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| | issue date = 06/28/1996 | | | issue date = 06/28/1996 |
| | title = Forwards Rev 14 to NMP Unit 1 Updated Fsar,Including Changes to QA Program Description & Annual 10CFR50.59 Safety Evaluation Summary Rept | | | title = Forwards Rev 14 to NMP Unit 1 Updated Fsar,Including Changes to QA Program Description & Annual 10CFR50.59 Safety Evaluation Summary Rept |
| | author name = SYLVIA B R | | | author name = Sylvia B |
| | author affiliation = NIAGARA MOHAWK POWER CORP. | | | author affiliation = NIAGARA MOHAWK POWER CORP. |
| | addressee name = | | | addressee name = |
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| =Text= | | =Text= |
| {{#Wiki_filter:REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9607020103 DOC.DATE: 96/06/28 NOTARIZED: | | {{#Wiki_filter:}} |
| YES DOCKET FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara owe 05000220 AUTH.NAME AUTHOR AFFILIATION SYLVIA,B.R.
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| ;Niagara Mohawk Power Corp./of p RECAP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
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| ==SUBJECT:==
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| Forwards Rev 14 to NMP Unit 1 Updated FSAR,including changes to QA program description 6 annual 10CFR50.59 safety eva 1ua t ion summa ry r ep t.DISTRIBDTION CODE: A053D COPIES RECEIVED!LTR I ENCL J I SIZE: TITLE: OR Submittal:
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| Updated FSAR (50.71)and Amendments NOTES: RECIPIENT ID CODE/NAME PD1-1 PD INTERNAL: AYERS FILE CENTER 01 EXTERNAL: IHS NRC PDR COPIES LTTR ENCL 1 0 2 2'2 2 1 1 1 1 RECIPIENT ID CODE/NAME HOOD,D AEOD/DOA/IRB RGN1 NOAC COPIES LTTR ENCL 1 1 1 1 1 1 1 1 D U N NOTE TO ALL"RIDS" RECIPIENTS:
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| PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT.415-2083)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!TOTAL NUMBER OF COPIES REQUIRED: LTTR 11 ENCL lo N
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| NIAGARA MOHAWK C EN ERATION BUSINESS CROUP B.RALPH SYLVIA Executive Vice President Generation Business Group Chief Nuclear Officer 300 ERIE BOULEVARD WEST.SYRACUSE, NEW YORK 1 3202/TELEPHONE (31 5)4284983 June 28, 1996 NMP1L 1090 U.S.Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 10 C.F.R.550.71(e)10 C.F.R.$50.54(a)(3) 10 C.F.R.$50.59(b)(2)
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| RE: Nine Mile Point Unit 1 Docket No.50-220
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| ==Subject:==
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| Submittal of Revision 14 to the Nine Mile Point Nuclear Station Unit 1 Final Safety Analysis Report (Updated), Including Changes to the Quality Assurance Program Descnption, and the Annual 10 C.F.R.5$0.$9 Safety Evaluation Summary Report Gentlemen:
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| Pursuant to the requirements of 10 C.F.R.$50.71(e), 10 C.F.R.$50.54(a)(3), and 10 C.F.R.$50.59(b)(2), Niagara Mohawk Power Corporation hereby submits Revision 14 to the Nine Mile Point Nuclear Station Unit 1 Final Safety Analysis Report (Updated), including changes to the Niagara Mohawk Power Corporation Quality Assurance Topical Report, and the annual Safety Evaluation Summary Report.One (1)signed original and ten (10)copies of the Unit 1 FSAR (Updated), Revision 14, are enclosed.Copies are also being sent directly to the Regional Administrator, Region I, and the Senior Resident Inspector at Nine Mile Point.The Unit 1 FSAR (Updated)revision contains changes made since the submittal of Revision 13 in June 1995.In addition, many of the Unit 1 FSAR (Updated)Sections have been reformatted in their entirety to eliminate blank pages, establish a uniform left-margin justification format, and to reorganize the information into"Text/Table/Figure" order.The certification required by 10 C.F.R.$50.71(e)is attached.Enclosure A'provides the identification, reason, and basis for each change to the quality assurance program description, Unit 1 FSAR (Updated)Appendix B, in accordance with 10 C.F.R.$50.54(a)(3)(ii).
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| 9607020i03 960628 PDR ADQCK'5000220 K PDR
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| Page 2 The enclosed annual Safety Evaluation Summary Report (Enclosure B)contains brief descriptions of changes to the facility design, procedures, tests, and experiments.
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| None of the Safety Evaluations involved an unreviewed safety question as defined in 10 C.F.R.$50.59(a)(2).
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| Very truly yours, B.Ralph Sylvia Chief Nuclear Officer BRS/JJL/kap Enclosures pc: Mr.T.T.Martin, Regional Administrator, Region I Mr.D.S.Hood, Senior Project Manager, NRR Mr.B.S.Norris, Senior Resident Inspector Records Management l'.i'l l NIACARA MOHAWK G E N E RAT I 0 N BUSINESS GROUP B.RALPH SYLVIA Executive Vice President Generation Business Group Chief Nuclear Officer 300 ERIE BOULEVARD WEST.SYRACUSE, NEW YORK 13202/TELEPHONE (3 1 5)428-6983 June 28, 1996 NMP1L 1090 U.S.Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555'0 C.F.R.$50.71(e), 10 C.F.R.550.54(a)(3) 10 C.F.R.$50.59(b)(2)
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| RE: Nine Mile Point Unit 1 Docket No.50-220
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| ==Subject:==
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| Submittal of Revision I4 to the Nine Mile Point Nuclear Station Unit I Final Safety Analysis Report (Updated), Including Changes to the Quality Assurance Program Description, and the Annual 10 C.F.R.550.59 Safety Evaluation Summary Report Gentlemen:
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| Pursuant to the requirements of 10 C.F.R.$50.71(e), 10 C.F.R.$50.54(a)(3), and 10 C.F.R.$50.59(b)(2), Niagara Mohawk Power Corporation hereby submits Revision 14 to the Nine Mile Point Nuclear Station Unit 1 Final Safety Analysis Report (Updated), including changes to the Niagara Mohawk Power Corporation Quality Assurance Topical Report, and the annual Safety Evaluation Summary Report.One (1)signed original and ten (10)copies of the Unit 1 FSAR (Updated), Revision 14, are enclosed.Copies are also being sent directly to the Regional Administrator, Region I, and the Senior Resident Inspector at Nine Mile Point.The Unit 1 FSAR (Updated)revision contains changes made since the submittal of Revision 13 in June 1995.In addition, many of the Unit 1 FSAR (Updated)Sections have been reformatted in their entirety to eliminate blank pages, establish a uniform left-margin justification format, and to reorganize the information into"Text/Table/Figure" order.The certification required by 10 C.F.R.$50.71(e)is attached.Enclosure A provides the identification, reason, and basis for each change to the quality assurance program description, Unit 1 FSAR (Updated)Appendix B, in accordance with 10 C.F.R.$50.54(a)(3)(ii).
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| Page 2 The enclosed annual Safety Evaluation Summary Report (Enclosure B)contains brief descriptions of changes to the facility design, procedures, tests, and experiments.
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| None of the Safety Evaluations involved an unreviewed safety question as defined in 10 C.F.R.$50.59(a)(2).
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| Very truly yours, B.Ralph Sylvia Chief Nuclear Officer BRS/JJL/kap Enclosures pc: Mr.T.T.Martin, Regional Administrator, Region I Mr.D.S.Hood, Senior Project Manager, NRR Mr.B.S.Norris, Senior Resident Inspector Records Management J
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| UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Niagara Mohawk Power Corporation (Nine Mle Point Unit 1)Docket No.50-220 CERTIFICATION B.Ralph Sylvia, being duly sworn, states that he is Chief Nuclear Officer of Niagara Mohawk Power Corporation; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this certification; and that, in accordance with 10 C.F.R.$50.71(e)(2), the information contained in the attached letter and updated Final Safety Analysis Report accurately presents changes made since the previous submittal necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirement and contains an identification of changes made under the provisions of$50.59 but not previously submitted to the Commission.
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| B.Ralph ylvia Chief Nuclear 0 ficer Subscribed and sworn to before me, a Notary Public in and for the State of New York and County of , this~~dayof , 1996.Notary Public in and for County, New York My Commission Expires: c LtNA M.'LANDERS Notery Public, State of New Yoh Registration No.i908015 Quahfied fn jefferson Cool nq Conrrnission Expires October 13, 19 r
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| ENCLOSE'IDENTIFICATION OF CHANGES, REASONS Al'6)BASES FOR N)PC-QATR-1 (UFSAR APPENDIX B)
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| ENCLOSURE A IDENTIFICATION OF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT I UFSAR APPENDIX B)';:;":;-'.UFSARIA'ppe'ndh't::,B:,,:.".
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| :';:.,',"::.;';;'::;::;:'Page'/SeetIon':.,-'':':,'',::::,::",,::.::-:,:l
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| =,Si'ih:for,"Con'dttdlnj'.that,."the':Revised;Prig''am::-",-"',".',Gi'ntin'iies to'Stttis~fy)10CFRSO;'Appendh't'8''iiid",::,+3","";::,Contitihn'en'::PreyIously,'.'.A'p'pr'o'y'ed;by":
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| the':NRC::.'::',:";.', Page B.0-1, third paragraph Changed"Executive Vice President Nuclear" to"Chief Nuclear Of5cer" Reorganization.
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| Reorganization approved by the NRC via Unit 1 License Amendment 157 and Unit 2 License Amendment 71, dated Febnuuy 20, 1996.Page B.1-1, Section B.l.1, first paragraph Replaced"contractors and consultants" with"suppliers" Editorial.
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| NMPC uses the term"suppliers" as a synonym of"contractors" and"consultants," and prefers the term"suppliers." The use of an all~compassing term (i.e., using"suppliers" to include or describe contractors, consultants, or vendors)does not affect compliance with 10CFR50 Appendix B.Page B.1-1, Section B.l.1, second paragraph Changed"Each organizational department, including Nuclear Generation, Nuclear Engineering, and Nuclear Safety Assessment and Support (NSAS), is responsible for the quality of its own work." to read"Each organizational department is responsible for the quality of its own work." Editorial.
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| Editorial.
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| N/A Page B.1-1, Section B.1.2.1 a.Changed"is delegated by the President to corporate offlcers, as described herein" to read"is delegated by the President to corporate oKcers and the Manager Quality Assurance, as described herein" EditoriaL To reflect that the authority and responsibility of the Manager Quality Assurance is also described.
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| a.Editorial.
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| N/A b.Changed"Figure 13.1-1a" to read"Figure 13.1-1" b.Editorial.
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| b.Editorial.
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| N/A
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| "',:::%.""UFSAR''Appendix':B;:.:,,';;:
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| '':,: '":,;"-",".;;:
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| Page/Section';:"';,':.;"::;"-
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| Page B.1-2, Section B.1.2.1.1, first paragraph a.Changed"Executive Vice President Nuclear" to"Chief Nuclear Officer" a.Reorganization.
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| a.Reorganization approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Antendment 71, dated Febnuuy 20, 1996.Changed"...including all functions performed by Nuclear Generation, Nuclear Engineering, Nuclear Safety Assessment and Support, Nuclear Controller,..." to read"...including the Plant Generation and Engineering Functions under the Vice President and General Manager-Nuclear, Nuclear Safety Assessment and Support (NSAS), Business Management,..." b.Reorganization.
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| Same as Item a.Page B.1-2, Section B.1.2.1.1, second paragraph Changed"Controller Nuclear Division" to"General Manager Business Management" Reorganization.
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| Title change.Position title change to reflect management of business computers and firuume/accounting activities under the position of General Manager Business Management.
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| This is an administrative management position which does not perform QA related activities.
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| ".-;;:",.;UFSAR'"'Appendix 8"':i,"".;:".':
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| '-,".-"",:,:.'-::.,:,:PagelSectlon<'-',-'.-.',:l,-,:,l
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| '..',;-';:::",",$
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| '"',j'-"IderitIficatlon'oE;Cliaiige::.''
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| ""i'."i!,;.'::;-':.':::-;::;,"",:.'.;:!'Reason'for,,Change':".'.::;i,"-:::--''"'-:."'::.'-::"'.Bi'sls, for,'.,Co'n'eluding that;thee':Revtsed,&iograiiii::.-':::;:>,, NConIlriues'hi,Satisfy.:10CFR50!Appendix B;and'";i.-;.:.,Cotniiiitd1ents Previously',Approved by'.the NRCg~:;Page B.1-2, Section B.1.2.1.1, Item 1 Changed"The Vice President Nuclear Generation reports to the Executive Vice President Nuclear, and is responsible for safe and efficient operation, maintenance, and modification of the Station in compliance with Station licenses, applicable regulations, and the QA Program.The Vice President Nuclear Generation delegates to the Plant Managers and other appropriate personnel authority for perfonnance in accordance with the QA Program.See Table B-1 for QA Program element responsibilities.
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| Activities performed under the responsibility of the Vice President Nuclear Generation include:" to read"The Vice President and General Manager-Nuclear reports to the Chief Nuclear Officer, and has the overaH divisional responsibility for plant operation and engineering.
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| The Vice President Nuclear Engineering, Plant Managers, and the General Supervisor Labor Relations report directly to this Vice President.
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| See Table B-1 for QA Program element responsibilities.
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| Activities performed under the responsibility of the Vice President and General Manager-Nuclear include:" Corporate management reorganization.
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| Reorganization approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Amendment 71, dated February 20, 1996.
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| l i";":::::P UFSAR:'Appendh B.;:.;:$:;,"<'P<!'':;.';:;::Ide'i'itHication'ofChang'e,"-,<:,.",::;"::::.,".,::'.,:",:,,,;':
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| IBash for".,'Co'ndudliig that.the'.Revfsed,Program"'.
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| ",::".;''.:
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| ;Contliities'to,'Sathfy,';.jOCFR50;Appendht.8:aiilq.';,'':
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| ~Comntitm'ents',PrevIon'slyApproved,by',the'NRC!'-:::.v.,'age B.1-3, Item 2 Changed Item 2 to read: "Responsibilities and duties of the Vice President Nuclear Engineering and the Nuclear Engineering organization are described in Unit 1 UFSAR Section XIH.A.1 and Unit 2 USAR Section 13.1.1.See Table B-1 for QA Program element responsibilities." Corporate management reorganization.
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| Reorganization approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Amendmettt 71, dated Febrmuy 20, 1996.Page B.1-3, Item 3 Changed Item 3 to read: "The Vice President Nuclear Safety Assessment and Support reports to the Chief Nuclear Officer and is responsible for Quality Assurance, Licensing, Training/Emergency Preparedness, Security, and the Unit 2 Independent Safety Engineering Choup gSEO).See Table B-1 for QA Program element responsibilities.
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| Nuclear Safety Assessment and Support responsibilities are described in Unit 1 UFSAR Section XIH and Unit 2 USAR Section 13." Corporate management reorganization.
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| Reorganization approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Amendment 71, dated February 20, 1996.Page B.1-3, Item 4, first paragraph Changed"Executive Vice President Nuclear" to read"Chief Nuclear Officer" Corporate management reorganization.
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| Reorganization approved by NRC via Unit 1 License Amadment 157 and Unit 2 License Augment 71, dated February 20, 1996.
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| ,I
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| ',:'~'~UFSAR'!Appendli,B.~,'-";,;:;
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| =.j:",:~+:Page/Sectlon.:'..;,::;::.':;;,;'-:,',-, i,"::;;.:,';;";':;'.".,::,IdetItilici
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| ;:.Siishi'forConcluding tImt'thi"Revtse'd Piogrt'un'.'s,":...'.;:Con/biue's,,tYi:Sibsfy,'.10CFR50'Appendix B;ttnd"':g>:;:-
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| 'CotttrnItiiients'PievIottsly:Approv'ed by'the'.NRC:,"'-'-':
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| Page B.1-3, Item 4, second paragraph a.Changed"Tasks performed to fulfill these responsibilities include" to read"Tasks performed to fulfill these responsibilities are delineated in site procedures and include" Editorial.
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| a.Editorial.
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| N/A Combined Inspections and NDE Examinations as one task and removed the following identified tasks:~Coordinating and Reporting Internal and External QA Assessments
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| ~Operations Experience A88essnlent
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| ~Administering the Evaluation and Corrective Action Program for Deviation Event Reports (DERs)~DER Trend Analysis~Preparing and Processing QA Organization Documents b.Editorial.
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| Many of these tasks are also described under the responsibilities of supervisors or in other sections of the QATR.Administering the Evaluation and Corrective Action Program for Deviation/Event Reports (DERs)is the responsibility of the Plant Managers.b.Editorial.
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| N/A C, Added the following tasks:~Records Management
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| ~Document Control c.Reorganization.
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| c.Reorganization approved by NRC via letter dated July 13, 1995.
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| ';':"':":jUFSAR'.Appeii'dbr'8-:.'',:;::::
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| :,';:.-,'::;:,'''.",IderiIdicaIIon ofChange+~,;:;,:;:::~:,:.;-;,-.::, ,".,",';-'Reison foi",Cliaiigi.",;";."..''',:.","."..'':':.'';"',i:
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| ':, iBisls,t'or,'Coaduding'that the.Revtse'0 Piogiasnt"-'"..'."'.'.-
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| ~Coritliiiies.,'to,Siibsfy.';10CFR50'Appendix:B,and
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| ':;:;"::::;>
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| lCon'ti'nitntents Pk'vionsiy".Appr'oved bythe NRC.":,=;";
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| Page B.14, Item 4.b Changed"...determining applicability of industry and in-plant experience" to read"~..assessments determining applicability of industry and in-plant operating experience" Editorial clarification.
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| Incorporates the term"assessments" associated with operations experience assessments into supervisor responsibilities.
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| Editorial.
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| N/A Page B.IQ, Item 4.d Changed"...performing source surveillances of selected procurements" to read"...performing supplier evaluations and source surveillances of selected procurements" Editorial clarification.
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| Qualifies the type of activities that are done by the Procurement Quality Assurance Group.Editorial.
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| N/A Page B.14, Item 4.e Added Item 4.e to describe the responsibilities of the General Supervisor Quality Services Reorganization established new position of Supervisor Quality Services which was later changed to General Supervisor Quality Services.Reorganization approved by NRC via letter dated July 13, 1995.Title change is administrative in nature and does not affect position functions or responsibiTities.
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| Page B.2-2, Section B.2.2.3, first paragraph Changed"Executive Vice President Nuclear" to'Chief Nuclear OKcer" Reorganization title change.Reorganization reviewed and approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Amendment 71, dated February'20, 1996.
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| 'J I
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| ';':".<<;.:-',',":
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| '"',.Id 4Basb" for',,Coiidiidbig that the.'Revfsed,Piog'ram'!"-.;,,''",'--Continues'to Satbfy':,10CN50;Appen'dix'.B:arid:.'-',5$
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| ='Commitinents Pr'evIously,"Appi'oved by,.the NRC:.;-.Page B.2-5, Section B.2.2.15, Item 1 Changed wording from"The Manager Quality Assurance is responsible for reporting on the status, adequacy and effectiveness of the NMPC QA Program through the Nuclear Division Internal SALP Type Assessment Reports" to read"The Chief Nuclear Officer is responsible for reporting on the status, adequacy and effectiveness of the NMPC QA Program" Clarification and reorganization title change.Although the Manager Quality Assurance is responsible for reporting on the status, adequacy and effectiveness of the QA Program to the Chief Nuclear Officer, it is the Chief Nuclear Officer that reports to the President or Chief Executive Officer (CEO).The Chief Nuclear OKcer reports to the President of NMPC as described in Unit 1 License Amendment 157 and Unit 2 License Anieixlnient 71, dated February 20, 1996.Page B.2-5, Section B.2.2.15, Item 2 Changed"Executive Vice President Nuclear" to"Chief Nuclear Officer" Reorganization title change.Reorganization of corporate management approved by NRC via Unit 1 License Amendmerit 157 and Unit 2 License Amendmeiit 71, dated Febnuuy 20, 1996.Page B.24, Section B.2.2.16 Changed wording from"The SRAB is a standing committee chaired by the Vice President Nuclear Engineering and reports to the Executive Vice President Nuclear regarding designated QA functions at the Nine Mile Point Nuclear Station" to read'The SRAB is a standing committee reporting to the Chief Nuclear Officer regarding designated QA functions at the Nine Mile Point Nuclear Station" Clarification.
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| To more clearly reflect Plant Administrative Technical Specifications.
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| The change more clearly reflects Plant Technical Specifications, Administrative Controls Section, and also reflects corporate management position title changes associated with Unit I License Amendment 157 and Unit 2 License Amendment 71.Page B.24, Section B.2.2.17 Changed wording from"The SORC is an independent review committee responsible to the Vice President Nuclear Generation and transmits reports to the SRAB" to read"The SORC is an independent review committee responsible to the Plant Managers and transmits reports to the SRAB" Clarification.
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| To more clearly reflect Plant Administrative Technical Specifications.
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| The change more clearly reflects Plant Technical Specifications, Administrative Controls Section, and also reflects corporate management position title changes associated with Unit 1 License Amendment 157 and Unit 2 License Amendnient
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| : 71.
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| ''":;,j,:UFSAR;:Appe'adh'r",B::9'i::.".:
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| I-'.,>,::::.-,'';~,,',,",'.:.:.:'.,Ideiitific>>aIIon'of,Change;,',,~;-'.,"".,s'..''",',':.:
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| .,":,::i,;:,;'.:;:'-;,'..:':='."';.';;.Reasori;for,,:Changers,-;..-'":,';;;:-::>"::
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| -,-.".',Cori(i'ues:,to',Satisfy,'10CPR50'Appeadh;B.'iiil"".4i>
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| >Commitments Prevloiisly",.":A'jipx'oved by,': the',NRC-">>." Pago B.24, Section B.2.2.18 Changed wording from"...and actions are verified by Q1P personnel prior to closeout" to read"...and the actions are verified prior to closeout" Clarification.
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| While Q1P personnel verify the overall closure of all items, other groups may be used to do some of the actual technical verifications for completeness.
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| The overall indepeadence and confidentiality of Q1P have not changed.The technical abiTity of other departmeats is used to review somo of tho concerns.Page B.4-2, Section B.4.2.7 Pago B.5-1, Section B.5.2.6 Chaaged wording from"NQA or Procurement personnel other than the person who generated the procurement dociiment, but cpalifiod in QA,..." to red"Personnel other than the person who generated the procurement document, but with adecpate uaderstaiding of the recpirements and intent of the procurement documents,..." Added Section B.5.2.6 to descnbo procedure review process Clarification.
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| As an alteraative to performing procedure reviews no less frequently than every two years to determine if changes are necessary or desirable (ANS-3.2).
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| Niagara Mohawk has programmatic controls in place to continually identify procedure revisions which may bo needed to ensure procedures are appropriate for the circumstances aad are maintained current e There is no specific repireaiont for any particular group to perform these reviews, only that the individuals doing the review adecpatoly understand the mpirements and intent of tho procurement documeats.
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| This is in accordance with NQA-l, 4S-1, section 3, which is our stated program for meeting 10CFR50 Appendix B.This does not constitute a reduction of commitment since whoever does the review fuaction is required to be cpalified.
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| This qualiT!cation is accomplished through training.NRC approval per 10CFR50.54 granted via letter dated January 30, 1996.Page B.7-1, Section B.7.2.2 Changed wording from"When contractors perform work under their own QA programs..." to read"When suppliers perform work under their own QA pfogfaias Editorial.
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| NMPC uses the term"suppliers" as a synonym of"contractors," aad prefers tbe term"suppliers." Tho use of an all~passing tenn (i.o., using"suppliers" to include or descnbe contractors, consultants, or vendors)does not affect compliance with 10CFR50 Appendix B.
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| I (
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| UFSAR.:"Appiiidlx,B,"'.";'"-,'..
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| ~+~';':-',~IdeiitifiaitIo'ii of,Chaiige'';;:,":;;'.-".:,;;.P,,;.,Contlniies,4i'SatlsfyiiOCH60,::Appendix'8
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| 'an'6=;,';.'"'~j (Coiniiiitinents Pit.vlo'iisly'!hpproved,by,that.
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| NRC:;.'-:,:.'age B.7-1, Section B.7.2.3, Item 1 a.Changed wording from"...result in the supplier being placed on the QualiTied Contractor List Database (QCLD)as a qualified vendor" to"...result in the supplier being placed on the Qualified Supplier List Database (QSLD)" a.Editorial clarification.
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| To reflect use of the term"supplier" rather than"contractor." a.Editorial.
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| N/A b.Changed wording from"...by virtue of this ability" to read"...by virtue of their ability" b.Editorial.
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| b.EditoriaL N/A Changed wording from"...characteristics identified by Nuclear Engineering and NQA" to read"...Characteristics identified by Nuclear Engineering" c.Clarification.
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| To fully reflect Nuclear Engineering responsibilities.
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| c Nuclear Engineering is responsible for maintaining the design basis of systems, structures, and comporlents and tfallslates design requirements to suppliers which are deemed critical for a particular item/service.
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| The identification of critical manufacturing and functional processes and characteristics by Nuclear Engineering continues to satisfy 10CFR50 Appendix B, Criterion 7.Changed wording from"...methods have been identified and documented by which NQA will verify conformance to these iequlfelnents to read re@methods have been$Ient tfiled and documented which will verify conformance to these requirements" d.ClariTication.
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| d.NQA is involved with verification of supplier programs with attention to critical processes/characteristics selected, unless they can be verified onsite via test and/or inspection.
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| These responsibilities have not changed and, therefore, continue to satisfy 10CFRSO hppendix B.
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| ';"~~":.="-'::::<
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| IdentIfimtloti':ot,.Change/!''-"::.:-':;:.';';;:::::.::::,;;:,:.'.-'.::,'-.:."'%~Reasosn.for,Chatige,",':;:',;:,,'':-:-',::,'.,:::Bash for',Coiicliidlng that the':Revtsed,Prog'rain',-':;;~j," NConttriues.'to';Sahfy':10CFR50:App'endlx.:B:aciid'-.::""4':
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| -.Coi'iimitrneeriti':Prevto'tisly,::Approve'd by,'the NRC';-" Page B.7-2, Section B.7.2.3, Item 2 a.Changed wording from"NMPC-qualified suppliers involved in active procurement are surveyed every 3 yr to maintain..." to read"NMPC~ified suppliers involved in active procurement are surveyed every 3 years~to maintain..." The change from 3 yr to 3 years is editorial.
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| The addition of a note to reflect a tolerance of one quarter of a year is also editorial as this reflects Regulatory Guide 1.28, paragraph 3.2, as described in QATR Table B-3, sheet 1 of 8.a.Editorial.
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| N/A"<<With a tolerance of one quarter of a year" b.Changed wording from"Supplier 3-yr surveys..." to read"Supplier 3-year surveys" b.Editorial.
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| b.Editorial.
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| N/A Page B.7-2, Section B.7.2.3, Item 3 Added Item 3 to identify suppliers/organizations that are not required to be evaluated or listed on the Qualified Supplier List Database (QSLD).Clarification.
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| These statements clarify the use of the National Institute of Shuxlards and Technology and other NRC licensed utilities that meet the requirements of 10CFRSO Appendix B.Page B.7-3, Section B.7.2.6 Changed wording from"...purchased in accordance with Nuclear Engineering Procedures that provide.o to read"purchased in accordance with procedures that provide..." Clarification.
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| These controls are not limited to Nuclear Engineering paicedures.
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| Several types of procedures are used to make sure that design criteria is included in purchase requirements.
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| Although Engineering procedures provide controls to assure that items satisfy design requirements, these controls may also be found in Nuclear Interface Procedures or department procedures other than Engineering.'hese procedural controls continue to satisfy IOCFRSO Appendix B Criterion 7.Page B.9-2, Section B.9.2.9 Changed wording from'...kept by vendors and/or forwarded to NMPC" to read"...kept by suppliers and/or forwarded to NMPC" EditoriaL NMPC uses the term"suppliers" as a synonym of"vendors," and prefers the term'suppliers." The use of an all~conipassing term (i.e., using'suppliers" to include or describe contractors, consultants, or vendors)does not affect compliance with 10CFR50 Appendix B.10
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| | |
| ;-:;;.~".';UFSAR::Appendix 8;-g;.;g(~:A:,:;."::".-:."Identdici'tlon,of,:Cha'nge=;..:::;::";-.;.'".'-::c.-:.':j';::-
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| ;".Bash for, Concluding that, the Reytsed,Prograni:"".:,'-',:
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| ;Coritliiites.
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| 4Sash y,;,10CFR50':Appendh;B aril;.';.':;..-,'',;.ComnIitmeit9 Prevlou'sly':Appr'oved.by.Ihe NRC<"-"-, Page B.10-1, Section B.10.1 Replaced policy statement with wording from NQA-1 Editorial clarification to reflect wording provided in NQA-1.The restructuring of the paragraph to reflect NQA-1 is consistent with 10CFR50 Appendix B, Criterion 10.All areas continue to be reviewed except for the deletion of witness points.Witness points have either been upgraded to hold points or deleted because they were not needed.Page B.10-1, Section B.10.2.2, Item 4 Changed from"Hold points and/or witness points" to read"Hold points" Witness points are no longer used at Nine Mile Point.All witness points have been converted to hold points or deleted from NMPC procedures.
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| The removal of witness points continues to satisfy Appendix B, Criterion 10, since those witness points that were required have been upgraded to hold liits o Page B.10-2, Section B.10.2.5 B.10.2,6 B.10.2.7 B.10.2.8 Deleted previous Section B.10.2.5, which stated"Witness points require sufficient notification of the specifying organuation prior to performance of the specified activity" and renumbered reinaining sections accordingly.
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| Witness points are no longer used at Nine Mile Point.All witness points have been converted to hold points or deleted from NMPC procedures.
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| The removal of witness points continues to satisfy Appendix B, Criterion 10, since those witness points that were required have been upgraded to hold points'age B.10-3, Section B.10.2.9 Changed wording from"A program for inspection and surveillance of activities affecting fire protection is established..." to read"A program for inspection and surveillance, as required, for activities affecting fire protection is established..." Editorial clarification for ease of reading aud sentence structure.
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| Editorial.
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| N/A Page B.11-2, Section B.11.2.3, Item 5 Page B.18-1, Section B.18.1 Changed wonling from"Any witness and hold points" to read"Any hold points" Changed wonling from"...including those elements of the program implemented by suppliers and contractors" to read'including those elements of the program implemented by suppliers" Witness points are no longer used at Nine Mile Point.All witness points have been converted to hold points or deleted from NMPC procedures.
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| Editorial.
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| NMPC uses the term'suppliers" as a synonym of"contractors," and prefers the tenn'suppliers." The removal of witness points continues to satisfy Appendix B, Criterion 10, since those witness points that were required have been upgraded to hold polrlts.The use of an allmicompassing term (i.e., using"suppliers" to include or describe contractors, consultants, or vendors)does not affect compliance with 10CFR50 Appendix B.11
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| | |
| ,'"::i';:UFSAR A'p'peti'dlx.S,:;;".-.';.'-,.;
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| .,'.':;.'-:)-:.''-,""-PIdeiitificatlon of,Ch'ange';:-,;,,:j,'=,:;:;g,'.<<'::.~;."'~"',",:".;,";."., Rea'son for;Change;..":''j
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| ':'.;.':::.-',::@
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| IBasIs for,Coiicluding'that'the Revfsed.Prograii'i'...':::;;-Contlriue<<s'to Satisfy'10CFR50'App<<eindiIx'B'aiid'..';~";.~
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| >,'.Conimsitinents:Prevlot'isly"'Approv'ed,by, the NRC-"-,";Page B.18-1, Section B.18.2.3 Changed wording from"once every 2 yr" to read"once every 2 years" Editorial.
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| Editorial N/A Table B-l, sheet I of 2 a.b.Under Procedures column, identiTied Quality Assurance (QA), Nuclear Licensing (NL), and Nuclear Training (N'I)under the Vice President Nuclear Safety Assessment and Support (VP-NSAS), and identified Nuclear Engineering (NE)and Nuclear Generation (NG)under the Vice President and General Manager-Nuclear (VPGM-N).Removed Nuclear Procurement (NP)from the NSAS Procedures column to reflect transfer of the Nuclear Procurement function to Nuclear Engineering.
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| Identified Nuclear Engineering as responsible for QA Program elements associated with Criterion IV, accordingly.
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| Editorial, to reflect corporate management restmcturing.
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| b.Reorganization.
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| a.Editorial.
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| N/A b.The Nuclear Procurement organization was transferred from Nuclear Safety Assessment and Support (NSAS)to Nuclear Engineering.
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| The duties, functions, and responsibilities of Nuclear Procurement have not been altered.c, Removed Technical Services (TS)and Information Management (IM)from the NSAS Procedures cotunm.co Reorganization.
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| c The duties, responsibilities, and functions performed by Technical Services (TS)and Information Management (IM)have been reassigned to other branches, as appropriate.
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| The QA Program elements once implemented by TS and IM have been integrated into the appropriate branch and are identified on the res nsibili matrix.12 l,
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| +:,",,"..,'"'UFSAR'Ap'pend@
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| 8'',:"..;;";
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| "'.Bi'sh,'fot'.
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| Con'du'din'g'that,'the".Revts'ed Pi'o'g'ram".,",;:.',:,",'~
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| ',.::,.ContlntIes,to.Satisfy',:jKFR50,''Appendfix 8;and:,,",:.'"":;;.';
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| jCommttmerits',Prevlou'sty'."Appr'oved by,''thi".'RC:;""k Table B.l, sheet 1 of 2 (cont'd.)Identified Quality Assurance responsibiTity for QA Program elements associated with Criteria VI (Document Control)to reflect transfer of responsibility from Nuclear Engineering.
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| d.Reorganization.
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| d.Reorganization approved by NRC via letter dated July 13, 1995.Table B-l, sheet 2 of 2 Removed Nuclear Procurement (NP), Technical Services (TS), and Information Management (IM), from under NSAS Procedures column and from listing of NMPC organizations.
