Similar Documents at Fermi |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217N4261999-10-25025 October 1999
Forwards Amend 17 to License DPR-9 & Safety Evaluation. Amend Revises TS by Deleting SR D.3.c,which Required Weekly Observation of Nitrogen Cover Gas Pressure within Sodium Storage Tanks Located in Sodium Building
NRC-99-0085, Forwards Rev 15 to Fermi 2 Security Personnel Training & Qualification Plan, IAW 10CFR50.54(p)(2).Without Encl1999-10-22022 October 1999
Forwards Rev 15 to Fermi 2 Security Personnel Training & Qualification Plan, IAW 10CFR50.54(p)(2).Without Encl
NRC-99-0094, Forwards Safeguards Event Rept 99-S01-00 for Reportable Event That Occurred on 990922.Commitments Made by Util Listed1999-10-22022 October 1999
Forwards Safeguards Event Rept 99-S01-00 for Reportable Event That Occurred on 990922.Commitments Made by Util Listed
ML20217P1661999-10-21021 October 1999
Forwards Insp Rept 50-341/99-17 on 990927-1005.No Violations Noted
ML20217M0771999-10-19019 October 1999
Confirms Discussion Between GL Shear,M Mitchell,E Kosky, D Williams,B Rumins,D Harmon & W Rutenberg in Public Meeting on 990923.Purpose of Meeting Was to Discuss Current Status RP Program at Enrico Fermi 2
ML20217N9231999-10-15015 October 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 40
ML20217E5671999-10-14014 October 1999
First Partial Response to FOIA Request for Documents.Records in App a Already Available in Pdr.App B Records Encl & Being Released in Entirety
NRC-99-0087, Responds to NRC Re Violations Noted in Insp Rept 50-341/99-10.Licensee Contests Subject Violation,Due to Listed Reasons1999-10-12012 October 1999
Responds to NRC Re Violations Noted in Insp Rept 50-341/99-10.Licensee Contests Subject Violation,Due to Listed Reasons
NRC-99-0083, Submits Requested Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal1999-10-0707 October 1999
Submits Requested Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal
ML20217D5331999-10-0707 October 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 39
ML20217F5931999-10-0707 October 1999
Forwards Insp Rept 50-341/99-11 on 990724-0908.Two Violations Re Failure to Follow Procedures While Sampling Standby Liquid Control Tank & Wear Dosimetry Properly Being Treated as Noncited Violations
ML20217B8261999-10-0404 October 1999
Forwards Summary of Decommissioning Insp Plan for Fy 2000. Plan Will Be Updated Annually & May Be Revised at Any Time Based Upon Future Insp Findings,Events & Resource Availability
ML20212J8121999-10-0101 October 1999
Discusses Plant,Unit 2 Completion of Licensing Action for GL 98-01 & Suppl 1, Y2K Readiness of Computer Sys at Npps. Ack Efforts That Licensee Have Completed to Date in Preparing Plant for Y2K Transition
ML20212K8881999-09-30030 September 1999
Refers to 990928 Meeting Conducted at Fermi Unit 2 to Discuss Initiatives in Risk Area & to Establish Dialog Between SRAs & PSA Staff
ML20217A3261999-09-30030 September 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 38
NRC-99-0091, Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Examinations1999-09-27027 September 1999
Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Examinations
ML20212J8331999-09-27027 September 1999
Responds to Licensee 990924 Request That NRC Exercise Discretion Not to Enforce Compliance with Actions Required in TS 3/4.6.1.8.Ltr Documents 990923 Telcon with Util in Which NRC Orally Issued NOED
ML20216G9791999-09-24024 September 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 37
NRC-99-0092, Documents Util Request for Enforcement Discretion with Respect to Requirements of TS 3/4.6.1.8 Re Drywell & Suppression Chamber Purge Sys1999-09-24024 September 1999
Documents Util Request for Enforcement Discretion with Respect to Requirements of TS 3/4.6.1.8 Re Drywell & Suppression Chamber Purge Sys
ML20216H3561999-09-23023 September 1999
Informs That During 990914 Telcon,L Sanders & M Bielby Made Arrangements with NRC to Inspect Licensed Operator Requalification Program at Fermi Nuclear Station for Week of 991129 Which Coincides with Util Regularly Scheduled Exam
ML20216H4541999-09-23023 September 1999
Informs That During 990914 Telcon L Sanders & M Bielby Made Arrangements for Administration of Licensing Exams at Fermi Nuclear Station for Week of Jan 31,2000.NRC Will Make Exam Validation Visit to Facility During Week of Jan 10,2000
ML20212F3241999-09-20020 September 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 36
NRC-99-0081, Submits Eight Relief Requests Associated with Containment Inservice Exam Program for Fermi 2 Plant,Per Provisions of 10CFR50.55a(g)(5)(iv).Requested Relief Does Not Adversely Affect Health & Safety of Public1999-09-17017 September 1999
Submits Eight Relief Requests Associated with Containment Inservice Exam Program for Fermi 2 Plant,Per Provisions of 10CFR50.55a(g)(5)(iv).Requested Relief Does Not Adversely Affect Health & Safety of Public
ML20217A9421999-09-0909 September 1999
Informs That NRC Plan to Conduct Addl Resident Insps Beyond Core Insp Program Over Next 6 Months to Assess Improvements in Areas of Procedural Compliance & Quality of Work. Historical Listing of Plant Issues Encl
NRC-99-0079, Forwards Rev 15 to Fermi 2 ITS Submittal,Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev1999-09-0808 September 1999
Forwards Rev 15 to Fermi 2 ITS Submittal,Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev
ML20211Q0991999-09-0303 September 1999
Forwards Rev 35 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0090, Forwards Proprietary Operator License Renewal Application, NRC Forms 398 & 396 for SW Burt,License SOP-30199-3. Proprietary Encls Withheld1999-09-0303 September 1999
Forwards Proprietary Operator License Renewal Application, NRC Forms 398 & 396 for SW Burt,License SOP-30199-3. Proprietary Encls Withheld
NRC-99-0066, Forwards semi-annual fitness-for-duty Rept for Fermi 2 for Period of 990101-0630,IAW 10CFR26.71(d)1999-08-30030 August 1999
Forwards semi-annual fitness-for-duty Rept for Fermi 2 for Period of 990101-0630,IAW 10CFR26.71(d)
ML20211G4951999-08-24024 August 1999
Forwards Rev 34 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0055, Forwards Rev 5 to Inservice Testing Program for Pumps & Valves for Fermi 2 Second 10-Year Interval, Per Provisions of 10CFR50.55a(f)(5)(iv).Rev Includes Four Pump Relief Requests & Ten Valve Relief Requests Requiring NRC Approval1999-08-19019 August 1999
Forwards Rev 5 to Inservice Testing Program for Pumps & Valves for Fermi 2 Second 10-Year Interval, Per Provisions of 10CFR50.55a(f)(5)(iv).Rev Includes Four Pump Relief Requests & Ten Valve Relief Requests Requiring NRC Approval
IR 05000341/19990131999-08-19019 August 1999
Discusses Insp Rept 50-341/99-13 & OI Rept 3-98-039 Conducted Into Potential Misconduct of Operator Formerly Employed by De.Nrc Determined That Violation of NRC Regulations Occurred.Synopsis of IO Rept Encl
NRC-99-0056, Forwards Rev 2,change 0 to ISI-Nondestructive Exam (ISI-NDE) Program (Plan) for Piping,Components & Integral Attachments, for Second ten-year Interval of Fermi 2 Plant Operation1999-08-19019 August 1999
Forwards Rev 2,change 0 to ISI-Nondestructive Exam (ISI-NDE) Program (Plan) for Piping,Components & Integral Attachments, for Second ten-year Interval of Fermi 2 Plant Operation
ML20211C0521999-08-17017 August 1999
Forwards Tables for Sections 3.6 & 3.7 of Draft Safety Evaluation of Proposed Conversion of Current TS to Improved Standard TS
NRC-99-0089, Forwards Rev 13 to Fermi 2 ITS Submittal Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev1999-08-17017 August 1999
Forwards Rev 13 to Fermi 2 ITS Submittal Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev
ML20217E6181999-08-13013 August 1999
FOIA Request for Documents Re Implementing Regulations in 10CFR & Requesting That Region III Task Interface Agreement (TIA) Related to Fermi 2 Plant Be Made Available
IR 05000341/19990011999-08-10010 August 1999
Forwards Insp Rept 50-341/99-01 on 990619-0723 & Nov. Violation of Concern Because Operators Did Not Understand License Requirements & Placed Plant in Configuration Where EDG 11 Was Removed from Service
IR 05000341/19990121999-08-0606 August 1999
Forwards Insp Rept 50-341/99-12 on 990712-0716.No Violations Noted.Insp Focused on Implementation of Plant Chemistry & Radiological Environ Monitoring Programs & Operation of post-accident Sampling Sys
ML20211C4761999-08-0606 August 1999
Forwards Draft Safety Evaluation Re Proposed Conversion of Current TS to Improved Standard TS
NRC-99-0057, Forwards Rev 12 to ITS 980403 Submittal,Providing Update to Submittal for ITS Section 3.3,including Changes Associated with NRC Review of Rev 6 ITS1999-08-0404 August 1999
Forwards Rev 12 to ITS 980403 Submittal,Providing Update to Submittal for ITS Section 3.3,including Changes Associated with NRC Review of Rev 6 ITS
ML20210N5451999-08-0303 August 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 33
NRC-99-0077, Forwards Rev 11 to Fermi ITS Submittal,Dtd 980403,providing Update to Submittal for ITS Sections 3.5,3.6 & 3.8.Update Includes Changes Associated with NRC Review of Util Transmittals of Revs 5,5a & 7 to ITS1999-07-26026 July 1999
