ML18041A058
| ML18041A058 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/07/1997 |
| From: | Terry C NIAGARA MOHAWK POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18041A059 | List: |
| References | |
| NMP1L-1265, NUDOCS 9711170030 | |
| Download: ML18041A058 (225) | |
Text
{{#Wiki_filter:I CATEGORY 1
" REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
'IV
~
ACCESSION NBR:9711170030 DOC.DATE: 97/11/07 NOTARIZED: YES DOCKET FACIL-50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe 0500022( AUTH;NAME AUTHOR AFFILIATION TERRY,C.D. Niagara Mohawk Power Corp. RECIP..NAME RECIPIENT AFFILIATION Document Control Branch (Document Control esk)
SUBJECT:
.Forwards rev 15 to NMPNS Unit 1 updated FSAR,including changes to QA program description & annual 10CFR50.59 safety evaluation summary rept. DISTRIBDTION CODE: A053D COPIES RE CEIVED:LTR 1 ENCL j SIZE: (268 TITLE: OR Submittal: Updated FSAR (50.71) and Amendments NOTES: RECIPIENT COPIES RECIPIENT COPIES ZD CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 0 HOOD,D 1 1 1 0 AEOD/DOA/IRB 1 1 ILE CENTER 0 2 2 RGN1 1 1 EXTERNAL: IHS 1 1' NOAC- 1 1 NRC PDR 1 NOTE TO ALL "RIDS" RECIPIENTS: PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LIS'. OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTRC DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 10 ENCL 8
l'-' t
NIACARA.MOHAWK NINE MILE POINT NUCLEAR STATIONAAKEROAO. P.O. BOX 63. LYCOMING.NEW YORK 13093/TELEPHONE (315) 349.7263 6 EN ER'ATION FAX (315) 349-4753 BUSINESS- CROUP CARL D. TERRY Vice President Nuclear Safety Assessment and Suppon November 7, 1997 NMP 1L 1265 U. S. Nuclear Regulatory Commission 10 C.F.R. $ 50.71(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 Subj ect: Submittal of Revision 1$ to the ¹ine Mile Point Nuclear Station Unit 1 Einal Safety Analysis Report (Updated), Including Changes to the Quality Assurance Program Description, and the Annual 10 C.F.R. 5$ 0.$ 9 Safety Evaluation Summary Report Gentlemen: P ursuant to the requirements of 10 C.F.R. $ 50.71(e), 10 C.F.R. 550.54(a)(3),.and 10 C.F.R. g50.59(b)(2), Niagara Mohawk Power Corporation hereby submits Revision 15 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 15, are enclosed. Copies are also being sent directly to the Regional Administrator, Region I, and the Senior Resident Inspector at Nine Mile Pc.'nt. The Unit 1 FSAR (Updated) revision contains changes made since the submittal of Revision 14 in June 1996. In addition, Chapter XVIIof the Unit 1 FSAR (Updated) has been reformatted in its entirety to eliminate blank pages, establish a uniform left-margin justification format, and to reorganize the information into "Text/Table/Figure" order. Also, many chapters have been reissued to change the header from "Nine Mile Point Unit 1 FSAR" to "Nine Mile Point Unit 1 UFSAR." The certification required by 10 C.F.R. $ 50.71(e) is attached. 97f i f70030 'gI7i i07 PDR ADOCK 05000220 IIIIIIIIIII!IIIIIIIIIJIIII!HIIIIIIIIII K PDR
~it r
'4 I w 1f p I'=
s ~ 'I I'4ii44
a UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION P In the Matter of =
)
+I ) ~
Docket No.'0-220 s Niagara Mohawk Powei Corp'oration V
)
(Nine Mile Point Unit 1) ) CERTIFICATION Carl D. Terry, being duly sworn, states that he is Vice President Nuclear Safety Assessment and Support 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. Carl D. Terry Vice President Nuclear Safety Assessment and Support Subscribed and sworn to before me, a Notary Public in and for the State of New York and County of , this &day of <<~~<<, 1997. Notary Public in and for County, New York UNA N. tANEfm My Commission Expires: OAry Pubtc, Stete ot Xew Yo4 Registretion No. i908015 8ekzM Ib l I9 CuaMied 4 Jefferson County Coorroission Expires October l3. 19 l t '
't I 1
Page 2 Enclosure A-provides'the identification, reason, and basis for each change to the quality i assurance program description,'nit FSAR (Updated) Appendix B, in,accordance with 10 C.F.R. $ 50.54(a)(3)(ii). The enclosed annual'Safety Evaluation Summa'ry Report.(Enclosure'B) contains brief- .- descriptions of changes to the facility design, piocedures, 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, Carl D. Terry Vice President Nuclear Safety Assessment and Support CDT/LWB/cmk Enclosures xc: Mr. H. J. Miller, Regional Administrator Mr. D. S. Hood, Senior Project Manager, NRR Mr. B, S. Norris, Senior Resident Inspector Records Management
~ p Ape, c
p '4
\ 4 n ~h
, ENCLOSURE A
'.TO NMP1L 1265 Q ~
IDENTIFICATIONOF CHANGES, REASONS AND BASES FOR NMPC-QATR-1 (UFSAR APPENDIX B)'
J
,~ 1
(,~f 4r
ENCLOSUREA IDENTIFICATIONOF CHANGES, REASONS, AND BASES FOR Qh PROGRAM DESCRIPTION CHANGES (UMT I UFSAR APPENDIX B)
-,-'.Hasii for.'Conciik " ButflIieRiVised UFSAR Appendix B=,;
"..:." Pa" Section'.'.". , Identi6catioiiofC e -" : .... Reistonfor, ~
' e",", '<<~6+>7~
'he B.1-2, Section 'age Changed "Manager Human Resource Reorganization that established the assiyunent of these respoaaMities to the B.1.2.1.1 second and third Development" to Director Human position Director Human Resource Director Human Resource Development pmvides paragraphs Resource Development". Development. This position reports to the dear management contml over related functional, Chief Nuclear OQicer and has areas. The rqiorting,of the functtons to the Chief,-"
Deleted "and the General Supervisor responsibility for Employee and Labor Nudear OIBcer ensures effectivi: 1mes of Labor Relations". Relations, Occupational Safety and communication.'hc Job functions and Health, Quality First Program (QIP) responaMities assigned to the diffeient groups'-" administrative issues, and the Fitness for remain the same. 'Zlienfore, the'revised pmgram-Duty Program. The Director Q1P continues to satisfy'the criteria of 10CFR50 continues to report to the Chief Nuclear B and the'QAIR commitments previously
'ppendix Officer on matters related to QlP accepted by the NRC.
concerns. These changes impmve NMPC's ability to maintain a safety conscious work lace. Page B. 14, Section Deleted previous Item b. Reorganization. The functions were Reorgmization iaipmves Quality Asstnance B.1.2.1.1.4.b moved to the other QA supervisors. and value to the Nudear Division. All
'ffectiveness reslensiMities associated with the position of the Supervisor Quality Veri6catioa/Safety Assessment were assumed by'tbo Supervisor Quality Assessment and/or General Supemsor Quality Services. The --
same qmdi6ed individuals continue to perform those functions. Also, the quali6cations'nece.mry those functfons'remain the same. to'erform
W. It '0 tW twv%plw prlwrr!I r I a g If r 0 t ~ 4 I T
- r. \
l 4" 4p ae H ~ r'! P ~4 4 f. r I I 'I I t ~ r
~Pghsis'for,,CoJic}u'di .thethe Revised.Pmgtasa-.-@ '-:.:<UFSAR Appendix B'-'.'; .+Conthiucs.to'Satisfy,'10 - O~ihx'8'i'(44
';.".::i:-'.Pa e/Section.. :,". '."-,'. Identification of Chari" '.-"-:":.:.. ':". .."':Reason for.CIiaii'* K~.>; -.~'~ Page B. 14, Sections Item c to b and Item d to c 'cnumbezed Reorganization. Combined surveillance Reorganization impmves Quality Assurance B.1.2.1.1.4.b and and audit functions into the single effcctivcziess and value to the Nuclear Division. B.1.2.1.1.4.c Changed "Supervisor Quality Assurance functional area "Quality Assessment". SuzveiHance responsibilities associated with the Audits" to "Supervisor Quality position of the Sulervisor Quality Assessment." Verification/Safety Assessment were assumed by the Supervisor Quality Assessmeat. The same quaiificd Added "and conducting performance- individuals continue to perform those functions. based suzveillances" after "QA audits". Also, the qualifications necessary to pezfozm those functions remain the same. Page B.14, Section Rcnumbcrcd Item e to d. Combined all plant support and Reorganization impmves Quality Assurance B.1.2.1.1.4.d administrative functions under Quality effectiveness and value to the Nuclear Division. Added "assessments determining Services. Plant support and administz3&e responsibilities applicability of industry and in-plant associatol with thc position of the Supervisor operating experience, assisting in mot Quality Verification/Safety Asscssmeat were cause evaluations when requested, DER assumed by the General Supervisor Quality Services. trend analysis," after "document The same qualificd individuals contimie to perform contml". those functions. Also, the qualifications necessary to orm those functions remain the same. Page B.24, Section Changed "Engineering" to Clarification. The criteria used to identify The pmcedure to determine the safety classification B.2.2.11.1 "Implementing". structures, systems and components for remained essentially the same aad continues to meet which the QA Pmgram applies was NMPC and 10CHt50 Appczuiix B criteria changed to a Nuclear Implementing Pmcedure fmm a Nuclear Engineering Pmccduze. Page B.24, Section Changed "Appendix B" to "Safety The title of the process changed fmm The pmceduze to dhteraiine the safety ciassification B.2.2.11.2 Classification". Appendix B Determination to Safety remaiaed essentially the same and continues to meet ClassiTication Determination. NMPC and 10CFR50 A dix B criteria Page B.5-2, Section Deleted "emergency plan implementing Clarification. Moved the emergency plan The periodic frequency was shortened; therefore, the B.5.2.6.3 pzoccdllfes . implementing pmccdures to the next level of commitmeat previously accepted by the paragraph. Periodic reviews require a full NRC was not reduced. A full revision is moie Added "full"between "A"and IevlsloIL restrictive and is required by NMPC procedures to "revision". qualify as a periodic review.
'ttwwkMAO taataattta~KBL&l~t I Mttt'wAwt,Jtahtth al ~ , I I 'l 7
ja I I I 4 4
~ I I
l r 4 ~ << 4 6<< LI -t I4. I ~ I
~ \
4 I,I r I hl f Ip g I ', '>>
~ " ~ ~
e Ir 4 C~ a 7
<<I 4
4 gh t'. tm t .p It ay1 g g, 47 r ha ti ' C f I 4 R j P tl ag ' tt P1
,~
I 4 I<<' )rg hr t aj
'I
~ \:
i'<< c~~~~~ c":".'.;";:le~<'~yBasjs'for.'Concluding'thatOeNkiJ'sckProgaua%$ <<-*-"',UFSARAppendIx B.':".'.
.;.;-::<<::.-':Pa'/Section".:.'-.";:"', ~', '..',:. Identlficatiori'ofChan':,; '.,' ',
.', '; '.: Reasei for Chaii Page B.5-2, Sections Added "Emergency plan implementing Implementation of the nquirements of The periodic &xgzuey was slertcncd; therefore, the B.5.2.6.4 procedures are reviewed at least annually NUREG4654 Revision 01 and Regulatory level of commitmcnt previously acccI~ by the and revised as appropriate. A full Guide 1. 101. NRC was not reduced.
revision of a pmcedure, or detailed scrutiny of a procedure as part of a documented training program, drill, simulator exercise or other such activity, constitutes a rocedure review". Page B.15-1, Section Deleted entire paragraph. Editorial. NMPC currently uses only one Nuclear Implementing Pnxedurcs were generated B.15.1, second paragraph type of system (Deviation/Event Report) to several years ago. MP-ECA41 "Deviation/Event identify, contml and disposition Report" (DER) was developed to incorporate the nonconforming conditions in materials, differen departmental systems. The retuimnents of arts corn nents or services. 10CFR50 A dix B contirme tobe met. Page B.15-1, Section Deleted "departmental". Editorial. NMPC currently uses only one Nuclear Implementing Pmcolures were generated B.15.2.2 type of system (Deviation/Event Report) to sevens years ago. MP-ECA41 "Deviation/Event identify, contml and disposition Report" (DER) was developed to incorporate the nonconforming conditions in materials, differen departmental sytNms. The rotuircruents of corn nents or services. IOCFR50 A dix B continue to be met. Page B.15-2, Section Deleted "departmental". Editorial. NMPC currently uses only onc Nuclear Implementing Pmcohres werc generated B.15.2.12 type of system (Deviation/Event Report) to several years ago. NIP-ECA41 "Deviation/Event identify, contml and disposition Report" (DER) was developed to incorporate the nonconforming conditions in materials, different departmental systems. The requirements of arts corn nents or services. 10CFR50 A dix B continue tobe met. Page B.15-2, Section Changed "senior nuclear division and Reorganization. To linc up with the Rcorgaruzation appmvcd by NRC via Unit 1 License B.15.2.13 corporate management" to "nuclear current management organization Amendment 157 and Unit 2 License Amendment division mana ement". described in Sections B. 1 and B.2. 71 dated Feb 20 1996. Page B.16-1, Section Deleted "departmental". Editorial. NMPC currently uses only one Nuclear Implementing Pmcedures were generated B.16.2.2 type of system (Deviation/Event Report) to several years ago. NIP-ECA41 "Deviation/Event identify, control and disposition Report" (DER) was developed to incorporate the nonconforming conditions in materials, differen departmental systcrM. The repnrements of parts, components or services. 10CFR50 Appendix B continue tobe met
f'.->f'k.f
~
v
~
f,b 1 c f ~
%4 I (5 gg V'~
fl f Pi af W r .'ei C',fC f [f, c C~~ ' I 3 A h ~
~ s f
~
f df
)~
I
~f ~ 4 f
f
.'-.-;.-~Basis:for-Coacludingth@iljeR'eviicKPxograxa~~
UFSAR Appendix B Pa e/Section Identification of Clian . Reason-for Chan B. 17-1, Section 'age Added "Quality Assurance" between Clarification. The addition of the words Adding "Quality Assmaace" between "considered" j, B.17.2.2 "considered" and "records". "Quality Assurance" provides a more and "records" is consistent with the wording in precise and accurate description of what 10CHt50 Appendix B Section XVK The change is Deleted 'These records include: 1. these documents are considered upon considered a clarification of an existiag commitmeut'nd, Results of ...calibration procedures and completion. The description of what types therefore, does aot contradict or alter any reports . of documents become records upon commitments previously apprmed by the NRG completion is contained in the first / Added "Additionally, the Records sentence of Section B.17.2.2. The specific The addition to the second statexaeat is consistent Management Program includes those list of records was removed since it was with 10CFR50 Appendix B Section XVIIand records identified in plant Technical not an all-inclusive list. The addition of ANSI/ASMBNQA-101983 (17, 17S-I). Inclusion Specifications." the statement "Additionally, the Records of a partial list of documents considered to M into Management ... in plant Technical this category allows the reader unaecessaxy mom for Specifications" ensures that those xecords uusintexpxetatioa. While a'reader may interpret that identified in Technical Specifications as a paxticuhr document need aotbe coatmlled by requiring retention, but which do not meet procedure because that document did not appear on the definition of a Quality Assurance the list of examples pmvided in the QATR, no such record, willbe captured under the Records misinterpretation can be made ifthe paxtial list is Management Pmgram. eliminated. Ifthe list is not all~usive and stand-alone it should aotbe inchided. ( I It The third statement ensuxei that those records identified in plant Technical Specifications as reqixiring retentioa, but which do not meet the definition of a Quality Assuiaace record, willbe ca under the Page B. 17-1, Section Changed "permanent" to "lifetime". Clarification. To be consistent with the The texns "lifetime" and "permanent," when B.17.2.3 terms used in NQA-1 to avoid any applied to Quality Asmaace records, are tential confusion. 0 ous. Page B.17-2, Section Changed "Except for records that are ClariTication by eliminating redundant The intent of this section was aot altered. This B.17.2.8 stored as originals, such as radiographs exception for records stored as originals. clarificati eliminates a redundant exception for
... or features are used" to "Records are When only a single original can be records stored as originals.
stored in appropriate fire rated facilities, retained, it willobviously not be stored in or in remote dual facilities to prevent a remote, dual facility. damage, deterioration, or loss due to natural or unnatural causes."
4W W <
- JF~~y~4aabSJUFFkk~R X hXFX'FX
'F tt, F i >~ C f 1 x C". x xXF VF A>> 'F F 5 iP F JF NI 4 ">> L F F e x ~
~
A
'" 'Uk('ReRiiRPiogQik g.Basid;for';Conch'Hing.
<~:;"-UFSAR-Appendix B',:::;
<...'::~,:..'Identification'of Chan '"-.'." ';;" ';;: ",';.";=:....;:"- Reason for C~Ci "7:."".~$;~,"'2 %
Table B-3, Sheet 4 of 8 Changed Exception wording in Item 3.r This was part of the correctivefpreventive The use of the PMST database for in~lant to "Installed plant instrumentation actions from a DER written during an equipment allows for.better taichng and scheduling calibration status is tracked through the ISEG assessment. The site was not of the calibration ofthis equipmerit. This database 's PMST database. Calibration status of implementing the exception as it was addressed in the procertures and used in training." portable measurement dt test equipment written. The portable MkTBsti11 are required to maintain (MATE) may be labeled on the case or the same type of calibration hibeling as the original attached to the device. For instances exception. 'Hie reqmrenients of ANSI/ANS-3-2 and where size or application precludes 10CFR50 Appendix; B contimie tobe met. attaching the calibration labels on the device, the device shall be uniquely identified and traceable to its calibration record. Table B-3, Sheet 5 of 8 Changed Exception in Item 4.c from Clarification. Some of the SRAB raImred Clarification. Some of the SRAB rotuired audits are "Personnel who perform audits for the audits are in the scope of Section B.18 of in the scope of Section B.18 ofthe QA'IK SRAB are not required to be so qualified, the QATR. since these audits are outside the scope of the audit program described in Section B.18 of this QATR" to "Personnel who perform SRAB audits that are outside the ~ \' scope of 10CFR50 Appendix B are not uired to be so ualified."