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| a.Reorganization.
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| a.The function of Nuclear Procurement was transferred fmm the Nuclear Safety Assessment and Support organization to Nuclear Engineering.
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| This transfer does not affect duties or functional responsibilities and, therefore, continues to satisfy 10CFR50 Appendix B criteria.The duties, responsibilities and functions of Technical Services and Information Management have been transferred to other organizations as appropriate.
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| b.Identified Quality Assurance responsibility for QA Program elements associated with Criteria XVH (Quality Assurance Records)to reflect transfer of responsibility from Nuclear Engineering.
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| b.Reorganization.
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| b.Reorganization approved by NRC via letter dated July 13, 1995.Table B-3, sheet 2 of 8 Changed Document column row"d" from Para.4 to read"Section 4" EditoriaL Editorial.
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| N/A 13 1
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| Enclosure B to NMP1L 1090 NINE MILE POINT-UNIT 1 SAFETY EVALUATION
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| | |
| ==SUMMARY==
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| REPORT 1996 Docket No.50-220 License No.DPR-63
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| Safety Evaluation Summary Report Page 1 of 60 Safety Evaluation No.: 89-019 Rev.0, 1, 2 8c 3 Implementation Document No.: USAR Affected Pages: System: Title of Change: Mod.N1-89-174 XI-14 Low-Pressure Reactor Feedwater System Low-Pressure Reactor Feedwater System Design Pressure Reduction
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| == Description:==
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| The pipe wall thickness originally specified for the low-pressure reactor feedwater system was very marginal for a design pressure of 600 psig.If the standard manufacturing tolerance and a reasonable corrosion allowance, were considered, the system piping was not adequate for 600 psig.For 16" pipe, the available corrosion allowance for the 40-year plant life is only.013", assuming the pipe wall thickness was supplied at the minimum of the manufacturer's tolerance.
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| The original plant specifications required a corrosion allowance of.088".In order to provide additional margin for corrosion over the remaining plant life, the design pressure was reduced from 600 psig to 530 psig.This increased the available corrosion allowance from.013" to.050".Safety Evaluation Summary: This modification will reduce the original design pressure of 600 psig to 530 psig.As a result of this change, pressure safety relief valves installed on feedwater pump suction piping and on the feedwater side of the feedwater drain coolers must be reset to the corresponding design pressure.Resetting the relief valves to the corresponding design pressure will not affect system operation because the valves provide negligible pressure relief during normal operation of the equipment/piping each valve protects.To demonstrate that the relief valves provide negligible pressure relief to the equipment/piping they protect during normal operation, the field-corrected pump curves were reviewed.A review of the pump curves indicates that during minimum flow conditions, a reduction of 150 gpm of flow due to all six relief valves lifting corresponds to approximately zero pressure reduction.
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| The purpose of the valves is to prevent excessive pressures in the system when the section of the system becomes isolated by valves and that section may be subjected to unexpected sources of heat.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 2 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 91-019 Mod.N1-88-052 N/A Spent Fuel Pool Installation of Poison High Density Spent Fuel Racks in the South Half of the Spent Fuel Pool Description of Change: Modification No.N1-82-013 and Safety Evaluation 84-003 Rev.1 encompassed the analyses design and installation of eight poison spent fuel racks, or 1710 storage locations, in the south half of the spent fuel pool for a'total installed capacity of 2776 locations.
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| This was approved by the NRC in Amendment 54 to the NMP1 Operating License DPR-63.The installation was planned in phases as described in the above safety evaluation.
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| Six of the eight racks were installed prior to this planned modification.
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| /This modification installed the seventh (216 spaces)storage rack in the southwest corner of the pool.The eighth rack (198 spaces)will be held in stores as contingency storage.The southwest work platform was permanently removed and replaced by a temporary seismic strut restraint.
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| The removal of the work platform uncovered the pool liner which was deformed from a prior installation.
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| The deformation (or bubble)was mapped and located to determine the extent to which one of four rack pedestal supports had to be modified so as to avoid the deformation.
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| Following exact information obtained from the mapping, the pedestal was reanalyzed and modified accordingly.
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| Following the pedestal modification, the temporary seismic restraint was removed and the rack installed along with its restraints and seismic beam.Safety Evaluation Summary: A seismic event occurring during the short period in which the racks are unrestrained is not considered a credible accident condition.
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| This condition was Safety Evaluation Summary Report Page 3 of 60 Safety Evaluation No.: Safety Evaluation Summary: 91-019 (cont'd.)(cont'd.)previously described in the June 1983 submittal to the NRC and was subsequently approved in Amendment 54.Heavy loads will not be handled over spent fuel.during reracking.
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| Prior criticality, thermal-hydraulics, and pool structure analyses will not require reanalyses as a result of this modification.
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| The rack mechanical analysis has-been revised to account for the pedestal modification and will be reviewed following the pool liner mapping.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 4 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 91-033 Rev.2 5.3 Gale.4.16KVAC-DG-ES, Rev.2 IX-18, Figure IX-6 Emergency Diesel Generator Emergency Diesel Generator Essential.Loading and Load Management Description of Change: This safety evaluation analyzed a change to the Unit 1 UFSAR that resulted from reconstitution of the diesel generator design basis loading analysis.Information contained in Section IX of the UFSAR pertaining to diesel generator ratings and accident loading was changed to show more complete and accurate information.
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| Safety Evaluation'ummary:
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| Maximum expected diesel generator load has been determined by calculation, compared to equipment capability, and found to be acceptable.
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| Informa'tion previously provided in the UFSAR shows that maximum diesel generator load is within the machine's capability.
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| The diesel generator load analysis has defined the maximum allowable design load based on vendor information and regulatory guidance documents.
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| The proposed UFSAR changes are consistent with the analysis and demonstrate that the diesel generators will be operated within their rating.With appropriate manual actions, total load can be maintained within the maximum allowable design basis load limit.Manual actions to be reflected in UFSAR Figure IX-6 agree with current operating procedures and include shutting down a core spray topping pump, control rod drive pump and a containment spray pump under appropriate plant conditions.
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| The impact on affected systems has been analyzed and it is concluded that these systems can continue to perform their intended functions as described in the UFSAR.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 5 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 93-056 Rev.1, 2&3 Simple Design Change SC2-0328-92 Figure III-1~-Title of Change: Construct a Spare Transformer Facility Description of Change: The spare transformer facility was constructed southwest of the Unit 2 345-kV switchyard.
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| This facility will be used for the storage of the additional spare transformer for Unit 2.Safety Evaluation Summary: The construction of the spare transformer facility does not impact the pertinent licensing issues that are associated with hydrological engineering; i.e., flooding, local intense precipitation (probable maximum precipitation), and the impact on the air intake accident X/0 (Chi/0), the atmospheric dispersion coefficient.i Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 6 of 60 Safety Evaluation No.: Implementation Document No.: 94-004 Simple Design Change SC1-0281-91/
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| SC1-01 58-93 UFSAR Affected Pages: Dwg.B-40143-C Overlay 1-3 System: CO2 Fire Protection, Fire Detection and Protection (FDP)Title of Change: Retire Styrene Fire Detection Zones D-8013, D-6053VP, and DA-6063HD Description of Change: This change retired in place three fire detection zones.Zone D-8013 is located in the Styrene (Binder)Pump House.The pump house is its own'building located north of the Waste Building.The flammable substance styrene has been removed and no other flammables exist at that location.The pumps are no longer operational and the building is designed explosion-proof.
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| Zone D-6053VP is used for styrene vapor detection located in the Pump House and the DOW System (Radioactive Waste Solidification) mixing area of the Waste Building.With styrene removed, these are not necessary.
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| Zone DA-6063HD is located in the mixing hood of the DOW mixing area, rendering it unnecessary.
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| The DOW System has been retired for years, making these selected retirements possible.Safety Evaluation Summary: This change will remove nuisance alarms presently on the system.With the alarms and panel engraving removed, this will aid operations.
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| This change will have no impact on the safe operation or shutdown of the plant.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 7 of 60 Safety Evaluation No.: Implementation Document No.: 94-007 Simple Design Change SC1-0156-92, SC1-0157-92, SC1-0158-92 UFSAR Affected Pages: System: Title of Change: Table Vl-1, Figure Vl-22;X-16 Containment Cutting and Capping of Unused Containment Piping Description of Change: The head spray piping, reference vessel leak rate piping, and the electrochemical piping are abandoned systems that penetrate the drywell.Since this unused piping contains valves or blind flanges, Appendix J Type B oriC testing is required.These simple design changes cut and capped unused piping and restored the drywell penetrations to a spare status.Safety Evaluation Summary:/The cutting and capping of unused drywell piping along with removal of valves or flanged ends will restore the drywell penetrations to a spare status.Capping near the penetration and elimination of isolation valves will eliminate Appendix J Type B or C testing.Potential leak paths through valves or flanges will be eliminated.
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| The capped lines will be considered spares and will be subject to Appendix J Type A testing to ensure overall containment integrity.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 8 of 60 Safety Evaluation No.: implementation Document No.: UFSAR Affected Pages: System: 94-008 Temporary Mod.94-019 X-34 Service Air Title of Change: Installation of a Portable House Service Air Compressor Description of Change: This temporary modification installed a portable air compressor at valve HSA-113 located at the Turbine Building elevation 300'.This temporary modification will prevent depressurization of the service air system during maintenance of compressor 95-01.The intertie between service air and instrument air will be closed.This will provide assurance of no flow of air to the instrument air system while maintenance is being performed on service air compressor 95-01, to prevent potential oil contamination of instrument air.Safety Evaluation Summary: The service air system does not have a backup air supply.If the service air system drops below 20 psig for the plant preaction fire sprinkler zones and 50 psig for the drypipe systems, the respective alarm will come in and the drypipe fire sprinkler zones will fill.The portable air compressor will prevent this from occurring.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 9 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-013 Mod.N1-93-018 Figure IX-2 24 VDC System Replace 24 VDC Battery Chargers This modification replaced the existing 24 VDC battery chargers with new battery charging units of modern design.This modification also provided a disconnect switch to isolate the 48 VDC, center-tapped, grounded neutral battery from the battery chargers and its connected loads.Safety Evaluation Summary: The failure of 24 VDC battery chargers is not an initiating event for any design basis accident.The replacement of the existing 24 VDC battery chargers with similar equipment is functionally a one-for-one equipment substitution with equivalent electrical characteristics and greater reliability.
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| Consequently, this will not increase the probability or consequences of an accident.No new malfunction or failure mode has been created that could cause a new unanalyzed event.System reliability, system characteristics, equipment qualifications, and compliance with fire protection and Appendix R requirements are unchanged or are improved by this modification.
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| Installation will not result in a reduction of the plant safety margin.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 10 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 94-019 Rev.0 L 1 IVlod.N1-92-005 10A-96, 10A-114, 10A-118 4160 VAC, 600 VAC and 125 VDC Systems Title of Change: Description of Change: Improve Electrical Coordination The emergency diesel generator 51-V relay and its miscoordination with downstream protective devices was resolved by defeating the tripping function of the 51-V relay when an automatic core spray injection signal is initiated.
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| /An alarm circuit was added so that actuation of the 51-V relay annunciates in the control room.A common annunciator window is used for 51-V relays located at power boards (PB)102 and 103;however, a separate computer point is provided for each 51-V relay.i For circuits associated with PB 16 and 17, electrical protective device coordination was improved as follows: 1.For PB16B/PB17B main supply breakers, the long time settings for existing electromechanical trip devices were revised.2.For the feeder breaker to lighting voltage regulator 16, the existing electromechanical trip devices were replaced with new electromechanical
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| 'rip devices.3.In PB161B/PB171B, one motor case circuit breaker (MCCB)was replaced with a new MCCB.4, In PB16B/PB17B, for the feeder breakers to PB167, the existing electromechanical trip devices were replaced with new electromechanical trip devices.5.In PB167, two MCCBs were replaced with two new MCCBs.
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| Safety Evaluation Summary Report Page 11 of 60 Safety Evaluation No.: Description of Change: 94-019 (cont'd.)(cont'd.)6.In PB16B/PB17B, for the feeder breakers to PB1671A/PB1671C, the long time and short time (PB17B only)settings for the existing electromechanical trip devices were revised.7.In PB1671A/PB'1671C, one existing MCCB was replaced with a new MCCB.For circuits associated with battery boards (BB)11 and 12, electrical protective device coordination was improved as follows: 1.In BB11/12 battery supply breakers, the existing electromechanical trip devices were replaced with new electromechanical trip devices.2.In BB11/12, replaced selected existing MCCBs with fuses and fuse blocks.3.For fuses located at or downstream of BB11/BB12, replaced selected existing fuses with new fuses.For some circuits, fuse block replacement was required./Safety Evaluation Summary: The replacement of electrical protective devices with similar equipment is schematically a one-for-one substitution of components with functionally equivalent characteristics.
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| The modification of the emergency diesel generator's output circuit breaker's tripping scheme will improve the availability of emergency power when it is most needed by reducing the possibility of a spurious breaker trip.This modification will also improve the availability of safety-related equipment when an electrical fault occurs.Consequently, this will not increase the probability or consequences of an accident.No new malfunction or failure mode has been created that could cause a new unanalyzed event.System reliability, system characteristics, equipment qualifications, and compliance with fire protection and Appendix R requirements are unchanged or are improved by this modification.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 12 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-037 Rev.1 Mod.N1-89-079 N/A Emergency Cooling System No.39 Emergency Cooling Appendix J Modification This modification cut out and replaced existing valves 39-03 and 39-04 and modified valves 39-05 and 39-06 with the reactor in the cold shutdown condition and single reactor pressure boundary isolation via safety-related manual gate valves 39-01 and 39-02./Safety Evaluation Summary: This modification effectively provides valve replacement and modification.
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| This will result in enhanced containment integrity and testability and does not increase the probability of occurrence or the consequences of an accident or malfunction to safety previously evaluated in the FSAR.The replacement and modification of these valves to enhance their ability to isolate the containment does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.It also does not reduce the margin of safety as defined in the basis for the Technical Specifications.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| gl Safety Evaluation Summary Report Page 13 of 60 Safety Evaluation No.: Implementation Document No.: 94-038 Rev.1 Simple Design Change SC1-0151-93 Mod.N1-93-023 UFSAR Affected Pages:~System: Title of Change: Description of Change: N/A Main Steam System 4'SIV Brake Installation and Poppet Upgrade This modification eliminated repeated leak rate test failures on main steam isolation valves (MSIV)to enhance their testability.
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| The modification consisted of installing vendor recommended MSIV poppet retrofit packages, and electric brakes and new motors with extended shafts for the operators on the motor-operated valves.Safety Evaluation Summary: This modification effectively provides changes to the MSIVs to eliminate leak rate test failures and enhance their testability.
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| The change has been designed to minimize stem susceptibility to cracking, provide improved anti-rotation mechanism, poppet self aligning features and additional hardfacing.
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| Additionally, installation of the new motors with brakes on the inboard MSIV motor operators will prevent excessive sliding of the valves during closure testing.This will result in enhanced containment integrity and testability.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 14 of 60 Safety Evaluation No.: Implementation Document No.: 94-039 Rev.0 Bc 1 Mod.N1-88-052, Phase IIA UFSAR Affected Pages: N/A System: Title of Change: Spent Fuel Pool Southwest Corner Rack Modifications and Addition of Rack Top Platform Description of Change: This change provided a repair solution to facilitate the installation of the spent fuel storage rack located in the southwest corner of the fuel pool prior to refueling outage RFO13.This rack was one of two remaining to be installed from the rerack campaign during 1984.A new work platform was installed on top of this rack.Safety Evaluation.
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| Summary: Revision 0 of this safety evaluation required that previously discharged fuel (i.e., fuel with low decay heat generation) be used when loading this spent fuel rack and did not allow for the loading of newly discharged fuel from RFO13.This restriction was established by NMPC to provide additional thermal margin with the rack top platform installed.
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| Revision 1 of this safety evaluation removes this restriction (allowing RFO13 newly discharged fuel to be loaded into the rack)and instead requires that the rack top platform not be installed for a minimum period of 180 days after the rack has had irradiated fuel transferred to it.The 180 days is bounded by the Unit 1 UFSAR Figure X-7, Decay Heat Generation vs.Days After Reactor Shutdown Curve, which indicates that 110 days following a full core discharge, decay heat generation is stabilized.
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| Since the addition of the 180-day waiting period prior to rack top platform installation will ensure that the intent of the original restriction is met (i.e., the rack top platform will not be installed over fuel with a high decay heat generation rate), this change is bounded by the original evaluation and has no impact on the spent fuel pool heat removal capability.
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| This modification installs one of the two remaining Boraflex design high-density spent fuel storage racks necessary to facilitate a full core discharge capability during our next refueling outage.This modification will have no impact on the safe operation or shutdown of the plant.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 15 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-051 Mod.N1-91-021 10A-118 Reactor Instrument Replace ACUREX This modification replaced the ACUREX Autodata Ten/5 dataloggers located in instrument and control cabinets 1S10 and 1S69, with new safety-related processing hardware located in instrument and control cabinets 1S16 and 1S17.It also replaced two existing recorders with one.Logging of torus temperature individual RTD values has been transferred to the process computer.RPS circuits to 1S16 and 1S17 have been changed for consistency of power sources.The Lo-Lo-Lo inputs to the fuel zone indication were removed and the normalization constant (k factor)was eliminated from the compensated water level calculation.
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| Safety Evaluation Summary: Nuclear safety is improved by this modification.
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| The RPS circuits are better balanced, thus improving their load characteristics.
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| The installation of more reliable electronic hardware improves the probability that the system will be available during an event where it is needed, and the information it provides will be more reliable.The removal of the normalization constant will provide more reliable indication throughout the entire instrument range.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 16 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-052 Mod.N1-87-042 VIII-11;10A-118 Plant Process Computer RIS Isolator Replacement This modification installed a safety-related state-of-the-art data acquisition system (DAS)to provide signal isolation for APRM/LPRM and feedwater system inputs to the plant process computer.I The modification replaced the existing Rochester Instrument isolators used to isolate safety-related signals input to the plant process computer.The isolators were replaced with an Input/Output (I/O)computer system and interfaced with the existing plant process computer system via a fiber optic cable.The fiber optic cable isolates all maximum credible faults from the safety-related inputs.The new.I/O system, cabinets, I/O cards and cables are safety related and seismically qualified.
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| Safety Evaluation Summary: The DAS loads will be appropriately isolated and protected to prevent malfunctions from impacting RPS.Also, this modification will not cause a change to any system interface in a way that would increase the likelihood of an accident.All safety-related equipment and materials will be procured and installed to applicable regulatory and industry codes and standards to ensure system integrity.
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| The new DAS performs the same design function as the currently installed isolators and is being procured and installed to equal to or higher design standards.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 17 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 94-053 Rev.0 5 1 Mod.N1-94-002 Figure Vill-4 Reactor Recirculation GEMAC Recirculation Pump Control Modification Description of Change: This modification removed from service five obsolete and unreliable GEMAC function generators (FGs)used in the reactor recirculation pump motor generator set control circuits.The FGs aid in linearizing the relationship between pump speed and reactor recirculation pump motor generator set output.Their signal processing function will be replaced by modifying the cams in each Bailey pneumatic positioner.
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| The proposed modification is defined in General Electric's proposal for the reactor recirculation control system enhancements.
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| Safety Evaluation Summary: The proposed design change is functionally equivalent to the present FG/Bailey actuator configuration.
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| The system response for each reactor recirculation control loop will remain linear;the Bailey positioner, with the A characterizer cam installed, will perform the required signal processing instead of the FGs.Elimination of the FGs from the control loops will provide additional system reliability and relief from calibrating, refurbishing or procuring replacement FGs.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 18 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 94-056 Rev.0, 1, 5 3 Mod.N1-88-153 Table Vl-3a Sh 2 Bc 3;Vll-5, Figure Vll-1;X-2, Figure X-1 System: Title of Change: Shutdown Cooling and Core Spray Systems Containment Isolation Valve (Appendix J)Modification
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| -Shutdown Cooling Water Seal Description of Change: This modification provided a qualified, 30-day water seal for the shutdown cooling system isolation valves utilizing a nominal flow of 22 gpm frorrl either loop 11 or 12 of the core spray system.This allows Unit 1 to achieve compliance with 10CFR50 Appendix J, as required by the NRC, without the replacement of isolation valves 38-01, 38-13, and 38-12.Safety Evaluation Summary:/This water seal will result in enhanced containment integrity.
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| The addition of check valves 38-165, 166, 167, 168, 169, 170, 171, and 172 provides isolation between the high-pressure reactor coolant:system and the low-pressure core spray system.The seal water system design up to the check valves meets the same design criteria as the reactor coolant system with respect to safety classification, temperature, and pressure.Leak testing for these check valves will be in accordance with Specification 3.2.7.1.Therefore, adequate assurance is provided such that the low-pressure core spray system will not be damaged by overpressurization and result in potential loss of integrity with subsequent release of radioactivity.
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| The acceptability of the design regarding single active failure under certain scenarios was found acceptable based on the PRA study and guidelines provided by NUREG-0800 and Generic Letter 88-20.It has been determined that no exemption is needed from the Appendix J compliance standpoint during this scenario based on guidance provided by NUREG-0800 and Generic Letter 88-20.The NRC has issued a Safety Evaluation Report (SER)on the Technical Specification Amendment to add the new water seal check valves to the Pressure Isolation Valve Table 3.2.7.1.This SER has reviewed the new design configuration for water-sealing the shutdown cooling isolation valves using the Safety Evaluation Summary Report Page 19 of 60 Safety Evaluation No.: Safety Evaluation Summary: 94-056 Rev.0, 1,&3 (cont'd.)(cont'd.)core spray water in order to meet 10CFR50 Appendix J criteria, including the two scenarios under which the single failure criteria is not being met.No significant adverse impact on shutdown cooling and core spray system performance will result as a consequence of this modification.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 20 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 94-057 Rev.1 Mod.N1-91-009 Figure Vll-3;10B-58;Table XV-4, Containment Spray Replace Operators on Containment Spray Intertie Valves EPN 80-40 and 80-45 Description of Change: This modification replaced the handwheels on valves 80-40 and 80-45 with pneumatic operators designed to fail-open.
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| The removal of the handwheels gives operators greater flexibility while the fail-open design ensures a water seal for 10CFR50 Appendix J.l Safety Evaluation'Summary:
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| The design function of intertie valves 80-40 and 80-45 is to be in the normally open position, ensuring interconnection of the primary and secondary loops of containment spray in order to provide a water seal in accordance with 10CFR50 Appendix J.The addition of pneumatic operators classified as safety-related active, and designed so that on a loss of motive power the air will be released, allowing the spring to open the valve, will not change the design function of the valves.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 21 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-060 Simple Design Change SC1-0087-93 N/A Service Water Emergency Service Water Supports This simple design change added new supports to the service water inlet and outlet piping to the Reactor Building closed loop cooling heat exchangers.
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| Existing supports were reworked and replaced.Safety Evaluation Summary: The new support arrangement will be in accordance with the original design basis criteria.AII piping stress will be less than the allowable specified in B31.1-55 and all supports will have allowable loads less than the manufacturer's allowable or less than AISC 8th Edition allowable.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 22 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-062 Rev.0 5.1 Simple Design Change SC1-0059-94 III-23;10B-69 N/A Foam Room Wall Replacement This simple design change removed the equipment supports attached to the Foam Room west and south precast concrete exterior wall panels and provided new supports.These wall panels were removed and replaced with a temporary enclosure until a masonry wall was constructed.
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| Welds were repaired to DC valve board¹11 to the existing embedded floor channels.An 8-inch reinforced block wall with a brick veneer was then constructed.
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| Safety Evaluation Summary: The design of the equipment supports, the temporary enclosure, the block wall barrier, and the masonry wall considers the requirements described in the Updated Final Safety Analysis Report (UFSARj.The design satisfies the licensing basis requirements described in the UFSAR.The weld repair of this valve board will satisfy the requirements of the SQUG Generic Implementation Procedure.
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| This change will have no impact on the safe operation or shutdown of the plant.Based on the evaluation performed, it is concluded that removal of the precast concrete panels and replacement with a reinforced concrete block wall does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 23 of 60 Safety Evaluation No.: Implementation Document No.: USAR Affected Pages: System: Title of Change: 94-068 Simple Design Change SC1-0062-94 10A-118 120 VAC Distribution
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| 'DC SC1-0062-94, Installation of Data Acquisition System
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| == Description:==
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| This modification installed a permanent Data Acquisition System (DAS)to replace the individual DAAS computers previously in use in the Auxiliary Control Room.The scope of this modification was limited to the installation of all computer equipment associated with the DAS, along with an asynchronous communications link to be utilized for data transfer to a Network File Service.This included the mounting of the computer cabinet, and providing a permanent power supply for the system MG set 167 computer panelboard.
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| A new cable was routed in the Auxiliary Control Room between MG set 167 computer panelboard through existing raceway and newly installed raceway./Safety Evaluation Summary: The function of the DAS, without inputs from process control instrumentation, is nonsafety related and does not.have an electrical interface with any safety-related systems or components; therefore, an electrical failure of the system will have no impact on plant safety.In addition, the system receives its power from MG set 167 computer panelboard, and since the loads associated with this panelboard are not required for safe plant shutdown, the DAS is also not required for plant shutdown.Structurally, the installation details conform to the requirements of a seismic Class I area.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 24 of 60 Safety Evaluation No.: Implementation Document No.: USAR Affected Pages: 94-069 Simple Design Change SC1-0039-94 Tables Vl-1, Vl-3b Sh 2;Vll-34, Figure Vll-13 System: Title of Change: Hydrogen/Oxygen Monitoring H,/O~Sample Line Reduction
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| == Description:==
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| This safety evaluation addresses the reduction in the number of sample points for the hydrogen/oxygen monitoring system.Specifically, this simple design change retired one sample line, stream A for H,/0, channel¹11 and two lines, streams A and C for H,/0, channel¹12.This was accomplished by cutting and welding a cap on the sample lines outside the drywell before the first isolation valve.The electrical connections were disconnected and the isolation valves and piping were retired in place.Control Room indication was blanked and deleted.This reduced the number of sample points from the drywell to two, one for each channel, and two from the torus, one for each channel.Since this safety evaluation, is generic with respect to the reduction in the number of sample points for H~/0isolation valve closure, as directed by OP-9, up to no less than one sample from the drywell and torus for each channel of H,/Owas evaluated and justified by this safety evaluation.
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| Safety Evaluation Summary: The H,/0, monitoring system is classified as safety related as a result of Regulatory Guide 1.97.However, the system is a passive monitoring system and is neither the initiator nor the contributor to any accidents evaluated in the UFSAR.The reduction of the number of sample points will not increase the probability of occurrence of any of these accidents.
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| Based on the results of numerous containment mixing studies, the response to the design basis event will be unchanged.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 25 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 94-070 Simple Design Change SC1-0070-94 10A-118 Reactor Recirculation Control System (RRCS)Title of Change: Installation of Cable&Terminations for the Data Acquisition System (DAS)Description of Change: The new data acquisition system (DAS)consists of a PC, video monitor, printer and an enclosure which will be located in the Auxiliary Control Room.The new DAS will provide information for trending and troubleshooting:that will increase both the reliability and capacity factor at Unit 1.This simple design change electrically connected the reactor recirculation control system (RRCS)to the DAS.In addition, several cables were run from various cabinets to the DAS in preparation for other installations.
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| These cables'were stored in the existing cable trays until final installation.
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| The scope of this change involved routing cable throughout the Auxiliary Control Room and terminating cables at the DAS and the RRCS.The installation required routing cable through existing cable trays and terminating these conductors at various points in the logic.The cable is multiple-conductor, twisted, shielded pair 016.The shields of each pair were grounded to prevent extraneous noise from entering the DAS.All terminations were made on terminal blocks with ring tongue terminals.
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| A safety-related signal isolator was installed to isolate the safety-related reactor recirculation total flow system from the nonsafety-related DAS.Safety Evaluation Summary: The logic of the RRCS will be unchanged by this modification.
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| The credible failures that this change can produce are enveloped in Section XV of the Updated Final Safety Analysis Report.Credible failures include the loss or sudden reduction of the control signal in one or more reactor recirculation loops due to a shorted lead in the control circuit.The risk of a short is minimal due to the DAS being a passive system that will monitor system parameters only.The control system will operate as designed with no changes in response or control.
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| Safety Evaluation Summary Report Page 26 of 60 Safety Evaluation No.: Safety Evaluation Summary: 94-070 (cont'd.)(cont'd.)In addition, the installation of the safety-related isolation amplifier in the total reactor recirculation flow circuit will have no effect on the circuit.The input resistance of the isolator is great enough so that it will not adversely affect flow input to the average power range monitors.The Class 1E isolator is designed to adequately isolate the safety-related signal from any faults or transients caused by the nonsafety-related DAS.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 27 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 94-072 Rev.0 8c 1 Mod.N1-90-041 Table Vl-3b Sh 2;Vll-2, Vll-3, Vll-4, Vll-7, Figures Vll-1, Vll-2;10A-108, 10A-109, 10A-110, 10A-112, 10A-113, 10A-114, 10A-118;10B-63;Table XV-9a System: Title of Change: Core Spray System Core Spray Minimum Flow Recirculation Lines/Throttling Description of Change: This modification installed separate minimum flow recirculation lines for each core/spray pump set.In addition, the inboard isolation valves were throttled to slowly inject core spray during anticipated transient without scram and small break loss-of-coolant accident events.To accomplish this, EOP jumpers were installed in the Control Room which inhibits the initiation and interlock signals for the inboard, outboard and test return valves.The test return valves are required to be opened to support extended recirculation flow and core spray pump set operation for the shutdown cooling water seal.Safety Evaluation Summary: The analysis section clearly demonstrates that the applicable design and licensing criteria have been satisfied.
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| This includes both normal design basis functions, manual EOP-directed operation, and shutdown cooling system water seal requirements.
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| The impact of separate minimum flow recirculation lines and inboard valve throttling on required core spray injection flows has been calculated.
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| Based on the calculation, the core spray system provides adequate flow to satisfy 10CFR50.46.
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| Redundancy, separation, Appendix R, seismic qualification and environmental qualification have been incorporated into the design and confirmed by analysis.This change does not increase the probability, consequences or create a different type of accident or malfunction of equipment important to safety.The margin of safety is not decreased.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 28 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages:-System: 94-074 DER 1-94-0224 VI-32 Containment Spray, Core Spray, Containment Spray Raw Water, Nitrogen Supply System¹12, Nitrogen Supply System¹11 Title of Change: Licensing Document Change Request 1-93-IST-006 to the NMP1 IST Program Plan and LDCR 1-94-UFS-056 to the UFSAR Description of Change: I/This safety evaluation evaluated additions to and deletions from the In-Service Testing (IST)Program Plan, based on the recent development and/or revisions of Safety Class determinations which necessitate changes to testing requirements.
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| Safety Evaluation Summary://It has been determined that overpressure of the containment due to malfunction of pressure control valves (PCV)in the nitrogen supply system is not a design basis safety concern as it would require the failure of 1)a PCV and 2)a failure open of an isolation valve.This would constitute two active failures of safety-related components and is beyond the design basis of the plant.Therefore, the relief valves do not need to be safety-related active and inclusion of the relief valves in the program is not required.Section VI-F.3 of the Updated Final Safety Analysis Report, Containment Ventilation System, contains language which is vague and requires clarification.
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| This section states that the relief valves and overpressure regulators in the nitrogen makeup supply line"were tested prior to initial startup and periodically thereafter for operability and setpoint." Note that this statement is in the past tense (it does not commit to continuous testing), while all other testing described in Section Vl-F and references to IST for other systems in Section Vll are in the present tense.The statement is to be revised prior to deleting the relief valves from the IST Program Plan.Further, the PCVs are not tested periodically at this time;PCVs are exempt from testing in accordance with ASME XI IWV-1200.Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 29 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: 94-075 Mod.N1-93-013 Table IX-1 Sh 1-3;X-18;10A-115 Containment Atmosphere Monitoring System, H~/0, Analyzer Title of Change: Description of Change: Replace H,/0, Monitoring System This modification replaced the hydrogen/oxygen (H,/O~)monitoring system, cabinets, and analyzers.
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| The previous system was produced by Beckman instruments and installed prior to Regulatory Guide 1.97 and NUREG-0737 requirements.
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| Due to age and the reclassification as safety related, the maintenance had become extensive.
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| A new Teledyne Analytical Instruments H,/0, monitoring system was installed in place of the Beckman.The cabinets are mounted in the same location, with the control cabinet mounted on the column across the aisle.The Teledyne system was procured safety related and is fully seismically and environmentally qualified.
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| /Safety Evaluation Summary: The analysis clearly demonstrates compliance with all applicable criteria including safety classification, seismic qualification, environmental qualification, power requirements and separation.
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| Calculations have been performed for the transportation time, response time and accuracy and assure compliance with design basis functions and licensing commitments.
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| Specifically, the requirements of 10CFR50 Appendix A, General, Design Criterion 41, Appendix B, Quality Assurance requirements, IEEE-344, Regulatory Guide 1.97, NUREG-0737, Fire Protection program and IST program have been satisfied.
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| This change does not increase the probability, consequences or create a different type of accident or malfunction of equipment important to safety.The margin of safety is not decreased.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 30 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 94-079 DER 1-93-0921 VI-21, VI-26, Table Vl-3b Sh 4 Traversing In-core Probe (TIP)TIP System Containment Isolation-LDCR This safety evaluation analyzed a method to more accurately represent the existing'esign of TIP system containment isolation features in the Unit 1 UFSAR.Valve configuration, isolation logic and valve motive power are addressed.
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| Wording was added to the UFSAR to describe the TIP ball and shear valves and how the ball valves are prevented from reopening after an isolation signal clears.I Safety Evaluation Summary: Applicable criteria have been reviewed to verify that the TIP containment isolation does comply with plant licensing and design basis requirements.
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| The ability of the TIP system to remain isolated following a containment isolation is assured by system design, normal operating configuration and procedural controls.In the unlikely event the TIP guide tubes fail to isolate, calculated accident doses are not significantly impacted and are well within 10CFR100 limits.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question..
