-June 1999,IAW 10CFR26.71(d)1999-08-20020 August 1999 Forwards fitness-for-duty Program Performance Data for Six Month Period from Jan-June 1999,IAW 10CFR26.71(d) LIC-99-0078, Forwards Rev 16C to FCS Site Security Plan,Per Provisions of 10CFR50.54(p).Rev Has Been Determined Not to Degrade Effectiveness of Current Security Plans.List of Changes, Provided.Encl Withheld1999-08-20020 August 1999 Forwards Rev 16C to FCS Site Security Plan,Per Provisions of 10CFR50.54(p).Rev Has Been Determined Not to Degrade Effectiveness of Current Security Plans.List of Changes, Provided.Encl Withheld ML20210R1491999-08-13013 August 1999 Forwards Insp Rept 50-285/99-07 on 990719-23.No Violations Noted.Nrc Identified One Issue Which Was Evaluated Under Risk Significance Determination Process & Was Determined to Be of Low Risk Significance LIC-99-0075, Forwards Monthly Operating Rept for July 1999 for Fcs,Unit 1,per Plant TS 5.9.1.Revised Rept for June 1999,encl Due to Inadvertent Reporting of Gross Electrical Energy Data When Net Electrical Energy Data Was Required1999-08-13013 August 1999 Forwards Monthly Operating Rept for July 1999 for Fcs,Unit 1,per Plant TS 5.9.1.Revised Rept for June 1999,encl Due to Inadvertent Reporting of Gross Electrical Energy Data When Net Electrical Energy Data Was Required 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 ML20211B7881999-08-10010 August 1999 Transmits Summary of Two Meetings with Risk-Informed TS Task Force in Rockville,Md on 990514 & 0714 ML20210N1991999-08-0909 August 1999 Informs That as Result of NRC Review of Licensee Responses to GL 92-01,rev 1,suppl 1 & Suppl 1 Rai,Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2 ML20210P8991999-08-0909 August 1999 Forwards FEMA Final Rept for 990630 Offsite Medical Drill. No Deficiencies or Areas Requiring Corrective Action Identified During Drill LIC-99-0072, Requests That Licensee 990526 Application for Amend & Related Documents,Be Withdrawn.Util Requests That NRC Review Attached Responses to Ensure That Upcoming Submittal Will Be Acceptable1999-08-0606 August 1999 Requests That Licensee 990526 Application for Amend & Related Documents,Be Withdrawn.Util Requests That NRC Review Attached Responses to Ensure That Upcoming Submittal Will Be Acceptable 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 LIC-99-0068, Documents OPPD Response to Requested Actions in GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal1999-08-0202 August 1999 Documents OPPD Response to Requested Actions in GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal LIC-99-0040, Forwards Rept of Changes,Tests & Experiments Performed Per 10CFR50.59 & Revs to USAR Other than Those Resulting from 10CFR50.59.Encl Info Covers Period of 981101-9903311999-07-28028 July 1999 Forwards Rept of Changes,Tests & Experiments Performed Per 10CFR50.59 & Revs to USAR Other than Those Resulting from 10CFR50.59.Encl Info Covers Period of 981101-990331 LIC-99-0070, Requests Reinstatement of SRO License Previously Held by K Grant-Leanna (License SOP-43569).Operator Is Current on All requalification-training Topics & Has Completed License Reactivation Program1999-07-28028 July 1999 Requests Reinstatement of SRO License Previously Held by K Grant-Leanna (License SOP-43569).Operator Is Current on All requalification-training Topics & Has Completed License Reactivation Program ML20210G1851999-07-27027 July 1999 Forwards Amend 192 to License DPR-40 & Safety Evaluation. Amend Consists of Changes to TS in Response to 990129 Application LIC-99-0057, Forwards 1999/2000 Statement of Cash Flow from Operations, as Guarantee of Payment of Deferred Premiums for Period of 990630 to 0006301999-07-26026 July 1999 Forwards 1999/2000 Statement of Cash Flow from Operations, as Guarantee of Payment of Deferred Premiums for Period of 990630 to 000630 ML20216D6951999-07-26026 July 1999 Informs That During 990614 Review of Fort Calhoun,Nrc Noted That Drill/Exercise Performance Indicator Has Entered White Band,Which Indicated Level of Licensee Performance That Warranted Increased Regulatory Response LIC-99-0071, Informs That Util Expects to Submit Four Licensing Actions in Each of Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates1999-07-26026 July 1999 Informs That Util Expects to Submit Four Licensing Actions in Each of Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates LIC-99-0069, Forwards non-proprietary & Proprietary Epips,Including FC- Epf Index Page & Page 2 ,rev 0 to FC- EPF-42,EPIP Index (Page 1 of 2) & EPIP-OSC-2, Cover & Pages 19 & 20.Proprietary Encls Withheld1999-07-23023 July 1999 Forwards non-proprietary & Proprietary Epips,Including FC- Epf Index Page & Page 2 ,rev 0 to FC- EPF-42,EPIP Index (Page 1 of 2) & EPIP-OSC-2, Cover & Pages 19 & 20.Proprietary Encls Withheld ML20210D9471999-07-22022 July 1999 Forwards Amend 191 to DPR-40 & Safety Evaluation.Amend Authorizes Rev to Licensing Basis as Described in Updated Safety Analysis Rept to Incorporate Mod for Overriding Containment Isolation ML20210C2041999-07-20020 July 1999 Discusses 990715 Mgt Meeting with Midamerican Energy Co Re risk-informed Baseline Insp Program Recently Implemented at Cooper & Fort Calhoun Stations ML20210C0201999-07-20020 July 1999 Informs That Due to Administrative Oversight,Revised TS Bases Provided in to Fort Calhoun Station Were Not Dated.Dated Pages to TS Encl ML20210B0761999-07-19019 July 1999 Forwards Insp Rept 50-285/99-05 on 990601-0710.No Violations Noted ML20209G5351999-07-15015 July 1999 Informs That Staff Incorporated Rev of TS Bases Provided by Licensee Ltr Dtd 980904,into Fort Calhoun Station Tss. Staff Finds Revs to Associated TS Bases to Be Acceptable. Revised Bases Pages Encl ML20209F5741999-07-12012 July 1999 Ack Receipt of Transmitting Exercise Scenario for Upcoming Biennial Offsite Exercise Scheduled 990810.Scenario Reviewed & Acceptable ML20209G5591999-07-0808 July 1999 Forwards Detailed Listing of Staff Concerns with C-E TR NPSD-683,rev 3 & RCS pressure-temperature Limits Rept LIC-99-0066, Forwards Fort Calhoun Performance Indicators Rept for May 19991999-07-0707 July 1999 Forwards Fort Calhoun Performance Indicators Rept for May 1999 LIC-99-0060, Forwards Proprietary & non-proprietary Revs to Fort Calhoun Station Emergency Plan Forms (Epf),Including FC-EPF,FC-EPF-1 & FC-EPF-17.Proprietary Info Withheld1999-07-0101 July 1999 Forwards Proprietary & non-proprietary Revs to Fort Calhoun Station Emergency Plan Forms (Epf),Including FC-EPF,FC-EPF-1 & FC-EPF-17.Proprietary Info Withheld ML20209D8771999-07-0101 July 1999 Summarizes 990630 Public Meeting Conducted in Fort Calhoun, Ne Re Results of Completed Culture Survey,Plant Status, Facility Activities Re New NRC Oversight Process & License Renewal.Licensee Handout & Attendance List Encl LIC-99-0063, Requests Termination of License Held by MR Anderson (License SOP 43812) for Fort Calhoun Station.Anderson Ended Employment Effective 9906241999-07-0101 July 1999 Requests Termination of License Held by MR Anderson (License SOP 43812) for Fort Calhoun Station.Anderson Ended Employment Effective 990624 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARLIC-99-0098, Forwards Revs to Fort Calhoun Station EPIPs & Emergency Planning Forms Manual.Document Update