|
|
| Line 17: |
Line 17: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) | | {{#Wiki_filter:}} |
| ACCESSION NBR:9607020103 DOC.DATE: 96/06/28 NOTARIZED: YES DOCKET FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara owe 05000220 AUTH. NAME AUTHOR AFFILIATION SYLVIA,B.R.
| |
| RECAP.NAME
| |
| ; Niagara Mohawk Power Corp.
| |
| RECIPIENT AFFILIATION /ofp Document Control Branch (Document Control Desk)
| |
| | |
| ==SUBJECT:==
| |
| 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 TITLE: OR Submittal: Updated COPIES RECEIVED!LTR FSAR I ENCL (50.71) and Amendments J I SIZE:
| |
| NOTES:
| |
| RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 0 HOOD,D 1 1 INTERNAL: AYERS 2 2 AEOD/DOA/IRB 1 1 FILE CENTER 01 '2 2 RGN1 1 1 EXTERNAL: IHS 1 1 NOAC 1 1 NRC PDR 1 1 D
| |
| U N
| |
| NOTE TO ALL "RIDS" RECIPIENTS:
| |
| 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 NIAGARA MOHAWK C EN ERATION 300 ERIE BOULEVARDWEST. SYRACUSE, NEW YORK 1 3202/TELEPHONE (31 5) 4284983 BUSINESS CROUP June 28, 1996 B. RALPH SYLVIA NMP1L 1090 Executive Vice President Generation Business Group Chief Nuclear Officer U. S. Nuclear Regulatory Commission 10 C.F.R. 550.71(e)
| |
| Attn: Document Control Desk 10 C.F.R. $ 50.54(a)(3)
| |
| Washington, DC 20555 10 C.F.R. $ 50.59(b)(2)
| |
| RE: Nine Mile Point Unit 1 Docket No. 50-220
| |
| | |
| ==Subject:==
| |
| 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:
| |
| 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).
| |
| 9607020i03 960628 PDR ADQCK'5000220 K PDR
| |
| | |
| Page 2 The enclosed annual Safety Evaluation Summary Report (Enclosure B) contains brief descriptions of changes to the facility design, procedures, tests, and experiments. None of the Safety Evaluations involved an unreviewed safety question as defined in 10 C.F.R. $ 50.59(a)(2).
| |
| 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 300 ERIE BOULEVARDWEST. SYRACUSE, NEW YORK 13202/TELEPHONE (3 1 5) 428-6983 BUSINESS GROUP June 28, 1996 B. RALPH SYLVIA Executive Vice President NMP1L 1090 Generation Business Group Chief Nuclear Officer U. S. Nuclear Regulatory Commission '0 C.F.R. $ 50.71(e),
| |
| Attn: Document Control Desk 10 C.F.R. 550.54(a)(3)
| |
| Washington, DC 20555 10 C.F.R. $ 50.59(b)(2)
| |
| RE: Nine Mile Point Unit 1 Docket No. 50-220
| |
| | |
| ==Subject:==
| |
| Submittal of Revision I4 to the Nine Mile Point Nuclear Station Unit Final I 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:
| |
| 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).
| |
| | |
| Page 2 The enclosed annual Safety Evaluation Summary Report (Enclosure B) contains brief descriptions of changes to the facility design, procedures, tests, and experiments. None of the Safety Evaluations involved an unreviewed safety question as defined in 10 C.F.R. $ 50.59(a)(2).
| |
| 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 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Niagara Mohawk Power Corporation Docket No. 50-220 (Nine Mle Point Unit 1)
| |
| 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.
| |
| 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 LtNAM.'LANDERS Notery Public, State of New Yoh Registration No. i908015 Quahfied fn jefferson Cool nq Conrrnission Expires October 13, 19 r
| |
| | |
| ENCLOSE' IDENTIFICATIONOF CHANGES, REASONS Al'6) BASES FOR N)PC-QATR-1 (UFSAR APPENDIX B)
| |
| | |
| ENCLOSURE A IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR QA PROGRAM DESCRIPTION CHANGES (UNIT I UFSAR APPENDIX B)
| |
| =,Si'ih:for,"Con'dttdlnj'.that,."the':Revised;Prig''am::-",
| |
| ';:;":;-'.UFSARIA'ppe'ndh't::,B:,,:.". to'Stttis~fy)10CFRSO;'Appendh't'8''iiid",::,+3"," -"',".',Gi'ntin'iies
| |
| :';:.,',"::.;';;'::;::;:'Page'/SeetIon':.,-'':':,'',::::,::",,::.::-:,:l ";::,Contitihn'en'::PreyIously,'.'.A'p'pr'o'y'ed;by": the':NRC::.'::',:";.',
| |
| Page B.0-1, third paragraph Changed "Executive Vice President Reorganization. Reorganization approved by the NRC via Unit 1 Nuclear" to "Chief Nuclear Of5cer" License Amendment 157 and Unit 2 License Amendment 71, dated Febnuuy 20, 1996.
| |
| Page B. 1-1, Section B.l. 1, Replaced "contractors and consultants" Editorial. NMPC uses the term The use of an all~compassing term (i.e., using first paragraph with "suppliers" "suppliers" as a synonym of "contractors" "suppliers" to include or describe contractors, and "consultants," and prefers the term consultants, or vendors) does not affect compliance "suppliers." with 10CFR50 Appendix B.
| |
| Page B. 1-1, Section B.l. 1, Changed "Each organizational department, Editorial. Editorial. N/A second paragraph 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."
| |
| Page B. 1-1, Section a. Changed "is delegated by the EditoriaL To reflect that the a. Editorial. N/A B.1.2.1 President to corporate offlcers, as authority and responsibility of the described herein" to read "is Manager Quality Assurance is delegated by the President to also described.
| |
| corporate oKcers and the Manager Quality Assurance, as described herein"
| |
| : b. Changed "Figure 13.1-1a" to read b. Editorial. b. Editorial. N/A "Figure 13.1-1"
| |
| | |
| "',:::%.""UFSAR''Appendix':B;:.:,,';;:'':,:
| |
| '":,;"-",".;;: Page/Section';:"';,':.;"::;"-
| |
| Page B. 1-2, Section a. Changed "Executive Vice a. Reorganization. a. Reorganization approved by NRC via Unit B.1.2.1.1, first paragraph President Nuclear" to "Chief 1 License Amendment 157 and Unit 2 Nuclear Officer" License Antendment 71, dated Febnuuy 20, 1996.
| |
| Changed "...including all b. Reorganization. Same as Item a.
| |
| 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,..."
| |
| Page B.1-2, Section Changed "Controller Nuclear Division" to Reorganization. Title change. Position title change to reflect management of B.1.2.1.1, second "General Manager Business Management" business computers and firuume/accounting activities paragraph under the position of General Manager Business Management. This is an administrative management position which does not perform QA related activities.
| |
| | |
| '.Bi'sls, for,'.,Co'n'eluding that;thee':Revtsed,&iograiiii::.-':::;:>,,
| |
| ".-;;:",.;UFSAR'"'Appendix 8"':i,"".;:".': NConIlriues'hi,Satisfy.:10CFR50!Appendix B;and'";i.-;.:.
| |
| '-,".-"",:,:.'-::.,:,:PagelSectlon<'-',-'.-.',:l,-,:,l '..',;-';:::",",$'"',j'-"IderitIficatlon'oE;Cliaiige::.'' ""i'."i!,;.'::;-':.':::-;::;,"",:.'.;:!'Reason'for,,Change':".'.::;i,"-:::--''"'-:."'::.'-::",Cotniiiitd1ents Previously',Approved by'.the NRCg~:;
| |
| Page B.1-2, Section Changed "The Vice President Nuclear Corporate management reorganization. Reorganization approved by NRC via Unit 1 License B.1.2.1.1, Item 1 Generation reports to the Executive Vice Amendment 157 and Unit 2 License Amendment 71, President Nuclear, and is responsible for dated February 20, 1996.
| |
| safe and efficient operation, maintenance, and modification of the Station in compliance with Station licenses, applicable regulations, and the QA Program. The Vice President Nuclear Generation delegates to the Plant Managers and other appropriate personnel authority for perfonnance in accordance with the QA Program. See Table B-1 for QA Program element responsibilities.
| |
| Activities performed under the responsibility of the Vice President Nuclear Generation include:" to read "The Vice President and General Manager-Nuclear reports to the Chief Nuclear Officer, and has the overaH divisional responsibility for plant operation and engineering. The Vice President Nuclear Engineering, Plant Managers, and the General Supervisor Labor Relations report directly to this Vice President. See Table B-1 for QA Program element responsibilities. Activities performed under the responsibility of the Vice President and General Manager - Nuclear include:"
| |
| | |
| l IBash for".,'Co'ndudliig that. the'.Revfsed,Program"'. ",::".;''.:
| |
| i";":::::P UFSAR:'Appendh B.;:.;:$:;,"
| |
| ;Contliities'to,'Sathfy,';.jOCFR50;Appendht.8:aiilq.';,'':
| |
| <'P<!'':;.';:;::Ide'i'itHication'ofChang'e,"-,<:,.",::;"::::.,".,::'.,:",:,,,;': ~ Comntitm'ents',PrevIon'slyApproved,by',the'NRC!'-:::.v.,'age B.1-3, Item 2 Changed Item 2 to read: Corporate management reorganization. Reorganization approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Amendmettt 71, "Responsibilities and duties of the Vice dated Febrmuy 20, 1996.
| |
| 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."
| |
| Page B.1-3, Item 3 Changed Item 3 to read: Corporate management reorganization. Reorganization approved by NRC via Unit 1 License Amendment 157 and Unit 2 License Amendment 71, "The Vice President Nuclear Safety dated February 20, 1996.
| |
| 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. Nuclear Safety Assessment and Support responsibilities are described in Unit 1 UFSAR Section XIH and Unit 2 USAR Section 13."
| |
| Page B.1-3, Item 4, first Changed "Executive Vice President Corporate management reorganization. Reorganization approved by NRC via Unit 1 License paragraph Nuclear" to read "Chief Nuclear Officer" Amadment 157 and Unit 2 License Augment 71, dated February 20, 1996.
| |
| | |
| ,I
| |
| ;:.Siishi'forConcluding tImt'thi"Revtse'd
| |
| ',:'~'~UFSAR'!Appendli,B.~,'-";,;:; B;ttnd"':g>:;:-
| |
| Piogrt'un'.'s,":...'.;:Con/biue's,,tYi:Sibsfy,'.10CFR50'Appendix
| |
| 'CotttrnItiiients'PievIottsly:Approv'ed by'the'.NRC:,"'-'-':
| |
| =.j:",: ~+:Page/Sectlon.:'..;,::;::.':;;,;'-:,',-, i,"::;;.:,';;";':;'.".,::,IdetItilici Page B.1-3, Item 4, second a. Changed "Tasks performed to Editorial. a. Editorial. N/A paragraph fulfillthese responsibilities include" to read "Tasks performed to fulfillthese responsibilities are delineated in site procedures and include" Combined Inspections and NDE b. Editorial. Many of these tasks b. Editorial. N/A Examinations as one task and are also described under the removed the following identified responsibilities of supervisors or tasks: in other sections of the QATR.
| |
| ~ Coordinating and Reporting Administering the Evaluation and Internal and External QA Corrective Action Program for Assessments Deviation/Event Reports (DERs)
| |
| ~ Operations Experience is the responsibility of the Plant A88essnlent Managers.
| |
| ~ Administering the Evaluation and Corrective Action Program for Deviation Event Reports (DERs)
| |
| ~ DER Trend Analysis
| |
| ~ Preparing and Processing QA Organization Documents C, Added the following tasks: c. Reorganization. c. Reorganization approved by NRC via letter
| |
| ~ Records Management dated July 13, 1995.
| |
| ~ Document Control
| |
| | |
| iBisls,t'or,'Coaduding'that the. Revtse'0 Piogiasnt"-'"..'."'.'.-
| |
| '; ':"':":jUFSAR'.Appeii'dbr'8-:.'',:;:::: ~Coritliiiies.,'to,Siibsfy.';10CFR50'Appendix:B,and ':;:;"::::;>
| |
| :,';:.-,'::;:,'''.",IderiIdicaIIon ofChange+~,;:;,:;:::~:,:.;-;,-.::, ,".,",';-'Reison foi",Cliaiigi.",;";."..''',:.","."..'':':.'';"',i:':, lCon'ti'nitntents Pk'vionsiy".Appr'oved bythe NRC.":,=;";
| |
| Page B. 14, Item 4.b Changed "...determining applicability of Editorial clarification. Incorporates the Editorial. N/A industry and in-plant experience" to read term "assessments" associated with
| |
| "~ ..assessments determining applicability of operations experience assessments into industry and in-plant operating experience" supervisor responsibilities.
| |
| Page B.IQ, Item 4.d Changed "...performing source Editorial clarification. Qualifies the type Editorial. N/A surveillances of selected procurements" to of activities that are done by the read "...performing supplier evaluations Procurement Quality Assurance Group.
| |
| and source surveillances of selected procurements" Page B. 14, Item 4.e Added Item 4.e to describe the Reorganization established new position of Reorganization approved by NRC via letter dated responsibilities of the General Supervisor Supervisor Quality Services which was July 13, 1995. Title change is administrative in Quality Services later changed to General Supervisor nature and does not affect position functions or Quality Services. responsibiTities.
| |
| Page B.2-2, Section Changed "Executive Vice President Reorganization title change. Reorganization reviewed and approved by NRC via B.2.2.3, first paragraph Nuclear" to 'Chief Nuclear OKcer" Unit 1 License Amendment 157 and Unit 2 License Amendment 71, dated February'20, 1996.
| |
| | |
| 'J I
| |
| | |
| 4Basb" for',,Coiidiidbig that the.'Revfsed,Piog'ram'!"-.;,,''",'--Continues'to Satbfy':,10CN50;Appen'dix'.B:arid:.'-',5$
| |
| ';':".<<;.:-',',": '"',.Id ='Commitinents Pr'evIously,"Appi'oved by,.the NRC:.;-.
| |
| Page B.2-5, Section Changed wording from "The Manager Clarification and reorganization title The Chief Nuclear OKcer reports to the President of B.2.2.15, Item 1 Quality Assurance is responsible for change. Although the Manager Quality NMPC as described in Unit 1 License Amendment reporting on the status, adequacy and Assurance is responsible for reporting on 157 and Unit 2 License Anieixlnient 71, dated effectiveness of the NMPC QA Program the status, adequacy and effectiveness of February 20, 1996.
| |
| through the Nuclear Division Internal the QA Program to the Chief Nuclear SALP Type Assessment Reports" to read Officer, it is the Chief Nuclear Officer "The Chief Nuclear Officer is responsible that reports to the President or Chief for reporting on the status, adequacy and Executive Officer (CEO).
| |
| effectiveness of the NMPC QA Program" Page B.2-5, Section Changed "Executive Vice President Reorganization title change. Reorganization of corporate management approved Nuclear" to "Chief Nuclear Officer" by NRC via Unit 1 License Amendmerit 157 and B.2.2.15, Item 2 Unit 2 License Amendmeiit 71, dated Febnuuy 20, 1996.
| |
| Page B.24, Section Changed wording from "The SRAB is a Clarification. To more clearly reflect The change more clearly reflects Plant Technical B.2.2.16 standing committee chaired by the Vice Plant Administrative Technical Specifications, Administrative Controls Section, and President Nuclear Engineering and reports Specifications. also reflects corporate management position title to the Executive Vice President Nuclear changes associated with Unit I License Amendment regarding designated QA functions at the 157 and Unit 2 License Amendment 71.
| |
| 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" Page B.24, Section Changed wording from "The SORC is an Clarification. To more clearly reflect The change more clearly reflects Plant Technical B.2.2.17 independent review committee responsible Plant Administrative Technical Specifications, Administrative Controls Section, and to the Vice President Nuclear Generation Specifications. also reflects corporate management position title and transmits reports to the SRAB" to read changes associated with Unit 1 License Amendment "The SORC is an independent review 157 and Unit 2 License Amendnient 71.
| |
| committee responsible to the Plant Managers and transmits reports to the SRAB"
| |
| | |
| ''":;,j,:UFSAR;:Appe'adh'r",B::9'i::.".: ',Cori(i'ues:,to',Satisfy,'10CPR50'Appeadh;B.'iiil"".4i>
| |
| I-'.,>,::::.-,'';~,,',,",'.:.:.:'.,Ideiitific>>aIIon'of,Change;,',,~;-'.,"".,s'..''",',':.: .,":,::i,;:,;'.:;:'-;,'..:':='."';.';;.Reasori;for,,:Changers,-;..-'":,';;;:-::>":: -,-.".
| |
| Commitments Prevloiisly",.":A'jipx'oved by,': the',NRC-">>."
| |
| Pago B.24, Section Changed wording from "...and actions are Clarification. While Q1P personnel verify The overall indepeadence and confidentiality of Q1P B.2.2.18 verified by Q1P personnel prior to the overall closure of all items, other have not changed. The technical abiTity of other closeout" to read "...and the actions are groups may be used to do some of the departmeats is used to review somo of tho concerns.
| |
| verified prior to closeout" actual technical verifications for completeness.
| |
| Page B.4-2, Section Chaaged wording from "NQA or Clarification. There is no specific repireaiont for any particular B.4.2.7 Procurement personnel other than the group to perform these reviews, only that the person who generated the procurement individuals doing the review adecpatoly understand dociiment, but cpalifiod in QA,..." to red the mpirements and intent of tho procurement "Personnel other than the person who documeats. This is in accordance with NQA-l, generated the procurement document, but 4S-1, section 3, which is our stated program for with adecpate uaderstaiding of the meeting 10CFR50 Appendix B. This does not recpirements and intent of the procurement constitute a reduction of commitment since whoever documents,..." does the review fuaction is required to be cpalified.
| |
| This qualiT!cation is accomplished through training.
| |
| Pago B.5-1, Section Added Section B.5.2.6 to descnbo As an alteraative to performing procedure NRC approval per 10CFR50.54 granted via letter B.5.2.6 procedure review process reviews no less frequently than every two dated January 30, 1996.
| |
| years to determine ifchanges are necessary or desirable (ANS-3.2).
| |
| 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 Page B.7-1, Section Changed wording from "When contractors Editorial. NMPC uses the term Tho use of an all~passing tenn (i.o., using B.7.2.2 perform work under their own QA "suppliers" as a synonym of "contractors," "suppliers" to include or descnbe contractors, programs..." to read "When suppliers aad prefers tbe term "suppliers." consultants, or vendors) does not affect compliance perform work under their own QA with 10CFR50 Appendix B.
| |
| pfogfaias
| |
| | |
| I
| |
| (
| |
| | |
| ;.,Contlniies,4i'SatlsfyiiOCH60,::Appendix'8 'an'6=;,';.'"'~j UFSAR.:"Appiiidlx,B,"'.";'"-,'..
| |
| ~+~';':-',~IdeiitifiaitIo'iiof,Chaiige'';;:,":;;'.-".:,;;.P,, (Coiniiiitinents Pit.vlo'iisly'!hpproved,by,that. NRC:;.'-:,:.'age B.7-1, Section a. Changed wording from "...result a. Editorial clarification. To reflect a. Editorial. N/A B.7.2.3, Item 1 in the supplier being placed on the use of the term "supplier" rather QualiTied Contractor List than "contractor."
| |
| Database (QCLD) as a qualified vendor" to "...result in the supplier being placed on the Qualified Supplier List Database (QSLD)"
| |
| : b. Changed wording from "...by b. Editorial. b. EditoriaL N/A virtue of this ability" to read
| |
| "...by virtue of their ability" Changed wording from c. Clarification. To fully reflect c Nuclear Engineering is responsible for
| |
| "...characteristics identified by Nuclear Engineering maintaining the design basis of systems, Nuclear Engineering and NQA" responsibilities. structures, and comporlents and tfallslates to read "...Characteristics design requirements to suppliers which are identified by Nuclear deemed critical for a particular Engineering" item/service. The identification of critical manufacturing and functional processes and characteristics by Nuclear Engineering continues to satisfy 10CFR50 Appendix B, Criterion 7.
| |
| Changed wording from d. ClariTication. d. NQA is involved with verification of
| |
| "...methods have been identified supplier programs with attention to critical and documented by which NQA processes/characteristics selected, unless will verify conformance to these they can be verified onsite via test and/or iequlfelnents to read re@methods inspection. These responsibilities have not have been $ Ient tfiled and changed and, therefore, continue to satisfy documented which willverify 10CFRSO hppendix B.
| |
| conformance to these requirements"
| |
| | |
| '.,:::Bash for',Coiicliidlng that the':Revtsed,Prog'rain',-':;;~j,"
| |
| NConttriues.'to';Sahfy':10CFR50:App'endlx.:B:aciid'-.::""4':
| |
| ';"~~":.="-'::::<IdentIfimtloti':ot,.Change/!''-"::.:-':;:.';';;:::::.::::,;;:,:.'.-'.::,'-.:."'%~Reasosn.for,Chatige,",':;:',;:,,'':-:-',::, -.Coi'iimitrneeriti':Prevto'tisly,::Approve'd by,'the NRC Page B.7-2, Section a. Changed wording from "NMPC- The change from 3 yr to 3 years a. Editorial. N/A B.7.2.3, Item 2 qualified suppliers involved in is editorial. The addition of a active procurement are surveyed note to reflect a tolerance of one every 3 yr to maintain..." to read quarter of a year is also editorial "NMPC~ified suppliers as this reflects Regulatory Guide involved in active procurement 1.28, paragraph 3.2, as described are surveyed every 3 years~ to in QATR Table B-3, sheet 1 of 8.
| |
| maintain..."
| |
| "<<With a tolerance of one quarter of a year"
| |
| : b. Changed wording from "Supplier b. Editorial. b. Editorial. N/A 3-yr surveys..." to read "Supplier 3-year surveys" Page B.7-2, Section Added Item 3 to identify Clarification. These statements clarify the use of the National B.7.2.3, Item 3 suppliers/organizations that are not Institute of Shuxlards and Technology and other required to be evaluated or listed on the NRC licensed utilities that meet the requirements of Qualified Supplier List Database (QSLD). 10CFRSO Appendix B.
| |
| Page B.7-3, Section Changed wording from "...purchased in Clarification. These controls are not Although Engineering procedures provide controls to B.7.2.6 accordance with Nuclear Engineering limited to Nuclear Engineering assure that items satisfy design requirements, these Procedures that provide. o to read paicedures. Several types of procedures controls may also be found in Nuclear Interface "purchased in accordance with procedures are used to make sure that design criteria Procedures or department procedures other than that provide..." is included in purchase requirements. Engineering.'hese procedural controls continue to satisfy IOCFRSO Appendix B Criterion 7.
| |
| Page B.9-2, Section Changed wording from '...kept by vendors EditoriaL NMPC uses the term The use of an all~conipassing term (i.e., using B.9.2.9 and/or forwarded to NMPC" to read "suppliers" as a synonym of "vendors," 'suppliers" to include or describe contractors,
| |
| "...kept by suppliers and/or forwarded to and prefers the term 'suppliers." consultants, or vendors) does not affect compliance NMPC" with 10CFR50 Appendix B.
| |
| 10
| |
| | |
| ;".Bash for, Concluding that, the Reytsed,Prograni:" ".:,'-',:
| |
| ;-:;;.~".';UFSAR::Appendix 8;-g;.; ;Coritliiites. 4Sash y,;,10CFR50':Appendh;B aril;.';.':;..-,'',
| |
| g(~:A:,:;."::".-:."Identdici'tlon,of,:Cha'nge=;..:::;::";-.;.'".'-::c.-:.':j';::- ;.ComnIitmeit9 Prevlou'sly':Appr'oved.by.Ihe NRC<"-"-,
| |
| Page B.10-1, Section Replaced policy statement with wording Editorial clarification to reflect wording The restructuring of the paragraph to reflect NQA-1 B.10.1 from NQA-1 provided in 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 Changed from "Hold points and/or witness Witness points are no longer used at Nine The removal of witness points continues to satisfy B.10.2.2, Item 4 points" to read "Hold points" Mile Point. All witness points have been Appendix B, Criterion 10, since those witness points converted to hold points or deleted from that were required have been upgraded to hold NMPC procedures. liits o Page B.10-2, Section Deleted previous Section B.10.2.5, which Witness points are no longer used at Nine The removal of witness points continues to satisfy B.10.2.5 stated "Witness points require sufficient Mile Point. All witness points have been Appendix B, Criterion 10, since those witness points B.10.2,6 notification of the specifying organuation converted to hold points or deleted from that were required have been upgraded to hold B.10.2.7 prior to performance of the specified NMPC procedures.
| |
| B.10.2.8 activity" and renumbered reinaining points'age sections accordingly.
| |
| B.10-3, Section Changed wording from "A program for Editorial clarification for ease of reading Editorial. N/A B.10.2.9 inspection and surveillance of activities aud sentence structure.
| |
| affecting fire protection is established..." to read "A program for inspection and surveillance, as required, for activities affecting fire protection is established..."
| |
| Page B.11-2, Section Changed wonling from "Any witness and Witness points are no longer used at Nine The removal of witness points continues to satisfy B.11.2.3, Item 5 hold points" to read "Any hold points" Mile Point. All witness points have been Appendix B, Criterion 10, since those witness points converted to hold points or deleted from that were required have been upgraded to hold NMPC procedures. polrlts.
| |
| Page B.18-1, Section Changed wonling from "...including those Editorial. NMPC uses the term The use of an allmicompassing term (i.e., using B.18.1 elements of the program implemented by 'suppliers" as a synonym of "contractors," "suppliers" to include or describe contractors, suppliers and contractors" to read and prefers the tenn 'suppliers." consultants, or vendors) does not affect compliance
| |
| 'including those elements of the program with 10CFR50 Appendix B.
| |
| implemented by suppliers" 11
| |
| | |
| IBasIs for,Coiicluding'that'the Revfsed.Prograii'i'...':::;;
| |
| ,'"::i';:UFSAR A'p'peti'dlx.S,:;;".-.';.'-,.; -Contlriue<<s'to Satisfy'10CFR50'App<<eindiIx'B'aiid'..';~";.~
| |
| .,'.':;.'-:)-:.''-,""-PIdeiitificatlon of,Ch'ange';:-,;,,:j,'=,:;:;g,'.<<'::.~;."'~"',",:".;,";."., Rea'son for; Change;..":''j ':'.;.':::.-',::@ >,'.Conimsitinents:Prevlot'isly"'Approv'ed,by, the NRC-"-,";
| |
| Page B.18-1, Section Changed wording from "once every 2 yr" Editorial. Editorial N/A B.18.2.3 to read "once every 2 years" Table B-l, sheet I of 2 a. Under Procedures column, Editorial, to reflect corporate a. Editorial. N/A identiTied Quality Assurance management restmcturing.