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| Safety Evaluation Summary Report Page 31 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 94-080 Rev.0&1 Mod.N1-94-003 IV-27, IV-28, IV-29, Figure IV-9;XVI-5, XVI-12, XVI-21, XVI-122, Tables XVI-2 Sh 1 5 2, XVI-9a, Figures XVI-12a, XVI-12b System: Title of Change: Description of Change: Reactor Vessel Reactor Core Shroud Repair The shroud modification was designed to provide an alternative load path for all Type 304 stainless steel circumferential welds (welds H1-H7).The modification ensures the structural integrity of the core shroud by replacing the function of core shroud welds H1 through H7 with four stabilizer assemblies.
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| 'The stabilizer assemblies are comprised of four tie-rod assemblies and four'ore plate wedges.Safety Evaluation Summary: This evaluation has investigated the installation of core shroud stabilizer's at Unit 1.The evaluation of the shroud modification hardware included design, code, materials, fabrication, structural, systems, installation and inspection considerations.
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| The evaluation concluded that the proposed modification is in accordance with the BWR VIP Core Shroud Repair Design Criteria and the NRC Safety Evaluation Report (SER)on the BWR VIP Shroud Repair Criteria.The Unit 1 repair modification of the core shroud is to be performed as an alternative to ASME Section XI, as permitted by 10CFR50.55a(a)(3).
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| Consequently, NRC approval of this repair approach is required.This safety evaluation documents the NIVIPC review of the repair in accordance with the provisions of 10CFR50.59.
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| A separate safety evaluation (95-013)was perfoimed to evaluate the acceptability of the installation of the core shroud stabilizer assemblies prior to NRC approval.The evaluation concluded that there are no unreviewed safety questions associated with installing the repair prior to NRC approval, provided the reactor remains in the cold shutdown condition or the hot shutdown condition for the Safety Evaluation Summary Report Page 32 of 60 Safety Evaluation No.: 94-080 Rev.0&1 (cont'd.)Safety Evaluation Summary: (cont'd.)performance of noncritical hydro testing above 212'F and/or the performance of CRD scram time testing until NRC approval of the repair is obtained.Additionally, SE 95-007 was performed to review potential safety impacts of the installation activities associated with the repair.The evaluation concluded that there are no unreviewed safety questions relative to the repair installation activities.
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| The NRC issued a SER, dated March 31, 1995, entitled"Safety Evaluation of the Repair Proposal for the NMP1 Core Shroud." The NRC SER has reviewed all of the repair design aspects and has concluded the repair design is acceptable.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 33 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 95-002 Rev.0 L 1 Simple Design Change SC1-0067-94 Figure Vll-1 Core Spray Core Spray IV Pressure Binding Relief This modification added a pressure binding relief path to core spray valves 40-01, 40-09, 40-10, 40-11, and test return valves 40-05 and 40-06, to ensure the valves will open under all postulated conditions.
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| For the core spray valves, an instrument line was connected to the existing 1/4-inch tap for the lantern gland stuffing box and connected to the reactor side drain valves.The drain valves were locked open to provide the relief path to the reactor side process piping.The valve packing was modified to allow pressure in the bonnet to relieve through the 1/4-inch tap.For the test return valves, a small hole was drilled into the disk on the reactor side of the valve to provide the relief path.Also included in this modification, the drain valve configuration of valve 40-05 was revised to provide two valves and a threaded cap.This change was required to enable the attachment of DP measurement equipment across the valve during Generic Letter (GLI 89-10 dynamic flow testing.These drain valves are locked closed during normal plant operation.
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| Safety Evaluation Summary: The core spray system is required to operate to prevent overheating of the fuel following a postulated loss-of-coolant accident.The inside valves are required to open when reactor pressure is 365 psig or less and the test return valves will be required to open to facilitate throttling of the core spray flow to the reactor to maintain water level.The addition of a pressure relief path to these valves will prevent the potential for pressure locking in the redundant core spray loops, and the addition of double isolation drain valves to one valve will enable GL 89-10 dynamic testing.This modification will have no effect on accidents or malfunctions previously evaluated in the UFSAR.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 34 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-003 Simple Design Change SC1-0089-94 X-35, X-36, Figure X-9 Instrument Air System Instrument Air Upgrade-Valve 94-91 Logic Change Description of Change: This simple design change added a pressure switch contact in the control circuit of air-operated valve 94-91 so that if low pressure is sensed downstream of the'alve, the valve will close.This enhances availability of the safety-related portion of the instrument air system should the nonsafety-related piping downstream of valve 94-91 fail.Additionally, a pressure indicator was added to monitor the same process pressure as the pressure switch.Safety Evaluation Summary: The applicable criteria from the UFSAR, Chapter X, XV and 10A, have been satisfied by this simple design change.Single failure criteria has been addressed and safety class determination No.93-084 evaluated for impact.The components added are nonsafety related and not required to be redundant.
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| The added cable is in conduit and, therefore, the fire load in this area is not increased.
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| The valve will perform its safety-related function to fail close.The availability of the instrument air system is enhanced by this change.This change does not increase the probability, consequences or create a different type of accident or malfunction of equipment important to safety.The margin of safety is not decreased.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 35 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 95-007 Mod.N1-94-003 N/A RXVE Core Shroud Repair Installation The NRC issued Generic Letter 94-03 due to observed cracking in the core shrouds of several BWRs.This generic letter requires inspection of the shroud and/or repair, if necessary.
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| NMPC will either inspect and possibly perform a repair if the weld inspection necessitates it or perform a preemptive repair of the shroud during RFO13.The Unit 1 reactor core shroud repair is designed to structurally replace shroud welds H1 through H8.The installation of the entire repair involves electrical discharge machining (EDM)of the shroud support cone and shroud itself, which will generate very fine particles called swarf, the attachment of a trolley/buggy to the Refuel Bridge, the addition of an auxiliary bridge on Reactor Building El.340, and other special considerations for the shroud repair.,This safety evaluation (SE)covers the shroud repair installation activities and will supplement SE 94-080,"Core Shroud Repair." Safety Evaluation Summary: The installation of the core shroud repair requires that special equipment and processes be used to minimize the in-vessel debris generation and provide minimal impact on other work being performed on Reactor Building El.340.The design and function of the spent fuel pool cooling (SFP)and the reactor water cleanup systems are not being altered during the repair installation.
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| Both systems have been evaluated and will continue to perform as designed during and after the repair installation.
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| The SFP system is designed to remove particles as small as 1 micron.The swarf particles from the EDM process which enter the skimmers from the tank overflow will be almost entirely removed in the filters.The remaining particles will be less than 1 micron in size and will not affect the function of the SFP system.
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| Safety Evaluation Summary Report Page 36 of 60 Safety Evaluation No.: Safety Evaluation Summary: 95-007 (cont'd.)(cont'd.)The cleanup system is designed to maintain high reactor water purity by continuously purifying a portion of the recirculation flow.The debris size expected from the shroud repair is 1 to 50 micron;therefore, any particles that the cleanup system cannot remove are assumed to be small enough that a particle of that size.could currently be in the system and is not a concern.The volume of particles expected to remain in the vessel and SFP system following the repair, after filtering, is considered insignificant when compared to the total volume of water in the vessel.The auxiliary bridge and refuel bridge buggy will not be used for moving fuel.The-auxiliary bridge has been analyzed and is acceptable for use over irradiated fuel.The refuel bridge buggy will not be moved over fuel unless it is tied off to the refuel bridge.The requirements of NUREG-0612 will be met through the use of N1-MMP-GEN-914, which is referenced in the GE shroud repair procedures.
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| The tooling for"heavy loads" has been designed and will be used in accordance with NUREG-061 2.Based on the evaluation performed, it is concluded that the special equipment and processes required for the installation of the core shroud repair do not constitute an unreviewed safety question.
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| Safety Evaluation Summary Report Page 37 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 95-009 Simple Design Change SC1-0102-93 Table Vl-3a Sh 1&3;VIII-1, Vill-2, Vill-9, VIII-32, Figures Vill-1, Vill-2;XV-48 System: Main Steam (MSS), Reactor Protection (RPS), Offgas (OFG), Postaccident Sampling, Emergency Cooling Title of Change: Removal of Main Steam Line High Radiation Scram/MSIV Isolation Signal to Implement Technical Specification Amendment 149 Description of Change: J This safety evaluation analyzed plant changes associated with Technical Specification Amendment (TSA)149 that are not explicitly described in the amendment and, therefore, are not covered by the NRC's safety evaluation for the amendment.
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| TSA 149 deleted the automatic reactor scram and main steam line isolation functions of the main steam line radiation monitors.TSA 149 only states that the main steam isolation valves (MSIV)isolation function will be removed;however, other reactor vessel isolation valves are coupled with the MSIVs logic circuits.Therefore, when the MSIV isolation signal is removed, these other valves will also not isolate on high radiation.
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| Safety Evaluation Summary: This safety evaluation demonstrates that all necessary safeguards have been taken to identify and address all licensing and design basis requirements for the design and operation of the systems impacted by this design change, and for installation of the design change.Based on review of NEDO 31400A, various system designs and system failure criteria, the change will not result in a radiological release beyond existing limits nor affect operation of safety-related equipment.
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| Based on the evaluation performed, it is concluded that these changes do not invoive an unreviewed safety question.
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| Safety Evaluation Summary Report Page 38 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 95-010 DER 1-95-0133 X-39 Fuel Spent Fuel Storage This safety evaluation evaluated a clarification change to the UFSAR where a"maximum fuel enrichment" of 3.75 percent is specified for the boraflex racks.It has been established that this value is the peak design lattice enrichment (nominal value).Safety Evaluation Summary: 1 Storing fuel with a peak design lattice enrichment of 3.75 percent (nominal value)in the boraflex racks is consistent with maintaining the K-effective below 0.95.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 39 of 60 Safety Evaluation No.: Implementation Document No.: 95-011 NEDE-24011-P-A-10 NEDE-2401 1-P-A-10-US (GESTAR II)UFSAR Affected Pages: I-9, I-10, I-15;IV-7, IV-12, IV-13, IV-30;V-20;XV-2, XV-3, XV-5, XV-6, XV-7, XV-15, XV-68, XV-79, Tables XV-2, XV-9 Sh 1 h 2, XV-9a System: Title of Change: Description of Change: Various Operation of NMP1 Reload 13/Cycle 12 This change consisted of the addition of new fuel bundles and the establishment of a new core loading pattern for Reload 13/Cycle 12 operation of Unit 1.Two Hundred (200)new fuel bundles of the GE11 design were loaded.All 164 of the PSxSR bundles from Cycle 10, and 36 of the GE Sx8EB bundles from Cycle 11, were discharged to the spent fuel pool.Various evaluations and analyses were performed to establish appropriate operating limits for the.reload core.These cycle-specific limits were documented in the Core Operating Limits Report.Safety Evaluation Summary: The reload analyses and evaluations are performed based on the General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-10 and NEDE-24011-P-A-10-US (GESTAR II).This document describes the fuel licensing acceptance criteria;the fuel thermal-mechanical, nuclear, and thermal-hydraulic analyses bases;and the safety analysis methodology.
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| For Reload 13, the evaluations included transients and accidents likely to limit operation because of minimum critical power ratio considerations; overpressurization events;loss-of-coolant accident;and stability analysis.Appropriate consideration of equipment-out-of-service was included.Limits on plant operation were established to assure that applicable fuel and reactor coolant system safety limits are not exceeded.Based on the evaluation performed, it is concluded that Unit 1 can be safely operated during Reload 13/Cycle 12, and that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 40 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-013 Rev.0 6 1 Mod.N1-94-003 N/A Reactor Vessel Reactor Core Shroud Repair Installation Prior to NRC Approval of Adequacy Under 10CFR50.55a(a)
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| (3)Description of Change: This modification installed the core shroud repair components prior to NRC approval of the adequacy of these components as alternative repairs for the core shroud horizontal welds H1 through H8./Safety Evaluation Summary: These shroud repair components are not required to perform any design basis function during the time period required for NRC review of the adequacy of the proposed modification.
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| The existing shroud H1 through H8 weld structural integrity which existed prior to plant shutdown is maintained as discussed in the analysis.The restriction that the reactor remain in cold shutdown except for entry into the hot shutdown condition for performance of the portion of the noncritical hydrotest and/or CRD scram time testing above 212'F, in accordance with Technical Specification 3.2.2.e, until NRC approval is obtained guarantees that Unit 1 will remain in compliance with the NRC Safety Evaluation Report dated January 13, 1995.This restriction plus the restriction to maintain total flow less than rated mass flow (67.5 Mlb/hr)during reactor noncritical hydrotesting ensures that no significant thermal or pressure loads are carried by the shroud or tie-rod assemblies.
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| The analysis has demonstrated that the tie-rod preload does not increase the stress significantly in any of the shroud welds such that the probability of failure is increased.
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| The analysis has demonstrated that the bypass leakage is less than 1'/o at rated recirculation flow conditions which bounds the cold and hot shutdown required core cooling conditions.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 41 of 60 Safety Evaluation No.: Implementation Document No.: 95-014 Rev.1 Temporary Mod.95-009 Temporary Mod.95-012 UFSAR Affected Pages: System: Title of Change: N/A Reactor Core Spray Removal of Core Spray Minimum Flow Relief Valves Description of Change: These temporary modifications removed the minimum flow relief valves for each of the core spray pump sets and replaced them with blind flanges.Flanges were installed at the existing end connections for each relief valve./This disabled the minimum flow capability for each of the pump sets of the core spray system.This change was required to support reload activities while the Core Spray Minimum Flow Modification, N1-90-041, was being re-evaluated for problems associated with the relief valves.This safety evaluation is only applicable while/the reactor is depressurized.
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| Safety Evaluation Summary: Removal of the minimum flow relief valves and replacing them with blank flanges for refuel/cold shutdown operation does not affect the ability of the core spray system to perform its function to cool the core.This change does not adversely affect the core spray system performance in a manner which would increase the probability of occurrence of an accident previously evaluated in the safety analysis report.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 42 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 95-015 N/A N/A System: N/A Title of Change: CO, Pellet Decontamination Facility at,Off Gas Building-261 Description of Change: Operation of the CO, (dry ice)pellet cleaning facility is for decontamination of miscellaneous tools and equipment post-RFO-13.
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| The actual CO, pellet blaster is located inside of a modular decontamination building (a portable kelly building)located in the Offgas Building on elevation 261.A CO, storage tank and compressor are located outdoors at the east side of the Offgas Building.\Safety Evaluation Summary: The use and operation of the CO, cleaning facility at Offgas Building elevation 261 with a CO, storage tank and compressor immediately adjacent to the east side of the Unit 1 Offgas Building does not create the possibility of an accident or malfunction not previously analyzed, increase the probability of an accident or malfunction already analyzed, reduce the margin of safety in any Technical Specifications, or increase the consequences of any accident or malfunction already analyzed.In addition, the proposed controls and operation of the facility will not create a new radioactive effluent pathway or create an unmonitored release of radioactivity.
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| Based on the evaluation performed, it is concluded that the temporary use of the CO~cleaning facility does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 43 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 95-016 LDCR No.1-95-UFS-019 X-47;10A-7, 10A-8, 10A-9, 1)A-24, 10A-32, 10A-41, 10A-44, 10A-46a, 10A-48, 10A-52a, 10A-55a, 10A-57;10B-196;XIII-1 thru Xll!-6, Xlll-10, Table Xlll-1, Figures XIII-1 thru XIII-4, System: Title of Change: N/A Deletion of Fire Protection Report Requirements in the Fire Hazards Analysis Description of Change:/This change deleted the requirement for sending fire protection program reports to the NRC for inoperable fire protection systems and components as required by the Fire Hazards Analysis.Safety Evaluation Summary: The Fire Hazards Analysis contains reporting requirements for inoperable fire protection systems and components.
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| These reports are made upon meeting specific criteria delineated within the Fire Hazards Analysis.In 1992, when License Amendment 132 was issued by the NRC, an additional Technical Specification (6.9.2)was added to require that fire protection program noncompliances be reviewed under the provisions of 10CFR50.72 and 10CFR50.73.
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| Reports are now submitted under these provisions, as determined through the plant procedures governing Deviation/Event Reporting.
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| The preparation and submittal of fire protection reports in addition to those required under Technical Specification 6.9.2 is no longer required for licensees who have removed the fire protection program elements from the Technical Specifications per the guidance of NRC Generic Letters 86-10 and 88-12.Unit 1 completed this process through the granting of License Amendment 132.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 44 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 95-017 Dwg.B-69007-C VI-21 58;201.2 LT 58-05 Isolation Capabilities LT 58-05 instrument lines do not contain two manual valves.Instead, isolation is provided by a single manual valve and diaphragm seal assemblies.
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| Safety Evaluation Summary:/The diaphragm seal assembly exceeds design ratings of the torus under normal and accident conditions.
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| The seal assembly effectively isolates the transmitter from process fluid.The seal assembly is located close to the torus room penetration and the first manual valve.The diaphragm seal assembly is an Appendix J testable configuration.
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| The valve/seal assembly combination provides sufficient isolation capability to mitigate potential instrument line leakage.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 45 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-018 Gale.S7-RX340-W01 XVI-70, Table XVI-31 Sh 1 N/A UFSAR Changes for Pressure Relief Panel Discrepancies Description of Change: UFSAR Sections III.A.1.2 and VI.C.1.2 stated that the pressure relief panels in the Turbine and Reactor Buildings blow out at 45 psf.UFSAR Table XVI-31, page XVI-185, and the discussion on pa'ge XVI-187, stated that the pressure relief panels blow out at 40 psf in both the Reactor and Turbine Buildings.
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| Calculation S7-RX340-W01 and DCR N1-95001LG329, documenting existing as-built conditions for field work done to the relief panels in Refueling Outage 13, indicate that the correct blowout pressure is 45 psf.This change to the UFSAR corrected discrepancies relating to the blowout pressures for the pressure relief panels in the Reactor and Turbine Buildings.
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| Safety Evaluation Summary: The above-referenced documents show that the correct blowout pressure is 45 psf for the pressure relief panels in both buildings.
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| This change corrects stated pressures of 40 psf to 45 psf on the UFSAR affected pages.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 46 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-019 Gale.S15-72-F009 X-25, Figure X-6 Service Water Revision to NMP1 FSAR Chapter X, Pages X-31, X-32, and X-33 Description of Change: The following changes were made to UFSAR Chapter X to reconcile a discrepancy between text and figure pertaining to the position of the blocking valves on the supply header.1.Text in Section X-2.0 was changed from: "During normal operation of the turbine building system only one service water supply is used;the other line is valved off as a spare.However, on the reactor building system both service water supply lines are engaged at all times for maximum reliability." to: "During normal operation both the supply headers on the RBCLC side and TBCLC side are engaged by keeping the blocking valves open;however, to perform maintenance or other plant activities one of the RBCLC side and one of the TBCLC side blocking valves can be secured." 2.Figure X-6 was revised to show both sets of supply header blocking valves to be open in the normal operating configuration.
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| Safety Evaluation Summary: The service water system can adequately meet its flow demand in a configuration where both of the valves on the RBCLC side and TBCLC side are open or in a configuration where one of the two valves on the RBCLC and TBCLC side are secured.This safety evaluation analyzes both sets of blocking valves to be open under normal operating condition.
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| This will reduce potential clogging and corrosion of the portions of the supply header where stagnant flow conditions would occur if the blocking valve was secured.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 47 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-020 DER 1-94-0354 lll-7, III-23, III-24 N/A Turbine and Administration Building Description Changes Description of Change: This safety evaluation evaluated changes in the configuration of the Administration and Turbine Buildings as follows: 1.The new personnel access control is at el.248'-0" of the Administration
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| /Building.2.The Administration Building basement has an entrance to the Turbine Building.3.j The personnel locker room was added to the Administration Building ground floor description in the UFSAR.A description of the office for shops and stores was deleted, as was the reference to the monitor room, the personnel decontamination room, laundry room, and lunch rooms.4.The reception area and the location of the lunch room, locker rooms, radiation protection offices, and access to the radiologically-controlled area (RCA)were deleted from the UFSAR.Safety Evaluation Summary: Changing the personnel RCA access, shop and radiation protection office, lunch rooms, laundry room, and deleting the reception area does not create the possibility of accident or malfunction not previously analyzed, affect plant equipment or systems, or reduce the margin of safety in any Technical Specification.
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| Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 48 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: 95-021 Procedures GAP-POL-01, NSAS-POL-01 X-47;10A-7, 10A-8, 10A-9;10B-196;XIII-1 thru Xlll-4, Xlll-8, Figures.XIII-1 thru'III-4;B.1-2, B.1-3, Table B-1 Sh 1&2 System: Title of Change: N/A Reorganization; Changes to GAP-POL-01 Bc NSAS-POL-01 to Establish Nuclear Business Management Organization Description of Change: Procedures GAP-POL-01 and NSAS-POL-01 have been revised'to reorganize the functions of Finance, Computer Software Development, Business Planning, and Nuclear Procurem'ent under a new organization titled,"Nuclear Business Management," reporting to the Vice President Nuclear Generation.
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| Safety Evaluation Summary: These procedure changes establish departmental responsibilities and lines of authority, responsibility, and communication for the Nuclear Business Management organization.
| |
| The proposed organization structure satisfies the criteria of SRP 13.1.1 and conforms with the requirements of Section 6.1.2 of the Plant Technical Specifications.
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| The proposed changes do not impact accident or malfunction initiation or consequences.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 49 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-022 Procedures GAP-POL-01, NSAS-POL-01 XIII-3, XIII-7 N/A Reorganization; Changes to GAP-POL-01
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| &NSAS-POL-01 to Transfer Management and Operational Responsibility for the Site Sewage Treatment Facility from Technical Services (Environmental) to Unit 1 Chemistry Description of Change:!Procedures GAP-POL-01 and NSAS-POL-01 have been revised to transfer management and'operational responsibility for the Site Sewage Treatment Facility from the Technical Services Branch (Environmental) of the Nuclear Safety Assessment
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| &Support Department to the Unit 1 Chemistry Branch of Nuclear Generation.
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| Safety Evaluation Summary: These procedure changes establish departmental responsibilities and lines of authority, responsibility, and communication for management and operation of the Site Sewage Treatment Facility.The proposed organizational structure satisfies the criteria of SRP 13.1.1 and conforms with the requirements of Section 6.1.2 of the Plant Technical Specifications.
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| The proposed changes do not impact accident or malfunction initiation or consequences.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 50 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-023 Procedure NSAS-POL-01 Xlll-3, Figure XIII-4 N/A Reorganization; Changes to NSAS-POL-01 to Transfer Procedure Program Coordination from Technical Services to Quality Assurance Description of Change: Procedure NSAS-POL-01 has been revised to delete the responsibility assigned to the Manager Technical Services to manage implementation of.the procedure program including publication.
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| The"managing" function assigned to the Manager Technical Services was to provide overall coordination of the procedure program.Responsibility for overall coordination of the procedure program has been transferred to the Manager Quality Assurance as a Quality Assurance administrative service function./Safety Evaluation Summary: These procedure changes establish departmental responsibilities and lines of authority, responsibility, and communication for implementation of the procedure program including publication.
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| The proposed organizational structure satisfies the criteria of SRP 13.1.1 and conforms with the requirements of Section 6.1.2 of the Plant Technical Specifications.
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| The proposed changes do not impact accident or malfunction initiation or consequences.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 51 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-025 Procedure NSAS-POL-01 Xlll-3, Xlll-4, Figure XIII-4 N/A Reorganization; Changes to NSAS-POL-01 to Transfer Environmental Protection Functions from Technical Services to Licensing and Emergency Preparedness Description of Change: Procedure NSAS-POL-01 has been revised to reorganize (transfer) responsibility for the functional areas of environmental monitoring (including control of hazardous and industrial wastes, and assessing effects of radioactive effluent)from the Manager Technical Services to the Manager Licensing, and meteorological monitoring from the Manager Technical Services to the Director Emergency Preparedness.
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| /Safety Evaluation Summary: These procedure changes establish departmental responsibilities and lines of authority, responsibility, and communication for implementation of the procedure program including publication.
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| The proposed organizational structure satisfies the criteria of SRP 13.1.1 and conforms with the requirements of Section 6.1.2 of the Plant Technical Specifications.
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| The proposed changes do not impact accident or malfunction initiation or consequences.
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| Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 52 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-058 GAP-POL-01 N/A N/A Dissolution of the Unit 2 Technical Support Branch Support Section Description of Change: The Nuclear Division is organized into departments with departments being subdivided into branches and branches being subdivided into sections.Sections compose the lowest organizational tier of the Nuclear Division.This safety evaluation analyzes changes to the Technical Support Branch.as a result of dissolving the Support Section.The organizational change involves dissolution of the Unit 2 Technical Support Branch's Support Section by:/~Eliminating the position of Lead Support Engineer.~Converting the Administrative Technician position to a supervisory position with the title of Supervisor Administrative Support.~Having the Supervisor Administrative Support report directly to the Manager Technical Support and to be responsible for: 1.Administration of Station Operations Review Committee (SORC).2.Administration of the plant technical review program.3.Supervision of the Branch's clerical staff.~Redistributing the remaining personnel and functions (including coordination of plant modifications) of the Support Section to the System Engineering Sections under the supervision of the Lead System Engineers.
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| Safety f valuation Summary Report Page 53 of 60 Safety Evaluation No.: Safety Evaluation Summary: 95-058 (cont'd.)(cont'd.)Dissolution of the Unit 2 Technical Support Branch's Support Section does not involve a change to the established responsibilities of the Technical Support Branch as described in the UFSAR;only the reporting structure within the Branch is being affected.The organization continues to provide for the integrated management of activities that support the operation of the facility and maintains clear management control and effective lines of authority and communication; Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 54 of 60 Safety Evaluation No.: Implementation Document No.: 95-102 Rev.0 5 1 Procedures NSAS-POL-01, GAP-POL-01, NEP-POL-0101, NIP-FPP-01, NIP-TQS-01, GAP-OPS-01 UFSAR Affected Pages: X-47, 10A-7, 10A-S, 10A-9;10B-196;XIII-1 through XIII-6, Xlll-10, Table Xlll-l, Figures XIII-1 through XIII-4;B.1-3, Table B-1 Sh 1 System: Title of Change: N/A Restructuring of Nuclear SBU in Accordance with Revised Procedures NSAS-POL-01, GAP-POL-01, NEP-POL-01, NIP-FPP-01, NIP-TQS-01 and GAP-OPS-01 Description of Change: NSAS-POL-01,"Composition and Responsibility of the Nuclear Safety Assessment 8c Support Organization," GAP-POL-01,"Composition and'Responsibility of the Nuclear Generation Organization," NEP-POL-01,"Nuclear Engineering Department Organization," NIP-FPP-01,"Fire Protection Program," NIP-TQS-01,"Qualification and Certification," and GAP-OPS-01,"Administration of Operations," have been revised to:~Transfer responsibility for Office Administration activities from Nuclear Safety Assessment L Support (NSAS)Site Services to Business Management,-and remove the Business Management organization from the Nuclear Generation Department (the General Manager Business Management reports directly to the Executive Vice President Nuclear).~Transfer responsibility for Procurement and Integrated Planning functions from the Business Management Organization to the Engineering Department.
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| ~"Unitize" the coordination of contractor maintenance/modification activities previously performed by NSAS Site Services and transfer responsibility for the functions to the Maintenance Branch at each unit.
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| Safety Evaluation Summary Report Page 55 of 60 Safety Evaluation No.: Description of Change: 95-102 Rev.0&1 (cont'd.)(cont'd.)Transfer responsibility for administration and implementation of.the Fire Protection Program from NSAS Technical Services to Unit 1 Operations.
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| ~Transfer responsibility for administration of Central Maintenance activities (a site function that includes M&TE calibration, security system support,.material testing, and warehouse preventive maintenance) from NSAS Technical Services to Unit 2 Maintenance.
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| Transfer responsibility for administration of Buildings&Grounds/Facilities Planning activities (a site function)from the NSAS Site Services to Unit 1 Maintenance.
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| Abolish the positions of Manager Technical Services and Manager Site Services.i~Transfer responsibility for In-service Testing at Unit 1 from Operations to Maintenance.
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| /~Consolidate the Unit 1 Operations Engineering and Planning Sections and combine with the Fire Protection Section (currently in NSAS Technical Services)into a new Operations Support Section to be headed by a new position, General Supervisor Operations Support.Assign I&C Technicians to Unit 2 Technical Support Branch-Lead Performance Engineer.Consolidate Unit 1 and Unit 2 Operations Training organizations into one common section under the supervision of the General Supervisor Operations Training.~Assign responsibility for management of Site Relay&Control Testing activities (formerly a Corporate support function)to the Manager Maintenance Unit 2.~Transfer responsibility for Maintenance Planning at Unit 2 from Maintenance to Work Control/Outage.
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| Safety Evaluation Summary Report Page 56 of 60 Safety Evaluation No.: Safety Evaluation Summary: 95-102 Rev.0 5 1 (cont'd.)These procedure changes establish departmental responsibilities and lines of authority, responsibility, and communication within the Nuclear SBU.The proposed organizational structure satisfies the criteria of SRPs 9.5.1, 13.1.1 and 13.1.2-13.1.3, and conforms with the requirements of Section 6.2 of the Plant Technical Specifications.
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| The proposed changes do not impact accident or malfunction initiation or consequences.
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| Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 57 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-106 N/A Figure III-1-N/A Demolition of Temporary Structures Inside the Protected Area, East of the Unit 2 Structures Description of Change: This safety evaluation addresses the demolition of the following buildings located east of the Unit 2 plant structures.
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| 1.2.3.4.5.Carpenter's shop Paint shop Electric fab shop Insulators fab shop Maintenance storage building All of these buildings were built for use as temporary buildings during the construction of Unit 2.These buildings have been demolished and activities consolidated within the remaining buildings.
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| Safety Evaluation Summary: All of the buildings to be demolished are located in an area that was not used as a flow channel for the Probable Maximum Precipitation analysis.Removal of these buildings and the consequent reduction in the runoff coefficient would make the analysis more conservative.
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| These buildings have no impact on the previously
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| .calculated X/0 values.The design margins for the control room fresh air intakes are not compromised.
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| Location of demolition activities are adequately separated from safety-related systems and structures to preclude any adverse impact from construction activities.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 58 of 60 Safety Evaluation No.: implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-107 NTP-TQS-102 N/A N/A NTP-TQS-102, Licensed Operator Requalification Training Changes to Reflect the Requirements of the NRC Approved Systems Approach to Training Program Description of Change: This change more clearly defines a Systems Approach to Training (SAT)-based Requalification program./The SAT-based program allows flexibility in addressing identified weaknesses and current issues while satisfying required training specified in 10CFR55.Safety Evaluation Summary: Unit 1 and 2 Licensed Operator Training Programs have been developed using a Systems Approach to Training and are accredited by the National Nuclear Accrediting Board.Based on this certification and NRC approval, this change satisfies 10CFR55 requirements for Licensed Operator Requalification Training.Changing the Licensed Operator Requalification Training program to more cleariy define a SAT-based program does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 59 of 60 Safety Evaluation No.: implementation Document No.: UFSAR Affected Pages: System: Title of Change: Description of Change: 95-128 N/A N/A N/A Aerial Radiation Survey of NMP Site Section 2.2.3.1.7 of the Unit 2 USAR discusses aircraft crashes, but the discussion is only in reference to airplane flights associated with nearby airports and helicopter flights to and from the site.The NRC contracted EG&G to perform an aerial radiation survey of the Nine Mile Point site.This survey involved helicopter flights directly over the site and, as stated above, tlie Unit 2 USAR only evaluated flights to and from the site.Safety Evaluation Summary: The helicopter flights directly over the site, for the purpose of performing an aerial radiation survey, were evaluated and found to present an insignificant risk of an aircraft crash on site.The helicopter accident rates, a previous Stone&Webster Engineering Corp.calculation, an Argonne National Laboratory Study of Aircraft Crash Hazards, and the NRC Standard Review Plan (SRP)were used to assess the risk associated with the Survey Plan described by the EG&G pilot.The assessment resulted in a crash probability between 8AE-8 and 7E-7.Per the SRP, this probability is sufficiently low enough that crashes need not be considered as design basis events.Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| Safety Evaluation Summary Report Page 60 of 60 Safety Evaluation No.: Implementation Document No.: UFSAR Affected Pages: System: Title of Change: 95-130 Nuclear Division Policy POL B.1-2 N/A Reorganization; Change to Nuclear Division Policy to Reflect Establishment of the'orporate Officer Position"Executive Vice President-Generation Business Group/Chief Nuclear Officer" Description of Change: The Nuclear Division Policy"POL" has been revised to reflect.the establishment of the corporate officer position Executive Vice President-Generation Business Group/Chief Nuclear Officer.This Executive Vice President reports directly to the Niagara Mohawk Power Corporation President.
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| The Executive Vice President-Nuclear is subordinate to the Executive Vice President-Generation Business Group/Chief Nuclear Officer and continues to have overall responsibility for the admjnistration and operation of the Nuclear SBU.Safety Evaluation Summary: The proposed upper management organizational structure satisfies applicable acceptance criteria, and does not impact accident or malfunction initiation or consequences.