Instructions & Summary of Changes Are Included on Confirmation of Transmittal Form Attached1999-10-15015 October 1999 Forwards Revs to Fort Calhoun Station EPIPs & Emergency Planning Forms Manual.Document Update Instructions & Summary of Changes Are Included on Confirmation of Transmittal Form Attached LIC-99-0091, Submits Suppl Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, in Response to NRC RAI During 990817 Meeting with Util1999-10-0808 October 1999 Submits Suppl Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, in Response to NRC RAI During 990817 Meeting with Util LIC-99-0097, Forwards Final RO & SRO License Applications,As Requested in ,Certifying That Training Completed for Six RO Candidates,Three Instant SRO Candidates & One Upgrade SRO Candidate.Without Encls1999-10-0707 October 1999 Forwards Final RO & SRO License Applications,As Requested in ,Certifying That Training Completed for Six RO Candidates,Three Instant SRO Candidates & One Upgrade SRO Candidate.Without Encls LIC-99-0089, Forwards Preliminary License Exam Applications for Six Ros, Three Instant SROs & One Upgrade SRO Candidate,Per Preparation for Ro/Sro Licensing Exams to Be Administered on 991025-29 at Fcs.Without Encl1999-09-24024 September 1999 Forwards Preliminary License Exam Applications for Six Ros, Three Instant SROs & One Upgrade SRO Candidate,Per Preparation for Ro/Sro Licensing Exams to Be Administered on 991025-29 at Fcs.Without Encl LIC-99-0092, Forwards Revised Epips,Including non-proprietary Rev 8 to EPIP-RR-66,rev 3 to EPIP-EOF-24,rev 14 to RERP-Section F,Rev 11 to RERP-Section I,Rev 12 to RERP-Appendix C & RERP & EPIP Indexes.Proprietary Rev 3 to EPIP-EOF-24,pages 1-3,encl1999-09-21021 September 1999 Forwards Revised Epips,Including non-proprietary Rev 8 to EPIP-RR-66,rev 3 to EPIP-EOF-24,rev 14 to RERP-Section F,Rev 11 to RERP-Section I,Rev 12 to RERP-Appendix C & RERP & EPIP Indexes.Proprietary Rev 3 to EPIP-EOF-24,pages 1-3,encl LIC-99-0087, Forwards Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams1999-09-17017 September 1999 Forwards Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams LIC-99-0088, Forwards Documentation Associated with SRO License Renewal Application for Cd Darrow.Applicant Has Satisfactorily Completed Applicable FCS Requalification Program.Without Encls1999-09-16016 September 1999 Forwards Documentation Associated with SRO License Renewal Application for Cd Darrow.Applicant Has Satisfactorily Completed Applicable FCS Requalification Program.Without Encls LIC-99-0080, Requests Approval of Attached Rev to QAP Contained in App a of Usar,Revising Commitment Re Qualifications of Personnel Performing Quality Control Insps of Activities Affecting Fire Protection1999-09-16016 September 1999 Requests Approval of Attached Rev to QAP Contained in App a of Usar,Revising Commitment Re Qualifications of Personnel Performing Quality Control Insps of Activities Affecting Fire Protection LIC-99-0085, Forwards Revs 18 & 19 to TDB-VI, COLR for FCS Unit 1,IAW TS 5.9.5.Rev 18 Resulted from Reevaluation of Cycle 18 Setpoints & Rev 19 Resulted from Changes in Cycle 19 Refueling Boron Concentrations & Addition of TS Parameters1999-09-0808 September 1999 Forwards Revs 18 & 19 to TDB-VI, COLR for FCS Unit 1,IAW TS 5.9.5.Rev 18 Resulted from Reevaluation of Cycle 18 Setpoints & Rev 19 Resulted from Changes in Cycle 19 Refueling Boron Concentrations & Addition of TS Parameters LIC-99-0083, Requests Participation in internet-based Y2K Early Warning Sys (Yews),As Noted in NRC Info Notice 99-25.Plant Point of Contact Will Be JW Chase1999-09-0202 September 1999 Requests Participation in internet-based Y2K Early Warning Sys (Yews),As Noted in NRC Info Notice 99-25.Plant Point of Contact Will Be JW Chase LIC-99-0077, Forwards non-proprietary & Proprietary Revised EP Forms, Including FC-EPF Index,Pp 1 Dtd 990805,pp 2 Dtd 990812,rev 11 to FC-EPF-9,rev 12 to FC-EPF-10,rev 9 to FC-EPF-11,rev 2 to FC-EPF-12 & Rev 3 to FC-EPF-38.Proprietary EP Withheld1999-08-27027 August 1999 Forwards non-proprietary & Proprietary Revised EP Forms, Including FC-EPF Index,Pp 1 Dtd 990805,pp 2 Dtd 990812,rev 11 to FC-EPF-9,rev 12 to FC-EPF-10,rev 9 to FC-EPF-11,rev 2 to FC-EPF-12 & Rev 3 to FC-EPF-38.Proprietary EP Withheld LIC-99-0078, Forwards Rev 16C to FCS Site Security Plan,Per Provisions of 10CFR50.54(p).Rev Has Been Determined Not to Degrade Effectiveness of Current Security Plans.List of Changes, Provided.Encl Withheld1999-08-20020 August 1999 Forwards Rev 16C to FCS Site Security Plan,Per Provisions of 10CFR50.54(p).Rev Has Been Determined Not to Degrade Effectiveness of Current Security Plans.List of Changes, Provided.Encl Withheld LIC-99-0076, Forwards fitness-for-duty Program Performance Data for Six Month Period from Jan-June 1999,IAW 10CFR26.71(d)1999-08-20020 August 1999 Forwards fitness-for-duty Program Performance Data for Six Month Period from Jan-June 1999,IAW 10CFR26.71(d) LIC-99-0075, Forwards Monthly Operating Rept for July 1999 for Fcs,Unit 1,per Plant TS 5.9.1.Revised Rept for June 1999,encl Due to Inadvertent Reporting of Gross Electrical Energy Data When Net Electrical Energy Data Was Required1999-08-13013 August 1999 Forwards Monthly Operating Rept for July 1999 for Fcs,Unit 1,per Plant TS 5.9.1.Revised Rept for June 1999,encl Due to Inadvertent Reporting of Gross Electrical Energy Data When Net Electrical Energy Data Was Required LIC-99-0072, Requests That Licensee 990526 Application for Amend & Related Documents,Be Withdrawn.Util Requests That NRC Review Attached Responses to Ensure That Upcoming Submittal Will Be Acceptable1999-08-0606 August 1999 Requests That Licensee 990526 Application for Amend & Related Documents,Be Withdrawn.Util Requests That NRC Review Attached Responses to Ensure That Upcoming Submittal Will Be Acceptable LIC-99-0068, Documents OPPD Response to Requested Actions in GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal1999-08-0202 August 1999 Documents OPPD Response to Requested Actions in GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal LIC-99-0040, Forwards Rept of Changes,Tests & Experiments Performed Per 10CFR50.59 & Revs to USAR Other than Those Resulting from 10CFR50.59.Encl Info Covers Period of 981101-9903311999-07-28028 July 1999 Forwards Rept of Changes,Tests & Experiments Performed Per 10CFR50.59 & Revs to USAR Other than Those Resulting from 10CFR50.59.Encl Info Covers Period of 981101-990331 LIC-99-0070, Requests Reinstatement of SRO License Previously Held by K Grant-Leanna (License SOP-43569).Operator Is Current on All requalification-training Topics & Has Completed License Reactivation Program1999-07-28028 July 1999 Requests Reinstatement of SRO License Previously Held by K Grant-Leanna (License SOP-43569).Operator Is Current on All requalification-training Topics & Has Completed License Reactivation Program LIC-99-0071, Informs That Util Expects to Submit Four Licensing Actions in Each of Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates1999-07-26026 July 1999 Informs That Util Expects to Submit Four Licensing Actions in Each of Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates LIC-99-0057, Forwards 1999/2000 Statement of Cash Flow from Operations, as Guarantee of Payment