| |
| (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).
| |
| : b. Removed Nuclear Procurement b. Reorganization. b. The Nuclear Procurement organization was (NP) from the NSAS Procedures transferred from Nuclear Safety column to reflect transfer of the Assessment and Support (NSAS) to Nuclear Nuclear Procurement function to Engineering. The duties, functions, and Nuclear Engineering. Identified responsibilities of Nuclear Procurement Nuclear Engineering as have not been altered.
| |
| responsible for QA Program elements associated with Criterion IV, accordingly.
| |
| c, Removed Technical Services (TS) co Reorganization. c The duties, responsibilities, and functions and Information Management performed by Technical Services (TS) and (IM) from the NSAS Procedures Information Management (IM) have been cotunm. reassigned to other branches, as appropriate. 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,
| |
| "'.Bi'sh,'fot'. Con'du'din'g'that,'the".Revts'ed Pi'o'g'ram".,",;:.',:,",'~
| |
| +:,",,"..,'"'UFSAR'Ap'pend@ 8'',:"..;;"; ',.::,.ContlntIes,to.Satisfy',:jKFR50,''Appendfix 8;and:,,",:.'"":;;.';
| |
| jCommttmerits',Prevlou'sty'."Appr'oved by,''thi".'RC:;""k Table B.l, sheet 1 of 2 Identified Quality Assurance d. Reorganization. d. Reorganization approved by NRC via letter (cont'd.) responsibiTity for QA Program dated July 13, 1995.
| |
| elements associated with Criteria VI (Document Control) to reflect transfer of responsibility from Nuclear Engineering.
| |
| Table B-l, sheet 2 of 2 Removed Nuclear Procurement a. Reorganization. a. The function of Nuclear Procurement was (NP), Technical Services (TS), transferred fmm the Nuclear Safety and Information Management Assessment and Support organization to (IM), from under NSAS Nuclear Engineering. This transfer does Procedures column and from not affect duties or functional listing of NMPC organizations. 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.
| |
| : b. Identified Quality Assurance b. Reorganization. b. Reorganization approved by NRC via letter responsibility for QA Program dated July 13, 1995.
| |
| elements associated with Criteria XVH (Quality Assurance Records) to reflect transfer of responsibility from Nuclear Engineering.
| |
| Table B-3, sheet 2 of 8 Changed Document column row "d" from EditoriaL Editorial. N/A Para. 4 to read "Section 4" 13
| |
| | |
| 1 Enclosure B to NMP1L 1090 NINE MILE POINT - UNIT 1 SAFETY EVALUATION
| |
| | |
| ==SUMMARY==
| |
| REPORT 1996 Docket No. 50-220 License No. DPR-63
| |
| | |
| Safety Evaluation Summary Report Page 1 of 60 Safety Evaluation No.: 89-019 Rev. 0, 1, 2 8c 3 Implementation Document No.: Mod. N1-89-174 USAR Affected Pages: XI-14 System: Low-Pressure Reactor Feedwater System Title of Change: Low-Pressure Reactor Feedwater System Design Pressure Reduction
| |
| | |
| ==
| |
| Description:==
| |
| | |
| 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. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 2 of 60 Safety Evaluation No.: 91-019 Implementation Document No.: Mod. N1-88-052 UFSAR Affected Pages: N/A System: Spent Fuel Pool Title of Change: 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. 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. Six of the eight racks were installed prior to this planned modification.
| |
| /
| |
| 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. The removal of the work platform uncovered the pool liner which was deformed from a prior installation. 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.
| |
| Following exact information obtained from the mapping, the pedestal was reanalyzed and modified accordingly. 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. This condition was
| |
| | |
| Safety Evaluation Summary Report Page 3 of 60 Safety Evaluation No.: 91-019 (cont'd.)
| |
| Safety Evaluation Summary: (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.
| |
| Prior criticality, thermal-hydraulics, and pool structure analyses will not require reanalyses as a result of this modification. 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.
| |
| | |
| Safety Evaluation Summary Report Page 4 of 60 Safety Evaluation No.: 91-033 Rev. 2 5. 3 Implementation Document No.: Gale. 4.16KVAC-DG-ES, Rev. 2 UFSAR Affected Pages: IX-18, Figure IX-6 System: Emergency Diesel Generator Title of Change: 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.
| |
| Safety Evaluation'ummary:
| |
| Maximum expected diesel generator load has been determined by calculation, compared to equipment capability, and found to be acceptable. Informa'tion previously provided in the UFSAR shows that maximum diesel generator load is within the machine's capability. The diesel generator load analysis has defined the maximum allowable design load based on vendor information and regulatory guidance documents. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 5 of 60 Safety Evaluation No.: 93-056 Rev. 1, 2 & 3 Implementation Document No.: Simple Design Change SC2-0328-92 UFSAR Affected Pages: Figure III-1 System:
| |
| ~ -
| |
| 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 6 of 60 Safety Evaluation No.: 94-004 Implementation Document No.: Simple Design Change SC1-0281-91/
| |
| 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. 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. Zone DA-6063HD is located in the mixing hood of the DOW mixing area, rendering it unnecessary. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 7 of 60 Safety Evaluation No.: 94-007 Implementation Document No.: Simple Design Change SC1-0156-92, SC1-0157-92, SC1-0158-92 UFSAR Affected Pages: Table Vl-1, Figure Vl-22; X-16 System: Containment Title of Change: 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.
| |
| The capped lines will be considered spares and will be subject to Appendix J Type A testing to ensure overall containment integrity.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 8 of 60 Safety Evaluation No.: 94-008 implementation Document No.: Temporary Mod. 94-019 UFSAR Affected Pages: X-34 System: 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 9 of 60 Safety Evaluation No.: 94-013 Implementation Document No.: Mod. N1-93-018 UFSAR Affected Pages: Figure IX-2 System: 24 VDC System Title of Change: Replace 24 VDC Battery Chargers Description of Change:
| |
| 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. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 10 of 60 Safety Evaluation No.: 94-019 Rev. 0 L 1 Implementation Document No.: IVlod. N1-92-005 UFSAR Affected Pages: 10A-96, 10A-114, 10A-118 System: 4160 VAC, 600 VAC and 125 VDC Systems Title of Change: Improve Electrical Coordination Description of Change:
| |
| 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.
| |
| /
| |
| 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
| |
| '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.
| |
| | |
| Safety Evaluation Summary Report Page 11 of 60 Safety Evaluation No.: 94-019 (cont'd.)
| |
| Description of Change: (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. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 12 of 60 Safety Evaluation No.: 94-037 Rev. 1 Implementation Document No.: Mod. N1-89-079 UFSAR Affected Pages: N/A System: Emergency Cooling System No. 39 Title of Change: Emergency Cooling Appendix J Modification Description of Change:
| |
| 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. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| gl Safety Evaluation Summary Report Page 13 of 60 Safety Evaluation No.: 94-038 Rev. 1 Implementation Document No.: Simple Design Change SC1-0151-93 Mod. N1-93-023 UFSAR Affected Pages: N/A
| |
| ~
| |
| System: Main Steam System 4
| |
| Title of Change: 'SIV Brake Installation and Poppet Upgrade Description of Change:
| |
| This modification eliminated repeated leak rate test failures on main steam isolation valves (MSIV) to enhance their testability. 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. The change has been designed to minimize stem susceptibility to cracking, provide improved anti-rotation mechanism, poppet self aligning features and additional hardfacing. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 14 of 60 Safety Evaluation No.: 94-039 Rev. 0 Bc 1 Implementation Document No.: Mod. N1-88-052, Phase IIA UFSAR Affected Pages: N/A System: Spent Fuel Pool Title of Change: 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. 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. 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. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 15 of 60 Safety Evaluation No.: 94-051 Implementation Document No.: Mod. N1-91-021 UFSAR Affected Pages: 10A-118 System: Reactor Instrument Title of Change: Replace ACUREX Description of Change:
| |
| 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.
| |
| Safety Evaluation Summary:
| |
| Nuclear safety is improved by this modification. The RPS circuits are better balanced, thus improving their load characteristics. 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.
| |
| | |
| Safety Evaluation Summary Report Page 16 of 60 Safety Evaluation No.: 94-052 Implementation Document No.: Mod. N1-87-042 UFSAR Affected Pages: VIII-11; 10A-118 System: Plant Process Computer Title of Change: RIS Isolator Replacement Description of Change:
| |
| 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.
| |
| 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.
| |
| 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 17 of 60 Safety Evaluation No.: 94-053 Rev. 0 5 1 Implementation Document No.: Mod. N1-94-002 UFSAR Affected Pages: Figure Vill-4 System: Reactor Recirculation Title of Change: 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. The proposed modification is defined in General Electric's proposal for the reactor recirculation control system enhancements.
| |
| Safety Evaluation Summary:
| |
| The proposed design change is functionally equivalent to the present FG/Bailey actuator configuration. 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.
| |
| | |
| Safety Evaluation Summary Report Page 18 of 60 Safety Evaluation No.: 94-056 Rev. 0, 1, 5 3 Implementation Document No.: Mod. N1-88-153 UFSAR Affected Pages: Table Vl-3a Sh 2 Bc 3; Vll-5, Figure Vll-1; X-2, Figure X-1 System: Shutdown Cooling and Core Spray Systems Title of Change: Containment Isolation Valve (Appendix J)
| |
| Modification - 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. 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. 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.: 94-056 Rev. 0, 1, & 3 (cont'd.)
| |
| Safety Evaluation Summary: (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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 20 of 60 Safety Evaluation No.: 94-057 Rev. 1 Implementation Document No.: Mod. N1-91-009 UFSAR Affected Pages: Figure Vll-3; 10B-58; Table XV-4, System: Containment Spray Title of Change: 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. 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:
| |
| 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.
| |
| | |
| Safety Evaluation Summary Report Page 21 of 60 Safety Evaluation No.: 94-060 Implementation Document No.: Simple Design Change SC1-0087-93 UFSAR Affected Pages: N/A System: Service Water Title of Change: Emergency Service Water Supports Description of Change:
| |
| This simple design change added new supports to the service water inlet and outlet piping to the Reactor Building closed loop cooling heat exchangers. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 22 of 60 Safety Evaluation No.: 94-062 Rev. 0 5. 1 Implementation Document No.: Simple Design Change SC1-0059-94 UFSAR Affected Pages: III-23; 10B-69 System: N/A Title of Change: Foam Room Wall Replacement Description of Change:
| |
| 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. Welds were repaired to DC valve board ¹11 to the existing embedded floor channels. An 8-inch reinforced block wall with a brick veneer was then constructed.
| |
| 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 23 of 60 Safety Evaluation No.: 94-068 Implementation Document No.: Simple Design Change SC1-0062-94 USAR Affected Pages: 10A-118 System: 120 VAC Distribution
| |
| 'DC Title of Change: SC1-0062-94, Installation of Data Acquisition System
| |
| | |
| ==
| |
| Description:==
| |
| | |
| 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 24 of 60 Safety Evaluation No.: 94-069 Implementation Document No.: Simple Design Change SC1-0039-94 USAR Affected Pages: Tables Vl-1, Vl-3b Sh 2; Vll-34, Figure Vll-13 System: Hydrogen/Oxygen Monitoring Title of Change: H,/O~ Sample Line Reduction
| |
| | |
| ==
| |
| Description:==
| |
| | |
| 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, channel ¹11 and two lines, streams A and C for H,/0, channel ¹12. 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~/0 isolation 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.
| |
| 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. Based on the results of numerous containment mixing studies, the response to the design basis event will be unchanged.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 25 of 60 Safety Evaluation No.: 94-070 Implementation Document No.: Simple Design Change SC1-0070-94 UFSAR Affected Pages: 10A-118 System: 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. These cables'were stored in the existing cable trays until final installation. 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. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 26 of 60 Safety Evaluation No.: 94-070 (cont'd.)
| |
| Safety Evaluation Summary: (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.
| |
| | |
| Safety Evaluation Summary Report Page 27 of 60 Safety Evaluation No.: 94-072 Rev. 0 8c 1 Implementation Document No.: Mod. N1-90-041 UFSAR Affected Pages: 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: Core Spray System Title of Change: 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. This includes both normal design basis functions, manual EOP-directed operation, and shutdown cooling system water seal requirements. The impact of separate minimum flow recirculation lines and inboard valve throttling on required core spray injection flows has been calculated.
| |
| Based on the calculation, the core spray system provides adequate flow to satisfy 10CFR50.46. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 28 of 60 Safety Evaluation No.: 94-074 Implementation Document No.: DER 1-94-0224 UFSAR Affected Pages: VI-32
| |
| | |
| System: Containment Spray, Core Spray, Containment Spray Raw Water, Nitrogen Supply System ¹12, Nitrogen Supply System ¹11 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.
| |
| 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 29 of 60 Safety Evaluation No.: 94-075 Implementation Document No.: Mod. N1-93-013 UFSAR Affected Pages: Table IX-1 Sh 1-3; X-18; 10A-115 System: Containment Atmosphere Monitoring System, H~/0, Analyzer Title of Change: Replace H,/0, Monitoring System Description of Change:
| |
| This modification replaced the hydrogen/oxygen (H,/O~) monitoring system, cabinets, and analyzers. The previous system was produced by Beckman instruments and installed prior to Regulatory Guide 1.97 and NUREG-0737 requirements. Due to age and the reclassification as safety related, the maintenance had become extensive. 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.
| |
| /
| |
| Safety Evaluation Summary:
| |
| The analysis clearly demonstrates compliance with all applicable criteria including safety classification, seismic qualification, environmental qualification, power requirements and separation. Calculations have been performed for the transportation time, response time and accuracy and assure compliance with design basis functions and licensing commitments. 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. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 30 of 60 Safety Evaluation No.: 94-079 Implementation Document No.: DER 1-93-0921 UFSAR Affected Pages: VI-21, VI-26, Table Vl-3b Sh 4 System: Traversing In-core Probe (TIP)
| |
| Title of Change: TIP System Containment Isolation - LDCR Description of Change:
| |
| 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. 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. 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..
| |
| | |
| Safety Evaluation Summary Report Page 31 of 60 Safety Evaluation No.: 94-080 Rev. 0 & 1 Implementation Document No.: Mod. N1-94-003 UFSAR Affected Pages: 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: Reactor Vessel Title of Change: Reactor Core Shroud Repair Description of Change:
| |
| 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. '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. 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).
| |
| 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.
| |
| 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.
| |
| 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 33 of 60 Safety Evaluation No.: 95-002 Rev. 0 L 1 Implementation Document No.: Simple Design Change SC1-0067-94 UFSAR Affected Pages: Figure Vll-1 System: Core Spray Title of Change: Core Spray IV Pressure Binding Relief Description of Change:
| |
| 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. 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.
| |
| 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.
| |
| | |
| Safety Evaluation Summary Report Page 34 of 60 Safety Evaluation No.: 95-003 Implementation Document No.: Simple Design Change SC1-0089-94 UFSAR Affected Pages: X-35, X-36, Figure X-9 System: Instrument Air System Title of Change: 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 the valve will close. This enhances availability of the safety-related portion
| |
| 'alve, 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. The added cable is in conduit and, therefore, the fire load in this area is not increased. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 35 of 60 Safety Evaluation No.: 95-007 Implementation Document No.: Mod. N1-94-003 UFSAR Affected Pages: N/A System: RXVE Title of Change: Core Shroud Repair Installation Description of Change:
| |
| 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. 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. Both systems have been evaluated and will continue to perform as designed during and after the repair installation.
| |
| 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.
| |
| | |
| Safety Evaluation Summary Report Page 36 of 60 Safety Evaluation No.: 95-007 (cont'd.)
| |
| Safety Evaluation Summary: (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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 37 of 60 Safety Evaluation No.: 95-009 Implementation Document No.: Simple Design Change SC1-0102-93 UFSAR Affected Pages: 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.
| |
| 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.
| |
| 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.
| |
| Based on the evaluation performed, it is concluded that these changes do not invoive an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 38 of 60 Safety Evaluation No.: 95-010 Implementation Document No.: DER 1-95-0133 UFSAR Affected Pages: X-39 System: Fuel Title of Change: Spent Fuel Storage Description of Change:
| |
| 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.
| |
| | |
| Safety Evaluation Summary Report Page 39 of 60 Safety Evaluation No.: 95-011 Implementation Document No.: 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: Various Title of Change: Operation of NMP1 Reload 13/Cycle 12 Description of Change:
| |
| 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 40 of 60 Safety Evaluation No.: 95-013 Rev. 0 6 1 Implementation Document No.: Mod. N1-94-003 UFSAR Affected Pages: N/A System: Reactor Vessel Title of Change: Reactor Core Shroud Repair Installation Prior to NRC Approval of Adequacy Under 10CFR50.55a(a) (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. 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. 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. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 41 of 60 Safety Evaluation No.: 95-014 Rev. 1 Implementation Document No.: Temporary Mod. 95-009 Temporary Mod. 95-012 UFSAR Affected Pages: N/A System: Reactor Core Spray Title of Change: 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.
| |
| 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.
| |
| | |
| Safety Evaluation Summary Report Page 42 of 60 Safety Evaluation No.: 95-015 Implementation Document No.: N/A UFSAR Affected Pages: 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. 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.
| |
| Based on the evaluation performed, it is concluded that the temporary use of the CO~ cleaning facility does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 43 of 60 Safety Evaluation No.: 95-016 Implementation Document No.: LDCR No. 1-95-UFS-019 UFSAR Affected Pages: 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: N/A Title of Change: 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. 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. Reports are now submitted under these provisions, as determined through the plant procedures governing Deviation/Event Reporting. 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.
| |
| | |
| Safety Evaluation Summary Report Page 44 of 60 Safety Evaluation No.: 95-017 Implementation Document No.: Dwg. B-69007-C UFSAR Affected Pages: VI-21 System: 58; 201.2 Title of Change: LT 58-05 Isolation Capabilities Description of Change:
| |
| LT 58-05 instrument lines do not contain two manual valves. Instead, isolation is provided by a single manual valve and diaphragm seal assemblies.
| |
| Safety Evaluation Summary:
| |
| /
| |
| The diaphragm seal assembly exceeds design ratings of the torus under normal and accident conditions. 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 45 of 60 Safety Evaluation No.: 95-018 Implementation Document No.: Gale. S7-RX340-W01 UFSAR Affected Pages: XVI-70, Table XVI-31 Sh 1 System: N/A Title of Change: 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.
| |
| 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.
| |
| Safety Evaluation Summary:
| |
| The above-referenced documents show that the correct blowout pressure is 45 psf for the pressure relief panels in both buildings. 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.
| |
| | |
| Safety Evaluation Summary Report Page 46 of 60 Safety Evaluation No.: 95-019 Implementation Document No.: Gale. S15-72-F009 UFSAR Affected Pages: X-25, Figure X-6 System: Service Water Title of Change: 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.
| |
| 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. 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.
| |
| | |
| Safety Evaluation Summary Report Page 47 of 60 Safety Evaluation No.: 95-020 Implementation Document No.: DER 1-94-0354 UFSAR Affected Pages: lll-7, III-23, III-24 System: N/A Title of Change: 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 Building.
| |
| : 2. The Administration Building basement has an entrance to the Turbine Building.
| |
| j
| |
| : 3. 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.
| |
| Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 48 of 60 Safety Evaluation No.: 95-021 Implementation Document No.: Procedures GAP-POL-01, NSAS-POL-01 UFSAR Affected Pages: 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: N/A Title of Change: 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.
| |
| 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. The proposed changes do not impact accident or malfunction initiation or consequences.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 49 of 60 Safety Evaluation No.: 95-022 Implementation Document No.: Procedures GAP-POL-01, NSAS-POL-01 UFSAR Affected Pages: XIII-3, XIII-7 System: N/A Title of Change: Reorganization; Changes to GAP-POL-01 &
| |
| 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 & Support Department to the Unit 1 Chemistry Branch of Nuclear Generation.
| |
| 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. The proposed changes do not impact accident or malfunction initiation or consequences.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 50 of 60 Safety Evaluation No.: 95-023 Implementation Document No.: Procedure NSAS-POL-01 UFSAR Affected Pages: Xlll-3, Figure XIII-4 System: N/A Title of Change: 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. 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. 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. The proposed changes do not impact accident or malfunction initiation or consequences.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 51 of 60 Safety Evaluation No.: 95-025 Implementation Document No.: Procedure NSAS-POL-01 UFSAR Affected Pages: Xlll-3, Xlll-4, Figure XIII-4 System: N/A Title of Change: 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.
| |
| /
| |
| Safety Evaluation Summary:
| |
| These procedure changes establish departmental responsibilities and lines of authority, responsibility, and communication for implementation of the procedure program including publication. 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. The proposed changes do not impact accident or malfunction initiation or consequences.
| |
| Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 52 of 60 Safety Evaluation No.: 95-058 Implementation Document No.: GAP-POL-01 UFSAR Affected Pages: N/A System: N/A Title of Change: 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.
| |
| | |
| f Safety valuation Summary Report Page 53 of 60 Safety Evaluation No.: 95-058 (cont'd.)
| |
| Safety Evaluation Summary: (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.
| |
| | |
| Safety Evaluation Summary Report Page 54 of 60 Safety Evaluation No.: 95-102 Rev. 0 5 1 Implementation Document No.: 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: N/A Title of Change: 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.
| |
| ~ "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.
| |
| | |
| Safety Evaluation Summary Report Page 55 of 60 Safety Evaluation No.: 95-102 Rev. 0 & 1 (cont'd.)
| |
| Description of Change: (cont'd.)
| |
| Transfer responsibility for administration and implementation of. the Fire Protection Program from NSAS Technical Services to Unit 1 Operations.
| |
| ~ 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.
| |
| Transfer responsibility for administration of Buildings & Grounds/Facilities Planning activities (a site function) from the NSAS Site Services to Unit 1 Maintenance.
| |
| 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.
| |
| /
| |
| ~ 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.
| |
| | |
| Safety Evaluation Summary Report Page 56 of 60 Safety Evaluation No.: 95-102 Rev. 0 5 1 (cont'd.)
| |
| Safety Evaluation Summary:
| |
| 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. The proposed changes do not impact accident or malfunction initiation or consequences.
| |
| Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 57 of 60 Safety Evaluation No.: 95-106 Implementation Document No.: N/A UFSAR Affected Pages: Figure III-1 System: N/A Title of Change: 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.
| |
| : 1. Carpenter's shop
| |
| : 2. Paint shop
| |
| : 3. Electric fab shop
| |
| : 4. Insulators fab shop
| |
| : 5. 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.
| |
| 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. These buildings have no impact on the previously .
| |
| calculated X/0 values. The design margins for the control room fresh air intakes are not compromised. Location of demolition activities are adequately separated from safety-related systems and structures to preclude any adverse impact from construction activities.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Safety Evaluation Summary Report Page 58 of 60 Safety Evaluation No.: 95-107 implementation Document No.: NTP-TQS-102 UFSAR Affected Pages: N/A System: N/A Title of Change: 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.
| |
| | |
| Safety Evaluation Summary Report Page 59 of 60 Safety Evaluation No.: 95-128 implementation Document No.: N/A UFSAR Affected Pages: N/A System: N/A Title of Change: Aerial Radiation Survey of NMP Site Description of Change:
| |
| 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.
| |
| | |
| Safety Evaluation Summary Report Page 60 of 60 Safety Evaluation No.: 95-130 Implementation Document No.: Nuclear Division Policy POL UFSAR Affected Pages: B.1-2 System: N/A Title of Change: 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. 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.
| |
| Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
| |
| | |
| Page 1 of NRC FORM 35 U.S. NUCLEAR REGULATORY COMMISSION (11-93)
| |
| NRCMD 3.53 T~NSFER ggyy yS'gS' 'coms I.P 73.
| |
| : 1. JOB NUMBER 2. FRC ACCESSION NUMBER 3. DATE JOB RECEIVED 3911 08/13/1997
| |
| : 4. VOLUMEIN CUBIC FEET 5. LOCATION NUMBER(S) 6. DATE ELIGIBLEFOR DESTRUCTION 5 17127 - 17131 10/01/2029
| |
| : 7. DATE ELIGIBLEFOR TRANSFER TO NARA 8. ORIGINATINGOFFICE CODE 9. FACILITYCODE FC/NRR/ [X] AF Q AF-V5 Q AF-V3 Q FRC
| |
| : 10. COMMENTS LICENSE EXPIRES 08/22/2009 G 0 <T< 'e ' 8-Z
| |
| : 11. ORIGINATINGOFFICE(Office/Division/Rranch) 12. FILE CUSTODIAN (Name/Telephone) 13. LOCATION OF RECORDS(ffuffdfng/Room)
| |
| Tyrone Greene TSC3 FC/NRR/ (301)415-6281
| |
| : 14. T(TLR AF RFCARDSFREFS (RefertoNUREG0910) 15.RFCARDSCHFA11LENUMBER (Complete a separate form for each serfes) (Refer to NUREG.0910)
| |
| Nuclear Power Plant Docket Files(E)
| |
| NRCS 02-20.09E
| |
| : 16. CLASSIFICATION OF RECORDS 17. RECORD MEDIUM
| |
| [X] UNCLASSIFIED [X] PAPER Q AUDIOVISUAL+
| |
| Q OTHER(Specify below) Q MICROFORMo Q MACEHNE READABLEo HIGHEST CLASSIFICATION *PROVIDE SPECIFIC MEDIUM
| |
| : 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 ofthe records using items commonly used by the stofffamfNar with thc records.)
| |
| 50-220 NIAGRAMOHAWKPOWER CORP. NINE MILE POINT UNIT 1 NRC FORM 35 (11-93) s
| |
| | |
| 1
| |
| 'I
| |
| | |
| Page 2 of DETAILEDINVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container. lf more space is required, use NRC for 35A, 'Records Transfer Continuation'.