~ I, 4~ ~'.Vali C
MW:x..c rw r '
\ P>rL~, ~
Q fr t> 14$ 9ih
~ -~, ~ Ilf ,,i ~ ~
I
> ~,
p k
~ y s- r r~ C ml Y
1 4
~ ~
- 4. s4 f~
~ ~ s
Enclosure B to NMP1L 1265
~ -i NINE MILE POINT - UNIT 1 SAFETY EVALUATION
SUMMARY
REPORT 1997 Docket No. 50-220 License No. DPR-63
' - I I 3 b 4 f I 4 ~ t ll '"i I".. Ig I t h I
Safety Evaluation Summary Report Page 1 of 68 .,- Safety Evaluation No.: Mod;:N1-86-085~ ..-.~.~--.g;-,,> -,-=,.>--.; 91:-'002'mplementation.Document No.-.'-> ....>; UFSAR Affected Pages: N/A'~".-'1, g \ Qp System: 600 VAC and 480 VAC Distribution Systems 4 ~ ~ g Title of Change: AK Breaker Overcurrent Trip Device: Replacement T O&l ~ ~ ~ Description of Change: g This modification replaced the General. Electric EC electromechanical overcurrent. trip devices in the AK breakers with Westinghouse solid-state Amptector overcurrent devices. Due to the age of the EC devices and the inherent design principle of the electromechanical type trip device, these EC devices had experienced an unusually high failure rate during testing of approximately 50 percent. Safety Evaluation Summary: The overcurrent trip function already exists and the modification only changes. the method of performing the function. The failure modes and effects were found to be identical to the modes and effects of the currently installed devices, and the new overcurrent trip devices are much more reliable. The new devices also permit greater flexibility in trip settings, allowing better achievement of proper selectivity and coordination in the low-voltage distribution system. 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 68
-. ~= '=Safety Evaluation No.: 9'2-041
-Implementation Document No -'"~--'IST Program.Plan'-"'.::"-'."'-~-'=-.
tg
- t:- "'< ~i~...
QFSAR Affected Pages: '""X-'16 System: Contr'ol Rod Drive (CRD) Title of Change: Update of FSAR to Reflect Revised Testing
'equirements of CRD Pumps 011 and 012. '
Description of Change:
'he CRD pumps are not safety related; therefore, the In-Service Testing Program does not need to test and t'rend these pumps in accordance with ASME Section "XI. The only requirements fo'r the pumps with respect to Technical Specifications
=-is that they be capable of delivering 40 gpm to the reactor vessel as makeup flow.'
This change updated the UFSAR to state that monitoring will be done under the quarterly surveillance test. The purpose of the surveillance test (N1-ST-02) is to assure that the Technical Specification requirement is met. Safety Evaluation Summary: The quarterly surveillance test will provide an opportunity to determine if and when pump degradation is occurring. Also, it will assure performance in accordance with Technical Specification requirements. This change in the mechanism used for trending has in no way had any impact on system availability or capability. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety EvaluatIon Summary Report. Page 3 of 68 ..', .,:.;,. Safety Evaluation No.:
~,'
-" P. ', <" 94-066 tmplementation Document No:-" '-~:-'".'Procedures N1-RTP-31,.N1-OP-50A..',;.:
",";.'FSAR Affected Pages: Table Xll-8 System: Area Radiation Monitoring Title of Change: Justification for Removal of,ARM-13 From Service Description of Change:
Area radiation monitor (ARM) number 13 has been retired in place in Radwaste Pump Room El. 225'..The pump room was used as a "drumming operation", whereby drums were fitted and elevated to a loading dock for transport;,All equipment associated with that operation has been removed as part of the cleanup effort. This ARM has not been required for service since 1981 when El. 225'f the radwaste contamination level became too high for further use. All radwaste operations were ceased at that time. When the decontamination effort was completed in 1993, an attempt was made to return ARM 13 to normal service, but it was discovered that the cables to the ARM were severed and that the ARM itself was painted over. Safety Evaluation Summary: Only two ARMs are credited during or following an accident; they are the Control Room vent and Refuel Floor high range monitors. The ARMs located in the Reactor Building are employed in executing Emergency Operating Procedures to monitor secondary containment radiation levels. The purpose of ARM 13 is to detect high rates of exposure during radwaste operations (existing or planned). Since the Radwaste Pump Room (El. 225') is no longer used for radwaste operations and ARM 13 is not credited for any accident, this change does not increase the probability of any accident previously evaluated in the SAR. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
,, Safety Evaluation
'ummary Report
<.-'Page.4 of 68
.Safety. Evaluation No.: 95-007 Rev. 1 &2 Implementation'ocument No. ~.: '<<.,Mod. N1-94-003l'"; .*i:n; ~: -0;:,:-;~i:-.;i.e~,,r;--.,
~'UFSAR Affected Pages: .. "=<<'N/A System: Reactor Vessel (RXVE)
Title of Change: Core Shroud Repair Installation Description of Change: cgkl 4'1. I ~ ~ This safety evaluation evaluated the shroud repair installation activities and supplements Safety Evaluations 94-080 "Core Shroud Repair and 96-018 - - ':->-',
"Modification to the. Core Shroud Repair Tie Rod Assemblies."
The NRC issued Generic Letter 94-03 due to observed cracking in the core shrouds of several boiling water reactors. This generic letter required inspection of
'the shroud and/or repair, if necessary. Revision 0 of this safety evaluation evaluated work performed during RFO13 and Revision 1 evaluated work performed during RFO14. Revision 2 evaluated the use of the 25-ton auxiliary hoist. NMPC performed a preemptive repair of the shroud during RFO13. The NMP1 reactor core shroud repair was designed to structurally replace shroud welds H1 through H8. The installation of the entire repair involved electrical discharge machining (EDM) of the shroud support cone and shroud itself, which generated 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. During RFO14, the 270 azimuthal tie rod assembly installed during RFO13 was removed and replaced with a modified spare tie rod assembly. Also, the lower spring contact against the shroud was modified to extend beyond the H6A weld on all four tie rod assemblies.
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.
I g
-Safety Evaluation Summary Report
."'Page 5 of 68
;,,;:-; Safety Evaluation-No.: ';=.;- 95-;007 Rev. 1 8c 2 (cont'd.) ~:~ ....
al
'afety Evaluation Summary: (cont'.d.) 'gt y ~t t+~A ~C'h AQQ/P j., Eral &
~ ~ ',
"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:
1 micron in size and will not affect the function of the SFP system. 'han
~ ~ ~ I a 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 General Electric shroud repair procedures. The tooling for "heavy loads" has been designed and will be used in accordance with NUREG-0612. During RFO14, the removal and installation of the 270'ie rod meets the requirements of NUREG-0612 by using lifting devices which meet NUREG-0612. The dose rates resulting from the removal of the 270'ie rod assembly and the installation of extension pieces will have minimal radiological impact and the radiological controls used during the removal and installation will ensure that there are no adverse impacts on the 10CFR20 limits. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 6 of 68 Safety Evaluation No.: .. -. I '= 95-01'1 Rev. 1 3mplementatlon Document No.:; N/A-'~ UFSAR Affected Pages: - -,'.1-15, IV-12, IV-32, V-,21'; XV-79 ""- ~ '
~
System: Various-Title of Change: Operation of NMP1.Reload 13/Cycle 12
>I y ~ t 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 NMP1. Two, i" Hundred'(200) new fuel bundles of the GE11 design were loaded. All 164 of the P8x8R bundles from Cycle 10, and 36 of the GE8x8EB 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. Revision 1 of this Safety Evaluation incorporated the changes necessary to the operating limits as a result of the revised General Electric Supplemental Reload Licensing 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 MCPR 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 NlVIP1 can be safely operated during Reload 13/Cycle 12 and that this change does not involve an unreviewed safety question.
C
'.-~ Safety Evaluation
=--. ==.-Summary Report Page'7 of 68
,;, ,.;,;..';-: Safety Evaluation No.: 95-012
> ':-'-= Implementation Document No'.: -=~'- -". Procedure N'i-MMP-GEN-904 .;-..":::-.-.::;-;-
"" UFSAR Affected Pages: X-'38, XKO, X&2 >-
System: N/A Title of Change: Reactor Servicing Platform Description of Change: This change removed references in the UFSAR regarding the use of the reactor servicing platform for disassembling/assembling the. steam separator assembly from the core structure during refueling activities; The reactor servicing platform was provided by General Electric Company to facilitate refueling. activities during the original construction of the plant. Safety Evaluation Summary: The ability to remove/install the steam separator without the use of the reactor servicing platform will not be affected. Not using the platform will not contribute to the initiation of any accident 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 8 of 68 ,, Safety Evaluation No.:- '95-024 -: *-'. r <<<<<<
~
Implementation Docume'nt No.':
" '~ Mod::N1~95-003 -;~=,;: ~;:< n~.'.-<a;=.".~; ~.'.
UFSAB,Affected Pages: ~t<<~ ggi~<<s lg"i,mqygl.P.5g~'jap '
~ ~
g System: Sewage Treatment Plant r<<
~ <<<<
Title of Change: Sewage Treatment Plant Dechlorination-- System
<<<<> ~ <<<<v <<(,
<<e<<+w vi<<<<<<i Description of Change:
4<<; I This modification installed new metering pumps, flow controllers, tanks and mixers to provide sodium sulfite to the Sewage Treatment Plant effluent to dechlorinate -. ~ the effluent and comply with SPDES permit levels for chlorine. This was required due to the decrease in permitted effluent chlorine levels as delineated in the revised SPDES permit issued December 1994. Safety Evaluation Summary: The design and operation of the new equipment associated with the injection of sodium sulfite to reduce the total residual chlorine level in the sewage plant effluent is in accordance with applicable criteria. The metering pumps will be automatically controlled by the total plant effluent signal and cover the full range of effluent flow from 0-120,000 gpd. The sodium sulfite solution concentration and calibrated flow rate are determined by the Sewage Treatment Plant Operator to produce the desired concentration in the process stream. The material used to manufacture the pumps, tubing and tanks is designed for mild chemical usage, which includes hypochlorite and sodium sulfite at the concentrations used in the facility. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
';-.-'=':~;:Safety Evaluation
'".'=,:Summary Report
- "-:>. Pigs 9 of 68
,,-,-, '"e!
. -,.: Safety'-Evaluation No.: -'-:95-101 I ';; ','.
"-'-.::.. Implementation Document No.= :s'5-:"";-'-
DER 95-2151" ~'~ <<~-.. >; .( .".
!'UFSAR Affected Pages: I-'12, IX-22, IX-24, IX-26, 10A-61 System: '125 'VDC System Title of Change: Reclassification of Battery 14 and Battery
. Board 14 from Non-Safety Related to 0-Related for Station Blackout Description of Change:
~ I I The control room dc emergency. lighting circuit 12 and paging system inverter are loads which are required.to cope with a station blackout event. Although these loads are nonsafety related, their power supplies are required to be quality related (0).
This change reclassifies Battery 14 and Battery Board 14 main breakers, bus, and feeder breakers which feed these two loads, as 0 related. Safety Evaluation Summary: The reclassification of Battery 14 and Battery Board 14 from nonsafety related to 0 related ensures that the future procurement of replacement components or parts and the installation, maintenance and testing are completed in conformance with design requirements. This change also assures Battery 14 has sufficient capacity to cope with a station blackout event in accordance with applicable design criteria. This change does not increase the probability, consequences or create a different type of accident or malfunction of equipment important to safety. 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 '68 ...,,, '.Safety Evaluation No.: 95-103 Implementation Document No.:. r - . 'DER':1.-95-2643.-..: f:":t:>u ':C":-, '--.;. -,
'UFSAR Affected Pages: ' ". t:. - Figure Vill-14 .i:.;;;m.-; ..~;;.;p ",;..': ',.:
System: Neutron Monitoring t NMS) Title of Change: ~ APRM Rod Block Calibration Description of Change: UFSAR Figure Vill-14 previously showed. the Technical Specification rod block as a horizontal line between 100% and 120% of recirculation core flow. This was being interpreted to require that the rod block setpoint be demonstrated to be calibrated to within the nominal trip setpoint, as described in Specification E133, at 100% and 120% of recirculation flow. In addition, the hardware was not capable of producing a horizontal line (setpolnt). There is a positive slope; i.e., the setpoint increases with increasing recirculation flow. Because of'this slope, the setpoint at 100% flow was lower than necessary, so that the setpoint at 120% flow could be set within the tolerance described in the specification. Hence, the setpoint at 100% flow caused unnecessary rod blocks. This change revised the UFSAR figure to allow for calibration of the APRM rod block setpoint at 107.1% recirculation flow. This was a change to the method of calibration only and did not require a hardware change. Safety Evaluation Summary: The APRM rod block responds to accidents and transients and, therefore, by design cannot initiate an accident or transient. The APRM rod block is not taken credit for in any accidents or transients described in the UFSAR. In addition, the scram setpoint is not affected. The APRM rod block will still provide margin to ensure fuel design limits are satisfied. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
.Safety Evaluation Summary Report Page 11 of 68 Safety'Evaluation No.: :95-106 implementation Document No=..", . a>'~.'N/A" =; OFSAR 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: a: ~ This safety evaluation addresses the demolition of the following buildings'located east of the. Unit 2 plant structures. t Carpenter's shop
- 2. Paint shop
- 3. Electric fab shop 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/Q 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 12 ef 68-.,- ., Safety Evaluation No.: ., .95-108 w gA 4 S'> ( ~ r~r Implementation Document No.: Procedure,GAP.-.RPP.-01;:u- .....~~t>'~i:,-'FSAR Affected Pages: - ..- ';. N/A ~~=.-.'ystem: N/A , ~
"~ il of Change: ;-10CFR19 Required Training For Personnel
~
Outside the Restricted Area
!'itle Description of Change: 1
<<ya~<~y i'lW
~ 5 ~
This safety evaluation evaluated the change to. Procedure GAP-RPP-01.which now requires training be provided for all individuals who, in.the course of their employment, are likely to receive an occupational dose in. excess of 100 mRem per year. This change complies with the revised requirements identified in 10CFR19. Safety Evaluation Summary: The proposed change involves training for personnel in the Unrestricted Area of the site and will meet the intent of the revised 10CFR19 and satisfy applicable portions of regulatory guidelines. Training of personnel outside the Restricted Area who are likely to receive an occupational dose of 100 m/Rem will not increase the probability of occurrence or the consequences of an accident or malfunction of a different type than already analyzed in the SAR. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety. Evaluation ,Summary Report Page 13 of 68 Safety:.Evaluation No.: :96-001 Implementation Document No.:.'-. t'.DER.1-94-0462; .,; .-,<~"... .: ~ ia "UFSAR Affected Pages: Xll-17, XII-18; Figure IIIX System: N/A Title of Change: 'Changes to RP- Facilities, Section XII and Section III Description of Change: This safety evaluation evaluated the following changes to the UFSAR: 4 ~ ~ ~ 1 ~
- 1. The instrument storage room. is now;in the administration building near the main access point. =
- 2. An auxiliary counting laboratory for portable count-rate instruments is now located in the old instrument storage room.
- 3. The current instrument storage room is also used for analysis of radiation protection samples using count-rate and gamma spectroscopy instruments.
- a. The auxiliary counting room is now being used to house a panoramic irradiator for calibration of dosimetry devices and testing of radiation detection instruments.
Safety Evaluation Summary: The changes to the UFSAR describe the current configuration of radiation protection facilities in the Turbine Building. Storage of portable radiation protection instruments, calibration of count-rate instruments, analysis of radiation protection samples, and location of the panoramic irradiator in the auxiliary counting room do not affect any equipment malfunctions or procedural errors that initiate any of the accidents analyzed in the SAR, and thus would not increase their probability of occurrence. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety&valuation .. Summary Report ~ Page 14 of 68'afety Evaluation No;: 9&402-P
.-'-"-.-- Procedure NIP-FPP-.01~:..
Implementation Document No.: UFSAR Affected Pages: - -;.i -> '- '=
-X&6;~10A-13; 10A-%8plQA-'56,:10B-196.'='ystem:
N/A il Title of'Change: Fire Brigade Membership Requirements and Revision of NlP-FPP-01 Description of Change:
\~
This evaluation examined the requirements for Fire Brigade membership-and the staff which may be qualified for membership in the Fire Brigade. Previously, the Fire Brigade leader and two of the Fire Brigade members. were required to be part of the fire protection staff. This change allows plant staff members who are qualified in accordance with the Fire Brigade training program to serve as Fire Brigade members at the level to which they are assigned. Safety Evaluation Summary: Niagara Mohawk Power Corporation has traditionally staffed the Fire Brigade at Nine Mile Point with "professional" firafighters, based on the concept that personnel assigned to the Fire Brigade were dedicated to fire protection duties. ln 1994, the composition of the Fire Brigade was modified to allow two of the Fire Brigade members to be non-fire protection staff personnel. Part of the philosophy for that modification was that each fire attack team could still have one full-time fire protection staff member, a "professional" firefighter, assigned to lead the fire hose attack in fire suppression activities. As these teams consisting of fire protection and non-fire protection staff personnel have practiced as teams and matured as Fire Brigade members, it has become apparent that non-fire protection personnel can perform fire suppression activities effectively, given adequate training and practice sessions (drills). Based on this, the Fire Brigade membership requirements are being revised to allow any individual receiving adequate training and practice to be assigned to the Fire Brigade. 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 68'afety Evaluation No:- 96-004'ER.1-'96-'0418. implementation Document No.: -:"..:,:,". ~'n":.;-.: '>" )~-=.--,.<. f7 4 UFSAR'Affected Pages N/A ) 8 "
~ ~
System: Liquid Radwaste Processing Systems (The rmex) Title of Change: Treatment of Sanitary Waste by Radwaste Systems Description of Change: After completing a city water outage for routine maintenance, water was discovered coming out of the top of the sewage line located on Turbine Building El. 250 between the cable spreading room and the remote shutdown panel. Further inspection revealed the pipe had developed a crack approximately three feet in length and up to three inches wide on the top of the pipe. Due to the initial surge of water and continued water usage (because of fixture valves not closing), the sanitary waste leaked from the pipe onto the floor. The sanitary waste/water mixture entered plant floor drains and was pumped from the turbine building sumps into the utility collector tank in the radwaste facility where radwaste operators were able to prevent it from processing through the Thermex
'ystem.
This safety evaluation evaluated treatment of the sanitary waste which entered the plant with existing radwaste equipment. Safety Evaluation Summary: The water/sewage mixture is contained in the utility collector tank. The treatment scheme will be to raise the pH of the tank's contents for the purpose of dissolving the organic and inorganic matter and for killing any biological organisms which may be growing in the tank. The solution will be maintained at a pH of approximately 10-10.5. The solution's pH will then be adjusted downward to eliminate depletion of radwaste resin. Any solids which do not dissolve will be removed by filtration. Soluble material will be removed by a combination of filtration by charcoal, reverse osmosis membranes, and by demineralization. Ultraviolet lights are available and can be used if necessary to oxidize organic material for easier removal. The effluent water will be evaluated using existing chemistry procedures before the water is released to the condensate storage tanks for reuse.
--Safety Evaluation
-"Summary" Report "Page 16'of 68
.-.;-,.- .,~=-"=-Safety Evaluation No.: ::96-004..{cont d.)
'- Safety Evaluation Summary: = ;.{cont.d.i: ..- -'." "-.'i'."i~ -"",~.~ ~Q nou::-sr~arr,--';.<<=,',
'The'resulting waste will be in a form which will allow for disposal in accordance-;
with current license basis documents.