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| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
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| NRC FORM 35 (11-93)NRCMD 3.53 Page 1 of U.S.NUCLEAR REGULATORY COMMISSION
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| 'coms T~NSFER ggyy yS'gS'1.JOB NUMBER 3911 4.VOLUME IN CUBIC FEET 5 I.P 2.FRC ACCESSION NUMBER 5.LOCATION NUMBER(S)17127-17131 73.3.DATE JOB RECEIVED 08/13/1997 6.DATE ELIGIBLE FOR DESTRUCTION 10/01/2029 7.DATE ELIGIBLE FOR TRANSFER TO NARA 10.COMMENTS LICENSE EXPIRES 08/22/2009 8.ORIGINATING OFFICE CODE FC/NRR/[X]AF Q AF-V3 9.FACILITY CODE Q AF-V5 Q FRC G 0<T<'e'8-Z 11.ORIGINATING OFFICE(Office/Division/Rranch)
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| FC/NRR/12.FILE CUSTODIAN (Name/Telephone)
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| Tyrone Greene (301)415-6281 13.LOCATION OF RECORDS(ffuffdfng/Room)
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| TSC3 14.T(TLR AF RFCARDSFREFS (RefertoNUREG0910)(Complete a separate form for each serfes)Nuclear Power Plant Docket Files(E)15.RFCARDSCHFA11LENUMBER (Refer to NUREG.0910)
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| NRCS 02-20.09E 16.CLASSIFICATION OF RECORDS[X]UNCLASSIFIED Q OTHER(Specify below)HIGHEST CLASSIFICATION
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| [X]PAPER Q MICROFORMo
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| *PROVIDE SPECIFIC MEDIUM 17.RECORD MEDIUM Q AUDIOVISUAL+
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| Q MACEHNE READABLEo 18.TOTAL NUMBER OF CONTAINERS 19.TYPE OF CONTAINER(S)(NOTEt Paper reconfs must be retired ln fwubfe foot record center boxes)[X]RECORD CENTER BOX Q OTHER (Specify)20 GENERAL DESCRIPTION OF RECORDS (Prorhfc a genera(dcserfptfon of the records using items commonly used by the stoff famfNar with thc records.)50-220 NIAGRA MOHAWK POWER CORP.NINE MILE POINT UNIT 1 NRC FORM 35 (11-93)s 1'I Page 2 of DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| lf more space is required, use NRC for 35A,'Records Transfer Continuation'.
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| 21.CONTAINER NUMBER~m:M 23.RECORDS DESCRIPTION 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)24.DATE SPAN (Month/Yrnrs) 10/1996-10/1996 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)09/1996-09/1996 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)08/1996-08/1996 50-220 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY)06/1996-06/1996 50-220 NINEMILEPI'I REVISION TO EPIPS (SECOND COPY)04/1996-04/1996 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)03/1996-03/1996 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)02/1996-02/1996 25.RECORDS LIAISON OFFICFR-Qprg/rrnne and Signa/urn Tyrone Greene 27.REVIEWING ANALYST..S/gna/nrc Latravetta Lee 26.DATE 08/13/1997 28.DATE 08/13/1997 29.DATE TRANSFERRED TO FRC Q 30.DATE DESTROYED 31.DATE TRANSFERRED TO NARA NRC FORM 35 (11.93)
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| ~5 NRC PORN 35 II 1-93I NRCMD 3.53 Page 3 of U.S.NUCLEAR REGULATORY COMMISSION RECORDS TRANSFER Continuation DETAILED INVENTORY OF RECORDS MING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated OI/3I/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAINER NUAIBER W~@M 23.RECORDS DESCRIPTION 24.DATE SPAN (Month/Years) 50-220 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY)12/1995-12/1995 50-220 NINEMILEPTI REVISION 35 TO SITE EMERGENCY PLAN AND EPIPS (SECOND COPY)01/1997-01/1997 50-220 NINEMILEPT1 REVISION 34 TO SITE EMERGENCY PLAN AND EPIPS (SECOND COPY)05/1996-05/1996 50-220 NINEMILEPT1 REVISION 33 TO SITE EMERGENCY PLAN AND EPIPS (SECOND COPY)04/1996-04/1996 50-220 NINEMILEPTI EVACUATION TRAVEL TIME ESTIMATES FOR THE NINE MILE POINT EMERGENCY PLANNING ZONE 06/1992-06/1992 50-220 NINEMILEPTI REVISION 14 TO UFSAR (SECOND COPY)06/1996-06/1996 2 50-220 NINEMILEPT1 REPORT TO ACRS 03/1969-03/1969.NRC FORM 35 (11-93)
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| NRC FORM 35/1I 931 NRCMD 3.53 Page 4 of U.S.NUCLEAR REGULATORY COMMISSION'ECORDS TRANSFER COntinuatiOn DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/9G, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAlNER NUSIBER 23.RECORDS DESCRIP1'lON 24.DATE SPAN (Month/Yenrs) 50-220 NINEMILEPTI ANNUAL FINANCIAL REPORT 1992 RG&EC 05/1993-05/1993 50-220 NINEMILEPT1 ANNUAL FINANCIAL REPORT 1992 LILCO 05/1993-05/1993 50-220 NINEMILEPTI ANNUAL FINANCIAL REPORT 1992 NYSE&GC 05/1993-05/1993 50-220 NINEMILEPT1 ANNUAL FINANCIAL REPORT 1992 CHG&EC 05/1993-05/1993 50-220 NINEMILEPT1 ANNUAL FINANCIAL REPORT 1992 NMPC 05/1993-05/1993 50-220 NINEMILEPT1 ANNUAL FINANCIAL REPORT 1991 NMPC I V1992-11/1992 50.220 NINEMILEPT1 ANNUAL FINANCIAL REPORT 1991 RG&EC 05/1992-05/1992 50-220 NINEMILEPTI ANNUAL FINANCIAL REPORT 1991 LILCO 05/1992-05/1992 NRC FORM 35 (11-93)t I
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| NRC FORM 35/I 1.9.Q NRCMD 3.53 Page 5 of U.S.NUCLEAR REGULATORY COMMISSION RECORDS TRANSFER Continuation DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAINER NUSIBER~%;;M 23.RECORDS DESCRIPI'ION 24.DATE SPAN (Month/Y<<nrs) 50-220 NINEMILEPTI ANNUAL FINANCIAL REPORT 1991 NYSE&GC 05/1992-05/1992 50-220 NINEMILEPTI ANNUAL FINANCIAL REPORT 1991 CHG&EC 05/1992-05/1992 50-220 NINEMILEPTI SAFETY EVALUATION OF THREE TMI ACTION PLAN ITEMS (NUREG-0737)>
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| H.F.'1.4, H.F.1.5 AND H.F.1.6, BY CONTAINMENT 06/1982 06/1982 50-220 NINEMILEPT1 SAFETYEVALUATIONREPORT 06/1981-06/1981 50-220 NINEM ILEPTI SAFETY EVALUATION REPORT REGARDING FIRE PROTECTION 07/1979-07/1979 50-220 NINEMILEPT1 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS REGARDING ADMINISTRATIVE CHANGES 06/1993-06/1993 50-220 NINEMILEPT1 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS CONSISTING OF ADMINISTRATIVE CHANGES 05/1993-05/1993 NRC FORM 35 (11-93)
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| I I NRC FORM 35 (11-931 NRCMD 3.53 Pape 6 of U.S.NUCLEAR REGULATORY COMMISSION'ECORDS TRANSFER Continuation DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAINER NUMBER MXM 23.RECORDS DESCRIPTION 24.DATESPAN (Monrh/Years) 50-220 NINEMILEPTI PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS TO INCREASE SURVEILLANCE TEST INTERVALS AND ADD ALLOIVABLE OUT-OFS 12/1992-12/1992 50-220 NINEMILEPTI PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS REGARDING COOLANT SYSTEM 02/1992-02/1992 50-220 NINEMILEPT1 DRAFT TECHNICAL SPECIFICATIONS (FSAR)(REVISED)03/1969-03/1969 50-220 NINEMILEPTI SAMPLE TECHNICAL SPECIFICATIONS 12/1965-12/1965 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)06/1997-06/1997 50-220 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)05/1997-05/1997 50-220 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY)04/1997-04/1997 50-220 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY)03/1997-03/1997 NRC FORM 35 (11-93)
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| I NRC FORM 35/I 1-931 NRCMD 3.53 Page 7 of U.S.NUCLEAR REGULATORY COMMISSION'ECORDS TRANSFER Continuation DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAINER NUi%1BER W>A~M 23.RECORDS DESCRIPTION 24.DATE SPAN (Month/Years) 50-220 NINEMILEPf1 REVISION TO EPIPS (SECOND COPY)10/1996-10/1996 50-220 NINEMILEPTI REACTOR CONTAINMENT BUILDING INTEGRATED LEAKAGE RATE TEST FINAL REPORT 04/1993-04/1993 50-220 NINEMILEPTI SAFETY EVALUATION
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| ==SUMMARY==
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| REPORT, 1992 06/1992-06/1992 50-220 NINEMILEPT1 SAFETY EVALUATION
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| ==SUMMARY==
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| REPORT 1990 SUBMITTED WITH UPDATED FSAR, REVISION 8 AND AMENDMENTS 06/1990-06/1990 3 50-220 NINEMILEPT1 APPROVED PROCEDURES POWER ASCENSION PROGRAM 01/1990-01/1990 W 50-220 NINEMILEPT1 CONTAINMENT LEAK TEST 05/1984-05/1984 50-220 NINEMILEPI'1 EQUIPMENT QUALIFICATION PROGRAM 05/1984-05/1984 NRC FORM 35 (11.93)
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| ~h NRC FORM 35/11-9:41 NRCMD 3.53 Page 8 of U.S.NUCLEAR REGULATORY COMMISSION'ECORDS TRANSFER Continuation DETAILED INVENTORY OF RECORDS MING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide ail identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAINER NUMBER 23.RECORDS DESCRIPTION 24.DATE SPAN (Month/Yeats) 50.220 NINEMILEPT1 TECHNICAL SUPPLEMENT TO PETITION FOR CONVERSION FROM PROVISIONAL OPERATING LICENSE TO FULL-TERM OPERATING LICE 07/1972-07/1972 50-220 NINEMILEPTI FIFTH ADDENDUM TO TECHNICAL SUPPLEMENT TO PETITION TO INCREASE POWER LEVEL OV1971-OV1971 50-220 NINEMILEPT1 FOURTH ADDENDUM TO TECHNICAL SUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 12/1970-12/1970 50-220 NINEMILEPT1 THIRD ADDENDUM TO TECHNICAL SUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 12/1970-12/1970 0-220 NINEMILEPT1 SECOND ADDENDUM TO TECHNICAL SUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 10/1970-10/1970 50-220 NINEMILEPTI FIRST ADDENDUM TO TECHNICAL SUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 10/1970-10/1970 50-220 NINEMILEPT1 RESPONSE TO B.L.RIDINGS OF KINSTON, TN'S LETTER OF OCTOBER 13, 1993 11/1993-11/1993 50-220 09/1993-09/1993 NRC FORM 35 (11-93)
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| WRC FORM 35 (11-931 NRCMD 3.53 Page ilof U.S.NUCLEAR REGULATORY COMMISSION RECORDS TRANSFER Continuation DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval.
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| Provide all identifying numbers that apply to the material (e.g.docket number, case number, contract number, form number, etc.).Provide the date span of each item as well as any date needed to apply the disposition instructions (e.g.license terminated 01/31/90, case closed 01/31/90, etc.).Double space between items in the same container.
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| Triple space when beginning a new container.
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| 21.CONTAINER NUMBER 23.RECORDS DESCRIPTION 24.DATE SPAN (MonthA'ears) 50-220 NINEMILEPTI SUPERSEDED PAGES PER REVISION TO EPP 10/1996-10/1996 50-220 NINEMILEPTI SUPERSEDED PAGES PER REVISION TO EPP 10/1996-10/1996 50-220 NINEMILEPT1 SUPERSEDED PAGES PER REVISION TO EPP 09/1996-09/1996 50-220 NINEMILEFf1 SUPERSEDED PAGES PER REVISION TO EPP 11/1995-11/1995 50-220 NINEMILEPTI SUPERSEDED PAGES PER REVISION TO SITE EMERGENCY PLAN 01/1997-01/1991 50-220 NINEMILEPI'1 SUPERSEDED PAGES PER REVISION 13 TO UFSAR 06/1995-06/1995~,NRC FORM 35 (11.93) l NINE MILE POINT UN3Z 1 FSAR-(UPDATED)
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| LIST OF EFFECTIVE PAGES GENERAL TABLE OF CONTENTS PAGE NUMBER 1 li ill iV V Vi Vii V3.3.'1 Viiia Viiib 3.X X Xl X3.i Xiii X3.V XV XVl XVl 1 XVlil XiX XX XXi XX3.1 XXiii XXiV XXV XXVi XXVll, XXV3.1 i XX3.X XXX XXXi XXxii XXX 3.3.3.XXX 3.V XXXV XXXVl.XXXV3.i XXXViii XXXiX xl Xli xiii xliii T=TABLE F=FIGURE NUMBER REVISION NUMBER 12 12 12 12 13 12 12 13 13 13 13 12 13 12 13 13 13 12 12 13 12~12 12 13-c)13 12 12 12 12 12 12 PAGE NUMBER xliv Xlv xlvi xlvii xlviii il l T=TABLE F=FIGURE NUMBER REVISION NUMBER 12 12~13 12 12 12.12 UFSAR Revision 13 EP i June 1995&
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| 0 NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES GENERAL TABLE OF CONTENTS PAGE NUMBER I-1 I-2 I-3 I-4 I-5 I-6 I-7 I-8 I-9 I-10 I-11 I-12 I-13 I-14 I-15 I-16 I-17 T=TABLE F=FIGURE NUMBER F I-1 REVISION NUMBER 0 9 9 0 11 0 0 0 0 13 13 13 12 0 0 0 3 PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER UFSAR Revision 13 EP 1-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES GENERAL TABLE OF CONTENTS T=TABLE.PAGE F=FIGURE NUMBER NUMBER REVISION NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER II-1 II-2 II-3 II-4 II-5 II-6 II-7 II-8 II-9 II-10 II-11 II-12-II-13 II-14 II-15 II-16 II-17 II-18 II-19 II-20 II-21 II-22 II-23 II-24 II-25 F II-1 F II-2 F II-3 T II-1 F II-4 F II-5 T II-2 F II-6 T II-3 T II-4 T II-5 T II-6 T II-7 T II-8 T II-8 T II-8 13 0 0 6 13 0 0 0 0 0 13 0 13 6 6 6 10 0 0 0 9 3 3 0 4 UFSAR Revision 13 EP 2-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION III PAGE NUMBER III-1 III-2 III-3 III-4 III-5 III-6 III-7 III-8 III-9 III-10.III-11 III-12 III-13 III-14 III-15 III-16 III-17 III-18 III-19 III-20 III-21 III-22 III-23 III-23a III-23b III-24 III-25 III-26 III-27 III-28 III-29 III-30 III-31 III-32 III-33 III-34 III-35 III-36 III-37 III-38 III-39 III-40 III-41 III-42 III-43 T=TABLE'=FIGURE NUMBER F III-1 F III-2 F III-3 F III-4~F III-5 F III-6 F III-7 F III-8 F III-9 F III-10 F III-11 F III-12 F III-13 F III-14 F III-15 F III-16" F III-17 REVISION NUMBER 9 0 13 1 9 0 6 12 12 12 12 6 6 6 2 2 5 1 13 1 13 13 12 12 12 , 1 10 11 12 10 0 0 0 0 2 0 i 2 0 11 0 0 0 0 0 9 PAGE NUMBER III-44 III-45 III-46 III-47 III-47a III-48 III-48a III-48b III-49 III-50 III-51 III-52 III-53 III-54 III-54a III-54b III-55 III-56 III-57 III-58 III-59 III-60 III-61 III-62 III-63 III-64 III-65 III-66 III-67 III-68 III-69 III-70 III-71 III-72 III-73 III-74 III-75 T=TABLE F=FIGURE NUMBER F III-18 F III-19 F III-20 F III-21 F III-22 F III-23 REVISION NUMBER 9 13 12 12 3 12 10 10 10 0 0 0 0 12 12 12 0 0 12 0 0 12 10 0 13 0 0 0 0 0 12 12 12'2 12 12 12 UFSAR Revision 13 EP 3-1 June 1995 0 h NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION IV PAGE NUMBER IV-1 IV-2 IV-3 IV-4 IV-5 IV-6 IV-7 IV-8 IV-9 IV-10 IV-11 IV-12 IV-13 IV-14 IV-15 IV-16 IV-17 IV-18 IV-19 IV-19a IV-20 IV-21 IV-22 IV-23 IV-24 IV-25 IV-26 IV-27 IV-28 IV-29 IV-30 IV-3 1 IV-32 IV-33 IV-34 IV-35 IV-35a IV-35b IV-36 IV-37 IV-38 IV-39 IV-40 IV-40a IV-40b T=TABLE F=FIGURE NUMBER F IV-1 F IV-2 F IV-3 F IV-4 F IV-5 F IV-6 F IV-7 F IV-8 REVISION NUMBER 12 13 6 6 13 12 0 13 0 13 13 13 13 12 12 0 0 0 i13 8 0 8 13 13 12 13 6 13 13 13 13 13 13 10 13 11'11 10 0 0 0 10 13 12 ,12 PAGE NUMBER IV-4 1 IV-42 IV-43 IV-44 IV-45 IV-46 IV-46a IV-46b IV-47 IV-48 IV-49 IV-50 IV-51 IV-52 T=TABLE F=FIGURE NUMBER F IV-9 REVISION NUMBER 0 1 0 0 0 10 10 10 0 0 0 0 0 0 UFSAR Revision 13 EP 4-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION V T=TABLE PAGE F=FIGURE NUMBER NUMBER REVISION NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER V-1 V-2 V-3 V-4 V=5 V-6 V-7 V-8 V-9 V-10 V-11 V-12 V-12a V-13.V-14 V-15 V-16 V-17 V-18 V-19 V-20 V-21 V-22 V-23 V-23a V-23b V-24 V-25 V-26 V-27 V-28 V-29 V-30 V-30a V-31 V-31a V-31b V-32 V-33 T V-1 T V-1 T V-1 T V-2 T V-3, T V-4 F V-1 T V-5 F V-2 F V-3 F V-4 F V-5 F V-6'F V-7 F V-8 0 0 11 8 10 0 0 0 0 0 11 1 1 8 9 0 0 13 13 10 0 11 11 11 11 11 11 11 0 0 12 9 4 4 12 12 12 0 3 UFSAR Revision 13 EP 5-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION VI PAGE NUMBER VI-1 VI-2 VI-3 VI-4 VI-5 VI-6 VI-7 VI-8 VI-9 V.I-10 VI-11 VI-12 VI-13 VI-14 VI-15-VZ-16 VI-17 VI-18 VI-19 VI-19a VI-20 VI-21 VI-22 VI-22a VI-22b VI-23 VI-24 VI-25 VI-26 VI-27 VI-28 VI-29 VI-30 VI-31 VI-32 VI-33 VI-34 VI-35 VI-36 VI-37 VI-38 VI-39 VI-40 VI-41 VI-42 T=TABLE F=FIGURE NUMBER F VI-1 F VI-2 F VI-3 F VI-4 F VI-4a F VI-5 F VI-6 F VI-7 F VI-8 F VI-9 F VI-10 F VI-11 F VI-12 F VI-13 F VI-14 F VI-15 F VI-16 F VI-17 F VI-18 F VI-19 REVISION NUMBER 0 4 11 11 11 0 0 9 0 10 0 0.0 0 0 0.PAGE NUMBER VI-43 VI-43a VI-43b VI-44 VI-45 VI-46 VI-47 VI-48 VI-48a VI-49 VI-50 VI-50a VI-50b VI-51 VI-52 VI-53 VI-54 VI-55 VI-56 VI-56a VI-56b VI-57 VI-58 VI-59 VI-60 VI-61 VI-62 VI-63 VI-64 VI-65 VI-66 VI-67 VI-68 VI-69 VI-70 T=TABLE F=FIGURE NUMBER T VI-1 T VI-1 T VI-2 T VI-3a T VI-3a T VI-3a T VI-3b T VI-3b T VI-3b T VI-3b T VI-3c F VI-20 F VI-21 T VI-4 F VI-22 F VI-23 F VI-24 T VI-5 T VI-5 REVISION NUMBER 11 11 11 11 11 11 13 13 13 11 11 13 11 11 0 0 0 12 11 13 13 3 11 12 0 0 0 11 12 13 13'3 13 0 0 UFSAR Revision 13 EP 6-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION VII PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER VII-1 VII-2 VII-3 VII-4 VII-5 VZI-,6 VII-6a VII-7 VII-8 VII-9 VII-10 VII-11 VII-11a VII-11b VII-12 VII-13 VIZ-13a VII-14 VII-15 VII-16 VII-17 VII-18 VII-18a VII-18b VII-19 VII-20 VII-21 VII-21a VII-21b VII-22 VII-23 VII-24 VII-25 VII-26 VII-26a VII-26b VII-27 VII-28 VII-29 VII-30 VII-30a VII-30b VII-31 VII-32 VII-33'VII-1 F VZI-2 F VII-3 F VII-4 F VII-4a F VII-5 F VII-6 F VIZ-7 F VII-8.F VII-9 9 10 10 8 9 8 10 6 8 8 0 13 13 10 11 9 10 9 9 10 9 12 10 10 13 13 11 11 11 10 7 0 7 11 10 10 0 11 0 10 10 10 10 10 10 VZI-33a VIZ-33b VII-34 VII-35 VII-36 VZI-37 VZI-38 VII-39 VII-40 VII-41 VII-42 VII-43 VII-44 VII-45 VII-46 VZI-46a VII-46b VII-47 VII-48 VII-49 VII-53 VII-54 VIZ-55 VII-55a VZI-55b VZI-56 VII-57 VZZ-58 VZI-59 VII-60 VII-61 VII-61a VII-62 VII-62a VII-63 T VII-1 F VII-10 F VIZ-11 F VII-12 F VII-16 F VII-17 VII-50 F VII-13 VII-51 F VII-14 VII-52 F VII-15 10 10 0 0 0 0 0 0 9 0 3 0 0 0 12 12 12 12 0 3 9 0 0 0 0 12 12 12 0 0 0 0 0 12 9 8 8 9 UFSAR Revision 13 EP 7-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION VIII PAGE NUMBER VIII-1 VIIX-2 VIII-3 VIII-4 VIII-5 VIII-6 VIII-7 VIIZ-8 VIII-9 VIII-10 VIII-11 VIII-12 VIII-13 VIII-14 VIII-15 VIIZ-16 VIII-17 VIII-18 VIII-19 VIII;20 VIII-21 VIII-22 VIII-23 VIII-24 VIII-25 VIII-26 VIII-27 VIII-28 VIII-29 VIII-30 VIII-31 VIII-32 VIII-33 VIII-34 VIII-35 VIII-36 VIII-37 VIII-38 VIII-39 VIII-40 VIII-40a VIII-41 VIII-42 VIII-43 VIII-44 T=TABLE F=FIGURE NUMBER F VIII-1 F VIII-2 F VIXI-3 F VIII-4 F VIXI-5 F VIII-6 F VIIZ-7 F VIII-8 F VIIZ-9 F VIII-10 F VIII-11 F VIII-12 F VIXI-13 F VIII-14 F VIII-15 REVISION NUMBER 12 12 5 9 5 7 0.13 13 0 11 7 13 7 4 5 0 0 0 0 4 0 0 0 0 8 0'8 10 0 0 0 0 0 10 0 0 12'1 8 8 0 0 0 0 PAGE NUMBER VIII-45 VIII-46 VIII-47 VIII-48 VIII-49 VIII-50 VIII-51 VIII-52 VIII-53 VIII-54 VIII-55 VIII-56 VIII-57 VIII-58'III-59 VIIX-60 VIII-61 VIII-62 VIII-62a VIII-62b VIII-63 VIII-64 VIII-65 VIII-66 VIII-67 VIII-68 VIII-69, VIII-70 VIII-71 VIII-71a VIII-71b VIII-72 VIII-72a VIII-72b VIII-73 VIII-74 VIII-75 VIII-76 VIII-77 VIIZ-78 VIIZ-79 VIIX-80'III-81 VIII-81a VIII-81b T=TABLE F=FIGURE NUMBER F VIII-16 F VIII-17 F VIII-18 F VIII-19 F VIII-20 F VIII-21 F VIII-22 F VIII-23 F VIII-24 F VIII-25 F VIII-26 F VIII-27 F VIII-28 F VIII-29 REVISION NUMBER 0 0 0 0 0 11 0 0 0 0 0 13 13 13 13 13 13 13 13 13 0 0 10 10 0~4 3 10 12 12 12 10 10 10 10 0 3 0 0 0 0 9 12 12 12 UFSAR Revision 13 EP 8-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION VIII PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER i NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER VIII-82 VIII-82a VIII-82b VIII-82c VIII-826 VIII-83 VIII-84 VIII-85 VIII-86 VIII-87 VIII-88 VIII-89 VIII-90 VIII-91 VIII-92 VIII-93 VIII-94 VIII-95 VIII-96 VIII-97 VIII-98 VIII-98a VIII-98b VIII-99 VIII-100 VIII-101 VIII-102 VIII-103 VIII-104 VIII-105 VIII-106 VIII-107 VIII-108 VIII-109 VIII-110 VIII-111 VIII-112 T VIII-1 T VIII-2 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 T VIII-3 9 9 12 12 12 12 12 12 12 13 12 12 12 12 12 12 9 12 12 12 13 13 13'12 12 12 12 12 12 12 12 12 12 12 12 12 13 UFSAR Revision 13 EP 8-2 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES, SECTION IX PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER IX-1 IX-2 IX-2a IX-2b IX-3 IX-4 IX-4a IX-4b IX-5 IX-6 IX-7 IX-8 IX-8a IX-8b IX-9 IX-10 IX-11 IX-12 IX-13 IX-14 IX-15 IX-16 IX-16a ZX-16b IX-17 IX-18 IX-19 IX-20 IX-20a IX-20b IX-21 IX-22 IX-23 IX-24 IX-24a IX-24b IX-25 IX-26 IX-27 IX-28 IX-29 IX-30 IX-31 IX-31a IX-31b F IX-1 F IX-2 F IX-3 F IX-4 F IX-5 F IX-6 F IX-7 0 11 11 11 6 12 11 11 0 6 0 13 13 13 12 12 6 12 0 12 12 12 12 12 10 10 12 12 10 10 6 6 6 13 13 11 13 8 13 12 12 12 12 12 12 IX-32 IX-33 IX-33a IX-34 IX-35 IX-35a IX-35b IX-36 IX-37 IX-38 IX-39 IX-40 T IX-1 T IX-1 T IX-1 12 12 12 0 12 12 12 12 12 11 11 11 UFSAR Revision 13 EP 9-1 June 1995 1