| |
| : 21. CONTAINER NUMBER ~m:M 23. RECORDS DESCRIPTION
| |
| : 24. DATE SPAN (Month/Yrnrs) 50-220 10/1996 - 10/1996 NINEMILEPTI REVISION TO EPIPS (SECOND COPY) 50-220 09/1996 - 09/1996 NINEMILEPTI REVISION TO EPIPS (SECOND COPY) 50-220 08/1996 - 08/1996 NINEMILEPTI REVISION TO EPIPS (SECOND COPY) 50-220 06/1996 - 06/1996 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY) 50-220 04/1996- 04/1996 NINEMILEPI'I REVISION TO EPIPS (SECOND COPY) 50-220 03/1996 - 03/1996 NINEMILEPTI REVISION TO EPIPS (SECOND COPY) 50-220 02/1996 - 02/1996 NINEMILEPTI REVISION TO EPIPS (SECOND COPY)
| |
| : 25. RECORDS LIAISON OFFICFR -Qprg/rrnne and Signa/urn 26. DATE Tyrone Greene 08/13/1997
| |
| : 27. REVIEWING ANALYST..S/gna/nrc 28. DATE Latravetta Lee 08/13/1997 Q
| |
| : 29. DATE TRANSFERRED TO FRC 30. DATE DESTROYED 31. DATE TRANSFERRED TO NARA NRC FORM 35 (11.93)
| |
| | |
| 5
| |
| ~
| |
| | |
| Page 3 of NRC PORN 35 U.S. NUCLEAR REGULATORY COMMISSION II 1-93I NRCMD 3.53 RECORDS TRANSFER Continuation DETAILEDINVENTORY OF RECORDS MINGTRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 21. CONTAINER NUAIBER W~@M 23. RECORDS DESCRIPTION
| |
| : 24. DATE SPAN (Month/Years) 50-220 12/1995 - 12/1995 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY) 50-220 01/1997 - 01/1997 NINEMILEPTI REVISION 35 TO SITE EMERGENCY PLAN AND EPIPS (SECOND COPY) 50-220 05/1996 - 05/1996 NINEMILEPT1 REVISION 34 TO SITE EMERGENCY PLAN AND EPIPS (SECOND COPY) 50-220 04/1996 - 04/1996 NINEMILEPT1 REVISION 33 TO SITE EMERGENCY PLAN AND EPIPS (SECOND COPY) 50-220 06/1992 - 06/1992 NINEMILEPTI EVACUATIONTRAVELTIME ESTIMATES FOR THE NINE MILEPOINT EMERGENCY PLANNING ZONE 50-220 06/1996 - 06/1996 NINEMILEPTI REVISION 14 TO UFSAR (SECOND COPY) 2 50-220 03/1969 - 03/1969 NINEMILEPT1 REPORT TO ACRS
| |
| . NRC FORM 35 (11-93)
| |
| | |
| Page 4 of NRC FORM 35 U.S. NUCLEAR REGULATORY
| |
| /1I 931 COMMISSION'ECORDS NRCMD 3.53 TRANSFER COntinuatiOn DETAILED INVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 21. CONTAlNER 24. DATE SPAN
| |
| : 23. RECORDS DESCRIP1'lON (Month/Yenrs)
| |
| NUSIBER 50-220 05/1993 - 05/1993 NINEMILEPTI ANNUALFINANCIALREPORT 1992 RG&EC 50-220 05/1993 - 05/1993 NINEMILEPT1 ANNUALFINANCIALREPORT 1992 LILCO 50-220 05/1993 - 05/1993 NINEMILEPTI ANNUALFINANCIALREPORT 1992 NYSE&GC 50-220 05/1993 - 05/1993 NINEMILEPT1 ANNUALFINANCIALREPORT 1992 CHG&EC 50-220 05/1993 - 05/1993 NINEMILEPT1 ANNUALFINANCIALREPORT 1992 NMPC 50-220 IV1992 - 11/1992 NINEMILEPT1 ANNUALFINANCIALREPORT 1991 NMPC 50.220 05/1992 - 05/1992 NINEMILEPT1 ANNUALFINANCIALREPORT 1991 RG&EC 50-220 05/1992 - 05/1992 NINEMILEPTI ANNUALFINANCIALREPORT 1991 LILCO NRC FORM 35 (11-93) t
| |
| | |
| I Page 5 of NRC FORM 35 U.S. NUCLEAR REGULATORY COMMISSION
| |
| /I 1.9.Q NRCMD 3.53 RECORDS TRANSFER Continuation DETAILEDINVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 24. DATE SPAN
| |
| : 21. CONTAINER NUSIBER ~%;;M 23. RECORDS DESCRIPI'ION (Month/Y<<nrs) 50-220 05/1992 - 05/1992 NINEMILEPTI ANNUALFINANCIALREPORT 1991 NYSE&GC 50-220 05/1992 - 05/1992 NINEMILEPTI ANNUALFINANCIALREPORT 1991 CHG&EC 50-220 06/1982 06/1982 NINEMILEPTI SAFETY EVALUATIONOF THREE TMI ACTION PLAN ITEMS (NUREG-0737)> H.F.'1.4, H.F.1.5 AND H.F.1.6, BY CONTAINMENT 50-220 06/1981 - 06/1981 NINEMILEPT1 SAFETYEVALUATIONREPORT 50-220 07/1979 - 07/1979 NINEMILEPTI SAFETY EVALUATIONREPORT REGARDING FIRE PROTECTION 50-220 06/1993 - 06/1993 NINEMILEPT1 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS REGARDING ADMINISTRATIVE CHANGES 50-220 05/1993 - 05/1993 NINEMILEPT1 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS CONSISTING OF ADMINISTRATIVECHANGES NRC FORM 35 (11-93)
| |
| | |
| I I
| |
| | |
| Pape 6 of NRC FORM 35 U.S. NUCLEAR REGULATORY (11-931 COMMISSION'ECORDS NRCMD 3.53 TRANSFER Continuation DETAILEDINVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 21. CONTAINER NUMBER MXM 23. RECORDS DESCRIPTION 24.DATESPAN (Monrh/Years) 50-220 12/1992 - 12/1992 NINEMILEPTI PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS TO INCREASE SURVEILLANCE TEST INTERVALSANDADD ALLOIVABLEOUT-OFS 50-220 02/1992 - 02/1992 NINEMILEPTI PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS REGARDING COOLANT SYSTEM 50-220 03/1969 - 03/1969 NINEMILEPT1 DRAFT TECHNICALSPECIFICATIONS (FSAR)
| |
| (REVISED) 50-220 12/1965 - 12/1965 NINEMILEPTI SAMPLE TECHNICALSPECIFICATIONS 50-220 06/1997 - 06/1997 NINEMILEPTI REVISION TO EPIPS (SECOND COPY) 50-220 05/1997 - 05/1997 NINEMILEPTI REVISION TO EPIPS (SECOND COPY) 50-220 04/1997 - 04/1997 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY) 50-220 03/1997 - 03/1997 NINEMILEPT1 REVISION TO EPIPS (SECOND COPY)
| |
| NRC FORM 35 (11-93)
| |
| | |
| I Page 7 of NRC FORM 35 U.S. NUCLEAR REGULATORY
| |
| /I 1-931 COMMISSION'ECORDS NRCMD 3.53 TRANSFER Continuation DETAILEDINVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 24. DATE SPAN
| |
| : 21. CONTAINER NUi%1BER W>A~M 23. RECORDS DESCRIPTION (Month/Years) 50-220 10/1996- 10/1996 NINEMILEPf1 REVISION TO EPIPS (SECOND COPY) 50-220 04/1993 - 04/1993 NINEMILEPTI REACTOR CONTAINMENTBUILDINGINTEGRATED LEAKAGERATE TEST FINALREPORT 50-220 06/1992 - 06/1992 NINEMILEPTI SAFETY EVALUATION
| |
| | |
| ==SUMMARY==
| |
| REPORT, 1992 50-220 06/1990 - 06/1990 NINEMILEPT1 SAFETY EVALUATION
| |
| | |
| ==SUMMARY==
| |
| REPORT 1990 SUBMITTED WITH UPDATED FSAR, REVISION 8 AND AMENDMENTS 3 50-220 01/1990 - 01/1990 W
| |
| NINEMILEPT1 APPROVED PROCEDURES POWER ASCENSION PROGRAM 50-220 05/1984 - 05/1984 NINEMILEPT1 CONTAINMENTLEAKTEST 50-220 05/1984 - 05/1984 NINEMILEPI'1 EQUIPMENT QUALIFICATIONPROGRAM NRC FORM 35 (11.93)
| |
| | |
| ~ h Page 8 of NRC FORM 35 U.S. NUCLEAR REGULATORY
| |
| /11-9:41 COMMISSION'ECORDS NRCMD 3.53 TRANSFER Continuation DETAILED INVENTORY OF RECORDS MING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 21. CONTAINER 24. DATE SPAN NUMBER 23. RECORDS DESCRIPTION (Month/Yeats) 50.220 07/1972 - 07/1972 NINEMILEPT1 TECHNICALSUPPLEMENT TO PETITION FOR CONVERSION FROM PROVISIONAL OPERATING LICENSE TO FULL-TERMOPERATING LICE 50-220 OV1971 - OV1971 NINEMILEPTI FIFTH ADDENDUMTO TECHNICALSUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 50-220 12/1970 - 12/1970 NINEMILEPT1 FOURTH ADDENDUMTO TECHNICALSUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 50-220 12/1970 - 12/1970 NINEMILEPT1 THIRD ADDENDUMTO TECHNICALSUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 0-220 10/1970 - 10/1970 NINEMILEPT1 SECOND ADDENDUMTO TECHNICALSUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 50-220 10/1970 - 10/1970 NINEMILEPTI FIRST ADDENDUMTO TECHNICALSUPPLEMENT TO PETITION TO INCREASE POWER LEVEL 50-220 11/1993 - 11/1993 NINEMILEPT1 RESPONSE TO B.L. RIDINGS OF KINSTON, TN'S LETTER OF OCTOBER 13, 1993 50-220 09/1993 - 09/1993 NRC FORM 35 (11-93)
| |
| | |
| Page ilof WRC FORM 35 U.S. NUCLEAR REGULATORY COMMISSION (11-931 NRCMD 3.53 RECORDS TRANSFER Continuation DETAILEDINVENTORY OF RECORDS BEING TRANSFERRED For each individual folder, binder, report, tape, etc., sufficient description to permit retrieval. 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. Triple space when beginning a new container.
| |
| : 21. CONTAINER 24. DATE SPAN
| |
| : 23. RECORDS DESCRIPTION (MonthA'ears)
| |
| NUMBER 50-220 10/1996- 10/1996 NINEMILEPTI SUPERSEDED PAGES PER REVISION TO EPP 50-220 10/1996- 10/1996 NINEMILEPTI SUPERSEDED PAGES PER REVISION TO EPP 50-220 09/1996 - 09/1996 NINEMILEPT1 SUPERSEDED PAGES PER REVISION TO EPP 50-220 11/1995 - 11/1995 NINEMILEFf1 SUPERSEDED PAGES PER REVISION TO EPP 50-220 01/1997 - 01/1991 NINEMILEPTI SUPERSEDED PAGES PER REVISION TO SITE EMERGENCY PLAN 50-220 06/1995 - 06/1995 NINEMILEPI'1 SUPERSEDED PAGES PER REVISION 13 TO UFSAR
| |
| ~,NRC FORM 35 (11.93)
| |
| | |
| l NINE MILE POINT UN3Z 1 FSAR -(UPDATED)
| |
| LIST OF EFFECTIVE PAGES GENERAL TABLE OF CONTENTS T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER 12 xliv 12 li 1
| |
| ill 12 12 Xlv xlvi 12 13
| |
| ~
| |
| iV 12 xlvii 12 xlviii 12 V
| |
| Vi Vii 13 12 12 il l .
| |
| 12 12 V3.3.'1 13 Viiia 13 Viiib 13 3.X 13 X 12 Xl 13 X3.
| |
| Xiii i 12 13 X3.V 13 XV 13 XVl 12 XVl1 12 XVlil 13 XiX XX XXi XX3.1 XXiii XXiV XXV XXVi
| |
| : XXVll, 12 12
| |
| ~
| |
| XXV3.1 XX3.X i
| |
| XXX XXXi 12 XXxii XXX3.3.3.
| |
| XXX3.V
| |
| - c) 13 XXXV XXXVl.
| |
| XXXV3.
| |
| XXXViii i 13 12 XXXiX 12 xl 12 Xli 12 xiii xliii 12 12 UFSAR Revision 13 EP i June 1995&
| |
| | |
| 0 NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES GENERAL TABLE OF CONTENTS T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER I-1 0 I-2 9 I-3 9 I-4 0 I-5 11 I-6 0 I-7 0 I-8 0 I-9 0 I-10 13 I-11 13 I-12 13 I-13 12 I-14 0 I-15 0 I-16 0 I-17 F I-1 3 UFSAR Revision 13 EP 1-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES GENERAL TABLE OF CONTENTS T=TABLE. T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER II-1 13 II-2 F II-1 0 II-3 F II-2 0 II-4 F II-3 6 II-5 13 II-6 T II-1 0 II-7 F II-4 0 II-8 F II-5 0 II-9 T II-2 0 II-10 F II-6 0 II-11 T II-3 13 II-12 T II-4 0
| |
| -II-13 T II-5 13 II-14 6 II-15 6 II-16 6 II-17 T II-6 10 II-18 T II-7 0 II-19 T II-8 0 II-20 T II-8 0 II-21 T II-8 9 II-22 3 II-23 3 II-24 0 II-25 4 UFSAR Revision 13 EP 2-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION III T=TABLE T=TABLE PAGE '=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER III-1 9 III-44 9 III-2 0 III-45 13 III-3 F III-1 13 III-46 F III-18 12 III-4 1 III-47 12 III-5 9 III-47a 3 III-6 0 III-48 12 III-7 F III-2 6 III-48a 10 III-8 F III-3 12 III-48b 10 III-9 F III-4 12 III-49 10 III-10 F III-5 12 III-50 0 III-6 III-51
| |
| ~
| |
| .III-11 F 12 0 III-12 F III-7 6 III-52 0 III-13 F III-8 6 III-53 0 III-14 F III-9 6 III-54 12 III-15 F III-10 2 III-54a 12 III-16 F III-11 2 III-54b 12 III-17 5 III-55 F III-19 0 III-18 F III-12 1 III-56 F III-20 0 III-19 13 III-57 12 III-20 F III-13 1 III-58 F III-21 0 III-21 13 III-59 0 III-22 13 III-60 12 III-23 12 III-61 10 III-23a 12 III-62 F III-22 0 III-23b 12 III-63 13 III-24 , 1 III-64 0 III-25 10 III-65 F III-23 0 III-26 F III-14 11 III-66 0 III-27 12 III-67 0 III-28 10 III-68 0 III-29 0 III-69 12 III-30 0 III-70 12 III-31 0 III-71 '2 12 III-32 0 III-72 III-33 F III-15 2 III-73 12 III-34 0 i III-74 12 III-35 F III-16" 2 III-75 12 III-36 0 III-37 11 III-38 0 III-39 0 III-40 F III-17 0 III-41 0 III-42 0 III-43 9 UFSAR Revision 13 EP 3-1 June 1995
| |
| | |
| 0 h
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION IV T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER IV-1 12 IV-41 0 IV-2 13 IV-42 1 IV-3 6 IV-43 0 IV-4 6 IV-44 0 IV-5 13 IV-45 0 IV-6 12 IV-46 10 IV-7 0 IV-46a 10 IV-8 13 IV-46b 10 IV-9 0 IV-47 0 IV-10 F IV-1 13 IV-48 F IV-9 0 IV-11 13 IV-49 0 IV-12 13 IV-50 0 IV-13 13 IV-51 0 IV-14 12 IV-52 0 IV-15 12 IV-16 0 IV-17 0 IV-18 0 IV-19 i13 IV-19a 8 IV-20 0 IV-21 8 IV-22 13 IV-23 13 IV-24 12 IV-25 13 IV-26 F IV-2 6 IV-27 13 IV-28 F IV-3 13 IV-29 13 IV-30 13 IV-31 13 IV-32 13 IV-33 F IV-4 10 IV-34 F IV-5 13 IV-35 11 IV-35a '11 IV-35b 10 IV-36 0 IV-37 F IV-6 0 IV-38 F IV-7 0 IV-39 F IV-8 10 IV-40 13 IV-40a 12 IV-40b ,12 UFSAR Revision 13 EP 4-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION V T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER V-1 0 V-2 0 V-3 T V-1 11 V-4 T V-1 8 V=5 T V-1 10 V-6 0 V-7 T V-2 0 V-8 0 V-9 T V-3, 0 T V-4 V-10 0 V-11 F V-1 11 V-12 1 V-12a T V-5 1 V-13. F V-2 8 V-14 9 V-15 F V-3 0 V-16 0 V-17 13 V-18 F V-4 13 V-19 10 V-20 0 V-21 11 V-22 11 V-23 F V-5 11 V-23a 11 V-23b 11 V-24 F V-6 11 V-25 'F V-7 11 V-26 0 V-27 0 V-28 12 V-29 9 V-30 4 V-30a 4 V-31 12 V-31a 12 V-31b 12 V-32 F V-8 0 V-33 3 UFSAR Revision 13 EP 5-1 June 1995
| |
| | |
| /
| |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION VI T=TABLE T=TABLE PAGE F=FIGURE REVISION . PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER VI-1 VI-43 11 VI-2 VI-43a 11 VI-3 VI-43b 11 VI-4 VI-44 T VI-1 11 VI-5 VI-45 T VI-1 11 VI-6 VI-46 T VI-2 11 VI-7 VI-47 T VI-3a 13 VI-8 VI-48 T VI-3a 13 VI-9 VI-48a T VI-3a 13 V.I-10 VI-49 T VI-3b 11 VI-11 VI-50 T VI-3b 11 VI-12 VI-50a T VI-3b 13 VI-13 VI-50b T VI-3b 11 VI-14 F VI-1 VI-51 T VI-3c 11 VI VI-52 0 VZ-16 VI-53 F VI-20 0 VI-17 F VI-2 VI-54 F VI-21 0 VI-18 F VI-3 VI-55 12 VI-19 F VI-4 VI-56 T VI-4 11 VI-19a F VI-4a VI-56a 13 VI-20 F VI-5 0 VI-56b 13 VI-21 4 VI-57 3 VI-22 11 VI-58 F VI-22 11 VI-22a 11 VI-59 12 VI-22b 11 VI-60 0 VI-23 0 VI-61 F VI-23 0 VI-24 0 VI-62 0 VI-25 9 VI-63 F VI-24 11 VI-26 0 VI-64 12 VI-27 10 VI-65 13 VI-28 F VI-6 0 VI-66 13'3 VI-29 F VI-7 0. VI-67 VI-30 F VI-8 0 VI-68 13 VI-31 F VI-9 VI-69 T VI-5 0 VI-32 F VI-10 VI-70 T VI-5 0 VI-33 F VI-11 0 VI-34 F VI-12 0 VI-35 F VI-13 0 VI-36 F VI-14 VI-37 F VI-15 VI-38 F VI-16 VI-39 F VI-17 VI-40 VI-41 F VI-18 VI-42 F VI-19 UFSAR Revision 13 EP 6-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION VII T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER VII-1 9 VZI-33a 10 VII-2 10 VIZ-33b 10 VII-3 VII-1 10 VII-34 0 VII-4 8 VII-35 T VII-1 0 VII-5 F VZI-2 9 VII-36 0 VZI-,6 8 VZI-37 0 VII-6a 10 VZI-38 0 VII-7 6 VII-39 F VII-10 0 VII-8 8 VII-40 9 VII-9 8 VII-41 0 VII-10 0 VII-42 3 VII-11 13 VII-43 0 VII-11a 13 VII-44 0 VII-11b 10 VII-45 F VIZ-11 0 VII-12 F VII-3 11 VII-46 12 VII-13 F VII-4 9 VZI-46a 12 VIZ-13a F VII-4a 10 VII-46b 12 VII-14 F VII-5 9 VII-47 F VII-12 12 VII-15 9 VII-48 0 VII-16 10 VII-49 3 VII-17 9 VII-50 F VII-13 9 VII-18 12 VII-51 F VII-14 0 VII-18a 10 VII-52 F VII-15 0 VII-18b 10 VII-53 0 VII-19 13 VII-54 F VII-16 0 VII-20 13 VIZ-55 12 VII-21 11 VII-55a 12 VII-21a 11 VZI-55b 12 VII-21b 11 VZI-56 0 VII-22 F VII-6 10 VII-57 0 VII-23 7 VZZ-58 0 VII-24 F VIZ-7 0 VZI-59 0 VII-25 F VII-8 7 VII-60 0 VII-26 11 VII-61 12 VII-26a 10 VII-61a 9 VII-26b 10 VII-62 8 VII-27 0 VII-62a 8 VII-28 11 VII-63 F VII-17 9 VII-29 0 VII-30 10 VII-30a 10 VII-30b 10 VII-31 10 VII-32 . F VII-9 10 VII-33' 10 UFSAR Revision 13 EP 7-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION VIII T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER VIII-1 12 VIII-45 F VIII-16 0 VIIX-2 F VIII-1 12 VIII-46 F VIII-17 0 VIII-3 5 VIII-47 F VIII-18 0 VIII-4 9 VIII-48 0 VIII-5 5 VIII-49 F VIII-19 0 VIII-6 7 VIII-50 11 VIII-7 0. VIII-51 0 VIIZ-8 F VIII-2 13 VIII-52 F VIII-20 0 VIII-9 13 VIII-53 F VIII-21 0 VIII-10 0 VIII-54 0 VIII-11 11 VIII-55 0 VIII-12 7 VIII-56 13 VIII-13 13 VIII-57 13 VIII-14 7 VIII-58 13 VIII-15 4 'III-59 13 VIIZ-16 F VIXI-3 5 VIIX-60 13 VIII-17 0 VIII-61 13 VIII-18 0 VIII-62 13 VIII-19 F VIII-4 0 VIII-62a 13 VIII;20 0 VIII-62b 13 VIII-21 4 VIII-63 F VIII-22 0 VIII-22 0 VIII-64 0 VIII-23 0 VIII-65 F VIII-23 10 VIII-24 F VIXI-5 0 VIII-66 F VIII-24 10 VIII-25 0 VIII-67 0 VIII-26 F VIII-6 8 VIII-68 ~
| |
| 4 VIII-27 F VIIZ-7 0 VIII-69, F VIII-25 3 VIII-28 '8 VIII-70 F VIII-26 10 VIII-29 F VIII-8 10 VIII-71 12 VIII-30 F VIIZ-9 0 VIII-71a 12 VIII-31 0 VIII-71b 12 VIII-32 0 VIII-72 10 VIII-33 F VIII-10 0 VIII-72a 10 VIII-34 F VIII-11 0 VIII-72b 10 VIII-35 F VIII-12 10 VIII-73 F VIII-27 10 VIII-36 0 VIII-74 0 VIII-37 F VIXI-13 0 VIII-75 F VIII-28 3 VIII-38 F VIII-14 12'1 VIII-76 0 VIII-39 VIII-77 0 VIII-40 8 VIIZ-78 F VIII-29 0 VIII-40a 8 VIIZ-79 0 VIII-41 F VIII-15 0 VIIX-80'III-81 9
| |