Forwards Rev 11 to Fermi ITS Submittal,Dtd 980403,providing Update to Submittal for ITS Sections 3.5,3.6 & 3.8.Update Includes Changes Associated with NRC Review of Util Transmittals of Revs 5,5a & 7 to ITS
ML20210J2121999-07-23023 July 1999
Discusses NRC OI Investigation Rept 3-98-039 Completed on 990524 & Forwards Notice of Violation
NRC-99-0053, Forwards Rev 14 to Fermi 2 Security Personnel Training & Qualification Plan, Per 10CFR50.54(p)(2).Changes Do Not Decrease Effectiveness of Plan.Summary of Changes Also Encl1999-07-15015 July 1999
Forwards Rev 14 to Fermi 2 Security Personnel Training & Qualification Plan, Per 10CFR50.54(p)(2).Changes Do Not Decrease Effectiveness of Plan.Summary of Changes Also Encl
ML20209H7211999-07-14014 July 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 32
NRC-99-0054, Forwards Rev 10 to Fermi 2 ITS Submittal, , Providing Update to Submittal for ITS Section 3.4,RCS.Update Includes Changes Associated with NRC Review of Util Transmittal of Rev 4 to ITS Submittal1999-07-13013 July 1999
Forwards Rev 10 to Fermi 2 ITS Submittal, , Providing Update to Submittal for ITS Section 3.4,RCS.Update Includes Changes Associated with NRC Review of Util Transmittal of Rev 4 to ITS Submittal
ML20209F6571999-07-12012 July 1999
Forwards Insp Rept 50-341/99-09 on 990516-0618.No Violations Noted.Weak Procedural Guidance,Communications Issues & Configuration Control Weaknessess Were Evident
NRC-99-0052, Forwards Rev 9 to Fermi 2 ITS Submittal,Dtd 980403. Attachment 1 Contains Brief Abstract of Changes Included in Rev.Attachment 2 Contains Revised Pages for Submittal Along with Rev Instructions1999-07-0707 July 1999
Forwards Rev 9 to Fermi 2 ITS Submittal,Dtd 980403. Attachment 1 Contains Brief Abstract of Changes Included in Rev.Attachment 2 Contains Revised Pages for Submittal Along with Rev Instructions
ML20196J5511999-07-0202 July 1999
Discusses GL 92-01,Rev 1,Supp 1, Rv Integrity, Issued by NRC on 950519 & DE Responses & 980729 for Fermi 2.Informs That Staff Revised Info in Rvid & Released Info as Rvid Version 2 as Result of Review of Responses
ML20209C6141999-07-0101 July 1999
Rev 31 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0076, Forwards Rev 8 to ITS Submittal to Provide Update to ITS Submittal for ITS Section 3.7.Brief Abstract of Changes Included1999-06-30030 June 1999
Forwards Rev 8 to ITS Submittal to Provide Update to ITS Submittal for ITS Section 3.7.Brief Abstract of Changes Included
1999-09-09
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Category:INCOMING CORRESPONDENCE
MONTHYEARNRC-99-0094, Forwards Safeguards Event Rept 99-S01-00 for Reportable Event That Occurred on 990922.Commitments Made by Util Listed1999-10-22022 October 1999
Forwards Safeguards Event Rept 99-S01-00 for Reportable Event That Occurred on 990922.Commitments Made by Util Listed
NRC-99-0085, Forwards Rev 15 to Fermi 2 Security Personnel Training & Qualification Plan, IAW 10CFR50.54(p)(2).Without Encl1999-10-22022 October 1999
Forwards Rev 15 to Fermi 2 Security Personnel Training & Qualification Plan, IAW 10CFR50.54(p)(2).Without Encl
ML20217N9231999-10-15015 October 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 40
NRC-99-0087, Responds to NRC Re Violations Noted in Insp Rept 50-341/99-10.Licensee Contests Subject Violation,Due to Listed Reasons1999-10-12012 October 1999
Responds to NRC Re Violations Noted in Insp Rept 50-341/99-10.Licensee Contests Subject Violation,Due to Listed Reasons
NRC-99-0083, Submits Requested Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal1999-10-0707 October 1999
Submits Requested Response to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal
ML20217D5331999-10-0707 October 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 39
ML20217A3261999-09-30030 September 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 38
NRC-99-0091, Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Examinations1999-09-27027 September 1999
Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Examinations
ML20216G9791999-09-24024 September 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 37
NRC-99-0092, Documents Util Request for Enforcement Discretion with Respect to Requirements of TS 3/4.6.1.8 Re Drywell & Suppression Chamber Purge Sys1999-09-24024 September 1999
Documents Util Request for Enforcement Discretion with Respect to Requirements of TS 3/4.6.1.8 Re Drywell & Suppression Chamber Purge Sys
ML20212F3241999-09-20020 September 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 36
NRC-99-0081, Submits Eight Relief Requests Associated with Containment Inservice Exam Program for Fermi 2 Plant,Per Provisions of 10CFR50.55a(g)(5)(iv).Requested Relief Does Not Adversely Affect Health & Safety of Public1999-09-17017 September 1999
Submits Eight Relief Requests Associated with Containment Inservice Exam Program for Fermi 2 Plant,Per Provisions of 10CFR50.55a(g)(5)(iv).Requested Relief Does Not Adversely Affect Health & Safety of Public
NRC-99-0079, Forwards Rev 15 to Fermi 2 ITS Submittal,Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev1999-09-0808 September 1999
Forwards Rev 15 to Fermi 2 ITS Submittal,Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev
NRC-99-0090, Forwards Proprietary Operator License Renewal Application, NRC Forms 398 & 396 for SW Burt,License SOP-30199-3. Proprietary Encls Withheld1999-09-0303 September 1999
Forwards Proprietary Operator License Renewal Application, NRC Forms 398 & 396 for SW Burt,License SOP-30199-3. Proprietary Encls Withheld
ML20211Q0991999-09-0303 September 1999
Forwards Rev 35 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0066, Forwards semi-annual fitness-for-duty Rept for Fermi 2 for Period of 990101-0630,IAW 10CFR26.71(d)1999-08-30030 August 1999
Forwards semi-annual fitness-for-duty Rept for Fermi 2 for Period of 990101-0630,IAW 10CFR26.71(d)
ML20211G4951999-08-24024 August 1999
Forwards Rev 34 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0056, Forwards Rev 2,change 0 to ISI-Nondestructive Exam (ISI-NDE) Program (Plan) for Piping,Components & Integral Attachments, for Second ten-year Interval of Fermi 2 Plant Operation1999-08-19019 August 1999
Forwards Rev 2,change 0 to ISI-Nondestructive Exam (ISI-NDE) Program (Plan) for Piping,Components & Integral Attachments, for Second ten-year Interval of Fermi 2 Plant Operation
NRC-99-0055, Forwards Rev 5 to Inservice Testing Program for Pumps & Valves for Fermi 2 Second 10-Year Interval, Per Provisions of 10CFR50.55a(f)(5)(iv).Rev Includes Four Pump Relief Requests & Ten Valve Relief Requests Requiring NRC Approval1999-08-19019 August 1999
Forwards Rev 5 to Inservice Testing Program for Pumps & Valves for Fermi 2 Second 10-Year Interval, Per Provisions of 10CFR50.55a(f)(5)(iv).Rev Includes Four Pump Relief Requests & Ten Valve Relief Requests Requiring NRC Approval
NRC-99-0089, Forwards Rev 13 to Fermi 2 ITS Submittal Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev1999-08-17017 August 1999
Forwards Rev 13 to Fermi 2 ITS Submittal Dtd 980403.Rev Provide Update for Remaining Open Issues Associated with NRC Review of ITS Submittal.Attachment 1 Contains Brief Abstract of Changes Included in Rev
ML20217E6181999-08-13013 August 1999
FOIA Request for Documents Re Implementing Regulations in 10CFR & Requesting That Region III Task Interface Agreement (TIA) Related to Fermi 2 Plant Be Made Available
NRC-99-0057, Forwards Rev 12 to ITS 980403 Submittal,Providing Update to Submittal for ITS Section 3.3,including Changes Associated with NRC Review of Rev 6 ITS1999-08-0404 August 1999
Forwards Rev 12 to ITS 980403 Submittal,Providing Update to Submittal for ITS Section 3.3,including Changes Associated with NRC Review of Rev 6 ITS
ML20210N5451999-08-0303 August 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 33
NRC-99-0077, Forwards Rev 11 to Fermi ITS Submittal,Dtd 980403,providing Update to Submittal for ITS Sections 3.5,3.6 & 3.8.Update Includes Changes Associated with NRC Review of Util Transmittals of Revs 5,5a & 7 to ITS1999-07-26026 July 1999
Forwards Rev 11 to Fermi ITS Submittal,Dtd 980403,providing Update to Submittal for ITS Sections 3.5,3.6 & 3.8.Update Includes Changes Associated with NRC Review of Util Transmittals of Revs 5,5a & 7 to ITS
NRC-99-0053, Forwards Rev 14 to Fermi 2 Security Personnel Training & Qualification Plan, Per 10CFR50.54(p)(2).Changes Do Not Decrease Effectiveness of Plan.Summary of Changes Also Encl1999-07-15015 July 1999
Forwards Rev 14 to Fermi 2 Security Personnel Training & Qualification Plan, Per 10CFR50.54(p)(2).Changes Do Not Decrease Effectiveness of Plan.Summary of Changes Also Encl
ML20209H7211999-07-14014 July 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 32
NRC-99-0054, Forwards Rev 10 to Fermi 2 ITS Submittal, , Providing Update to Submittal for ITS Section 3.4,RCS.Update Includes Changes Associated with NRC Review of Util Transmittal of Rev 4 to ITS Submittal1999-07-13013 July 1999
Forwards Rev 10 to Fermi 2 ITS Submittal, , Providing Update to Submittal for ITS Section 3.4,RCS.Update Includes Changes Associated with NRC Review of Util Transmittal of Rev 4 to ITS Submittal