of Deferred Premiums for Period of 990630 to 0006301999-07-26026 July 1999 Forwards 1999/2000 Statement of Cash Flow from Operations, as Guarantee of Payment of Deferred Premiums for Period of 990630 to 000630 LIC-99-0069, Forwards non-proprietary & Proprietary Epips,Including FC- Epf Index Page & Page 2 ,rev 0 to FC- EPF-42,EPIP Index (Page 1 of 2) & EPIP-OSC-2, Cover & Pages 19 & 20.Proprietary Encls Withheld1999-07-23023 July 1999 Forwards non-proprietary & Proprietary Epips,Including FC- Epf Index Page & Page 2 ,rev 0 to FC- EPF-42,EPIP Index (Page 1 of 2) & EPIP-OSC-2, Cover & Pages 19 & 20.Proprietary Encls Withheld LIC-99-0066, Forwards Fort Calhoun Performance Indicators Rept for May 19991999-07-0707 July 1999 Forwards Fort Calhoun Performance Indicators Rept for May 1999 LIC-99-0060, Forwards Proprietary & non-proprietary Revs to Fort Calhoun Station Emergency Plan Forms (Epf),Including FC-EPF,FC-EPF-1 & FC-EPF-17.Proprietary Info Withheld1999-07-0101 July 1999 Forwards Proprietary & non-proprietary Revs to Fort Calhoun Station Emergency Plan Forms (Epf),Including FC-EPF,FC-EPF-1 & FC-EPF-17.Proprietary Info Withheld LIC-99-0063, Requests Termination of License Held by MR Anderson (License SOP 43812) for Fort Calhoun Station.Anderson Ended Employment Effective 9906241999-07-0101 July 1999 Requests Termination of License Held by MR Anderson (License SOP 43812) for Fort Calhoun Station.Anderson Ended Employment Effective 990624 LIC-99-0061, Submits Response to GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. Disclosure Encl1999-06-30030 June 1999 Submits Response to GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. Disclosure Encl LIC-99-0056, Forwards Rev 2 to EA-FC-90-082, Potential Over- Pressurization of Containment Penetration Piping Following Main Steam Line Break in Containment, to Facilitate Review of Response to GL 96-061999-06-22022 June 1999 Forwards Rev 2 to EA-FC-90-082, Potential Over- Pressurization of Containment Penetration Piping Following Main Steam Line Break in Containment, to Facilitate Review of Response to GL 96-06 LIC-99-0059, Provides Formal Notification to NRC of Plans Re Renewal of Operating License DPR-40 for Unit 1.Util Will Continue to Inform NRC of Decisions Relative to License Renewal in Order to Facilitate NRC Resource Planning1999-06-22022 June 1999 Provides Formal Notification to NRC of Plans Re Renewal of Operating License DPR-40 for Unit 1.Util Will Continue to Inform NRC of Decisions Relative to License Renewal in Order to Facilitate NRC Resource Planning LIC-99-0054, Forwards Manual Containing Description of Fort Calhoun Station 1999 Emergency Preparedness Exercise, to Be Conducted on 9908101999-06-0808 June 1999 Forwards Manual Containing Description of Fort Calhoun Station 1999 Emergency Preparedness Exercise, to Be Conducted on 990810 LIC-99-0052, Forwards Fort Calhoun Station Performance Indicators, for Apr 19991999-06-0202 June 1999 Forwards Fort Calhoun Station Performance Indicators, for Apr 1999 LIC-99-0045, Forwards Application for Amend to License DPR-40,proposing to Relocate pressure-temp Curves,Predicated NDTT Shift Curve & LTOP Limits & Values from FCS TS to OPPD Controlled Document.Rev 3 to TR CE NPSD-683 & Supporting TSs Encl1999-05-26026 May 1999 Forwards Application for Amend to License DPR-40,proposing to Relocate pressure-temp Curves,Predicated NDTT Shift Curve & LTOP Limits & Values from FCS TS to OPPD Controlled Document.Rev 3 to TR CE NPSD-683 & Supporting TSs Encl LIC-99-0046, Forwards Scope & Objectives for Plant 1999 EP Exercise to Be Conducted 990810.Info Provided Ninety Days Prior to Exercise IAW Established Stds.Complete Scenario Package Will Be Submitted to NRC & FEMA Sixty Days Prior to Exercise1999-05-12012 May 1999 Forwards Scope & Objectives for Plant 1999 EP Exercise to Be Conducted 990810.Info Provided Ninety Days Prior to Exercise IAW Established Stds.Complete Scenario Package Will Be Submitted to NRC & FEMA Sixty Days Prior to Exercise LIC-99-0028, Provides NRC with Status of Corrective Actions Re Plant Fire Protection Program,As Result of NRC Insp Rept 50-285/97-06. Mod to MOV Is in Design Phase & Currently Scheduled for Installation During 1999 Refueling Outage1999-04-30030 April 1999 Provides NRC with Status of Corrective Actions Re Plant Fire Protection Program,As Result of NRC Insp Rept 50-285/97-06. Mod to MOV Is in Design Phase & Currently Scheduled for Installation During 1999 Refueling Outage LIC-99-0036, Submits Annual Rept of 1998 Loca/Eccs Models Per 10CFR50.46. Small & Large Break LOCA PCT Values Continue to Remain Less than 10CFR50.46(b)(1) Acceptance Criterion of 2200 F1999-04-28028 April 1999 Submits Annual Rept of 1998 Loca/Eccs Models Per 10CFR50.46. Small & Large Break LOCA PCT Values Continue to Remain Less than 10CFR50.46(b)(1) Acceptance Criterion of 2200 F LIC-99-0043, Forwards Fort Calhoun Performance Indicators Rept for Mar 19991999-04-28028 April 1999 Forwards Fort Calhoun Performance Indicators Rept for Mar 1999 LIC-99-0041, Informs That Util Has Implemented Necessary Actions Associated with Open TACs M92926,M99885,M98619 & M98933 Issues.Items Considered Ready to Close,Subject to Agreement by Nrc/Nrr Project Manager1999-04-27027 April 1999 Informs That Util Has Implemented Necessary Actions Associated with Open TACs M92926,M99885,M98619 & M98933 Issues.Items Considered Ready to Close,Subject to Agreement by Nrc/Nrr Project Manager LIC-99-0037, Responds to NRC 990126 RAI Re GL 97-01, Degradation of CRDM Nozzle & Other Vessel Closure Head Penetrations. Response Confirms with NRC & NEI Recommendations to Utilize Generic Industry Response to Issue1999-04-26026 April 1999 Responds to NRC 990126 RAI Re GL 97-01, Degradation of CRDM Nozzle & Other Vessel Closure Head Penetrations. Response Confirms with NRC & NEI Recommendations to Utilize Generic Industry Response to Issue LIC-99-0039, Forwards Diskette Containing 1998 Annual Rept of Individual Monitoring for Fcs,Per 10CFR20.2206.Diskette Follows File Structure of Reg Guide 8.7,App a1999-04-20020 April 1999 Forwards Diskette Containing 1998 Annual Rept of Individual Monitoring for Fcs,Per 10CFR20.2206.Diskette Follows File Structure of Reg Guide 8.7,App a LIC-99-0038, Forwards 1998 Radiological Environ Operating Rept & Annual Rept for TS Section 5.9.4.A for 980101-981231, for Fcs,Unit 1.Rept Is Presented in Format Outlined in Reg Guide 1.21,rev 1.Results of Quarterly Dose Calculations,Encl1999-04-20020 April 1999 Forwards 1998 Radiological Environ Operating Rept & Annual Rept for TS Section 5.9.4.A for 980101-981231, for Fcs,Unit 1.Rept Is Presented in Format Outlined in Reg Guide 1.21,rev 1.Results of Quarterly Dose Calculations,Encl LIC-99-0033, Forwards Addl Info to Support NRC Review of Response to GL 97-04, Assurance of Sufficient Net Positive Suction Head for ECC & Containment Heat Removal Pumps1999-04-15015 April 1999 Forwards Addl Info to Support NRC Review of Response to GL 97-04, Assurance of Sufficient Net Positive Suction Head for ECC & Containment Heat Removal Pumps LIC-99-0035, Forwards Proprietary Rev 2b to FCS Emergency Plan Form FC-EPF-17 & non-proprietary Revised FC-EPF Forms Index,Rev 5a to FC-EPF-1 & Rev 4 to FC-EPF-15.Proprietary Info Withheld from