| VIII-42 0 12 VIII-43 0 VIII-81a 12 VIII-44 0 VIII-81b 12 UFSAR Revision 13 EP 8-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION VIII T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER i NUMBER NUMBER NUMBER NUMBER VIII-82 9 VIII-82a 9 VIII-82b 12 VIII-82c 12 VIII-826 12 VIII-83 T VIII-1 12 VIII-84 12 VIII-85 T VIII-2 12 VIII-86 12 VIII-87 T VIII-3 13 VIII-88 T VIII-3 12 VIII-89 T VIII-3 12 VIII-90 T VIII-3 12 VIII-91 T VIII-3 12 VIII-92 12 VIII-93 12 VIII-94 T VIII-3 9 VIII-95 T VIII-3 12 VIII-96 T VIII-3 12 VIII-97 T VIII-3 12 VIII-98 T VIII-3 13 VIII-98a T VIII-3 13 VIII-98b 13 VIII-99 '12 VIII-100 12 VIII-101 12 VIII-102 12 VIII-103 12 VIII-104 12 VIII-105 12 VIII-106 12 VIII-107 12 VIII-108 12 VIII-109 12 VIII-110 12 VIII-111 12 VIII-112 13 UFSAR Revision 13 EP 8-2 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES, SECTION IX T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER IX-1 0 IX-32 T IX-1 12 IX-2 11 IX-33 T IX-1 12 IX-2a 11 IX-33a T IX-1 12 IX-2b 11 IX-34 0 IX-3 6 IX-35 12 IX-4 12 IX-35a 12 IX-4a 11 IX-35b 12 IX-4b 11 IX-36 12 IX-5 0 IX-37 12 IX-6 6 IX-38 11 IX-7 0 IX-39 11 IX-8 13 IX-40 11 IX-8a 13 IX-8b 13 IX-9 F IX-1 12 IX-10 12 IX-11 6 IX-12 12 IX-13 0 IX-14 F IX-2 12 IX-15 12 IX-16 12 IX-16a 12 ZX-16b 12 IX-17 10 IX-18 10 IX-19 12 IX-20 12 IX-20a 10 IX-20b 10 IX-21 F IX-3 6 IX-22 F IX-4 6 IX-23 F IX-5 6 IX-24 13 IX-24a 13 IX-24b 11 IX-25 13 IX-26 8 IX-27 F IX-6 13 IX-28 F IX-7 12 IX-29 12 IX-30 12 IX-31 12 IX-31a 12 IX-31b 12 UFSAR Revision 13 EP 9-1 June 1995
| |
| | |
| 1 NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION X T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER X-1 X-40 F X-7 4 X-2 F X-1 9 X-41 F X-8 11 X-3 8 X-42 10 X-4 0 X-43 0 X-5 F X-2 12 X-44 0 X-6 0 X-45 11 X-7 0 X-46 F X-9 13 X-8 13 X-47 11 X-9 F X-3 13 X-48 13 X-10 0 X-49 13 X-11 0 X-50 13 X-12 0 X-51 F X-10 7 X-13 0 X-52 13 X-14 7 X-52a 13 X-15 9 X-52b 13 X-16 0 X-53 7 X-'17 0 X-54 13 X-18 0 X-55 13 X-19 0 X-55a 13 X-20 0 X-55b 10 X-21 13 X-56 F X-11 0 X-21a 11 X-57 11 X-21b 11 X-58 11 X-22 10 X-59 10 X-23 F X-4 10 X-60 10 X-24 10 X-61 10 X-25 10 X-62 11, X-25a 10 X-63 10 X-25b 10 X-64 10 X-26 9 X-65 10 X-27 0 X-66 10 X-28 F X-5 1 X-67 10 X-29 11 X-68 10 X-30 0 X-69 10 X-3 1 11 X-70 10 X-32 F X-6 12 X-71 10 X-.3 3 11 X-72 10 X-34 9 X-73 10 X-35 12 X-74 10 X-36 12 X-75 10 X-37 12 X-76 1 X-38 10 X-77 3 X-39 13 X-78 5 X-39a 13 X-79 5 X-39b 12
| |
| -UFSAR Revision 13 EP 10-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION X T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER 10A 10 10A-34 T 2 ' ' '-2 10 10A-i 10 10A-35 T 2.5.1 '-3 10 10A-ii 13 10A-36 T 2.5 ' '-4 12 10A-iii 10 10A-37 T 2 ' ' '-5 10 10A-iv 10 10A-38 T 2 ' ' '-6 10 10A-v 13 10A-39 T 2 ' ' '-7 10 10A-vi 10 10A-40 10 10A-vii 12 10A-41 13 10A-viii . 12 10A-41a 13
| |
| '10A-ix 10 10A-41b 13 10A-1 12 10A-42 10 10A-2 10 10A-43 10 10A-3 10 10A-44 10 10A-4 12 10A-45 10 10A-5 13 10A-46 13 10A-6 13 10A-46a 13 10A-7 12 10A-46b 13 10A-8 12 10A-47 10 10A-9 12 10A-48 10 10A-10 11 10A-49 T 2. 5 3 4-1 10 10A-11 12 10A-50 T 2
| |
| ~
| |
| 5.3 '-1
| |
| ~
| |
| 10 10A-12 11 10A-51 10 10A-13 12 10A-52 13 10A-14 10 10A-52a 13 10A-15 10 10A-52b 13 10A-16 10 10A-53 10 10A-17 10 10A-54. 10 10A-18 10 10A-55 13 10A-19 11 10A-55a 13 10A-20 T 2' 11 10A-55b 13 10A-21 10 10A-56 12 10A-22 10 10A-57 10 10A-23 10 10A-58 10 10A-24 13. 10A-59 10 10A-24a 13 10A-60 10 10A-24b 13 10A-61 12 10A-25 10 10A-61a 12 10A-26 12 10A-61b 12 10A-27 10 10A-62 10 10A-28 10 10A-63 10 10A-29 10 10A-64 10 10A-30. 10 10A-65 10 10A-31 10 10A-66 10 10A-32 12 10A-67 10 10A-33 T 2. 5 '.1-1 10 10A-68 10 UFSAR Revision 13 EP 10-2 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION X T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER
| |
| '2 10A-69 10 10A-108 T 3.1.1-1 10A-70 12 10A-109 T 3 '.1-1 12 10A-71 10 10A-110 T 3.1.1-1 12 10A-72 10 10A-111 T 3 '.1-1 12 10A-73 10 10A-112 T 3.1 '-2
| |
| '-2 12 10A-74 10 10A-113 T 3.1 12 10A-75 10 10A-114 T 3.1.1-2 12 10A-76 10 10A-115 T 3 '.1-2 12 10A-77 10 10A-116 T 3 '.1-2 12 10A-78 13 10A-117 T 3.1 '-2 3.1.1-3 12 10A-78a 13 10A-118 T 12 10A-78b 13 10A-119 T 3.1.1-4 12 10A-79 10 10A-120 T 3.1.1-4 12 10A-80. 10 10A-121 T 3 '.1-5 12 10A-81 10 10A-122 T 3 '.1-6 12 10A-82 12 10A-123 T 3 '.1-6 12 10A-83 12 10A-124 T 3 '.1-7 12 10A-83a 12 10A-125 T 3 '.1-8 12 10A-83b 12 10A-126 T 3.1.1-8 12 10A-84 10 10A-127 T 3 '.1-9 12 10A-85 10 10A-128 T 3 '.1-9 12 10A-86 10 10A-129 T 3.1.1-9 12 10A-87 12 10A-130 T 3 '.1-9 12 10A-88 10A-89 T
| |
| T 1
| |
| 1
| |
| '.2
| |
| '.2 10 10 Legend B-40141-C 10 11 10A-90 T 1.2.2 13 Overlay 1-1 10A-91 T 1.2.2 13 Overlay 2-1 10A-92 T 1.2.2 13 B-40142-C 12 10A-93 T 1.2.2 13 Overlay 1-2 10A-94 10A-95 T
| |
| T 1 '.2 3.2-1 13 10 Overlay Overlay 2-2 3-2 10A-96 T 3.3-1 12 Overlay 4-2 10A-97 T 3.3-1 10 B-40143-C 12 10A-98 T 3.4-1 10 Overlay 1-3 10A-99 T 3.5-1 10 Overlay 2-3 10A-100 T 3.6-1 10 Overlay 3-3 10A-101 T 3.7-1 10 Overlay 4-3 10A-102 T 3.8-1 10 B-40144-C 12 10A-103 T 3.9-1 10 Overlay 1-4 10A-104 T 3.10-1 10 Overlay 2-4 10A-105 T 3.10-1 10 Overlay 3-4 10A-106 T 3 a 1 1 10 Overlay 4-4 10A-107 T 3 1 1 12 10A-107a 10A-107b T 3 '-1
| |
| ~
| |
| 3.1-1 12 T 12 UFSAR Revision 13 EP 10-3 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION X T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER , NUMBER NUMBER NUMBER NUMBER B-40145-C 12 10B-25 T 1 12 Overlay 1-5 10B-26 T 1 12 Overlay 2-5 10B-27 T 1 12 Overlay 3-5 10B-28 T 1 12 B-40146-C 12 10B-29 T 1 12 Overlay 1-6 10B-30 T 1 12 Overlay 2-6 10B-31 T 1 12 Overlay 3-6 10B-32 T 1 12 B-40147-C 10B-33 T 1 12 Overlay 1-7 10B-34 T 1 12 Overlay 2-7 10B-35 12 Overlay 3-7 10B-36 12 B-40148-C 10B-37 12 Overlay 1-8 10B-38 12 Overlay 2-8 10B-39 12 Overlay 3-8 10B-40 12 10B-41 T 2A 12 10B 10 10B-42 T 2B 12 10B-i 12 10B-43 T 3 12 10B-ii 12 10B-44 T 3 12 10B-iii '12 10B-45 T 3 12
| |
| '10B-1 12 10B-46 T 3 12 10B-2 12 10B-47 T 3 12 10B-3 12 10B-48 T 3 12 10B-4 ~
| |
| 12 10B-49 12 10B-5 12 10B-50 12 10B-6 12 10B-51 12 10B-7 12 10B-52 12 10B-8 12 10B-53 12 10B-9 12 10B-54 12 10B-10 12 10B-55 12 10B-11 12 10B-56 12 10B-12 12 10B-57 12 10B-13 12 10B-58 12 10B-14 12 10B-59 12 10B-15 12 10B-60 12 10B-16 12 10B-61 12 10B-17 12 10B-62 12 10B-18 12 10B-63 12 10B-19 12 10B-64 12 10B-20 12 10B-65 12 10B-21 12 10B-66 12 10B-22 12 10B-67 12 10B-23 12 10B-68 12 10B-24 12 10B-69 12 UFSAR Revision 13 EP 10-4 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION X T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER 10B-70 12 10B-115 12 10B-71 12 10B-116 12 10B-72 12 10B-117 12 10B-73 12 10B-118 12 10B-74 12 10B-119 12 10B-75 12 10B-120 12 10B-76 12 10B-121 12 10B-77 12 10B-122 12 10B-78 12 10B-123 12 10B-79 12 10B-124 12 10B-80 12 10B-125 12 10B-81 12 10B-126 12 10B-82 12 10B-127 12 10B-83 1,2 10B-128 12 10B-84 1'2 10B-129 12 10B-85 12 10B-130 12 10B-86 12 10B-131 12 10B-87 12 10B-132 12 10B-88 12 10B-133 12 10B-89 12 10B-134 12 10B-90 12 10B-135 12 10B-91 12 10B-136 12 10B-92 12 10B-137 12 10B-93 12 10B-138 12 10B-94 12 10B-139 12 10B-95 12 10B-140 12 10B-96 12 10B-141 12 10B-97 12 10B-142 12 10B-98 12 10B-143 12 10B-99 12 10B-144 12 10B-100 12 10B-145 12 10B-101 12 10B-146 12 10B-102 12 10B-147 12 10B-103 12 10B-148 12 10B-104 12 10B-149 12 10B-105 12 10B-150 12 10B-106 12 10B-151 '12 10B-107 12 10B-152 12 10B-108 12 10B-153 12 10B-109 12 10B-154 12'2 10B-110 12 10B-155 10B-111 12 10B-156 12 10B-112 12 10B-157- 12 10B-113 12 10B-158 12 10B-114 12 10B-159 12 UFSAR Revision 13 EP 10-5 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION X T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER 10B-160 12 10B-205 12 10B-161 12 10B-206 12 10B-162 12 10B-207 12 10B-163 12 10B-208 12 10B-164 12 10B-209 12 10B-165 12 10B-210 12 10B-166 12 10B-211 12 10B-167 12 10B-212 12 10B-168 12 10B-213 12 10B-169 12 10B-214 12 10B-170 12 10B-215 12 10B-171 12 10B-216 12 10B-172 12 10B-217 12 10B-173 12 10B-218 12 10B-174 12 10B-219 12 10B-175 12 10B-220 12 10B-176 12 10B-221 12 10B-177 12 10B-222 12 10B-178 12 10B-223 12 10B-179 12 10B-180 12 10B-181 12 10B-182 12 10B-183 12 10B-184 12 10B-185 12 10B-186 12 10B-187 12 10B-188 12 10B-189 12 10B-190 12 10B-191 12 10B-192 12 10B-193 12 10B-194 12 10B-195 12 10B-196 12 10B-197 12 10B-198 12 10B-199 12 10B-200 12 10B-201 12 10B-202 12 10B-203 12 10B-204 12 UFSAR Revision 13 EP 10-6 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XI T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XI-1 0 XI-2 F XI-1 0 XI-3 F XI-2 1 XI-4 F XI 13 XI-5 F XI-4 2 XI-6 F XI-5 2 XI-7 F XI-6 0 XI-8 F XI-7 12 XI-9 10 XI-10 8 XI-11 8 XI-12 9 XI-13 0 XI-14 0 XI-15 0 XI-16 0 XI-17 13 XI-18 0 XI-19 0 XI-20 0 XI-21 0 XI-22 0 XI-23 13 XI-24 0 XI-25 0 UFSAR Revision 13 EP 11-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XII T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER 'UMBER XIZ-1 0 XII-42 13 XII-2 T XII-1 0 XII-43 13 XII-3 0 XII-44 13 XII-4 T XII-2 0 XII-45 13 XII-5 T XII-2 0 XII-46 13 XII-6 T XII-3 0 XII-7 0 XII-8 6 XII-9 F XII-1 13 XII-10 6 XII-11 6 XIZ-12 6 XII-13 13 XII-14 T XII-4 6 XII-15 13 XII-16. 13 XII-17 13 XII-18 T XII-5 9 XII-19 6 XII-20 0 XII-21 T XII-6 0 XII-22 T XII-7 0 XII-23 0 XII-24 0 XII-25 0 XII-26 T XII-8 11 XII-27 T XII-8 11 XII-28 11 XII-29 0 XII-30 11 XII-31 0 XII-32 12 XII-32a 12 XII-32b 12 XII-33 12 XII-34'II-35 12
| |
| '12 XII-36 12 XII-37 13 XII-38 12 XII-39 0 XII-40 12 XII-41 13 XII-41a 13 XII-41b 11 UFSAR Revision 13 EP 12-1 June 1995
| |
| | |
| )
| |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XIII T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XIII-1 12 XIII-2 13 XIII-3 13 XIII-4 12 XIII-5 12 XIII-6 13 XIII-6a 13 XIII-6b 13 XIII-7 12 XIII-8 12 XIII-9 12 XIII-10 12 XIII-11 12 XIII-12 12 XIII-13 12 XIII-14 12 XIII-15 12 XIII-16 13 XIII-17 13 XIII-18 13 XIII-19 13 XIII-20 12 XIII-21 12 XIII-22 12 XIII-23 12 XIII-24 12 T XIII-1 13 F XIII-1 13 F XIII-2 13 V XIII-3 13 F XIII-4 13 F XIII-5 13 UFSAR Revision 13 EP 13-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XIV T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER 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 UFSAR Revision 13 EP 14-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XV T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER 13 XV-39 F XV-15 13 XV-2 0 XV-40 F XV-16 3 XV-3 F XV-1 12 XV-41 3 XV-4 T XV-1 0 XV-42 F XV-17 0 XV-5 T XV-1 12 XV-43 0 XV-5a 11 XV-44 0 XV-5b 11 XV-45 F XV-18 0 XV-6 T XV-2 13 XV-46 9 13 XV-47 F XV-19 0 XV-7a 13 XV-48 9 XV-7b 12 XV-49 F XV-20 0 XV-8 F XV-2 13 XV-50 12 XV-8a F XV-2a 6 XV-50a 12 XV-8b F XV-2b 6 XV-50b 12 XV-9 13 XV-51 F XV-21 0 XV-10 F XV-3 6 XV-52 9 XV-11 13 XV-52a 8 XV-11a 13 XV-53 F XV-22 0 XV-12 F XV-4 13 XV-54 9 XV-13 13 XV-55 F XV-23 0 XV-14 T XV-3 13 XV-56 0 XV-15 6 XV-57 0 XV-16 F XV-5 13 XV-58 F XV-24 0 XV-17 F XV-6 6 XV-59 1 XV-18 13 XV-59a 1 XV-19 F XV-7 13 XV-60 T XV-4 8 XV-20 0 XV-61 T XV-4 9 XV-21 13 XV-62 1 XV-22 13 XV-63 0 XV-23 F XV-8 XV-64 9 XV-24 XV-65 13 XV-25 F XV-9 XV-66 7 XV-26 F XV-10 XV-67 0 XV-27 0 XV-68 0 XV-28 0 XV-69 0 XV-29 F XV-11 0 XV-70 8 XV-30 0 XV-71 8 XV-31 F XV-12 0 XV-72 0 XV-32 0 XV-73 T XV-5 0 XV-33 F XV-13 0 XV-74 F XV-25 0 XV-34 0 XV-75 0 XV-35 F XV-14 0 XV-76 T XV-6 0 XV-36 13 XV-77 T XV-7 0 XV-37 12 XV-78 T XV-8 0 XV-38 13 XV-79 8 UFSAR Revision 13 EP 15-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XV T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XV-80 8 XV-120 F XV-51 8 XV-81 12 XV-12 1 F XV-52 8 XV-81a 12 XV-122 F XV-53 8 XV-81b 12 XV-123 F XV-54 8 XV-82 T XV-9 9 XV-124 T XV-14 12 XV-82 a T XV-9 12 XV-125 T XV-15 9 XV-82b T XV-9A 12 XV-12 6 F XV-55 9 XV-83 8 XV-127 F XV-56 12 XV-84 8 XV-128 8 XV-85 F XV-26 8 XV-129 8 XV-86 F XV-27 8 XV-13 0 8 XV-87 8 XV-13 1 T XV-16 8 XV-88 8 XV-132 8 XV-89 F XV-28 8 XV-133 8 XV-90 F XV-29 8 XV-134 T XV-17, 12 XV-91 F XV-30 8 T XV-18 XV-92 F XV-3 1 8 XV-135 8 XV-93 F XV-32 8 XV-13 6 T XV-19, 12 XV-94 F, XV-33 8 T XV-20, XV-94a XV-33a 6. T XV-21 XV-95 8 XV-137 XV-96 12 XV-137a XV-97 XV-10 XV-137b XV-98 XV-11 XV-137c 12 XV-99 XV-137d 12 XV-100 XV-137d1 12 XV-101 T XV-12 XV-137d2 12 XV-102 T XV-13 XV-137e T XV-21A 12 XV-103 F XV-34 XV-137 f T XV-21B XV-21C 12 XV-104 F XV-35 XV-137g T 12 XV-105 F XV-3 6 XV-137h T XV-21D, 12 XV-106 F XV-37 T XV-21E XV-107 F XV-38 8 XV-137i F XV-56a 8 XV-108 F XV-39 '8 XV-137) F XV-56b 12 XV-109 F XV-40 8 XV-137k F XV-56c 12 XV-110 F XV-4 1 XV-138 8 XV-111 F XV-42 XV-139 12 XV-112 F XV-43 XV-139a 12 XV-113 F XV-44 XV-139b 12 XV-114 F XV-45 XV-140 0 XV-115 F XV-46 XV-141 0 XV-116 F XV-47 XV-142 12 XV-117 F XV.-48 XV-143 0 XV-118 F XV-49 XV-144 T XV-22 0 XV-119 F XV-50 XV-145 T XV-23 0 UFSAR Revision 13 EP 15-2 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION,XV T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XV-146 T XV-24 XV-164 XV-147 T XV-25 XV-164a 11 XV-148 T XV-26, XV-164b 10 T XV-27 XV-165 F XV-59 0 XV-149 12 "
| |
| XV-166 F XV-60 10 XV-150 8 XV-167 T XV-30 11 XV-151 8 XV-168 T XV-31, XV-152 8 T XV-32 XV-153 0 XV-169 11 XV-154 0 XV-169a 10 XV-155 0 XV-169b 11 XV-156 0 XV-169c 11 XV-157 0 XV-169d 11 XV-158 T XV-28 0 XV-169e 10 XV-159 T XV-29 11 XV-169f 11 XV-159a 11 XV-169g 11 XV-159b 11 XV-169h 11 XV-159c 11 XV-169i T XV-32a 10 XV-159d 11 XV-169> 10 XV-159e F XV-56D 11 XV-169k 13 XV-159f T XV-29A 11 XV-169L F XV-60a 10 XV-159g F XV-56E 11 XV-169m F XV-60b 10 XV-159h F XV-56F 11 XV-170 8 XV-159i 11 XV-171 13 XV-159j 11 XV-172 T XV-33 0 XV-159k 11 XV-173 T XV-34 0 XV-159L F XV-56G 11 XV-174 0 XV-159m 11 XV-175 0 XV-159n F XV-56H 11 XV-176 T XV-35 0 XV-159o 11 XV-177 0 XV-159p 11 XV-178 F XV-61 0 11 XV-179 0 XV-159'V-159r 11 XV-180 0 XV-159s T XV-29B 11 XV-181 0 XV-159t T XV-29B 11 XV-182 0 XV-159u T XV-29C 11 XV-183 0 XV-159v 11 XV-184 F XV-62 0 XV-159w 11 XV-185 F XV-63 10 XV-159x T XV-29D 11 XV-186 F XV-64 0 XV-160 11 XV-187 0 XV-160a 10 XV-188 F XV-65 0 XV-160b 10 XV-189 F XV-66 0 XV-161 F XV-57 11 XV-190 F XV-67 0 XV-162 0 XV-191 F XV-68 0 XV-163 F XV-58 0 XV-192 F XV-69 0 UFSAR Revision 13 EP 15-3 June 1995
| |
| | |
| 0 NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XV T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XV-193 F XV-70 0 XV-194 F XV-71 0 XV-195 T XV-36 0 XV-196 F XV-72 10 XV-197 0 XV-198 0 XV-199 0 XV-200 0 UFSAR Revision 13 EP 15-4 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XVI T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XVI-1 11 XVI-46 0 XVI-2 0 XVI-46a 13 XVI-3 0 XVI-46b 13 XVI-4 T XVI-1 0 XVI-47 0 XVI-5 T XVI-1 0 XVI-48 0 XVI-6 T XVI-2 0 XVI-49 0 XVI-7 T XVI-2 9 XVI-50 0 XVI-8 T XVI-2 0 XVI-51 11 XVI-9 0 XVI-52 .0 XVI-10 9 XVI-53 0 XVI-11 T XVI-3 0 XVI-54 F XVI-13 0 XVI-12 F XVI-1 3 XVI-55 T XVI-10 0 XVI-13 F XVI-2 3 XVI-56 T XVI-11 0 XVI-14 0 XVI-57 0 XVI-15 F XVI-3 3 XVI-58 10 XVI-16 F XVI-4 3 XVZ-59 T XVI-12 0 XVI-17 F XVI-5 3 XVI-60 F XVI-14 10 XVI-18 F XVI-6 3 XVI-61 0 XVI-19 F XVI-7 3 XVI-62 F XVI-15 0 XVI-20 F XVI-8 3 XVI-63 0 XVI-21 F XVI-9 0 XVI-64 F XVI-16 0 XVI-22 F XVI-10 0 XVI-65 T XVI-13 9 XVI-23 9 XVI-66 F XVI-17 0 XVI-24 0 XVI-67 0 XVI-25 0) XVI-68 0 XVI-26 0 XVI-69 0 XVI-27 0 XVI-70 F XVI-18 0 XVI-28 F XVI-11 0 XVI-71 F XVI-19 0 XVI-29 0 XVI-72 0
| |
| .XVI-30 0 XVI-73 F XVI-20 0' XVI-31 T XVI-4 0 XVI-74 XVI-32 F XVI-12 0 XVI-75 F XVI-21 0 XVI-33 T XVI-5 0 XVI-76 0 XVI-34 ~
| |
| T XVI-6 0 XVI-77 0 XVI-35 T XVI-7 0 XVI-78 0 XVI-36 0 XVI-79 0 XVI-37 T XVI-8 '0 XVI-80 F XVI-22 0 XVI-38 0 XVI-81 F XVI-23 0 XVI-39 0 XVI-82 F XVI-24 0 XVI-40 T XVI-9 0 XVI-83 0 XVI-41 0 XVI-84 0 XVI-42 0 XVI-85 0 XVI-43 0 XVI-86 T XVI-14 0 XVI-44 0 XVI-87 0 13'P XVI-45 0 XVI-88 0
| |
| ~
| |
| UFSAR Revision 16-1 June 1995
| |
| | |
| ~ .