NRC-99-0052, Forwards Rev 9 to Fermi 2 ITS Submittal,Dtd 980403. Attachment 1 Contains Brief Abstract of Changes Included in Rev.Attachment 2 Contains Revised Pages for Submittal Along with Rev Instructions1999-07-0707 July 1999
Forwards Rev 9 to Fermi 2 ITS Submittal,Dtd 980403. Attachment 1 Contains Brief Abstract of Changes Included in Rev.Attachment 2 Contains Revised Pages for Submittal Along with Rev Instructions
ML20209C6141999-07-0101 July 1999
Rev 31 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0049, Submits Response to NRC GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps. Disclosure Encl1999-06-30030 June 1999
Submits Response to NRC GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps. Disclosure Encl
NRC-99-0076, Forwards Rev 8 to ITS Submittal to Provide Update to ITS Submittal for ITS Section 3.7.Brief Abstract of Changes Included1999-06-30030 June 1999
Forwards Rev 8 to ITS Submittal to Provide Update to ITS Submittal for ITS Section 3.7.Brief Abstract of Changes Included
ML20209C1081999-06-29029 June 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section), Rev 30
NRC-99-0050, Forwards Rev 7 to ITS Submittal of 980403,to Provide Remaining Responses to NRC RAIs Re ITS Sections 3.5 & 3.8. Revision Instructions Included1999-06-24024 June 1999
Forwards Rev 7 to ITS Submittal of 980403,to Provide Remaining Responses to NRC RAIs Re ITS Sections 3.5 & 3.8. Revision Instructions Included
NRC-99-0075, Forwards Operator License Renewal Application,Nrc Forms 398 & 396 for CR Effan,License SOP-30186-3.Encl Withheld1999-06-22022 June 1999
Forwards Operator License Renewal Application,Nrc Forms 398 & 396 for CR Effan,License SOP-30186-3.Encl Withheld
ML20212J3861999-06-18018 June 1999
Forwards Rev 29 to Approved UFSAR LCR That Have Not Been Incorporated Into UFSAR Hardcopy (Sorted by Section)
NRC-99-0047, Responds to NRC 990513 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions. Revised Design Calculation DC-5957, GL 96-06 Calculations, Encl1999-06-0909 June 1999
Responds to NRC 990513 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions. Revised Design Calculation DC-5957, GL 96-06 Calculations, Encl
ML20195G2001999-06-0808 June 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy & Approved TRM Lcrs (Sorted by Section), Rev 28
ML20195E7391999-06-0303 June 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy & Approved TRM Lcrs (Sorted by Section), Rev 27
NRC-99-0046, Forwards Rev 6 to Fermi 2 ITS Submittal ,to Provide Remaining Responses to NRC RAI Re ITS Section 3.3, & Update ITS Submittal to Reflect Responses.Brief Abstract of Changes Contained in Attachment 11999-06-0202 June 1999
Forwards Rev 6 to Fermi 2 ITS Submittal ,to Provide Remaining Responses to NRC RAI Re ITS Section 3.3, & Update ITS Submittal to Reflect Responses.Brief Abstract of Changes Contained in Attachment 1
ML20207D9631999-05-26026 May 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy & Approved TRM Lcrs (Sorted by Section)
ML20195B3381999-05-19019 May 1999
Forwards Rev 25 to Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy & Approved TRM Lcrs (Sorted by Section)
NRC-99-0040, Requests NRC Approval of Revs to Weld Tables of Encl Relief Requests RR-A21,rev 1 & RR-A22,rev 1.Revs Do Not Impact Basis for Original Requests for Relief Because There Is No Change to Technical Basis or Proposed Alternative1999-05-14014 May 1999
Requests NRC Approval of Revs to Weld Tables of Encl Relief Requests RR-A21,rev 1 & RR-A22,rev 1.Revs Do Not Impact Basis for Original Requests for Relief Because There Is No Change to Technical Basis or Proposed Alternative
NRC-99-0039, Requests Relief from ASME Section XI Requirement to Examine Essentially 100% of Selected Category B-J Pipe Welds,Per 10CFR50.55a(g)(5)(iii).Relief Request RR-A23,encl1999-05-14014 May 1999
Requests Relief from ASME Section XI Requirement to Examine Essentially 100% of Selected Category B-J Pipe Welds,Per 10CFR50.55a(g)(5)(iii).Relief Request RR-A23,encl
NRC-99-0041, Requests Relief from ASME Section XI Requirements to Examine Essentially 100% of Accessible Category B-D Nozzle Welds,Per 10CFR50.55a(g)(5)(iii).Rev 1 to Relief Request RR-A6,encl1999-05-14014 May 1999
Requests Relief from ASME Section XI Requirements to Examine Essentially 100% of Accessible Category B-D Nozzle Welds,Per 10CFR50.55a(g)(5)(iii).Rev 1 to Relief Request RR-A6,encl
NRC-99-0038, Requests Relief from ASME Section XI 1980 Requirement to Examine Essentially 100% If Category B-A RPV Shell Weld Lengths,Per 10CFR50.55a(g)(5)(iii)1999-05-14014 May 1999
Requests Relief from ASME Section XI 1980 Requirement to Examine Essentially 100% If Category B-A RPV Shell Weld Lengths,Per 10CFR50.55a(g)(5)(iii)
NRC-99-0045, Forwards Rev 5a to Fermi 980403 Submittal,Converting TS to Its.Encl Rev Provides Remaining Responses to NRC RAIs Re ITS Section 3.6 Contained in NRC 990211 Rai.Brief Abstract of Changes & Responses to Rai,Encl1999-05-0707 May 1999
Forwards Rev 5a to Fermi 980403 Submittal,Converting TS to Its.Encl Rev Provides Remaining Responses to NRC RAIs Re ITS Section 3.6 Contained in NRC 990211 Rai.Brief Abstract of Changes & Responses to Rai,Encl
NRC-99-0044, Forwards Rev 5 to ITS to Provide Responses to RAI Re ITS Section 3.6 (Except ITS Section 3.6.1.3) Contained in & Update in ITS Submittal to Reflect Responses1999-04-30030 April 1999
Forwards Rev 5 to ITS to Provide Responses to RAI Re ITS Section 3.6 (Except ITS Section 3.6.1.3) Contained in & Update in ITS Submittal to Reflect Responses
NRC-99-0059, Provides Notification Required by 10CFR50.82(a)(7) of Changes to Fermi 1 Decommissioning Cost Estimate1999-04-29029 April 1999
Provides Notification Required by 10CFR50.82(a)(7) of Changes to Fermi 1 Decommissioning Cost Estimate
ML20206E2941999-04-28028 April 1999
Forwards Approved UFSAR Lcrs That Have Not Been Incorporated Into UFSAR Hardcopy & Approved TRM Lcrs (Sorted by Section), Rev 24
NRC-99-0037, Forwards Rev 9 to Fermi 2 Ufsar,Per 10CFR50.71(e) & 10CFR50.4(b)(6).Summary of Changes to QA Program & Vol 1 of Technical Requirements Manual,Encl1999-04-26026 April 1999
Forwards Rev 9 to Fermi 2 Ufsar,Per 10CFR50.71(e) & 10CFR50.4(b)(6).Summary of Changes to QA Program & Vol 1 of Technical Requirements Manual,Encl
1999-09-08
[Table view]
Category:UTILITY TO NRC
MONTHYEARNRC-90-0142, Forwards Revised Semiannual Fitness for Duty Rept,Consisting of Statement Re Percentage of Individuals Tested & Revised Graph,Per G Pirtle Request1990-09-0707 September 1990
Forwards Revised Semiannual Fitness for Duty Rept,Consisting of Statement Re Percentage of Individuals Tested & Revised Graph,Per G Pirtle Request
NRC-90-0152, Forwards Rev 3 to Radiological Emergency Response Preparedness Plan.W/O Encl1990-09-0505 September 1990
Forwards Rev 3 to Radiological Emergency Response Preparedness Plan.W/O Encl
NRC-90-0139, Responds to NRC Re Memorization of Immediate Operator Actions in Abnormal Operating Procedures (Aop). Definition of Immediate Action Under Review to Provide Mechanism for Consistent Application to All AOPs1990-08-31031 August 1990
Responds to NRC Re Memorization of Immediate Operator Actions in Abnormal Operating Procedures (Aop). Definition of Immediate Action Under Review to Provide Mechanism for Consistent Application to All AOPs
NRC-90-0116, Forwards Semiannual Fitness for Duty Rept1990-08-29029 August 1990
Forwards Semiannual Fitness for Duty Rept
NRC-90-0117, Forwards, Fermi 2 Nuclear Power Plant Semiannual Radioactive Effluent Release Rept for Jan-June 1990. No Releases Occurred1990-08-28028 August 1990
Forwards, Fermi 2 Nuclear Power Plant Semiannual Radioactive Effluent Release Rept for Jan-June 1990. No Releases Occurred
NRC-90-0118, Forwards Endorsement 35 to Maelu Policy MF-120 & Endorsement 139 to Nelia Policy NF-921990-08-24024 August 1990
Forwards Endorsement 35 to Maelu Policy MF-120 & Endorsement 139 to Nelia Policy NF-92
NRC-90-0135, Affirms That Util Intends to Conduct Insp of Low Pressure Turbine Rotor Disks During Upcoming Refueling Outage.Insp of Low Pressure Turbine 2 Carried Out During First Refueling Outage.No Evidence of Stress Corrosion Found1990-08-17017 August 1990
Affirms That Util Intends to Conduct Insp of Low Pressure Turbine Rotor Disks During Upcoming Refueling Outage.Insp of Low Pressure Turbine 2 Carried Out During First Refueling Outage.No Evidence of Stress Corrosion Found
NRC-90-0137, Forwards NPDES Permit MI0037028 for Plant,Approved by State of Mi Water Resources Commission During 900719 Meeting1990-08-17017 August 1990
Forwards NPDES Permit MI0037028 for Plant,Approved by State of Mi Water Resources Commission During 900719 Meeting
NRC-90-0136, Requests That NRC Administer Operator Licensing Exam in Dec 1990 & Postpone Operator Requalification Exam1990-08-17017 August 1990
Requests That NRC Administer Operator Licensing Exam in Dec 1990 & Postpone Operator Requalification Exam
NRC-90-0133, Forwards Rev 5 to Security Personnel Training & Qualification Plan.Rev Establishes App D & Records Retention Matrix to Provide Central Location in Plan for Info on Records Retention Pages 3-4,3-8 & 3-9.Rev Withheld1990-08-0909 August 1990