Public Disclosure1999-04-0808 April 1999 Forwards Proprietary Rev 2b to FCS Emergency Plan Form FC-EPF-17 & non-proprietary Revised FC-EPF Forms Index,Rev 5a to FC-EPF-1 & Rev 4 to FC-EPF-15.Proprietary Info Withheld from Public Disclosure LIC-99-0031, Forwards Application for Amend to License DPR-40,increasing Min Required RCS Flow Rate & Changing SRs for RCS Flow Rate1999-03-31031 March 1999 Forwards Application for Amend to License DPR-40,increasing Min Required RCS Flow Rate & Changing SRs for RCS Flow Rate LIC-99-0032, Forwards Fort Calhoun Station Performance Indicators, for Feb 19991999-03-31031 March 1999 Forwards Fort Calhoun Station Performance Indicators, for Feb 1999 LIC-99-0029, Forwards Response to 990226 RAI for GL 97-05, SG Tube Insp Techniques1999-03-30030 March 1999 Forwards Response to 990226 RAI for GL 97-05, SG Tube Insp Techniques LIC-99-0027, Forwards Rev 16B to FCS Site Security Plan,Rev 2 to FCS Safeguards Contingency Plan & Rev 6 to FCS Training & Qualification Plan.Encls Withheld1999-03-16016 March 1999 Forwards Rev 16B to FCS Site Security Plan,Rev 2 to FCS Safeguards Contingency Plan & Rev 6 to FCS Training & Qualification Plan.Encls Withheld LIC-99-0022, Agrees with NRC Request That Proposed Wording of USAR Section 9.2.5 Included in Be Revised to Indicate That Use of Letdown Sys Following Transient.Revs of Appropriate Pages to ,Encl1999-03-12012 March 1999 Agrees with NRC Request That Proposed Wording of USAR Section 9.2.5 Included in Be Revised to Indicate That Use of Letdown Sys Following Transient.Revs of Appropriate Pages to ,Encl LIC-99-0024, Informs That OPPD Withdraws Applications Submitted by 950626 &1999-03-12012 March 1999 Informs That OPPD Withdraws Applications Submitted by 950626 & LIC-99-0016, Requests Relief from Certain Requirements of Paragraphs IWA- 5242(a) & IWA-5250(a)(2) of ASME Section Xi,For Remainder of Third 10-year Insp Interval on Basis of Attached Proposed Alternatives1999-03-0101 March 1999 Requests Relief from Certain Requirements of Paragraphs IWA- 5242(a) & IWA-5250(a)(2) of ASME Section Xi,For Remainder of Third 10-year Insp Interval on Basis of Attached Proposed Alternatives LIC-99-0023, Forwards FCS Performance Indicators for Jan 19991999-03-0101 March 1999 Forwards FCS Performance Indicators for Jan 1999 LIC-99-0020, Forwards non-proprietary EPIP Index,Page 1 of 2,rev 31 to EPIP-OSC-1, Emergency Classification, FC-EPF Forms Index, Page 1 of 2, Rev 8 to FC-EPF-14 & Rev 15 to RERP-Appendix a, Rerp. Proprietary Version of Rev 15 to RERP, Included1999-02-24024 February 1999 Forwards non-proprietary EPIP Index,Page 1 of 2,rev 31 to EPIP-OSC-1, Emergency Classification, FC-EPF Forms Index, Page 1 of 2, Rev 8 to FC-EPF-14 & Rev 15 to RERP-Appendix a, Rerp. Proprietary Version of Rev 15 to RERP, Included LIC-99-0011, Forwards 1999 Biennial Decommissioning Funding Status Rept for Fort Calhoun Station,Unit 11999-02-0505 February 1999 Forwards 1999 Biennial Decommissioning Funding Status Rept for Fort Calhoun Station,Unit 1 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARLIC-90-0705, Forwards Suppl to 900906 Application for Amend to License DPR-40,making Administrative Changes to Tech Specs. Justification & NSHC Analysis for Changes Encl1990-09-17017 September 1990 Forwards Suppl to 900906 Application for Amend to License DPR-40,making Administrative Changes to Tech Specs. Justification & NSHC Analysis for Changes Encl LIC-90-0745, Responds to NRC Re Violations Noted in Insp Rept 50-285/90-32.Corrective Actions:Enhancements Made to Facility Mod Procedures & Engineering Instructions1990-09-17017 September 1990 Responds to NRC Re Violations Noted in Insp Rept 50-285/90-32.Corrective Actions:Enhancements Made to Facility Mod Procedures & Engineering Instructions LIC-90-0717, Forwards Operator Licensing Natl Exam Schedule for FY91 Through FY94,per Generic Ltr 90-071990-09-13013 September 1990 Forwards Operator Licensing Natl Exam Schedule for FY91 Through FY94,per Generic Ltr 90-07 LIC-90-0716, Forwards Ref Matl for Preparing Reactor Operator Licensing Exams Scheduled for 901113.W/o Encls1990-09-13013 September 1990 Forwards Ref Matl for Preparing Reactor Operator Licensing Exams Scheduled for 901113.W/o Encls LIC-90-0740, Forwards Rev 0 to EA-FC-90-062, Diesel Generator Upper Temp Operating Limits,Engineering Analysis. Both Generators Operable & Will Perform Safety Function Consistent W/ Specified Ambient Air Temp Limits1990-09-12012 September 1990 Forwards Rev 0 to EA-FC-90-062, Diesel Generator Upper Temp Operating Limits,Engineering Analysis. Both Generators Operable & Will Perform Safety Function Consistent W/ Specified Ambient Air Temp Limits LIC-90-0741, Submits Plan for Replacement of Remaining Two Control Element Assemblies Which Contain All B4C Pellets1990-09-0707 September 1990 Submits Plan for Replacement of Remaining Two Control Element Assemblies Which Contain All B4C Pellets LIC-90-0440, Forwards Application for Amend to License DPR-40,revising Tech Specs to Make Administrative Changes1990-09-0606 September 1990 Forwards Application for Amend to License DPR-40,revising Tech Specs to Make Administrative Changes LIC-90-0667, Responds to NRC Re Violations Noted in Insp Rept 50-285/90-30 Re Inadequate post-maint Testing Instructions. Corrective Actions:Maint Work Order Written & Performed to Tighten Actuator Diaphragm Bolts on 9005291990-09-0404 September 1990 Responds to NRC Re Violations Noted in Insp Rept 50-285/90-30 Re Inadequate post-maint Testing Instructions. Corrective Actions:Maint Work Order Written & Performed to Tighten Actuator Diaphragm Bolts on 900529 LIC-90-0712, Forwards fitness-for-duty Program Performance Data for Period of Jan-June 19901990-08-31031 August 1990 Forwards fitness-for-duty Program Performance Data for Period of Jan-June 1990 LIC-90-0683, Forwards Responses to Recommendations in SALP Rept 50-285/90-24.Util Currently Developing Corporate Level Procedure to Provide Guidance on Operability Determinations1990-08-27027 August 1990 Forwards Responses to Recommendations in SALP Rept 50-285/90-24.Util Currently Developing Corporate Level Procedure to Provide Guidance on Operability Determinations LIC-90-0627, Responds to Generic Ltr 89-19 Re USI A-47 on Safety Implications of Control Sys in Lwrs.Util Reconsidered Original Commitment Noted in & Does Not Plan to Implement Automatic Steam Generator Overfill Protection1990-08-22022 August 1990 Responds to Generic Ltr 89-19 Re USI A-47 on Safety Implications of Control Sys in Lwrs.Util Reconsidered Original Commitment Noted in & Does Not Plan to Implement Automatic Steam Generator Overfill Protection LIC-90-0670, Forwards Fort Calhoun Station Performance Indicators for Jul 19901990-08-22022 August 1990 Forwards Fort Calhoun Station Performance Indicators for Jul 1990 LIC-90-0662, Forwards Rev 1 to Monthly Operating Rept for June 1990 for Fort Calhoun Station Unit 1,for Proper Distribution1990-08-22022 August 1990 Forwards Rev 1 to Monthly Operating Rept for June 1990 for Fort Calhoun Station Unit 1,for Proper Distribution LIC-90-0673, Retransmits Original Response of Util Responding to Violations Noted in Insp Rept 50-285/90-25.W/o Encl1990-08-22022 August 1990 Retransmits Original Response of Util