,<<rg<> ~ >><<$ <<I
<<~ ~ ' >>>> C>> ~ [ ~ ~
I
'Safety Evaluation Summary'Report Page 17 of 68 z...,--Safety EvaluatIon No.: , ~"..".; =.:"96-005 Implementation Document No.: ;Procedut'e 'N1-STP.-56 q;~.....-..:>; -: =
'UFSAR Affected 'Pages: .'N/A~". ~
System: Feedwater ~- Title of Change: Procedure N1-STP-56, Feedwater {Rhf) Heater Leak Test-Descrlptlon of Change: Tracer technology has been used to calibrate feedwater.flow venturis and to conduct steam purity evaluations. Due to.the radiological concerns associated with the use of the radioactive tracer, sodium-24, potassium nitrate (KNO,) was selected for use at NMP1 in order to quantify the feedwater heat exchanger tube leaks; Potassium nitrate is a neutral salt which is soluble in water and completely dissociates. The use of this nonradioactive tracer provided the necessary level of detection without the radiological challenges of a radioactive tracer. Procedure N1-STP-56 determined the FW heaters which had tube leaks. The test also was able to estimate the size of leaks. The location and amount of the leak was needed to determine the best economical solution to the problem. The injection point was through sample valves downstream of FW booster pumps. The sample points are downstream of the sample system heat exchangers. Cooling water was supplied from service water, and mixing water was supplied from demineralized water. The waste cooling water and sample water was released to the floor drains. The equipment required a 5 gpm cooling water flow rate. The power requirements were supplied by 240 VAC welding outlet for the vendor-supplied injection equipment and 110 VAC for the vendor-supplied control equipment. Safety Evaluation Summary: The plant will not be significantly affected by this test and the margin of safety is unchanged.
.Safety Evaluation
-.:Summary Report Page=18 of 68
,'.".,;..':;.'..-Safety Evaluation No.: '.005 (cont'd.):."Fi:.ai'~ '~., -.',...
";Safety Evaluation Summary!', ':= ll'r<~>"(cont-'4-) ' .',~ ~n-.'.nua=~.n".lzr.-:-;ia~r.'i<..-.:.
~
~Potassium nitrate is readily available Qrlth extremely low chemical contamlnants.:-'
Thls material ls ideal for tracer quantification since it is nonvolatile and forms no harmful by-products In a nuclear environment.~ This type of test has been done.-- successfully at other boiling water reactor plants..
~ '
The injection and sampling equipment will be attached to nonsafety-related sample and drain connections. If any problems occur, the equipment can be isolated from the plant systems. The flow of cooling water from service water may be ..: ..:: ... approximately 5 gpm and radwaste is able to receive and process this water. There is no significant increased=risk to the plant systems from installation of the -.: test equipment or to the fuel frominjection of the tracer chemical. The ability of:.- the plant to shut down, and remain shut down, will not be'impacted by injection of the chemical tracer. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report 19 of 'age 68Mod. .. Safety, Evaluation No.:,. l .",';,;"; '96-007 implementation Document No.: N1-.95-006;;e~ ~~:= UFSAR Affected Pages: ~ '..', N/A-:"."..;.:-', ".::---;. "-Ii.-~-~a<::~: ~::.~a~
-=,
System: Spent Fuel Pool Spent Fuel Storage Rack - 8202-11 t Description of Change: I ~ ~ This design change relocated the control rod blade (CRB) holders to the cask drop protection system (CDPS); removed the work table (WP1), 1000-lb. test weight,. and seismic restraints in front of the spent fuel gates; installed the 198-cell spent fuel rack, 8202-11, as a freestanding. structure in this location; and installed the Holtec overhead platform (HOP) on top of the southwest corner spent fuel rack. Engineering workscope included the seismic qualification of the rack as a freestanding uncoupled structure, evaluation of localized nucleate. boiling within the rack and calculation of maximum cladding temperature, and calculation for all the rigging required to ensure compliance with NUREG-0612. Safety Evaluation Summary: In order to remove the work platform WP1, the CRB holders currently bolted t'o the table need to be relocated to the CDPS temporarily and as required to support future blade exchanges. The CRB holders have been evaluated in Calculation No. S10RX340SPRIG23 as freestanding structures in the CDPS, either loaded or unloaded with control blades. The analysis completed in accordance with the applicable criteria concludes that no damage will occur to the spent fuel pool or CDPS due to a seismic event or other abnormal transient. No damage to fuel or fuel racks will occur as the CDPS is isolated from the remainder of the spent fuel pool. The CRB holders in the CDPS will be used as required to support control blade exchanges to and from the reactor and to and from the single blade holders on the spent fuel pool curb. The duration that control blades will actually be stored in the CRB holders in the CDPS is small, and the consequences of a transient or accident involving control blades is insignificant. Calculation No. S10RX340SPRIG23 demonstrates that the CRB holders will not overturn during a seismic event and no damage to the CDPS can occur. The work table, restraints and test weight will be pressure washed during removal from the pool to minimize contamination and exposure. The equipment will be
Safety Evaluation Summary Report
'Page 20 of 68
'96-007 (cont'd.) w- i iM f p j i i
...:- .:,-. Safety EvaloaSon No.:
l If\
'L ~ 'l ~
1 \a 1 Safety- Evaluation Summary':-~-""'-"'-=---'-'-'<(corit'd ) +v< ZeDAWp i w is gt tP 6w i3 fi%3 ifi ~ placed in the designated laydown area and wrapped 'at the direction of Radiation=.".: Protection. The accidents relevant to a spent fuel rack and the spent fuel pool include'a fuel bundle. drop, an Inadvertent criticality, and a loss of spent fuel pool cooling. Heavy loads will not be handled over spent fuel with the exception of the HOP. The HOP will be installed utilizing the 125-ton crane. In addition, all heavy loads be handled in accordance with NUREG-0612 and applicable NMPC procedures.
.'ill As such, a heavy load drop is highly improbable, and does not increase the probability of an accident evaluateddn'the-UFSAR. The spentfuel pool activities;.,
required to install.'the~198-cell spent fuel rack and HOP include the relocation of the CRB holders, the removal of. the existing'work table; seismic'restraints and ' 1000-lb. test weight, and associated preoperational testing requirements for the rack. None of these activities are initiators of the accidents described in the UFSAR. While spent fuel will be relocated prior to and after the installation of this design change, this will be completed in accordance with the applicable fuel handling procedures and has been previously evaluated. The design codes, calculations, materials, installation procedures and post-installation testing assure that the probability of occurrence of an accident associated with the spent fuel and spent fuel pool will not be increased. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report ,Page 21 of 68 .Safety Evaluation No.: . 96-.008'.i-:
.Implementation Document No'.. " DDC ~1 E00045: .;."; ", q.'""<. <<,:Q:~ . <".:~ ",~;
UFSAR Affected Pages: ~;,:III-16'-:" -'."=:,>a;..~;~:~'~ z';-.."'~.i.'ystem: Radwaste Building Heating 8c Ventilation (HVW) Title of Change: Waste Building Control Room Alarm Description of Change: This change retired in place the Radwaste Building high radiation alarm. The continuous air monitoring system warns personnel occupying or entering the Radwaste Building of significant airborne contamination levels, and a high radiation signal still alarms in the main control room. Safety Evaluation Summary: The proposed change removes only the requirement for the Radwaste Building ventilation radiation alarm in the waste control room. The ability to detect high radiation levels is provided in the main control room and via local alarms. Deleting the alarm cannot increase the probability. of an accident because its function is alarm only. It does not provide a trip, nor does it control other components, i.e., valves, pumps, etc. It is not discussed in the SAR as part of any transient or accident analysis. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluatlon-Summary. Report ". Page 22of 68 ' l ... Safety Evaluation No.: 96-0.'IO ~ ~>>= 'll ~ 4(Va>> - >aj'<<AI.>>
"~a 'mplementation Document No;:",'<9.Procedure NEP-POL&1:..-'.-'G ri-"-.'4.-name! qm..
UFSAR.Affected Pages; Fig ul'e"Xlll-3 N/A
"'ystem:
Title of Change: Restructuring of Unit 1 Engineering in Accordance with Revised Procedure NEP-POL-01
>> o I ~ ~
k - R [ g>>>> ~ Description of Change: a ~
~
=
I ~ J l( ~ >>
~ ~ , ~ f. "~
Procedure NEP-POL-01, "Nuclear Engineering Department Organization," has been revised-to reflect organizational changes in Unit.1 Engineering. The Unit.1 Plant. Evaluation group, consisting of a supervisor and one engineer, has been merged with the Unit 1 Project Management group. The Supervisor - Plant Evaluation position has been eliminated. Both individuals in the Plant Evaluation group now report to the Unit 1 Supervisor - Project Management. 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 SRP 13.1.1. 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 23 of 68. Safety Evaluation No.: 96-011 .;- E~,(h 1 Implementation Document No.:,.:.,DER:,1-94-1980';, i,: --., -.;.;.;.:r: ", -..:,,;-, VFSAR Affected Pages: ~ N/A -"-< '"<. System: Control Room Air Treatment, Reactor Building Emergency Ventilation Title of Change: Revision to the Bases for Technical Specification 3.4 4/4.4 4 and 3.4.5/4.4.5 Description of Change: This safety evaluation evaluated updating the charcoal sampling technique currently described in the Technical Specification Bases for Technical Specification 3.4.4/4.4.4, Emergency Ventilation System, and.Technical Specification 3.4.5/4 4.5, Control Room Air Treatment System. The collection method previously described in these Technical Specification. bases was not possible on the control room air treatment system, and was not practical for the Reactor Building emergency ventilation system. The change to the Technical Specification Bases allows for performance of alternate charcoal sampling techniques. Safety Evaluation Summary: Changing the collection technique to alternate methods endorsed by ANSI/ASME N510-1980 is within the licensing basis of the system. The proposed alternative techniques sample the charcoal beds with minimal disturbance of the filter media. This results in samples which are representative of the condition of the charcoal beds, thus ensuring that the test results accurately reflect the ability of the filter trains to remove the potential release of particulates from the air stream. This provides an accurate check of the efficiency of the charcoal filters. When the efficiencies of the filter trains are maintained as specified, the resulting doses will be less than the 10CFR100 guidelines for the accidents analyzed. 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-68 .;
Safety Evaluation No.: 96-012- Rev; 05 1 '$ ~NT' Implementation Document No.'-"""-""-Procedure Nl-TTP~"""""-"ao"-'. '>" -'-:: .- ',=':. UFSAR Affected Pages: "-:XI-9'r~ s<'-.=.= "i- s
~ ~ ~
System: Circulating Water System, Condenser Offgas, Condensate/Feedwater Title of Change: Sulfur Hexafluoride (SFo) Injection to Detect Condenser Tube Leaks Description of Change:. * +
~
R I
=I
~ 4 This safety evaluation evaluated injection of sulfur hexafluoride gas (SF,) and-helium into the circulating water and turbine building service water to locate condenser tube leaks or offgas vent cooler leaks. It was also dispersed in the vicinity of the main condenser to detect air in-leakage.
Safety Evaluation Summary: Sulfur hexafluoride, fluoride and helium do not have concentration limits for the reactor coolant since these chemicals are not normally expected and present in detectable concentrations. No adverse consequences are expected from the concentrations calculated in S1.1-74-F002. This calculation assumes a maximum usage of SF, of 250 SCF and a postulated tube leak of up to 5 gpm. Helium use up to 250 SCF is permitted. Should additional SF6 or helium be required, engineering shall be contacted to evaluate its use, Reactor water sulfate concentration action level 1 is 5 ppb. By calculation the expected increase in sulfates due to dissolution of SF6 will be less than 5 ppb. In addition, sulfates will be removed by the reactor water cleanup system. Feedwater and reactor water conductivity should be unaffected by the use of SF6 or helium and can be monitored during this test. Technical Specification limits for chlorides and conductivity shall still be monitored and adhered to. Conformance to NDD-CHE guidelines assures that intergranular stress corrosion cracking (IGSCC) is not increased by this test. Sulfur hexafluoride and helium, at the concentration expected, have a negligible impact on the production, moderation or absorption of neutrons. Reactivity will be unaffected by the presence of these chemicals. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation
- Summary Report-..
- ...Page 25of 68-':-
, Safety Evaluation No;: . '96-013:
Implementation Docuinent No.'.:""~ ','. DDC 1F00109.t;.'='-;:". UFSAR'Affected Pages: Figure X-8 System: Spent Fuel Pool Title of Change: Replace BV-54-70, 3" Chapman-Crane Gate Valve with 3" Worcester Controls Ball Valve
~ I" Description of Change:
The valve stem nut failed on suction valve BV-54-70 for the spent fuel pool filter precoat tank. The failure was assumed to be caused by resins being packed between the valve seats. When the valve did not close properly, the handle may have been over-tightened causing the stem nut to fail. I This change replaced the 3-inch, 150-pound flanged Chapman-Crane aluminum gate valve with a 3-inch, 150-pound flanged Worcester Controls stainless steel ball valve. This replacement valve bolted into the system without any piping or support changes. Safety Evaluation Summary: The function and operating characteristics of the system are unchanged. The gate valve and ball valves are fully ported and the flow characteristics are unchanged. The ball valve increases the weight at this location to 43 pounds, which is an insignificant change for the design of the piping. The ball valve meets or exceeds the design requirements of the spent fuel pool system. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
;Safety Evaluation 'Summary Report .Page'26 of'68.
~ Safety Evaluation:No:: -, - ., l96-01:4 -' Implementation:Document No.30l'OORDesfgn Change.N1-.9M)30;(< riolgp;ri~;;,- 'q:...
'-' '" B.i: IV-17;SIV-18, IV-23, IVe24<iVll.-.22;. Vil-23;;-
UFSAR Affected'Pages:
~ . -'. -', -- ." F '." VII~, XV-46, XV-47, XV-82; Figures IV-4, IV~
System: Control Rod Drive 1
~ t, ~ t 4 Title of Change: Use of Modified BWR-6 Original Equipment Control Blades at NMP1; l
Description of Change: 1 ~ ~ 4 r This safety evaluation evaluated the use of modified General Electric BWR-6 ~ Original Equipment control blades (M6CB) as standard replacement control blades at NMP1. These control blades were modified by replacing the existing rollers with rollers of correct diameter for use in the BWR-2 lattice at NMP1. This modification was performed'by General Electric. Browns Ferry Nuclear Plant (GE BWR-4), with a "D" lattice water gap dimension equal to NMP1, has operated 20 M6CBs in control cell core locations since June 1993. This safety evaluation also evaluated changing the NMP1 UFSAR maximum control rod drop velocity from 5 ft/sec to 3.11 ft/sec consistent with Technical Specification Basis 3.1.1.b.3. Safety Evaluation Summary: The M6CB nominal sheath thickness, absorber tube outside diameter, roller dimensions, and wing thickness are equivalent to those dimensions used in the Duralife 230 control blade. The Duralife 230 control blade was evaluated to ensure that it could be inserted during normal, abnormal, emergency and faulted modes of operation within the limits assumed in the plant analyses. The analyses considered the effects of manufacturing tolerances, swelling and irradiation growth and includes the time-dependent effects of corrosion. The Duralife 230 control blade was approved for use in a BWR-2 by the NRC and several are currently in use at NMP1. Additionally, the weight of the M6CB is equivalent to the BWR-2 Original Equipment control blade design. Therefore, the mechanical performance of the M6CB will not differ from control blades currently used at NMP1. The M6CB control blades have approximately the same hot and cold reactivity worth as the BWR-2 Original Equipment control blade (matched worth). Therefore, the M6CB has the same nuclear performance properties as blades currently installed in NMP1.
.Safety Evaluation
- Summary Report
.Page 27 of 68 Safety Evaluation No.: :96-014 (cont'd.) Safety Evaluation Summary: -": " '. -l:(cont'd;:) '.=.-" .<':-"~"...'" ".:-"'.-', 'i:-"::-v': "-:. Based on'the evaluation performed, it'iswoncluded that these changes do not .:, involve an unreviewed safety question.