| |
| I NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XVI T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XVI-89 XVI-130 T XVI-23 9 XVI-90 XVI-131 T XVI-24 0 XVI-91 T XVI-15 XVI-132 T XVI-25 0 XVI-92 XVI-133 T XVI-26 9 XVI-93 XVI-134 F XVI-34 0 XVI-94 XVI-135 F XVI-35 0 XVI-95 XVI-136 F XVI-36 0 XVI-96 0 XVI-137 F XVI-37 0 XVI-97 0 XVI-138 F XVI-38 0 XVI-98 0 XVI-139 F XVI-39 0 XVI-99 0 XVI-140 F XVI-40 0 XVI-100 F XVI-25 0 XVI-141 F XVI-41 0 XVI-101 T XVI-16 0 XVI-142 9 XVI-102 0 XVI-143 0 XVI-103 F XVI-26 0 XVI-144 F XVI-42 0 XVI-104 10 XVI-145 0 XVI-104a 10 XVI-146 0 XVI-104b 10 XVI-147 F XVI-43 ,0 XVI-105 0 XVI-148 F XVI-44 0 XVI-106 0 XVI-149 F XVI-45 0 XVI-107 0 XVI-150 0 XVI-108 0 XVI-151 T XVI-27 0 XVI-109 F XVI-27 0 XVI-152 T XVI-27 0 XVI-110 T XVI-17 0 XVI-153 0 XVI-111 T XVI-18 0 XVI-154 0 XVI-112 0 XVI-155 12 XVI-113 0 XVI-155a T XVI-28 12 XVI-114 10 XVI-155b 12 XVI-114a 10 XVI-156 0 XVI-114b 10 XVI-157 0 XVI-115 F XVI-28 0 XVI-158 0 XVI-116 0 XVI-159 0, XVI-117 F XVI-29 0 XVI-160 0 XVI-118 F XVI-30 0 XVI-161 0 XVI-119 F XVI-31 0 XVI-162 0 XVI-120 F XVI-32 0 XVI-163 0 XVI-121 F XVI-33 0 XVI-164 12 XVI-122 T XVI-19 12 XVI-165 0 XVI-123 0 XVI-166 T XVI-29 0 XVI-124 .0 XVI-167 T XVI-30 0 XVI-125 'a 0 XVI-168 0 XVI-126 XVI-169 F XVI-46 9 XVI-127 T XVI-20 XVI-170 F XVI-47 9 XVI-128 T XVI-21 XVI-171 F XVI-48 0 XVI-129 T XVI-22 XVI-172 F XVI-49 0 UFSAR Revision 13 EP 16-2 June 1995
| |
| | |
| 0 NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF 'EFFECTIVE PAGES SECTION XVI T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XVI-173 F XVI-50 XVI-218 EXHIBIT 10 0 XVI-174 F XVI-51 XVI-219 EXHIBIT 10 0 XVI-175 F XVI-52 XVI-220 EXHIBIT 11 0 XVI-176 F XVI-53 XVI-221 EXHIBIT 11 0 XVI-177 F XVI-54 XVI-222 EXHIBIT 12 0 XVI-178 F XVI-55 XVI-223 EXHIBIT 12 0 XVI-179 F XVI-56 XVI-224 EXHIBIT 12 0 XVI-180 F XVI-57 XVI-225 EXH1BIT 13 0 XVI-181 F XVI-58 XVI-226 EXHIBIT 13 0 XVI-182 F 'XVI-59 XVI-227 0 XVI-183 F XVI-60 XVI-228 EXHIBIT 14 0 XVI-184 F XVI-61 XVI-229 EXHIBIT 14 0 XVI-185 T XVI-31 XVI-230 EXHIBIT 14 0 XVI-186 T XVI-31 XVI-231 EXHIBIT 14 0 XVI-187 XVI-232 EXHIBIT 14 0 XVI-188 XVI-233 EXHIBIT 14 0 XVI-189 XVI-234 EXHIBIT 14 0 XVI-190 EXHIBIT 1 XVI-235 EXHIBIT 14 0 XVI-191 EXHIBIT 1 XVI-236 EXHIBIT 14 0 XVI-192 EXHIBIT 1 XVZ-237 EXHIBIT 14 0 XVI-193 EXHIBIT 1 XVI-238 9 XVI-194 EXHIBIT 2 XVI-195 EXHIBIT 2 XVI-196 EXHIBIT 2 XVI-197 EXHIBIT 2 XVI-198 EXHIBIT 2 XVI-199 EXHIBIT 2 XVI-200 EXHIBIT 2 XVI-201 EXHIBIT 2 XVI-202 EXHIBIT 2 XVI-203 EXHIBIT 2 XVI-204 EXHIBIT 3 XVI-205 EXHIBIT 3 XVI-206 EXHIBIT 4 XVI-207 EXHIBIT 4 XVI-208 EXHIBIT 5 XVI-209 EXHIBIT 6
| |
| , XVI-210 EXHIBIT 7, XVI-211 EXHIBIT 8 XVI-212 EXHIBIT 8 XVI-213 EXHIBIT 8 XVI-214 EXHIBIT 8 XVI-215 EXHIBIT 9 XVI-216 EXHIBIT 9 XVI-217 EXHIBIT 10 UFSAR Revision 13 EP 16-3 June 1995
| |
| | |
| I NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XVII T=TABLE STABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XVII-1 0 XVII-46 XVII-2 0 XVII-47 F XVII-33 XVII-3 F XVIZ-1 0 XVII-48 F XVII-34 XVII-4 F XVII-2 0 XVZI-49 F XVII-35 XVII-5 F XVII-3 0 XVII-50 F XVII-36 XVII-6 F XVII-4 0 XVII-51 F XVII-37 XVII-7 F XVZI-5 0 XVII-52 F XVII-38 XVII-8 F XVII-6 0 -XVII-53 XVII-9 F XVIZ-7 0 XVII-54 XVII-39 XVII-10 . F XVII-8 0 XVII-55 XVII-11 F XVII-9 0 XVII-56 T XVZI-5 XVII-12 F XVZI-10 0 XVII-57 F XVZI-40 XVII-13 F XVII-11 0 XVII-58 F XVIZ-41 XVII-14 F XVII-12 0 XVII-59 F XVII-42 XVII-15 F XVZI-13 0 XVII-60 F XVII-43 XVII-16 0 XVII-61 F XVII-44 XVII-17 XVII-1 0 XVII-62 F XVII-45 XVII-18 XVII-2 0 XVII-63 F XVII-46 XVII-19 0 XVII-64 T XVII-6 XVII-20 F XVII-14 0 XVII-65 XVII-21 F XVZZ-15 0 XVII-66 T XVII-7 XVII-22 F XVII-16 0 XVII-67 T XVII-8 XVII-23 F XVZI-17 0 XVII-68 T XVII-9 XVII-24 F XVII-18 0 XVII-69 'T XVII-10 XVII-25 F XVII-19 0 XVIZ-70 T XVII-11 XVII-26 T XVII-3 0 XVIZ-71 T XVII-12 XVII-27 F XVII-20 0 XVII-72 T XVII-13 XVII-28 F XVII-21 0 XVII-73 T XVII-14 XVII-29 F XVII-22 0 XVII-74
| |
| .XVII-30 F XVII-23 0 XVII-75 T XVII-15 XVII-31 F XVII-24 0 XVII-76 T XVII-16 XVII-32 F XVII-25 0 XVII-77 T XVII-17 XVII-33 F XVII-26 0 XVII-78 T XVII-18 XVII-34 F XVII-27 0 XVII-79 T XVZI-19 XVII-35 F XVII-28 0 XVII-80 T XVII-20 XVII-36 F XVII-29 0 XVIZ-81 T XVII-21 XVII-37 F XVII-30 0 XVII-82 T XVII-22 XVII-38 F XVII-31 0 XVZI-83 XVII-39 0 XVII-84 XVII-23 XVII-40 XVII-4 0 XVII-85 XVII-24 XVII-41 0 XVII-86 XVII-42 0 XVIZ-87 XVII-43 0 XVII-88 XVII-25 XVII-44 XVII-3 0 XVII-89 XVII-26 XVII-45 0 XVII-90 UFSAR Revision 13 EP 17-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XVII T=TABLE STABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XVII-91 0 XVII-136 F XVII-63 XVII-92 T XVII-27 0 XVII-137 F XVII-64 XVII-93 F XVII-47 0 XVII-138 XVII-94 F XVII-48 0 XVII-139 XVII-95 F XVII-49 0 XVII-140 XVII-96 F XVII-50 0 XVII-141 F XVII-65 XVII-97 F XVII-51 0 XVII-142 XVII-98 F XVII-52 0 XVII-99 F XVII-53 0 XVII-10Q 0 XVII-101 F XVII-54 0 XVII-102 F XVII-55 0 XVII-103 0 XVII-104 T XVII-28 0 XVII-105 T XVII-29 0 XVII-106 0 XVII-107 0 XVII-108 F XVII-56 0 XVII-109 0 XVII-110 0 XVII-111 0 XVII-112 F XVII-57 0 XVII-113 F XVII-58 0 XVII-114 0 XVII-115 0 XVII-116 0 XVZI-117 0 XVZI-118 0 XVII-119 T XVII-30 0 XVII-120 0 XVII-121 0 XVII-122 0 XVII-123 0 XVII-124 0 XVII-125 F XVII-59 0 XVII-126 0 XVII-127 0 XVIZ-128 0 XVII-129 0 XVII-130 0 XVII-131 0 XVIZ-132 0 XVII-133 F XVII-60 0 XVII-134 F XVII-61 0 XVII-135 F XVII-62 0 UFSAR Revision 13 EP 17-2 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES SECTION XVIII T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER XVIII-1 10 XVIII-2 10 XVIII-3 10 XVIII-4 10 XVIII-5 10 XVIII-6 10 XVIII-7 10 XVIII-8 10 XVIII-9 10 XVIII-10 10 XVIII-11 10 XVIII-12 10 XVIII-13 10 XVIII-14 10 XVIII-15 10 XVIII-16 10 XVIII-17 10 XVIII-18 10 XVIII-19 t 10 XVIII-20 T XVIII-1 12 UFSAR Revision 13 EP 18-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES APPENDIX A T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER A-1 UFSAR Revision 13 EP A-1 June 1995
| |
| | |
| NINE MILE POINT UNIT 1 FSAR (UPDATED)
| |
| LIST OF EFFECTIVE PAGES APPENDIX B T=TABLE T=TABLE PAGE F=FIGURE REVISION PAGE F=FIGURE REVISION NUMBER NUMBER NUMBER NUMBER NUMBER NUMBER B-i 11 B.10-3 B-ii 12 B.11-1 11 B.0-1 12 B.11-2 12 B. 1-1 12 B. 12-1 BE 1-2 12 B. 12-2 B.1-3 13 B.1-4 13 B. 13-1 12 B.1-5 12 B. 14-1 B.2-1 12 B.2-2 .12 B.15-1 12 B.2-3 12 B.15-2 11 B.2-4 12 B.2-5 12 B.16-1 B.2-6 12 B. 17-1 B.3-1 12 B. 17-2 B.3-2 12 B.3-3 11 B. 18-1 12 B.3-4 11 B.18-2 B.18-3 '1 12 B.4-1 11 B.4-2 12 T B-1 S?1 1 13 B.4-3 12 T B-1 S?1 2 13 T B-2 12 B.5-1 12 T,B-3 Sh 1 12 T B-3 S11 2 12 B.6-1 T B-3 Sh 3 12 B.6-2 T B-3 Sh 4 12 T B-3 Sh 5 12 B. 12 T B-3 Sh 6 12 B-3 7-1'.
| |
| 7-2 12 T Sh 7 12 B.7-3 11 T B-3 Sh 8 12 B.7-4 11 B.8-1 B.9-1 11 B.9-2 12 B.9-3 11 B.10-1 11 B.10-2 12 UFSAR Revision 13 EP B-1 June 1995
| |
| | |
| '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&MOHAWKPOWER CORPORATION SYRACUSE, NEW YORK
| |
| | |
| 0 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS Section Title P acae TABLE OF CONTENTS LIST OF FIGURES XXX LIST OF TABLES xiii SECTION I INTRODUCTION AND
| |
| | |
| ==SUMMARY==
| |
| | |
| A. PRINCIPAL DESIGN CRITERIA I-2 1.0 General I-2 2.0 Buildings and Structures I-2 3.0 Reactor I-3 4.0 Reactor Vessel I-4 5.0 Containment I-5 6.0 Control and Instrumentation Z-7 7.0 Electrical Power I-8 8.0 Radioactive Waste Disposal I-8 9.0 Shielding and Access Control I-9 10.0 Fuel Handling and Storage I-9 B. CHARACTERISTICS I-10 1.0 Site I-10 2.0 Reactor I-10 3.0 Core I-10 4.0 Fuel Assembly I-10.
| |
| 5.0 Control System I-11 6.0 Core Design and Operating Conditions I-11 7.0 Design Power Peaking Factor I-11 8.0 Nuclear Design Data I-12 9.0 Reactor Vessel I-12 10.0 Coolant Recirculation Loops I-12 11.0 Primary Containment I-12 12.0 Secondary Containment I-13 13.0 Structural Design I-13 14.0 Station Electrical System I-13 15.0 Reactor Instrumentation System I-13 16.0 Reactor Protection System I-14 C. IDENTIFICATION OF CONTRACTORS I-15 D. GENERAL CONCLUSIONS X-16 SECTION II STATION SITE AND ENVIRONMENT IX-1 A. SITE DESCRIPTION II-1 1.0 General II-1 2.0 Physical Features XI-1 3.0 Property Use and Development II-5 UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae B. DESCRIPTION OF AREA ADJ'ACENT TO THE SITE II-6 1.0 General II-6 1.1 Population ZZ-6 2.0 Agricultural, Industrial and Recreational Use XI-9 2.1 Agricultural Use II-9 2.2 Industrial Use IZ-11 2.3 Recreational Use ZZ-17 C. METEOROLOGY II-22 D. LIMNOLOGY XX-23 E. EARTH SCIENCES II-24 F. ENVIRONMENTAL RADIOLOGY II-25 SECTION IXI BUILDINGS AND STRUCTURES IIZ-1 A. TURBINE BUILDING III-4 1.0 Design Bases IZZ-4 1.1 Wind and Snow Loadings IIZ-4 1.2 Pressure Relief Design XII-4 1.3 Seismic Design and Internal Loadings IXZ-4 1.4 Heating and Ventilation III-6 1.5 Shielding and Access Control ZZZ-6 2.0 Structure Design IIZ-6 2.1 General Structural Features II1-6 2.2 Heating and Ventilation System III-17 2.3 Smoke and Heat Removal IZX-21 2.4 Shielding and Access Control III-21 3.0 Safety Analysis III-21 B. CONTROL ROOM IZI-23 1.0 Design Bases ZII-23 1.1 Wind and Snow Loadings III-23 1.2 Pressure Relief Design III-23 1.3 Seismic Design and Internal Loadings XIZ-23 1.4 Heating and Ventilation III-23 1.5 Shielding and Access Control III-23a 2.0 Structure Design III-24 2.1 General Structural Features ZZZ-24 2.2 Heating, Ventilation and Air Conditioning System ZIZ-25 2.3 Smoke and Heat. Removal ZZZ-27 2.4 Shielding and Access Control IXI-27 3.0 Safety Analysis ZZZ-28 UFSAR Revision 0'une 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae C. WASTE DISPOSAL BUILDING III-29 1.0 Design Bases IIZ-29 1.1 Wind and Snow Loadings ZI1-29 1.2 Pressure Relief Design IZZ-29 1.3 Seismic Design and Internal Loadings III-29 1.4 Heating and Ventilation III-30 1.5 Shielding and Access Control ZZI-30 2.0 Structure Design III-30 2.1 General Structural Features ZII-30 2.2 Heating and Ventilation System IIZ-32 2.3 Shielding and Access Control III-34 3.0 Safety Analysis IIZ-36 D. OFFGAS BUILDING ZII-38 1.0 Design Bases ZII-38 1.1 Wind and Snow Loadings III-38 1.2 Pressure Relief Design III-38 1.3 Seismic Design and Internal Loadings ZII-38 1.4 Heating and Ventilation ZZI-38 1.5 Shielding and Access Control III-38 2.0 Structure Design III-38 2.1 General Structural Features III-39 2.2 Heating and Ventilation System ZII-39 2.3 Shielding and Access Control ZII-41 3.0 Safety Analysis III-41 E. NONCONTROLLED BUILDINGS ZII-42 1.0 Administration Building III-42 1.1 Design Bases IIZ-42 1.1.1 Wind and Snow Loadings IZI-42 1.1.2 Pressure Relief Design III-42 1.1.3 Seismic Design and Internal Loadings III-42 1.1.4 Heating, Cooling and Ventilation ZII-43 1.1.5 Shielding and Access Control IIZ-43 1.2 Structure Design IZZ-43 1.2.1 General Structural Features III-43 1.2.2 Heating, Ventilation and Air Conditioning ZII-45 1.2.3 Access Control IZI-45 1.3 Safety Analysis IIZ-47 2.0 Sewage Treatment Building III-47 2.1 Design Bases IZZ-47a 2.1.1 Wind and Snow Loadings III-47a 2.1.2 Pressure Relief Design IZI-47a 2.1.3 Seismic Design and Internal Loadings IZZ-47a 2.1.4 Electrical Design ZII-47a 2.1.5 Fire and Explosive Gas Detection ZII-47a 2.1.6 Heating and Ventilation III-48 2.1.7 Shielding and Access Control ZZI-48 UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae 2.2 Structure Design III-48 2.2.1 General Structural Features IZX-48 2.2.2 Ventilation System ZII-49 2.2.3 Access Control III-49 3.0 Energy Information Center III-49 3.1 Design Bases IZI-49 3.1.1 Wind and Snow Loadings IXI-49 3.1.2 Pressure Relief Design III-50 3.1.3 Seismic Design and Internal Loadings XII-50 3.1.4 Heating and Ventilation ZZI-50 3.1.5 Shielding and Access Control III-50 3.2 Structure Design III-51 3.2.1 General Structural Features ZII-51 3.2.2 Heating and Ventilation System II 1-51 3.2.3 Access Control Zll-52 F. SCREENHOUSE~ INTAKE AND DISCHARGE TUNNELS II1-53 1.0 Screenhouse ZII-53 1.1 Design Basis III-53 1.1.1 Wind and Snow Loadings ZIZ-53 1.1.2 Pressure Relief Design ZZZ-53 1.1.3 Seismic Design and Internal Loadings III-53 1.1.4 Heating and Ventilation IIZ-53 1.1.5 Shielding and Access Control III-53 1.2 Structure Design III-53 2.0 Intake and Discharge Tunnels IIZ-57 2.1 Design Bases XII-57 2.2 Structure Design III-57 3.0 Safety Analysis III-59 G. STACK III-61 1.0 Design Bases ZZZ-61 1.1 General III-61 1.2 Wind Loading ZZZ-61 1.3 Seismic Design ZZI-61 1.4 Shielding and Access Control ZIZ-61 2.0 Structure Design ZZZ-61 3.0 Safety Analysis IZX-64 3.1 Radiology ZII-64 3.2 Stack Failure Analysis ZII-64 3.2.1 Reactor Building ZIX-66 3.2.2 Diesel-Generator Building ZZZ-66 3.2.3 Screen and Pump House ZZZ-67 H. SECURITY BUILDING AND SECURITY BUILDING ANNEX III-68 UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| ~Sectio Pacae RADWASTE SOLIDIFICATION AND STORAGE BUILDING III-69 1~0 Design Bases III-69 1.1 Wind and Snow Loadings XII-69 1.2 Pressure Relief Design III-69 1.3 Seismic Design and Internal Loadings III-69 1.4 Heating, Ventilation and Air Conditioning XXI-69 1~5 Shielding and Access Control III-70 2' Structure and Design III-70 2~1 General Structural Features III-70 2~2 Heating, Ventilation and Air Conditioning III-71 2 ' Shielding and Access Control III-73 3.0 Use III-73 4' References III-75 SECTION IV REACTOR IV-1 A. DESIGN BASES IV-1 1.0 General IV-1 2.0 Performance Ob)ecti.ves IV-1 3.0 Design Limits and Targets IV-2 B. REACTOR DESIGN IV-3 1.0 General IV-3 2' Nuclear Design Technique IV-5 2.1 Reference Loading Pattern XV-6 2' Final Loading Pattern IV-6 2.2.1 Acceptable Deviation From Reference Loading Pattern IV-6 2.2 ' Reexamination of Licensing Basis IV-7 2.3 Refueling Cycle Reactivity Balance IV-8 3.0 Thermal and Hydraulic Characteristics IV-8 3.1 Thermal and Hydraulic Design IV-8 3.1 1
| |
| ~ Recirculation Flow Control IV-8
| |
| : 3. 1.2 Core Thermal Limits IV-9 3.2 Thermal and Hydraulic Analyses IV-15 3.2.1 Hydraulic Analysis IV-15 3.2.2 Thermal IV-17 3.3 Transients Analysis'eactor IV-23 4.0 Stability Analysis IV-23 4.1 Design Bases IV-23 4' Stability Analysis Method IV-24 5 ' Mechanical Design and Evaluation IV-25 5.1 Fuel Mechanical Design IV-25 5' ' Design Bases IV-25 UFSAR Revision 13 June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title
| |
| 'L Pacae 5.1.2 Fuel Rods IV-27 5.1.3 Water Rods ZV-29 5.1.4 Fuel Assemblies IV-29 5.1.5 Mechanical Design Limits and Stress Analysis IV-30 5.1.6 Relationship Between Fuel Design Limits and Fuel Damage Limits ZV-31 5.1.7 Surveillance and Testing IV-31 6.0 Control Rod Mechanical Design and Evaluation ZV-32 6.1 Design IV-32 6.1.1 Control Rods and Drives ZV-32 6.1.2 Standby Liquid Poison System ZV-40 6.2 Control System Evaluation ZV-40 6.2.1 Rod Withdrawal Errors Evaluation IV-40 6.2.2 Overall Control System Evaluation ZV-41 6.3 Limiting Conditions for Operation and Surveillance ZV-45 6.4 Control Rod Lifetime IV-45 7.0 Reactor Vessel Internal Structure ZV-46 7.1 Design Bases IV-46 7.1.1 Core Shroud IV-47 7.1.2 Core Support IV-49 7.1.3 Top Grid IV-49 7.1.4 Control Rod Guide Tubes ZV-49 7.1.5 Feedwater Sparger IV-50 7.1.6 Core Spray Spargers ZV-50 7.1.7 Liquid Poison Sparger IV-50 7.1.8 Steam Separator and Dryer ZV-50 7.2 Design Evaluation IV-51 7.3 Surveillance and Testing IV-52 SECTION V REACTOR COOLANT SYSTEM A. DESIGN BASES V-1 1.0 General V-1 2.0 Performance Objectives V-1 3.0 Design Pressure V-6 4.0 Cyclic Loads (Mechanical and Thermal) V-6 5;0 Codes V-8 B. SYSTEM DESIGN AND OPERATION v-10 1.0 General V-10 1.1 Drawings V-10 1.2 Materials of Construction V-10 1.3 Thermal Stresses V-10 1.4 Primary Coolant Leakage V-12 1.5 Coolant Chemistry V-12a UFSAR Revision 12 Vi June 1994
| |
| | |
| Nine Mile Point Unit, 1 FSAR TABLE OF CONTENTS (Cont'.)
| |
| Section Title PBcee 2.0 Reactor Vessel V-12a 3.0 Reactor Recirculation Loops V-14 4.0 Reactor Steam and Auxiliary Systems Piping V-16 5.0 Relief Devices V-16 C. SYSTEM DESIGN EVALUATION V-19 1.0 General V-19 2.0 Pressure V-19 3.0 Design Heatup and Cooldown Rates V-21 4.0 Materials Radiation Exposure V-21 4.1 Pressure-Temperature Limit Curves V-21 4.2 Temperature Limits for Boltup V-22 4.3 Temperature Limits for In-Service System Pressure Tests V-22 4.4 Operating Limits During Heatup, Cooldown, and Core Operation V-23a 4.5 Predicted Shift in RT~~ V-23a 5.0 Mechanical Considerations V-23a 5.1 Jet Reaction Forces V-23a 5.2 Seismic Forces V-23b 5.3 Piping Failure Studies v-26 6.0 Safety Limits, Limiting Safety Settings and Minimum Conditions for Operation V-2 6 D. TESTS AND INSPECTIONS V-27
| |
| : 1. 0. Prestartup Testing V-27 2.0 Inspection and Testing Following Startup V-27 2.1 Hydro Pressure V-27 2.2 Pressure Vessel Irradiation V-27 E. EMERGENCY COOLING SYSTEM V-28 1.0 Design Bases V-28 2.0 System Design and Operation V-28 3.0 Design Evaluation v-30 3.1 Redundancy V-30 3.2 Makeup Water v-30 3.3 System Leaks V-30 3.4 Containment Isolation V-31 4.0 Tests and Inspections V-31 4.1 Prestartup Test V-31 4.2 Subsequent Inspections and Tests V-31 SECTION VI CONTAXNMENT SYSTEM VI-1 PRXMARY CONTAINMENT-MARK I CONTAXNMENT PROGRAM VI-2 UFSAR Revision 12 Vii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| ~Sectic T~it e ~acae 1~0 General Structure VI-2 2.0 Pressure Suppression Hydrodynamic Loads VI-2 2.1 Safety/Relief Valve Discharge VI-3 2.2 Loss-of-Coolant Accident VI-3 2.3 Summary of Loading Phenomena VI-5 3.0 Plant Unique Modifications VI-6 B. PRIMARY CONTAINMENT PRESSURE SUPPRESSION SYSTEM VI-8 1.0 Design Bases VI-8 1~1 General VI-8 1~2 Design Basis Accident VI-8 1' Containment Heat Removal VI-11 1' Isolation Criteria VI-11 1.5 Vacuum Relief Criteria VI-11 1.6 Flooding Criteria VI-12 1.7 Shielding VI-12 2.0 Structure Design VI-13 2.1 -General VI-13 2.2 Penetrations and Access Openings VI-15 2.3 Jet and Missile Protection VI-21 2.4 Materials VI-22 2.5 Shielding VI-22a
| |
| ,- VI-22a 2.6, Vacuum Relief 2.7 Containment Flooding VI-23 C. SECONDARY CONTAINMENT REACTOR BUILDING VI-24 1~0 Design Bases VI-24 1~1 Wind and Snow Loadings VI-24 1' Pressure Relief Design VI-24 1.3 Seismic Design VI-25 1.4 Shielding VI-25 2~0 Structure Design VI-25 2.1 General Structural Features VI-26 D. CONTAINMENT ISOLATION SYSTEM VI-43 1.0 Design Bases VI-43 1.1 Containment Spray Appendix J Water Seal Requirements VI-56a 2.0 System Design VI-57 3.0 Tests and Inspections VI-59 E., CONTAINMENT VENTILATION SYSTEM VI-60 1.0 Pri'mary Containment VI-60 1 1 ,Design Bases VI-60 1.2 System Design VI-60 2.0 Secondary Containment VI-62 UFSAR Revision 13 viii June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| ~Sectio ~ae 2~1 Design Bases VI-62 2' System Design 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 ~ae F. TEST AND INSPECTIONS VI-65 1.0 Drywell and Suppression Chamber VI-65 1.1 Preoperational Testing VI-65 1.2 Postoperational Testing VI-65 2.0 Containment Penetrations and Isolation Valves VI-66 2.1 Penetration and Valve Leakage VI-66 2' Valve Operability Test VI-67 3.0 Containment Ventilation System VI-67 4.0 Other Containment Tests VI-67 5.0 Reactor Building VI-68 5.1 Reactor Building Normal Ventilation System VI-68
| |
| '5 ~ 2 Reactor Building Isolation Valves VI-68 5.3 Emergency Ventilation System VI-68 SECTION VII ENGINEERED SAFEGUARDS VII-1 A. CORE SPRAY SYSTEM VII-2 1.0 Design Bases VII-2 2.0 System Design VII-2 2.1 General VII-2 2.2 Operator Assessment VII-7 3.0 Design Evaluation VII-8 4.0 Tests and Inspections VII-9 B. CONTAINMENT SPRAY SYSTEM VII-11 1.0 Design Bases VII-11 2.0 System Design VII-11a 2' Operator Assessment VII-17 3.0 Design Evaluation VII-18 4.0 Tests and Inspections VII-19 C. LIQUID POISON INJECTION SYSTEM VII-21 1.0 Design Bases VII-21 2.0 System Design VII-21a 2~1 Operator Assessment VII-27 3.0 Design Evaluation VII-28 4.0 Tests and Inspections VII-29 5.0 Alternate Boron Injection VII-30 D. CONTROL ROD VELOCITY LIMITER VII-31 1.0 Design Bases VII-31 2.0 System Design VII-31 3.0 Design Evaluation VII-34 3.1 General VII-34 3.2 Design Sensitivity VII-34 3.3 Normal Operation VII-35 4.0 Tests and Inspections VII-36 UFSAR Revision 13 iX June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae E. CONTROL ROD HOUSING SUPPORT VZI-37 1.0 Design Bases VZI-37 2.0 System Design VI1-38 2.1 Loads and Deflections VZZ-40 3.0 Design Evaluation VII-41 4.0 Tests and Inspections VII-41 F. FLOW RESTRICTORS VZZ-42 1.0 Design Bases VZI-42 2.0 System Design VIZ-42 3.0 Design Evaluation VII-42 4.0 Tests and Inspections VZZ-43 G. COMBUSTIBLE GAS CONTROL SYSTEM VZZ-44 1.0 Design Bases VZZ-44 2.0 Containment Inerting System VII-44 2.1 System Design VZZ-44 2.2 Design Evaluation VZZ-46a 3.' Containment Atmospheric Dilution System VZZ-48 3.1 System Design VZZ-48 3.2 Design Evaluation VIX-49 4.0 Tests and Inspections VZI-53 H. EMERGENCY VENTILATION SYSTEM VZI-55 1.0 Design Bases VII-55 2.0 System Design VII-55a 2.1 Operator Assessment VZI-58 3.0 Design Evaluation vZZ-58 4.0 Tests and Inspections VII-59 I. HXGH-PRESSURE COOLANT ZNJECTXON VII-61 1.0 Design Bases VZZ-61 2.0 System Design VII-61 3.0 Design Evaluation VZI-61a 4.0 Tests and Xnspections VZZ-62a SECTION VIII INSTRUMENTATZON AND CONTROL VIXX-1 A. PROTECTIVE SYSTEMS VIIZ-1 1.0 Design Bases VIII-1 1.1 Reactor Protection System VXII-1 1.2 Anticipated Transients Without Scram Mitigation System VIII-6 2.0 System Design VIZ1-6 2.1" Reactor Protection System VZZZ-6 2.2 Anticipated Transients Without Scram Mitigation System VXXI-14 3.0 System Evaluation VXII-15 UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section ~ae B. REGULATING SYSTEMS VIII-18 1.0 Design Bases VIII-18 2~0 System Design VIII-18 2~1 Control Rod Adjustment Control VIII-18 2.2 Recirculation Flow Control VIII-18 2.3 Pressure and Turbine Control VIII-20 2.4 Reactor Feedwater Control VIII-21 3.0 System Evaluation VIII-21 3' Control Rod Adjustment Control VIII-21 3.2 Recirculation Flow Control VIII-22 3 ' Pressure and Turbine Control VIII-22 3.4 Reactor Feedwater Control VIII-22 C. INSTRUMENTATION SYSTEMS VIII-23 1.0 Nuclear Instrumentation VIII-23 1.1 Design VIII-23 1~1~1 Source Range Monitors VIII-25 1.1 ~ 2 Intermediate Range Monitors VIII-28 1.1 ~ 3 Local Power Range Monitors VIII-32 1.1 ~ 4 Average Power Range Monitors VIII-32
| |
| : 1. 1.5 Traversing In-Core Probe System VIII-40 1.2 Evaluation VIII-40 1.2.1 Source Range Monitors VIII-42 1.2.2 Intermediate Range Monitors VIII-43 1.2.3 Local Power Range Monitors VIII-50 1.2.4 Average Power Range Monitors VIII-50 2.0 Non-Nuclear Process Instrumentation VIII-54 2.1 Design Bases VIII-54 2.1.1 Non-Nuclear Process Instruments in-Protective System VIII-54
| |
| : 2. 1.2 Nonnuclear Process Instruments in Regulating Systems VIII-58
| |
| : 2. 1.3 Other Nonnuclear Process Instruments VIII-60 2 ' Evaluation VIII-62 2 '.1 Nonnuclear Process Instruments in Protective System VIII-62 2.2.2 Nonnuclear Process Instruments in Regulating Systems VIII-62 2 '.3 Other Nonnuclear Process Instruments VIII-62a 3.0 Radioactivity Instrumentation VIII-62a 3 ' Design Bases VIII-62a 3 ~ 1~1 Radiation Monitors in Protective Systems VIII-62a
| |
| : 3. 1.2 Other Radiation Monitors VIII-68 3.2 Evaluation VIII-76 4.0 Other Instrumentation VIII-76 4.1 Rod Worth Minimizer VIII-76 UFSAR Revision 13 June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR STABLE OF CONTENTS (Cont'd.)