Responding to Violations Noted in Insp Rept 50-285/90-25.W/o Encl LIC-90-0636, Forwards Results of 1990 Refueling Outage Inservice Insp & Rept for Repairs or Replacement1990-08-13013 August 1990 Forwards Results of 1990 Refueling Outage Inservice Insp & Rept for Repairs or Replacement LIC-90-0591, Provides Info Re RCS Leak Detection Sys to Detect Leaks from Primary Piping Sys to Containment,Per Generic Ltr 84-04 & USI A-21990-08-13013 August 1990 Provides Info Re RCS Leak Detection Sys to Detect Leaks from Primary Piping Sys to Containment,Per Generic Ltr 84-04 & USI A-2 LIC-90-0631, Responds to Generic Ltr 89-15 Re NRC Solicitation of Utils to Voluntarily Participate Emergency Response Data Sys Program.Util Prepared to Work W/Nrc to Investigate Actions Necessary to Implement Program at Plant1990-08-0808 August 1990 Responds to Generic Ltr 89-15 Re NRC Solicitation of Utils to Voluntarily Participate Emergency Response Data Sys Program.Util Prepared to Work W/Nrc to Investigate Actions Necessary to Implement Program at Plant LIC-90-0485, Confirms Interim Resolution of Identified Problem Associated W/Use of Input Variable for PORV Opening Time in Calculation of Low Temp Overpressure Protection Sys Setpoints1990-08-0707 August 1990 Confirms Interim Resolution of Identified Problem Associated W/Use of Input Variable for PORV Opening Time in Calculation of Low Temp Overpressure Protection Sys Setpoints LIC-90-0645, Forwards Application for Amend to License DPR-40,making Typo & Administrative Corrections to Tech Specs1990-08-0202 August 1990 Forwards Application for Amend to License DPR-40,making Typo & Administrative Corrections to Tech Specs LIC-90-0647, Informs That Tl Patterson Assumed Position of Plant Manager, Effective 9008011990-08-0101 August 1990 Informs That Tl Patterson Assumed Position of Plant Manager, Effective 900801 LIC-90-0659, Advises of Plans to Achieve Operability of Third Auxiliary Feedwater Pump by 9008101990-07-31031 July 1990 Advises of Plans to Achieve Operability of Third Auxiliary Feedwater Pump by 900810 LIC-90-0449, Forwards Rev 1 to EA-FC-90-037, Control Room Habitability Evaluation for NUREG-0737,Item III.D.3.41990-07-20020 July 1990 Forwards Rev 1 to EA-FC-90-037, Control Room Habitability Evaluation for NUREG-0737,Item III.D.3.4 LIC-90-0588, Provides Notification of SPDS Implementation at Plant,Per1990-07-19019 July 1990 Provides Notification of SPDS Implementation at Plant,Per LIC-90-0585, Forwards Simulator Malfunction Cause & Effects1990-07-18018 July 1990 Forwards Simulator Malfunction Cause & Effects ML20055D9351990-07-0202 July 1990 Forwards Fort Calhoun Station Performance Indicators,May 1990 LIC-90-0564, Forwards Updated, Position Paper Re Seismic Capability of Welded Steel Piping. Documentation Demonstrates No Generic Thermal or Seismic Safety Issue Exists for Welded Steel Small Diameter Piping at Facility1990-06-29029 June 1990 Forwards Updated, Position Paper Re Seismic Capability of Welded Steel Piping. Documentation Demonstrates No Generic Thermal or Seismic Safety Issue Exists for Welded Steel Small Diameter Piping at Facility LIC-90-0522, Forwards Revised Response to Notice of Deviation from Insp Rept 50-285/90-02.Corrective Actions:Cable Trays 5-4A & 5-4B Reworked to Neatly Tie Down Cables & Remove Crossovers & Engineering Instruction GEI-9 on Electrical Sys Updated1990-06-29029 June 1990 Forwards Revised Response to Notice of Deviation from Insp Rept 50-285/90-02.Corrective Actions:Cable Trays 5-4A & 5-4B Reworked to Neatly Tie Down Cables & Remove Crossovers & Engineering Instruction GEI-9 on Electrical Sys Updated LIC-90-0566, Forwards 1990/1991 Statement of Cash Flow from Operations. Cash Flow Statement Deviates from Reg Guide 9.4 Due to Util Being Political Subdivision of State of Ne1990-06-29029 June 1990 Forwards 1990/1991 Statement of Cash Flow from Operations. Cash Flow Statement Deviates from Reg Guide 9.4 Due to Util Being Political Subdivision of State of Ne LIC-90-0488, Responds to Request for Info Re Status of Implementation of Generic Safety Issue Requirements1990-06-29029 June 1990 Responds to Request for Info Re Status of Implementation of Generic Safety Issue Requirements LIC-90-0446, Forwards Application for Amend to License DPR-40,to Restrict Containment Spray Use as Backup for Shutdown Cooling in Tech Specs1990-06-28028 June 1990 Forwards Application for Amend to License DPR-40,to Restrict Containment Spray Use as Backup for Shutdown Cooling in Tech Specs LIC-90-0429, Forwards Application to Amend License DPR-40,adding Hydrogen Purge Sys to Tech Spec.Discussion,Justification & NSHC Analysis for Amend,Provided1990-06-28028 June 1990 Forwards Application to Amend License DPR-40,adding Hydrogen Purge Sys to Tech Spec.Discussion,Justification & NSHC Analysis for Amend,Provided ML20043J0611990-06-22022 June 1990 Responds to NRC 900523 Ltr Re Violations Noted in Insp Rept 50-285/90-29.Corrective Actions:Five Chemistry Technicians Provided Refresher Training on 900427 on Existing EPIP-EOF-6 ML20044A1631990-06-22022 June 1990 Requests That NRC Continue Processing Exemption Request for Configuration of wide-range Nuclear Instrumentation Cables in Fire Area 34B (Upper Electrical Penetration Room). Emergency Boration Added to Achieve Safe Shutdown ML20044A4961990-06-18018 June 1990 Forwards Supplemental Application for Amend to License DPR-40,making Administrative Corrections to Tech Specs. Procedure SO-G-70, Chemical Control Also Encl ML20043G2981990-06-14014 June 1990 Provides Followup to Special Rept on Inoperable Fire Barriers.Eight Inoperable Dampers Replaced Under Mod MR-FC-87-036, Fire Damper Replacement. Remaining 38 Dampers Found to Be Acceptable After Damper Design Review ML20043E6451990-06-0808 June 1990 Provides Supplemental Response to Generic Ltr 88-17 & NRC Bulletin 80-12.Plant Pumps & Suction Header Piping Not Originally Constructed for Use as Backup to LPSI Sys for Shutdown Cooling ML20043E0071990-06-0101 June 1990 Discusses Rev to Previous Commitments Made in Response to Requirements of 10CFR50,App R,Section Iii.G Re post-fire Repairs to Equipment Required for Cold Shutdown Following Fire in Fire Area 6 ML20043G3281990-06-0101 June 1990 Advises That Personnel at Util Reviewed Reactor Operator Licensing Exam Administered on 900529.Comments on Exam Encl ML20043C1811990-05-29029 May 1990 Advises That Util Has Cleaned RW HXs AC-1A & AC-1D & Will Clean AC-1C by 900615 ML20043B4981990-05-24024 May 1990 Updates Schedule for Submitting Results of PORV Block Valve Prototype Testing.Establishment of Test Parameters & Development of Test Procedures Tentatively Scheduled for Submittal in Nov 1990.Final Rept Expected by 910630 ML20043B5061990-05-23023 May 1990 Advises That Testing to Verify Operability of Hydrogen Purge Sys Successfully Completed During 1990 Refueling Outage.Proposed Tech Spec Rev Will Be Submitted by 900701 ML20043B4731990-05-23023 May 1990 Forwards Fort Calhoun Station Performance Indicators, Monthly Rept for Apr 1990 ML20043A7821990-05-18018 May 1990 Advises That Licensee Plans to Replace Valves HCV-247 & HCV-2988 During 1991 Refueling Outage Pending Successful Redesign & Retesting of New Valves by Valcor ML20043A8021990-05-18018 May 1990 Advises That Walkdown of safety-related Portions of Plant to Verify Accuracy of Fire Barrier Penetration Drawings & Surveillance Tests Will Be Completed by 900831 ML20043B0391990-05-16016 May 1990 Informs of Current Status of Post Accident Sampling Sys. Program Meets Min Requirements & Fulfills Conditions of Tech Spec 5.15 ML20042G9861990-05-11011 May 1990 Forwards Rev 0 to Interim Operability Criteria (Ioc) Criteria for Performing Safety Analysis for Interim Operability of Piping, Resulting from Discussion at 900306 & 0418 Meetings ML20042G9571990-05-10010 May 1990 Forwards Suppl to Application for Amend to License DPR-40, Making Administrative Clarification to Chloride Sampling Requirements.Standing Order Procedure S.O.