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION X PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER X-1 X-2 X-3 X-4 X-5 X-6 X-7 X-8 X-9 X-10 X-11 X-12 X-13 X-14 X-15 X-16 X-'17 X-18 X-19 X-20 X-21 X-21a X-21b X-22 X-23 X-24 X-25 X-25a X-25b X-26 X-27 X-28 X-29 X-30 X-3 1 X-32 X-.3 3 X-34 X-35 X-36 X-37 X-38 X-39 X-39a X-39b F X-1 F X-2 F X-3 F X-4 F X-5 F X-6 9 8 0 12 0 0 13 13 0 0 0 0 7 9 0 0 0 0 0 13 11 11 10 10 10 10 10 10 9 0 1 11 0 11 12 11 9 12 12 12 10 13 13 12 X-40 X-41 X-42 X-43 X-44 X-45 X-46 X-47 X-48 X-49 X-50 X-51 X-52 X-52a X-52b X-53 X-54 X-55 X-55a X-55b X-56 X-57 X-58 X-59 X-60 X-61 X-62 X-63 X-64 X-65 X-66 X-67 X-68 X-69 X-70 X-71 X-72 X-73 X-74 X-75 X-76 X-77 X-78 X-79 F X-7 F X-8 F X-9 F X-10 F X-11 4 11 10 0 0 11 13 11 13 13 13 7 13 13 13 7 13 13 13 10 0 11 11 10 10 10 11, 10 10 10 10 10 10 10 10 10 10 10 10 10 1 3 5 5-UFSAR Revision 13 EP 10-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION X PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER 10A 10A-i 10A-ii 10A-iii 10A-iv 10A-v 10A-vi 10A-vii 10A-viii.'10A-ix 10A-1 10A-2 10A-3 10A-4 10A-5 10A-6 10A-7 10A-8 10A-9 10A-10 10A-11 10A-12 10A-13 10A-14 10A-15 10A-16 10A-17 10A-18 10A-19 10A-20 10A-21 10A-22 10A-23 10A-24 10A-24a 10A-24b 10A-25 10A-26 10A-27 10A-28 10A-29 10A-30.10A-31 10A-32 10A-33 T 2'T 2.5'.1-1 10 10 13 10 10 13 10 12 12 10 12 10 10 12 13 13 12 12 12 11 12 11 12 10 10 10 10 10 11 11 10 10 10 13.13 13 10 12 10 10 10 10 10 12 10 10A-34 10A-35 10A-36 10A-37 10A-38 10A-39 10A-40 10A-41 10A-41a 10A-41b 10A-42 10A-43 10A-44 10A-45 10A-46 10A-46a 10A-46b 10A-47 10A-48 10A-49 10A-50 10A-51 10A-52 10A-52a 10A-52b 10A-53 10A-54.10A-55 10A-55a 10A-55b 10A-56 10A-57 10A-58 10A-59 10A-60 10A-61 10A-61a 10A-61b 10A-62 10A-63 10A-64 10A-65 10A-66 10A-67 10A-68 T 2'''-2 T 2.5.1'-3 T 2.5''-4 T 2'''-5 T 2'''-6 T 2'''-7 T 2.5~3~4-1 T 2 5.3'-1 10 10 12 10 10 10 10 13 13 13 10 10 10 10 13 13 13 10 10 10 10 10 13 13 13 10 10 13 13 13 12 10 10 10 10 12 12 12 10 10 10 10 10 10 10 UFSAR Revision 13 EP 10-2 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION X PAGE NUMBER 10A-69 10A-70 10A-71 10A-72 10A-73 10A-74 10A-75 10A-76 10A-77 10A-78 10A-78a 10A-78b 10A-79 10A-80.10A-81 10A-82 10A-83 10A-83a 10A-83b 10A-84 10A-85 10A-86 10A-87 10A-88 10A-89 10A-90 10A-91 10A-92 10A-93 10A-94 10A-95 10A-96 10A-97 10A-98 10A-99 10A-100 10A-101 10A-102 10A-103 10A-104 10A-105 10A-106 10A-107 10A-107a 10A-107b T=TABLE F=FIGURE NUMBER T 1'.2 T 1'.2 T 1.2.2 T 1.2.2 T 1.2.2 T 1.2.2 T 1'.2 T 3.2-1 T 3.3-1 T 3.3-1 T 3.4-1 T 3.5-1 T 3.6-1 T 3.7-1 T 3.8-1 T 3.9-1 T 3.10-1 T 3.10-1 T 3 a 1 1 T 3~1 1 T 3'-1 T 3.1-1 REVISION NUMBER 10 12 10 10 10 10 10 10 10 13 13 13 10 10 10 12 12 12 12 10 10 10 12 10 10 13 13 13 13 13 10 12 10 10 10 10 10 10 10 10 10 10 12 12 12 PAGE NUMBER 10A-108 10A-109 10A-110 10A-111 10A-112 10A-113 10A-114 10A-115 10A-116 10A-117 10A-118 10A-119 10A-120 10A-121 10A-122 10A-123 10A-124 10A-125 10A-126 10A-127 10A-128 10A-129 10A-130 Legend B-40141-C Overlay Overlay B-40142-C Overlay Overlay Overlay Overlay B-40143-C Overlay Overlay Overlay Overlay B-40144-C Overlay Overlay Overlay Overlay T=TABLE F=FIGURE NUMBER T 3.1.1-1 T 3'.1-1 T 3.1.1-1 T 3'.1-1 T 3.1'-2 T 3.1'-2 T 3.1.1-2 T 3'.1-2 T 3'.1-2 T 3.1'-2 T 3.1.1-3 T 3.1.1-4 T 3.1.1-4 T 3'.1-5 T 3'.1-6 T 3'.1-6 T 3'.1-7 T 3'.1-8 T 3.1.1-8 T 3'.1-9 T 3'.1-9 T 3.1.1-9 T 3'.1-9 1-1 2-1 1-2 2-2 3-2 4-2 1-3 2-3 3-3 4-3 1-4 2-4 3-4 4-4 REVISION NUMBER'2 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 10 11 12 12 12 UFSAR Revision 13 EP 10-3 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION X PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER , NUMBER PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER B-40145-C Overlay Overlay Overlay B-40146-C Overlay Overlay Overlay B-40147-C Overlay Overlay Overlay B-40148-C Overlay Overlay Overlay 10B 10B-i 10B-ii 10B-iii'10B-1 10B-2 10B-3 10B-4 10B-5 10B-6 10B-7 10B-8 10B-9 10B-10 10B-11 10B-12 10B-13 10B-14 10B-15 10B-16 10B-17 10B-18 10B-19 10B-20 10B-21 10B-22 10B-23 10B-24 1-5 2-5 3-5 1-6 2-6 3-6 1-7 2-7 3-7 1-8 2-8 3-8 12 12 10 12 12'12 12 12 12~12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 10B-25 10B-26 10B-27 10B-28 10B-29 10B-30 10B-31 10B-32 10B-33 10B-34 10B-35 10B-36 10B-37 10B-38 10B-39 10B-40 10B-41 10B-42 10B-43 10B-44 10B-45 10B-46 10B-47 10B-48 10B-49 10B-50 10B-51 10B-52 10B-53 10B-54 10B-55 10B-56 10B-57 10B-58 10B-59 10B-60 10B-61 10B-62 10B-63 10B-64 10B-65 10B-66 10B-67 10B-68 10B-69 T 1 T 1 T 1 T 1 T 1 T 1 T 1 T 1 T 1 T 1 T 2A T 2B T 3 T 3 T 3 T 3 T 3 T 3 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 UFSAR Revision 13 EP 10-4 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION X PAGE NUMBER 10B-70 10B-71 10B-72 10B-73 10B-74 10B-75 10B-76 10B-77 10B-78 10B-79 10B-80 10B-81 10B-82 10B-83 10B-84 10B-85 10B-86 10B-87 10B-88 10B-89 10B-90 10B-91 10B-92 10B-93 10B-94 10B-95 10B-96 10B-97 10B-98 10B-99 10B-100 10B-101 10B-102 10B-103 10B-104 10B-105 10B-106 10B-107 10B-108 10B-109 10B-110 10B-111 10B-112 10B-113 10B-114 T=TABLE F=FIGURE REVISION NUMBER NUMBER 12 12 12 12 12 12 12 12 12 12 12 12 12 1,2 1'2 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 PAGE NUMBER 10B-115 10B-116 10B-117 10B-118 10B-119 10B-120 10B-121 10B-122 10B-123 10B-124 10B-125 10B-126 10B-127 10B-128 10B-129 10B-130 10B-131 10B-132 10B-133 10B-134 10B-135 10B-136 10B-137 10B-138 10B-139 10B-140 10B-141 10B-142 10B-143 10B-144 10B-145 10B-146 10B-147 10B-148 10B-149 10B-150 10B-151 10B-152 10B-153 10B-154 10B-155 10B-156 10B-157-10B-158 10B-159 T=TABLE F=FIGURE NUMBER REVISION NUMBER 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12'12 12 12 12'2 12 12 12 12 UFSAR Revision 13 EP 10-5 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION X PAGE NUMBER 10B-160 10B-161 10B-162 10B-163 10B-164 10B-165 10B-166 10B-167 10B-168 10B-169 10B-170 10B-171 10B-172 10B-173 10B-174 10B-175 10B-176 10B-177 10B-178 10B-179 10B-180 10B-181 10B-182 10B-183 10B-184 10B-185 10B-186 10B-187 10B-188 10B-189 10B-190 10B-191 10B-192 10B-193 10B-194 10B-195 10B-196 10B-197 10B-198 10B-199 10B-200 10B-201 10B-202 10B-203 10B-204 T=TABLE F=FIGURE NUMBER REVISION NUMBER 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 PAGE NUMBER 10B-205 10B-206 10B-207 10B-208 10B-209 10B-210 10B-211 10B-212 10B-213 10B-214 10B-215 10B-216 10B-217 10B-218 10B-219 10B-220 10B-221 10B-222 10B-223 T=TABLE F=FIGURE NUMBER REVISION NUMBER 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 12 UFSAR Revision 13 EP 10-6 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XI PAGE NUMBER XI-1 XI-2 XI-3 XI-4 XI-5 XI-6 XI-7 XI-8 XI-9 XI-10 XI-11 XI-12 XI-13 XI-14 XI-15 XI-16 XI-17 XI-18 XI-19 XI-20 XI-21 XI-22 XI-23 XI-24 XI-25 T=TABLE F=FIGURE NUMBER F XI-1 F XI-2 F XI-3-F XI-4 F XI-5 F XI-6 F XI-7 REVISION NUMBER 0 0 1 13 2 2 0 12 10 8 8 9 0 0 0 0 13 0 0 0 0 0 13 0 0 T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER UFSAR Revision 13 EP 11-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XII PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER'UMBER XIZ-1 XII-2 XII-3 XII-4 XII-5 XII-6 XII-7 XII-8 XII-9 XII-10 XII-11 XIZ-12 XII-13 XII-14 XII-15 XII-16.XII-17 XII-18 XII-19 XII-20 XII-21 XII-22 XII-23 XII-24 XII-25 XII-26 XII-27 XII-28 XII-29 XII-30 XII-31 XII-32 XII-32a XII-32b XII-33 XII-34'II-35 XII-36 XII-37 XII-38 XII-39 XII-40 XII-41 XII-41a XII-41b T XII-1 T XII-2 T XII-2 T XII-3 F XII-1 T XII-4 T XII-5 T XII-6 T XII-7 T XII-8 T XII-8 0 0 0 0 0 0 0 6 13 6 6 6 13 6 13 13 13 9 6 0 0 0 0 0 0 11 11 11 0 11 0 12 12 12 12 12'12 12 13 12 0 12 13 13 11 XII-42 XII-43 XII-44 XII-45 XII-46 13 13 13 13 13 UFSAR Revision 13 EP 12-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XIII T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER XIII-1 XIII-2 XIII-3 XIII-4 XIII-5 XIII-6 XIII-6a XIII-6b XIII-7 XIII-8 XIII-9 XIII-10 XIII-11 XIII-12 XIII-13 XIII-14 XIII-15 XIII-16 XIII-17 XIII-18 XIII-19 XIII-20 XIII-21 XIII-22 XIII-23 XIII-24 T XIII-1 F XIII-1 F XIII-2 V XIII-3 F XIII-4 F XIII-5 12 13 13 12 12 13 13 13 12 12 12 12 12 12 12 12 12 13 13 13 13 12 12 12 12 12 13 13 13 13 13 13 UFSAR Revision 13 EP 13-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XIV PAGE NUMBER XIV-1 XIV-2 XIV-3 XIV-4 XIV-5 XIV-6 XIV-7 XIV-8 XIV-9 XIV-10 XIV-11 XIV-12 XIV-13 XIV-14 XIV-15 T=TABLE F=FIGURE NUMBER REVISION NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER UFSAR Revision 13 EP 14-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XV PAGE NUMBER XV-2 XV-3 XV-4 XV-5 XV-5a XV-5b XV-6 XV-7a XV-7b XV-8 XV-8a XV-8b XV-9 XV-10 XV-11 XV-11a XV-12 XV-13 XV-14 XV-15 XV-16 XV-17 XV-18 XV-19 XV-20 XV-2 1 XV-22 XV-23 XV-24 XV-25 XV-26 XV-27 XV-28 XV-29 XV-30 XV-3 1 XV-32 XV-3 3 XV-3 4 XV-35 XV-3 6 XV-37 XV-38 T=TABLE F=FIGURE NUMBER F XV-1 T XV-1 T XV-1 T XV-2 F XV-2 F XV-2a F XV-2b F XV-3 F XV-4 T XV-3 F XV-5 F XV-6 F XV-7 F XV-8 F XV-9 F XV-10 F XV-11 F XV-12 F XV-13 F XV-14 REVISION NUMBER 13 0 12 0 12 11 11 13 13 13 12 13 6 6 13 6 13 13 13 13 13 6 13 6 13 13 0 13 13 0 0 0 0 0 0 0 0 0 13 12 13 PAGE NUMBER XV-3 9 XV-4 0 XV-4 1 XV-42 XV-43 XV-44 XV-45 XV-4 6 XV-47 XV-48 XV-49 XV-50 XV-50a XV-50b XV-51 XV-52 XV-52 a XV-53 XV-54 XV-55 XV-56 XV-57 XV-58 XV-59 XV-59a XV-60 XV-6 1 XV-62 XV-63 XV-64 XV-65 XV-66 XV-67 XV-68 XV-69 XV-7 0 XV-7 1 XV-72 XV-73 XV-74 XV-75 XV-76 XV-77 XV-78 XV-79 T=TABLE F=FIGURE NUMBER F XV-15 F XV-16 F XV-17 F XV-18 F XV-19 F XV-20 F XV-21 F XV-22 F XV-23 F XV-24 T XV-4 T XV-4 T XV-5 F XV-25 T XV-6 T XV-7 T XV-8 REVISION NUMBER 13 3 3 0 0 0 0 9 0 9 0 12 12 12 0 9 8 0 9 0 0 0 0 1 1 8 9 1 0 9 13 7 0 0 0 8 8 0 0 0 0 0 0 0 8 UFSAR Revision 13 EP 15-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XV PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER XV-80 XV-8 1 XV-81a XV-8 1b XV-82 XV-82 a XV-82b XV-83 XV-84 XV-85 XV-86 XV-87 XV-88 XV-89 XV-90 XV-9 1 XV-92 XV-93 XV-94 XV-94a XV-95 XV-96 XV-97 XV-98 XV-99 XV-100 XV-101 XV-102 XV-103 XV-104 XV-105 XV-106 XV-107 XV-108 XV-109 XV-110 XV-111 XV-112 XV-113 XV-114 XV-115 XV-116 XV-117 XV-118 XV-119 T T T XV-9 XV-9 XV-9A F F XV-26 XV-27 F F F F XV-28 XV-29 XV-3 0 XV-3 1 XV-10 XV-11 T T F F F F F F F F F F F F F F F F F XV-12 XV-13 XV-34 XV-3 5 XV-3 6 XV-37 XV-38 XV-3 9 XV-40 XV-4 1 XV-42 XV-43 XV-44 XV-45 XV-4 6 XV-47 XV.-48 XV-49 XV-50 F XV-32 F, XV-33 XV-33a 8 12 12 12 9 12 12 8 8 8 8 8 8 8 8 8 8 8 8 6.8 12 8'8 8 XV-120 XV-12 1 XV-122 XV-123 XV-124 XV-125 XV-12 6 XV-127 XV-128 XV-129 XV-13 0 XV-13 1 XV-132 XV-133 XV-134 XV-135 XV-13 6 XV-137 XV-137a XV-137b XV-137c XV-137d XV-137d1 XV-137d2 XV-137e XV-137 f XV-137g XV-137h XV-137i XV-137)XV-137k XV-138 XV-139 XV-139a XV-139b XV-140 XV-141 XV-142 XV-143 XV-144 XV-145 F XV-51 F XV-52 F XV-53 F XV-54 T XV-14 T XV-15 F XV-55 F XV-56 T XV-16 T XV-17, T XV-18 T XV-19, T XV-20, T XV-21 T XV-21A T XV-21B T XV-21C T XV-21D, T XV-21E F XV-56a F XV-56b F XV-56c T XV-22 T XV-23 8 8 8 8 12 9 9 12 8 8 8 8 8 8 12 8 12 12 12 12 12 12 12 12 12 8 12 12 8 12 12 12 0 0 12 0 0 0 UFSAR Revision 13 EP 15-2 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION,XV PAGE NUMBER XV-146 XV-147 XV-148 XV-149 XV-150 XV-151 XV-152 XV-153 XV-154 XV-155 XV-156 XV-157 XV-158 XV-159 XV-159a XV-159b XV-159c XV-159d XV-159e XV-159f XV-159g XV-159h XV-159i XV-159j XV-159k XV-159L XV-159m XV-159n XV-159o XV-159p XV-159'V-159r XV-159s XV-159t XV-159u XV-159v XV-159w XV-159x XV-160 XV-160a XV-160b XV-161 XV-162 XV-163 T=TABLE F=FIGURE NUMBER T XV-24 T XV-25 T XV-26, T XV-27 T XV-28 T XV-29 F XV-56D T XV-29A F XV-56E F XV-56F F XV-56G F XV-56H T XV-29B T XV-29B T XV-29C T XV-29D F XV-57 F XV-58 REVISION NUMBER 12 8 8 8 0 0 0 0 0 0" 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 11 10 10 11 0 0 PAGE NUMBER XV-164 XV-164a XV-164b XV-165" XV-166 XV-167 XV-168 XV-169 XV-169a XV-169b XV-169c XV-169d XV-169e XV-169f XV-169g XV-169h XV-169i XV-169>XV-169k XV-169L XV-169m XV-170 XV-171 XV-172 XV-173 XV-174 XV-175 XV-176 XV-177 XV-178 XV-179 XV-180 XV-181 XV-182 XV-183 XV-184 XV-185 XV-186 XV-187 XV-188 XV-189 XV-190 XV-191 XV-192 T=TABLE F=FIGURE NUMBER F XV-59 F XV-60 T XV-30 T XV-31, T XV-32 T XV-32a F XV-60a F XV-60b T XV-33 T XV-34 T XV-35 F XV-61 F XV-62 F XV-63 F XV-64 F XV-65 F XV-66 F XV-67 F XV-68 F XV-69 REVISION NUMBER 11 10 0 10 11 11 10 11 11 11 10 11 11 11 10 10 13 10 10 8 13 0 0 0 0 0 0 0 0 0 0 0 0 0 10 0 0 0 0 0 0 0 UFSAR Revision 13 EP 15-3 June 1995 0
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XV PAGE NUMBER XV-193 XV-194 XV-195 XV-196 XV-197 XV-198 XV-199 XV-200 T=TABLE F=FIGURE NUMBER F XV-70 F XV-71 T XV-36 F XV-72 REVISION NUMBER 0 0 0 10 0 0 0 0 T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER UFSAR Revision 13 EP 15-4 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XVI PAGE NUMBER XVI-1 XVI-2 XVI-3 XVI-4 XVI-5 XVI-6 XVI-7 XVI-8 XVI-9 XVI-10 XVI-11 XVI-12 XVI-13 XVI-14 XVI-15 XVI-16 XVI-17 XVI-18 XVI-19 XVI-2 0 XVI-2 1 XVI-22 XVI-23 XVI-24 XVI-25 XVI-26 XVI-27 XVI-28 XVI-29.XVI-3 0 XVI-3 1 XVI-32 XVI-3 3 XVI-34~XVI-35 XVI-3 6 XVI-37 XVI-38 XVI-39 XVI-40 XVI-4 1 XVI-42 XVI-43 XVI-44 XVI-45 T=TABLE F=FIGURE NUMBER T XVI-1 T XVI-1 T XVI-2 T XVI-2 T XVI-2 T XVI-3 F XVI-1 F XVI-2 F XVI-3 F XVI-4 F XVI-5 F XVI-6 F XVI-7 F XVI-8 F XVI-9 F XVI-10 F XVI-11 T XVI-4 F XVI-12 T XVI-5 T XVI-6 T XVI-7 T XVI-8 T XVI-9 REVISION NUMBER 11 0 0 0 0 0 9 0 0 9 0 3 3 0 3 3 3 3 3 3 0 0 9 0 0)0 0 0 0 0 0 0 0 0 0 0'0 0 0 0 0 0 0 0 0 PAGE NUMBER XVI-4 6 XVI-46a XVI-46b XVI-47 XVI-48 XVI-49 XVI-50 XVI-51 XVI-52 XVI-53 XVI-54 XVI-55 XVI-56 XVI-57 XVI-58 XVZ-59 XVI-60 XVI-6 1 XVI-62 XVI-63 XVI-67 XVI-68 XVI-69 XVI-70 XVI-71 XVI-72 XVI-73 XVI-74 XVI-75 XVI-76 XVI-77 XVI-78 XVI-79 XVI-80 XVI-81 XVI-82 XVI-83 XVI-84 XVI-85 XVI-86 XVI-87 XVI-88 F XVI-13 T XVI-10 T XVI-11 T XVI-12 F XVI-14 F XVI-15 XVI-64 F XVI-16 XVI-65 T XVI-13 XVI-66 F XVI-17 F XVI-18 F XVI-19 F XVI-20 F XVI-21 F XVI-22 F XVI-23 F XVI-24 T XVI-14 0 13 13 0 0 0 0 11.0 0 0 0 0 0 10 0 10 0 0 0 0 9 0 0 0 0 0 0 0 0'0 0 0 0 0 0 0 0 0 0 0 0 0 0 T=TABLE F=FIGURE REVISION NUMBER NUMBER~UFSAR Revision 13'P 16-1 June 1995
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| I NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XVI PAGE NUMBER T=TABLE F=FIGURE NUMBER REVISION NUMBER PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER XVI-89 XVI-90 XVI-9 1 XVI-92 XVI-93 XVI-94 XVI-95 XVI-96 XVI-97 XVI-98 XVI-99 XVI-100 XVI-101 XVI-102 XVI-103 XVI-104 XVI-104a XVI-104 b XVI-105 XVI-106 XVI-107 XVI-108 XVI-109 XVI-110 XVI-111 XVI-112 XVI-113 XVI-114 XVI-114a T XVI-15 F XVI-25 T XVI-16 F XVI-26 F XVI-27 T XVI-17 T XVI-18 XVI-116 XVI-117 XVI-118 XVI-119 XVI-120 XVI-121 XVI-122 XVI-123 XVI-124 XVI-125 XVI-126 XVI-127 XVI-128 XVI-129 F XVI-29 F XVI-30 F XVI-31 F XVI-32 F XVI-33 T XVI-19 T XVI-20 T XVI-21 T XVI-22 XVI-114b XVI-115 F XVI-28 0 0 0 0 0 0 0 0 10 10 10 0 0 0 0 0 0 0 0 0 10 10 10 0 0 0 0 0 0 0 12 0.0'a 0 XVI-130 XVI-131 XVI-132 XVI-133 XVI-134 XVI-135 XVI-136 XVI-137 XVI-138 XVI-139 XVI-140 XVI-141 XVI-142 XVI-143 XVI-144 XVI-145 XVI-146 XVI-147 XVI-148 XVI-149 XVI-150 XVI-151 XVI-152 XVI-153 XVI-154 XVI-155 T XVI-23 T XVI-24 T XVI-25 T XVI-26 F XVI-34 F XVI-35 F XVI-36 F XVI-37 F XVI-38 F XVI-39 F XVI-40 F XVI-41 F XVI-42 F XVI-43 F XVI-44 F XVI-45 T XVI-27 T XVI-27 XVI-156 XVI-157 XVI-158 XVI-159 XVI-160 XVI-161 XVI-162 XVI-163 XVI-164 XVI-165 XVI-166 XVI-167 XVI-168 XVI-169 XVI-170 XVI-171 XVI-172 T XVI-29 T XVI-30 F XVI-46 F XVI-47 F XVI-48 F XVI-49 XVI-155a T XVI-28 XVI-155b 9 0 0 9 0 0 0 0 0 0 0 0 9 0 0 0 0 ,0 0 0 0 0 0 0 0 12 12 12 0 0 0 0, 0 0 0 0 12 0 0 0 0 9 9 0 0 UFSAR Revision 13 EP 16-2 June 1995 0
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF'EFFECTIVE PAGES SECTION XVI PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER T=TABLE PAGE F=FIGURE NUMBER NUMBER REVISION NUMBER XVI-173 XVI-174 XVI-175 XVI-176 XVI-177 XVI-178 XVI-179 XVI-180 XVI-181 XVI-182 XVI-183 XVI-184 XVI-185 XVI-186 XVI-187 XVI-188 XVI-189 XVI-190 XVI-191 XVI-192 XVI-193 XVI-194 XVI-195 XVI-196 XVI-197 XVI-198 XVI-199 XVI-200 XVI-201 XVI-202 XVI-203 XVI-204 XVI-205 XVI-206 XVI-207 XVI-208 XVI-209 , XVI-210 XVI-211 XVI-212 XVI-213 XVI-214 XVI-215 XVI-216 XVI-217 F XVI-50 F XVI-51 F XVI-52 F XVI-53 F XVI-54 F XVI-55 F XVI-56 F XVI-57 F XVI-58 F'XVI-59 F XVI-60 F XVI-61 T XVI-31 T XVI-31 EXHIBIT 1 EXHIBIT 1 EXHIBIT 1 EXHIBIT 1 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 2 EXHIBIT 3 EXHIBIT 3 EXHIBIT 4 EXHIBIT 4EXHIBIT 5 EXHIBIT 6 EXHIBIT 7, EXHIBIT 8 EXHIBIT 8 EXHIBIT 8 EXHIBIT 8 EXHIBIT 9 EXHIBIT 9 EXHIBIT 10 XVI-218 XVI-219 XVI-220 XVI-221 XVI-222 XVI-223 XVI-224 XVI-225 XVI-226 XVI-227 XVI-228 XVI-229 XVI-230 XVI-231 XVI-232 XVI-233 XVI-234 XVI-235 XVI-236 XVZ-237 XVI-238 EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXH1BIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT EXHIBIT 10 0 10 0 11 0 11 0 12 0 12 0 12 0 13 0 13 0 0 14 0 14 0 14 0 14 0 14 0 14 0 14 0 14 0 14 0 14 0 9 UFSAR Revision 13 EP 16-3 June 1995 I
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XVII PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER PAGE NUMBER STABLE F=FIGURE NUMBER REVISION NUMBER XVII-1 XVII-2 XVII-3 XVII-4 XVII-5 XVII-6 XVII-7 XVII-8 XVII-9 XVII-10.XVII-11 XVII-12 XVII-13 XVII-14 XVII-15 XVII-16 XVII-17 XVII-18 XVII-19 XVII-20 XVII-21 XVII-22 XVII-23 XVII-24 XVII-25 XVII-26 XVII-27 XVII-28 XVII-29.XVII-30 XVII-31 XVII-32 XVII-33 XVII-34 XVII-35 XVII-36 XVII-37 XVII-38 XVII-39 XVII-40 XVII-41 XVII-42 XVII-43 XVII-44 XVII-45 F F F F F F F F F F F F F XVIZ-1 XVII-2 XVII-3 XVII-4 XVZI-5 XVII-6 XVI Z-7 XVII-8 XVII-9 XVZI-10 XVII-11 XVII-12 XVZI-13 F F F F F F T F F XVII-1 XVII-2 XVII-14 XVZZ-15 XVII-16 XVZI-17 XVII-18 XVII-19 XVII-3 XVII-20 XVII-21 F F F F F F F F F XVII-23 XVII-24 XVII-25 XVII-26 XVII-27 XVII-28 XVII-29 XVII-30 XVII-31 XVII-4 XVII-3 F XVII-22 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 XVII-46 XVII-47 XVII-48 XVZI-49 XVII-50 XVII-51 XVII-52-XVII-53 XVII-54 XVII-55 XVII-56 XVII-57 XVII-58 XVII-59 XVII-60 XVII-61 XVII-62 XVII-63 XVII-64 XVII-65 XVII-66 XVII-67 XVII-68 XVII-69 XVIZ-70 XVIZ-71 XVII-72 XVII-73 XVII-74 XVII-75 XVII-76 XVII-77 XVII-78 XVII-79 XVII-80 XVIZ-81 XVII-82 XVZI-83 XVII-84 XVII-85 XVII-86 XVIZ-87 XVII-88 XVII-89 XVII-90 F F F F F F T F F F F F F F T T T T'T T T T T T T T T T T T T XVII-33 XVII-34 XVII-35 XVII-36 XVII-37 XVII-38 XVII-39 XVZI-5 XVZI-40 XVIZ-41 XVII-42 XVII-43 XVII-44 XVII-45 XVII-46 XVII-6 XVII-7 XVII-8 XVII-9 XVII-10 XVII-11 XVII-12 XVII-13 XVII-14 XVII-15 XVII-16 XVII-17 XVII-18 XVZI-19 XVII-20 XVII-21 XVII-22 XVII-23 XVII-24 XVII-25 XVII-26 UFSAR Revision 13 EP 17-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XVII PAGE NUMBER XVII-9 1 XVII-92 XVII-93 XVII-94 XVII-95 XVII-96 XVII-97 XVII-98 XVII-99 XVII-10Q XVII-101 XVII-102 XVII-103 XVII-104 XVII-105 XVII-106 XVII-107 XVII-108 XVII-109 XVII-110 XVII-111 XVII-112 XVII-113 XVII-114 XVII-115 XVII-116 XVZI-117 XVZI-118 XVII-119 XVII-120 XVII-121 XVII-122 XVII-123 XVII-124 XVII-125 XVII-126 XVII-127 XVIZ-128 XVII-129 XVII-130 XVII-131 XVIZ-132 XVII-133 XVII-134 XVII-135 T=TABLE F=FIGURE NUMBER T XVII-27 F XVII-47 F XVII-48 F XVII-49 F XVII-50 F XVII-51 F XVII-52 F XVII-53 F XVII-54 F XVII-55 T XVII-28 T XVII-29 F XVII-56 F XVII-57 F XVII-58 T XVII-30 F XVII-59 F XVII-60 F XVII-61 F XVII-62 REVISION NUMBER 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 PAGE NUMBER XVII-136 XVII-137 XVII-138 XVII-139 XVII-140 XVII-141 XVII-142 STABLE F=FIGURE NUMBER F XVII-63 F XVII-64 F XVII-65 REVISION NUMBER UFSAR Revision 13 EP 17-2 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES SECTION XVIII PAGE NUMBER XVIII-1 XVIII-2 XVIII-3 XVIII-4 XVIII-5 XVIII-6 XVIII-7 XVIII-8 XVIII-9 XVIII-10 XVIII-11 XVIII-12 XVIII-13 XVIII-14 XVIII-15 XVIII-16 XVIII-17 XVIII-18 XVIII-19 XVIII-20 T=TABLE F=FIGURE NUMBER t T XVIII-1 REVISION NUMBER 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 12 T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER UFSAR Revision 13 EP 18-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES APPENDIX A PAGE NUMBER T=TABLE F=FIGURE REVISION NUMBER NUMBER T=TABLE PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER A-1 UFSAR Revision 13 EP A-1 June 1995
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| NINE MILE POINT UNIT 1 FSAR (UPDATED)LIST OF EFFECTIVE PAGES APPENDIX B PAGE NUMBER B-i B-ii B.0-1 T=TABLE F=FIGURE NUMBER REVISION NUMBER 11 12 12 PAGE NUMBER B.10-3 B.11-1 B.11-2 T=TABLE F=FIGURE REVISION NUMBER NUMBER 11 12 B.1-1 BE 1-2 B.1-3 B.1-4 B.1-5 B.2-1 B.2-2 B.2-3 B.2-4 B.2-5 B.2-6 B.3-1 B.3-2 B.3-3 B.3-4 B.4-1 B.4-2 B.4-3 B.5-1 B.6-1 B.6-2 B.7-1'.7-2 B.7-3 B.7-4 12 12 13 13 12 12.12 12 12 12 12 12 12 11 11 11 12 12 12 12 12 11 11 B.12-1 B.12-2 B.13-1 B.14-1 B.15-1 B.15-2 B.16-1 B.17-1 B.17-2 B.18-1 B.18-2 B.18-3 T B-1 S?1 1 T B-1 S?1 2 T B-2 T,B-3 Sh 1 T B-3 S11 2 T B-3 Sh 3 T B-3 Sh 4 T B-3 Sh 5 T B-3 Sh 6 T B-3 Sh 7 T B-3 Sh 8 12 12 11 12 12'1 13 13 12 12 12 12 12 12 12 12 12 B.8-1 B.9-1 B.9-2 B.9-3 11 12 11 B.10-1 B.10-2 11 12 UFSAR Revision 13 EP B-1 June 1995
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| '1 t U.S.NUCLEAR REGULATORY COMMISSION DOCKET 50-220 LICENSE DPR-63.NINE MILE POINT NUCLEAR STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT (UPDATED)VOLUME 1'UNE 1995-REVISION 13 NIAGAI&MOHAWK POWER CORPORATION SYRACUSE, NEW YORK 0
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| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS Section Title P acae SECTION I TABLE OF CONTENTS LIST OF FIGURES LIST OF TABLES INTRODUCTION AND
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| ==SUMMARY==
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| XXX xiii A.1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 B.1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 14.0 15.0 16.0 C.D.SECTION II PRINCIPAL DESIGN CRITERIA General Buildings and Structures Reactor Reactor Vessel Containment Control and Instrumentation Electrical Power Radioactive Waste Disposal Shielding and Access Control Fuel Handling and Storage CHARACTERISTICS Site Reactor Core Fuel Assembly Control System Core Design and Operating Conditions Design Power Peaking Factor Nuclear Design Data Reactor Vessel Coolant Recirculation Loops Primary Containment Secondary Containment Structural Design Station Electrical System Reactor Instrumentation System Reactor Protection System IDENTIFICATION OF CONTRACTORS GENERAL CONCLUSIONS STATION SITE AND ENVIRONMENT I-2 I-2 I-2 I-3 I-4 I-5 Z-7 I-8 I-8 I-9 I-9 I-10 I-10 I-10 I-10 I-10.I-11 I-11 I-11 I-12 I-12 I-12 I-12 I-13 I-13 I-13 I-13 I-14 I-15 X-16 IX-1 A.1.0 2.0 3.0 SITE DESCRIPTION General Physical Features Property Use and Development II-1 II-1 XI-1 II-5 UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section B.1.0 1.1 2.0 2.1 2.2 2.3 Title DESCRIPTION OF AREA ADJ'ACENT TO THE SITE General Population Agricultural, Industrial and Recreational Use Agricultural Use Industrial Use Recreational Use Pacae II-6 II-6 ZZ-6 XI-9 II-9 IZ-11 ZZ-17 C.D.E.F.SECTION IXI METEOROLOGY LIMNOLOGY EARTH SCIENCES ENVIRONMENTAL RADIOLOGY BUILDINGS AND STRUCTURES II-22 XX-23 II-24 II-25 IIZ-1 A.1.0 1.1 1.2 1.3 1.4 1.5 2.0 2.1 2.2 2.3 2.4 3.0 B.1.0 1.1 1.2 1.3 1.4 1.5 2.0 2.1 2.2 2.3 2.4 3.0 TURBINE BUILDING Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating and Ventilation Shielding and Access Control Structure Design General Structural Features Heating and Ventilation System Smoke and Heat Removal Shielding and Access Control Safety Analysis CONTROL ROOM Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating and Ventilation Shielding and Access Control Structure Design General Structural Features Heating, Ventilation and Air Conditioning System Smoke and Heat.Removal Shielding and Access Control Safety Analysis III-4 IZZ-4 IIZ-4 XII-4 IXZ-4 III-6 ZZZ-6 IIZ-6 II1-6 III-17 IZX-21 III-21 III-21 IZI-23 ZII-23 III-23 III-23 XIZ-23 III-23 III-23a III-24 ZZZ-24 ZIZ-25 ZZZ-27 IXI-27 ZZZ-28 UFSAR Revision 0'une 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section C.1.0 1.1 1.2 1.3 1.4 1.5 2.0 2.1 2.2 2.3 3.0 D.1.0 1.1 1.2 1.3 1.4 1.5 2.0 2.1 2.2 2.3 3.0 E.1.0 1.1 1.1.1 1.1.2 1.1.3 1.1.4 1.1.5 1.2 1.2.1 1.2.2 1.2.3 1.3 2.0 2.1 2.1.1 2.1.2 2.1.3 2.1.4 2.1.5 2.1.6 2.1.7 Title WASTE DISPOSAL BUILDING Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating and Ventilation Shielding and Access Control Structure Design General Structural Features Heating and Ventilation System Shielding and Access Control Safety Analysis OFFGAS BUILDING Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating and Ventilation Shielding and Access Control Structure Design General Structural Features Heating and Ventilation System Shielding and Access Control Safety Analysis NONCONTROLLED BUILDINGS Administration Building Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating, Cooling and Ventilation Shielding and Access Control Structure Design General Structural Features Heating, Ventilation and Air Conditioning Access Control Safety Analysis Sewage Treatment Building Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Electrical Design Fire and Explosive Gas Detection Heating and Ventilation Shielding and Access Control Pacae III-29 IIZ-29 ZI1-29 IZZ-29 III-29 III-30 ZZI-30 III-30 ZII-30 IIZ-32 III-34 IIZ-36 ZII-38 ZII-38 III-38 III-38 ZII-38 ZZI-38 III-38 III-38 III-39 ZII-39 ZII-41 III-41 ZII-42 III-42 IIZ-42 IZI-42 III-42 III-42 ZII-43 IIZ-43 IZZ-43 III-43 ZII-45 IZI-45 IIZ-47 III-47 IZZ-47a III-47a IZI-47a IZZ-47a ZII-47a ZII-47a III-48 ZZI-48 UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 2.2 2.2.1 2.2.2 2.2.3 3.0 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.2 3.2.1 3.2.2 3.2.3 Title Structure Design General Structural Features Ventilation System Access Control Energy Information Center Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating and Ventilation Shielding and Access Control Structure Design General Structural Features Heating and Ventilation System Access Control Pacae III-48 IZX-48 ZII-49 III-49 III-49 IZI-49 IXI-49 III-50 XII-50 ZZI-50 III-50 III-51 ZII-51 I I 1-51 Zll-52 F.1.0 1.1 1.1.1 1.1.2 1.1.3 1.1.4 1.1.5 1.2 2.0 2.1 2.2 3.0 G.1.0 1.1 1.2 1.3 1.4 2.0 3.0 3.1 3.2 3.2.1 3.2.2 3.2.3 H.SCREENHOUSE~