-".- Safety Evaluation
.;.Summary Report " Page 28 of;68 ='=-
Safety.Evaluation No.: :>>96-015 ~ > 4 r'5 ge" ~
~ ~
I"Implementation Document', No.: ,':.Procedure EPMP;EPP;,02~;3 no:qi.;.:.~..=:- -.-",-'r UFSAR Affected Pages:,:: .:~,>..10A-1?.,':.-,-,:-:pq,-0,'=;~-..:-.'=
~
+ p<~a ~ \ytr<e 'i +a+ ' ->a
>t'g ~
>>y i ,; ~
System: N/A Title of Change: Description of Fire Brigade Equipment Location in Unit 1 FSAR -
~ ".-
Description of Change: Appendix 10A (Fire Hazards Analysis) of the Unit 1 UFSAR listed areas within the plant where firefighting equipment is stored. Specifically, the UFSAR identified locations in the Turbine, Reactor, Offgas, and Administration Buildings, as well as the Unit 1 Maintenance Shop, as storage locations for firefighting equipment. This change removed reference to these specific locations from the UFSAR,.thus allowing the Fire Brigade more flexibility in choosing the best storage location for firefighting equipment. Safety Evaluation Summary: ln accordance with industry codes, standards, and guidelines, references to. specific plant locations regarding storage of firefighting equipment has been from the Unit 1 UFSAR. Appendix 10A of the UFSAR provides specific 'emoved equipment locations in detail far exceeding the industry norm. Equipment inventory and locations are administratively controlled via approved NMPC Procedure EPMP-EPP-02, "Emergency Equipment Inventories and Checklists." The change provides the Fire Brigade with more flexibility in choosing the best storage location for firefighting equipment based on improved firefighting technology, training and site-run drills, and site-specific knowledge. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report
..Page 29 of-68 ,. Safety Evaluation No.: -.~ 96.-016/
~:; - Implementation Document No; ~ NFPA 16, DER.1-95;2856,
'-UFSAR Affected Pages: ~ 10A-51, 10A-52, 10B-78 System: Foam-Water Title of Change: Clarification of Foam-Water Fire Suppression System Arrangement Description of Change:
The foam-water system at NMP1 provides protection around, the turbine generator in the event of an oil fire. Six foam-water deluge spray systems exist as follows: four protect the Turbine Building El. 300 area around the turbine, and one each in the turbine oil reservoir and hydrogen seal oil rooms. The water portion of the four turbine area systems is automatically initiated by cross-zoned thermal detection. Actuation of these open head deluge systems provides WATER ONLY to the covered areas. While detector actuation opens the supply motor-operated valve (foam and water) to these lines, the foam pump must be manually started in order to get foam concentrate injection. This mode of operation is in compliance with National Fire Protection Association Code 16 (NFPA '16), "Installation of Deluge Foam-Water Sprinkler and Foam-Water Spray System," and is per the. original system design. Discrepancies existed between the system description sections of the Unit 1 UFSAR and NRC Safety Evaluation Report (SER) regarding automatic vs. manual starting of the foam injection pumps. These discrepancies were minor in nature and did not affect the Fire Protection Program at NMP1. The NMP1 UFSAR has been revised to indicate the foam injection pumps can only be started manually, and the NRC SER discrepancies have been identified and discussed. Safety Evaluation Summary: The proposed changes clarify and update the UFSAR and reconcile the UFSAR and NRC SER. The changes are strictly editorial in nature and reflect what has always been the design basis for the foam-water system. This clarification and reconciliation have no physical effect on any plant structure, system or component, or any design basis or accident. This update will clarify the method and mode of operation of the NMP1 foam-water system as described in the Fire Hazard Analysis and Safe Shutdown Analysis. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
;Safety Evaluation Nummary. Report Page 30 of 68
...-Safety Evaluation No.: 9&417 0 ~
-'Implementation Document No.: -"~<" l '"GER "1-96-0738o<"'">r <."-" - '-':"-'-'r '>" "'"-'-"-
UFSAR Affected Pages System: Screen Wash Title of Change: Closure of Standby Screen Wash Pump ~ Intertie Valves Description of Change: ~,I I 41 i wI\ %P iThe configuration of the header'intertie valves for the standby screen wash pump was changed from open to closed. The position change was reqvested to reduce-backflow through pump 13 and avoid.simvltaneous feed of the vpper and lower screen wash headers'in the event pump 13 initiated. Screen wash pump 13 is a standby pump used as a backup to either pump '11 or 12. Safety Evaluation Summary: Since the intertie valves are manual, isolation of upstream eqvipment can be obtained as necessary by closing the valves. With the valves normally closed, backflow throvgh pump 13 is prevented, assuring full flow to the screens from pump 11 and 12 and reducing the potential for damage to pump 13. Closing both intertie valves will require manual action to open either the upper header valve or the lower header valve before putting the pump in service. This is preferable to running with the valves open since: 1) running with the valves open causes recirculation of flow from pump 11 or 12, resulting in less flow to the screens and potential damage to seals; and 2) the development of differential pressure across the screen is not expected to occur rapidly (by engineering judgment and operating experience), allowing sufficient time for operator action. 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 68
~,.',,; ..- Safety Evaluatiori No'.: " ' ""96-'018 Rev. 0 & 1 Implementation Document No.: 'i Mod.:N1-94-003
'= UFSAR:Affecte'd'Pagesr -,'~'IV-29,g XVI-12,g XVI-14;g~Table XVI-9a
~
System: Reactor Vessel Title of Change: Modification to the Core Shroud Repair Tie Rod Assemblies " . Description of Change:
~ = 4 The as-built configuration of the lower spring contact on each of the four core shroud repair stabilizers (tie rods) did not encompass shroud weld H6A as was intended by the original shroud repair design. This change modified the lower spring contact to extend beyond the H6A weld. This modification restored the contact to its intended design condition. Also, the lower spring of the stabilizer was bearing on the blend radius of a recirculation nozzle. An 270'zimuth additional change replaced the 270'ie rod and spring assembly without having a spring on the opposite side of the tie rod. This modification relocated the spring to bear on the reactor pressure vessel as intended.
During RFO14, clearance was found between the toggle bolts and the shroud support cone that could affect the axial tightness of the stabilizer assemblies.'he clearance between the toggle bolts and the shroud support cone was removed, restoring the stabilizer assemblies to their originally intended design. The lower wedge latches had the potential to become loaded due to differential vertical displacement greater than intended by the original design of the latches. New modified latches were installed which are more tolerant of differential vertical displacement. Safety Evaluation Summary: GE Safety Evaluation GE-NE-B13-01739-5 and NMPC Safety Evaluation 94-080 evaluated the design, fabrication and construction of the core shroud stabilizers at NMP1, 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 Boiling Water Reactor Vessel & Internals Project (BWRVIP) Core Shroud Repair Design Criteria. The shroud repair design analyses were also reviewed and approved by the NRC as documented in the Commission's safety evaluation report (SER) dated March 31, 1995; however, the NRC SER required
...>i Safety Evaluation
- .-.,'='- Summary. Report
=,.::::Page 32'of 68 j
...>'"Safety=Evaluation No.:
."~ -r~J 96-..018 Rev. 0 h1 (cont'd.), -'.
I i1 i (t%.
""..-"Safety EvatuatIon Summary: >'-::"'-'E.-" '(cont'd:) =-' ~c>>
Oe+ ~
'L 's P> 4<8~
$ 1% 4VJ Pf4 ~
0>> I'
$ ~
<'" 'that correctly" actions be implemented to address the lack of coverage of weld ==:,.
H6A. The NRC provided NMPC with a SER on March 3, 1997, which approved
'the modifications to capture weld H6A and to remove the'lower wedge from the-recirculation nozzle.
The NMP1 repair modification of the core shroud was performed as an alternative to ASME Section Xl as permitted by 10CFR50.55a(a)(3). Consequently, NRC approval of this repair approach was required. The BWRVIP Report (EPRI .; TR-105692, BWRVIP-04), entitled Guide for Format and Content of Core Shroud
~
-'epair Design Submittals," requires that a safety evaluation-of core shroud repairs .'
be made and that the conclusions be provided to the NRC:. This safety evaluation documents. the NMPC review of the repair in accordance with. the provisions of 10CFR50.59. The evaluation included a review of the plant licensing'bases.'- The evaluation demonstrates that the proposed modifications can be implemented 1) without an increase in the probability or consequences of an accident or malfunction previously evaluated, 2) without creating the possibility of an accident or malfunction of a new or different kind from any previously evaluated, and 3) without reducing the margin of safety. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary-Report Page 33 of 68
, -Safety-Evaluation No.: 96-'039 c.*k>. ~'
Implementation Document No.:, "..~. ':Site~Emergency.Plan.o;.".& ~9r."" ~".:e .".~ r;"->>,:; ~
~ I d 'r A p~
g y$ UFSAR Affected Pages: 'Xlll-13 P System: Emergency Operations Facility (EOF) Title of Change: Emergency Operations Facility (EOF) Move From the Nuclear Learning Center at 9 Mile Point to the Existing EOF on Route 176 in Fulton, New York Description of Change: The EOF is a support facility for the management of overall licensee emergency response, coordination of radiological and environmental assessments, and determination of recommended public protective actions. The EOF is equipped with administrative, communication, and computer equipment that meet the requirements of license basis documents including NUREG-0696, Site Emergency Plan (SEP), Unit 1 UFSAR, Unit 2 USAR, and Technical Specifications. The EOF has been relocated from the Nuclear Learning Center (NLC) to a new facility located on Route 176 by the Oswego County Airport in Fulton, New York, approximately 11 miles from Nine Mile Point. The new location is also used as the New York Power Authority EOF. Safety Evaluation Summary: Relocation of the EOF will satisfy the NRC recommendation that the EOF be located outside the 10 Mile Emergency Planning Zone (EPZ). This will also eliminate the need for NMPC to maintain an Alternate EOF outside the 10 Mile EPZ. The EOF located at the NLC does not provide plant control functions and is not connected to any system used to mitigate an accident. The EOF operates in accordance with design configuration and site procedures to comply with NUREG-0696, SEP, Unit 1 UFSAR and Unit 2 USAR. Changes to the SEP and Unit 1 UFSAR, as a result of relocating the EOF, will not affect any plant system used to mitigate an accident or any system associated with accidents previously analyzed. 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 68
.'=Safety Evaluation =No.: - 96-021 i) vy
~ 0 <<)
Document No.: vi:;,~~.'alculations'-S7-RX340-W01,,-,~i: ~,=;~'>>,; .-",.
-.4Implementation S4RX340BLDG01, S4TB300BLDG01'
~ I UFSAR Affected Pages: lll-3, VI-17, XVI-70; Table XVI-31 Sh 1 N
* ~ ~ ~ t <<
System: N/A Title of Change: UFSAR Changes for Reactor Building and Turbine Building Pressure Relief Panel Failure Loads. Description of Change: The UFSAR has been revised to show the new blowout:load.of Turbine Building (TB) pressure relief panels as 62 psf, new wall panel area of 1900 sq. ft., and the failure load of superstructures as at least 135 psf. This change also shows the new blowout load of Reactor Building (RB) pressure relief panels as 65 psf, new walt panel area of 2400 sq. ft., and the failure load of superstructures as at least 117 psf (internal pressure). The UFSAR has also been revised to indicate the ratio of relief area to building volume as 1.6 ft'/1000 ft'or the Reactor Building and 0.21 ft /1000 ft'or the Turbine Building. Safety Evaluation Summary: The failure blowout pressures (internal pressure) of 65 psf (RB) and 62 psf (TB) are sufficiently <117 psf (RB) and <135 psf (TB) and provide adequate protection of Reactor Building/Turbine Building superstructures against internal pressure, where 117 psf and 135 psf are the minimum internal pressures that should reach inside the Reactor Building and Turbine Building superstructures, respectively, for failure, as documented in Calculations S4RX340BLDG01 and S4TB300BLDG01. The blowout panels have been returned to a configuration functionally equivalent to the original design, i.e., 3/16" diameter bolts spaced at 12" O.C. have the equivalent strength of 1/4" diameter bolts spaced at 24" O.C., with the same tensile strength. 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 68
. Safety'Evaluation
~ I No.: 96-023 iie' pA. >'w "ry,, it ~
;.'mplementation Document No.::"-".-"'~~'"DER.1.-.95-3438 J me
=
'-""-.UFSAR.-Affected Pages: ""
.- X-24
=
System: Service Water (SW) Title of Change: Service Water Strainers Mesh Size Discrepancy . Description of Change: The. UFSAR previously stated that each SW pump was provided with a .010-inch, .
. mesh automatic self-cleaning strainer. Although the initial mesh size chosen for these strainers was .01 inch,.due to frequent clogging'of the strainers,"the mesh size was changed to .03 inch; This change provides clarification in the UFSAR of the SW strainer mesh size to conform to the as-built condition of the strainer.
Safety Evaluation Summary: There is no defined industry criteria for the selection of strainer mesh sizes. The decision regarding the size is primarily based on past experience and engineering judgment. Factors such as amount and size of particulate matter in the fluid, flow velocities in piping and components, and propensity of any equipment to develop clogging, normally forms the basis for engineering judgment regarding selection of the strainer mesh size. The present installed mesh size of .03 inch on the normal SW pump strainers is of appropriate design and does not adversely impact nuclear safety. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
.'afety Evaluation ;,,Summary Report- , Page 36 of 68=
.- Safety Evaluation No':, implementation Document No:: 4..>> ~ .PCE.to Procedure.N1-OP-6;:. r " -: ., =, Calculation S14-54-HX08 f 4+~ Ag ~ 'll p
~
Affected Pages: X-33
]a~'FSAR System: Spent Fuel Pool Cooling (SFC)
P Title of Change: Securing the Spent Fuel Storage Pool Filtering and Cooling System for Maintenance
.~
Description of Change:
~ t
- 4 ~ ~ N ~~
This safety evaluation evaluated changes to Procedure.N1-OP-06, Spent Fuel -'. Storage Pool Filtering and Cooling System, to allow securing spent fuel cooling for maintenance, provided SFC temperatures are alternately monitored and controlled below 125 F, and incorporated a description of spectacle flanges downstream of the heat exchangers which allow independent isolation of the SFC subsystems. There are common components in the system. Further, the effluent of each of the redundant cooling and filtering trains is bounded by a check valve and a spectacle flange. The system must be secured to do corrective maintenance on a common component. The system must also be secured for a short period for maintenance on each redundant train to allow time to reverse the spectacle flange, because the associated discharge check valve cannot be considered a secure boundary for personnel safety. According to the UFSAR, the SFC system must maintain the pool temperature below 125'F and maintain acceptable water clarity. This safety evaluation considered power operation, not refueling outages; therefore, reactor cavity and equipment storage pit level functions are unaffected. Safety Evaluation Summary: The SFC system may be secured for a limited time for maintenance on common components, or components which require securing common components for personnel safety. During this period, the pool temperature will be monitored so the temperature of the pool will not exceed the design limit of 125 F. Evaporative, radiative, and conductive heat losses are not considered in Calculation S14-54-HX08. These heat losses are not expected to actually cool the pool; therefore, pool temperature will remain above 68 F and K~ for the high-density racks will remain <0.95. K, in the low-density racks increases with
,'-.'.:Safety EvaluatIon
.";; Summary Report i-.-Page'.37'of 68
...-':". Safety. Evaluatioii No.: 96-'1 04 (cont'd.) '<<I
':-: "- '=='-: (cont'd.)
Safety Evaluation Summary: but is <0.91 at 125'F and so meets the <0.95 criterion;;;; ',.
'temperature; Evaporative inventory losses are considered negligible; however, the fire and condensate transfer systems will be available as makeup water supplies while the system is secured. The proposed maintenance on the SFC system will have no bearing on other equipment important to safety and, specifically, will have no effect on the RBEV system, which is 'necessary to mitigate the effects of the most relevant analyzed accident, a dropped fuel bundle.
Securing the SFC system for a limited period can be accomplished while remaining within the design limit of 125'F and will ensure there is no negative. effect on other equipment important to safety. Based on the 'evaluation performed; it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report-Page 38 of 68 ... Safety, Evaluation No.: .: .."~-. 96-196. Implementation Document No.: ~ Temporary Mod. 96-008,, UFSAR Affected Pages: . N/A. ~ r P
~
g System: Main Turbine, Feedwater Title of Change: Plant- Operation with Feedwater.Pump 13 Stub Shaft Uncoupled Description of Change: 1,J ~ >I ~ s>,S =
~q I'>> ~ <<y = iG 1 )>) ~ ~ p g ~ps ~
The ¹1:3 feedwater pump ls mechantcally connected to, and driven. by, the, main .:-., turbine generator. The mechanical. connection includes-a clutch assembly .. comprised of a. fluid friction clutch and,a geared (dental) clutch which work in parallel. Damage was sustained to the dental clutch and removal for repair of the rotating element was required. This temporary modification installed a stub shaft as a replacement part within the clutch housing. The stub shaft is coupled to the turbine at the shear shaft and may be coupled to the ¹13 feedwater pump gear set at a later date. Safety Evaluation Summary: installation of the stub shaft in lieu of the clutch rotating element is an original feature of the clutch in the event of major mechanical failure. The shaft is 'esign capable of transmitting 10,000 hp at 1800 rpm from the main turbine through the clutch housing to the ¹13 feedwater pump step-up gear. The input end of the shaft is equipped with a coupling flange to mate to the shear shaft at the turbine. The output end of the stub shaft is suitable for mounting the existing Thomas flexible half coupling. This mounting is a shrink fit. This type of mounting assures the coupling will not detach from the stub shaft at rated speed. The Thomas coupling between the clutch and step-up gear is removed to defeat operation of the ¹13 feedwater pump. Removal of the coupling does not pose a safety hazard, as the housing cover will be installed as designed. The thrust bearing in the clutch ensures stability of the free coupling hub. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
à Safety Evaluation Summary Report Page 39 of 68 97-.'901 '".
..=,,, Safety Evaluation No.: ".
E
~ ~
Implementation Document No.'.. '",.'.Simple"Design Change SC1-0122-92.,'..-,= UFSAR Affected Pages: V-5:~8 ~~
~
~
' A System: Floor Drains, Equipment Drains Title of Change: DWEDT Level Instrument Upgrade Description of Change:.
This simple design change installed new level sensors for the drywell equipment and floor drain tanks that will provide input to new programmable logic controllers (PLC), which will calculate. the rate of rise of water in the floor drain tanks and perform the alarm function based on that rate. This safety evaluation also evaluated the changes required to the existing PLCs installed in drywell leak detection cabinets A & B. Additional circuit boards were installed to accommodate the signals supplied by the new level sensors and to provide output signals to the Control Room chart recorders. Software changes were required to support the new hardware and functions. All changes were transparent to Control Room operations. Safety Evaluation Summary: The excessive leakage detection system provides the Control Room with an annunciator warning of an incipient reactor coolant system (RCS) failure. This is determined by the measurement of identified and unidentified leakage inside the drywell. This leakage is collected in tanks where level changes are used to determine the rate of RCS leakage. An annunciator is alarmed if the rate of leakage exceeds limits set in Technical Specification 3.2.5. A secondary function is control of the tank sump pumps. The present system consists of many original plant components that are at or near the end of useful life. The method of calculating the rate of rise in tank level will differ slightly from the original method, but full conformance with all Technical Specification requirements is demonstrated in this safety evaluation. The change will have no effect on the safe operation or shutdown of the plant. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report-Page 40 of 68-97-.'002;= ' t Safety Evaluation-No.: . py C' 0
~
~ p ~
Implementation Document No.:,"..>~;. !:DER'1=96-2795.'oM
~ jgp%h, ~E. ~t fl.-es~ L9)PI %.
c
~ ..n.a UFSAR Affected Pages: Vl-32 -'
System: Reactor Building Normal Ventilation Title of Change: Reactor Building Normal Ventilation Intake and Exhaust Local Flow Indication
'~- Cr~-
IW 4" W 14 Description of Change: I ~ *~ This change updated UFSAR Section VI-F.5.1 to indicate that flow switches in:the-supply and exhaust lines provide for low flow alarms in the Control Room for the . reactor building normal ventilation system flow. Previously, the UFSAR indicated: local flow rate indication was provided in the supply and exhaust lines. Neither . the current design nor the original completed plant design provide for this flow rate indication. The flow indication was removed during plant construction. Flow rate indication is only required for the emergency ventilation system. Reactor building normal ventilation system including flow indication/alarm is not safety related. Safety Evaluation Summary: Local flow indication was originally discussed in the FSAR because the original plant design once included local flow indication in both the intake and exhaust of the reactor building normal ventilation system. The indication was removed before original construction was completed. Flow indications were also added to the emergency ventilation system such that both trains of the system would have flow monitoring capability, while the local flow indications for the normal reactor ventilation system were removed. No justification or documented modification was found for the removal of the flow indications. The final as-built design did not include them and there is no evidence that local flow indication was ever actually installed in 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 41 of 68 Safety Evaluation No.: 97-003
\
Implementation Document No.: Procedure NIP-FFD-02,:; ~ ~ kP~ ~
"p " ~ 'h
~
l
~
~a t liglv ~
~ ~
Affected Pages: 'FSAR . N/A =;:- 4 System: N/A " ~ Title of Change: Change to NIP-FFD-02 Which Extends Respirator Physicals to Once Per 2 Years for . Select Groups of Personnel Description of Change: This change revised the UFSAR to reflect the 10CFR Part 20 changes made in February 1995 regarding respirator qualifications. It is now required that respirator qualifications include a physician's determination prior to initial fitting of respirators and periodically at a frequency determined by a physician that the individual is medically fit to use the respiratory protection equipment. Safety Evaluation Summary: The changes to NIP-FFD-02 are based on the current regulations of 10CFR Part 20 as prescribed by the company physician. These changes meet or exceed all current requirements for respirator qualification physical frequency. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
,Safety Evaluation
.Summary Report
==Page 42-of 68
. '. -.Safety EvaluatIon No.: .97-005 Implementation Document No:: -:-" < =.! NEDE '24011-'P-'A-"I 0'~ ~"<;>""'~ A4l>~" >~:<"-i;" ~ NEDE-24011-P.-A-10-US {GESTAR II) j RNc i 5 9Js~iB jig ~4 lw4l'4 UFSAR.Affected Pages I-10, I-15; IV-7, IV-12, IV-32, V-2'I,'VII-20, XV-3, XV-5, XV-6, XV-7, XV-13, XV-15,-: XV-68, XV-79, XV-82; Table V-1 Sh 2 System: Various-Title of Change: Operation of NMP1 Reload 14/Cycle 13 ll ~ Y ~ 1 Description of Change: . This change consisted of 1he addition of new fuel bundles and the establishment of a new core loading pattern for Reload 14/Cycle 13 operation of NMP1. One hundred eighty eight (188) new fuel b'undies of the GE11 design were loaded. 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: r 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 14, 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 NMP1 can be safely operated during Reload 14/Cycle 13 and that this change does not involve an unreviewed safety question.