-M-10, Tool Accountability Also Encl ML20042G9981990-05-10010 May 1990 Forwards Proprietary & Nonproprietary Rev 13 to EPIP-EOF-18, Offsite Radiological Surveys. ML20042H0021990-05-10010 May 1990 Fowards Fort Calhoun Station Performance Indicators,Mar 1990, Monthly Rept ML20042G4671990-05-0909 May 1990 Forwards Application for Amend to License DPR-40,changing Chemical & Vol Control Sys 1990-09-07
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Omaha Public Power District 1623 HARNEY I O M A H A. NEBRASKA 68102 a TELEPt CN E 536 4000 AREA CODE 402 June 26, 1981 Mr. Robert A. Clark, Chief U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Licensing Operating Reactors Branch No. 3 Washington, D.C. 205o5
Reference:
Docket Ilo. 50-285
Dear Mr. Clark:
Omaha Public Power District's response to the Commission's letter dated April 23, 1981, requesting resubmittal of an application for amendment to Facility License DPR-40, regarding iodine spiking is attcched. Please note that the District has addressed the specific concerns identified in the Commission's letter. As supported by the attached discussions, the District identified only two minor changes to the District's amendment application dated November 17, 1980. Accord-ingly, Attachments 5 and 6 are forwarded to replace pages 2-8 and 3-18 of the November 17, 1980, application.
Sincerely,
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W. C. Jones i / .- ,
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Attachments A'f , /
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. Attachment 1-Concern 1 The equilibrium primary coolant activity limit was essentially removed (i.e., the absence of a limit on the fraction of time that the plant can -
operate with an iodine " spike", in effect, negated the equilibrium value and substituted the " spike" maximum as the only real limit).
1
Response
The equilibrium primary coolant activity limit of 2 pCi/gm Dose Equiva-lent I-131 was included in the District's application for " Amendment of Operating License", dated November 17, 1980. The District had pre-viously. agreed to adding to the Technical Specifications a cumulative operating time limit of 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> above the 2 uCi/gm limit for any 12 consentive months. A revised page 2-8 of the Fort Calhoun Station 2
Technical. Specifications is attached (Attachment 5) for incorporation of
, this new requirement. It should be noted that the safety significance of this operating limit is not clear, since the protection of the public <
health and safety is insured following any design basis accident co-incident with iodine spiking, as demonstrated in Attachments 2 and 3.
Concern 2 The proposed sampling frequencies would not assure detection of iodine spikes in the reactor coolant.
- Response The sampling requirement assuring the detection of iodine spikes in the reactor coolant is Item 4 b), Table 4-4-4 of the STS. The District,-
after reviewing the 7-year operating history of the plant, proposed a Iampling requirement of one sample within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, following a thermal power change greater than 15 percent per hour. This rampling frequency has been changed in Attachment 6 to "within 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />". Using 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as an upper limit, instead of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> as called out in the STS, would provide operating f! txibility to accommodate changes of personnel between shifts.
Concern 3 The proposed equilibrium value was twice the STS value. i i Response
_ .The radioactivity of the reactor coolant for DOSE EQUIVALENT I-131 is 2.0 uCi/gm instead of 1.0 Ci/gm based upon the following:
~
- 1. The 2.0 pCi/gm limit represents approximately 1% failed fue1 as referenced in Table 11.1.5 of the FSAR, l
- 2. The 2.0 pCi/gm limit was determined using the methodology presented in NUREG-0017 for 10 CFR Part 50, Appendix I evalu-j ations, 1
- 3. The thyroid doses under accident conditions using 2.0 pCi/gm are less than 1% of 10 CFR Part 100 value, and
- 4. The proposed equilibrium value will not have any adverse effect on the transient limit of ~60 uCi/gm, specifically under
" Iodine Spiking" conditions. Please also see the response to Concern 5.
Concern 4 Your proposed Technical Specification as_ submitted could result in postulated off-site doses substantially in excess of 10 CFR Part 100 guidelines in the event of a steam generator tube rupture under iodine spik;.19 conditions (i.e. , with coolant activity up to 60 uCi/gm).
Response
The coolant concentration of 60 uCi/gm I-131 DOSE EQUIVALENT was not proposed by the District. This concentration was derived from Figure 3.4.1 of the STS. The District used the limiting value of 60 uCi/gm for 0-100 percent of rated thermal power instead of Figure 3.4.1 of the STS.
Figure 3.4.1 of the STS allows the radioactivity to exceed 60 uCi/gm whenever the reactor thermal power is less than 80%; therefore, the District's approach is more conservative than the STS.
Two analyses, pertinent to the proposed Technical Specifications, were performed to' determine the radiological conseque'nces of a' steam genera-tor tube rupture and a main steam line break 'with* the assumptions of coincident iodine spiking and an accident (initial activity of 2.0 uCi/gm I-131 DOSE EQUIVALENT and a final' activity, following the acci-dent, of 60 uCi/gm I-131 DOSE EQUIVALENT). The results of these analy-ses show that the thyroid and whole body doses would be well within the limits of 10 CFR Part 100.
The analysis and results of the main steam line break are provided as Attachment 2. The analysis and results of a steam generato tube rupture are provided as Attachment 3.
Concern 5 Your proposal to double the STS maximum (equilibrium *) value could lead to " spiked" coolant activity levels twice the transient limit of 60 uCi/gm.
Response
The reasons for proposing the equilibrium reactor coolant concentration of 2.0 uCi/gm instead of 1.0 pCi/gm are provided in response to Concern
- 3. After reviewing the plant operating data and the data presented in Table II of the NRC paper on " Iodine Spiking" and in Combustion En-gineering's topical report, "CENPD-180", the District finds no evidence
leading to the conclusion that by doubling the equilibrium values the i spiked coolant activity levels would be twice the transient limit of 60 I uCf/gm. Specifically, the iodine spikes experienced by the Fort Calhoun !