| |
| Section Title P RCRe 4.1.1 Design Bases VIXI-76 4.1.2 Evaluation VIXI-80 5.0 Regulatory Guide 1.97 (Revision 2)
| |
| Instrumentation VIXI-80 5.1 Licensing Activities Background VZ1Z-81 5.2 Definition of RG 1.97 Variable Types and Instrument Categories VIII-81 5.3 Determination of RG 1.97 Type A Variables for NMP1 VIII-82a 5.4 Determination of "EOP Key Parameters" for NMP1" VIII-82b 5.4.1 Determination Basis/Approach VIIX-82b 5.4.2 Definition of Primary Safety Functions VZII-82c 5.4.3 Association of EOPs to Primary Safety Functions VIII-82d 5.4.4 Identification of EOP Key Parameters VIII-84 5.5 NMP1 RG 1.97 Variables, Variable Type, and Associated Instrument Category Designations VIII-86 5.6 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 VIII-99 5.6.1 No Type A Variables VIXI-99 5.6.2 EOP Key Parameters VI1I-100 5.6.3 Single Tap for the Fuel Zone RPV Water Level Instrument VIII-101 5.6.4 Nonredundant Wide-Range RPV Water Level Indication VIXI-102 5.6.5 Upgrading EOP Key Parameter Category 1 Instrument Loop Components to Safety Related Classification VIIZ-104 5.6.6 Safety Related Classification of Instrumentation for RG 1.97 Variable Types Other than the EOP Key Parameters VIIX-105 Routing and Separation of Channelized Category 1 Instrument Loop Cables VIII-105 I 5 ' ' Electrical Isolation of Category 1 Instrument Loops from Associated Components that are not Safety Related VIII-106 5.6.9 Power Source Information for Category 1 Instruments VIII-107 UFSAR Revision 12 Xii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section ~ae 5.6.10 Marking of Instruments of Control Room Panels VZIZ-108 5.6.11 "Alternate" Instruments for Monitoring EOP Key Parameters VIIZ-108 5.6.12 Indicating Ranges of Monitoring Instruments VIII-109 5.7 References VIII-110 SECTION IX ELECTRICAL SYSTEMS IX-1 A. DESIGN BASES IX-1 B. ELECTRICAL SYSTEM DESIGN IX-2 1.0 Network Interconnections IX-2 1.1 345-kV System IX-2 1.2 115-kV System IX-3 2.0 Station Distribution System IX-8a 2.1 Two f24-Volt DC Systems IX-13 2.2 Two 120-V, 60-Hz, Single-Phase, Uninterruptible Power Supply Systems IX-15 2.3 Two 120-V, 57-60 Hz, One-Phase, Reactor Trip Power Supplies IX-16 2.4 One 120/208-V, 60-Hz, Instrument and Control Transformer IX-16 2.5 One 120/240-V, 60-Hz, Three-Phase, Computer Power Supply IX-16a 3.0 Cables and Cable Trays IX-16a 3.1 Cable Separation IX-'16a 3' Cable Penetrations IX-17 3.3 Protection in Hazardous Areas IX-17 3.4 Types of Cables ZX-18 3.4.1 Power Cable IX-18 3.4.2 Control Cable IX-18 3.4.3 Special Cable IX-19 3.5 Design and Spacing of Cable Trays IX-19 3.5.1 Tray Design Specifications IX-19 3.5.2 Tray Spacing IX-20 4.0 Emergency Power IX-20 4.1 Diesel Generator System IX-20 4.2 Station Batteries IX-26 4.3 Nonsafety Battery System IX-31 5.0 Tests and Inspections IX-34 5.1 Diesel-Generator IX-34 5.2 Station Batteries IX-35 5.3 Nonsafety Batteries IX-35 6.0 Conformance with 10CFR50.63, Station Blackout Rule IX-35 6.1 Station Blackout Duration IX-35 UFSAR Revision 13 xiii June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section ~ae 6.2 Station Blackout Coping Capability IX-36 6.3 Procedures and Training IX-38 6.4 Quality Assurance IX-39 6.5 EDG Reliability Program IX-39 6.6 References IX-40 SECTION X REACTOR AUXILIARY AND EMERGENCY SYSTEMS X-1 A. REACTOR SHUTDOWN COOLING SYSTEM X-1 1' Design Bases X-1 2.0 System Design X-1 3.0 System Evaluation X-3 4.0 Tests and Inspections X-3 1
| |
| B. REACTOR CLEANUP SYSTEM X-4 1.0 Design Bases X-4 2.0 System Design X-4 3.0 System Evaluation X-7 4.0 Tests and Inspections X-7 C. CONTROL ROD DRIVE HYDRAULIC SYSTEM X-8 1.0 Design Bases X-8 2.0 System Design X-8 2~1 Pumps X-10 2.2 Filters X-10 2.3 First Pressure Stage X-11 2.4 Second Pressure Stage X-11 2.5 Third Pressure Stage X-12 2.6 Exhaust Header X-13 2.7 Accumulator X-13 2.8 Scram Pilot Valves X-14 2.9 Scram Valves X-14 2.10 Scram Dump Volume X-15 2 '1 2.12 Control Rod Drive Cooling System Directional Control and Speed X-16 Control Valves X-16
| |
| : 2. 13 Rod Insertion and Withdrawal X-17
| |
| : 2. 14 Scram Actuation X-17 3.0 System Evaluation X-18 3.1 Normal Withdrawal Speed X-18 3.2'.3 Accidental Multiple Operation X-19 Scram Reliability X-19 3.4 Operational Reliability X-20 3.5 Alternate Rod Injection (ARI) X-21 4.0 Reactor Vessel Level Instrumentation Reference Leg Backfill X-21 5.0 Tests and Inspections X-21 UFSAR Revision 13 xiv June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Sect o D. REACTOR BUILDING CLOSED LOOP COOLING WATER SYSTEM X-22 1' Design Bases X-22 2.0 System Design X-22 3.0 Design Evaluation X-25 4.0 Tests and Inspections X-26 E. TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM X-27 1.0 Design Bases X-27 2 ' System Design X-27 3.0 Design Evaluation X-29 4.0 Tests and Inspections X-30 F. SERVICE WATER SYSTEM X-31 1.0 Design Bases X-31 2.0 System Design X-31 3.0 Design Evaluation X-33 4.0 Tests and Inspections X-34 G. MAKEUP WATER SYSTEM X-35 1.0 Design Bases X-35 2 ' System Design X-35 3.0 System Evaluation X-37 4.0 Tests and Inspections X-38 H. SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM X-39 1 ' Design Bases X-39 2.0 System Design X-39a 3.0 Design Evaluation X-43 4.0 Tests and Inspections X-44 BREATHING, INSTRUMENT AND SERVICE AIR SYSTEM X-45 1.0 Design Bases X-45 2.0 System Design X-45 3.0 Design Evaluation X-48 4.0 Tests and Inspections X-49 FUEL AND REACTOR COMPONENTS HANDLING SYSTEM X-50 1~0 Design Bases X-50 2.0 System Design X-50 2.1 Description of Facility X-50 2.1.1 Cask Drop Protection System X-54 2.2 Operation of the Facility X-55 3.0 Design Evaluation X-57 4.0 Tests and Inspections X-58 UFSAR Revision 13 xv June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae K. FIRE PROTECTION PROGRAM X-59 1.0 Program Bases X-59 1.1 Nuclear Division Directive Fire Protection Program X-59 1.2 Nuclear Division Interface Procedure Fixe Protection Program X-59 1.3 Fire Hazards Analysis (FHA) X-59 1.4 Appendix R Review Safe Shutdown Analysis (SSA) X-60 1.5 Fire Protection and Appendix R Related Portions of Operations Procedures (OPs, SOPs, and EOPs) and Damage Repair Procedures X-60 1.6 Fire Protection Portions of the Emergency Plan X-60 2.0 Program Implementation and Design Aspects X-60 2.1 Fire Protection Implementing Procedures X-60 2.2 Fire Protection Administrative Controls X-61 2.3 Fire Protection System Drawings and Calculations X-61 2.4 Fire Protection Engineering Evaluations (FPEEs) X-61 3.0 Monitoring and Evaluating Program Implementation X-62 3.1 Fire Protection Quality Assurance Program (FPQAP) x-62 3.2 Fire Brigade Manning, Training, Drills and Responsibilities X-62 4.0 Surveillance and Tests x-62 L. REMOTE SHUTDOWN SYSTEM X-76 1.0 Design Bases X-76 2.0 System Design X-76 3.0 System Evaluation X-76 4.0 Tests and Inspections X-77 M. SAFETY PARAMETER DISPLAY SYSTEM X-78 1.0 Design Bases X-78 2.0 System Design X-78 3.0 System Evaluation X-78 4.0 Tests and Inspections X-79 I APPENDIX 1 OA FIRE HAZARDS ANALYSIS I APPENDIX 10B SAFE SHUTDOWN ANALYSIS UFSAR Revision 12 Xvi June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Ti.tie P RcRB SECTION XZ STEAM TO POWER CONVERSION SYSTEM XZ-,1 A. DESIGN BASES XI-1, B. SYSTEM DESIGN AND OPERATION XI-9 1.0 Turbine-Generator XZ-9 2.0 Turbine Condenser XI-11 3.0 Condenser Air Removal and Offgas System XI-12 4.0 Circulating Water System XI-17 5.0 Condensate Pumps XZ-17 6.0 Condensate Demineralirer System XI-18 7.0 Condensate Transfer System XI-19 8.0 Feedwater Booster Pumps XI-20 9.0 Feedwater Pumps XI-20 10.0 Feedwater Heaters XZ-20 C. SYSTEM ANALYSIS XZ-22 D. TESTS AND INSPECTIONS XZ-25 SECTION XZZ RADIOLOGICAL CONTROLS XII-1 RAD10ACTZVE WASTES XZI-1 1.0 Design Bases XII-1 1.1 Objectives XZZ-1 1.2 Types of Radioactive Wastes XIZ-1 1.2.1 Gaseous Wastes XZZ-1 1.2.2 Liquid Wastes XZI-2 1.2.3 Solid Wastes XZZ-3 2.0 System Design and Evaluation XII-3 2.1 Gaseous Waste System XII-3 2.1.1 Offgas System XII-7 2.1.2 , Steam Packing Exhauster System XII-7 2.1.3 Buildup Ventilation Systems XZZ-7 2.1.4 Stack XII-8 2.2 Liquid Waste System XII-8 2.2.1 Liquid Waste Handling Processes XZI-8 2.2.2 Sampling and Monitoring Liquid Wastes XII-12 2.2.3 Liquid Waste Equipment. Arrangement XII-13 2.2.4 Liquid Radioactive Waste System Control XII-13 2~3 Solid Waste System XII-15 2.3.1 Solid Waste Handling Processes XZZ-15 2.3.2 Solid Waste System Equipment XIZ-17 3.0 Safety Limits XII-17 4.0 Tests and Inspections XZI-18 4.1 Waste Process Systems XIZ-18 i
| |
| U UFSAR Revision 12 XV3. June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| ~sectic ~ae 4.2 Filters XII-18 4.3 Effluent Monitors XII-19 4.3.1 Offgas and Stack Monitors XII-19 4.3.2 Liquid Waste Effluent Monitor XII-19 B. RADIATION PROTECTION XII-20 1.0 Primary and Secondary Shielding XII-20 1.1 Design Bases XII-20 1.2 Design XII-22 1.2.1 Reactor Shield Wall XII-22 1' ' Biological Shield XII-23 1.'2. 3 Miscellaneous XII-23 1' Evaluation XII-23 2.0 Area Radioactivity Monitoring Systems XII-24 2.1 Area Radiation Monitoring System XII-24 2.1.1 Design Bases XII-24 2.1.2 Design XII-25 2.1 ~ 3 Evaluation XII-30 2.2 Area Air Contamination Monitoring System XII-31 2.2.1 Design Bases XII-31 2.2.2 Design XII-31 2 2 ' Evaluation XII-31 3.0 Radiation Protection XII-32 3 ~ 1 Facilities XII-32a
| |
| : 3. 1.1 Laboratory, Counting Room and Calibration Facilities XII-32a
| |
| : 3. 1.2 Change Room and Laundry Facilities XII-33
| |
| : 3. 1.3 Personnel Decontamination Facility XII-34 3.1.4 Tool and Equipment Decontamination Facility XII-34 3 ' Radiation, Control XII-35 3.2.1 Shielding XII-35 3.2.2 Access Control XII-36 3.3 Contamination Control XII-37 3.3.1 Facility Contamination Control XII-38 3.3 ' Personnel Contamination Control XII-38 3.3.3 Airborne Contamination Control XII-39 3.4 Personnel. Dose Determinations XII-41 3.4.1 Radiation Dose XII-41 3.5 Radiation Protection Instrumentation XII-41a 3.5.1 Counting Room Instrumentation XII-41a 3.5.2 Portable Radiation Instrumentation XII-42 3.5.3 Air Sampling Instrumentation XII-42 3.5 ' Personnel Monitoring Instruments XII-43 3.5.5 Emergency Instrumentation XII-43 4.0 Tests and Inspections XII-44 UFSAR Revision 13 xvi1 1 June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Pacae 4.1 Shielding XII-44 4.2 Area Radiation Monitors XII-44 4.3 Area Air Contamination Monitors XII-45 4' Radiation Protection Facilities XII-45 4.4 ' Ventilation Air Flows XII-45 4.4.2 Instrument Calibration Well Shielding 'XII-45 4.5 Radiation Protection Instrumentation XII-45 SECTION XIII CONDUCT OF OPERATIONS XIII-1 A. ORGANIZATION AND RESPONSIBILITY XIII-1 1.0 Management and Technical Support Organization XIII-1 Nuclear Division XIII-1 Vice President Nuclear Generation XIII-1 Vice President Nuclear Engineering XIII-2'III-3
| |
| : 1. 1.3 Vice President Nuclear Safety Assessment and Support 1.1 4 Director Nuclear Communications and
| |
| ~
| |
| Public Affairs XIII-6
| |
| : 1. 1.5 Director Human Resource Development XIII-7 1 '.6 1.2 Controller Nuclear Division Corporate Support Departments XIII-7 XIII-7 2.0 Operating Organization XIII-8 2.1 Plant Manager XIII-8 3.0 Quality Assurance XIII-11 4.0 Facility Staff Qualifications XIII-12 B. QUALIFICATIONS AND TRAINING OF PERSONNEL, XIII-13 1.0 This Section Deleted XIII-13 2.0 This Section Deleted XIII-13 3.0 This Section Deleted XIII-13 4.0 Training of Personnel= XIII-13 4.1 General Responsibility XIII-13 4.2 Implementation XIII-13 4.3 Quality XIII-13 4.3. 1 For Operator Training XIII-13 4.3.2 For Maintenance XZZZ-14 4.3.3 For Technicians XIII-14 4.3.4 For General Employee Training/Radiation Protection and Emergency Plan XIII-14 4 ' ' For Industrial Safety XIII-14 4.3.6 For Nuclear Quality Assurance XIII-14 4.3.7 For Fire Brigade XIII-14 UFSAR Revision 13 XiX June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title P acae Training of Licensed Operator Candidates/Licensed NRC Operator Retraining XIII-14 5.0 Cooperative Training with Local, State and Federal Officials XIII-15 C. OPERATING PROCEDURES XIII-16 D. EMERGENCY PLAN AND PROCEDURES XIII-17 E. SECURITY XIII-19 F. RECORDS XIII-20 1.0 Operations XIII-20 1.1 Control Room Log Book XIII-20 1.2 Station Shift Supervisor's Book XIII-20 1.3 Radwaste Log Book XZZI-20 1.4 Waste Quantity Level Shipped XIII-20 2.0 Maintenance XIII-21 3.0 Radiation Protection XIII-21 3.1 Personnel Exposure XIII-21 3.2 By-Product Material as Required by 10CFR30 XIII-21 3.3 Meter Calibrations XIII-21 3.4 Station Radiological Conditions in
| |
| - Accessible Areas XIII-21 3.5 Administration of the Radiation Protection Program and Procedures XTII-21 4.0 Chemistry and Radiochemistry XIII-21 5.0 Special Nuclear Materials XIII-21 6.0 Calibration of Instruments XZII-22 7.0 Administrative Records and Reports XIII-22 G. REVIEW AND AUDIT OF OPERATIONS XIII-23 1..0 Station Operations Review Committee XIII-23 1 ' Function XZZZ-23 2.0 Safety Review and Audit Board XIII-23 2.1 Function XIII-23 3.0 Review of Operating Experience XIII-24 SECTION XZV INITIAL TESTING AND OPERATIONS XZV-1 A. TESTS PRIOR TO INITIAL REACTOR FUELING XIV-1 B. INITIAL CRITICALITY AND POSTCRITICALZTY TESTS XIV-5 1.0 Initial Fuel Loading and Near-Zero XZV-5 Power Tests at: Atmospheric Pressure UFSAR Revision 12 XX June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR
| |
| 'L TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae 1.1 General Recpxirements XZV-5 1.2 General Procedures XIV-5 1.3 Core Loading and Critical Test Program XZV-8 2.0 Heatup from Ambient to Rated Temperature XIV-10 2.1 General XZV-10 2.2 Tests Conducted XIV-10 3.0 From Zero to 100 Percent Initial Reactor Rating XIV-12 4.0 Full-Power Demonstration Run XIV-14 5.0 Comparison of Base Conditions XZV-14 6.0 Additional Tests at Design Rating XIV-14 SECTION XV SAFETY ANALYSIS XV-1 A. INTRODUCTION XV-1 B. BOUNDARY PROTECTION SYSTEMS XV-2 1.0 Transients Considered XV-2 2.0 Methods and Assumptions XV-5a 3.0 Transient Analysis XV-6 3.1 Turbine Trip Without Bypass XV-6 3.2 Loss of 100'F Feedwater Heating XV-7a 3.3 Feedwater Controller Failure Maximum Demand XV-9 3.4 Control Rod Withdrawal Error XV-11a 3.5 Main Steam Line Isolation Valve Closure (With Scram) XV-18 3.6 Inadvertent Startup of Cold Recirculation Loop XV-20 3.7 Recirculation Pump Trips XV-24 3.8 Recirculation Pump Stall XV-27 3.9 Recirculation Flow Controller Malfunction Increasing Flow XV-28 3.10 Flow Controller Malfunction Decrease Flow XV-32 3.11 Inadvertent Actuation of One Solenoid Relief Valve XV-34 3.12 Safety Valve Actuation XV-36 3'. 12. 1 Objectives XV-36 3.12.2 Assumptions and Initial Conditions XV-36 XV-37 3.12.3 Comments 3.12.4 Results XV-37 3.13 Feedwater Controller Malfunction (Zero Demand) XV-41 3.14 Turbine Trip with Partial Bypass (Low Power) XV-43 UFSAR Revision 12 xx1 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae 3.15 Turbine Trip with Partial Bypass (Full Power) XV-44 3.16 Inadvertent Actuation of One Bypass Valve XV-48 3,17 One Feedwater Pump Trip and Restart xv-50 3.18 Loss of Main Condenser Vacuum xv-50 3.19 Loss of Electrical Load (Generator Trip) XV-52 3.20 Loss of Auxiliary Power XV-54 3.21 Pressure Regulator Malfunction XV-56 3.22 Instrument Air Failure XV-57 3.23 D-C Power Interruptions XV-66 3.24 Failure of One Diesel-Generator to Start XV-67 3.25 Power Bus Loss of Voltage xv-68 C. STANDBY SAFEGUARDS ANALYSIS XV-70 1.0 Main Steam Line Break Outside the Drywell XV-70 1.1 Identification of Causes XV-70 1.2 Accident Analysis XV-70 1.2.1 Valve Closure Initiation XV-71 1.2.2 Feedwater Flow XV-71 1.2.3 Core Shutdown XV-72 1.2.4 Mixture Level XV-72 1.2.5 Subcooled Liquid XV-72 1.2.6 System Pressure and Steam-Water Mass XV-72 1.2.7 Mixture Impact Forces XV-73 1.2.8 Core Internal Forces XV-75 1.3 Radiological Effects XV-75 1.3.1 Radioactivity Releases XV-75 1.3.2 Meteorology and Dose Rates XV-77 1.3.3 Comparison with Regulatory Guide 1.5 XV-78 2.0 Loss-of-Coolant Accident XV-81 2.1 Introduction XV-81 I 2.2 Input to Analysis xv-81 a 2.2.1 Operational and ECCS Input Parameters XV-81a 2.2.2 Single Failure Study on ECCS Manually-Controlled Electrically-Operated Valves xv-81a 2.2.3 Single Failure Basis xV-96 2.2.4 Pipe Whip Basis xv-96 2.3 Appendix K LOCA Performance Analysis for P8DNB277 XV-96 2.4 Appendix K LOCA Performance Analysis XV-134 UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Ti.tie PRcRB 2.4.1 Computer Codes XV-137a 2.4.2 Description of Model Changes XV-137a 2.4.3 Analysis Procedure XV-137b 2.4.4 Analysis Results XV-137d 2.5 Appendix K LOCA MAPLHGR Evaluation with Core Spray Flow Through One Sparger XV-137d 3.0 Refueling Accident XV-137d2 3.1 Identification of Causes XV-137d2 3.2 Accident Analysis XV-13 9 3.3 Radiological Effects XV-142 3.3.1 Fission Product Releases XV-142 3.3.2 Meteorology and Dose Rates XV-145 3.3.3 Comparison to Regulatory Guide 1.25 XV-145 4.0 Control Rod Drop Accident XV-14 6 4.1 Identification of Causes XV-14 6 4.2 Accident Analysis XV-14 9 4 ' Designed Safeguards XV-152 4.4 Procedural Safeguards XV-153 4,5 Radiological Effects XV-154 4.5.1 Fission Product Releases XV-154 4.5.2 Meteorology and Dose Rates XV-15 9 5.0 Containment Design Basis Accident XV-15 9 5.1 Original Recirculation Line Rupture Analysis With Core Spray XV-159a 5.1.1 Purpose XV-159a 5.1.2 Analysis Method and Assumptions XV-159a 5.1.3 Core Heat Buildup XV-15 9b 5.1.4 Core Spray System XV-159c 5.1.5 Containment Pressure Immediately Following Blowdown XV-159f 5.1.6 Containment Spray XV-159i 5.1.7 Blowdown Effects on Core Components XV-159k 5.1.8 Radiological Effects XV-159m 5.2 Original Containment Design Basis Accident Analysis Without Core Spray XV-160 5.2. 1 Purpose XV-160 5.2.2 Core Heatup XV-160 5.2.3 Containment Response XV-164 5.2.4 Fission Product Release from the Fuel XV-164a 5.2.5 Fission Product Release from the Reactor and Containment XV-167 5.2.6 Meteorology and Dose Rates XV-169 5.3 Design Basis Reconstitution (DBR)
| |
| Suppression Chamber Heatup Analysis XV-169 5.3.1 Introduction XV-169 5.3.2 Input to Analysis XV-169b UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae 5.3.3 DBR Suppression Chamber Heatup Analysis XV-169b 5.3.4 Conclusions XV-169g 6.0 New Fuel Bundle Loading Errox Analysis XV-169k 6.1 Identification of Causes XV-1 69k 6.2 Accident Analysis XV-1 69k 6.3 Safety Requirements XV-171 7.0 Meteorological Models Used in Accident Analyses XV-171 7~1 Ground Releases XV-171 7.2 Stack Releases XV-171 7.3 Variability XV-173 7.4 Exfiltration XV-175 7.5 Ground Deposition XV-197 7.6 Thyroid Dose XV-198 7.7 Whole Body Dose 'V-199 SECTION XVZ SPECIAL TOPICAL REPORTS XVZ-1 A. REACTOR VESSEL XVZ-1 1.0 Applicability of Formal Codes and Pertinent Certifications XVI-1 2.0 Design Analysis XVZ-3 2.1 Code Approval Analysis XVI-3 2.2 Steady-State Analysis XVI-3 2.2.1 Basis for Determining Stresses XVZ-3 2.3 Pipe Reaction Calculations XVI-10 2.4 Earthquake Loading Criteria and Analysis XVI-10 2.5 Reactor Vessel Support Stress Design Criteria and Analysis XVZ-14 2.6 Strain Safety Margin for Reactor Vessels XVI-24 2.6.1 Introduction XVZ-24 2.6.2 Strain Margin XVZ-26
| |
| : 2. 6.3 Failure Probability XVI-27
| |