I ~ Safety 'Evaluation Summary Report Page 43 of 68
.,:.=- -Safety Evaluation No.: 97-006 I
'. Implementation Document No..'.."." '" DER 1-.'96-2971....:i';"; <.... -,~ iii.;i~<
UFSAR Affected Pages: ~ XI-11 Feedwater System'."'itle of Change: Shaft- and Motor-Driven Feedwater Pump Capacities Description of Change: This change updated UFSAR Section XI-B.9.0 to change the stated capacity of the shaft-driven feedwater pump from 6,400,000 Ib/hr to 5,500,000 Ib/hr; and to change the stated capacity for the motor-driven pumps from 1,900,000 Ib/hr to 1,250,000 Ib/hr. These values were incorrectly changed in UFSAR Rev. 0. Safety Evaluation Summary: The proposed changes make the UFSAR consistent with the as-built plant feedwater pump capacities. The proposed changes do not increase the probability of occurrence of an accident previously evaluated in the UFSAR, since high-pressure coolant injection (HPCI) system performance was based on the as-built capacities of the motor-driven pumps. The shaft-driven pump does not perform a HPCI function; therefore, the change to the shaft-driven pump rating has no impact on HPCI performance. Further, HPCI is not an engineered safeguards system and is not considered in any loss-of-coolant accident analyses. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Siimmary Report Page 44 of 68 Safety Evaluation No.: 97-'007. - i '-.~ qi'.S Implementation Document No;: DER 1 96-3180'=: UFSAR A'ffected Pag'es Vll-'7
%,l ~
System: Core Spray (CSS) Title of Change: Core Spray System Pump and Valve Testing 4 Description of Change: This change updated UFSAR Section VII-A.4.0 to delete the reference to testing of the CSS'ump and valve shaft seals by applying pressure to a lantern ring between sections of packing and visually observing leakage. -Testing of the core spray pump and valve shaft seals was never performed in the manner previously described in the UFSAR. Safety Evaluation Summary: Testing of the CSS pump and valve seals is governed by Technical Specification 4.2.6, "ISI/IST," and 6.14, "Systems Integrity," and their respective implementing programs (Second Ten-Year Pressure Testing Program Plan and Leakage Reduction Program). These testing requirements have been reviewed and determined . adequate by the NRC. The proposed change to the UFSAR will result in a more accurate description of actual CSS pump and valve testing. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Reyort Page 45 of 68 ,, Safety Evaluation No.:
- 97-008 Implementation Document No.:"..." ~ DER 1-97-0002 .;;"
UFSAR Affected Pages: VII-2: I System: Core Spray (CSS) Title of Change: Core Spray System Design Pressures... Description of Change: This change updated UFSAR Section Vll-A.2.1 to correct the CSS equipment and piping design pressures to reflect the original design specifications and as-built construction of the system. The design pressure of'CSS equipment and piping between the suppression chamber and the topping pumps has been. changed from 340 psig to 310 psig. The design pressure of CSS equipment and piping from the suction of the topping pump has been changed from 465 psig to 470 psig, and clarified to indicate after the topping pump. The UFSAR has also been revised to clarify that the core spray pump motor cooling coils are designed to 100 psig. Safety Evaluation Summary: The primary function of the CSS is accident mitigation. The system is not identified in the UFSAR as an initiator to any of the accidents described in the UFSAR. The proposed changes will correct the UFSAR CSS equipment and design pressures to make them consistent with original design specifications and as-built construction of the CSS. Therefore, the proposed changes do not increase the probability of occurrence of an accident 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 46 of 68
...,. Safety Evaluation No.: 97.-01$ q ." ... Implementation Document No..'."'" ";. DDC:.IM00336.;p '; ..;"~;;; .. UFSAR Affected Pages: Figure 1V-7 ~ 4 l t System: Control Rod Drive (CRD) . Title of Change: FSAR Update for Change in CRD internals . Description of Change: 4 The control rod drive mechanism (CRDM) is used to rapidly insert (scram) the control rods in response to manual. or-automatic signals from the-reactor protection system (RPS). The CRDM is also used to change the position of the control rods
~
within the core in response to the reactor manual control system for the control of reactivity. The CRDMs are provided by General Electric, the original equipment manufacturer. This safety evaluation evaluated a redesign of the inner filter and spud; a change in material to XG-M stainless steel for the construction of the index tube and piston tube assemblies; a change in design of the uncoupling rod and 0-ring spacer; and an upgrade to a multi-port cooling water orifice. These changes were made to improve reliability of the CRD and minimize CRD installation errors. Safety Evaluation Summary: These changes were made to the CRD by General Electric to incorporate plant operating and maintenance experience. The changes do not adversely affect the ability of the CRD to scram the reactor in response to signals from the RPS, nor do they adversely affect the ability of the CRDs to control reactivity. The results of these changes included increased CRD reliability and minimized installation errors after CRD refurbishment, such that there is continued assurance that the CRD will continue to be able to perform these design functions. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 47 of 68 Safety Evaluadon No.: 97-015 Implementation Documerit No..6" 8!: DER'.1'-'96-1894 "i';:-.. -. ~
~:. o.;-,. :. "~ L t' r p}m rg~
p UFSAR Affected Pages: Vill-38 System: Rod Worth Minimizer (RWM) Tide of Change: Revision to UFSAR Section Vill, Description of RWM Description of Change: ~ ~ t ~ This change revised the UFSAR to agree with the as-built plant; The UFSAR, in describing the function of the bypassing of RWM control above the reactor power level called the "low power setpoint," previously stated that only feedwater flow provides the low power setpoint trip, whereas both feedwater flow and steam flow provide redundant inputs to the RWM as indirect measurements of reactor power. On decreasing power, either the steam flow input or the feedwater flow input will trip to low power setpoint above 20% reactor power to enable the RWM. On increasing power, both steam flow and feedwater flow inputs are required to disable the RWM above the low power setpoint. After the low power setpoint has been exceeded, the RWM does not inhibit rod selection or movement. Safety Evaluation Summary: The RWM system supplements procedural controls to prevent an inadvertent control rod drop accident. The proposed change only corrects the UFSAR description of the inputs to the RWM and does not change the design function of the system. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation. Summary'Report Page 48 of 68=-- -- Safety Evaluation No.: 97:016' "I Implementation 'Document No;i '~": f -B'DERs'f-96-2933j 1~9'6-2947~ 4-96-.2948,;~.-... 1-96-2949 t( <~v .~ > ~ ~ j;y UFSAR Affected Pages: X-7, X-8, X-'!0, X-11, X-52; Figure X-3 System: Control Rod Drive (CRD) 5 Title of Change: CRD System UFSAR Changes Description of Change: The UFSAR has been revised as=-follows: w4 ~ a I ., Section X-C.2.1 has been revised to state: "One pump is rated et 85 gpm at a head of 3,760 ft.'ith a 250 HP motor." A sentence.has been added to read: "The other is rated at 87 gpm at a head of 3,740 ft. with a 250 HP motor." Section X-C.2.2 has been revised to state: "The two parallel filters will remove 99 percent of foreign material larger than 40 microns from the hydraulic system water." g A Nl Sections X-C.2.0 and X-C.2.4 have been revised to indicate the second-stage pressure is maintained at approximately "250-270" psi above reactor pressure. Section X-C.2.10 has been revised to state: "The scram dump volume has a capacity to accommodate a free volume of 3.34 gal. per drive up to an in-leakage of approximately 0.5 gpm per drive. A sentence has been added to state: "For an in-leakage of greater than 0.5 gpm per drive, the free volume will fall below 3.34 gallons per drive; however, the system function will be maintained." Safety Evaluation Summary: The CRD system is not identified as an initiator of any transients or accident previously evaluated in the SAR. The CRD pumps are not designated as an element of the emergency core cooling system (ECCS), even though they may aid in mitigation of small high-pressure line breaks. The proposed changes will not impact CRD performance, and will provide licensing basis consistency with the
Safety, Evaluation = Summary Report Page 49 of 68.. Safety Evaluation No.: 97.-'0'l6'(cont'-d.) "-" ~"...;.>, r ~-., Safety Evaluation,Summary: '-.. " ' (cont d.)'- ',;: ~:;.'n'<:;,-*a-.-, .,-;-~ .;. ~ 1 as-built design.' Therefore, the'.proposed'changes <do,not~increase the:probability;- * '.- of occurrence of an accident previously evaluated in the SAR. ~ ' a Based on the evaluation performed, it is concluded that this change does not-an unreviewed safety question. 'nvolve a ~
~ ~
~
Safety. Evaluation.=-: Sumrrlary Report Page 50-of 68" Evaluation'No.: ; .." > .:: . '97-'018 -'a'fety
~ 0 4 ~ 4" +
Implementation Document No.: . I Alod."N1-97-005':4;.< ..",rli:<c. v> r:".",.i!e;, 9 '~g.: .: UFSAR 'Affected'Pages:.' :: " ." Vl-'2G," Vl-'25;:Table Vl-3a Sh 2 8c 3; .
; Figure Vl-22 System: Shutdown Cooling (SDC), Postaccident Sampling (PASS)
Title of Change: Addition of Thermal Overpressure Protection on Penetrations X-7, X-8 and X-139 Description of Change: This change added a rupture disk to PASS penetration X-139. The rupture disk discharges into an enclosed expansion chamber located outside primary containment. The expansion chamber is attached to existing support steel and piped into the cavity between isolation valves 110-127 and 110-128. The expansion chamber is flanged to accommodate periodic replacement of the rupture disk. The new valve between the rupture disk and the process piping was locked open after installation was complete. Overpressure protection of the SDC penetrations was provided by adding a bypass line with a flow restricting orifice and a check valve. The new line is used to vent fluid from the isolated penetrations to the upstream side of inboard isolation valve 38-01. The seal ties penetrations X-7 and X-8 together via the common seal piping. SDC'ater This allows the use of a single bypass loop to accommodate thermal expansion in both penetrations. The use of a single bypass loop minimizes the loss of seal water through the line. The flow restricting orifice is sized to: 1) pass the flow rate required to offset thermal expansion in both SDC penetrations, 2) maintain the integrity of the SDC water seal, and 3) pass the largest expected debris to preclude plugging. A check valve is installed in the bypass loop to maintain containment and reactor coolant isolation. The bypass loop is flanged to allow removal for decontamination. Safety Evaluation Summary: This modification provides overpressure protection for penetrations X-7, X-8 and X-139. Containment and reactor coolant isolation is still maintained for the SDC bypass line via a check valve. These modifications insure that proper thermal relief is provided as required by Generic Letter 96-06. Appendix J and Section XI requirements are instituted into the physical design of the two changes. The PASS and SDC system configurations meet or exceed the design criteria for the existing systems and the reactor coolant system.
Safety Evaluation Summary Report Page 51 of 68 Safety Evaluation No.: 97-018 (cont'd.)
~ 1 P
Safety Eva'luatlon Summary: (cont'd.) j ' e Based on the evaluation performed, it is concluded that this change does not involve an Unreviewed safety question. ~
.:...Safety Evaluation
'- Summary Report Page 52of 68
.: .! Safety Evaluatlbn No.: -. ~ 97-019 Rev. 0 L1
' gg1ic
.4 v'i +f j "t 'Implementation Document No.: 'rocedure S-MMP-GEN-014 "ye > vg
~~ I ~ \ ~ q . L ~ 4 "1 $ (, ~ g
- l" UFSAR Affected Pages: "s" N/A
~ '
Z
- System: Reactor Water Cleanup (RWCU)
Title of Change: .....-installation of Freeze Seal For IV 33-01R or IV 33-02R. Description of Change: This temporary change installed freeze seals on sections of RWCU piping to assist in completion of the testing and repair of IV 33-01R and IV 33-02R in the reactor vessel and reactor recirculation loop ¹11, respectively. Revision 1 of this safety evaluation clarified that carbon steel pipe is brittle below-40'F. Safety Evaluation Summary: The proposed activity will be performed during RFO14 when the reactor head is removed and the entire reactor core will be offloaded to the spent fuel pool., With the fuel offloaded and the inner or the outer spent fuel pool gate closed, the fuel is sufficiently protected and cannot be uncovered. Additionally, freeze seals have been shown to be effective up to 10,000 psi. Since the fuel is safeguarded and freeze seals have been proven reliable, the probability of fuel damage due to a loss of water inventory is not increased. Secondary Containment will be available and implemented if required in accordance with Technical Specifications. Although containment isolation is not in effect for this work, maintenance of the water inventory in the reactor cavity, internals storage pit, and spent fuel pool is necessary. The freeze seals will perform the vessel isolation function while installed. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation =-
'Summary Report ~
Page 63 of 68
-Safety Evaluation No.: -,97-022
~Implementation Document No.:
7 4
'..: Mod. N1-97-012=
'UFSAR Affected Pages: "Vll-42 System: Fee dwater/HPCI Title of Change: RPV Overfill Prevention Backup Time Delay Trip of HPCI Pumps Description of Change:
This modification installed backup time delay relays in the breaker trip. circuitry of. the high-pressure coolant injection (HPCI) pump motors. This provides a trip of the HPCI pumps if reactor pressure vessel (RPV) level is sustained above 95 inches. Safety Evaluation Summary: The new/additional trip logic has a delay, which is set in accordance with the analysis documented in Calculation S22.1-XX-G025NF, to prevent RPV overfill. The new trip logic will not be interlocked with the flow control valve position switches. Therefore, an improperly adjusted or faulty valve position switch will not prevent a trip of the motor-driven feedwater pump if RPV water level is sustained above 95 inches. The existing trip logic, including the high level reset logic, will not be altered. 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 68 "."=...Safety Evaluation No.: -=. 97.-'025 Rev. 1 J~
I t 'i~* Implementation.Document No.':.".. '==": I GE-NE-523-B13-01869-043 Rev. 0;; .:;.'.. - ~ GE-NE-523-113-0894 Rev. 1, BWRVlP-07 4 JI + t UFSAR Affected Pages: IV-25, IV-26, IV-32
~ I System: Reactor Vessel Internals Title of Change: Core Shroud Vertical Weld Cracking Description of Change:
Inspection of the core shroud vertical welds identified intergranular stress corrosion cracking (IGSCC) of the-vertical welds. The inspections revealed fairly significant cracking on welds V-4, V-9,.and V-10; relatively minor cracking on welds V-3, V-12, V-15 and V-16; no cracking on the accessible portions of V-7, V-8, and V-11. Safety Evaluation Summary: The vertical weld cracking has been analyzed and determined to provide the required ASME Section XI margins considering both fracture and limit load mechanisms for 10,600 hours of operation before the next required inspection. This margin is maintained with allowance for the following: This margin is maintained with no credit for any of the horizontal welds H1 through H7 which are structurally replaced by the shroud stabilizer assemblies. A bounding crack growth of 5E-5 inches per hour is used to define the next inspection interval. The General Electric analysis has demonstrated that the 5E-5 growth rate is applicable and conservatively bounding for the NMP1 core shroud vertical weld cracking. Allowance is made for crack sizing uncertainty consistent with the NRC-approved BWRVIP-03 requirements. All uninspected regions are assumed cracked through wall. In addition to the structural margin, all the design basis requirements and criteria have been demonstrated to be satisfied. The NDE inspections performed of the core shroud vertical welds and adjacent base metal have demonstrated that the
Safety Evaluation
'ummary Report Page 55 of 68 4
Safety Evaluation No.: '7-'025 Rev;.1 tcont'd.) . f Safety Evaluation Summary: .,': ';,".'i."'!:.(cont',d.) . '>'.==';>p~.;--."".;-,",-,.; ~ '..~...- -., vertical weld.cracking'is IGSCC b'ounded by NRG review.ofithe core shroud IGSCC cracking addressed by the BWRVlP core shroud inspection and evaluation documents. The bounding core shroud crack growth rate of 5E-5, approved by the NRC for generic application, is applicable to the core shroud vertical weld cracking. The NMP'l Technical Specification regarding reactor coolant chemistry has been reviewed and determined to be consistent with the application of the bounding crack growth rate.- Based on this review, no unreviewed safety question exists associated with the vertical weld cracking identified in the RF014 shroud..., vertical weld inspections, provided an inspection interval of 10;600 hours is established for the vertical welds. The 10,600 hour inspection interval is based. on hot o P eratin g time above 200'F. Based on the evaluation performed, it is concluded that the vertical weld cracking identified in the RFO14 shroud vertical weld inspections does not involve an. unreviewed safety question.
~ .*- Safety'Evaluation ~-" Summary. Report ~-'<'age 66 of 68
- = -. Safety Evaluation No.:
'-a.",':<<.
-': 97-100 a ijigig i
'a ~
Implementation Docume'nt No.:
" 'alculation SO-GOTHIG RB01.Rev,. 01 UFSAR Affected Pages: '"i-" -XV-68 XV-76'-: >'" a-.:;~:~,-;:,"a +
~h v
W I System:=- Reactor Water Cleanup Title of Change: Reactor Water Cleanup System High Energy Line Break Re-Analyses tl ~ ~ Description of Change:
~ g~
The following changes to the plant configuration have been performed:
- 1. All resistance temperature detectors (RTD) in the cleanup system areas have been added to the Equipment Qualification (EQ) program.
- 2. Eight of the twelve cleanup area RTDs,-which were originally MINCO Model S1255, have been replaced with PYCO Model 122-7026.
- 3. Two RTDs have been relocated to the auxiliary cleanup pump room. One was relocated from the cleanup pump area and the other from the heat exchanger room area.
- 4. High-energy line break (HELB) temperature and pressure profiles in the Reactor Building have been revised.