Station have been in the range of 12 to 16 pCi/gm with variations in I equilibrium leveu between 0.09 and 1.9 uCi/gm. Therefore, tbc District has observed no relationship between equilibrium and spike activity levels.
Concern 6 If you still believe that operation above steady state limit for up to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> is necessary and justified (versus 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per the STS), we request that you provide justification beyond that offered in your November 17, 1980 submittal.
Response
The longer duration (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer) to restore the coolant ac-tivity within acceptable values is based upon the specific design of the Fort Calhoun Station purification system. The coolant activity removal constant is dependent upon the purification system flow rates and the total _ volume of the reactor coolant system. Again, the bases for "48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> operation", as per the STS, are not known but it is believed that "48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />" are possibly based on a much more efficient purification system.
The justification for "up to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> operation" is provided in the form of graphs in Attachment 4. The values for coolant concentrations for April 1 through 6,1977, September 28 through October 3,1977, June 4 through 15, 1979, and December 14 through 24, 1979 have been normalized to postulate the initial steady state concentrations of 2.0 uCi/gm I-131 DOSE EQUIVALENT. The data for August 15 through 18, 1976 were not changed as the coolant concentrations were already in the vicinity of 2.0 uCi/gm. The time f atervals, shown on Attachment 4, ranging from 67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> are the durations from the beginning of a spike till the activity returns to steady-state conditions. There-fore, the District requests the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> limit be retained in the Amendment Application.
. Attachment 2 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE Physical Model The evaluation of the radiological consequences of a postulated steam generator tube rupture assumes a complete severance of a single steam generator tube while the reactor is operating at full rated power and a coincident loss of offsite power at the time of reactor trip. Occur-rence of the accident leads to an increase in contamination of the secondary system due to reactor coolant leakage through the tube break.
A reactor trip occurs z.utomatically as a result of low pressurizer pressure after the tube rupture occurs. The reactor trip automatically trips the turbine.
The resulting increase in radioactivity in the secondary system is detected by radiation monitors. A coincident loss of offsite station power causes closure of the turbine bypass valves to protect the con-denser. The steam generator pressure will increase rapidly, resulting in steam discharge as well as activity release through the main steam safety valves. Venting from the affected steam generator, i.e., the steam generator which experiences tube rupture, continues until the secondary system pressure is below the main steam safety valve setpoint.
At this time, the affected steam generator is effectively isolated, and thereafter, no steam or activity is assumed to be released from the affected steam generator. The remaining unaffected steam generators remove core decay heat by venting _ G._m through the main safety valves, atmospheric dump valve, and steam driven auxiliary turbine until cool- ,
down can be accomplished witn the shutdown cooling system.
o The analysis of the radiological consequences of a steam generator tube rupture considers the most severe release of secondary activity as well as reactor activity leaked from the tube break. The inventory of iodine and noble gas fission product activity available for release to the environment is a function of the primary-to-secondary coolant ledage rate, the percentage of defective fuel in the core, and the mass of steam discharged to the environment. Conservative assumptions are made for all these parameters.
In this evaluation, a case with coincident iodine spike which already exists due to a previous power transient was considered. The mathema-tical models, assumptions, and parameters used in this analysis were identical with the design basis SGTR without an iodine spike as des-cribed in the above paragraphs with the following exception:
The reactor coolant system inventory was assumed to be 60 pC/g dose equivalent Iodine 131. This 60 pC/g is the Technical Specification limit for full power operation following an iodine spike for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Assumptions and Conditions The major assumptions and parameters assumed in the analysis are item-ized in Table I.
RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE (Continue'd)
Assumpyons and Conditions (Continued)
The following assumptions and parameters are used to calculate the activity releases and offsite doses for a steam generator tube rupture (SGTR):
- 1. The reactor coo 7 ant system-equilibrium activity is 60 uC/cc DEC I-131.
- 2. The steam generator equilibrium activity for both steam gen-erators is assumed to be 0.1 uC/g dose equivalent I-131 prior to the accident.
- 3. 0ffsite power is lost; the main condenser is not available for steam relief via the turbine bypass system.
- 4. Following the accident, no additional steam and radioactivity are released to the~ environment when the shutdown cooling system is placed in operation.
- 5. There is no main condenser evacuation system release and no steam generator blowdown during the accident.
- 6. Only 1 steam generator is' affected.
- 7. The amount of noble gas activity released is equal to the amount present in the rdactor coolant discharged into the secondcry side following the tube ruptere. The amount of noble gas activity contained in the seccndary system is negligible in comparison.
- 8. Iodine activity released is based on the equilibrium activity present in the steam generators (0.1 uC/g dose equivalent I-131) and the amount of activity present in the reactor coolant discharged into the affected steam generator.
- 9. Thirty minutes after the accident, the affected unit is iso-lated. No steam and fission product activities are released from the affected steam generator thereafter.
- 10. The total amount of discharge of reactor coolant into the secondar i minutes)y system through the rupture is 42,300 pounds (in 30
- 11. A post-accident DF of 10 was used in the steam generator between the water and steam phases.
- 12. The. primary-to-secondary leakage of 8640 lbm/d (1.0 gal / min)
! is assumed to be applicable to the unaffected steam generator.
The portion of the noble gas activity from the primary-to-secondary leakage attributed to the unaffected steam generator l
is assumed to be released during the course of the accident.
RADIOLOGICAL CONSEQUENCES OF A STEAti GENERATOR TUBE RUPTURE (Continuad)
Assumptions and Conditions (Continued)
- 13. The amount of discharge of steam from'the affected steam gene'rator is assumed to be 3.55 x 106 pounds.
- 14. The activity released from the affected and unaffected steam generators is ininediately vented to the atmosphere. No credit for radioactive decay for isotopes in transit to dose points.
Uncertainties and Conservatisms The uncertainties and conservatisms in the assumptions used to. evaluate the radiological consequences of a steam generator tube rupture are as follows:
- 1. Reactor coolant equilibrium activities are based on 1% failed fuel, which is greater by a factor of 2 to 8 than that nor-mally observed in past PWR operation.
- 2. Steam generator equilibrium activity for both steam generators is assumed to be equal to the Technical Specification limit.
The Technical Specifica. ion limits are conservatively derived
- based on accidents suc.) as the SGTR.
- 3. Tube rupture of the 3. team generator is assumed to be a double-ended severance of a single steam generator tube. This is a conservative assumption since the steam generator tubes are constructed of highly ductile materials. The more probable mode of tube failure is 1 of minor leaks of undetermined origin. Activity in the secondary steam system is subject to continual surveillance, and the accumulation of activity from minor leaks that exceed the limits established in the Technical Specificatims would lead to reactor shutdown. Therefore, it is unlikely that the total amount of activity considered available for release in this analysis would ever be realized.
- 4. The coincident loss of offsite power with the occurrence of the reactor trip following the steam generator tube rupture is a conservative assumption. In the event of availability of offsite power, the turbine bypass valves will open, relieving steam to the main condenser. This will reduce the amount of steam and entrained activity discharged directly to the envi-ronment from the unaffected steam generators.
- 5. The meteorological conditions assumed to be present at the site during the course of the accident are based on x/Q values for the exclusion area boundary or LPZ outer boundary.
Furthermore, no credit has been taken for the transit time required for activity to travel from the point of release to the exclusion area boundary or LPZ outer boundary. Hence, the radiological consequences evaluated under these conditions are conservatP;e.
l
. RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTl".', C t ontinued)
- 6. A conservative steam generator decontamination factor (DF) of 10 is used in the cooldown phase (release to atmospheric dump valve).
Results The casults of radiological consequences of a steam generator tube rupture are presented in Table III. The values for thyroid and whole body g',-ma dose show that the doses calculated using the above defined concervative assumptions are well within the limits of 10CFR100.