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| INTAKE AND DISCHARGE TUNNELS Screenhouse Design Basis Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating and Ventilation Shielding and Access Control Structure Design Intake and Discharge Tunnels Design Bases Structure Design Safety Analysis STACK Design Bases General Wind Loading Seismic Design Shielding and Access Control Structure Design Safety Analysis Radiology Stack Failure Analysis Reactor Building Diesel-Generator Building Screen and Pump House SECURITY BUILDING AND SECURITY BUILDING ANNEX II1-53 ZII-53 III-53 ZIZ-53 ZZZ-53 III-53 IIZ-53 III-53 III-53 IIZ-57 XI I-57 III-57 III-59 III-61 ZZZ-61 III-61 ZZZ-61 ZZI-61 ZIZ-61 ZZZ-61 IZX-64 ZII-64 ZII-64 ZIX-66 ZZZ-66 ZZZ-67 III-68 UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)~Sectio 1~0 1.1 1.2 1.3 1.4 1~5 2'2~1 2~2 2'3.0 4'SECTION IV A.1.0 2.0 3.0 B.1.0 2'2.1 2'2.2.1 2.2'2.3 3.0 3.1 3.1~1 3.1.2 3.2 3.2.1 3.2.2 3.3 4.0 4.1 4'5'5.1 5''RADWASTE SOLIDIFICATION AND STORAGE BUILDING Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design and Internal Loadings Heating, Ventilation and Air Conditioning Shielding and Access Control Structure and Design General Structural Features Heating, Ventilation and Air Conditioning Shielding and Access Control Use References REACTOR DESIGN BASES General Performance Ob)ecti.ves Design Limits and Targets REACTOR DESIGN General Nuclear Design Technique Reference Loading Pattern Final Loading Pattern Acceptable Deviation From Reference Loading Pattern Reexamination of Licensing Basis Refueling Cycle Reactivity Balance Thermal and Hydraulic Characteristics Thermal and Hydraulic Design Recirculation Flow Control Core Thermal Limits Thermal and Hydraulic Analyses Hydraulic Analysis Thermal Analysis'eactor Transients Stability Analysis Design Bases Stability Analysis Method Mechanical Design and Evaluation Fuel Mechanical Design Design Bases Pacae III-69 III-69 XII-69 III-69 III-69 XXI-69 III-70 III-70 III-70 III-71 III-73 III-73 III-75 IV-1 IV-1 IV-1 IV-1 IV-2 IV-3 IV-3 IV-5 XV-6 IV-6 IV-6 IV-7 IV-8 IV-8 IV-8 IV-8 IV-9 IV-15 IV-15 IV-17 IV-23 IV-23 IV-23 IV-24 IV-25 IV-25 IV-25 UFSAR Revision 13 June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 5.1.2 5.1.3 5.1.4 5.1.5 5.1.6 5.1.7 6.0 6.1 6.1.1 6.1.2 6.2 6.2.1 6.2.2 6.3 6.4 7.0 7.1 7.1.1 7.1.2 7.1.3 7.1.4 7.1.5 7.1.6 7.1.7 7.1.8 7.2 7.3 SECTION V Title'L Fuel Rods Water Rods Fuel Assemblies Mechanical Design Limits and Stress Analysis Relationship Between Fuel Design Limits and Fuel Damage Limits Surveillance and Testing Control Rod Mechanical Design and Evaluation Design Control Rods and Drives Standby Liquid Poison System Control System Evaluation Rod Withdrawal Errors Evaluation Overall Control System Evaluation Limiting Conditions for Operation and Surveillance Control Rod Lifetime Reactor Vessel Internal Structure Design Bases Core Shroud Core Support Top Grid Control Rod Guide Tubes Feedwater Sparger Core Spray Spargers Liquid Poison Sparger Steam Separator and Dryer Design Evaluation Surveillance and Testing REACTOR COOLANT SYSTEM Pacae IV-27 ZV-29 IV-29 IV-30 ZV-31 IV-31 ZV-32 IV-32 ZV-32 ZV-40 ZV-40 IV-40 ZV-41 ZV-45 IV-45 ZV-46 IV-46 IV-47 IV-49 IV-49 ZV-49 IV-50 ZV-50 IV-50 ZV-50 IV-51 IV-52 A.1.0 2.0 3.0 4.0 5;0 B.1.0 1.1 1.2 1.3 1.4 1.5 DESIGN BASES General Performance Objectives Design Pressure Cyclic Loads (Mechanical and Thermal)Codes SYSTEM DESIGN AND OPERATION General Drawings Materials of Construction Thermal Stresses Primary Coolant Leakage Coolant Chemistry V-1 V-1 V-1 V-6 V-6 V-8 v-10 V-10 V-10 V-10 V-10 V-12 V-12a UFSAR Revision 12 Vi June 1994 Nine Mile Point Unit, 1 FSAR TABLE OF CONTENTS (Cont'.)Section 2.0 3.0 4.0 5.0 C.1.0 2.0 3.0 4.0 4.1 4.2 4.3 4.4 4.5 5.0 5.1 5.2 5.3 6.0 D.1.0.2.0 2.1 2.2 Title Reactor Vessel Reactor Recirculation Loops Reactor Steam and Auxiliary Systems Piping Relief Devices SYSTEM DESIGN EVALUATION General Pressure Design Heatup and Cooldown Rates Materials Radiation Exposure Pressure-Temperature Limit Curves Temperature Limits for Boltup Temperature Limits for In-Service System Pressure Tests Operating Limits During Heatup, Cooldown, and Core Operation Predicted Shift in RT~~Mechanical Considerations Jet Reaction Forces Seismic Forces Piping Failure Studies Safety Limits, Limiting Safety Settings and Minimum Conditions for Operation TESTS AND INSPECTIONS Prestartup Testing Inspection and Testing Following Startup Hydro Pressure Pressure Vessel Irradiation PBcee V-12a V-14 V-16 V-16 V-19 V-19 V-19 V-21 V-21 V-21 V-22 V-22 V-23a V-23a V-23a V-23a V-23b v-26 V-2 6 V-27 V-27 V-27 V-27 V-27 E.1.0 2.0 3.0 3.1 3.2 3.3 3.4 4.0 4.1 4.2 SECTION VI EMERGENCY COOLING SYSTEM Design Bases System Design and Operation Design Evaluation Redundancy Makeup Water System Leaks Containment Isolation Tests and Inspections Prestartup Test Subsequent Inspections and Tests CONTAXNMENT SYSTEM PRXMARY CONTAINMENT-MARK I CONTAXNMENT PROGRAM V-28 V-28 V-28 v-30 V-30 v-30 V-30 V-31 V-31 V-31 V-31 VI-1 VI-2 UFSAR Revision 12 Vii June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)~Sectic 1~0 2.0 2.1 2.2 2.3 3.0 B.1.0 1~1 1~2 1'1'1.5 1.6 1.7 2.0 2.1 2.2 2.3 2.4 2.5 2.6, 2.7 C.1~0 1~1 1'1.3 1.4 2~0 2.1 T~it e General Structure Pressure Suppression Hydrodynamic Loads Safety/Relief Valve Discharge Loss-of-Coolant Accident Summary of Loading Phenomena Plant Unique Modifications PRIMARY CONTAINMENT
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| -PRESSURE SUPPRESSION SYSTEM Design Bases General Design Basis Accident Containment Heat Removal Isolation Criteria Vacuum Relief Criteria Flooding Criteria Shielding Structure Design-General Penetrations and Access Openings Jet and Missile Protection Materials Shielding Vacuum Relief Containment Flooding SECONDARY CONTAINMENT
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| -REACTOR BUILDING Design Bases Wind and Snow Loadings Pressure Relief Design Seismic Design Shielding Structure Design General Structural Features~acae VI-2 VI-2 VI-3 VI-3 VI-5 VI-6 VI-8 VI-8 VI-8 VI-8 VI-11 VI-11 VI-11 VI-12 VI-12 VI-13 VI-13 VI-15 VI-21 VI-22 VI-22a ,-VI-22a VI-23 VI-24 VI-24 VI-24 VI-24 VI-25 VI-25 VI-25 VI-26 D.1.0 1.1 2.0 3.0 E., 1.0 1 1 1.2 2.0 CONTAINMENT ISOLATION SYSTEM Design Bases Containment Spray Appendix J Water Seal Requirements System Design Tests and Inspections CONTAINMENT VENTILATION SYSTEM Pri'mary Containment ,Design Bases System Design Secondary Containment VI-43 VI-43 VI-56a VI-57 VI-59 VI-60 VI-60 VI-60 VI-60 VI-62 UFSAR Revision 13 viii June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)~Sectio 2~1 2'Design Bases System Design~ae VI-62 VX-62 UFSAR Revision 13 viiia June 1995 Nine Mile Point Unit 1 FSAR THIS PAGE INTENTIONALLY BLANK UFSAR Revision 13 viiib June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)~Sectic F.1.0 1.1 1.2 2.0 2.1 2'3.0 4.0 5.0 5.1'5~2 5.3 SECTION VII TEST AND INSPECTIONS Drywell and Suppression Chamber Preoperational Testing Postoperational Testing Containment Penetrations and Isolation Valves Penetration and Valve Leakage Valve Operability Test Containment Ventilation System Other Containment Tests Reactor Building Reactor Building Normal Ventilation System Reactor Building Isolation Valves Emergency Ventilation System ENGINEERED SAFEGUARDS
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| ~ae VI-65 VI-65 VI-65 VI-65 VI-66 VI-66 VI-67 VI-67 VI-67 VI-68 VI-68 VI-68 VI-68 VII-1 A.1.0 2.0 2.1 2.2 3.0 4.0 B.1.0 2.0 2'3.0 4.0 C.1.0 2.0 2~1 3.0 4.0 5.0 D.1.0 2.0 3.0 3.1 3.2 3.3 4.0 CORE SPRAY SYSTEM Design Bases System Design General Operator Assessment Design Evaluation Tests and Inspections CONTAINMENT SPRAY SYSTEM Design Bases System Design Operator Assessment Design Evaluation Tests and Inspections LIQUID POISON INJECTION SYSTEM Design Bases System Design Operator Assessment Design Evaluation Tests and Inspections Alternate Boron Injection CONTROL ROD VELOCITY LIMITER Design Bases System Design Design Evaluation General Design Sensitivity Normal Operation Tests and Inspections VII-2 VII-2 VII-2 VII-2 VII-7 VII-8 VII-9 VII-11 VII-11 VII-11a VII-17 VII-18 VII-19 VII-21 VII-21 VII-21a VII-27 VII-28 VII-29 VII-30 VII-31 VII-31 VII-31 VII-34 VII-34 VII-34 VII-35 VII-36 UFSAR Revision 13 iX June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section E.1.0 2.0 2.1 3.0 4.0 F.1.0 2.0 3.0 4.0 G.1.0 2.0 2.1 2.2 3.'3.1 3.2 4.0 H.1.0 2.0 2.1 3.0 4.0 I.1.0 2.0 3.0 4.0 SECTION A.1.0 1.1 1.2 2.0 2.1" 2.2 3.0 VIII Title CONTROL ROD HOUSING SUPPORT Design Bases System Design Loads and Deflections Design Evaluation Tests and Inspections FLOW RESTRICTORS Design Bases System Design Design Evaluation Tests and Inspections COMBUSTIBLE GAS CONTROL SYSTEM Design Bases Containment Inerting System System Design Design Evaluation Containment Atmospheric Dilution System System Design Design Evaluation Tests and Inspections EMERGENCY VENTILATION SYSTEM Design Bases System Design Operator Assessment Design Evaluation Tests and Inspections HXGH-PRESSURE COOLANT ZN JECTXON Design Bases System Design Design Evaluation Tests and Xnspections INSTRUMENTATZON AND CONTROL PROTECTIVE SYSTEMS Design Bases Reactor Protection System Anticipated Transients Without Scram Mitigation System System Design Reactor Protection System Anticipated Transients Without Scram Mitigation System System Evaluation Pacae VZI-37 VZI-37 VI1-38 VZZ-40 VII-41 VII-41 VZZ-42 VZI-42 VIZ-42 VII-42 VZZ-43 VZZ-44 VZZ-44 VII-44 VZZ-44 VZZ-46a VZZ-48 VZZ-48 VIX-49 VZI-53 VZI-55 VII-55 VII-55a VZI-58 vZZ-58 VII-59 VII-61 VZZ-61 VII-61 VZI-61a VZZ-62a VIXX-1 VIIZ-1 VIII-1 VXII-1 VIII-6 VIZ1-6 VZZZ-6 VXXI-14 VXII-15 UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section B.1.0 2~0 2~1 2.2 2.3 2.4 3.0 3'3.2 3'3.4 REGULATING SYSTEMS Design Bases System Design Control Rod Adjustment Control Recirculation Flow Control Pressure and Turbine Control Reactor Feedwater Control System Evaluation Control Rod Adjustment Control Recirculation Flow Control Pressure and Turbine Control Reactor Feedwater Control~ae VIII-18 VIII-18 VIII-18 VIII-18 VIII-18 VIII-20 VIII-21 VIII-21 VIII-21 VIII-22 VIII-22 VIII-22 C.1.0 1.1 1~1~1 1.1~2 1.1~3 1.1~4 1.1.5 1.2 1.2.1 1.2.2 1.2.3 1.2.4 2.0 2.1 2.1.1 2.1.2 2.1.3 2'2'.1 2.2.2 2'.3 3.0 3'3~1~1 3.1.2 3.2 4.0 4.1 INSTRUMENTATION SYSTEMS Nuclear Instrumentation Design Source Range Monitors Intermediate Range Monitors Local Power Range Monitors Average Power Range Monitors Traversing In-Core Probe System Evaluation Source Range Monitors Intermediate Range Monitors Local Power Range Monitors Average Power Range Monitors Non-Nuclear Process Instrumentation Design Bases Non-Nuclear Process Instruments in-Protective System Nonnuclear Process Instruments in Regulating Systems Other Nonnuclear Process Instruments Evaluation Nonnuclear Process Instruments in Protective System Nonnuclear Process Instruments in Regulating Systems Other Nonnuclear Process Instruments Radioactivity Instrumentation Design Bases Radiation Monitors in Protective Systems Other Radiation Monitors Evaluation Other Instrumentation Rod Worth Minimizer VIII-23 VIII-23 VIII-23 VIII-25 VIII-28 VIII-32 VIII-32 VIII-40 VIII-40 VIII-42 VIII-43 VIII-50 VIII-50 VIII-54 VIII-54 VIII-54 VIII-58 VIII-60 VIII-62 VIII-62 VIII-62 VIII-62a VIII-62a VIII-62a VIII-62a VIII-68 VIII-76 VIII-76 VIII-76 UFSAR Revision 13 June 1995 Nine Mile Point Unit 1 FSAR STABLE OF CONTENTS (Cont'd.)Section 4.1.1 4.1.2 5.0 5.1 5.2 5.3 5.4 5.4.1 5.4.2 5.4.3 5.4.4 5.5 5.6 5.6.1 5.6.2 5.6.3 5.6.4 5.6.5 5.6.6 I 5''5.6.9 Title Design Bases Evaluation Regulatory Guide 1.97 (Revision 2)Instrumentation Licensing Activities
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| -Background Definition of RG 1.97 Variable Types and Instrument Categories Determination of RG 1.97 Type A Variables for NMP1 Determination of"EOP Key Parameters" for NMP1" Determination Basis/Approach Definition of Primary Safety Functions Association of EOPs to Primary Safety Functions Identification of EOP Key Parameters NMP1 RG 1.97 Variables, Variable Type, and Associated Instrument Category Designations Summary of the RG 1.97 Instrument Design and Implementation Criteria that were Established for NMP1 as Part of the Unit 1 1990 Restart Activities No Type A Variables EOP Key Parameters Single Tap for the Fuel Zone RPV Water Level Instrument Nonredundant Wide-Range RPV Water Level Indication Upgrading EOP Key Parameter Category 1 Instrument Loop Components to Safety Related Classification Safety Related Classification of Instrumentation for RG 1.97 Variable Types Other than the EOP Key Parameters Routing and Separation of Channelized Category 1 Instrument Loop Cables Electrical Isolation of Category 1 Instrument Loops from Associated Components that are not Safety Related Power Source Information for Category 1 Instruments P RCRe VIXI-76 VIXI-80 VIXI-80 VZ1Z-81 VIII-81 VIII-82a VIII-82b VIIX-82b VZII-82c VIII-82d VIII-84 VIII-86 VIII-99 VIXI-99 VI1I-100 VIII-101 VIXI-102 VIIZ-104 VIIX-105 VIII-105 VIII-106 VIII-107 UFSAR Revision 12 Xii June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 5.6.10 5.6.11 5.6.12 5.7 SECTION IX A.B.1.0 1.1 1.2 2.0 2.1 2.2 2.3 2.4 2.5 3.0 3.1 3'3.3 3.4 3.4.1 3.4.2 3.4.3 3.5 3.5.1 3.5.2 4.0 4.1 4.2 4.3 5.0 5.1 5.2 5.3 6.0 6.1 Marking of Instruments of Control Room Panels"Alternate" Instruments for Monitoring EOP Key Parameters Indicating Ranges of Monitoring Instruments References ELECTRICAL SYSTEMS DESIGN BASES ELECTRICAL SYSTEM DESIGN Network Interconnections 345-kV System 115-kV System Station Distribution System Two f24-Volt DC Systems Two 120-V, 60-Hz, Single-Phase, Uninterruptible Power Supply Systems Two 120-V, 57-60 Hz, One-Phase, Reactor Trip Power Supplies One 120/208-V, 60-Hz, Instrument and Control Transformer One 120/240-V, 60-Hz, Three-Phase, Computer Power Supply Cables and Cable Trays Cable Separation Cable Penetrations Protection in Hazardous Areas Types of Cables Power Cable Control Cable Special Cable Design and Spacing of Cable Trays Tray Design Specifications Tray Spacing Emergency Power Diesel Generator System Station Batteries Nonsafety Battery System Tests and Inspections Diesel-Generator Station Batteries Nonsafety Batteries Conformance with 10CFR50.63, Station Blackout Rule Station Blackout Duration~ae VZIZ-108 VIIZ-108 VIII-109 VIII-110 IX-1 IX-1 IX-2 IX-2 IX-2 IX-3 IX-8a IX-13 IX-15 IX-16 IX-16 IX-16a IX-16a IX-'16 a IX-17 IX-17 ZX-18 IX-18 IX-18 IX-19 IX-19 IX-19 IX-20 IX-20 IX-20 IX-26 IX-31 IX-34 IX-34 IX-35 IX-35 IX-35 IX-35 UFSAR Revision 13 xiii June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 6.2 6.3 6.4 6.5 6.6 Station Blackout Coping Capability Procedures and Training Quality Assurance EDG Reliability Program References
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| ~ae IX-36 IX-38 IX-39 IX-39 IX-40 SECTION X A.1'2.0 3.0 4.0 B.1.0 2.0 3.0 4.0 C.1.0 2.0 2~1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 2.10 2'1 2.12 2.13 2.14 3.0 3.1 3.2'.3 3.4 3.5 4.0 5.0 REACTOR AUXILIARY AND EMERGENCY SYSTEMS REACTOR SHUTDOWN COOLING SYSTEM Design Bases System Design System Evaluation Tests and Inspections 1 REACTOR CLEANUP SYSTEM Design Bases System Design System Evaluation Tests and Inspections CONTROL ROD DRIVE HYDRAULIC SYSTEM Design Bases System Design Pumps Filters First Pressure Stage Second Pressure Stage Third Pressure Stage Exhaust Header Accumulator Scram Pilot Valves Scram Valves Scram Dump Volume Control Rod Drive Cooling System Directional Control and Speed Control Valves Rod Insertion and Withdrawal Scram Actuation System Evaluation Normal Withdrawal Speed Accidental Multiple Operation Scram Reliability Operational Reliability Alternate Rod Injection (ARI)Reactor Vessel Level Instrumentation Reference Leg Backfill Tests and Inspections X-1 X-1 X-1 X-1 X-3X-3 X-4 X-4 X-4 X-7 X-7 X-8 X-8 X-8 X-10 X-10 X-11 X-11 X-12 X-13 X-13 X-14 X-14 X-15 X-16 X-16 X-17 X-17 X-18 X-18 X-19 X-19 X-20 X-21 X-21 X-21 UFSAR Revision 13 xiv June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Sect o D.1'2.0 3.0 4.0 E.1.0 2'3.0 4.0 F.1.0 2.0 3.0 4.0 G.1.0 2'3.0 4.0 H.1'2.0 3.0 4.0 1.0 2.0 3.0 4.0 1~0 2.0 2.1 2.1.1 2.2 3.0 4.0 REACTOR BUILDING CLOSED LOOP COOLING WATER SYSTEM Design Bases System Design Design Evaluation Tests and Inspections TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM Design Bases System Design Design Evaluation Tests and Inspections SERVICE WATER SYSTEM Design Bases System Design Design Evaluation Tests and Inspections MAKEUP WATER SYSTEM Design Bases System Design System Evaluation Tests and Inspections SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM Design Bases System Design Design Evaluation Tests and Inspections BREATHING, INSTRUMENT AND SERVICE AIR SYSTEM Design Bases System Design Design Evaluation Tests and Inspections FUEL AND REACTOR COMPONENTS HANDLING SYSTEM Design Bases System Design Description of Facility Cask Drop Protection System Operation of the Facility Design Evaluation Tests and Inspections X-22 X-22 X-22 X-25 X-26 X-27 X-27 X-27 X-29 X-30 X-31 X-31 X-31 X-33 X-34 X-35 X-35 X-35 X-37 X-38 X-39 X-39 X-39a X-43 X-44 X-45 X-45 X-45 X-48 X-49 X-50 X-50 X-50 X-50 X-54 X-55 X-57 X-58 UFSAR Revision 13 xv June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section K.1.0 1.1 1.2 1.3 1.4 1.5 1.6 2.0 2.1 2.2 2.3 2.4 3.0 3.1 3.2 4.0 Title FIRE PROTECTION PROGRAM Program Bases Nuclear Division Directive-Fire Protection Program Nuclear Division Interface Procedure-Fixe Protection Program Fire Hazards Analysis (FHA)Appendix R Review Safe Shutdown Analysis (SSA)Fire Protection and Appendix R Related Portions of Operations Procedures (OPs, SOPs, and EOPs)and Damage Repair Procedures Fire Protection Portions of the Emergency Plan Program Implementation and Design Aspects Fire Protection Implementing Procedures Fire Protection Administrative Controls Fire Protection System Drawings and Calculations Fire Protection Engineering Evaluations (FPEEs)Monitoring and Evaluating Program Implementation Fire Protection Quality Assurance Program (FPQAP)Fire Brigade Manning, Training, Drills and Responsibilities Surveillance and Tests Pacae X-59 X-59 X-59 X-59 X-59 X-60 X-60 X-60 X-60 X-60 X-61 X-61 X-61 X-62 x-62 X-62 x-62 L.1.0 2.0 3.0 4.0 M.1.0 2.0 3.0 4.0 I APPENDIX 1 OA I APPENDIX 10B REMOTE SHUTDOWN SYSTEM Design Bases System Design System Evaluation Tests and Inspections SAFETY PARAMETER DISPLAY SYSTEM Design Bases System Design System Evaluation Tests and Inspections FIRE HAZARDS ANALYSIS SAFE SHUTDOWN ANALYSIS X-76 X-76 X-76 X-76 X-77 X-78 X-78 X-78 X-78 X-79 UFSAR Revision 12 Xvi June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section SECTION XZ A.Ti.tie STEAM TO POWER CONVERSION SYSTEM DESIGN BASES P RcRB XZ-,1 XI-1, B.1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 SYSTEM DESIGN AND OPERATION Turbine-Generator Turbine Condenser Condenser Air Removal and Offgas System Circulating Water System Condensate Pumps Condensate Demineralirer System Condensate Transfer System Feedwater Booster Pumps Feedwater Pumps Feedwater Heaters XI-9 XZ-9 XI-11 XI-12 XI-17 XZ-17 XI-18 XI-19 XI-20 XI-20 XZ-20 C.D.SECTION XZZ SYSTEM ANALYSIS TESTS AND INSPECTIONS RADIOLOGICAL CONTROLS XZ-22 XZ-25 XII-1 1.0 1.1 1.2 1.2.1 1.2.2 1.2.3 2.0 2.1 2.1.1 2.1.2 2.1.3 2.1.4 2.2 2.2.1 2.2.2 2.2.3 2.2.4 2~3 2.3.1 2.3.2 3.0 4.0 4.1 RAD10ACTZVE WASTES Design Bases Objectives Types of Radioactive Wastes Gaseous Wastes Liquid Wastes Solid Wastes System Design and Evaluation Gaseous Waste System Offgas System , Steam Packing Exhauster System Buildup Ventilation Systems Stack Liquid Waste System Liquid Waste Handling Processes Sampling and Monitoring Liquid Wastes Liquid Waste Equipment.
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| Arrangement Liquid Radioactive Waste System Control Solid Waste System Solid Waste Handling Processes Solid Waste System Equipment Safety Limits Tests and Inspections Waste Process Systems XZI-1 XII-1 XZZ-1 XIZ-1 XZZ-1 XZI-2 XZZ-3 XII-3 XII-3 XII-7 XII-7 XZZ-7 XII-8 XII-8 XZI-8 XII-12 XII-13 XII-13 XII-15 XZZ-15 XIZ-17 XI I-17 XZI-18 XIZ-18 UFSAR Revision 12 XV3.i U June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)~sectic 4.2 4.3 4.3.1 4.3.2 B.1.0 1.1 1.2 1.2.1 1''1.'2.3 1'2.0 2.1 2.1.1 2.1.2 2.1~3 2.2 2.2.1 2.2.2 2 2'3.0 3~1 3.1.1 3.1.2 3.1.3 3.1.4 3'3.2.1 3.2.2 3.3 3.3.1 3.3'3.3.3 3.4 3.4.1 3.5 3.5.1 3.5.2 3.5.3 3.5'3.5.5 4.0 Filters Effluent Monitors Offgas and Stack Monitors Liquid Waste Effluent Monitor RADIATION PROTECTION Primary and Secondary Shielding Design Bases Design Reactor Shield Wall Biological Shield Miscellaneous Evaluation Area Radioactivity Monitoring Systems Area Radiation Monitoring System Design Bases Design Evaluation Area Air Contamination Monitoring System Design Bases Design Evaluation Radiation Protection Facilities Laboratory, Counting Room and Calibration Facilities Change Room and Laundry Facilities Personnel Decontamination Facility Tool and Equipment Decontamination Facility Radiation, Control Shielding Access Control Contamination Control Facility Contamination Control Personnel Contamination Control Airborne Contamination Control Personnel.