| : 2. 6.4 Results of Probability Analysis XVI-31 2.6.5 Conclusions XVI-31 2.7 Components Required for Safe Reactor Shutdown XVZ-36 2.7.1 Design Basis Load Combinations XVI-36 2.7.2 Expected Stress and Deformation XVI-36 2.7.3 Stresses and Deformations at Which the Component is Unable to Function and Margin of Safety XVI-41 3.0 Inspection and Test Report Summary XVI-47 3.1 Materials XVZ-47 3.2 Fabrication and Inspection XVI-47 UFSAR Revision 12 xxiv June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title P acae 4.0 Surveillance Provisions XVI-51 4.1 Coupon Surveillance Program XVI-51 4.2 Periodic Inspection XVI-51 B. PRESSURE SUPPRESSION CONTAINMENT XVI-52 1.0 Applicability of Formal Codes and Pertinent Certifications XVI-52 2.0 Design Analysis XVI-53 2.1 Code Approval Calculations Under Rated Conditions XVI-53 2.2 Ultimate Capability Under Accident Conditions XVI-53 2.3 Capability to Withstand Internal Missiles and Jet Forces XVI-53 2.4 Flooding Capabilities of the Containment XVZ-57 2.5 Drywell Air Gap XVZ-61 2 '.1 2.6 Tests and Inspections Biological Shield Wall XVI-63 XVZ-65 2.7 Compatibility of Dynamic Deformations Occurring in the Drywell, Torus, and Connecting Vent Pipes XVI-69 2.8 Containment Penetrations XVI-74 2 '.1 2.8.2 Classification of Penetrations Design Bases XVI-74 XVZ-74 2.8.3 Method of Stress Analysis XVZ-76 2.8.4 Leak Test Capability XVI-76 2.8.5 Fatigue Design XVZ-77 2.8.6 Material Specification XVI-77 2.8.7 Applicable Codes XVI-77 2.8.8 Jet and Reaction Loads XVI-78 2.9 Drywell Shear Resistance Capability and Support Skirt Junction Stresses XVI-79 3.0 Inspection and Test Report Summary XVI-83 3.1 Fabrication and Inspection XVZ-83 3.2 Tests Conducted XVZ-83 3.3 Discussion of Results XVI-85 3.3.1 Results XVI-85 3.3.2 Effect of Various Transients XVI-87 C. ENGINEERED SAFEGUARDS XVZ-92 1.0 Seismic Analysis and Stress Report XVI-92 1.1 Introduction XVI-92 1.2 Mathematical Model XVZ-93 1.3 Method of Analysis XVI-94 1.3.1 Flexibility or.Influence Coefficient Matrix 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 Normal Mode Frequencies and Mode Shapes XVZ-96 1.3.3 The Seismic Spectrum Values XVI-96 1.3.4 Dynamic Modal Loads XVI-97 1.3.5 Modal Response Quantities XVZ-98 1.3.6 The Combined Response Quantities XVI-98 1.3.7 Basic Criteria for Analysis XVZ-99 1.4 Discussion of Results XVI-99 2.0 Containment Spray System XVI-104 2.1 Design Adequacy at Rated Conditions XVZ-104 2.1.1 General XVI-104 2.1.2 Condensation and Heat Removal Mechanisms XVI-104 2.1.3 Mechanical Design XVI-113 2.1.4 Loss-of-Coolant Accident XVI-113 2.2 Summary of Test Results XVZ-116 2.2.1 Spray Tests Conducted XVI-116 D. DESIGN OF STRUCTURES, COMPONENTS/
| |
| EQUIPMENT AND SY'STEMS XVI-123 1.0 Classification and Seismic Criteria XVI-123 1.1 Design Techniques XVI-126 1.1.1 Structures XVZ-126 1.1.2 Systems and Components XVI-146 1.2 Pipe Supports XVZ-153 1.3 Seismic Exposure Assumptions XVI-154 2.0 Plant Design for Protection Against Postulated Piping Failures in High Energy Lines XVI-155 2.1 Inside Primary Containment XVZ-155 I F 1-1 Containment Integrity Analysis XVI-155a 2.1.2 Systems Affected by Line Break XVZ-158 2.1.3 Engineered Safeguards Protection XVI-163 2.2 Outside Primary Containment XVI-165 3.0 Building Separation Analysis XVZ-168 4.0 Tornado Protection XVZ-168 E. EXHIBITS XVI-189 CONTAINMENT DESIGN REVIEW XVI-227 G. USAGE OF CODES/STANDARDS FOR STRUCTURAL STEEL AND CONCRETE XVZ-238 SECTION XVII ORIGINAL ENVIRONMENTAL STUDIES XVII-1 A. METEOROLOGX XVII-1 1.0 General XVII-1 2.0 Synoptic Meteorological Factors XVII-2 UFSAR Revision 12 xxvi June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae 3.0 Micrometeoro logy XVII-2 3.1 Wind Patterns XVII-2 3.1.1 200-Foot Wind Roses XVII-2 3.1 ~ 2 Estimates of Winds at the 350-Foot Level XVII-2 3.1.3 Comparison Between Tower and Satellite Winds XVII-16 3.2 Lapse Rate Distributions XVII-19 3.3 Turbulence Classes XVIZ-19 3.4 Dispersion Parameters XVII-19 3.4.1 Changes in Dispersion Parameters XVII-39 4.0 Applications to Release Problems XVIZ-45 4.1 Concentrations from a Ground-Level Source XVZZ-46 4.2 Concentrations from an Elevated Source XVZZ-53 4.3 Radial Concentrations XVZZ-55 4.3.1 Monthly and Annual Sector Concentrations XVZZ-55 4.4 Least Favorable Concentrations Over an Extended Period XVII-83 4.4.1 Ground-Level Release XVZZ-83 4.4.2 Elevated Release XVIZ-86 4.5 Mean Annual Sector Deposition XVI1-87 4.6 Dose Rates from a Plume of Gamma Emitters XVII-90 4.6.1 RADOS Program XVIZ-90 4.6;2 Centerline Dose Rates XVII-91
| |
| : 4. 6.3 Sector Dose Rates XVZZ-100 4.7 Concentrations from a Major Steam Line Break XVII-103 5.0 Conclusions XVZZ-106 B. L IMNOLOGY XVZZ-107 1.0 Introduction XVZZ-107 2.0 Summary Report of Cruises XVII-107 3.0 Dilution of Station Effluent in Selected Areas XVZZ-109 3.1 Dilution of Effluent at the Lake Surface Above the Discharge XVII-109 3.2 Dilution of Effluent at the Site Boundaries XVZZ-114 3.2.1 General XVIZ-114 3.2.2 Dilution of Effluent at the Eastern Site Boundary XVZI-116 3.2.3 Dilution of Effluent West of the Station Site XVII-122 3.3 Dilution of Effluent at the City of Oswego Intake XVZZ-123 UFSAR Revision 12 Xxvii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title P acae 3.3.1 Tilting of the Isothermal Planes and Subsequent Dilution XVZI-123 3.3.2 Dilution as a Function of Current Velocity XVII-124 3.3.3 Percent of Time Effluent Will Be Carried to the Oswego Area XVII-127 3.3.4 Mixing with Distance XVII-127 3.3.5 Oswego River Water as a Buffer to Prevent Effluent From Passing Over the Intake XVII-127 3.3.6 Summary of Annual Dilution Factors for the City of Oswego Intake XVII-127 3.4 Dilution of Effluent at the Nine Mile Point Intake XVZI-128 3.5 Summary of Dilution in the Nine Mile Point Area XVII-128 4.0 Preliminary Study of Lake Biota Off Nine Mile Point XVII-129 4.1 Biological Studies XVII-129 4.1.1 Plankton Study XVII-129 4.1.2 Bottom Study XVII-129 4.2 Summary of Biological Studies XVII-130 5.0 Conclusions XVZZ-130 C. EARTH SCIENCES XVZI-132 1.0 Introduction XVII-132 2.0 Additional Subsurface Studies XVII-132 3.0 Construction Experience XVII-138 3.1 Station Area XVII-138 3.2 Intake and Discharge Tunnels XVII-139 4.0 Correlation With Previous Studies XVII-140 4.1 General XVII-140 4.2 Geological Conditions XVII-140 4.3 Hydrological Conditions XVII-142 4.4 Seismological Conditions XVZI-142 4.5 Conclusion XVZZ-142 SECTION XVZZI HUMAN FACTORS ENGINEERING/SAFETY PARAMETER DISPLAY SYSTEM XVZZZ-1 A. DETAILED CONTROL ROOM DESIGN REVIEW XVIIZ-1 1.0 General XVIII-1 2.0 Planning Requirements for the DCRDR XVIII-2 3.0 DCRDR Review Process XVIII-2 3.1 Operator Survey XVII I-2 3.2 Historical Review XVIII-2 3.3 Task'nalysis XVIII-3 3.4 Control Room Inventory XVIZZ-3 3.5 Control Room Survey 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 Verification of Task Performance Capabilities XVIII-4 3.7 Validation o f Control Room Functions XVIII-4 3.8 Compilation of Discrepancy Findings XVIZZ-5 I I-5
| |
| ~
| |
| 4.0 Assessment and Implementation XVZ 4.1 Assessment XVII I-5 4.2 Implementation XVIZI-5 4.2.1 Integrated Cosmetic Package XVIZ1-6 4.2.2. Functional Fixes XVIII-7 5.0 Reporting XVIII-7 6.0 Continuing Human Factors Program XVIII-8 6.1 Fix Verifications XVIII-8 6.2 Multidisciplinary Review Team Assessments XVZ I I-8 6.3 Human Factors Manual for Future Design Change XVIII-8 6.4 Outstanding Human Factors Items XVIII-8 7.0 References XVIII-9 B. SAFETY PARAMETER DISPLAY SYSTEM XVZII-11 1.0 Introduction to the Safety Parameter Display System XVIII-11 2.0 System Description XVIII-11 3.0 Role of the SPDS XVIII-12 4.0 Human Factors Engineering Guidelines XVIII-13 5.0 Human Factors Engineering Principles Applied to the SPDS Design XVIII-13 5.1 NUREG-0737, Supplement 1, Section 4.1.a XVIZZ-13 5.1.1 Concise Display XVIII-13 5.1.2 Criteria Plant Variables XVIII-13 5.1.3 Rapid and Reliable Determination of Safety Status XVIIZ-14 5.1.4 Aid to Control Room Personnel XVIII-14 5.2 NUREG-0737, Supplement 1, Section 4.1.b XVIII-14 5.2.1 Convenient Location XVIII-14 5.2.2 Continuous Display XVZII-15 5.3 NUREG-0737, 'Supplement 1, Section 4.1.c XVIII-15 5.3.1 Procedures and Training XVII I-15 5.3.2 Isolation of SPDS from Safety-Related Systems XVIII-15 5.4 NUREG-0737, Supplement 1, Section 4.1.e XVIII-16 UFSAR Revision 12 xxix June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.)
| |
| Section Title Pacae 5.4.1 Incorporation of Accepted Human Factors Engineering Principles XVIII-16 5.4.2 Information Can be Readily Perceived and Comprehended XVIII-17 5.5 NUREG-0737, Supplement 1, Section
| |
| : 4. 1. f, Sufficient Information XVIII-17 6.0 Procedures XVIII-17 6.1 Operating Procedures XVIII-17 6.2 Surveillance Procedures XVIII-18 7.0 References XVIII-19 APPENDIX A Unused APPENDIX B NIAGARA MOHAWK POWER CORPORATION QUALITY ASSURANCE PROGRAM TOPICAL REPORT (NMPC-QATR-1) ~ 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 Title Pacae Piping, Instrument and Equipment Symbols I-17 II-1 Station Location II-2 ZI-2 Area Map IZ-3 II-3 Site Topography XZ-4 II-4 Population Distribution Within a Twelve Mile Radius of the Station ZI-7 ZI-5 Counties and Towns Within Twelve Miles of the Station IZ-8 XI-6 1980 Population Distribution Within a Fifty Mile Radius of the Station IZ-10 ZZI-1 Plot Plan III-3 III-2 Station Floor Plan Elevation 225-6 IZZ-7 IIX-3 Station Floor Plan Elevations 237 and 250 ZZI-8 IXI-4 Station Floor Plan Elevation 261 III-9 Xjj-5 Station Floor Plan Elevations 277 and 281 III-10 Ijj-6 Station Floor Plan - Elevations 281 and 291 XIZ-11 IXZ-7 Station Floor Plan - Elevations 298 and 300 ZII-12 ZZI-8 Station Floor Plan --Elevations 317-6 and 318 XIZ-13 IIX-9 Station Floor Plan Elevations 320, 333-8, 340 and 369 XII-14 XXX-10 Section Between Column Rows 7 'and 8 Ijj-15 IZI-11 Section Between Column Rows 12 and 14 XII-16 III-12 Turbine Building Ventilation System Xlj-18 IIX-13 Laboratory and Radiation Protection Facility Ventilation System XIX-20 IIZ-14 Control Room Ventilation System IZI-26 IIX-15 Waste Disposal Building Ventilation System ZII-33 IIZ-16 Waste Disposal Building Extension Ventilation System XXX-35 IIX-17 Offgas Building Ventilation System Ijj-40 IZI-18 Technical Support Center Ventilation System IZI-46 III-19 Circulating Water Channels Under Screen and Pump House Normal Operation IXI-55 UFSAR Revision XXXi June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure
| |
| ~umber xMt1R ~ae III-20 Circulating Water Channels Under Screen and Pump House Special Operations III-56 III-21 Intake and Discharge Tunnels Plan III-58 and Profile III-22 Stack Plan and Elevation III-62 III-23 Stack Failure Critical
| |
| /
| |
| Directions III-65 IV-1 Limiting Power/Flow Line IV-10 IV-2 Figure Deleted IV-26 IV-3 Figure Deleted IV-28 IV-3a GE11 Fuel Assembly Design IV-28a IV-4 Typical Control Rod Isometric IV-33 I
| |
| IV-5 Figure Deleted IV-34 IV-5a Control Rod Positioning Within Core Cell GE8X8EB IV-34a IV-5b Control Rod Positioning Within Core Cell With Typical GE11 Fuel Assembly IV-34b IV-6 Control Rod Drive and Hydraulic System IV-37 IV-7 Control Rod Drive Assembly IV-38 IV-8 Typical Control Rod to Drive Coupling Isometric IV-39 IV-9 Reactor Vessel Isometric IV-48 V-1 Reactor Emergency Coolant System V-11 V-2 Reactor Vessel Nozzle Location V-13 V-3 . Reactor Vessel Support V-15 V-4 Figure Deleted V-18 V-5 Pressure Vessel Embrittlement Trend V-23 V-6 Figure Deleted V-24 V-7 Figure Deleted V-25 V-8 Emergency Condenser Supply Isolation Valves (Typical of 2) V-32 VI-1 Drywell and Suppression Chamber VI-2 Electrical Penetrations High VI-14'I-17 Voltage VI-3 Electrical Penetrations Low Voltage VI-18 VI-4 Pipe Penetrations Hot VI-19 VI-5 Typical Penetration For Instrument Lines VI-20 VI-6 Reactor Building Dynamic Analysis Acceleration East-West Direction VI-28 UFSAR Revision 13 XXXii June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title Pacae VI-7 Reactor Building Dynamic Analysis Deflections East-West Direction vx-29 vz-8 Reactor Building Dynamic Analysis Elevation vs. Building Shear East-West Direction VI-30 VI-9 Reactor Building Dynamic Analysis Elevation vs. Building Moment East-West Direction VI-31 VX-10 Reactor Building Dynamic Analysis Acceleration North-South Direction VI-32 VZ-11 Reactor Building Dynamic Analysis-Deflections North-South Direction VI-33 VI-12 Reactor Building Dynamic Analysis Elevation vs. Building Shear North-South Direction VI-34 Vz-13 Reactor Building Dynamic Analysis' Elevation vs. Building Moment North-South Direction VX-35 VZ-14 Reactor Support Dynamic Analysis Elevation vs. Acceleration Vz-36 VI-15 Reactor Support Dynamic Analysis Elevation vs. Deflection VZ-37 VI-16 Reactor Support Dynamic Analysis Elevation vs. Shear vz-38 VZ-17 Reactor Support Dynamic Analysis Elevation vs. Moment VI-39 VI-18 Typical Door Seals VZ-41 VI-19 Details of Reactor Building Air Locks VX-42 vz-20 Instrument Line Isolation Valve Arrangement vz-53 VZ-21 Typical Flow Check Valve VI-54 VX-22 Isolation Valve System vz-58 VI-23 Drywell Cooling System VI-61 VI-24 Reactor Building Ventilation System vz-63 VII-1 Core Spray System VII-3 VIX-2 Core Spray Sparger Flow, Per Sparger, for One Core Spray Pump and One Topping Pump VXI-5 VXI-3 Containment Spray System VIZ-12 VII-4 Figure Deleted VII-13 VII-4a Figure Deleted VZZ-13a VIZ-5 Figure Deleted Vzz-14 vzz-6 Liquid Poison System VII-22 VII-7 Minimum Allowable Solution Temperature VII-24 vzx-8 Figure Deleted VII-25 UFSAR Revision 12 ZXXiii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title Pacae VII-9 Typical Control Rod Velocity Limiter VII-32 VZI-10 Control Rod Housing Support VI1-39 VZI-11 Hydrogen Flammability Limits VII-45 VZI-12 Combustible Gas Control System VII-47 VII-13 H,-O, Sampling System VII-50 VXI-14 H,-O, Concentrations in Containment Following LOCA VZZ-51 VZI-15 N, Added by CAD Operation Following LOCA VII-52 VII-16 Containment Pressure with CAD Operation Zero Containment Leakage VII-54 VII-17 Feedwater Delivery Capability (Shaft Driven Pump) VII-63 VIIZ-1 Protective System Function VIII-2 VIII-2 Reactor Protection System Elementary Diagram VIII-8 VIII-3 Protective System Typical Sensor Arrangement VIII-16 VXII-4 Recirculation Flow and Turbine Control VXIZ-19 VIII-5 Neutron Monitoring Instrument Ranges VII1-24 VIII-6 Source Range Monitor (SRM) VIII-26 VIII-7 SRM Detector Location VIII-27 VXII-8 Intermediate Range Monitor (IRM) VII I-29 VIII-9 IRM Core Location VIII-30 VIIX-10 LPRM Location Within Core Lattice VIIX-33 VXII-11 LPRM and APRM Core Location VIII-34 VIII-12 Local Power Range Monitor (LPRM) and Average Power Range Monitors (APRM) VIXI-35 VXII-13 Typical VIXI-37 VIXI-14 APRM System Trip Logic f'r APRM Scram and Rod VIII-38 Block VIII-15 Traversing In-Core Probe VIXI-41 VIII-16 Rod Pattern During Startup VIII-45 VIIZ-17 Radial Power Distribution for Control Rod Pattern Shown in Figure VXII-16 VIII-46 VIXX-18 Distance from Worst Control Rod to Nearest Active IRM Monitor VIIZ-47 VIII-19 Measured Response Time of Intermediate Range Safety Instrumentation VIXI-49 UFSAR Revision 12 XXX'.V June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title P acae VIXI-20 Envelope of Maximum APRM Deviation by Flow Control Reduction in Power VIIX-52 VIIZ-21 Envelope of Maximum APRM Deviation for APRM Tracking With On Units Control Rod Withdrawal VXZZ-53 VIIZ-22 Main Steam Line Radiation Monitor VIII-63 VIII-23 Reactor Building Ventilation Radiation Monitor VIII-65 VIII-24 Offgas System Radiation Monitor VXII-66 VIXI-25 Emergency Condenser Vent Radiation Monitor VXIZ-69 VIIZ-26 Stack Effluent and Liquid Effluent Radiation Monitors VXII-70 VIII-27 Containment Spray Heat Exchanger Raw Water Effluent Radiation Monitor VXIX-73 VIII-28 Containment Atmospheric Monitoring System VIXI-75 VIII-29 Rod Worth Minimizer VIIZ-78 IX-1 A.C. Station Power Distribution Ix-9 IX-2 Control and Instrument Power XX-14 IX-3 Trays Below Elevation 261 IX-21 IX-4 Trays Below Elevation 277 Ix-22 IX-5 Trays Below Elevation 300 Ix-23 IX-6 Diesel Generator Loading Following Loss-of-Coolant Accident Zx-27 Ix-7 Diesel Generator Loading for Orderly Shutdown Ix-28 X-1 Reactor Shutdown Cooling System X-2 X-2 Reactor Cleanup System x-5
| |
| -X-3 Control Rod Drive Hydraulic System X-9 X-4 Reactor Building Closed Loop Cooling System X-23 Turbine Building Closed Loop Cooling System x-28 X-6 Service Water System X-32 X-7 Decay Heat Generation vs. Days After Reactor Shutdown X-40 X-8 Spent Fuel Storage Pool Filtering and Cooling System X-41 Breathing, Instrument, and Service Air X-46 X-10 Reactor Refueling System Pictorial X-51 X-11 Cask Drop Protection System x-56 UFSAR Revision 12 XXXV June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number ~T't e ~ae XI-1 Steam Flow and Reheater Ventilation System XI-2 XI-2 Extraction Steam Flow XI-3 XI-3 Main Condenser Air Removal and Offgas System XI-4 XI-4 Circulating Water System XI-5 XI-5 Condensate Flow XI-6 XI-6 Condensate Transfer System XI-7 XI-7 Feedwater Flow System XI-8 XII-1 Radioactive Waste Disposal System XII-9 XIII-1 NMPC Upper Management Nuclear I Organization XIII-2 Nine Mile Point Nuclear Site Organization XIII-3 Nuclear Engineering Organization XIII-4 Nuclear Safety Assessment and Support Organization XIII-5 Safety Organization XV-1 Station Transient Diagram XV-3 I XV 2 Figure Deleted XV-8 XV-2a Figure Deleted XV-8a XV-2b Figure Deleted XV-8b XV-3 Plant Response to Loss of 100 F Feedwater Heating XV-10 XV-4 Figure Deleted XV-12 XV-4a Plant Response to Feedwater Controller Failure (GE11 at EOC) XV-12a XV-4b Plant Response to Feedwater Controller Failure (GE11 at EOC Extended Load Line Limit) XV-12b XV-5 Figure Deleted XV-16 XV-6 Figure Deleted XV-17 XV-6a Figure Deleted XV-17a XV-7 Figure Deleted XV-19 XV-8 Startup of Cold Recirculation Loop Partial Power XV-23 XV-9 Recirculation Pump Trips (1 Pump) XV-25 XV-10 Recirculation Pump Trips (5 Pumps) XV-26 XV-11 Recirculation Pump Stall XV-29 UFSAR Revision 13 xxxvi June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR 1
| |
| LIST OF FIGURES (Cont,'d.)
| |
| Figure Number Title ~ae XV-12 Flow Controller Malfunction (Increased Flow) XV-3 1 XV-13 Flow Controller Malfunction Decreasing Flow XV-33 XV-14 Inadvertent Actuation of One Solenoid Relief Valve XV-35 XV-15 Figure Deleted XV-39 XV-16 Figure Deleted XV-40 XV-17 Feedwater Controller Malfunction-Zero Flow XV-42 XV-18 Turbine Trip With Partial Bypass Intermediate Power XV-45 XV-19 Turbine Trip With Partial Bypass XV-47 XV-20 Inadvertent Actuation of One Bypass Valve XV-49 XV-21 One Feedwater Pump Trip and Restart XV-51 XV-22 Loss of Electrical Load XV-53 XV-23 Loss of Auxiliary Power XV-55 XV-24 Pressure Regulator Malfunction XV-58 XV-25 Main Steam Line Break Coolant Loss XV-74 XV-26 Figure Deleted XV-85 XV-27 Figure Deleted XV-86 XV-28 Figure Deleted XV-89 XV-29 Figure Deleted XV-90 XV-3 0 Figure Deleted XV-91 XV-3 1 Figure Deleted XV-92 XV-32 Figure Deleted XV-93 XV-33 Figure Deleted XV-94 XV-33a Figure Deleted XV-94a XV-34 Figure Deleted XV-103 XV-3 5 Figure Deleted XV-104 XV-3 6 Figure Deleted XV-105 XV-37 Figure Deleted XV-106 XV-38 Figure Deleted XV-107 XV-39 Figure Deleted XV-108 XV-4 0 Figure Deleted XV-109 XV-4 1 Figure Deleted XV-110 XV-42 Figure Deleted XV-111 XV-43 Figure Deleted XV-112 XV-44 Figure Deleted XV-113 XV-45 Figure Deleted XV-114 XV-46 Figure Deleted XV-115 XV-47 Figure Deleted XV-116 XV-48 Figure Deleted XV-117 XV-49 Figure Deleted XV-118 XV-50 Figure Deleted XV-119 UFSAR Revision 13 xxxvii June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title P acae XV-51 Figure Deleted XV-120 XV-52 Figure Deleted XV-121 XV-53 Figure Deleted XV-122 XV-54 Figure Deleted XV-123 XV-55 Figure Deleted XV-12 6 xv-56 Figure Deleted XV-127 XV-56A Figure Deleted XV-137i XV-56B Figure Deleted XV-137j
| |
| ! XV-56C Figure Deleted XV-137k XV-56D Loss-of-Coolant Accident With Core Spray Cladding Temperature XV-159e XV-56E Loss-of-Coolant Accident Drywell Pressure XV-159g,,
| |
| XV-56F Loss-of-Coolant Accident Suppression Chamber Pressure XV-159h XV-56G Loss-of-Coolant Accident Containment Temperature With Core Spray XV-15 9L xv-56H Loss-of-Coolant Accident Clad Perforation With Core Spray XV-159n XV-57 Containment Design Basis Clad Temperature Response Without O.