- 5. Additional components have been included in the EQ program.
- 6. The backup SCRAM solenoid valves have been reclassified from safety-related active to safety-related passive.
- 7. The cleanup system HELB analysis has been revised; the new analysis assumes that the isolation is initiated by high temperature detection.
The cleanup system as configured and analyzed meets the design and licensing basis commitments as defined in the UFSAR and other design and licensing basis documents.
Safety Evaluation . Summary Report Page 67 of 68 Safety Evaluation No.: ....='7-1 00 (cont'd.) .- 'iir r-".fi~ g4 'I'II&II Safety Evaluation Summary i'i"~ ~II .;.:.-..='i'.". i '.
~ ~
)Ig I
i"~ '~..~"-i-."." i.:-.ii~i<~ -;el:~:. All equipment necessary'to mitigate the consequences of a cleanup system line-..:iI break or to initiate and maintain a safe shutdown during or following a cleanup ";, system line break have been verified to be qualified for the revised HELB profiles.
~ I I ~ P With high area temperature detectors located in appropriate locations, it can be concluded that the guillotine line break is a bounding event for the cleanup ..;:
system. The guillotine break at full power is bounded by the main steam line break. ~ 'I NMP1 has inherent features and capabilities which provide a basis for reasonable << assurance that leaks and small breaks will be detected within design basis limits. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation-Summary. Report Page:58'o'f 68 ~ . Safety-EvsiuatIon No.! " ':~'-'.; 97-101 Rev. 1 ted&, Implementation Document No.: 'rocedure No. NMP-SOT-001,",~'=,."w-.."v"-', -:::: c,:>" .v NMP BOT 002 v
~ ~ teel() ~ JP i )It< It'p ~'l~g~4+ Mr) a
)~ 6 g )ega ~ g~gpl gl" UFSAR Affected Pages: N/A I
System: Reactor Vessel, Core Shroud, Reactor Water Title of Change: Core Shroud Boat Sample Removal Description of Change: This safety evaluation analyzed the impact of removing two-boat-shaped samples from the Unit 1 core shroud. The boat samples were approximately 1.7" long,
- 1.13" wide and 0.85" deep. The core shroud has been structurally analyzed considering the removal of this sample and the remaining structural ligament and probability of intergranular stress corrosion cracking (IGSCC). In addition, the generation and impact of swarf, due to the electrical discharge machining (EDM) process, on the plant systems has been evaluated.
Safety Evaluation Summary: The EDM of two boat-shaped samples from the core shroud has been analyzed for . conformance to UFSAR and Technical Specification requirements. A structural analysis of the core shroud has been performed and demonstrates the structural adequacy of the core shroud. The generation and impact of swarf on plant systems, including reactor water cleanup, spent fuel pool filtering and cooling, reactor recirculation, control rod drive, and condensate and feedwater, has been considered and found acceptable. The integrity of the core shroud assures that the core spray spargers, core geometry, core flow distribution and control blades function as required. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation
, Summary; Report=
Page-59 of 68 . Safety Evaluation No . 97-,-'1.02 -'- Imptementation:Document, No '.'-'-"-.-: iDCR.',1)-97-'UFS-:043 i. -";;-"...;.'"-'i;>'.Jh>.: UFSAR Affected Pages: XII-14 System: Area Radiation Monitoring Title of Change: Change to Section XII-B.2.1.1.2 of Unit 1 UFSAR Description of Change: e The UFSAR has been updated to clarify the design basis of the area radiation monitor (ARIVI) in the new fresh fuel storage vault to show it is not subject to sudden changes in radiation levels and, therefore, does not require both an alarm in the Control Room and in the area where the monitor is located. Safety Evaluation Summary: The ARM in the fresh fuel storage vault is not subject to sudden changes in radiation levels due to the inherent design of the bundles in the rack where geometric spacing is used to preclude criticality. This change only provides clarification in the UFSAR regarding the use of the ARM already in place in the vault to show that it is within the NMP1 design basis. Based on the evaluation performed, it is-concluded that this change does not involve an unreviewed safety question.
~ ~
Safety Evaluation Summary Report.'age 80-of 88- , Safety Evaluation No.: ..*
'97-1.03 '".
Imple'mentation Document No;:.'. '-: Mod. N1-94-003:.i~ - .-;. -,->>..-,. UFSAR Affected Pages: '"i. iN/A if:;". * )i p ) 4<<1fl&.
~'ystem:
Reactor Vessel Title of. Change: Installation of Modified Shroud Repair Latches Prior to NRC Approval of Adequacy Under 10CFR50.55a(a)(3)
~ ~
Description of Change: The UFSAR describes the shroud tie rod lower lateral spring as being in contact with the shroud and the reactor pressure vessel (RPV), and is designed to restrain lateral movement of the shell between welds H5 and H6A via the core plate bolts and wedges, the ring between welds H6A and H6B, and the shell between H6B and H7. For this analysis, the lower lateral spring was presumed not to be in contact with the shroud and RPV and not capable of providing horizontal restraint. Lateral movement of the lower shroud is restrained by the remaining ligament of good metal at welds H4 through H7. This modification installed a modified latch design without prior NRC approval of the modification and, therefore, takes no credit for the latch to perform its design basis function. The analysis was performed with the restriction that the reactor remain in the cold shutdown or hot shutdown condition. Safety Evaluation Summary: This evaluation analyzed the ability of the tie rod assembly to provide restraint to the shroud differently than that currently described in the UFSAR. The analysis demonstrates that the modified lower wedge latches are not required to perform their intended design basis function during the cold shutdown and hot shutdown condition, i.e., the combination of the structural integrity provided by shroud horizontal welds H4 through H7, and the tie rod components credited in the analysis, has demonstrated that the shroud will perform its design basis functions during noncritical hydro testing above 212 F, and/or control rod drive (CRD) scram time testing with the reactor vessel beltline downcomer water temperature as required to satisfy Technical Specification 3.2.2.e. Compliance with the Technical Specification requires the reactor be considered in the hot shutdown condition. In addition, during hot shutdown several leak rate tests and CRD scram time tests are
Safety Evaluation Summary Report Page 61 of 68 --:;.'.:.." ..:-. Safety Evaluation No.-. . -'.. 97:l03 (cont'd.) r,
-Safety Evaluation 'Summa'ry:"-6 I'8-CRr"-{c'one!d.)-;....:.c.'V.,mam>8" vQ ~vgi;;,oem.-i,;,.-
>" 80 >
I F-CWV-i4->9. performed.'hese tests have no impact on the conditions evaluated in the analysis section. This review demonstrates that'during the shutdown conditions the shroud is operable and its repair assemblies are operable 1) without an Increase in the probability or consequences of an accident or malfunction previously evaluated, 2) without creating the possibility of an accident or malfunction of a new'r dNerent kind from any previously evaluated, and 3) without reducing the margin of safety in the bases of a Technical Specification. Based on the evaluation performed, it is concluded that this change does not " -.
-involve an unreviewed safety question.
Safety Evaluation Summary Report Page 52 of 58 ..: Safety Evaluation No.: '--"'7104 i ImplementatIon Document No.: l GE-)IE 523-B13-01869-043 Rev. 0,: =- ...... GE-NE-523-113-0894 Rev. 1, BWRVIP-07
"">>-'> tlljc ).'Ai'9":l5J)d. a'i's,".~~~-,~i" Nq~ip';cpu,' ~i!~,'=I<::.g~i;-q >:yl'~;-,;.:~
UFSAR Affected Pages: '-. ':~.- -N/A System:;, , Reactor Vessel Internals Title of Change: Core Shroud Vertical Weld Cracking, Cold and Hot Shutdown Description of Change: Inspection of the core shroud vertical wetds identified intergranular stress corrosion cracking (IGSCC) of the vertical welds. The inspections revealed fairly significant cracking on welds V-4, V-9, and V-10; relatively minor cracking on welds V-3, V-12, V-15 and V-16; no cracking on the accessible portions of V-7, V-8, and V-11. Safety Evaluation Summary: The vertical weld cracking has been analyzed and determined to provide the required ASME Section XI margins considering both fracture and limit load mechanisms for the reload condition. This margin is maintained with allowance for the following: This margin is maintained utilizing shroud stabilizer assemblies and horizontal welds as approved in Safety Evaluation 97-103. A bounding crack growth of 5E-5 inches per hour is used to define the next inspection interval. The General Electric analysis has demonstrated that the 5E-5 growth rate is applicable and conservatively bounding for the NMP1 core shroud vertical weld cracking. Crack growth rate is insignificant for the temperature and reactor water chemical conditions during these conditions. Even when considered, the resulting crack growth is immeasurable for the required duration of the testing. Allowance is made for crack sizing uncertainty consistent with the NRC-approved BWRVIP-03 requirements. All uninspected regions are assumed cracked through wall.
Safety Evaluation Summary Report Page 63 of 68 e I . 4 ~
~ 'p F'C" .'2' Safety Evaluation No.:
g 4 97-104 (cont'd.) av
......~"-:~". '::".
in,addition to the structural, margin, all the design basis requirements and criteria have been demonstrated to be satisfied. Based on the evaluation performed, it is concluded that vertical weld cracking identified in the RFO14 shroud vertical weld inspections for the cold and hot I shutdown modes, including noncritical hydro testing and CRD scram time testing, does not involve an unreviewed safety question.
~
~
Safety Evaluation Summary: "'"'"-""=" (cont'd.);-...5';,".e.i. ~ri.-'='".e.ts:."r~-.r-.=-'".,:
=
~
Y
~ g . o Safety Evaluation Summary Report Page 64 of 68 1
- ,,, ...',,Safety Evaluation No.
- ..;"~e a) 87-'107 Implementation Document No.: Nuclear Division Policy,(POL) Rev. 10,
,C Nuclear Safety Assessment 8c Support
'I I fa,~ . ~ 1T+ ~ ~ ~ l> t 0 u ~
~. Policy (NSAS-POL-01) Rev~)0 k~ .)
A C UFSAR Affected Pages: Xlll-1, Xlll-3, XIII-4; Figures Xlll-1, XIII-4
~ ~
System: .':N/A
't Title of Change: :Organization of Q1P, Labor Relations, HRD..
and Occupational Safety and Health Under the Newly Created Position of Director Human Resource Development Description of Change: The Nuclear Division Policy (POL) and NSAS-POL-01 have been revised to reorganize the functions of Employee/Labor Relations, Leadership/Career Development, Occupational Safety and Health, Quality First Program (Q1P) administrative issues, and the Fitness for Duty Program under the newly created position of "Director Human Resource Development." Safety Evaluation Summary: The proposed organizational changes establish responsibilities and lines of authority and communications for the newly created position of "Director Human Resource Development." The proposed organizational structure satisfies the criteria of SRP '13.1.1 and conforms with the requirements of Section 6.2.1.a of the plant Technical Specifications. The proposed changes do not impact accident or malfunction initiation, or radiological consequences. Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
I Safety Evaluation ~ ~ Summary Report Page 65 of 68 Safety Evaluation No;: 97-108 "c, Cft'-' I) 0 g4y1g
~ <
VV Implementation Document No.::.." >-,DER-1-97-1433 ";".-:,; .;;-;,.;;:.. UFSAR Affected Pages: IV-20, X-8, X-12, X-14 System: ~ Control Rod Drive (CRD) Title of Change: UFSAR Update for Control Rod Withdrawal Speed Description of Change: I 4p This safety evaluation evaluated a change to the UFSAR in the allowed tolerance for control rod withdrawal rate from 3 in/sec to 3~ 20% (i.e., 2 4- 3.6) in/sec,,- which corresponds to a full withdrawal time of 38.4- 57.6 seconds. Additionally, the change allows operation with withdraw speeds up to 5.0 in/sec corresponding to a 28-second stroke time. An analysis by General Electric concluded that such operation is bounded by the assumptions used in the rod withdrawal error (RWE) analysis and the minimum critical power ratio safety limit analyses.. This safety evaluation also evaluated operating with CRD drive water pressure less than 250 ps id. Safety Evaluation Summary: Addition of the bases used in the RWE for maximum control rod withdrawal time provides information which can be used to determine operability of a control rod if the stroke time is found out of specification. Lowering drive water pressure to compensate for degraded CRD seals or hydraulic control unit leakage is a conservative action which can be used to maintain CRD stroke time within design. The original design and function of the CRD system are unchanged; the ability of the CRD to function as described in the UFSAR is not affected; and the performance requirements as defined in the Technical Specifications are not affected by the proposed change. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
~ ~
Safety Evaluation Summary Report Page 66 of 68 J Safety Evaluation No.: ',-..".,, "97-121 tt' J l,'
~ (
ImplementatIon Document No.:""~' ', Mod.'N1-87-032:; '~ 4~ ~
/
UFSAR Affected Pages'.:" '-"'-:,:. "".BOA-34. e System: Smoke Detection Title of Change: Addition of Smoke Detector in Zone DA-2022S;
~
Description of Change:
's a result of walkdowns conducted by the Security Department to determine if unauthorized access may be obtained; it became evident that openings in the Uninterruptible Power Supply (UPS) Battery Room and UPS Room (TB El. 250'):
may permit unauthorized access to these rooms. This modification provided barriers designed to control access to these areas and.. installed an additional smoke detector inside the UPS Room. Safety Evaluation Summary: Addition of this extra smoke detector provides fire detection monitoring for the UPS Room. This enhances the ability of plant personnel to detect and respond to potential fires. Thus, this change has no adverse effect on the probability of occurrence of a fire in any plant area which is different from any fire or accident previously evaluated in the SAR. Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 67 of 68 Safety Evaluation No.: -'. '>rtr~,": '97-124 >.S aa aS
- y. ~
~ ~ ~
S
~
~ qa a
".an)pa, ~S a
a a a,. UFSAR Affected Pages: a Implemen'tation Document No.: a I p ~
~ ~,S V>>a S 1 ~
~ ~
- l. IGE-.NE-523-B'1 3-01869-043 Rev..0;>>:= ~-.": ~B
~
GE-NE-523;113-0894 Rev.. 1, BWRVIP-07 s.a>>a N
'a
/A>>*
s
~
~
f( . J sa,all)
~
~ ~
~
I ~
>>a>> >>
~
stst') aaas 'AL" a =
$~
wa as
~
~
a,Ah s) +4) f":
'a.
~
~
~
a ~
~-
g( a aa>> a
~
a ~ a qa
- =f 4 aaa a a~
SIP baa q* System: Reactor Vessel Internals . Title of Change: Core Shroud Vertical Weld Crack, Cold Shutdown (Refueling and Major Maintenance) Description of Change: Inspection of the core shroud vertical welds identified intergranular stress corrosion cracking (IGSCC) of the vertical welds. The inspections revealed fairly significant cracking on welds V-4, V-9, and V-10; relatively minor cracking on welds V-3, V-12, V-15 and V-16; no cracking on the accessible portions of V-7, V-S, and V-11. Safety Evaluation Summary: The vertical weld cracking has been analyzed and determined to provide the required ASME Section XI margins considering both fracture and limit load mechanisms for the reload condition. This margin is maintained with allowance for the following: This margin is maintained with no credit for any of the horizontal welds H1 through H7 which are structurally replaced by the shroud stabilizer assemblies. A bounding crack growth of 5E-5 inches per hour is used to define the next inspection interval. The General Electric analysis has demonstrated that the 5E-5 growth rate is applicable and conservatively bounding for the NMP1 core shroud vertical weld cracking. Crack growth rate need not be applied for the refueling mode. Allowance is made for crack sizing uncertainty consistent with the NRC-approved BWRVIP-03 requirements. All uninspected regions are assumed cracked through wall.
g Safety Evaluation Summary Repoit = -, Page 68 of 68 I C, ~
, Safety, EvaluatIon No.:. =- '-.-',:" .:-',.'97-,'I24 (cont'd.)
~ t
. Safety Evaluation".Summary -"~".:Dkr.-(cont'.A.) =;".!'"'r::n-":."~~.."-"iZ-';-..t,".'..=.;l In addition tb'the Structuraf'margin,-all the design basis requirements and criteria have been demonstrated to be satisfied.