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| Dose Determinations Radiation Dose Radiation Protection Instrumentation Counting Room Instrumentation Portable Radiation Instrumentation Air Sampling Instrumentation Personnel Monitoring Instruments Emergency Instrumentation Tests and Inspections
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| ~ae XII-18 XII-19 XII-19 XII-19 XII-20 XII-20 XII-20 XII-22 XII-22 XII-23 XII-23 XII-23 XII-24 XII-24 XII-24 XII-25 XII-30 XII-31 XII-31 XII-31 XII-31 XII-32 XII-32a XII-32a XII-33 XII-34 XII-34 XII-35 XII-35 XII-36 XII-37 XII-38 XII-38 XII-39 XII-41 XII-41 XII-41a XII-41a XII-42 XII-42 XII-43 XII-43 XII-44 UFSAR Revision 13 xvi1 1 June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 4.1 4.2 4.3 4'4.4'4.4.2 4.5 Shielding Area Radiation Monitors Area Air Contamination Monitors Radiation Protection Facilities Ventilation Air Flows Instrument Calibration Well Shielding Radiation Protection Instrumentation A.1.0 1.1.3 1.1~4 1.1.5 1'.6 1.2 2.0 2.1 3.0 4.0 ORGANIZATION AND RESPONSIBILITY Management and Technical Support Organization Nuclear Division Vice President Nuclear Generation Vice President Nuclear Engineering Vice President Nuclear Safety Assessment and Support Director Nuclear Communications and Public Affairs Director Human Resource Development Controller Nuclear Division Corporate Support Departments Operating Organization Plant Manager Quality Assurance Facility Staff Qualifications SECTION XIII CONDUCT OF OPERATIONS Pacae XII-44 XII-44 XII-45 XII-45 XII-45'XII-45 XII-45 XIII-1 XIII-1 XIII-1 XIII-1 XIII-1 XIII-2'III-3 XIII-6 XIII-7 XIII-7 XIII-7 XIII-8 XIII-8 XIII-11 XIII-12 B.1.0 2.0 3.0 4.0 4.1 4.2 4.3 4.3.1 4.3.2 4.3.3 4.3.4 4''4.3.6 4.3.7 QUALIFICATIONS AND TRAINING OF PERSONNEL, This Section Deleted This Section Deleted This Section Deleted Training of Personnel=
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| General Responsibility Implementation Quality For Operator Training For Maintenance For Technicians For General Employee Training/Radiation Protection and Emergency Plan For Industrial Safety For Nuclear Quality Assurance For Fire Brigade XIII-13 XIII-13 XIII-13 XIII-13 XIII-13 XIII-13 XIII-13 XIII-13 XIII-13 XZZZ-14 XIII-14 XIII-14 XIII-14 XIII-14 XIII-14 UFSAR Revision 13 XiX June 1995 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 5.0 C.D.E.F.1.0 1.1 1.2 1.3 1.4 2.0 3.0 3.1 3.2 3.3 3.4 3.5 4.0 5.0 6.0 7.0 G.1..0 1'2.0 2.1 3.0 SECTION XZV Title Training of Licensed Operator Candidates/Licensed NRC Operator Retraining Cooperative Training with Local, State and Federal Officials OPERATING PROCEDURES EMERGENCY PLAN AND PROCEDURES SECURITY RECORDS Operations
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| .Control Room Log Book Station Shift Supervisor's Book Radwaste Log Book Waste Quantity Level Shipped Maintenance Radiation Protection Personnel Exposure By-Product Material as Required by 10CFR30 Meter Calibrations Station Radiological Conditions in-Accessible Areas Administration of the Radiation Protection Program and Procedures Chemistry and Radiochemistry Special Nuclear Materials Calibration of Instruments Administrative Records and Reports REVIEW AND AUDIT OF OPERATIONS Station Operations Review Committee Function Safety Review and Audit Board Function Review of Operating Experience INITIAL TESTING AND OPERATIONS P acae XIII-14 XIII-15 XIII-16 XIII-17 XIII-19 XIII-20 XIII-20 XIII-20 XIII-20 XZZI-20 XIII-20 XIII-21 XIII-21 XIII-21 XIII-21 XIII-21 XIII-21 XTII-21 XIII-21 XIII-21 XZII-22 XIII-22 XIII-23 XIII-23 XZZZ-23 XIII-23 XIII-23 XIII-24 XZV-1 A.TESTS PRIOR TO INITIAL REACTOR FUELING XIV-1 B.1.0 INITIAL CRITICALITY AND POSTCRITICALZTY TESTS Initial Fuel Loading and Near-Zero Power Tests at: Atmospheric Pressure XIV-5 XZV-5 UFSAR Revision 12 XX June 1994 Nine Mile Point Unit 1 FSAR'L TABLE OF CONTENTS (Cont'd.)Section 1.1 1.2 1.3 2.0 2.1 2.2 3.0 4.0 5.0 6.0 SECTION XV A.B.1.0 2.0 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 3.10 3.11 3.12 3'.12.1 3.12.2 3.12.3 3.12.4 3.13 3.14 Title General Recpxirements General Procedures Core Loading and Critical Test Program Heatup from Ambient to Rated Temperature General Tests Conducted From Zero to 100 Percent Initial Reactor Rating Full-Power Demonstration Run Comparison of Base Conditions Additional Tests at Design Rating SAFETY ANALYSIS INTRODUCTION BOUNDARY PROTECTION SYSTEMS Transients Considered Methods and Assumptions Transient Analysis Turbine Trip Without Bypass Loss of 100'F Feedwater Heating Feedwater Controller Failure-Maximum Demand Control Rod Withdrawal Error Main Steam Line Isolation Valve Closure (With Scram)Inadvertent Startup of Cold Recirculation Loop Recirculation Pump Trips Recirculation Pump Stall Recirculation Flow Controller Malfunction
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| -Increasing Flow Flow Controller Malfunction-Decrease Flow Inadvertent Actuation of One Solenoid Relief Valve Safety Valve Actuation Objectives Assumptions and Initial Conditions Comments Results Feedwater Controller Malfunction (Zero Demand)Turbine Trip with Partial Bypass (Low Power)Pacae XZV-5 XIV-5 XZV-8 XIV-10 XZV-10 XIV-10 XIV-12 XIV-14 XZV-14 XIV-14 XV-1 XV-1 XV-2 XV-2 XV-5a XV-6 XV-6 XV-7a XV-9 XV-11a XV-18 XV-20 XV-24 XV-27 XV-28 XV-32 XV-34 XV-36 XV-36 XV-36 XV-37 XV-37 XV-4 1 XV-43 UFSAR Revision 12 xx1 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 3.15 3.16 3,17 3.18 3.19 3.20 3.21 3.22 3.23 3.24 3.25 C.1.0 1.1 1.2 1.2.1 1.2.2 1.2.3 1.2.4 1.2.5 1.2.6 1.2.7 1.2.8 1.3 1.3.1 1.3.2 1.3.3 2.0 2.1 I 2.2 2.2.1 2.2.2 2.2.3 2.2.4 2.3 2.4 Title Turbine Trip with Partial Bypass (Full Power)Inadvertent Actuation of One Bypass Valve One Feedwater Pump Trip and Restart Loss of Main Condenser Vacuum Loss of Electrical Load (Generator Trip)Loss of Auxiliary Power Pressure Regulator Malfunction Instrument Air Failure D-C Power Interruptions Failure of One Diesel-Generator to Start Power Bus Loss of Voltage STANDBY SAFEGUARDS ANALYSIS Main Steam Line Break Outside the Drywell Identification of Causes Accident Analysis Valve Closure Initiation Feedwater Flow Core Shutdown Mixture Level Subcooled Liquid System Pressure and Steam-Water Mass Mixture Impact Forces Core Internal Forces Radiological Effects Radioactivity Releases Meteorology and Dose Rates Comparison with Regulatory Guide 1.5 Loss-of-Coolant Accident Introduction Input to Analysis Operational and ECCS Input Parameters Single Failure Study on ECCS Manually-Controlled Electrically-Operated Valves Single Failure Basis Pipe Whip Basis Appendix K LOCA Performance Analysis for P8DNB277 Appendix K LOCA Performance Analysis Pacae XV-44 XV-4 8 xv-50 xv-50 XV-52 XV-54 XV-56 XV-57 XV-66 XV-67 xv-68 XV-70 XV-70 XV-70 XV-70 XV-7 1 XV-7 1 XV-72 XV-72 XV-72 XV-72 XV-73 XV-75 XV-75 XV-75 XV-77 XV-78 XV-81 XV-81 xv-81 a XV-81a xv-81a xV-96 xv-96 XV-96 XV-134 UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 2.4.1 2.4.2 2.4.3 2.4.4 2.5 3.0 3.1 3.2 3.3 3.3.1 3.3.2 3.3.3 4.0 4.1 4.2 4'4.4 4,5 4.5.1 4.5.2 5.0 5.1 5.1.1 5.1.2 5.1.3 5.1.4 5.1.5 5.1.6 5.1.7 5.1.8 5.2 5.2.1 5.2.2 5.2.3 5.2.4 5.2.5 5.2.6 5.3 5.3.1 5.3.2 Ti.tie Computer Codes Description of Model Changes Analysis Procedure Analysis Results Appendix K LOCA MAPLHGR Evaluation with Core Spray Flow Through One Sparger Refueling Accident Identification of Causes Accident Analysis Radiological Effects Fission Product Releases Meteorology and Dose Rates Comparison to Regulatory Guide 1.25 Control Rod Drop Accident Identification of Causes Accident Analysis Designed Safeguards Procedural Safeguards Radiological Effects Fission Product Releases Meteorology and Dose Rates Containment Design Basis Accident Original Recirculation Line Rupture Analysis-With Core Spray Purpose Analysis Method and Assumptions Core Heat Buildup Core Spray System Containment Pressure Immediately Following Blowdown Containment Spray Blowdown Effects on Core Components Radiological Effects Original Containment Design Basis Accident Analysis-Without Core Spray Purpose Core Heatup Containment Response Fission Product Release from the Fuel Fission Product Release from the Reactor and Containment Meteorology and Dose Rates Design Basis Reconstitution (DBR)Suppression Chamber Heatup Analysis Introduction Input to Analysis PRcRB XV-137a XV-137a XV-137b XV-137d XV-137d XV-137d2 XV-137d2 XV-13 9 XV-142 XV-142 XV-145 XV-145 XV-14 6 XV-14 6 XV-14 9 XV-152 XV-153 XV-154 XV-154 XV-15 9 XV-15 9 XV-159a XV-159a XV-159a XV-15 9b XV-159c XV-159f XV-159i XV-159k XV-159m XV-1 60 XV-160 XV-160 XV-1 64 XV-164a XV-167 XV-169 XV-169 XV-169 XV-169b UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 5.3.3 5.3.4 6.0 6.1 6.2 6.3 7.0 7~1 7.2 7.3 7.4 7.5 7.6 7.7 SECTION XVZ A.1.0 2.0 2.1 2.2 2.2.1 2.3 2.4 2.5 2.6 2.6.1 2.6.2 2.6.3 2.6.4 2.6.5 2.7 2.7.1 2.7.2 2.7.3 3.0 3.1 3.2 Title DBR Suppression Chamber Heatup Analysis Conclusions New Fuel Bundle Loading Errox Analysis Identification of Causes Accident Analysis Safety Requirements Meteorological Models Used in Accident Analyses Ground Releases Stack Releases Variability Exfiltration Ground Deposition Thyroid Dose Whole Body Dose SPECIAL TOPICAL REPORTS REACTOR VESSEL Applicability of Formal Codes and Pertinent Certifications Design Analysis Code Approval Analysis Steady-State Analysis Basis for Determining Stresses Pipe Reaction Calculations Earthquake Loading Criteria and Analysis Reactor Vessel Support Stress Design Criteria and Analysis Strain Safety Margin for Reactor Vessels Introduction Strain Margin Failure Probability Results of Probability Analysis Conclusions Components Required for Safe Reactor Shutdown Design Basis Load Combinations Expected Stress and Deformation Stresses and Deformations at Which the Component is Unable to Function and Margin of Safety Inspection and Test Report Summary Materials Fabrication and Inspection Pacae XV-169b XV-169g XV-169k XV-1 69k XV-1 69k XV-171 XV-171 XV-171 XV-171 XV-173 XV-175 XV-1 97 XV-1 98'V-199 XVZ-1 XVZ-1 XVI-1 XVZ-3 XVI-3 XVI-3 XVZ-3 XVI-10 XVI-10 XVZ-14 XVI-24 XVZ-24 XVZ-2 6 XVI-27 XVI-3 1 XVI-31 XVZ-36 XVI-36 XVI-36 XVI-41 XVI-47 XVZ-47 XVI-47 UFSAR Revision 12 xxiv June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 4.0 4.1 4.2 B.1.0 2.0 2.1 2.2 2.3 2.4 2.5 2'.1 2.6 2.7 2.8 2'.1 2.8.2 2.8.3 2.8.4 2.8.5 2.8.6 2.8.7 2.8.8 2.9 3.0 3.1 3.2 3.3 3.3.1 3.3.2 C.1.0 1.1 1.2 1.3 1.3.1 Title Surveillance Provisions Coupon Surveillance Program Periodic Inspection PRESSURE SUPPRESSION CONTAINMENT Applicability of Formal Codes and Pertinent Certifications Design Analysis Code Approval Calculations Under Rated Conditions Ultimate Capability Under Accident Conditions Capability to Withstand Internal Missiles and Jet Forces Flooding Capabilities of the Containment Drywell Air Gap Tests and Inspections Biological Shield Wall Compatibility of Dynamic Deformations Occurring in the Drywell, Torus, and Connecting Vent Pipes Containment Penetrations Classification of Penetrations Design Bases Method of Stress Analysis Leak Test Capability Fatigue Design Material Specification Applicable Codes Jet and Reaction Loads Drywell Shear Resistance Capability and Support Skirt Junction Stresses Inspection and Test Report Summary Fabrication and Inspection Tests Conducted Discussion of Results Results Effect of Various Transients ENGINEERED SAFEGUARDS Seismic Analysis and Stress Report Introduction Mathematical Model Method of Analysis Flexibility or.Influence Coefficient Matrix P acae XVI-5 1 XVI-51 XVI-51 XVI-52 XVI-52 XVI-53 XVI-53 XVI-53 XVI-53 XVZ-57 XVZ-61 XVI-63 XVZ-65 XVI-69 XVI-74 XVI-74 XVZ-74 XVZ-7 6 XVI-7 6 XVZ-77 XVI-77 XVI-77 XVI-7 8 XVI-79 XVI-83 XVZ-83 XVZ-83 XVI-85 XVI-85 XVI-87 XVZ-92 XVI-92 XVI-92 XVZ-93 XVI-94 XVZ-95 UFSAR Revision 12 xxv June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section Title Pacae 1.3.2 1.3.3 1.3.4 1.3.5 1.3.6 1.3.7 1.4 2.0 2.1 2.1.1 2.1.2 2.1.3 2.1.4 2.2 2.2.1 D.1.0 1.1 1.1.1 1.1.2 1.2 1.3 2.0 2.1 I F 1-1 2.1.2 2.1.3 2.2 3.0 4.0 Normal Mode Frequencies and Mode Shapes The Seismic Spectrum Values Dynamic Modal Loads Modal Response Quantities The Combined Response Quantities Basic Criteria for Analysis Discussion of Results Containment Spray System Design Adequacy at Rated Conditions General Condensation and Heat Removal Mechanisms Mechanical Design Loss-of-Coolant Accident Summary of Test Results Spray Tests Conducted DESIGN OF STRUCTURES, COMPONENTS/
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| EQUIPMENT AND SY'STEMS Classification and Seismic Criteria Design Techniques Structures Systems and Components Pipe Supports Seismic Exposure Assumptions Plant Design for Protection Against Postulated Piping Failures in High Energy Lines Inside Primary Containment Containment Integrity Analysis Systems Affected by Line Break Engineered Safeguards Protection Outside Primary Containment Building Separation Analysis Tornado Protection XVZ-96 XVI-96 XVI-97 XVZ-98 XVI-98 XVZ-99 XVI-99 XVI-1 04 XVZ-104 XVI-104 XVI-104 XVI-1 13 XVI-113 XVZ-116 XVI-116 XVI-123 XVI-123 XVI-126 XVZ-126 XVI-146 XVZ-153 XVI-154 XVI-155 XVZ-155 XVI-155a XVZ-158 XVI-163 XVI-1 65 XVZ-168 XVZ-168 E.G.EXHIBITS CONTAINMENT DESIGN REVIEW USAGE OF CODES/STANDARDS FOR STRUCTURAL STEEL AND CONCRETE SECTION XVII ORIGINAL ENVIRONMENTAL STUDIES XVI-189 XVI-227 XVZ-238 XVI I-1 A.1.0 2.0 METEOROLOGX General Synoptic Meteorological Factors XVII-1 XVI I-1 XVI I-2 UFSAR Revision 12 xxvi June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 3.0 3.1 3.1.1 3.1~2 3.1.3 3.2 3.3 3.4 3.4.1 4.0 4.1 4.2 4.3 4.3.1 4.4 4.4.1 4.4.2 4.5 4.6 4.6.1 4.6;2 4.6.3 4.7 5.0 B.1.0 2.0 3.0 3.1 3.2 3.2.1 3.2.2 3.2.3 3.3 Title Micrometeoro logy Wind Patterns 200-Foot Wind Roses Estimates of Winds at the 350-Foot Level Comparison Between Tower and Satellite Winds Lapse Rate Distributions Turbulence Classes Dispersion Parameters Changes in Dispersion Parameters Applications to Release Problems Concentrations from a Ground-Level Source Concentrations from an Elevated Source Radial Concentrations Monthly and Annual Sector Concentrations Least Favorable Concentrations Over an Extended Period Ground-Level Release Elevated Release Mean Annual Sector Deposition Dose Rates from a Plume of Gamma Emitters RADOS Program Centerline Dose Rates Sector Dose Rates Concentrations from a Major Steam Line Break Conclusions L IMNOLOGY Introduction Summary Report of Cruises Dilution of Station Effluent in Selected Areas Dilution of Effluent at the Lake Surface Above the Discharge Dilution of Effluent at the Site Boundaries General Dilution of Effluent at the Eastern Site Boundary Dilution of Effluent West of the Station Site Dilution of Effluent at the City of Oswego Intake Pacae XVI I-2 XVI I-2 XVI I-2 XVII-2 XVII-16 XVII-19 XVIZ-19 XVII-19 XVII-39 XVIZ-45 XVZZ-46 XVZZ-53 XVZZ-55 XVZZ-55 XVII-83 XVZZ-83 XVIZ-86 XVI1-87 XVII-90 XVIZ-90 XVII-91 XVZZ-100 XVII-103 XVZZ-106 XVZZ-107 XVZZ-107 XVII-107 XVZZ-109 XVII-109 XVZZ-114 XVIZ-114 XVZI-116 XVII-122 XVZZ-123 UFSAR Revision 12 Xxvii June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 3.3.6 3.4 3.5 4.0 4.1 4.1.1 4.1.2 4.2 5.0 Title Tilting of the Isothermal Planes and Subsequent Dilution Dilution as a Function of Current Velocity Percent of Time Effluent Will Be Carried to the Oswego Area Mixing with Distance Oswego River Water as a Buffer to Prevent Effluent From Passing Over the Intake Summary of Annual Dilution Factors for the City of Oswego Intake Dilution of Effluent at the Nine Mile Point Intake Summary of Dilution in the Nine Mile Point Area Preliminary Study of Lake Biota Off Nine Mile Point Biological Studies Plankton Study Bottom Study Summary of Biological Studies Conclusions P acae XVZI-123 XVII-124 XVII-127 XVII-127 XVII-127 XVII-127 XVZI-128 XVII-128 XVII-129 XVII-129 XVII-129 XVII-129 XVII-130 XVZZ-130 C.1.0 2.0 3.0 3.1 3.2 4.0 4.1 4.2 4.3 4.4 4.5 SECTION XVZZI EARTH SCIENCES Introduction Additional Subsurface Studies Construction Experience Station Area Intake and Discharge Tunnels Correlation With Previous Studies General Geological Conditions Hydrological Conditions Seismological Conditions Conclusion HUMAN FACTORS ENGINEERING/SAFETY PARAMETER DISPLAY SYSTEM XVZI-132 XVII-132 XVII-132 XVII-138 XVII-138 XVII-139 XVII-140 XVII-140 XVII-140 XVII-142 XVZI-142 XVZZ-142 XVZZZ-1 A.1.0 2.0 3.0 3.1 3.2 3.3 3.4 3.5 DETAILED CONTROL ROOM DESIGN REVIEW General Planning Requirements for the DCRDR DCRDR Review Process Operator Survey Historical Review Task'nalysis Control Room Inventory Control Room Survey XVIIZ-1 XVIII-1 XVIII-2 XVIII-2 XVI I I-2 XVIII-2 XVIII-3 XVIZZ-3 XVIII-4 UFSAR Revision 12 XXV'.3.3.June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section Title~acae 3.6 3.7 3.8 4.0 4.1 4.2 4.2.1 4.2.2.5.0 6.0 6.1 6.2 6.3 6.4 7.0 Verification of Task Performance Capabilities Validation o f Control Room Functions~Compilation of Discrepancy Findings Assessment and Implementation Assessment Implementation Integrated Cosmetic Package Functional Fixes Reporting Continuing Human Factors Program Fix Verifications Multidisciplinary Review Team Assessments Human Factors Manual for Future Design Change Outstanding Human Factors Items References XVI I I-4 XVII I-4 XVIZZ-5 XVZ I I-5 XVI I I-5 XVIZI-5 XVIZ1-6 XVIII-7 XVIII-7 XVIII-8 XVIII-8 XVZ I I-8 XVIII-8 XVIII-8 XVIII-9 B.1.0 2.0 3.0 4.0 5.0 5.1 5.1.1 5.1.2 5.1.3 5.1.4 5.2 5.2.1 5.2.2 5.3 5.3.1 5.3.2 5.4 SAFETY PARAMETER DISPLAY SYSTEM Introduction to the Safety Parameter Display System System Description Role of the SPDS Human Factors Engineering Guidelines Human Factors Engineering Principles Applied to the SPDS Design NUREG-0737, Supplement 1, Section 4.1.a Concise Display Criteria Plant Variables Rapid and Reliable Determination of Safety Status Aid to Control Room Personnel NUREG-0737, Supplement 1, Section 4.1.b Convenient Location Continuous Display NUREG-0737,'Supplement 1, Section 4.1.c Procedures and Training Isolation of SPDS from Safety-Related Systems NUREG-0737, Supplement 1, Section 4.1.e XVZII-11 XVIII-11 XVIII-11 XVIII-12 XVIII-13 XVIII-13 XVIZZ-13 XVIII-13 XVIII-13 XVIIZ-14 XVIII-14 XVIII-14 XVI I I-14 XVZII-15 XVIII-15 XVI I I-15 XVIII-15 XVIII-16 UFSAR Revision 12 xxix June 1994 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)Section 5.4.1 5.4.2 5.5 6.0 6.1 6.2 7.0 Title Incorporation of Accepted Human Factors Engineering Principles Information Can be Readily Perceived and Comprehended NUREG-0737, Supplement 1, Section 4.1.f, Sufficient Information Procedures Operating Procedures Surveillance Procedures References Pacae XVIII-16 XVIII-17 XVIII-17 XVIII-17 XVIII-17 XVIII-18 XVIII-19 APPENDIX A APPENDIX B Unused NIAGARA MOHAWK POWER CORPORATION QUALITY ASSURANCE PROGRAM TOPICAL REPORT (NMPC-QATR-1)
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| ~NINE MILE POINT NUCLEAR STATION UNITS 1 AND 2 OPERATIONS PHASE UFSAR Revision 12 XXZ June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES Figure Number II-1 ZI-2 II-3 I I-4 ZI-5 XI-6 ZZI-1 III-2 IIX-3 IXI-4 Xjj-5 Ijj-6 IXZ-7 ZZI-8 IIX-9 XXX-10 IZI-11 III-12 IIX-13 IIZ-14 IIX-15 IIZ-16 IIX-17 IZI-18 III-19 Title Piping, Instrument and Equipment Symbols Station Location Area Map Site Topography Population Distribution Within a Twelve Mile Radius of the Station Counties and Towns Within Twelve Miles of the Station 1980 Population Distribution Within a Fifty Mile Radius of the Station Plot Plan Station Floor Plan-Elevation 225-6 Station Floor Plan-Elevations 237 and 250 Station Floor Plan-Elevation 261 Station Floor Plan-Elevations 277 and 281 Station Floor Plan-Elevations 281 and 291 Station Floor Plan-Elevations 298 and 300 Station Floor Plan--Elevations 317-6 and 318 Station Floor Plan-Elevations 320, 333-8, 340 and 369 Section Between Column Rows 7'and 8 Section Between Column Rows 12 and 14 Turbine Building Ventilation System Laboratory and Radiation Protection Facility Ventilation System Control Room Ventilation System Waste Disposal Building Ventilation System Waste Disposal Building Extension Ventilation System Offgas Building Ventilation System Technical Support Center Ventilation System Circulating Water Channels Under Screen and Pump House-Normal Operation Pacae I-17 II-2 IZ-3 XZ-4 ZI-7 IZ-8 IZ-10 III-3 IZZ-7 ZZI-8 III-9 III-10 XIZ-11 ZII-12 XIZ-13 XII-14 Ijj-15 XII-16 Xlj-18 XIX-20 IZI-26 ZII-33 XXX-35 Ijj-40 IZI-46 IXI-55 UFSAR Revision XXXi June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure~umber III-20 III-21 III-22 III-23 IV-1 IV-2 IV-3 IV-3a IV-4 I IV-5 IV-5a IV-5b IV-6 IV-7 IV-8 IV-9 V-1 V-2 V-3 V-4 V-5 V-6 V-7 V-8 VI-1 VI-2 VI-3 VI-4 VI-5 VI-6 xMt1R Circulating Water Channels Under Screen and Pump House-Special Operations Intake and Discharge Tunnels Plan and Profile Stack-Plan and Elevation Stack Failure-Critical Directions
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| /Limiting Power/Flow Line Figure Deleted Figure Deleted GE11 Fuel Assembly Design Typical Control Rod-Isometric Figure Deleted Control Rod Positioning Within Core Cell GE8X8EB Control Rod Positioning Within Core Cell With Typical GE11 Fuel Assembly Control Rod Drive and Hydraulic System Control Rod Drive Assembly Typical Control Rod to Drive Coupling-Isometric Reactor Vessel Isometric Reactor Emergency Coolant System Reactor Vessel Nozzle Location.Reactor Vessel Support Figure Deleted Pressure Vessel Embrittlement Trend Figure Deleted Figure Deleted Emergency Condenser Supply Isolation Valves (Typical of 2)Drywell and Suppression Chamber Electrical Penetrations
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| -High Voltage Electrical Penetrations
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| -Low Voltage Pipe Penetrations
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| -Hot Typical Penetration For Instrument Lines Reactor Building Dynamic Analysis-Acceleration East-West Direction~ae III-56 III-58 III-62 III-65 IV-10 IV-26 IV-28 IV-28a IV-33 IV-34 IV-34a IV-34b IV-37 IV-38 IV-39 IV-48 V-11 V-13 V-15 V-18 V-23 V-24 V-25 V-32 VI-14'I-17 VI-18 VI-19 VI-20 VI-28 UFSAR Revision 13 XXXii June 1995 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number VI-7 vz-8 VI-9 VX-10 VZ-11 VI-12 Vz-13 VZ-14 VI-15 VI-16 VZ-17 VI-18 VI-19 vz-20 VZ-21 VX-22 VI-23 VI-24 VII-1 VIX-2 VXI-3 VII-4 VII-4a VIZ-5 vzz-6 VII-7 vzx-8 Title Reactor Building Dynamic Analysis Deflections East-West Direction Reactor Building Dynamic Analysis-Elevation vs.Building Shear East-West Direction Reactor Building Dynamic Analysis Elevation vs.Building Moment East-West Direction Reactor Building Dynamic Analysis-Acceleration North-South Direction Reactor Building Dynamic Analysis-Deflections North-South Direction Reactor Building Dynamic Analysis-Elevation vs.Building Shear-North-South Direction Reactor Building Dynamic Analysis'-
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| Elevation vs.Building Moment-North-South Direction Reactor Support Dynamic Analysis Elevation vs.Acceleration Reactor Support Dynamic Analysis-Elevation vs.Deflection Reactor Support Dynamic Analysis-Elevation vs.Shear Reactor Support Dynamic Analysis-Elevation vs.Moment Typical Door Seals Details of Reactor Building Air Locks Instrument Line Isolation Valve Arrangement Typical Flow Check Valve Isolation Valve System Drywell Cooling System Reactor Building Ventilation System Core Spray System Core Spray Sparger Flow, Per Sparger, for One Core Spray Pump and One Topping Pump Containment Spray System Figure Deleted Figure Deleted Figure Deleted Liquid Poison System Minimum Allowable Solution Temperature Figure Deleted Pacae vx-29 VI-30 VI-31 VI-32 VI-33 VI-34 VX-35 Vz-36 VZ-37 vz-38 VI-39 VZ-41 VX-42 vz-53 VI-54 vz-58 VI-61 vz-63 VII-3 VXI-5 VIZ-12 VII-13 VZZ-13a Vzz-14 VII-22 VII-24 VII-25 UFSAR Revision 12 ZXXiii June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number VII-9 VZI-10 VZI-11 VZI-12 VII-13 VXI-14 VZI-15 VII-16 VII-17 VIIZ-1 VIII-2 VIII-3 VXII-4 VIII-5 VIII-6 VIII-7 VXII-8 VIII-9 VIIX-10 VXII-11 VIII-12 VXII-13 VIXI-14 VIII-15 VIII-16 VIIZ-17 VIXX-18 VIII-19 Title Typical Control Rod Velocity Limiter Control Rod Housing Support Hydrogen Flammability Limits Combustible Gas Control System H,-O, Sampling System H,-O, Concentrations in Containment Following LOCA N, Added by CAD Operation Following LOCA Containment Pressure with CAD Operation-Zero Containment Leakage Feedwater Delivery Capability (Shaft Driven Pump)Protective System Function Reactor Protection System Elementary Diagram Protective System Typical Sensor Arrangement Recirculation Flow and Turbine Control Neutron Monitoring Instrument Ranges Source Range Monitor (SRM)SRM Detector Location Intermediate Range Monitor (IRM)IRM Core Location LPRM Location Within Core Lattice LPRM and APRM Core Location Local Power Range Monitor (LPRM)and Average Power Range Monitors (APRM)APRM System-Typical Trip Logic f'r APRM Scram and Rod Block Traversing In-Core Probe Rod Pattern During Startup Radial Power Distribution for Control Rod Pattern Shown in Figure VXII-16 Distance from Worst Control Rod to Nearest Active IRM Monitor Measured Response Time of Intermediate Range Safety Instrumentation Pacae VII-32 VI1-39 VII-45 VII-47 VII-50 VZZ-51 VII-52 VII-54 VII-63 VIII-2 VIII-8 VIII-16 VXIZ-19 VII1-24 VIII-26 VIII-27 VI I I-29 VIII-30 VIIX-33 VIII-34 VIXI-35 VIXI-37 VIII-38 VIXI-41 VIII-45 VIII-46 VIIZ-47 VIXI-49 UFSAR Revision 12 XXX'.V June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number VIXI-20 VIIZ-21 VIIZ-22 VIII-23 VIII-24 VIXI-25 VIIZ-26 VIII-27 VIII-28 VIII-29 IX-1 IX-2 IX-3 IX-4 IX-5 IX-6 Ix-7 X-1 X-2-X-3 X-4 X-6 X-7 X-8 X-10 X-11 Title Envelope of Maximum APRM Deviation by Flow Control Reduction in Power Envelope of Maximum APRM Deviation for APRM Tracking With On Units Control Rod Withdrawal Main Steam Line Radiation Monitor Reactor Building Ventilation Radiation Monitor Offgas System Radiation Monitor Emergency Condenser Vent Radiation Monitor Stack Effluent and Liquid Effluent Radiation Monitors Containment Spray Heat Exchanger Raw Water Effluent Radiation Monitor Containment Atmospheric Monitoring System Rod Worth Minimizer A.C.Station Power Distribution Control and Instrument Power Trays Below Elevation 261 Trays Below Elevation 277 Trays Below Elevation 300 Diesel Generator Loading Following Loss-of-Coolant Accident Diesel Generator Loading for Orderly Shutdown Reactor Shutdown Cooling System Reactor Cleanup System Control Rod Drive Hydraulic System Reactor Building Closed Loop Cooling System Turbine Building Closed Loop Cooling System Service Water System Decay Heat Generation vs.Days After Reactor Shutdown Spent Fuel Storage Pool Filtering and Cooling System Breathing, Instrument, and Service Air Reactor Refueling System Pictorial Cask Drop Protection System P acae VIIX-52 VXZZ-53 VIII-63 VIII-65 VXII-66 VXIZ-69 VXII-70 VXIX-73 VIXI-75 VIIZ-78 Ix-9 XX-14 IX-21 Ix-22 Ix-23 Zx-27 Ix-28 X-2 x-5 X-9 X-23 x-28 X-32 X-40 X-41 X-46 X-51 x-56 UFSAR Revision 12 XXXV June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number XI-1 XI-2 XI-3 XI-4 XI-5 XI-6 XI-7 XII-1 XIII-1 I XIII-2 XIII-3 XIII-4 XIII-5 XV-1 I XV 2 XV-2a XV-2b XV-3 XV-4 XV-4a XV-4b XV-5 XV-6 XV-6a XV-7 XV-8 XV-9 XV-10 XV-11~T't e Steam Flow and Reheater Ventilation System Extraction Steam Flow Main Condenser Air Removal and Offgas System Circulating Water System Condensate Flow Condensate Transfer System Feedwater Flow System Radioactive Waste Disposal System NMPC Upper Management Nuclear Organization Nine Mile Point Nuclear Site Organization Nuclear Engineering Organization Nuclear Safety Assessment and Support Organization Safety Organization Station Transient Diagram Figure Deleted Figure Deleted Figure Deleted Plant Response to Loss of 100 F Feedwater Heating Figure Deleted Plant Response to Feedwater Controller Failure (GE11 at EOC)Plant Response to Feedwater Controller Failure (GE11 at EOC Extended Load Line Limit)Figure Deleted Figure Deleted Figure Deleted Figure Deleted Startup of Cold Recirculation Loop-Partial Power Recirculation Pump Trips (1 Pump)Recirculation Pump Trips (5 Pumps)Recirculation Pump Stall~ae XI-2 XI-3 XI-4 XI-5 XI-6 XI-7 XI-8 XII-9 XV-3 XV-8 XV-8a XV-8b XV-10 XV-12 XV-12 a XV-12b XV-16 XV-17 XV-17a XV-19 XV-23 XV-25 XV-2 6 XV-2 9 UFSAR Revision 13 xxxvi June 1995 Nine Mile Point Unit 1 FSAR 1 LIST OF FIGURES (Cont,'d.)