M e
r
~- - - - - . - , , _ _
Table I Parameters Used in Evaluating the Radiological Consequences of a Steam, Generator Tube Rupture Parameter Assumption Reference t-
. muruc lata:
Power level, Mwt 1,500 FSA9 Steam generator tube leakage, 8,640 Assumption Ib/d (1 gal / min)
Equilibrium reactor coolant activity Coincident (existing) iodine spike uc/g dose equivalent I-131 60 Assumption Equilibrium sec nda.ry system 0.1 uc/g dose Assumption activity - equivalent I-131 Activity Release Data: -
Steam discharge, lb Affected steam generator Reactor coolant leakage to steam generator (0-30 min) 42,300 Mass of steam released 3.551 x 106 Dispersion Data:
Atmospheric dispersion factors, sec/M EAB 4.4 x 10-4 LPZ(0-8 hours) 1,57 x 10-5 Iodine decontamination factors for steam generators (between water and steam phase) 10 Assumption 1
l
, . Table II 4
Activity Released from Steam Generator 9-fluclide Activity (Curies)
DEC I-131 2.7423 (2)
Kr-83m 1.74 .
Kr-85m 9.61 Kr-85 1.67 (3)
Kr-87 4.71 Kr-88 17.17 Xe-131m 13.64 4
Xe-133m 21.08 Xe-133 1.90 (3) i Xe-135m 6.5 (-1)
Xe-135 ,
32.07 Xe-138 . 2.12 I
I l
I i l 1
Table III Radiological Consequences of a Postulated Steam Generator Tube Rupture Results Value Exclusion Area Boundary Dose (duration), Rem Coincident (existing) iodine spike Thyroid 62 Total Body Gamma 2.72 x 10-2 LPZ Outer Boundary Dose (duration)*, rem Coincident (existing) iodine spike Thyroid 2.21 Total Whole Body 9.7 x 10-4
~
- For conservatism, the LPZ dose calculations are based on a dispersion factor for 0-8 hours.
B l
l-Attachment 3 i
RADIOLOGICAL CONSEQUENCE.S OF A MAIN STEAM LINE BREAK .
Physical Model To evaluate the radiological consequences due to a postulated main steam line break (outside containment), it i; assumed that there is a complete severance of a main steam line outside the containment with the plant in a hot zero power condition where the transient is initiated shortly after full power operation. It is also assumed that there is a simul-taneous loss of offsite power. The hot zero power condition assures the maximum water inventory in the steam generators and the shutdown from full power (in conjunction with the loss of off-site power) assures the maximum decay heat which nust be removed by manual control of the atmos-pheric dump valve associatad with the intact steam generator.
The main steam isolation valve.s are installed in the main steam lines from each steam generator, downstream from the safety relief valves and atmospheric dump valves butside containment. The severance of the main steam line is assumed to be upstream of the main steam isolation valve.
A reactor trip is actuated by a low steam generator pressure signal. A main steam isolation signal (MSIS) is actuated to shut the main steam isolation valves from both steam generators. The affected steam gener-ator (steam generator connected to the severed steam lina) blows down completely. The steam is vented directly to the atmosphere. The atmos-pheric dump valve of the unaffected steam generator is used to initiate a 75F/h cooldown of the reactor coolant system 1800 seconds after initi-4 ation of the accident. The steam is vented directly to the atmosphnre.
Mass release from the unaffected steam generator is terminated wher. the shutdown cooling system is initiated at a reactor coolant system temper-
.ature of 3000F.
In this evaluation, a case with an iodine spike caused by the main steam line break accident was evaluated for radiological consequences. The mathematical models, assumptions, and parameters used in this analysic were identical with the design basis main steam line break accident without an iodine spike and described in the above paragraphs with the following exception:
Prior to the main steam line break accident the reactor coolant system activity is based on 1% failed fuel.
However, at the initiation of the MSLB accident, the I-131 equivalent sauce term (released from fuel) is assumed to increase. The iodine release rate is assumed to increase by a factor of 500. l l
I Assumptions and Conditions !
The major assumptions, parameters, and calculational methods used.in the design basis analysis are presented in Table I. Additional clarification is provided as follows:
l
i l'
12-RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE B' LEAK (Continued)
Assumptions and Conditions (Continued)
- 1. Reactor Coolant Activity The reactor coolant equilibrium activity is based on long term operation at 100% of the ultimate ctre power level of 1500 Mwt and 1% failed fuel. Source terms ara listed in Table I.
- 2. Secondary System Activity The activity in steam generators is conservatively assumed to be equal to 0.1 uCi/gm dose equivalent Iodine-131 (I-131).
- 3. Primary-to-Secondary Leakage
. The primary-to-secondary leakage of 1 gal / min (Technical Specification limit) was assumed to continue through the affected steam generator at a constant rate until shutdown cooling is initiated.
Uncertainties and Conservathms
- 1. An 8640 lbm/d (1 gal / min) steam generator primary-to-secondary leakage is assumed, which is greater by a factor of 50 to 200 than that normally observed in past PWR operation.
- 2. The steam generator equilibrium activity for both steam gener-ators is assumed to be equal to the Technical Specification limit (0.1 uCi/g dose equivalent I-131) for the duration of the accident.
- 3. The meteorological conditions assumed to be present at the site during the course of the accident are based on x/Q values for the exclusion area boundary or LPZ outer boundary. Fur-thennore, no credit has been taken for the transit time re-quired for activity to travel from the point of release to the exclusica area boundary or LPZ outer boundary. Hence, the radiological consequences evaluated under these conditions will be conservative.
Results The results of radiological consequences due to a postulated Main Steam Line Break are presented in Table III. The values for thyroid and whole body gama dose show that the doses calculated using the above defined conservative assumptions are well within the limits of 10CFR100.
l Table I l Parameters Used in Evaluating the Radiological Consequences I of a Main Steam Line Break Accident (HSLBA)
Parameter Assumptions Reference Data and assumptions used to estimate radioactive source General Power level, Mwt 1500 FSAR Burnup End of Cycle Assumption Reactor coolant activity after 60 uCi/gm accident Iodine spike caused by DEC I-131 Assumption accident Steam generator activity before 0.1 uCi/gm Assumption accident dose equiv.
I-131 General Loss of offsite power yes Assunption Credit for ridioactive decay no Assumptior in transit t.) dose point Affected steam generator -
Primary-to-secondary leakage rate, 8,640 Assu.ipt'an lb/d (1 gal / min)
Secondary mass release to atmosphere 233,498 (through severed line), lbm Mass of primary-to-secondary leakage, 491 Assumption Ibm Steam generator decont&mination factor 1 Assumptior between steam and water phase c
Unaffected steam generator Primary-to-secondary leakage rate, O Assumption Ib/d !
Dispersion Data:
Atmospheric dispersion factors, sec/m3 EAB 4.4 x 10-4 LPZ (0-8 hours) 1.57 x 10-5 l
l l i
.,.m . .--- -_,m.-. , , .__--..,%_ _
Table II
, Activity Released from the Steam Generator Nuclide Activity (Curies) !
I DEC I-131 23.93 Kr-83m 1.86 (-2) ,
Kr-85m 1.08 (-1)
Kr-85 1.93 Kr-87 4.8 (-2)
Kr-88 2.12 (-1)
Xe-131m 1.61 (-1)
Xe-133m 2.44 (-1)
~
2.20 (1)
Xe-133 Xe-135m 4.83 (-3)
Xe-135 3.63 (-1)
Xe-138 1.54 (-2) 4 I
4 L
I
i i
I i
Table III Radiological Consequences Due to a Postulated j Main Steam Line Break ,
Iodine Spike Caused Result by Accident Exclusion Area Boundary Dose (duration) rem:
Thyroid 5.41 Total Body Ganna 1.19 x 10-3 LPZ Outer Boundary Dose *
(duration), rem: -
Thyroid 1.19 x 10-1 Total 3ody Gamma 4.22 x 10-5 _
- For conservatism, the LPZ dose calculations are based on a dispersion factor for 0-8 hours.
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