Based on the evaluation performed, it is concluded that vertical weld crackirig identified in the RFO14 shroud vertical weld inspections for the refueling mode does not involve an unreviewed safety question. 4
pA ~/6W ii 7/<7 PP~~~~~<>" U.S." NUCLEAR REGULATORY COMMISSIO-DOCKET 0-220 LICENSE D - 3 NINE MILE POINT NUCL AR STATION U'T1 FINAL SAF TY ANALYSIS REP RT (UPDATED) VOLUME 1 JUNE 1996 REVISION 14 NIAG&&MOHAWKPOWER CORPORATION S&&CUSE, NEW YORK
0 Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS Section Title Pacae TABLE OF CONTENTS LIST OF TABLES LIST OF FIGURES 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-2 4.0 Reactor Vessel I-4 5.0 Containment I-5 6~0 Control and Instrumentation I-6 7.0 Electrical Power I-8 8.0 Radioactive Waste Disposal I-8 9.0 Shielding and Access Control I-8 10.0 Fuel Handling and Storage I-8 B. CHARACTERISTICS I-9 1.0 Site I-9 2.0 Reactor I-9 3.0 Core I-9 4.0 Fuel Assembly I-9 5.0 Control System I-9 6.0 Core Design and Operating Conditions I-10 7.0 Design Power Peaking Factor I-10 8.0 Nuclear Design Data I-10 9.0 Reactor Vessel I-11 10.0 Coolant Recirculation Loops I-11 11.0 Primary Containment I-11 12.0 Secondary Containment I-11 13.0 Structural Design I-11 14.0 Station Electrical System I-12 15.0 Reactor Instrumentation System I-12 16.0 Reactor Protection System I-12 C. IDENTIFICATION OF CONTRACTORS I-13 D. GENERAL CONCLUSIONS I-14 E. REFERENCES I-15 SECTION II STATION SITE AND ENVIRONMENT II-1 A. SITE DESCRIPTION II-1 1~0 General II-1 UFSAR Revision June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectio Title Pacae 2.0 Physical Features II-1 3.0 Property Use and Development II-2 B. DESCRIPTION OF AREA ADJACENT TO THE SITE II-3 1.0 General II-3 1~1 Population II-3 2.0 Agriculture, Industrial and Recreational Use II-3 2.1 Agricultural Use II-3 2.2 Industrial Use II-3 2.3 Recreational Use II-4 C. METEOROLOGY II-5 D. LIMNOLOGY II-6 E. EARTH SCIENCES II-7 F. ENVIRONMENTAL RADIOLOGY II-8 G. REFERENCES II-9 SECTION III BUILDINGS AND STRUCTURES III-1 A. TURBINE BUILDING III-3 1.0 Design Bases III-3 1.1 Wind and Snow Loadings III-3 1~2 Pressure Relief Design III-3 1~3 Seismic Design and Internal Loadings III-3 1.4 Heating and Ventilation III-4 1.5 Shielding and Access Control III-4 2.0 Structure Design III-4 2.1 General Structural Features III-5 2.2 Heating and Ventilation System III-5 2.3 Smoke and Heat Removal III-7 2.4 Shielding and Access Control III-7 3.0 Safety Analysis III-7 B. CONTROL ROOM III-9 1' Design Bases III-9 1.1 Wind and Snow Loadings III-9 1~2 Pressure Relief Design III-9 1.3 Seismic Design and Internal Loadings III-9 1.4 Heating and Ventilation III-9 1.5 Shielding and Access Control III-9 2' Structure Design III-10 2.1 General Structural Features III-10 UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 2.2 Heating, Ventilation and Air Conditioning System III-11 2.3 Smoke and Heat Removal III-11 2.4 Shielding and Access Control III-12 3.0 Safety Analysis III-12 C. WASTE DISPOSAL BUILDING XII-13 1.0 Design Bases III-13 1.1 Wind and Snow Loadings III-13 1.2 Pressure Relief Design IIX-13 1' Seismic Design and Internal Loadings III-13 1.4 Heating and Ventilation III-14 1.5 Shielding and Access Control III-'14 2.0 Structure Design III-14 2.1 General Structural Features III-14 2.2 Heating and Ventilation System III-15 2.3 Shielding and Access Control III-17 3.0 Safety Analysis III-17 D. OFFGAS BUILDING III-19 1.0 Design Bases III-19 1.1 Wind and Snow Loadings IXI-19 1.2 Pressure Relief Design IIX-19 1' Seismic Design and Internal Loadings III-19 1.4 Heating and Ventilation III-19 1.5 Shielding and Access Control III-19 2.0 Structure Design III-19 2.1 General Structural Features III-19 2.2 Heating and Ventilation System III-20 2.3 Shielding and Access Control III-20 3.0 Safety Analysis III-20 E. NONCONTROLLED BUILDINGS III-22 1.0 Administration Building III-22 1.1 Design Bases III-22 1' ' Wind and Snow Loadings III-22 1 '.2 1.1.3 Pressure Relief Design Seismic Design and Internal Loadings III-22 III-22 1' ' Heating, Cooling and Ventilation III-23 1.1.5 Shielding and Access Control III-23 1.2 Structure Design III-23 1 '.1 1.2.2 General Structural Features Heating, Ventilation and Air III-23 Conditioning III-24 1 '-3 1.3 Access Control Safety Analysis IXI-24 III-24 2.0 Sewage Treatment Building XII-25 F 1 Design Bases III-25 2.1.1 Wind and Snow Loadings III-25 UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 2' ' Pressure Relief Design III-25
- 2. 1-3 Seismic Design and Xnternal Loa dings III-25
- 2. 1.4 Electrical Design III-25 2.1.5 Fire and Explosive Gas Detectio IIX-25 2.1.6 Heating and Ventilation III-26 2.1.7 Shielding and Access Control III-26 2.2 Structure Design III-26 2~2 1 General Structural Features XII-26 2
2
'.2 '.3 Ventilation System Access Control III-27 III-28 3.0 Energy Information Center III-28 F 1 Design Bases III-28 3
3
'.1 Wind and Snow Loadings Pressure Relief Design III-28 III-28 3 '.3 Seismic Design and Internal Loadings III-28
- 3. 1.4 Heating and Ventilation III-29
- 3. 1.5 Shielding and Access Control III-29 3 ' Structure Design III-29 3.2.1 General Structural Features III-29 3.2.2 Heating and Ventilation System III-29 3.2.3 Access Control III-30 F. SCREENHOUSE, INTAKE AND DISCHARGE TUNNELS ZII-31 1.0 Screenhouse III-31 1.1 Design Basis III-31 1~1~1 Wind and Snow Loadings III-31
- 1. 1.2 Pressure Relief Design III-31
- 1. 1.3 Seismic Design and Internal Loadings III-31 1.1.4 Heating and Ventilation III-31
- 1. 1.5 Shielding and Access Control III-31 1.2 Structure Design III-31 2.0 Intake and Discharge Tunnels XII-33 2.1 Design Bases III-33 2 ' Structure Design XII-33 3.0 Safety Analysis III-34 G. STACK III-35 1.0 Design Bases III-35 1~1 General III-35 1.2 Wind Loading IXX-35 1.3 Seismic Design III-35 1.4 Shielding and Access Control III-35 2.0 Structure Design III-35 3.0 Safety Analysis III-36 3.1 Radiology III-36 3.2 Stack Failure Analysis III-37 UFSAR Revision 14 iv June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 3.2.1 Reactor Building III-37 3 '.2 Diesel Generator Building III-38 3.2.3 Screen and Pump House III-38 H. SECURITY BUILDING AND SECURITY BUILDING ANNEX III-39 RADWASTE SOLIDIFICATION AND STORAGE BUILDING IIX-40 1~0 Design Bases III-40 1~1 Wind and Snow Loadings III-40 1.2 Pressure Relief Design III-40 1' Seismic Design and Internal Loadings XXX-40 1.4 Heating, Ventilation and Air Conditioning IIX-40 1~5 Shielding and Access Control III-40 2.0 Structure and Design III-41 2' General Structural Features IIX-41 2.2 Heating, Ventilation and Air Conditioning IXI-41 2.3 Shielding and Access Control IXI-43 3.0 Use IIX-43 REFERENCES III-45 SECTION IV REACTOR IV-1 A. DESIGN BASES IV-1 1.0 General IV-1 2.0 Performance Objectives IV-1 3.0 Design Limits and Targets IV-2 B. REACTOR DESIGN IV-3 1.0 General IV-3 2.0 Nuclear Design Technique IV-4 2.1 Reference Loading Pattern IV-5 2' Final Loading Pattern IV-6 2~2~1 Acceptable Deviation From Reference Loading Pattern IV-6 2.2 ' Reexamination of Licensing Basis IV-6 2 ' Refueling Cycle Reactivity Balance .IV-7 3.0 Thermal and Hydraulic Characteristics IV-7 3 ~ 1 Thermal and Hydraulic Design IV-7 3.1.1 Recirculation Flow Control IV-7 3' ' Core Thermal Limits IV-7 3.1.2.1 Excessive Clad Temperature IV-8 3.1.2.2 Cladding Strain IV-9 UFSAR Revision 14 v June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 3.1.2.3 Coolant Flow IV-9 3 ' Thermal and Hydraulic Analyses IV-9 3.2.1 Hydraulic Analysis IV-9 3.2.2 Thermal Analysis IV-11 3.2.F 1 Fuel Cladding Integrity Safety Limit Analysis IV-11 3.2.2.2 MCPR Operating Limit Analysis IV-12 3' Reactor Transients IV-13 4.0 Stability Analysis IV-14 4.1 Design Bases IV-14 4.2 Stability Analysis Method IV-14 5.0 Mechanical Design and Evaluation IV-15 5.1 Fuel Mechanical Design IV-15 5.1.1 Design Bases IV-15
- 5. 1.2 Fuel Rods IV-15 5.1.3 Water Rods IV-16 5.1.4 Fuel Assemblies IV-16 5.1.5 Mechanical Design Limits and Stress Analysis IV-16 5.1.6 Relationship Between Fuel Design Limits and Fuel Damage Limits IV-16 5.1.7 Surveillance and Testing IV-16 6.0 Control Rod Mechanical Design and Evaluation IV-17 6.1 Design IV-17 6.1.1 Control Rods and Drives IV-17
- 6. 1.2 Standby Liquid Poison System IV-19 6.2 Control System Evaluation IV-20 6.2.1 Rod Withdrawal Errors Evaluation IV-20 6.2.2 Overall Control System Evaluation IV-21 6.3 Limiting Conditions for Operation and Surveillance IV-23 6.4 Control Rod Lifetime IV-23 7.0 Reactor Vessel Internal Structure IV-24 7.1 Design Bases IV-24 7.1.1 Core Shroud IV-25 7.1.2 Core Support IV-25 7 ~ l.. 3 Top Grid IV-26
- 7. 1.4 Control Rod Guide Tubes IV-26 7.1.5 Feedwater Sparger IV-26 7.1.6 Core Spray Spargers IV-26 7.1 ' Liquid Pois'on Sparger IV-26 7.1.8 Steam Separator and Dryer IV-26 7.1.9 Core Shroud Stabilizers IV-27 C. REFERENCES IV-30 7.2 Design Evaluation IV-29 7 ' Surveillance and Testing IV-29 UFSAR Revision 14 vi June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae SECTION V REACTOR COOLANT SYSTEM V-1 A. DESIGN BASES V-1 1.0 General V-1 2.0 Performance Objectives V-1 3.0 Design Pressure V-2 4.0 Cyclic Loads (Mechanical and Thermal) V-3 5.0 Codes V-3 B. SYSTEM DESIGN AND OPERATION V-4 1~0 General V-4 1.1 Drawings V-4 1.2 Materials of Construction V-4 1.3 Thermal Stresses V-4 1.4 Primary Coolant Leakage V-5 1.5 Coolant Chemistry V-5 2.0 Reactor Vessel V-5 3.0 Reactor Recirculation Loops V-6 4.0 Reactor Steam and Auxiliary Systems Piping V-7 5.0 Relief Devices V-7 C. SYSTEM DESIGN EVALUATION V-9 1' General V-9 2' Pressure V-9 3' Design Heatup and Cooldown Rates V-10 4.0 Materials Radiation Exposure V-11 4.1 Pressure-Temperature Limit Curves V-11 4' Temperature Limits for Boltup V-11 4.3 Temperature Limits for In-Service System Pressure Tests V-12 4 ' Operating Limits During Heatup, Cooldown, and Core Operation V-12 4.5 Predicted Shift in RT>>~ V-12 5.0 Mechanical Considerations V-12 5.1 Jet Reaction Forces V-12 5.2 Seismic Forces V-13 5.3 Piping Failure Studies V-13 6.0 Safety Limits, Limiting Safety Settings and Minimum Conditions for Operation V-13 D. TESTS AND INSPECTIONS V-15 1.0 Prestartup Testing V-15 2.0 Inspection and Testing Following Startup V-15 2.1 Hydro Pressure V-15 2.2 Pressure Vessel Irradiation V-15 UFSAR Revision 14 vii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae E. EMERGENCY COOLING SYSTEM V-16 1.0 Design Bases V-16 2 ' System Design and Operation V-16 3 ' Design Evaluation V-17 F 1 Redundancy V-17 3 ' Makeup Water V-18 3 ' System Leaks V-18 3.4 Containment Isolation V-18 4.0 Tests and Inspections V-19 4.1 Prestartup Test V-19 4' Subsequent Inspections and Tests V-19 F. REFERENCES V-20 SECTION VI CONTAINMENT SYSTEM VI-1 A. PRIMARY CONTAINMENT-NARK I VI-2 CONTAINMENT PROGRAM 1.0 General Structure VI-2 2.0 Pressure Suppression Hydrodynamic Loads VI-2 2.1 Safety/Relief Valve Discharge VI-2 2' Loss-of-Coolant Accident VI-3 2' Summary of Loading Phenomena VI-4 3.0 Plant-Unique Modifications VI-5 B. PRIMARY CONTAINMENT PRESSURE SUPPRESSION SYSTEM VI-6 1.0 Design Bases VI-6 1~1 General VI-6 1~2 Design Basis Accident (DBA) VI-6 1~3 Containment, Heat Removal VI-8 1.4 Isolation Criteria VI-8 1.5 Vacuum Relief Criteria VI-8 1.6 Flooding Criteria VI-9 1.7 Shielding VI-9 2.0 Structure Design VI-9 2 1 General VI-9 2.2 Penetrations and Access Openings VI-11 2.3 Jet and Missile Protection VI-12 2.4 Materials VI-13 2.5 Shielding VI-13 2.6 Vacuum Relief VI-14 2' Containment Flooding VI-14 C. SECONDARY CONTAINMENT REACTOR BUILDING VI-16 1.0 Design Bases VI-16 1.1 Wind and Snow Loadings VI-16 UFSAR Revision 14 viii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic Title Pacae 1' Pressure Relief Design VI-16 1.3 Seismic Design VI-17 1.4 Shielding VI-17 2.0 Structure Design VI-17 2.1 General Structural Features VI-17 D. CONTAINMENT ISOLATION SYSTEM VI-20 1.0 Design Bases VI-20 1.1 Containment Spray Appendix J Water Seal Requirements VI-23 2 ' System Design VI-24 3.0 Tests and Inspections VI-26 E. CONTAINMENT VENTILATION SYSTEM VI-27 1.0 Primary Containment VI-27 1.1 Design Bases VI-27 1' System Design VI-27 2.0 Secondary Containment VI-28 2.1 Design Bases VI-28 2.2 System Design VI-28 F. TEST AND INSPECTIONS VI-30 1.0 Drywell and Suppression Chamber VI-30 1.1 Preoperational Testing VI-30 1.2 Postoperational Testing VI-30 2.0 Containment Penetrations and Isolation Valves VI-31 2.1 Penetration and Valve Leakage VI-31 2.2 Valve Operability Test VI-31 3.0 Containment Ventilation System VI-32 4.0 Other Containment Tests VI-32 5.0 Reactor Building VI-32 5.1 Reactor Building Normal Ventilation System VI-32 5.2 Reactor Building Isolation Valves VI-33 5.3 Emergency Ventilation System VI-33 G. REFERENCES VI-33 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 ' General VII-2 2.2 Operator Assessment VII-5 3.0 Design Evaluation VII-6 4.0 Tests and Inspections VII-6 UFSAR Revision ix June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectio Title Pacae B. CONTAINMENT SPRAY SYSTEM VII-8 1.0 Design Bases VII-8 2.0 System Design VII-8 2.1 Operator Assessment VII-11 3.0 Design Evaluation VII-12 4.0 Tests and Inspections VII-13 C. LIQUID POISON XNJECTION SYSTEM VII-15 1.0 Design Bases VII-15
- 2. 0' System Design VII-15
~ 1 Operator Assessment VII-18 3.0 Design Evaluation VII-19 4.0 Tests and Inspections VII-20 5.0 Alternate Boron Injection VIX-20 D. CONTROL ROD VELOCITY LXMITER VII-22 1.0 Design Bases VII-22 2.0 System Design VIX-22 3.0 Design Evaluation VII-24 3' General VII-24 3.2 Design Sensitivity VII-24 3.3 Normal Operation VII-25 4.0 Tests and Inspections VII-25 E. CONTROL ROD HOUSING SUPPORT VII-26 1.0 Design Bases VII-26 2.0 System Design VII-26 2.1 Loads and Deflections VII-28 3.0 Design Evaluation VII-28 4.0 Tests and Inspections VII-29 F. FLOW RESTRICTORS VII-30 1.0 Design Bases VII-30 2.0 System Design VII-30 3.0 Design Evaluation VII-30 4.0 Tests and Inspections VII-31 G. COMBUSTIBLE GAS CONTROL SYSTEM VII-32 1.0 Design Bases VII-32 2.0 Containment Inerting System VII-32 2 ' System Design VIZ-32 2 ' Design Evaluation VII-33 3.0 Containment Atmospheric Dilution System VII-33 3.1 System Design VII-33 3.2 Design Evaluation VII-35 4.0 Tests and Inspections VII-35 UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic Title Pacae H. EMERGENCY VENTILATION SYSTEM VII-36 1.0 Design Bases VII-36 2.0 System Design VII-36 2.1 Operator Assessment VII-38 3.0 Design Evaluation VII-39 4.0 Tests and Inspections VII-39 I ~ HIGH-PRESSURE COOLANT INJECTION VII-41 VII-41 1.0 Design Bases 2.0 System Design VII-41 3.0 Design Evaluation VII-42 4.0 Tests and Inspections VII-43 REFERENCES VII-44 SECTION VIII INSTRUMENTATION AND CONTROL VIII-1 A. PROTECTIVE SYSTEMS VIII-1 1.0 Design Bases VIII-1 1~1 Reactor Protection System VIII-1 1.2 Anticipated Transients Without Scram Mitigation System VIII-4 2.0 System Design VIII-4 2' Reactor Protection System VIII-4 2.2 Anticipated Transients Without Scram Mitigation System VIII-10 3.0 System Evaluation VIII-10 B. REGULATING SYSTEMS VIII-12 1.0 Design Bases VIII-12 2.0 System Design VIII-12 2.1 Control Rod Adjustment Control VIII-12 2.2 Recirculation Flow Control VIII-12 2 ' Pressure and Turbine Control VIII-13 2.4 Reactor Feedwater Control VIII-14 3.0 System Evaluation VIII-14 3.1 Control Rod Adjustment Control VIII-14 3 ' Recirculation Flow Control VIII-14 3.3 Pressure and Turbine Control VIII-14 3.4 Reactor Feedwater Control VIII-14 C. INSTRUMENTATION SYSTEMS VIII-15 1.0 Nuclear Instrumentation VIII-15 1.1 Design VIII-15 1.1.1 Source Range Monitors VIII-17
- 1. 1.2 Intermediate Range Monitors VIII-18 1.1.3 Local Power Range Monitors VIII-19
- 1. 1.4 Average Power Range Monitors VIII-19