| |
| Core Spray XV-1 61 XV-58 Containment Design Basis Metal-Water Reaction XV-163 XV-59 Containment Design Basis Clad Perforation Without Core Spray XV-165 XV-60 Containment Design Basis Containment Temperature Without Core Spray XV-1 66 xv-60 a DBR Analysis Suppression Pool and Wetwell Airspace Temperature Response Containment Spray Design Basis Assumption XV-1 69L XV-60b DBR Analysis Suppression Pool and Wetwell Airspace Temperature Response EOP Operation Assumptions xv-169m XV-61 Reactor Building Model XV-62 Exfiltration vs. Wind Speed XV-178'v-184 Northerly Wind xv-63 Reactor Building Differential Pressure xv-185 XV-64 Exfiltration vs. Wind Speed Southerly Wind .XV-186 XV-65 Reactor Building Isometric XV-188 XV-66 Reactor Building Corner Sections XV-189 XV-67 React. or Building Roof Sections XV-190 UFSAR Revision 12 XXXViii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title Pacae XV-68 Reactor Building Panel to Concrete Sections XV-191 XV-69 Reactor Building Expansion Joint Sections XV-192 XV-70 Reactor Building Exfiltration Northerly Wind XV-193 XV-71 Reactor Building Exfiltration Southerly Wind XV-194 XV-72 Reactor Building Differential Pressure XV-196 XVI-1 Seismic Analysis of Reactor Vessel Geometric and Lumped Mass Representation XVI-12 XVZ-2 Reactor Support Dynamic Analysis Elevation vs. Moment XVI-13 XVI-3 Reactor Support Dynamic Analysis Elevation vs. Shear XVI-15 XVI-4 Reactor Support Dynamic Analysis Elevation vs. Deflection XVI-16 Reactor Support Dynamic Analysis O. XVI-6 XVI-7 Elevation vs. Acceleration Figure Deleted Figure Deleted XVI-17 XV1-1 8 XVZ-19 XVI-8 Figure Deleted XVZ-20 XVZ-9 Reactor Vessel Support Structure Stress Summary XVI-21 XVI-10 Thermal Analysis XVI-22 XVI-11 Failure Probability Density Function XVI-28 XVI-12 Addition Strains Past 4% Required to Exceed Defined Safety Margin XVI-32 XVZ-13 Loss of Coolant Accident Containment Pressure No Core or Containment .Sprays XVI-54 XVI-14 Figure Deleted XVZ-60 XVI-15 Drywell to Concrete Air Gap XVI-62 XVZ-16 Typical Penetrations XVZ-64 XVI-17 Biological Shield Wall Construction Details XVI-66 XVI-18 Vent Pipe and Suppression Chamber XVI-70 XVI-19 Primary Containment Support and Anchorage . XVI-71 XVI-20 Seal Details Drywell Shell Steel and Adjacent Concrete XVI-73 XVI-21 Drywell Sliding Acceleration, Shear, and Moment XVI-75 UFSAR Revision 12 Xxxix June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title P RcRB XVZ-22 Shear Resistance Capability Inside Drywell XVI-80 XVI-23 Shear Resistance Capability Outside Drywell XVI-81 XVZ-24 Drywell Support Skirt Junction Stresses xvI-82 XVI-25 Point Location for Containment Spray System Piping Heat Exchanger to Drywell XVZ-10 0 xvI-26 Comparison of Static and Dynamic Stresses (PSI) Seismic Conditions Containment Spray System Heat Exchanger to Drywell XVI-103 XVZ-27 Conduction in a Droplet XVI-10 9 XVZ-28 Loss of Coolant Accident Containment Pressure XVI-115 XVI-29 Loss of Coolant Accident Containment Pressure XVZ-117 XVZ-30 Nozzle Spray Test Pressure Drop of 80 psig XVI-118 XVI-31 Nozzle Spray Test Pressure Drop of 80 psig XVI-119 XVI-32 Nozzle Spray Test Pressure Drop of 30 psig XVZ-120 XVZ-33 Nozzle Spray Test Pressure Drop of 30 psig XVI-121 XVZ-34 Seismic Analysis Reactor Building XVI-134 XVZ-35 Dynamic Analysis Drywell XVI-135 xvz-36 Reactor Support Structure Seismic XVI-136 XVI-37 Seismic Analysis Waste Building XVI-137 xvI-38 Seismic Analysis Screenhouse XVI-138 XVZ-39 Seismic Analysis Turbine Building (North of Row C) XVI-139 XVI-40 Seismic Analysis Turbine Building (South of Row C) XVI-140 XVZ-41 Seismic Analysis Concrete Ventilation Stack XVI-141 XVI-42 Reactor Building Mathematical Model (North-South) XVI-144 XVZ-43 Reactor Support Structure Seismic XVI-147 XVI-44 Reactor Support Structure Reactor Building XVZ-148 xvI-45 Reactor Support Structure Reactor Building and Seismic XVI-149 XVI-46 Plan of Building XVZ-169 XVI-47 Wall Section 1 XVI-170 XVI-48 Wall Section 1 Detail "A" XVZ-171 XVZ-49 Wall Section 1 Detail "B" 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 Wall Section 1 Detail XVI-173 XVI-51 Wall Section 1 Detail "D" XVI-174 XVI-52 Wall Section 1 Detail "E" XVI-175 XVI-53 Wall Section 2 XVI-176 XVI-54 Wall Section 3 XVI-177 XVI-55 Wall Section 3A Details XVI-178 XVI-56 Wall Section 4 XVZ-179 XVI-57 Wall Section 4 Detail 1 XVI-180 XVZ-58 Wall Section 4 Detail 2 XVI-181 XVZ-59 Wall Section 5 XVI-182 XVI-60 Wall Section 6 XVI-183 XVI 61 Wall Section 7 XVI-184 XVZI-1 Average Wind Roses for January
| |
| '3-'4 for XVII-3 XVII-2 Average Wind Roses February
| |
| '3-'4 for XVII-4 XVII-3 Average Wind Roses March
| |
| '3-'4 for XVZI-5 XVII-4 Average Wind Roses April
| |
| '3-'4 for '3-'4 XVII-6 XVII-5 Average XVII-7 June '3-'4 Wind Roses May XVII-6 Average Wind Roses for XVZ I-8 XVII-7 Average Wind Roses for July '63-'64 XVII-9 XVII-8 Average Wind Roses for August
| |
| '3-'4 for XVII-10 XVZI-9 Average Wind Roses September XVZI-10
| |
| '3-'4 for October XVII-11 Average Wind Roses XVIZ-11
| |
| '3-'4 for XVII-12 Average Wind Roses November
| |
| '3-'4 XVIZ-13 XVII-12 Average Wind Roses for December XVII-13
| |
| '3-'4 for '63-'64 XVII-14 XVII-15 Average Wind Roses XVII-14 Diurnal Lapse Rate January Average
| |
| '3-'4 February '3-'4 XVII-20 XVII-15 Average Diurnal Lapse Rate March
| |
| '3-'4 April '3-'4 XVIZ-21 XVII-16 Diurnal Lapse Rate May June '3-'4 Average
| |
| '3-'4 Diurnal Lapse Rate July XVIZ-22 August '3-'4 Average XVZZ-17
| |
| '3-'4 XVII-23 XVII-18 Average Diurnal Lapse Rate Septemb er '3-'4, October '3-'4 XVIZ-24 Average Diurnal Lapse Rate November XVZZ-19
| |
| '3-'4 December '2-'3 XVII-25 UFSAR Revision xli June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title ~acae XVII-20 Lapse Rates by Wind Speed and Turbulence Classes for January XVII-21
| |
| '3-'4 XVZI-27 Lapse Rates by Wind Speed and Turbulence Classes for February XVII-22
| |
| '3-'4 XVZI-28 Lapse Rates by Wind Speed and Turbulence Classes for March XVIZ-23
| |
| '3-'4 XVZI-29 Lapse Rates by Wind Speed and Turbulence Classes for April XVII-24
| |
| '3-'4 XVII-30 Lapse Rates by Wind Speed and Turbulence Classes for May '3-'4 XVII-31 XVII-25 Lapse Rates by Wind Speed and Turbulence Classes for June '3-'4 XVII-32 XVII-26 Lapse Rates by Wind Speed and Turbulence Classes for July '3-'4 XVII-33 XVII-27 Lapse Rates by Wind Speed and Turbulence Classes for August
| |
| '3-'4 XVII-34 XVZZ-28 Lapse Rates by Wind Speed and Turbulence Classes for September XVIZ-29
| |
| '3-'4 XVZZ-35 Lapse Rates by Wind Speed and Turbulence Classes for October XVII-30
| |
| '3-'4 XVII-36 Lapse Rates by Wind Speed and Turbulence Classes for November XVII-31
| |
| '3-'4 XVZI-37 Lapse Rates by Wind Speed and Turbulence Classes for December
| |
| '3-'4 XVIZ-38 XVIZ-32 Sector Map XVZI-44 XVII-33 Centerline Concentrations Turbulence Class I XVII-47 XVII-34 Centerline Concentrations Turbulence Class II XVII-4 8 XVZI-35 Centerline Concentrations Turbulence Class III XVZI-49 XVII-36 Centerline Concentrations Turbulence Class IV XVII-50 XVIZ-37 Centerline Concentrations Turbulence Class IZ Becoming Class IV at 2 km and Class II at 23 km XVIZ-51 XVII-38 Centerline Concentrations Turbulence Class IV Becoming Class II at 16 km 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 Centerline Concentrations Turbulence Class IV Becoming Class II at 2 km XVZI-54 XVII-4 0 Radial Concentrations Turbulence Class I XVZI-57 XVII-41 Radial Concentrations Turbulence Class II XVII-58 XVII-42 Radial Concentrations Turbulence XVII-43 Class Ill Radial Concentrations Turbulence XVII-59 Class IV XVII-60 XVII-44 Radial Concentrations Turbulence Class II Becoming Class ZV at 2 km and Class II at 23 km XVZI-61 XVII-45 Radial Concentrations Turbulence Class ZV Becoming Class Iz at 16 km XVII-62 XVZI-46 Radial Concentrations Turbulence Class ZV Becoming Class ZZ at 2 km xvzZ-63 XVII-47 Centerline Gamma Dose Rates Turbulence Class I XVIZ-93 XVII-48 Centerline Gamma Dose Rates Turbulence Class ZZ XVIZ-94 XVII-49 Centerline Gamma Dose Rates Turbulence Class IZI XVIZ-95 XVII-50 Centerline Gamma Dose Rates Turbulence Class IV XVZI-96 XVII-51 Centerline Gamma Dose Rates Turbulence Class Iz Becoming Class ZV at 2 km and Class Iz at 23 km XVII-97 XVII-52 Centerline Gamma Dose Rates Turbulence Class IV Becoming Class II at 16 km XVZZ-98 XVII-53 Centerline Gamma Dose Rates Turbulence Class IV Becoming Class II at 2 km XVII-99 XVIZ-54 Assumed Concentration and Dose Rate Distributions Close to the Elevated Source Xvzz-101 XVII-55 Gamma Dose Rate as a Function of Gy at 1 km From the Source XVII-102 XVZI-56 Southeastern Lake Ontario XVII-108 XVII-57 Dilution of Rising Plume XVII-112 XVZI-58 Estimated Lake Currents at Cooling Water Discharge XVIZ-113 XVII-59 Temperature Profiles in an Eastward Current at the Oswego City Water Intake XVII-125 UFSAR Revision 12 xliii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.)
| |
| Figure Number Title Pacae XVZI-60 f Subsur ace Section Plot Plan XVII-133 XVII-61 Log of Boring (Boring CB-1) XVZI-134 XVZI-62 Log of Boring (Boring CB-2) XVII-135 XVIZ-63 Log of Boring (Boring CB-3) XVII-136 XVII-64 Log of Boring (Boring CB-4) XVZI-137 XVII-65 Attenuation Curves XVII-141 UFSAR Revision 12 xliv June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF TABLES Table Number Title Pacae XI-1 1980 Population and Population Density for Towns and Cities Within 12 Miles of Nine Mile. Point Unit 1 Ix-6 ZI-2 Cities Within a 50-mile Radius of the Station With Populations over 10,000 Ix-9 ZI-3 Regional Agricultural Use ZX-11 IZ-4 Regional Agricultural Statistics -
| |
| Cattle and Milk Production II-12 II-5 Industrial Firms Within 8 km (5 mi) of Unit 1 IX-13 xx-6 Public Utilities in Oswego County II-17 IX-7 Public Water Supply Data for Locations Within an Approximate 30-Mile Radius IZ-18 II-8 Recreational Areas in the Region II-19" V-1 Reactor Coolant System Data V-3 V-2 Operating Cycles and Transient Analysis Results V-7 V-3 Fatigue Resistance Analysis V-9 V-4 Codes for Systems Connected to the Reactor Coolant System V-9 V-5 Time to Automatic Blowdown V-12a VI-1 Drywell Penetrations VX-44 VI-2 Suppression Chamber Penetrations Vx-46 VI-3a Reactor Coolant System Isolation Valves Vx-47 VZ-3b Primary Containment Isolation Valves Lines Entering Free Space of the Containment VZ-49 VI-3c Table Deleted VZ-51 VI-4 Seismic Design Criteria for Isolation Valves VI-56 VZ-5 Xnitial Tests Prior to Station Operation vx-69 VII-1 Performance Tests VZI-35 VIXI-1 Association Between Primary Safety Functions and Emergency Operating Procedures VXXI-83 VIXI-2 List of EOP Key Parameters VIII-85 VIII-3 Type and Instrument Category for NMP1 RG 1.97 'Variables VIIZ-87 UFSAR Revision 12 xlv June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)
| |
| Table Number ~ice Pacae IX-1 Magnitude and Duty Cycle of Major Station Battery Loads IX-32 XII-1 Flows and Activities of Major Sources of Gaseous Activity XII-2 XII-2 Quantities and Activities of Liquid Radioactive Wastes XII-4 XII-3 Annual Solid Waste Accumulation and Activity XII-6 XII-4 Liquid Waste Disposal System-Major Components XII-14 XII-5 Solid Waste Disposal System Major Components XII-18 XII-6 Occupancy Times XII-21 XII-7 Gamma Energy Groups XII-22 XII-8 Area Radiation Monitor Detector Locations XII-26 XIII-1 ANSI Standard Cross-Reference Unit 1 XV-1 Transients Considered XV-2 Trip Points for Protective Functions XV-6 Table Deleted XV-14 XV-4 Instrument Air Failure XV-60 XV-5 Blowdown Rates XV-73 XV-6 'odine Concentrations XV-76 XV-7 Fractional Concentrations in Clouds XV-77 XV-8 Main Steam Line Break Accident Doses XV-78 XV-9 Significant Input Parameters to the Loss-of-Coolant Accident Analysis XV-82 XV-9A Core Spray System Flow Performance Assumed in LOCA Analysis XV-82b XV-10 ECCS Single Valve Failure Analysis XV-97 XV-11 Single Failures Considered in LOCA Analysis XV-98 XV-12 Table Deleted XV-101 XV-13 Table Deleted XV-102 XV-14 Table Deleted XV-124 XV-15 Table Deleted XV-125 XV-16 Table Deleted XV-13 1 XV-17 Table Deleted XV-134 XV-18 Table Deleted XV-134 XV-19 Table Deleted XV-13 6 UFSAR Revision 13 xlvi June 1995
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)
| |
| Table Number Title Pacae XV-20 Table Deleted XV-13 6 XV-21 Table Deleted XV-13 6 XV-21A Analysis Assumptions For Nine Mile Point-1 Calculations XV-137e XV-21B Table Deleted XV-137f XV-21C Table Deleted XV-137g XV-21D Table Deleted XV-137h XV-21E Table Deleted XV-137h XV-22 Reactor Building Airborne Fission Product Inventory XV-144 XV-23 Stack Discharge Rates XV-145 XV-24 Fuel Handling Accident Doses (REM) XV-14 6 XV-25 Fission Product Release Assumptions XV-147 XV-2 6 Atmospheric Dispersion and Dose Conversion Factors XV-14 8 XV-27 Effect on Dose of Factors Used in the Calculations XV-148 XV-28 Noble Gas Release XV-158 XV-29 Halogen Release XV-15 9 XV-29a Wetting of Fuel Cladding by Core Spray XV-15 9 XV-29b Airborne Drywell Fission Product Inventory f'V-159s XV-29c Reactor Building Airborne Fission Product Inventory XV-159u XV-29d Stack Discharge Rates XV-159x XV-30 Airborne Drywell Fission Product Inventory XV-167 Reactor Building Airborne Fission Product Inventory XV-168 XV-32 Stack Discharge Rates XV-168 XV-32a Significant Input Parameters to the DBR Containment Suppression Chamber Heatup Analysis XV-169i XV-33 Downwind Ground Concentrations XV-172 XV-34 Maximum Ground Concentrations XV-173 XV-35 Diversity Factors for "Least Favorable" Ground Concentrations from Stack Release, for Ground Release, for Maximum Ground Concentrations XV-17 6 XV-36 Reactor Building Leakage Paths XV-195 XVI-1 Code Calculation Summary XVI-4 XVI-2 Steady State (100% Full Power Normal Operation) Pertinent Stresses or Stress Intensities XVI-6 UFSAR Revision 12 xlvii June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)
| |
| Table Number Title P acae XVI-3 List of Reactions for Reactor Vessel Nozzles XVI-11 XVI-4 Effect of Value of Initial Failure Probability XVI-31 XVI-5 Single Transient Event for Reactor Pressure Vessel. XVI-33 XVI-6 Postulated Events XVI-34 XVI-7 Maximum Strains from Postulated Events xvz-35 XVZ-8 Core Structure Analysis Recirculation Line Break XVI-37 XVZ-9 Core Structure Analysis Steam Line Break XVI-40 xvz-10 Drywell Jet and Missile Hazard Analysis Data xvz-55 XVl-11 Drywell Jet and Missile Hazard Analysis Results xvz-56 XVI-12 Stress Due to Drywell Flooding XVI-59 XVZ-13 Allowable Weld Shear Stress xvz-65 XVI-14 Leak Rate Test Results xvz-86 XVI-15 Overpressure Test Plate Stresses xvz-91 XVI-16 Stress Summary xvz-101 XVI-17 Heat Transfer Coefficients as a Function of Drop Diameter XVI-110 XVI-18 Heat Transfer Coefficients as a Function of Pressure XVI-111 XVZ-19 Relationship Between Particle Size and Type of Spray Pattern XVI-122 XVZ-20 Allowable Stresses for Floor Slabs, Beams, Columns, Walls, Foundations, etc. XVI-127 XVI-21 Allowable Stresses for Structural Steel XVI-128 XVI-22 Allowable Stresses Reactor Vessel Concrete Pedestal xvz-129 XVI-23 Drywell Analyzed Design Load Combinations xvz-130 XVI-24 Suppression Chamber Analyzed Design Load Combinations XVI-131 XVI-25 ACI Code 505 Allowable Stresses and Actual Stresses for Concrete Ventilation Stack XVZ-132 xvz-26 Allowable Stresses for Concrete Slabs, Walls, Beams, Structural Steel, and Concrete Block Walls XVI-133 XVI-27 System Load Combinations XVZ-151 XVI-28 High Energy Systems Inside Containment xvz-155a UFSAR Revision 12 xlviii tune 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)
| |
| Table Number Title Pacae XVI-29 High Energy Systems Outside Containment XVI-166 XVZ-30 Systems Which May Be Affected by Pipe Whip XVI-167 XVZ-31 Capability to Resist Wind Pressure and Wind Velocity XVI-185 XVZI-1 Dispersion and Associated Meteorological Parameters XVII-17 XVZI-2 Relation of Satellite and Nine Mile Point Winds XVZI-18 XVZI-3 Frequency of Occurrence of Lapse Rates 1963 and 1964 XVII-26 XVZI-4 Relation Between Wind Direction Range and Turbulence Classes XVII-40 XVII-5 Stack Characteristics XVZI-56 XVII-6 Distribution of Turbulence Class es By Sectors XVII-64 XVII-7 Sector Concentrations 1963-64 Sector A Elev. 350 XVII-66 XVII-8 Sector Concentrations 1963-64 Sector B Elev. 350 XVII-67 XVII-9 Sector Concentrations 1963-64 Sector C Elev. 350 XVII-68 XVIZ-10 Sector Concentrations 1963-64 Sector D, Elev. 350 XVII-69 XVZI-11 Sector Concentrations 1963-64 Sector D, Elev. 350 XVII-7 0 XVII-12 Sector Concentrations 1963-64 Sector E Elev. 350 XVZI-71 XVII-13 Sector Concentrations 1963-64 Sector F Elev. 350 XVII-72 XVZI-14 Sector Concentrations 1963-64 Sector G Elev. 350 XVII-73 XVII-15 Sector Concentrations 1963-64 Sector A Ground Height XVII-75 XVII-16 Sector Concentrations 1963-64 Sector B Ground Height XVII-76 XVIZ-17 Sector Concentrations 1963-64 Sector C Ground Height XVII-77 XVII-18 Sector Concentrations 1963-64 Sector D, Ground Height XVII-78 XVZI-19 Sector Concentrations 1963-64 Sector D~ Ground Height XVIZ-79 XVII-20 Sector Concentrations 1963-64 Sector E Ground Height XVIZ-80 XVII-21 Sector Concentrations 1963-64 Sector F Ground Height XVZI-81 UFSAR Revision 12 June 1994
| |
| | |
| Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.)
| |
| Table Number Title Pacae XVII-22 Sector Concentrations 1963-64 Sector G Ground Height XVII-82 XVII-23 Estimates of the Least Favorable 30 Days in 100 Y'ears xvII-84 XVII-24 Concentrations in the Least Favorable Calendar Month 1963-64 XVII-85 XVII-25 Annual Average Sector Deposition Rates (Vg = 0.5 cm/sec) XVII-88 XVII-2 6 Annual Average Sector Deposition Rates (Vg = 2.5 cm/sec) xvII-89 XVII-27 Principal Radionuclides in Gaseous Waste Release XVII-92 XVII-28 Correction Factors to Obtain Adjusted Centerline Dose Rates for Sector Estimates XVII-104 XVII-29 Annual Average Gamma Dose Rates XVII-105 XVII-30 Dilution Calculation for Eastward Currents Based on Water Availability XVII-119 XVIII-1 SPDS Parameter Set XVIII-20 UFSAR Revision 12 June 1994
| |
| | |
| SECTION I INTRODUCTION AND
| |
| | |
| ==SUMMARY==
| |
| | |
| 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.
| |
| | |
| 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.
| |
| Careful attention has been given .to fabrication procedures and adherence to code requirements. 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.
| |
| W Periodic test programs have been developed for required engineered safeguards equipment. 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. 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. 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. 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
| |
| | |
| 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.
| |
| g, 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.
| |
| : h. Redundant reactivity shutdown capability is provided independent of normal reactivity control provisions. 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-.
| |
| 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 a0 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.
| |
| | |
| 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. 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.
| |
| 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. 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.
| |
| | |
| 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. 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.
| |
| 6.0 Control and Instrumentation a ~ 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.
| |
| : b. Interlocks or other protective features are provided to augment the reliability of procedural controls in preventing serious accidents.
| |
| Interlock systems are provided which block or prevent rod withdrawal from a multitude of abnormal conditions. The-control rod block logic is shown in Figures VIII-6 and VIII-S, respectively, for the SRM and IRM neutron instrumentation. 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.
| |
| | |
| A cold water addition reactivity excursion is prevented by the procedures and interlocks described in Section XV, Startup of Cold Recirculation Loop (Transient Analysis).
| |
| Security (keycard and alarms) and procedural controls for the drywell and reactor building airlocks are provided to ensure that containment integrity is maintained.
| |
| c 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 aloss reactor of scram in the event of loss of power or air supply to the scram solenoid valves. Periodic testing and calibration of individual subchannels is performed to assure system reliability. toThe ability of the reactor protection system 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. 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. 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.
| |
| | |
| I-9
| |
| : 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.
| |
| 9.0 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.
| |
| | |
| Nine Mile Point Unit 1 FSAR B. CHARACTERISTICS The following is a summary of design and operating characteristics.
| |
| 1.0 Site Location Oswego County, New York State Size of Site 900 Acres Site and Station Niagara Mohawk Power Corp.
| |
| Ownership Net Electrical 615 MW (Maximum)
| |
| Output 2.0 Reactor Reference Rated 1850 MW Thermal Output Dome Pressure 1030 psig Turbine Inlet 956 psig Pressure Total Core Coolant 67. 5 x 10~ lb/hr Flow Rate Steam Flow Rate 7.29 x 10'lb/hr 3.0 Core Circumscribed 167.16 in Core Diameter Active Core 171.125 in Height + Assembly 4.0 Fuel Assembl Number of Fuel 532 Assemblies Fuel Rod Array SRLR('i Fuel Rod Pitch Reference 3 Cladding Material Reference 3 Fuel Material UO~ and UO~-Gdz03 Active Fuel Length Reference 3 Cladding Outside >Reference 3 Diameter Cladding Thickness Reference 3 Fuel Channel Reference 3 Material GE Fuel Bundle Designs, General Electric Company Proprietary, NEDE-31152P, February 1993.
| |
| (4) 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 129 Control Rods Shape of Movable Cruciform Control Rods Pitch of Movable 12.0 in Control Rods -
| |
| Control Material in B4C 70% Theoretical Movable Control Density; Hafnium Rods Type of Control Bottom Entry, Hydraulic Drives Actuated Control of Reactor Movement of Control Rods and Output Variation of Coolant Flow Rate 6.0 Core Desi n and 0 eratin Conditio s Maximum Linear Heat Core Operating Limits Report Generation Rate Heat Transfer 50I 496 Surface Area Average Heat Flux 119,830 Btu/hr-ft~
| |
| Rated Power Xnitial Critical Core Operating Limits Report Power Ratio for Most Limiting Transients Core Average Void 0.280 Fraction Coolant within Assemblies Core Average Exit 10.715%
| |
| Quality Coolant within Assemblies 7.0 Desi n Power Peakin 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 Reference 3 Volume Metric Enrichment Beginning of Cycle 11 Core Effective Multiplication and Control System Worth-No Voids, 20C<4>
| |
| Uncontrolled 1. 098 Fully Controlled 0.958 Strongest Control 0.981 Rod Out Standby Liquid Control System Capability:
| |
| Shutdown Margin (cR) 20C Xe on Free SRLR~ ~ SRLR~ ~
| |
| 9.0 eactor Vessel Inside Diameter Internal Height 17 63 ft ft 9 in 10 in Design Pressure 1250 psig at 575F 10.0 Coolant Recirculation Loo s Location of Recircu- Containment'rywe11 lation Loops Number of Recircula-tion Loops and Pumps Pipe Size 28 in 11.0 Primar Containment Type Pressure Suppression Design Pressure of 62 psig Drywell Vessel Design Pressure of 35 psig Suppression Chamber Vessel Design Leakage Rate 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 Reinforced concrete and steel superstructure with metal siding Internal Design 40 Pressure lb/ft'00%
| |
| Design Leakage Rate 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 0.11g Acceleration Sustained Wind Loading 125 mph, 300 level ft above ground Control Room Shielding 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 Two 115-kV transmission lines Outgoing Power Lines Two 345-kV transmission lines Onsite Power Sources Two diesel generators Provided Two safety-related Station I batteries One nonsafety 125-V dc battery system 15.0 Reactor Instrumentation S stem Location of Neutron In-core Monitor Sensors Ranges of Nuclear Instrumentation:
| |
| Four Startup Range Source to 0.01% rated power Monitors and to 10% with chamber retraction Eight Intermediate 0.0003% to 10% rated power Range Monitors 120 Power Range 1% to 125% rated power Monitors 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)
| |
| | |
| 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. Currently, Niagara Mohawk utilizes various contractors to assist in continuous station modifications.
| |
| | |
| 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.}}
| |