Forwards Rev 5 to Security Personnel Training & Qualification Plan.Rev Establishes App D & Records Retention Matrix to Provide Central Location in Plan for Info on Records Retention Pages 3-4,3-8 & 3-9.Rev Withheld
NRC-90-0112, Responds to NRC Re Violations Noted in Insp Rept 50-341/90-07.Corrective Actions:Review of RF01 Turbine Work Packages to Identify Where Addl Controls Should Be Placed in Similar Future RF02 Packages Conducted1990-08-0303 August 1990
Responds to NRC Re Violations Noted in Insp Rept 50-341/90-07.Corrective Actions:Review of RF01 Turbine Work Packages to Identify Where Addl Controls Should Be Placed in Similar Future RF02 Packages Conducted
NRC-90-0111, Responds to Concerns Re Open Item 90-007-04,per NRC 900627 Request.Potential Enhancements to Drawings Being Reviewed & to Be Resolved by Dec 19901990-08-0303 August 1990
Responds to Concerns Re Open Item 90-007-04,per NRC 900627 Request.Potential Enhancements to Drawings Being Reviewed & to Be Resolved by Dec 1990
ML20055H8921990-07-26026 July 1990
Forwards Certification of Financial Assurance for Decommissioning of Facility,Per 10CFR50.33(k) & 50.75. Util Intends to Retain Unit 1 in Safe Storage Status Until Unit 2 Shut Down & Decommissioned
NRC-90-0131, Forwards Rev 15 to Physical Security Plan.Rev Withheld1990-07-24024 July 1990
Forwards Rev 15 to Physical Security Plan.Rev Withheld
NRC-90-0108, Discusses Deletion of Commitment for Temporary Mod Program, Per Insp Rept 50-341/84-25.Requirement for Operations Engineer to Approve Extensions of Temporary Mods Beyond 30 Days Being Deleted from Procedure FIP-OP1-021990-07-18018 July 1990
Discusses Deletion of Commitment for Temporary Mod Program, Per Insp Rept 50-341/84-25.Requirement for Operations Engineer to Approve Extensions of Temporary Mods Beyond 30 Days Being Deleted from Procedure FIP-OP1-02
NRC-90-0093, Submits Clarification of Design for Plant Suppression Pool Temp Monitoring Sys.Design Adequate to Ensure Suppression Pool Water Temp within Allowable Tech Spec Limits1990-07-18018 July 1990
Submits Clarification of Design for Plant Suppression Pool Temp Monitoring Sys.Design Adequate to Ensure Suppression Pool Water Temp within Allowable Tech Spec Limits
NRC-90-0091, Requests Relief from Commitment to Perform 100% Audits of Tech Spec Line Items.Tech Spec Improvement Program Has Been Completed Which Provides Extensive Review of Tech Spec Line Items1990-07-0202 July 1990
Requests Relief from Commitment to Perform 100% Audits of Tech Spec Line Items.Tech Spec Improvement Program Has Been Completed Which Provides Extensive Review of Tech Spec Line Items
NRC-90-0101, Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Table Listing Status for Each Issue Encl1990-06-29029 June 1990
Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions. Table Listing Status for Each Issue Encl
NRC-90-0100, Discusses Installation of Motor Operated Valve Limiter Plates,Per IE Bulletin 85-003.Commitment to Install Limiter Plates Deferred Until Time When Valve or Valve Operator Would Be Disassembled for Another Reason1990-06-27027 June 1990
Discusses Installation of Motor Operated Valve Limiter Plates,Per IE Bulletin 85-003.Commitment to Install Limiter Plates Deferred Until Time When Valve or Valve Operator Would Be Disassembled for Another Reason
NRC-90-0088, Discusses Changes Made to Commitment Addressing Calibr Frequency for Inservice Test Instrumentation,Per Violation 87009-03.Nine-month Calibr Frequency for Subj Instrumentation No Longer Appropriate or in Use1990-06-14014 June 1990
Discusses Changes Made to Commitment Addressing Calibr Frequency for Inservice Test Instrumentation,Per Violation 87009-03.Nine-month Calibr Frequency for Subj Instrumentation No Longer Appropriate or in Use
NRC-90-0095, Provides Schedule for Implementation of Mods Re Compliance W/Reg Guide 1.97,Rev 2,per 900502 Request1990-06-0606 June 1990
Provides Schedule for Implementation of Mods Re Compliance W/Reg Guide 1.97,Rev 2,per 900502 Request
NRC-90-0089, Submits Supplemental Response to NRC Concerns on RHR Operability from Insp Rept 50-341/89-17.Util Faced W/ Decision to Provide Addl Overheating Protection to RHR Sys Pump Vs Maintaining Necessary Water Coverage of Fuel1990-06-0404 June 1990
Submits Supplemental Response to NRC Concerns on RHR Operability from Insp Rept 50-341/89-17.Util Faced W/ Decision to Provide Addl Overheating Protection to RHR Sys Pump Vs Maintaining Necessary Water Coverage of Fuel
NRC-90-0085, Provides Plans for Eliminating Temporaary Mods at Plant,Per Insp Rept 50-341/90-051990-05-29029 May 1990
Provides Plans for Eliminating Temporaary Mods at Plant,Per Insp Rept 50-341/90-05
NRC-90-0083, Discusses Action Plan to Review Socket Weld Failures at Plant.Failure History of Socket Welds Will Be Investigated to Determine Whether or Not Significant Common Mode Failure Mechanism Evident.Plan Expected to Be Completed by 9005251990-05-18018 May 1990
Discusses Action Plan to Review Socket Weld Failures at Plant.Failure History of Socket Welds Will Be Investigated to Determine Whether or Not Significant Common Mode Failure Mechanism Evident.Plan Expected to Be Completed by 900525
NRC-90-0074, Discusses Insp of Flash Tanks & Reactor Feed Pump Strainers. No Further Evidence of Moisture Separator Reheater Debris Discovered in Strainers,Stub Tubes,Wire Mesh or Structural Members1990-05-18018 May 1990
Discusses Insp of Flash Tanks & Reactor Feed Pump Strainers. No Further Evidence of Moisture Separator Reheater Debris Discovered in Strainers,Stub Tubes,Wire Mesh or Structural Members
NRC-90-0082, Documents Commitments Made During Recent Meetings & Telcon on RHR Sys Small Bore Connections1990-05-11011 May 1990
Documents Commitments Made During Recent Meetings & Telcon on RHR Sys Small Bore Connections
NRC-90-0068, Forwards 1989 Annual Financial Repts for Detroit Edison Co & Wolverine Power Supply Cooperative,Inc1990-05-0101 May 1990
Forwards 1989 Annual Financial Repts for Detroit Edison Co & Wolverine Power Supply Cooperative,Inc
NRC-90-0071, Responds to NRC 900316 Ltr Re Violations Noted in Insp Rept 50-341/90-03 Re Emergency Preparedness Exercise Weaknesses. Corrective Actions:Case Studies Will Be Developed & Presented in Licensed Operator Requalification Training1990-04-30030 April 1990
Responds to NRC 900316 Ltr Re Violations Noted in Insp Rept 50-341/90-03 Re Emergency Preparedness Exercise Weaknesses. Corrective Actions:Case Studies Will Be Developed & Presented in Licensed Operator Requalification Training
NRC-90-0078, Responds to NRC Bulletin 90-002, Loss of Thermal Margin Caused by Channel Bow. Util Has Decided That Fuel Channel Boxes Will Not Be Used After First Bundle Lifetime1990-04-25025 April 1990
Responds to NRC Bulletin 90-002, Loss of Thermal Margin Caused by Channel Bow. Util Has Decided That Fuel Channel Boxes Will Not Be Used After First Bundle Lifetime
NRC-90-0051, Responds to SALP 11 Board Rept 50-341/90-01.Outage Critique Action Plan Issued for First Refueling Outage Identifying Lessons Learned & Areas Needing Improvement1990-04-12012 April 1990
Responds to SALP 11 Board Rept 50-341/90-01.Outage Critique Action Plan Issued for First Refueling Outage Identifying Lessons Learned & Areas Needing Improvement
NRC-90-0060, Responds to 900104 Request for Supplemental Station Blackout Submittal1990-03-29029 March 1990
Responds to 900104 Request for Supplemental Station Blackout Submittal
NRC-90-0061, Forwards Addl Info Re Proposed Drywell Level Instrument Line Configuration,As Followup to 890927 Tech Spec Change Request.Rept Provides Info Re Reasons for Designating Primary Containment Isolation Valve Normally Opened1990-03-27027 March 1990
Forwards Addl Info Re Proposed Drywell Level Instrument Line Configuration,As Followup to 890927 Tech Spec Change Request.Rept Provides Info Re Reasons for Designating Primary Containment Isolation Valve Normally Opened
NRC-90-0055, Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47 - Safety Implications of Control Sys in Lwrs. Overfill Protection Sys Adequately Separated from Main Feedwater Control Sys1990-03-19019 March 1990
Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47 - Safety Implications of Control Sys in Lwrs. Overfill Protection Sys Adequately Separated from Main Feedwater Control Sys
NRC-90-0024, Suppls Response to Violations Noted in Insp Rept 50-341/85-47 Re Security Records.Util Intends to Rely Solely on Historical Tapes to Retrieve Records Required by Physical Security Plan.Hard Copy Computer Logs Will Be Discarded1990-03-16016 March 1990
Suppls Response to Violations Noted in Insp Rept 50-341/85-47 Re Security Records.Util Intends to Rely Solely on Historical Tapes to Retrieve Records Required by Physical Security Plan.Hard Copy Computer Logs Will Be Discarded
NRC-90-0045, Forwards Inservice Insp Summary Rept of 1989 Inservice Insp at Fermi 2 Nuclear Power Plant.1990-03-13013 March 1990
Forwards Inservice Insp Summary Rept of 1989 Inservice Insp at Fermi 2 Nuclear Power Plant.