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| Figure Number XV-12 XV-13 XV-14 XV-15 XV-16 XV-17 XV-18 XV-19 XV-20 XV-2 1 XV-22 XV-23 XV-24 XV-2 5 XV-26 XV-27 XV-28 XV-2 9 XV-3 0 XV-3 1 XV-3 2 XV-3 3 XV-33a XV-34 XV-3 5 XV-3 6 XV-3 7 XV-3 8 XV-39 XV-4 0 XV-4 1 XV-4 2 XV-43 XV-44 XV-45 XV-46 XV-47 XV-48 XV-49 XV-50 Title Flow Controller Malfunction (Increased Flow)Flow Controller Malfunction Decreasing Flow Inadvertent Actuation of One Solenoid Relief Valve Figure Deleted Figure Deleted Feedwater Controller Malfunction-Zero Flow Turbine Trip With Partial Bypass Intermediate Power Turbine Trip With Partial Bypass Inadvertent Actuation of One Bypass Valve One Feedwater Pump Trip and Restart Loss of Electrical Load Loss of Auxiliary Power Pressure Regulator Malfunction Main Steam Line Break-Coolant Loss Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted~ae XV-3 1 XV-3 3 XV-35 XV-39 XV-40 XV-42 XV-45 XV-47 XV-49 XV-51 XV-53 XV-55 XV-58 XV-74 XV-85 XV-86 XV-89 XV-90 XV-9 1 XV-92 XV-93 XV-94 XV-94a XV-103 XV-104 XV-105 XV-106 XV-107 XV-108 XV-109 XV-110 XV-111 XV-112 XV-113 XV-114 XV-115 XV-1 16 XV-117 XV-118 XV-119 UFSAR Revision 13 xxxvii June 1995 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number XV-51 XV-52 XV-53 XV-54 XV-55 xv-56 XV-5 6A!XV-5 6B XV-5 6C XV-5 6D XV-5 6E XV-5 6F XV-5 6G xv-56H XV-57 XV-58 XV-59 XV-60 xv-60 a XV-60b XV-61 XV-62 xv-63 XV-64 XV-65 XV-66 XV-67 Title Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Figure Deleted Loss-of-Coolant Accident-With Core Spray Cladding Temperature Loss-of-Coolant Accident Drywell Pressure Loss-of-Coolant Accident Suppression Chamber Pressure Loss-of-Coolant Accident Containment Temperature
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| -With Core Spray Loss-of-Coolant Accident Clad Perforation With Core Spray Containment Design Basis Clad Temperature Response-Without Core Spray Containment Design Basis Metal-Water Reaction Containment Design Basis Clad Perforation Without Core Spray Containment Design Basis Containment Temperature
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| -Without Core Spray DBR Analysis Suppression Pool and Wetwell Airspace Temperature Response-Containment Spray Design Basis Assumption DBR Analysis Suppression Pool and Wetwell Airspace Temperature Response-EOP Operation Assumptions Reactor Building Model Exfiltration vs.Wind Speed-Northerly Wind Reactor Building Differential Pressure Exfiltration vs.Wind Speed-Southerly Wind Reactor Building-Isometric Reactor Building-Corner Sections React.or Building-Roof Sections P acae XV-120 XV-121 XV-122 XV-123 XV-12 6 XV-127 XV-137i XV-137j XV-137k XV-159e XV-159g,, XV-159h XV-15 9L XV-159n XV-1 61 XV-1 63 XV-165 XV-1 66 XV-1 69L xv-169m XV-178'v-184 xv-185.XV-186 XV-188 XV-189 XV-1 90 O.UFSAR Revision 12 XXXViii June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number Title Pacae XV-68 XV-69 XV-7 0 XV-7 1 XV-72 Reactor Building-Panel to Concrete Sections Reactor Building-Expansion Joint Sections Reactor Building Exfiltration-Northerly Wind Reactor Building Exfiltration-Southerly Wind Reactor Building Differential Pressure XV-191 XV-1 92 XV-193 XV-1 94 XV-196 O.XVI-1 XVZ-2 XVI-3 XVI-4 XVI-6 XVI-7 XVI-8 XVZ-9 XVI-10 XVI-11 XVI-12 XVZ-13 XVI-14 XVI-15 XVZ-1 6 XVI-17 XVI-1 8 XVI-1 9 XVI-20 XVI-21 Seismic Analysis of Reactor Vessel Geometric and Lumped Mass Representation Reactor Support Dynamic Analysis-Elevation vs.Moment Reactor Support Dynamic Analysis Elevation vs.Shear Reactor Support Dynamic Analysis-Elevation vs.Deflection Reactor Support Dynamic Analysis Elevation vs.Acceleration Figure Deleted Figure Deleted Figure Deleted Reactor Vessel Support Structure Stress Summary Thermal Analysis Failure Probability Density Function Addition Strains Past 4%Required to Exceed Defined Safety Margin Loss of Coolant Accident-Containment Pressure No Core or Containment.Sprays Figure Deleted Drywell to Concrete Air Gap Typical Penetrations Biological Shield Wall Construction Details Vent Pipe and Suppression Chamber Primary Containment Support and Anchorage.Seal Details-Drywell Shell Steel and Adjacent Concrete Drywell Sliding-Acceleration, Shear, and Moment XVI-12 XVI-13 XVI-15 XVI-1 6 XVI-17 XV1-1 8 XVZ-1 9 XVZ-20 XVI-21 XVI-22 XVI-28 XVI-32 XVI-54 XVZ-60 XVI-62 XVZ-64 XVI-66 XVI-70 XVI-7 1 XVI-73 XVI-75 UFSAR Revision 12 Xxxix June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number XVZ-22 XVI-23 XVZ-24 XVI-25 xvI-26 XVZ-27 XVZ-28 XVI-2 9 XVZ-30 XVI-31 XVI-32 XVZ-33 XVZ-34 XVZ-35 xvz-36 XVI-37 xvI-38 XVZ-39 XVI-4 0 XVZ-4 1 XVI-42 XVZ-43 XVI-44 xvI-45 XVI-4 6 XVI-47 XVI-4 8 XVZ-4 9 Title Shear Resistance Capability-Inside Drywell Shear Resistance Capability-Outside Drywell Drywell-Support Skirt Junction Stresses Point Location for Containment Spray System Piping Heat Exchanger to Drywell Comparison of Static and Dynamic Stresses (PSI)Seismic Conditions-Containment Spray System Heat Exchanger to Drywell Conduction in a Droplet Loss of Coolant Accident-Containment Pressure Loss of Coolant Accident-Containment Pressure Nozzle Spray Test-Pressure Drop of 80 psig Nozzle Spray Test-Pressure Drop of 80 psig Nozzle Spray Test-Pressure Drop of 30 psig Nozzle Spray Test-Pressure Drop of 30 psig Seismic Analysis-Reactor Building Dynamic Analysis-Drywell Reactor Support Structure-Seismic Seismic Analysis-Waste Building Seismic Analysis-Screenhouse Seismic Analysis-Turbine Building (North of Row C)Seismic Analysis-Turbine Building (South of Row C)Seismic Analysis-Concrete Ventilation Stack Reactor Building Mathematical Model (North-South)
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| Reactor Support Structure-Seismic Reactor Support Structure-Reactor Building Reactor Support Structure-Reactor Building and Seismic Plan of Building Wall Section 1 Wall Section 1-Detail"A" Wall Section 1-Detail"B" P RcRB XVI-8 0 XVI-81 xvI-82 XVZ-1 0 0 XVI-1 03 XVI-1 0 9 XVI-115 XVZ-117 XVI-118 XVI-119 XVZ-120 XVI-121 XVI-134 XVI-135 XVI-136 XVI-137 XVI-138 XVI-139 XVI-140 XVI-141 XVI-144 XVI-147 XVZ-148 XVI-149 XVZ-169 XVI-170 XVZ-171 XVI-172 UFSAR Revision 12 xl June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number Title Pacae XVI-50 XVI-51 XVI-52 XVI-53 XVI-54 XVI-55 XVI-5 6 XVI-57 XVZ-58 XVZ-5 9 XVI-60 XVI 61 Wall Wall Wall Wall Wall Wall Wall Wall Wall Wall Wall Wall Section Section Section Section Section Section Section Section Section Section Section Section 1-Detail 1-Detail"D" 1-Detail"E" 2 3 3A-Details 4 4-Detail 1 4-Detail 2 5 6 7 XVI-173 XVI-174 XVI-175 XVI-176 XVI-177 XVI-178 XVZ-179 XVI-180 XVI-181 XVI-182 XVI-183 XVI-1 84 XVZ I-1 XVI I-2 XVI I-3 XVI I-4 XVI I-5 XVI I-6 XVI I-7 XVI I-8 XVZ I-9 XVZI-10 XVIZ-11 XVII-12 XVII-13 XVII-14 XVII-15 XVII-16 XVZZ-17 XVII-18 XVZZ-19 Average'3-'4 Average'3-'4 Average'3-'4 Average'3-'4 Average Average Average Average'3-'4 Average'3-'4 Average'3-'4 Average'3-'4 Average'3-'4 Average Average'3-'4 Average'3-'4 Average'3-'4 Average'3-'4 Average Septemb Average'3-'4 Wind Roses for January Wind Roses for February Wind Roses for March Wind Roses for April Wind Roses for May'3-'4 Wind Roses for June'3-'4 Wind Roses for July'63-'64 Wind Roses for August Wind Roses for September Wind Roses for October Wind Roses for November Wind Roses for December Wind Roses for'63-'64 Diurnal Lapse Rate January February'3-'4 Diurnal Lapse Rate March April'3-'4 Diurnal Lapse Rate May June'3-'4 Diurnal Lapse Rate July August'3-'4 Diurnal Lapse Rate er'3-'4, October'3-'4 Diurnal Lapse Rate November December'2-'3 XVII-3 XVII-4 XVZI-5 XVII-6 XVII-7 XVZ I-8 XVI I-9 XVI I-10 XVII-11 XVII-12 XVIZ-13 XVII-14 XVII-15 XVII-20 XVIZ-21 XVIZ-22 XVII-23 XVIZ-24 XVI I-25 UFSAR Revision xli June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number XVII-20 XVII-21 XVII-22 XVIZ-23 XVII-24 XVII-25 XVII-26 XVII-27 XVZZ-28 XVIZ-29 XVII-30 XVII-31 XVIZ-32 XVII-33 XVII-34 XVZI-35 XVII-36 XVIZ-37 XVII-38 Title Lapse Rates by Wind Speed and Turbulence Classes for January'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for February'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for March'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for April'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for May'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for June'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for July'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for August'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for September'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for October'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for November'3-'4 Lapse Rates by Wind Speed and Turbulence Classes for December'3-'4 Sector Map Centerline Concentrations Turbulence Class I Centerline Concentrations-Turbulence Class II Centerline Concentrations-Turbulence Class III Centerline Concentrations-Turbulence Class IV Centerline Concentrations Turbulence Class IZ Becoming Class IV at 2 km and Class II at 23 km Centerline Concentrations-Turbulence Class IV Becoming Class II at 16 km~acae XVZI-27 XVZI-28 XVZI-29 XVII-30 XVII-31 XVII-32 XVII-33 XVII-34 XVZZ-35 XVII-36 XVZI-37 XVIZ-38 XVZI-44 XVII-47 XVI I-4 8 XVZI-49 XVII-50 XVIZ-51 XVZI-52 UFSAR Revision xiii June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number Title P acae xvzz-39 XVI I-4 0 XVII-41 XVII-42 XVII-43 XVII-44 XVII-45 XVZI-46 XVII-47 XVII-48 XVII-49 XVII-50 XVII-51 XVII-52 XVII-53 XVIZ-54 XVI I-55 XVZI-56 XVII-57 XVZI-58 XVI I-59 Centerline Concentrations-Turbulence Class IV Becoming Class II at 2 km Radial Concentrations
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| -Turbulence Class I Radial Concentrations
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| -Turbulence Class II Radial Concentrations
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| -Turbulence Class Ill Radial Concentrations
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| -Turbulence Class IV Radial Concentrations
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| -Turbulence Class II Becoming Class ZV at 2 km and Class II at 23 km Radial Concentrations
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| -Turbulence Class ZV Becoming Class Iz at 16 km Radial Concentrations
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| -Turbulence Class ZV Becoming Class ZZ at 2 km Centerline Gamma Dose Rates-Turbulence Class I Centerline Gamma Dose Rates-Turbulence Class ZZ Centerline Gamma Dose Rates Turbulence Class IZI Centerline Gamma Dose Rates-Turbulence Class IV Centerline Gamma Dose Rates Turbulence Class Iz Becoming Class ZV at 2 km and Class Iz at 23 km Centerline Gamma Dose Rates Turbulence Class IV Becoming Class II at 16 km Centerline Gamma Dose Rates-Turbulence Class IV Becoming Class II at 2 km Assumed Concentration and Dose Rate Distributions Close to the Elevated Source Gamma Dose Rate as a Function of Gy at 1 km From the Source Southeastern Lake Ontario Dilution of Rising Plume Estimated Lake Currents at Cooling Water Discharge Temperature Profiles in an Eastward Current at the Oswego City Water Intake XVZI-54 XVZ I-57 XVII-58 XVII-59 XVI I-60 XVZI-61 XVII-62 xvzZ-63 XVIZ-93 XVIZ-94 XVIZ-95 XVZI-96 XVII-97 XVZZ-98 XVII-99 Xvzz-101 XVII-102 XVII-108 XVII-112 XVIZ-113 XVII-125 UFSAR Revision 12 xliii June 1994 Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)Figure Number XVZ I-60 XVI I-61 XVZ I-62 XVIZ-63 XVII-64 XVI I-65 Title Subsur f ace Section Plot Plan Log of Boring (Boring CB-1)Log of Boring (Boring CB-2)Log of Boring (Boring CB-3)Log of Boring (Boring CB-4)Attenuation Curves Pacae XVII-133 XVZI-134 XVII-135 XVII-136 XVZI-137 XVII-141 UFSAR Revision 12 xliv June 1994 Nine Mile Point Unit 1 FSAR LIST OF TABLES Table Number XI-1 ZI-2 ZI-3 IZ-4 II-5 xx-6 IX-7 II-8 V-1 V-2 V-3 V-4 V-5 VI-1 VI-2 VI-3a VZ-3b VI-3c VI-4 VZ-5 VII-1 VIXI-1 VIXI-2 VIII-3 Title 1980 Population and Population Density for Towns and Cities Within 12 Miles of Nine Mile.Point-Unit 1 Cities Within a 50-mile Radius of the Station With Populations over 10,000 Regional Agricultural Use Regional Agricultural Statistics
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| -Cattle and Milk Production Industrial Firms Within 8 km (5 mi)of Unit 1 Public Utilities in Oswego County Public Water Supply Data for Locations Within an Approximate 30-Mile Radius Recreational Areas in the Region Reactor Coolant System Data Operating Cycles and Transient Analysis Results Fatigue Resistance Analysis Codes for Systems Connected to the Reactor Coolant System Time to Automatic Blowdown Drywell Penetrations Suppression Chamber Penetrations Reactor Coolant System Isolation Valves Primary Containment Isolation Valves-Lines Entering Free Space of the Containment Table Deleted Seismic Design Criteria for Isolation Valves Xnitial Tests Prior to Station Operation Performance Tests Association Between Primary Safety Functions and Emergency Operating Procedures List of EOP Key Parameters Type and Instrument Category for NMP1 RG 1.97'Variables Pacae Ix-6 Ix-9 ZX-11 II-12 IX-13 II-17 IZ-18 II-19" V-3 V-7 V-9 V-9 V-12a VX-44 Vx-46 Vx-47 VZ-49 VZ-51 VI-56 vx-69 VZI-35 VXXI-83 VIII-85 VIIZ-87 UFSAR Revision 12 xlv June 1994 Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)Table Number IX-1 XII-1 XII-2 XII-3 XII-4 XII-5 XII-6 XII-7 XII-8~ice Magnitude and Duty Cycle of Major Station Battery Loads Flows and Activities of Major Sources of Gaseous Activity Quantities and Activities of Liquid Radioactive Wastes Annual Solid Waste Accumulation and Activity Liquid Waste Disposal System-Major Components Solid Waste Disposal System-Major Components Occupancy Times Gamma Energy Groups Area Radiation Monitor Detector Locations Pacae IX-32 XII-2 XII-4 XII-6 XII-14 XII-18 XII-21 XII-22 XII-26 XIII-1 XV-1 XV-2 XV-4 XV-5 XV-6 XV-7 XV-8 XV-9 XV-9A XV-10 XV-11 XV-12 XV-13 XV-14 XV-15 XV-16 XV-17 XV-18 XV-19 ANSI Standard Cross-Reference Unit 1 Transients Considered Trip Points for Protective Functions Table Deleted Instrument Air Failure Blowdown Rates'odine Concentrations Fractional Concentrations in Clouds Main Steam Line Break Accident Doses Significant Input Parameters to the Loss-of-Coolant Accident Analysis Core Spray System Flow Performance Assumed in LOCA Analysis ECCS Single Valve Failure Analysis Single Failures Considered in LOCA Analysis Table Deleted Table Deleted Table Deleted Table Deleted Table Deleted Table Deleted Table Deleted Table Deleted XV-6 XV-14 XV-60 XV-73 XV-76 XV-77 XV-78 XV-82 XV-82b XV-97 XV-98 XV-101 XV-102 XV-124 XV-125 XV-13 1 XV-134 XV-134 XV-13 6 UFSAR Revision 13 xlvi June 1995 Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)Table Number XV-20 XV-2 1 XV-21A XV-2 1B XV-2 1C XV-2 1D XV-21E XV-22 XV-23 XV-24 XV-25 XV-2 6 XV-27 XV-28 XV-2 9 XV-29a XV-2 9b XV-2 9c XV-29d XV-30 XV-32 XV-32a XV-33 XV-34 XV-35 XV-3 6 XVI-1 XVI-2 Title Table Deleted Table Deleted Analysis Assumptions For Nine Mile Point-1 Calculations Table Deleted Table Deleted Table Deleted Table Deleted Reactor Building Airborne Fission Product Inventory Stack Discharge Rates Fuel Handling Accident Doses (REM)Fission Product Release Assumptions Atmospheric Dispersion and Dose Conversion Factors Effect on Dose of Factors Used in the Calculations Noble Gas Release Halogen Release Wetting of Fuel Cladding by Core Spray Airborne Drywell Fission Product Inventory Reactor Building Airborne Fission Product Inventory Stack Discharge Rates Airborne Drywell Fission Product Inventory Reactor Building Airborne Fission Product Inventory Stack Discharge Rates Significant Input Parameters to the DBR Containment Suppression Chamber Heatup Analysis Downwind Ground Concentrations Maximum Ground Concentrations Diversity Factors for"Least Favorable" Ground Concentrations from Stack Release, for Ground Release, for Maximum Ground Concentrations Reactor Building Leakage Paths Code Calculation Summary Steady State-(100%Full Power Normal Operation)
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| Pertinent Stresses or Stress Intensities Pacae XV-13 6 XV-13 6 XV-137e XV-137f XV-137g XV-137h XV-137h XV-144 XV-145 XV-14 6 XV-147 XV-14 8 XV-148 XV-158 XV-15 9 XV-15 9 f'V-159s XV-159u XV-159x XV-167 XV-168 XV-168 XV-169i XV-172 XV-173 XV-17 6 XV-1 95 XVI-4 XVI-6 UFSAR Revision 12 xlvii June 1994 Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)Table Number XVI-3 XVI-4 XVI-5 XVI-6 XVI-7 XVZ-8 XVZ-9 xvz-10 XVl-1 1 XVI-12 XVZ-13 XVI-14 XVI-15 XVI-1 6 XVI-1 7 XVI-1 8 XVZ-1 9 XVZ-20 XVI-21 XVI-22 XVI-23 XVI-2 4 XVI-25 xvz-26 XVI-27 XVI-28 Title List of Reactions for Reactor Vessel Nozzles Effect of Value of Initial Failure Probability Single Transient Event for Reactor Pressure Vessel.Postulated Events Maximum Strains from Postulated Events Core Structure Analysis Recirculation Line Break Core Structure Analysis-Steam Line Break Drywell Jet and Missile Hazard Analysis Data Drywell Jet and Missile Hazard Analysis Results Stress Due to Drywell Flooding Allowable Weld Shear Stress Leak Rate Test Results Overpressure Test-Plate Stresses Stress Summary Heat Transfer Coefficients as a Function of Drop Diameter Heat Transfer Coefficients as a Function of Pressure Relationship Between Particle Size and Type of Spray Pattern Allowable Stresses for Floor Slabs, Beams, Columns, Walls, Foundations, etc.Allowable Stresses for Structural Steel Allowable Stresses-Reactor Vessel Concrete Pedestal Drywell-Analyzed Design Load Combinations Suppression Chamber-Analyzed Design Load Combinations ACI Code 505 Allowable Stresses and Actual Stresses for Concrete Ventilation Stack Allowable Stresses for Concrete Slabs, Walls, Beams, Structural Steel, and Concrete Block Walls System Load Combinations High Energy Systems-Inside Containment P acae XVI-1 1 XVI-31 XVI-33 XVI-34 xvz-35 XVI-37 XVI-40 xvz-55 xvz-56 XVI-59 xvz-65 xvz-86 xvz-91 xvz-101 XVI-110 XVI-111 XVI-122 XVI-127 XVI-128 xvz-129 xvz-130 XVI-131 XVZ-132 XVI-133 XVZ-151 xvz-155a UFSAR Revision 12 xlviii tune 1994 Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)Table Number XVI-29 XVZ-30 XVZ-31 Title High Energy Systems-Outside Containment Systems Which May Be Affected by Pipe Whip Capability to Resist Wind Pressure and Wind Velocity Pacae XVI-166 XVI-167 XVI-185 XVZ I-1 XVZ I-2 XVZ I-3 XVZ I-4 XVII-5 XVII-6 XVI I-7 XVI I-8 XVI I-9 XVIZ-10 XVZI-11 XVII-12 XVII-13 XVZ I-14 XVII-15 XVII-16 XVIZ-17 XVII-18 XVZI-19 XVII-20 XVII-21 Dispersion and Associated Meteorological Parameters Relation of Satellite and Nine M Point Winds Frequency of Occurrence of Lapse Rates-1963 and 1964 Relation Between Wind Direction Range and Turbulence Classes Stack Characteristics Distribution of Turbulence Class By Sectors Sector Concentrations
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| -1963-64 Sector A Elev.350 Sector Concentrations
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| -1963-64 Sector B Elev.350 Sector Concentrations
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| -1963-64 Sector C Elev.350 Sector Concentrations
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| -1963-64 Sector D, Elev.350 Sector Concentrations
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| -1963-64 Sector D, Elev.350 Sector Concentrations
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| -1963-64 Sector E Elev.350 Sector Concentrations
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| -1963-64 Sector F Elev.350 Sector Concentrations
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| -1963-64 Sector G Elev.350 Sector Concentrations
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| -1963-64 Sector A Ground Height Sector Concentrations
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| -1963-64 Sector B Ground Height Sector Concentrations
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| -1963-64 Sector C Ground Height Sector Concentrations
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| -1963-64 Sector D, Ground Height Sector Concentrations
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| -1963-64 Sector D~Ground Height Sector Concentrations
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| -1963-64 Sector E Ground Height Sector Concentrations
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| -1963-64 Sector F Ground Height ile es XVII-17 XVZI-18 XVII-26 XVII-40 XVZI-56 XVII-64 XVI I-66 XVI I-67 XVII-68 XVII-69 XVI I-7 0 XVZI-71 XVI I-72 XVI I-73 XVII-75 XVII-76 XVII-77 XVII-78 XVIZ-79 XVIZ-80 XVZI-81 UFSAR Revision 12 June 1994 Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)Table Number XVII-22 XVII-23 XVII-24 XVII-25 XVI I-2 6 XVII-27 XVII-28 XVII-29 XVII-30 XVIII-1 Title Sector Concentrations
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| -1963-64 Sector G Ground Height Estimates of the Least Favorable 30 Days in 100 Y'ears Concentrations in the Least Favorable Calendar Month-1963-64 Annual Average Sector Deposition Rates (Vg=0.5 cm/sec)Annual Average Sector Deposition Rates (Vg=2.5 cm/sec)Principal Radionuclides in Gaseous Waste Release Correction Factors to Obtain Adjusted Centerline Dose Rates for Sector Estimates Annual Average Gamma Dose Rates Dilution Calculation for Eastward Currents Based on Water Availability SPDS Parameter Set Pacae XVII-82 xvII-84 XVII-85 XVII-88 xvII-89 XVII-92 XVII-104 XVII-105 XVII-119 XVIII-20 UFSAR Revision 12 June 1994 SECTION I INTRODUCTION AND
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| ==SUMMARY==
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| This report is submitted in accordance with 10 CFR Part 50.71(e)entitled"Periodic Updating of Final Safety Analysis Reports" for Niagara Mohawk Power Corporation's Nine Mile Point Unit 1 Nuclear Station.The Station is located on the southeast shore of Lake Ontario, in Oswego County, New York, 7 miles northeast of the city of Oswego.
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| I-2 A.PRINCIPAL DESIGN CRITERIA The following paragraphs describing the principal design criteria are oriented toward the twenty-seven criteria issued by the USAEC.1.0 General The Station is intended as a high load factor generating facility to be operated as an integral part of the Niagara Mohawk system.The recirculation flow control system described in Section VIII contributes to this objective by providing a relatively fast means for adjusting the Station output over a preselected power range.Overall reliability, routine and periodic test requirements, and other design considerations must also be compatible with this objective.
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| Careful attention has been given.to fabrication procedures and adherence to code requirements.
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| The rigid requirements of specific portions of various codes have been arbitrarily applied to some safety-related systems to ensure quality construction in such cases where the complete code does not apply.For piping, the ASA B31.1-1955 Code was used and where exceptions were taken, safety evaluations were performed to document that an adequate margin of safety was maintained.
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| W Periodic test programs have been developed for required engineered safeguards equipment.
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| These tests cover component testing such as pumps and valves and full system tests, duplicating as closely as possible the accident conditions under which a given system must perform.2.0 Buildin s and Structures The Station plot plan, design and arrangement of the various buildings and structures are described in Section III.Principal structures and equipment which may serve either to prevent accidents or to mitigate their consequences are designed, fabricated and erected in accordance with applicable codes to withstand the most severe earthquake, flooding condition, windstorm, ice condition, temperature and other deleterious natural phenomena which can be expected to occur at the site.'SAEC Press Release 8-252,"General Design Criteria for Nuclear Power Plant Construction Permits," November 22, 1965.Revision 9 June 1991 I-3 3.0 Reactor A direct-cycle boiling water system reactor, described in Section IV, is employed to produce steam (1030 psig in reactor vessel, 956 psig turbine inlet)for use in a steam-driven turbine-generator.
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| The rated thermal output of the reactor is 1850 Mw(t).b.The reactor is fueled with slightly enriched uranium dioxide contained in Zircaloy-clad fuel rods described in Section IV.Selected fuel rods also incorporate small amounts of gadolinium as burnable poison.C~To avoid fuel damage, the Minimum Critical Power Ratio is maintained greater than the Safety Limit Critical Power Ratio.d.The fuel rod cladding is designed to maintain its integrity throughout the anticipated fuel life as described in Section IV.Fission gas release within the rods and other factors affecting design life are considered for the maximum expected burnup.e.The reactor and associated systems are designed so that there is no inherent tendency for undamped oscillations.
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| A stability analysis evaluation is given in Section IV.Heat removal systems are provided which are capable of safely accommodating core decay heat under all credible circumstances, including isolation from the main condenser and loss-of-coolant, from the reactor.Each different system so provided has appropriate redundant features.Independent auxiliary cooling means are provided to cool the reactor under a variety of conditions.
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| The normal auxiliary cooling means during shutdown and refueling is the shutdown cooling system described in Section X-A.A redundant emergency cooling system, described in Section V-E, is provided to remove decay heat in the event the reactor is isolated from the main condenser while still under pressure.Additional cooling capability is also available from the high-pressure coolant injection system and the fire protection system.Revision 9 June 1991 g, h.Redundant and independent core spray systems are provided to cool the core in the event of a loss-of-coolant accident.Automatic depressurization is included to rapidly reduce pressure to assist with core spray operation (see Section VII-A).Operation of the core spray system assures that any metal-water reaction following a postulated loss-of-coolant accident will be limited to less than 1 percent of the Zircaloy clad.Reactivity shutdown capability is provided to make and hold the core adequately subcritical, by control rod action, from any point in the operating cycle and at any temperature down to room temperature, assuming that any one control rod is fully withdrawn and unavailable for use.This capability is demonstrated in Section IV-B.A physical description of the movable control rods is given in Section IV-B.The control rod drive hydraulic system is described in Section X-C.The force available to scram a control rod is approximately 3000 pounds at the beginning of a scram stroke.This is well in excess of the 440-pound force required in the event of fuel channel pinching of the control rod blade during a loss-of-coolant accident as discussed in Section XV.Even with scram accumulator failure a force of at least 1100 pounds from reactor pressure acting alone is available with reactor pressures in excess of 800 psig.Redundant reactivity shutdown capability is provided independent of normal reactivity control provisions.
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| This system has the capability, as shown in Section VII-C, to bring the reactor to a cold shutdown condition, K f<0.97, at any time in the core life, independenAf the control rod system capabilities-.
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| A flow restrictor in the main steam line limits coolant loss from the reactor vessel in the event of a main steam line break (Section VII-F).4.0 Reactor Vessel a 0 The reactor core and vessel are designed to accommodate tripping of the turbine generator, loss of power to the reactor recirculation system and other transients and maneuvers which can be expected without compromising safety and without fuel damage.
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| Nine Mile Point Unit 1 FSAR A bypass system having a capacity of approximately 40 percent of turbine steam flow for the throttle valves wide open condition partially mitigates the effects of sudden load rejection.
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| This and other transients and maneuvers which have been analyzed are detailed in Section XV.b.Separate systems to prevent serious reactor coolant system overpressure are incorporated in the design.These include an overpressure scram, solenoid-actuated relief valves, safety valves and the turbine bypass system.An analysis of the adequacy of reactor coolant.system pressure relief devices is included in Section V-C.c~Power excursions which could result from any credible reactivity addition accident will not cause damage, either by motion or rupture, to the pressure vessel or impair operation of required safeguards systems.The magnitude of credible reactivity addition accidents is curtailed by control rod velocity limiters (Section VII-D), by a control rod housing support structure (Section VII-E), and by procedural controls supplemented by a rod worth minimizer (Section VIII-C).Power excursion analyses for control rod dropout accidents are included in Section XV.d.The reactor vessel=will not be substantially pressurized until the vessel wall temperature is in excess of NDTT+60'F.The initial NDTT of the reactor vessel material is no greater than.40'F.
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| The change of NDTT with radiation exposure has been evaluated in accordance with Regulatory Guide 1.99, Revision 2.Vessel material surveillance samples are located within the reactor vessel to permit periodic verification of material properties with exposure.5.0 Containment a~The primary containment, including the drywell, pressure suppression chamber, associated access openings and penetrations, is designed,.fabricated and erected to accommodate, without failure, the pressures and temperatures resulting from or subsequent to the double-ended rupture or equivalent failure of any coolant pipe within the drywell.UFSAR Revision I-5 June 1993 I-6 The primary containment is designed to accommodate the pressures following a loss-of-coolant accident including the generation of hydrogen from a metal-water reaction.Pressure transients including hydrogen effects are presented in section XV.The initial NDTT for the.primary containment system is about-20F and is not expected to increase during the lifetime of the Station.These structures are described in Section VI-A, B and C.Additional details, particularly those related to design and fabrication are included in Section XVI.Provisions are made for the removal of heat from within the primary containment, for reasonable protection of the containment from fluid jets or missiles and such other measures as may be necessary to maintain the integrity of the containment system as long as necessary following a loss-of-coolant accident.I<Redundant containment spray systems, described in Section VII, pump water from the suppression chamber through independent heat exchangers to spray nozzles which discharge into the drywell and suppression chamber.Water sprayed into the drywell is returned by gravity to the suppression chamber to complete the cooling cycle.Studies performed to verify the capability of=the containment system to withstand potential fluid jets and missiles are summarized in Section XVI.Provision is made for periodic integrated leakage rate tests to be performed during each refueling and maintenance outage.Provision is also made for leak testing penetrations and access openings and for periodically demonstrating the integrity of the reactor building.These provisions are all describejl.
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| in.Section VI-F.The containment system and all other necessary engineered safeguards are designed and maintained such that off-site doses resulting from postulated.accidents are below the values stated in 10 CFR 100.Thc analysis results are detailed in Section XV.Double isolation valves are provided on all lines directly entering the primary containment freespace or penetrating the primary containment and connected to the reactor coolant system.Periodic testing of these valves will assure their capability to isola':c at all times.The isolation valve system is discussed in detail in Section VI-D.
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| I-7 The reactor building provides secondary containment when the pressure suppression system is in service and serves as the primary containment barrier during periods when the pressure suppression system.,is open, such as during refueling.
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| This structure is described in Section VI-C.An emergency ventilation system (Section VII-H)provides a means for controlled release of halogens and particulates via filters from the reactor building to the stack under accident conditions.
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| 6.0 Control and Instrumentation a~b.The Station is provided with a control room (Section III-B)which has adequate shielding and other emergency features to permit occupancy during all credible accident situations.
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| Interlocks or other protective features are provided to augment the reliability of procedural controls in preventing serious accidents.
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| Interlock systems are provided which block or prevent rod withdrawal from a multitude of abnormal conditions.
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| The-control rod block logic is shown in Figures VIII-6 and VIII-S, respectively, for the SRM and IRM neutron instrumentation.
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| In the power range, APRM instrumentation provides both control rod and recirculation flow control blocks, as shown in Figure VIII-14.Reactivity excursions involving the control rods are either prevented or their consequences substantially mitigated by a control rod worth minimizer (Section VIII-C.4.0) which supplements procedural controls in avoiding patterns of high rod worths, an LPRM neutron monitoring and alarm system (Section VIII-C.1.1.3) and a control rod position indicating system (Section IV-B.6.0)both of which enable the operator to observe rod movement, thus verifying his actions.A control rod overtravel position light verifies that the blade is coupled to a withdrawn control rod drive.A refueling platform operation interlock is discussed in Section XV, Refueling Accident, which, along with other procedures and supplemented by automatic interlocks, serves to prevent criticality accidents in the refueling mode.
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| c A cold water addition reactivity excursion is prevented by the procedures and interlocks described in Section XV, Startup of Cold Recirculation Loop (Transient Analysis).
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| Security (keycard and alarms)and procedural controls for the drywell and reactor building airlocks are provided to ensure that containment integrity is maintained.
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| A reliable, dual logic channel reactor protection system described in Section VIII-A is provided to automatically initiate appropriate action whenever various parameters exceed preset limits.Each logic channel contains two subchannels with completely independent sensors, each capable of tripping the logic channel.A trip of one-of-two subchannels in each logic channel results in a reactor scram.The trip in each logic channel may occur from unrelated parameters, i.e., high neutron flux in one logic channel coupled with high-pressure in the other logic channel will result in a scram.The reactor protective system circuitry fails in a direction to cause a reactor scram in the event of loss of power or loss of air supply to the scram solenoid valves.Periodic testing and calibration of individual subchannels is performed to assure system reliability.
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| The ability of the reactor protection system to safely terminate a variety of Station malfunctions is demonstrated in Section XV.d.Redundant sensors and circuitry are provided for the actuation of all equipment required to function under post-accident conditions.
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| This redundancy is described in the various sections of the text discussing system design., 7.0 Electrical Power Sufficient normal and standby auxiliary sources.of electrical power are provided to assure a capability for prompt shutdown and continued maintenance of the Station in a safe condition under all credible circumstances.
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| These features are discussed in Section IX.8.0 Radioactive Waste Dis osal a~Gaseous, liquid and solid waste disposal facilities are designed so that discharge of effluents are in accordance with 10 CFR 20 and 1Q CFR 50, Appendix I.The facility descriptions are given in Section XII-A while the development of appropriate limits is covered in Section II.
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| I-9 9.0 b.Gaseous dischax'ge from the Station is appropriately monitored, as discussed in Section VIII and automatic isolation features are incorporated to maintain releases below the limits of 10 CFR 20 and 10 CFR 50, Appendix I.Shieldin and Access Contxol Radiation shielding and access control patterns are such that doses will be less than those specified in 10 CFR 20.These features are described in Section XII-B.10.0 Fuel Handlin and Stora e Appropriate fuel handling and storage facilities which preclude accidental criticality and provide adequate cooling for spent fuel are described in Section X.
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| Nine Mile Point Unit 1 FSAR B.CHARACTERISTICS The following is a summary of design and operating characteristics.
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| 1.0 Site Location Size of Site Site and Station Ownership Net Electrical Output 2.0 Reactor Oswego County, New York State 900 Acres Niagara Mohawk Power Corp.615 MW (Maximum)Reference Rated Thermal Output Dome Pressure Turbine Inlet Pressure Total Core Coolant Flow Rate Steam Flow Rate 3.0 Core Circumscribed Core Diameter Active Core Height+Assembly 4.0 Fuel Assembl 1850 MW 1030 psig 956 psig 67.5 x 10~lb/hr 7.29 x 10'lb/hr 167.16 in 171.125 in Number of Fuel Assemblies Fuel Rod Array Fuel Rod Pitch Cladding Material Fuel Material Active Fuel Length Cladding Outside Diameter Cladding Thickness Fuel Channel Material 532 SRLR('i Reference 3 Reference 3 UO~and UO~-Gdz03 Reference 3>Reference 3 Reference 3 Reference 3 (4)GE Fuel Bundle Designs, General Electric Company Proprietary, NEDE-31152P, February 1993.GENE 23A7170, Revision 3,"Supplemental Reload Licensing Report for NMP1, Reload 12, Cycle 11," May 1994.UFSAR Revision 13 I-10 June 1995 Nine Mile Point Unit 1 FSAR 5.0 Control S stem Number of Movable Control Rods Shape of Movable Control Rods Pitch of Movable Control Rods Control Material in Movable Control Rods Type of Control Drives Control of Reactor Output 129 Cruciform 12.0 in B4C-70%Theoretical Density;Hafnium Bottom Entry, Hydraulic Actuated Movement of Control Rods and Variation of Coolant Flow Rate 6.0 Core Desi n and 0 eratin Conditio s 7.0 Maximum Linear Heat Generation Rate Heat Transfer Surface Area Average Heat Flux-Rated Power Xnitial Critical Power Ratio for Most Limiting Transients Core Average Void Fraction-Coolant within Assemblies Core Average Exit Quality-Coolant within Assemblies Desi n Power Peakin Core Operating Limits Report 50I 496 119,830 Btu/hr-ft~
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| Core Operating Limits Report 0.280 10.715%Factor Total Peaking Factor P8x8R-3.00 GE8x8EB-2.90 GE11-2.94(')-2.62+Maximum total peaking factor for the portion of the bundle containing part length rods (lattice types 1522, 1519 and 1520 per Reference 3).Maximum total peaking factor for the region above the part length rods (lattice types 1521, 1523 and 1524 per Reference 3).UFSAR Revision 13 I-11 June 1995 Nine Mile Point Unit 1 FSAR 8.0 Nuclear Desi n Data Average Initial Volume Metric Enrichment Beginning of Cycle 11-Core Effective Multiplication and Control System Worth-No Voids, 20C<4>Uncontrolled Fully Controlled Strongest Control Rod Out Reference 3 1.098 0.958 0.981 Standby Liquid Control System Capability:
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| Shutdown Margin (cR)20C Xe on Free SRLR~~SRLR~~9.0 eactor Vessel Inside Diameter Internal Height Design Pressure 17 ft-9 in 63 ft-10 in 1250 psig at 575F 10.0 Coolant Recirculation Loo s Location of Recircu-lation Loops Number of Recircula-tion Loops and Pumps Pipe Size 11.0 Primar Containment Containment'rywe11 28 in Type Design Pressure of Drywell Vessel Design Pressure of Suppression Chamber Vessel Design Leakage Rate Pressure Suppression 62 psig 35 psig 0.5 weight percent per day at 35 psig UFSAR Revision 13 I-,12 June 1995 Nine Mile Point Unit 1 FSAR 12.0 Secondar Containment Type Internal Design Pressure Design Leakage Rate Reinforced concrete and steel superstructure with metal siding 40 lb/ft'00%free volume per day discharged via stack while maintaining 0.25-in water negative pressure in the reactor building relative to atmosphere 13.0 Structural Desi n Seismic Ground Acceleration Sustained Wind Loading Control Room Shielding 0.11g 125 mph, 300 ft above ground level Dose not to exceed hourly equivalent (based on 40-hr week)of maximum permissible cgxarterly dose specified in 10CFR20 14.0 Station Electrical S stem Incoming Power Sources Outgoing Power Lines Onsite Power Sources Provided Two 115-kV transmission lines Two 345-kV transmission lines Two diesel generators Two safety-related Station I batteries One nonsafety 125-V dc battery system 15.0 Reactor Instrumentation S stem Location of Neutron Monitor Sensors In-core Ranges of Nuclear Instrumentation:
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| Four Startup Range Monitors Eight Intermediate Range Monitors 120 Power Range Monitors Source to 0.01%rated power and to 10%with chamber retraction 0.0003%to 10%rated power 1%to 125%rated power UFSAR Revision 12 I-13 June 1994 16.0 Reactor Protection S stem Number of Channels in Reactor Protection Spstem Number of Channels Reguired to Scram or Effect Other Protective Functions Number of Sensors per Monitored Variable in each Channel (Minimum for scram function)
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| IDENTIFICATION OF CONTRACTORS The General Electric Company was engaged to design, fabricate and deliver the nuclear steam supply system, turbine-generator and other major elements and systems.General Electric also furnished the complete core design and nuclear fuel supply for the initial core and is currently furnishing replacement cores.Niagara Mohawk Power Corporation, acting as its own architect engineer, specified and procured the remaining systems and components including the pressure suppression containment system, and coordinated the complete integrated Station.Stone and Webster Engineering Corporation was engaged by Niagara Mohawk to manage field construction.
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| Currently, Niagara Mohawk utilizes various contractors to assist in continuous station modifications.
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| GENERAL CONCLUSIONS The favorable site characteristics, criteria and design requirements of all the systems related to safety, the potential consequences of postulated accidents and the technical competence of the applicant and its contractors assure that the Nine Mile Point Unit 1 Nuclear Station can be operated without endangering the health and safety of the public.}}
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