NRC-90-0043, Responds to NRC 900224 Ltr Re Weaknesses & Violations Noted in Insp Rept 50-341/89-24.Corrective Actions:Nuclear Fire Protection Specialist Provided Training to Appropriate Personnel on 9002021990-03-0909 March 1990
Responds to NRC 900224 Ltr Re Weaknesses & Violations Noted in Insp Rept 50-341/89-24.Corrective Actions:Nuclear Fire Protection Specialist Provided Training to Appropriate Personnel on 900202
NRC-90-0054, Responds to NRC 900212 Order Modifying License & Notice of Violation.Corrective Actions:Former Director of Nuclear Security No Longer Works for Util & Extensive Actions Taken to Improve Communications W/Nrc1990-03-0909 March 1990
Responds to NRC 900212 Order Modifying License & Notice of Violation.Corrective Actions:Former Director of Nuclear Security No Longer Works for Util & Extensive Actions Taken to Improve Communications W/Nrc
NRC-90-0037, Forwards Completed Questionnaires for Types of Insps,Audits & Evaluations,In Response to Generic Ltr 90-01, Voluntary Participation in NRC Regulatory Impact Survey.1990-03-0505 March 1990
Forwards Completed Questionnaires for Types of Insps,Audits & Evaluations,In Response to Generic Ltr 90-01, Voluntary Participation in NRC Regulatory Impact Survey.
NRC-90-0029, Responds to Violations Noted in Insp Rept 50-341/89-36. Corrective Action:All Personnel in Operations & I&C Viewed Professionalism Video Which Included Presentation of Facts Concerning Event1990-03-0505 March 1990
Responds to Violations Noted in Insp Rept 50-341/89-36. Corrective Action:All Personnel in Operations & I&C Viewed Professionalism Video Which Included Presentation of Facts Concerning Event
NRC-90-0028, Forwards Semiannual Radioactive Effluent Release Rept for Jul-Dec 1989 & Revised Pages of Rept for Jul-Dec 1988, Containing Recalculated Noble Gas Releases & Doses1990-03-0101 March 1990
Forwards Semiannual Radioactive Effluent Release Rept for Jul-Dec 1989 & Revised Pages of Rept for Jul-Dec 1988, Containing Recalculated Noble Gas Releases & Doses
NRC-90-0034, Forwards, 1989 Annual Operating Rept for Fermi 2 & Main Steam Piping Dynamic Response Test Results1990-02-26026 February 1990
Forwards, 1989 Annual Operating Rept for Fermi 2 & Main Steam Piping Dynamic Response Test Results
NRC-90-0036, Forwards Response to NRC Survey on Nuclear Fuel Cycle Industry Future Activity,Per 900122 Request1990-02-26026 February 1990
Forwards Response to NRC Survey on Nuclear Fuel Cycle Industry Future Activity,Per 900122 Request
NRC-90-0026, Forwards Monthly Operating Rept for Jan 1990 for Fermi Unit 2 & Revised Rept for Dec 1989,correcting Error in Net Electrical Generation Figure1990-02-15015 February 1990
Forwards Monthly Operating Rept for Jan 1990 for Fermi Unit 2 & Revised Rept for Dec 1989,correcting Error in Net Electrical Generation Figure
NRC-90-0020, Forwards Response to Violations Noted in Insp Rept 50-341/89-31 Re Maint Support for Security Equipment.Addl Instrumentation & Control Repairmen Hired.Encl Withheld (Ref 10CFR73.21)1990-02-12012 February 1990
Forwards Response to Violations Noted in Insp Rept 50-341/89-31 Re Maint Support for Security Equipment.Addl Instrumentation & Control Repairmen Hired.Encl Withheld (Ref 10CFR73.21)
NRC-90-0010, Responds to NRC 900102 Ltr Re Violations Noted in Insp Rept 50-341/89-30.Corrective Actions:Storage Area in Machine Shop closed-out & QA1 Matl Returned to Warehouse Where Positive Controls in Effect & Shift Turnover Procedure Revised1990-02-0101 February 1990
Responds to NRC 900102 Ltr Re Violations Noted in Insp Rept 50-341/89-30.Corrective Actions:Storage Area in Machine Shop closed-out & QA1 Matl Returned to Warehouse Where Positive Controls in Effect & Shift Turnover Procedure Revised
NRC-90-0013, Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Design. Subj Valves Not Used by Util1990-01-26026 January 1990
Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Design. Subj Valves Not Used by Util
NRC-90-0012, Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Monitoring & Treatment Programs to Reduce Incidents of Flow Blockage Due to Biofouling Will Be Implemented During Early 19911990-01-26026 January 1990
Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Monitoring & Treatment Programs to Reduce Incidents of Flow Blockage Due to Biofouling Will Be Implemented During Early 1991
NRC-90-0016, Responds to 10CFR26 Re fitness-for-duty Programs.Util Conducting Testing for Drugs Described in 10CFR26,App a, Section 2.1(a) & at Cutoff Levels Defined in Sections 2.7(3)(1) & 2.7(f)(2) as of 9001031990-01-26026 January 1990
Responds to 10CFR26 Re fitness-for-duty Programs.Util Conducting Testing for Drugs Described in 10CFR26,App a, Section 2.1(a) & at Cutoff Levels Defined in Sections 2.7(3)(1) & 2.7(f)(2) as of 900103
NRC-89-0297, Forwards Supplemental Response to NRC Request for Addl Info on Generic Ltr 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping. All Welds Accessible for Ultrasonic Insp1990-01-12012 January 1990
Forwards Supplemental Response to NRC Request for Addl Info on Generic Ltr 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping. All Welds Accessible for Ultrasonic Insp
NRC-89-0240, Advises That Info in NRC 891010 Ser,Section 4.13 Re Alternate Rod Insertion/Recirculation Pump Trip Initiation Installed on Divisional Basis,Per ATWS Rule,Deemed Correct. Addl Alarm Also Provided for Arm Position of Switches1990-01-11011 January 1990
Advises That Info in NRC 891010 Ser,Section 4.13 Re Alternate Rod Insertion/Recirculation Pump Trip Initiation Installed on Divisional Basis,Per ATWS Rule,Deemed Correct. Addl Alarm Also Provided for Arm Position of Switches
1990-09-07
[Table view] |
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June 8, 1981 p(hp 19 EF2 - 53,492 S" A Mr. L. L. Kintner Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear 'Ir. Kintner:
Reference:
Enrico Ferr.'i Atomic Power Plant, Unit 2 NRC Docket No. 50-341
Subject:
Responses to Open NRC Questions Please find enclosed several items related to the Fermi 2 docket responding to NRC questions. This information will be included in a forthcoming FSAR amendment as appropriate.
Item 1 AEB/CSB Akstulewic / Lane Bypass Leakage Detroit Edison's response to this verbal equest of maximum bypass leakage and supplemented Tr.ble 6.2-2 information is enclosed as Attachment 1.
Item 2 SEB Q130.9 High Density Spent Fuel Rocks Detroit Edison's response to this SEB question is enclosed as Attachment 2.
Item 3 ETSB H.III.D.1.1 Primary Coolant Outside Containment Detroit Edison will revise theFSAR as detailed on Attachment 3 to further clarify this item.
go oI 1;,jo 1
8106 3o 0 5'V4 A -
Mr..L. L. Kintner June 8, 1981 EF2 - 53,492.
Page 2-1 Item 4 RSB. Savannah River Non-Qualified Equipment in Accidents Det'roit. Edison's response to this item is enclosed as l Attachment 4.
Item 5 ICSB II.D.3 V,alve Position Indication .
In response to a concern from Mr. Gerry Mauch, Detroit Jdison has revised its response to this item. Please see Attachment 5..
Sincerely,
. mA ESM William F. Colbert Technical Director j Enrico Fermi 2 l RMB/WFC:jl l Attachments l
cc: Mr. B. Little
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i Mr. L. L. Kintner ,
June 8,-1981 l EF2 - 53,492 i Page 3 !
bec:. T. A.-Alessi~
-E. L. Alexanderson
-R. M.. Berg..
W. F. Colbert' F.-E. Gregor .
~
E. P. Griffing W.-R. Holland J. W. Honkala W. H. Jens
. E. Lusis.
P. A. Marquardt L. E. Schuerman H. Tauber
, R. A. VArace A..E..Wegele Document' Control J. E.-Slider i
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,~ $TDV"WW f. .
s+2.- sa y12 June 8,1981 To:
/
R. M. Berg /L. E. Schuerman From: R. J. Beaudry/M. L. Batch Item
- 4. NRC is requesting that Edison provide its best estimate of the maximum bypass leakage that could be expected (NRC suggests 2%) . A sophisticated calculation is not necessary, but a number is required. NRC notes that FSAR Appendix A was apparently deleted. This appendix could be used as a basis for determining bypass leakage should Edison choose to do so. As a minimum, however, the NR (Fraak Akstulewicz and John Lane) requires that we pro-vide a supplement to Table 6.2-2 which identifies the lines connected to the primary containment and pass through the
- secondary containment which could be potential bypass paths.
Response
We estimate that the amount of primary - containment leakage expected to bypass the SGTS should be no greater than 4%;
moreover, we anticipate that the actual amount will be sign-ificantly below this value.
The attached table lists all potential bypass leakage paths in the broadest interpretation of Section B.5 of Branch Technical Position CSB 6-3. However, as accepted in CSB 6-3, Fermi 2 employs air or water sealing systems which el-iminate leakage through certain valves.
'1. Ths MSIVLCS eliminates leakage through main steam lines and steam line drains.
- 2. The IWMS suction lines - penetration X-213A and B - are sealed with water in the torus.
- 3. All leakages to the feedwater system will be sealed with water in that system.
The Bypaso Leakage Program will maintain a running total of
~
leak rate measurements through all other valves in the table and compare 'it with the maximum allowable. Decision on valve maintenance will be based on the closeness of these two num-bers.
RJB: HLB:sim
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+
BYPASS IFENE PHOCitiv ,
Primary P&ID Containnent Ntzabers Valves Size,
- Penetration Line 6M721- Tested In.dia. Test IEMAGE TO STEAM MANIEOID AT 'IURBINE-Gt24 ERA' LOR X-7A Main Steam Line 2089 V17-2003 26 MSIVICS
- V17-2007 26 V17-2101 26 X-7B Main Steam Line 2089 V17-2001 26 V17-2005 26 i V17-2102 26 X-7C Main Steam Line 2089 V17-2002 26 V17-2006 26 V17-2099 26 X-7D Main Steam Line 2089 V17-2004 26 V17-2008 26 V17-2100 26 LEAKAGE 'IO MAIN 00tn2JSER X-8 Steam Line Drains 2089 V17-2009 3 App. J Type C V17-2010 3 App. J Type C t
V10-2003 3 MSIVILS X-10 BCIC Steam Line 2044 V17-2036 1 Sect XI, Cat. B Drain V17-2037 1 Sect XI, Cat. B X-ll llFCI Steam Line 2035 V17-2024 1 Sect XI, Cat. B Drain V17-2025 1 Sect XI, Cat. B
- Testing Program for MSIVILS described in ESAR 9A.3.4 MIB/dk 6-5-81
BYPASS IEAKAGE PHOGRAM PAGE 2
' Prinary P&ID (bntaintment Nuobers Valves Size, I
Penetration 'Line 6M721- Tested In.dia. Test-IDEAGE 10 FEEIMATElt SYSlm X-9A' Feedwater 2023 V12-2002 20 App. J Type C V12-2008 20 . App. J Type C X-9B Feedwater 2023 V12-2001 20 App. J Type C V12-2007 20 App. J Type C X-21CB SWIS 4100 V8-3847 4 App. J Type C V8-3849 4 App. J Type C X-213A Tess 4100 V8-3832 6 App. J Type C V8-3834 6 App. J Type C X-213B SWE 4100 V8-3831 6 App. J Type C
.V8-3833 6 App. J Type C X-227A *1ws 4100 V8-3848 4 App. J Type C V8-3850 4 App. J Type C 4
IEAKAGE TO C0t0WSATE STORAGE TANK i
11ICI 2035 V8-2198 10 Sect. XI Cat. B X-9A 4
V8-2200 10 Sect. XI Cat. B X-98 ICIC 2044 V8-2232 24 Sect. XI Cat. B IIAKAGE 'IO RAD WASTE BUIIDLNG X-18 Drywell Floor Sunp 2032 V9-2044 3 App. J Type P V9-2005 3 App. J Type C Drywell BJuipment 2032 V9-2022 3 App. J. Type C X-19 S*C V9-2023 3 App. J Type C MLB/dk 6-5-81 ,
/P77)9CerMM JL-SF& - TJ'{12. -
RESPONSE TO STRUCTURAL ENGINEERING BRANCH QUESTIONS OF JUNE 1, 1981 130.9 HIGH DENSITY SPENT FUEL RACKS QUESTION 1 Indicate whether the proposed fuel racks modifications conform with the NRC position on " Fuel Pool Storage and Handling Application" dated April, 1978 and amended January 1979. If any deviations exist identify and justify these deviations.
RESPONSE
The proposed fuel rack modifications do conform with Sawg the NRC position dated April 1978 and amended 1979.
QUESTION 2 The responses to Questions 130.7.2 and 130.7.3 are not satis-factory. Provide the following:
- a. Justify why the spent fuel pool liner was designed in accordance with ASME Boiler and Pressure Vessel Code, Section VIII, Division I instead of Section III, Subsection NE and ACI 359.
- b. Indicate in detail,the methodology used to demonstrate the leak tight integrity of the fuel pool liner when i
subjected to the postulated fuel assembly drop over spent fuel racks .. directly falling over the fuel l l
pool Liner. The fuel assembly drop should be analyzed for tilted position and straight drop.
. RESPONSE TO STRUCTURAL ENGINEERING BRAHCN QUESTIONS OF JUNE 1, 1981 PAGE 2 I
RESPONSE
- a. The following codes and standards are used for .
the design of the spent fuel pool liner.
(i) ASME Boiler and Pressure Vessel Code,Section III, Division 2, also called as ACI 359 (Liner plate design).
(ii)
ASME Boiler and Pressure Vesse' Code, Sectior VIII, Division 1, Subsection B (for welds).
(iii) ACI 347, recommended practice for concrete formwork (for tolerances).
It is to be noted that ASME Boiler and Pressure
- Vessel Code,Section III, Division 1, Subsection NE is not applicable for spent fuel, pool liner since Subsection NE establishes rules for metal contain-ment vessels and their appurtenances.
~
- b. The effect of fuel assembly straight drop on to thefuelpoollinerwasevaluatd@Qingthe principles of conservation of energy and con-servation of momentum. The following assumptions were made in the analysis:
- - Fuel assembly weights 700~ pounds (600 lbs sub-merged) l
- Nose of assem'bly assumed 1.5 inches diameter (based upon ass'umed partial crushing of the nose, "TF 44W W WEWWW W** -g-- -@ -
RESPONSE TO STRUCTURAL ENGINEERING BRANCH QUESTIONS OF JUNE 1, 1981 PAGE 3 but this crushing energy neglected in the analysis).
Fuel assemtly assumed rigid as far as energy absorption (all kinetic energy transferred to liner and slab)
Dragcoefficientassumeck . 86 (based upon a cylinder)
Overall behavior of the slab was assumed to be formation of yield lines in a 34' x 42' fixed edge condition slab The. analysis used the Standford Research formula to investigate'the local penetration of the liner plate.
Yield line theory was used to evaluate the overall
~~
behavior of the slab.
Results of the analysis show that the fuel assembly
. can be dropped 6'-6" (nose of assembly 6'-6" from linct plate) without penetrating the liner or causing over-all slab instability.
The assumption of fuel assembly rigidity is a gross
, approximation, of course, which leads to extremely conservative results. The degree of conservatism, however, cannot be estimated unicas some experimental G
e W
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=v v6 -- - - - -v -r , -__ _v- . -,__ *Mg - , __ gg>get 7 ,
r
RESPONSE'TO STRUCTURAL ENGINEERING BRANCl! QUESTIONS OF JUNE 1, 1981 PAGE 4 work has been done measuring the energy in deforma-
' tion on impact; we are not aware of any such work.
The assumption of fuel assembly rigidity would give even more conservativ9 results for the tilted case.
Therefore, the straight drop case would be limiting and no additional work was done for the tilted case.
& shy o( N % 4 A %d nuw. An. ovw sy,n 7. t\ r ob h um '.yl 6 L < yout, t. 2, Mm 3.
QUESTION 3 .
With regard to the fuel assembly drop on top of the rack, provide the following: ,
- a. The acceptance criteria used for this case. -
- b. Detailed descriptions of the method used to satisfy the acceptance criteria. ,
ew
- c. Comparison between the drops in tne titled position, straight drop on the middle of the rack and on the edge of the rack.
- d. Indicate whether other modes of failure of the racks exist beside crushing.
- e. Discuss these same concerns for the stainless steel racks and for the. aluminum racks.
e RESPONSE The acceptance criteria used for the drop test is a calculation of the stress induced at the base of the rack, and a calculation of the local stress at the point of impact at the top of the rack. Also calcu-
l RESPONSE TO STRUCTURAL ENGn1EERING BRANCH QUESTIOMS OF JUNE 1, 1981 PAGE 5 lated is a conservative estimate of the deformation of the rack base for the case of a falling object impacting directly on the rack base. The method used involves calculating the total energy of the falling
object, and assuming that all of this energy is absorbed as strain energy in the rack. This item gives an esti-mate of the peak force developed by the impact. For a corner impact, this peak force leads to direct load plus ; two bending movements at the rack base. The stress induced in the rack is shown to be below yield.
4 The cage of corner impact leads to the highest rack stress and.is the only case investigated. The local impact stress at the point of impact is computed using wave propagation theory and a knowledge of the impact-ing body velocity. It is shown that the local stress at the point of impact is below yield.
The possibility of performation of the base plate by an object falling directly to the bottom is addressed.
An energy balance is'used to calculate an estimate
~.
The deformation is of local plate deformation.
' shown to be less than the plate thickness so that no perfor/ationoccurs. This calculatior. is made neglect-ing any ef fect of strain hardening and neglecting the retarding ef fect of ,the water so that the results are conse r vr.tive . _
RESPONSE TO STRUCTURAL ENGINEERING BRANCH QUESTIONS OF JUNE 1, 1981 PAGE 6
~
QUESTION 4 .
Indicate whether material, fabrication, installation and quality control of the racks conform with Subsection NE of the i
ASME Code.
RESPONSE
The fuel rack structural design is per ASME Section NF, the material specifications are per ASTM, and ASME Section III is followed for all items.
QUESTION 5 In.Section 6.2.4 of the Joseph Oat Corporation Report, it is stated that only fluid damping is included in the analysis and is simulated by inclusion of appropriate equivalent linear damping. Indicate what this damping value is and justify the damping value used in the analysis.
RESPONSE
. Section 6.2.4 is being revised to reflect the racks have been qualified by analysis without taking credit for any fluid damping. When and if structural damp-ing has been included, a maximum of 4% during SSE conditions has been used in accordance with URC Regulatory Guide 1.61. The seismic modelin the e
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RESPONSE TO STRUCTURAL ENGINEERING BRANCH QUESTIONS OF JUNE 1, 1981 PAGE 7 Joseph Oat Report contains provisions for inclusion of fluid damping. The damping model used is based on the losses incurr,ed in forcing a fluid in'and out of a narrow gap. As noted above, qualification runs have assumed zero fluid damping values. If fluid damping were to be included, the values chos .: would be such as to insure that the ectal damping from all contributions (structure, material, fluid)did not exceed 4% during SSE conditions.
QUESTION 6 .
Indicate why the empty or nearly empty rack case was,not used in the analysis to predict the maximum rack displacement.
RESPONSE
The Joseph Oat Report is being revised to show the results for the nearly empty rack.
QUESTION 7 Indicate how the increase,in the value of the plant design earthquake and consequently the change in the floor response spectrum could affect the seismic analysis done so far for .
1 the fuel pool racks.
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RESPONSE TO STRUCTURAL ENGZNEERING
, BRANCH' QUESTIONS OF JUNE 1, 1981 PAGE 8
RESPONSE
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RESPONSE TO STRUCTURAL ENGINEERING BRANCH QUESTIONS OF JUNE 1, 1981 PAGE 9 a
QUESTION 8 Because different type (169, 108 and 35 cells) modules were used in the proposed modification with different sizes and
' weights, indicate which type was used in the seismic and slicing analysis. Indicate also how other types were quali-fled for the postulated loadings.
. RESPONSE All three modules were qualified separately and completely for seismic and other postulated load-
~
ings.
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EF-2-FSAR k'IDW W W 3 us-c y n H.III.D.l.1 Primary Coolant Outside Containment H.III.D.l.l.1 Statement of Concern Parts 20 and 100 of Title 10 of the Code of Federal Regulations specify radiation limits and guidelines for licensed facilities to ensure the protection of public health and safety. In a power reactor, many systems that may or will contain significant radio-active liquid and/or gas inventories af ter a serious transient or accident have components located outside containment. At TMI-2, the major radioactive releases appear to have come from leaks in such systems. Leakage from the systems must be maintained as low as practicable to prevent releases of significant quantities of radioactive material when the systems are operated. The plant operating staff should know the leakage rate of each system and have positive control over them *.o ensure the maximum availability of the equipment.
H.III.D.l.l.2 NRC Position Applicants shall implement a program to reduce the leakage from systems outside containment that would or could contain signifi- ,
cant radioactive inventories during and after a serious transient :
or accident to as low as practicable rates. A summary description and the initial leak-test results of the program shall be provided.
Systems that should be leak tested include the following:
! a. Residual heat removal l
l b. Containment spray recirculation
- c. High-pressure injection recirculation
- d. Containment and primary coolant sampling
- e. Reactor core isolation cooling
- f. Waste gas including headers and cover gas system, decay system, or storage system l
The testing of gaseous systems should include helium leak detection or equivalent methods. Consideration should also be given to 3 i
design changes that reduce potential release paths.
i .
t H.III.D.l.l.3 Detroit Edison Position Table H.III.D.l.1-1 identifies those liquid and gaseous systems a containing radioactive fluids following a serious accident. Leak-test methods and procedures, described in Subsection H.III.D..l.l.4, and the baseline leak-test data are the bases for the program to maintain the leakage rates of all the applicable systems as low Q as practicable.
H.III.D.l.1-1 Am?ndment 33 - March 1981
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. EF.2-FSAR
'H.III.D.l.l.4 Leakage Reduction Program s
H.III.D.l.l.4.1 Design Basis
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The leakage reduction program described in this section is based on Requirement 2.1. 6.a of UUREG-0 578 (Reference 1). It also takes into account the requirements of i tem II I .D. l .1 of NUREG- /
0660, NUREG-0694, and NUREG-07 37, " Primary Coolant Sources Out- /
side Containment" (References 2, 3, and 4).
\
The program includes measures to reduce and maintain Icakage to as low as practical for systems outside primary containment that could contain highly radioactive fluids dur{ng,a serious - 3-transient or accident. 7Y'$ M#9 t.x;ac r&.s7J s47 m7.f4.M444 Ne7~ $Q'g%'8b;;r gpg# $ wg W,. &
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4 ypplicabh systems a+--incp :ted f or--+xte rs.2 ' ic axow = ne r* re a
leab t : : t 9 W'hJc,H =4 _ ; y. ; w a e n d p r " ; - ' ' 7 - i i' ~-
SJBA' ,'[. ,
aft ^r. Visual inspections are conducted on accessible portion) N ##/54"'
of systems during system operational testing or during normal U#4 ##j
K system operation. Leak-rate testing is performed on specific j((j7y-[a 4 valves on systems that provide an interface to equipment or 3 k systems located external to the secondary containment and which jgy,,g,g, ,
can bypass secondary containment. AW'G#4 ',','
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Records are maintained on the inspections and tests performed #6 J'u 9##7 and are used to identify chronic and generic leakage problems 77 WVE b
g in order to implement measures. These reconds modif.ications and/oravailable are also made corrective to maintenance the plant -
/X4fC (E operators.
H.III.D.l.l.4.2 Descriotion -
Table H.III.D.l.1-1 lists systems outside primary containment that could contain highly radioactiva fluids and that would be
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used in mitigation of an accident. Table H.III.D.L.1-2 lists systems outside primary containment.that should rot contain highly radioactive fluids or would not be used in the mitigs-tion of an accident. Only the systems in Table H.III.D.1.1-1 are included in the leakage reduction program.
Each system identified in Table H.III.D.l.1-1 has a surveillance test pack.' age containing the following items:
-a_ Rimolified engineering' sketch indi-'t!.g system test boundaries. into as o used only as an aid in visualizing tha _ omponents 6 " r em to be tested; detaile . st boundaries are verified tr .. ha current r 'oton of the system ' drawings before initial t+. 'a .
dLA. Statement explaining why that particular system is an impor tant factor in mitigating a serious tran.cient or accident.
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Me H.III.D.l.1-2 Amendment 33- - March 1981
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EFo2-FSAR b g. Description cf system conditions and plant operating
) conditions necessary to conduct each component leak-rate test, along with an explanation substantiating the test boundary. The division of the system into ,
parts for testing purposes and the emission of parts of the system from testing is included in this explanation. ,
- c. g. '
Elaboration of special test methods necessary to supplement the general test methods.
E g. Data sheet listing system components to be tested.
The data sheet identifies drawing numbers and provides
. for recording of measured leakage.
H.III.D.l.l.4.3 Evaluation The leakage reduction program is designed to minimize leakage from systems outside primary containment that could contain large radioactive inventories after a serious transient or accident; minimizing this leakage will diminish the potential for release of significant amounts of radioactive materials to the anviron-ment. Periodic leak testing, followed by preventive and cor-
.rective maintenance, ensures minimum leakage on a continuing O basis.
o E.III.D.l.l.4.4 General Test Methods
<c In order to ensure system integrity, two general types of tests 3 are conducted. The first type of test consists of periodic H visual inspection of systems and components during system opera-tional testing or during normal system operation. This inspection V is made to identify leakage into the secondary containment through D valve stems, pump seals, fittings, relief valve discharge lines, T drains, vents, and instrument loops. The second type of test s is a periodic individual component leak-rate test used to deter-l 4 mine leakage through valve seats into interfacing systems outside of g p g g ary containment. Leak-rate testing is accomplished in ccr,ur.0 :en with ASME XIstest requirements (Reference 5) .
The following general test methods are used to determine leakage from the systems identified in Table H.III.D.l.1-1.
. ~~p o -o a. Liquid Systems - Sy;tems or portions of systems that
-fej contain liquids during accident conditions are filled k
v7 y} with system operating liquidaand pressurized to opera-tional test pressure or normal operating pressure 7 .< 9 before conducting visual inspections for external
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leakage. Where evidence of leakage exists (e . g . ,
water stains on lc.gging), the source of leakage is
> } verified by visually observing the suspected leakage point for a minimum of 5 minutes. Where measurable
$c4b lj .. V leakage exists, specific techniques for collecting l ]%
Jfe gg the leakage are implemented (see below). Valve seat leaA rate measurements are cdhducted in ac.ordance
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H.III.D.l.1-3 Amendment 33 - March 1981
- *WPW * ,e ..eg,n e same, w .p = e . a == m * * +
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,$q y,I 1 ith ASME XI procedures. Measured Icakage is recorded and is used as a basis for impicmenting corrective
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3 action.
N' B t yh diaseous Systems - For systems or portions of systems
% % FIthatcontaingasesduringaccidentconditions,a f pressure drop test or measurement required to maintain system pressure provides of makeup gas rate an For such s
3 overall indication of system integrity.
3 tests, the system is pressurized to operational test 9g kNigpressure a clean or source normal of operating air.7) Specificpressure usingofnitrogen points leakage ggs got are located by helium Y leak detection or equivalent testing methods. Valve
)Sntotneseconcarycontatn=ent kgg Q seat leak-rate measurements are conducted in accordance jwithASMEXIprocedures.
. Specific Techniques for Collecting Liquid Leakagq
- a. Drains, Vents, and Reliefs is - Leakage from valves collected by connecting piped to open connections polyethylene tubing from the open end to a suitable collection container
- b. Pumo seals is
- For pumps with shaft leakoff drain lines, collected by connecting polyethylene tubing -
. leakage from the drain line to a suitably sized container. ,f'
- For pumps without drain lines, a temporary means of collecting leakage is used according to pump con-figuration and accessibility In all cases where polyethylene tubing is used, the length of
< tubing is minimizedWhere in order to collect collection the smallest of leakage amount of is impossible due leakage possible.
to physical configurations, or if the leakage is too small to be collected, a visual method for estimating leakage is used.
For example, 20 drops is equivalent to 1 ml.
H.III.D.1.1.5 References
- 1. U.S. Nuclear Regulatory Commission, TMI-2 Lessons Learned Task Force Status Repor t and Short-Term Recommendations,
' NUREG-057 8, J uly 19 79.
- 2. U.S. Nuclear Regulatory Commi'ssion, NRC Action Plan Developed
- as a Result of the TMI-2 Accident, NUREG-0660, Vols. 1 and 2, e May 1980.
'3 . U.S. Nuclear Regulatory Commission, TMI-Related Requirements for New Operating Licenses, NUREG-0694. J une 1980.
H.III.D.l.1-4 Amendment 33 - March 1981
~
EF-2-FSAR
--q _4. U.S. Nuclear Regulatory Commission, Clarification of TMI V Action Plan Recuirements, NUREG-0737, Octobot 1980.
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- 5. ASME Boiler and Pressure Vessel Code,Section XI.
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EF 2-FSAR SYS* EMS OUTSIDE PRIMARY T NTAfh?.D4T TABI.E H. III .O .1.1-1
'I1IAT CCCID CO*;!A!N HICHI.Y RADIOACTIVE Ft.UIDS O Reactor core isolation cooling Residual heat removal Contain=cnt spray
. Suppression pool cooling Ion-pressure coolant injection
' Shutdown cooling Core spray l Reactor water sample Reactor water cleanup Canbustible gas control High-pressure coolant injection ,
Standby gas treat.ent Control rod drive discharge header's,
- $W"44/WA/&hSANflW(y fl7GA-l .
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H.III.D.l.1-6 A:nendment 33 - March 1981 P-----,-r...._ . '~ ,,*E-*- e . ,-.. m .,, , _ , , == **
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. ' TAat.E H III .D. l.1-2 SYS n4S Or1TSIDE PRIMARY CD'ITA!.WD4T
- h. '
THAT WOU!D IKTl CC!r*AIN IttClit.Y MDICACTIVE FI.UIDS System Carnent RER fuel pool cooling Not directly af fected by accident.
Standby liquid control Injects fluid and does not circulate
- reactor coolant.
General service water / emergency Do not circulate reactor coolant and equipment service water could becom conta:ninsted only due to system leaks.
Reactor building closed cooling Do not circulate reactor coolant and water /cmergency equipment could become conta:ninated only due cooling water to system leaks.
Condencate storage Could beccrue contaminated only due to isolation valve leakage.
Domineralized water makeup Could tecoce contasinated only due to isolation valve leakage.
Torus water management Isolated during IOCA and not required for accident mitigation.
Control air / station air Would require system failure.
Fuel-pool cooling and cleanup Not directly affected by accident.
Main stea:s lines Would require f ailure of MSIVs and f ail-ute of MSIV leakage control system.
Feedwater lines Would require failure of isolation valves.
. Drywell cooling system Uses RSCCW or EECW and is not needed for safe shutdown of plant.
RER
- tea:s condensing Not required for accident mitigation.
Reac or building floor / equipment Not required for accident mitigation.
dr > ins Minimizing leakage frern systees in' Table H.III.D.l.1-1 minimizes input to this syste:n.
Padwaste Not required for accident :nitigation.
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H.III.D.l.1-7 Amendment 33 - March 1981
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- 16. USE OF NON-SAFETY GRADE EQUIPMENT IN' SHAFT SEIZURE ACCIDENT ANALYSIS (15)_
The applicant must submit additional information with regard to pump shaft seizure accident analysis without allowance for the use of non safety grade equipment.
RESPONSE -
The recirculation pump seizure event is considered to be en extremely unlikely event and as suca falls -
into the category generally classified as an accident.
The event is evaluated as a limiting fault. The potential eifects of the hypothetical pump seizure
" accident" are very conservatively bounded by the effects of the DBA-LOCA.
- This is easily verified by comparis.on of the two events. In both accidents, the recirculation driving-
- loop flow decreases extremely rapidly. In the case of seizure, stoppage of the pump occurs; for the DBA-LOCA, the severence of the line has a similar, but more rapid and severe influence. Following a pump seizure event, water level s maintained, the core remains submerged, and this provides a continuous core cooling mechanism. However, for the DBA-LOCA complete-flow stoppage occurs and water level decreases due to loss of coolant, thus resulting in uncovery 1
of the reactor core and subsequent overheating of the fuel-rod cladding. Also, complete depressuriza-tion occurs with the DBA-LOCA, while reactor pressure 1
9
does not significantly decrease for the pump s41zure event. Clearly, the increased temperature ,
of the fuel cladding and reduced reactor pressure for the DBA-LOCA both combine to yield a much more severe stress and potential for cladding perfora-tion for the DBA-LOCA than for the pump seizure.
Therefore, it is concluded that the potential effects
- of the hypothetical pump seizure accident are very '
conservatively bounded by the effects of the DBA-LOCA and a specific core performance analysis or ra'dio-Logical evaluation is not considered necessary. How-ever, to be completely responsive to the NRC question,.
the following narrative is provided to show the impact of not taking credit for non-safety-grade equipment to terminate this event:
- 1. Level 8 Turbine Trip The FSAR analysis of the pump seizure event assumes that the vessel water level swell due to pump seizure will cause high water level (Level 8) trips of the main turbine and the feedwater pumps, and indirectly initiates a reactor sc am as a result of the turbine trip. The level 8 trip for each feedpump and_ M O e, turbit.eggnerator are designed tObec-jinf71e--failbfe#'
Fftbr-- IT,'"for some unknown rea.~n,1tffr'! Walls c6 operat//e, high high moisture-revefHnter%F r, tEu~TdTIrine'would cause,
. l j
vibration to the point where mechanical vibration
. will trip the turbine. In case of the pump seizure without an L8 h ip, the event is less severe than the analysis in the'FSAR with the L8 trip for the following reason: A pump seizure, should it occur, would result in core flow reduction which reduces
' the core powcr and surface heat flux due to the .
effect of the negative void reactivity coefficient.
Hence, the surface heat flux existing when the turbine trip occurs is lower because the turbine trip occurs later. Therefore, a loss of Level 8 trip would result in a less severe event consequence -
from the fuel han that depicted in Subsection 15B.3.3.
- 2. Main Turbine Byoass System As a result of the NRC's concern regarding reactivity effects of pressure transients, GE and the NRC met on November 20 and 21, 1978 for a comprehensive review of turbine trip and load reject transients without bypass. The principal conclusion of that meeting was that the most limiting BWR transient
, event which takes credit for nonsafety-grade equip-ment is the feedwater controller failure. Analy. sis indicates that a ACPR increase of approximately 0.08 applies to this transient without a functioning main turbine bypass system.
~
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For recirculation pump seizure with a failure of turbine bypass system, the increase of ACPR would
.be less than that for the feedwater controller . .
failure for the follo ing reason. As this event occurs, the reactor power drops significantly within the first 2 seconds due to decreased core flow. Therefore, by the time of the turbine trip,
~
the reactor power is at e low level. The core power is he main parameter which relates to the fuel thermal limit. The effect of f ailure of the main turbine bypass system to stop the steam flow retains pressure on the core but contributes only a small positive reactivity feedback. This is a
' secondary effec't of much less significance than the reactivity decrease due to fluid flow decreasing
. through the core.
The increase of core power is more severe for feed-water controller. failure (increasing) event than for a recirculation pump failure.because it occurs at a higher power level.
- 3. Relief Function of Safety / Relief Valver The contribution of MCPR from taking credit for the relief function rather than the safety function of safety / relief valves is not significant because the MCPR always reaches its lowest value before opening of the relief valves.
9
+
Analyses of recirculation pump seizure where coolant flow rate drops rapidly have shown that MCPR does not increase
. significantly before fuel surface heat flux begins dropping enough to restore greater ther' mal margins as the plant intrinsically responds to the reduced flow rate. The effect of not taking credit for non-safety-grade equipment is a ACPR increase of 0.08. Therefore, the MCPR for pump seizure event -
is still well above the safety limit of 1.06.
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Integration into op2rator trcining --
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Other alarms during emergency and,nced for priori- kh>
H.II.D.3.3 Detroit Edison Position '
Tha Farmi 2 plant has a valve-position-indication system that uccc cne pressure switch on each of the 15 safety and relief volycc.
The existing tailpipe temperature-monitoring system provides a backup and monitors valve leakage as its normal ,
I function. .
H.II.D.3.4 s
- Modifiestions H.II.D.3.4.1 . Design Basis -
An in-containment, tailpipe-mounted, pressure-switch system provides of the relief status valves information and safety viavalves.
control room indicating lights -
This system provides the following information cnd obnor=al operating conditions: to the plant operator during normal
' a ." FO itiv indication of valv pe r i t ; a.. , .aciucing the i st'Luk v e s.. . 1+c ca..ditir- _
- b. Positive identification of the specific valve or valves that are open e r
e ?. .;a u uv i v u v:
depressuri:r.icn
.'.; c '~ S nn ne the au t o- e i e system (ADS) in the control room 3Y i ,U 5fL& 'f' By being provided with the immediate indication and annunciation of the valve opening and the identification of the valve, the plant operator can initiate recommended actions to control or rectify the situation.
The NRC has specified in NUREG-0578 and NUREG-0737' (References 1 and 2) that components of the safety / relief valve 5 cystem must (SRV) monitor be qualified for the appropriate environmental condi- i tions to plant operation. be experienced under normal and abnor=al conditions of
- These environmental conditions include temper-of the component or humidity, and also the seismic acceleration ature, pressure, and aystem are attached. system to which the components of the SRV The system instruments are qualified to IEEE 323-1974 and
- IEEE 344-1975. . N She' power for this system comes from a reliable source that *
-in not affected by the loss of offsite power.
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H. II . D. 3 Amendment 33 - March 1981 P00RORIGK1
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