- 1. 1.5 Traversing In-Core Probe System VIII-21 UFSAR Revision 14 Xi June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 1.2 Evaluation VIII-21 1.2.1 Source Range Monitors VIII-22 1' ' Intermediate Range Monitors VIII-23 1.2 ' Local Power Range Monitors VIII-25 1.2.4 Average Power Range Monitors VIII-25 2 ' Nonnuclear Process Instrumentation VIII-26 2.1 Design Bases VIII-26 2.1.1 Nonnuclear Process Instruments in Protective System VIII-26 2' ' Nonnuclear Process Instruments in Regulating Systems VIXI-28 2.1.3 Other Nonnuclear Process Instruments VIII-29 2.2 Evaluation VIII-31 2.2.1 Nonnuclear Process Instruments in Protective System VIII-31 2.2 ' Nonnuclear Process Instruments in Regulating Systems VIII-3g 2.2 ' Other Nonnuclear Process Instruments VIII-31 3.0 Radioactivity Instrumentation VIXI-32 3.1 Design Bases VIII-32 3~1~1 Radiation Monitors in Protective Systems VIIX-32 3.1.2 Other Radiation Monitors VIII-34 3 ' Evaluation VIXI-36 4.0 Other Instrumentation VIXI-37 4.1 Rod North Minimizer VIII-37 4.1.1 Design Bases VIII-37 4.1.2 Evaluation VIII-38 5.0 Regulatory Guide 1.97 (Revision 2) Instrumentation VIII-39 5.1 Licensing Activities Background VIII-39 5.2 Definition of RG 1.97 Variable Types and Instrument Categories VIII-39 5.3 Determination of RG 1.97 Type A Variables for Unit 1 VIII-41 5.4 Determination of EOP Key Parameters for Unit 1 VIII-42 5.4.1 Determination Basis/Approach VIII-42 5.4.2 Definition of Primary Safety Functions VIII-43 5.4.3 Association of EOPs to Primary Safety Functions VIII-43 5.4.4 Identification of EOP Key Parameters VXXX-44 5.5 Unit 1 RG 1.97 Variables, Variable Type, and Associated Instrument Category Designations VIII-44 UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Parcae 5.6 Summary of the RG 1.97 Instrument Design and Implementation Criteria that were Established for Unit 1 as Part of the Unit 1 1990 Restart Activities VIII-45 5.6 1
~ No Type A Variables VIII-46
- 5. 6.2 EOP Key Parameters VIII-46 5.6 ' Single Tap for the Fuel Zone RPV Water Level Instrument VIII-46 5.6 ' Nonredundant Wide-Range RPV Water Level Indication VIII-48 5.6.5 Upgrading EOP Key Parameter Category 1 Instrument Loop Components to Safety-Related Classification VIII-48 5.6.6 Safety-Related Classification of Instrumentation for RG 1.97 Variable Types Other than the EOP Key Parameters VIII-49 5.6.7 Routing and Separation of Channelized Category 1 Instrument Loop Cables VIII-49 5.6.8 Electrical Isolation of Category 1 Instrument Loops from Associated Components that are not Safety Related VIII-50
- 5. 6.9 Power Source Information for Category 1 Instruments VIII-51
- 5. 6. 10 Marking of Instruments of Control Room Panels VIII-51 5.6.11 "Alternate" Instruments for Monitoring EOP Key Parameters VIII-51 5.6.12 Indication Ranges of Monitoring Instruments VIII-52 D. REFERENCES VIII-53 SECTION IX ELECTRICAL SYSTEMS IX-1 A. DESIGN BASES IX-1 B. ELECTRICAL SYSTEM DESIGN ZX-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-9 2~1 Two +24-V Dc Systems IX-12 2.2 Two 120-V, 60-Hz, Single-Phase, Uninterruptible Power Supply Systems IX-12 UFSAR Revision 14 Xiii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectio Title Pacae 2.3 Two 120-V, 57-60 Hz, One-Phase, Reactor Trip Power Supplies IX-13 2.4 One 120/208-V, 60-Hz, Instrument and Control Transformer IX-14 2.5 One 120/240-V, 60-Hz, Three-Phase, Computer Power Supply IX-14 3.0 Cables and Cable Trays IX-14 3.1 Cable Separation IX-14 3.2 Cable Penetrations IX-15 3.3 Protection in Hazardous Areas IX-15 3.4 Types of Cables IX-15 3.4.1 Power Cable IX-16 3.4.2 Control Cable IX-16 3.4.3 Special Cable IX-16 3.5 Design and Spacing of Cable Trays IX-17 3.5.1 Tray Design Specifications IX-17 3.5.2 Tray Spacing IX-17 4.0 Emergency Power IX-17 4.1 Diesel Generator System IX-17 4.2 Station Batteries IX-20 4.3 Nonsafety Battery System IX-22 5.0 Tests and Inspections IX-23 5.1 Diesel Generator IX-23 5.2 Station Batteries IX-24 5.3 Nonsafety Batteries IX-24 6.0 Conformance with 10CFR50. 63 Station Blackout Rule IX-24 6.1 Station Blackout Duration IX-25 6.2 Station Blackout Coping Capability IX-25 6.3 Procedures and Training IX-27 6.4 Quality Assurance IX-27 6.5 Emergency Diesel Generator Reliability Program IX-28 6.6 References IX-29 SECTION X REACTOR AUXILIARY AND EMERGENCY SYSTEMS X-1 A. REACTOR SHUTDOWN COOLING SYSTEM X-1 1.0 Design Bases X-1 2.0 System Design X-1 3.0 System Evaluation X-2 4.0 Tests and Inspections X-2 B. REACTOR CLEANUP SYSTEM X-3 1.0 Design Bases X-3 2.0 System Design X-3 3.0 System Evaluation X-4 4.0 Tests and Inspections X-5 UFSAR Revision 14 xiv June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic Title Pacae C. CONTROL ROD DRIVE HYDRAULIC SYSTEM X-6 1~0 Design Bases X-6 2' System Design X-6 2' Pumps X-7 2' Filters X-7 2' First Pressure Stage X-7 X-8 2.4 Second Pressure Stage 2.5 Third Pressure Stage X-8 2.6 Exhaust Header X-9 2.7 Accumulator X-9 2.8 Scram Pilot Valves X-10 2.9 Scram Valves X-10 2.10 Scram Dump Volume X-10 2 2
'1 '2 Control Rod Drive Cooling System Directional Control and Speed X-11 Control Valves X-11 2 ~ 13 Rod Insertion and Withdrawal X-12
- 2. 14 Scram Actuation X-13 3.0 System Evaluation X-13 3' Normal Withdrawal Speed X-13 3.2 Accidental Multiple Operation X-14 3.3 Scram Reliability X-14 3.4 Operational Reliability X-15 3.5 Alternate Rod Injection X-15 4.0 Reactor Vessel Level Instrumentation Reference Leg Backfill X-15 5.0 Tests and Inspections X-16 D. REACTOR BUILDING CLOSED LOOP COOLING WATER SYSTEM X-17 1~0 Design Bases X-17 2' System Design X-17 3.0 Design Evaluation X-19 4.0 Tests and Inspections X-20 E. TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM X-21 1.0 Design Bases X-21 2.0 System Design X-21 3.0 Design Evaluation X-22 4.0 Tests and Inspections X-23 F. SERVICE WATER SYSTEM X-24 1.0 Design Bases X-24 2.0 System Design X-24 3.0 Design Evaluation X-25 4.0 Tests and Inspections X-26 UFSAR Revision 14 XV June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title pacae G. MAKEUP WATER SYSTEM X-27 1.0 Design Bases X-27 2.0 System Design X-27 3.0 System Evaluation X-28 4.0 Tests and Inspections X-29 H. SPENT FUEL STORAGE POOL FILTERING AND COOLING SYSTEM X-30 1.0 Design Bases X-30 2.0 System Design X-31 3.0 Design Evaluation X-33 4.0 Tests and Inspections X-33 BREATHING, INSTRUMENT AND SERVICE AIR SYSTEM X-34 1.0 Design Bases X-34 2.0 System Design X-34 3.0 Design Evaluation X-36 4.0 Tests and Inspections X-37 FUEL AND REACTOR COMPONENTS HANDLING SYSTEM X-38 1.0 Design Bases X-38 2.0 System Design X-38 2.1 Description of Facility X-38 2.1.1 Cask Drop Protection System X-41 2.2 Operation of the Facility X-42 3.0 Design Evaluation X-42 4.0 Tests and Inspections X-43 K. FIRE PROTECTION PROGRAM X-44 1.0 Program Bases X-44 1.1 Nuclear Division Directive Fire Protection Program X-44 1.2 Nuclear Division Interface Procedure Fire Protection Program X-44 1.3 Fire Hazards Analysis X-44 1.4 Appendix R Review Safe Shutdown Analysis X-45 1.5 Fire Protection and Appendix R Related Portions of Operations Procedures (OPs, SOPs, and EOPs) and Damage Repair Procedures X-45 1.6 Fire Protection Portions of the Emergency Plan X-45 2.0 Program Implementation and Design Aspects X-45 2.1 Fire Protection Implementing Procedures X-45 UFSAR Revision 14 xvi June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 2.2 Fire Protection Administrative Controls X-46 2.3 Fire Protection System Drawings and Calculations X-46 2.4 Fire Protection Engineering Evaluations (FPEEs) X-46 3.0 Monitoring and Evaluating Program Implementation X-46 3.1 Quality Assurance Topical Report X-46 3.2 Fire Brigade Manning, Training, Drills and Responsibilities X-46 X-47 4.0 Surveillance and Tests L. REMOTE SHUTDOWN SYSTEM X-48 1.0 Design Bases X-48 2.0 System Design X-48 3.0 System Evaluation X-48 4.0 Tests and Inspections X-49 M. SAFETY PARAMETER DISPLAY SYSTEM X-50 1.0 Design Bases X-50 2.0 System Design X-50 3.0 System Evaluation X-50 4.0 Tests and Inspections X-51 N. REFERENCES X-52 APPENDIX 10A FIRE HAZARDS ANALYSIS APPENDIX 10B SAFE SHUTDOWN ANALYSIS SECTION XI STEAM-TO-POWER CONVERSION SYSTEM XI-1 A. DESIGN BASES XI-1 B. SYSTEM DESIGN AND OPERATION XI-2 1.0 Turbine Generator XI-2 2.0 Turbine Condenser XI-4 3.0 Condenser Air Removal and Offgas System XI-5 4.0 Circulating Water System XI-9 5.0 Condensate Pumps XI-9 6.0 Condensate Demineralizer System XI-9 7.0 Condensate Transfer System XI-10 8.0 Feedwater Booster Pumps XI-11 9.0 Feedwater Pumps XI-11 10 ' Feedwater Heaters XI-11 C. SYSTEM ANALYSIS XI-13 UFSAR Revision Xvll June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae D. TESTS AND INSPECTIONS XI-16 SECTION XII RADIOLOGICAL CONTROLS XII-1 A. RADIOACTIVE WASTES XII-1 1.0 Design Bases XII-1 1.1 Objectives XII-1 1.2 Types of Radioactive Wastes XII-1 1.2.1 Gaseous Waste XII-1 1' ' Liquid Wastes XII-1 1' ' Solid Wastes XII-2 2.0 System Design and Evaluation XII-2 2~1 Gaseous Waste System XII-2 2~1~1 Offgas System XII-3 2~1~2 Steam-Packing Exhauster System XII-3
- 2. 1.3 Buildup Ventilation Systems XII-3
- 2. 1.4 Stack XII-3 2.2 Liquid Waste System XII-4 2.2.1 Liquid Waste Handling Processes XII-4 2.2 ' Sampling and Monitoring Liquid Wastes XII-6 2.2.3 Liquid Waste Equipment Arrangement XII-6 2.2.4 Liquid Radioactive Waste System Control XII-6 2.3 Solid Waste System XII-7 2.3.1 Solid Waste Handling Processes XZI-7 2.3.2 Solid Waste System Equipment XII-9 3.0 Safety Limits XII-9 4.0 Tests and Inspections XII-9 4.1 Waste Process Systems XII-9 4.2 Filters XII-9 4.3 Effluent Monitors XII-9 4.3.1 Offgas and Stack Monitors XII-9 4.3 ' Liquid Waste Effluent Monitor XII-10 B. RADIATION PROTECTION XII-11 1.0 Primary and Secondary Shielding XII-11 1 1 Design Bases XII-11 1.2 Design XII-12 1.2. 1 Reactor Shield Wall XII-12 1.2.2 Biological Shield XII-12 1.2.3 Miscellaneous XII-12 1 3
~ Evaluation XII-13 2.0 Area Radioactivity Monitoring Systems XII-13 2.1 Area Radiation Monitoring System XII-13 2 ~ 1~1 Design Bases XII-13
- 2. 1.2 Design XII-14
- 2. 1.3 Evaluation XII-15 UFSAR Revision 14 xviii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 2.2 Area Air Contamination Monitoring System XII-15 2 ~ 2.1 Design Bases XII-15 2 ' ' Design XII-16 2 3
'.3 Evaluation Radiation Protection XII-16 XII-16 F 1 Facilities XII-17 3.1.1 Laboratory, Counting Room and Calibration Facilities XII-17 3
3
'.2 Change Room and Laundry Facilities Personnel Decontamination Facility XII-18 XII-18 3 '.4 Tool and Equipment Decontamination Facility XII-18 3.2 Radiation Control XII-19 3
3
'.1 '.2 Shielding Access Control XII-19 XII-20 3 ' Contamination Control XII-21 3
3
'.1 3.2 Facility Contamination Control Personnel Contamination Control XII-21 XZI-21 3 '.3 3.4 Airborne Contamination Control Personnel Dose Determinations XII-22 XII-23 3 '.1 3.5 Radiation Dose Radiation Protection XII-23 Instrumentation XII-24 3 5.1 Counting Room Instrumentation XII-24 3 '.2 ~
3.5.3 Portable Radiation Instrumentation Air Sampling Instrumentation XII-24 XII-25 3.5.4 Personnel Monitoring Instruments XII-25 3.5.5 Emergency Instrumentation XII-25 4' Tests and Inspections XII-26 4.1 Shielding XII-26 4.2 Area Radiation Monitors XII-26 4.3 Area Air Contamination Monitors XII-27 4' Radiation Protection Facilities XII-27 4 4
'.1 '.2 Ventilation Air Flows Instrument Calibration Well XII-27 Shielding XII-27 4 ' Radiation Protection Instrumentation XII-27 SECTION XIII CONDUCT OF OPERATIONS XIII-1 A. ORGANIZATION AND RESPONSIBILITY XIII-1 1.0 Management and Technical Support Organization XIII-1 1~1 Nuclear Division XIII-1 1~1~1 Vice President and General Manager Nuclear XIII-1 1.1 ' Vice President Nuclear Engineering XIII-2 UFSAR Revision 14 X1X June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 1.1.3 Vice President Nuclear Safety Assessment and Support XIII-2 1.1.4 Director Nuclear Communications and Public Affairs XIII-4 1 1~5
~ Manager Human Resource Development XIII-4 1.1.6 General Manager Business Management XIII-4 1.2 Corporate Support Departments XIII-4 2.0 Operating Organization XZZI-5 2.1 Plant Manager XIII-5 2' General Manager Business Management XIII-8 3.0 Quality Assurance XIII-8 4.0 Facility Staff Qualifications XIII-8 B. QUALIFICATIONS AND TRAINING OF PERSONNEL XIII-9 1.0 This Section Deleted XIII-9 2.0 This Section Deleted XIII-9 3.0 This Section Deleted XIII-9 4.0 Training of Personnel XIII-9 4.1 General Responsibility XIII-9 4.2 Implementation XIII-9 4.3 Quality XIII-9 4.3.1 For Operator Training XIII-9 4.3.2 For Maintenance XIII-9 4.3.3 For Technicians XIII-10 4.3.4 For General Employee Training/Radiation Protection and Emergency Plan XIII-10 4.3.5 For Industrial Safety XIII-10 4.3.6 For Nuclear Quality Assurance XIII-10 4.3 ' For Fire Brigade XIII-10 4' Training of Licensed Operator Candidates/Licensed NRC Operator Retraining XIII-10 5.0 Cooperative Training with Local, State and Federal Officials XIII-11 C. OPERATING PROCEDURES XIII-12 D. EMERGENCY PLAN AND PROCEDURES XIII-13 E. SECURITY XIII-15 F. RECORDS XIII-16 1.0 Operations XIII-16 1.1 Control Room Log Book XIII-16 1' Station Shift Supervisor's Book XIII-16 1.3 Radwaste Log Book XIII-16 1.4 Waste Quantity Level Shipped XIII-16 UFSAR Revision 14 XX June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 2.0 Maintenance XIII-16 3.0 Radiation Protection XIII-17 3' Personnel Exposure XIII-17 3.2 By-Product Material as Required by 10CFR30 XIII-17 3.3 Meter Calibrations XIII-17 3.4 Station Radiological Conditions in Accessible Areas XIII-17 3.5 Administration of the Radiation Protection Program and Procedures XIII-17 4.0 Chemistry and Radiochemistry XIII-17 5.0 Special Nuclear Materials XIII-17 6.0 Calibration of Instruments XIII-17 7.0 Administrative Records and Reports XIII-17 G. REVIEW AND AUDIT OF OPERATIONS XIII-19 1.0 Station Operations Review Committee XIII-19 1.1 Function XIII-19 2 ' Safety Review and Audit Board XIII-19 F 1 Function XIII-19 3.0 Review of Operating Experience XIII-20 SECTION XIV INITIAL TESTING AND OPERATIONS XIV-1 A. TESTS PRIOR TO INITIAL REACTOR FUELING XIV-1 B. INITIAL CRITICALITY AND POSTCRITICALITY TESTS XIV-5 1' Initial Fuel Loading and Near-Zero Power Tests at Atmospheric Pressure XIV-5 1.1 General Requirements XIV-5 1.2 General Procedures XIV-5 1.3 Core Loading and Critical Test Program XIV-7 2.0 Heatup from Ambient to Rated Temperature XIV-9 2.1 General XIV-9 2' Tests Conducted XIV-9 3.0 From Zero to 100 Percent Initial Reactor Rating XIV-10 4.0 Full-Power Demonstration Run XIV-12 5.0 Comparison of Base Conditions XIV-12 6.0 Additional Tests at Design Rating XIV-13 SECTION XV SAFETY ANALYSIS XV-1 A. INTRODUCTION XV-1 UFSAR Revision 14 xxi June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic Title Pacae B. BOUNDARY PROTECTION SYSTEMS XV-2 1.0 Transients Considered XV-2 2.0 Methods and Assumptions XV-3 3.0 Transient Analysis XV-3 3.1 Turbine Trip Without Bypass XV-3 3' ' Objectives XV-3 3.1.2 Assumptions and Initial Conditions XV-3 3' ' Comments XV-3 3.1.4 Results XV-3 3.2 Loss of 100'F Feedwater Heating XV-4 3.2.1 Objectives XV-4 3' ' Assumptions and Initial Conditions XV-4 XV-4 3.2 ' Results 3' Feedwater Controller Failure Maximum Demand XV-5 3.3.1 Objectives XV-5 3.3.2 Assumptions and Initial Conditions XV-5 XV-5 3.3.3 Comments 3.3.4 Results XV-5 3' Control Rod Withdrawal Error XV-5 3.4.1 Objectives XV-5 3.4.2 Assumptions and Initial Conditions XV-5 3.4.3 Comments XV-6 3.4.4 Results XV-6 3.5 Main Steam Line Isolation Valve Closure (With Scram) XV-6 3.5.1 Objectives XV-6 3.5.2 Assumptions and Initial Conditions XV-6 XV-7 3.5.3 Comments 3.5.4 Results XV-7 3.6 Inadvertent Startup of Cold Recirculation Loop XV-7 3.6.1 Objectives XV-7 3.6 2
~ Assumptions and Initial Conditions XV-7 XV-8 3.6.3 Comments 3.6.4 Results XV-9 3.7 Recirculation Pump Trips XV-9 3.F 1 Objectives XV-9 3.7.2 Assumptions and Initial Conditions XV-9 XV-9 3' ' Comments 3.7.4 Results XV-10 3.8 Recirculation Pump Stall XV-10 3.8.1 Objectives XV-10 3.8.2 Assumptions and Initial Conditions XV-10 XV-10 3.8.3 Comments 3.8.4 Results XV-11 3.9 Recirculation Flow Controller Malfunction Increase Flow XV-11 3.9.1 Objectives XV-11 UFSAR Revision 14 xxii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectio Title Pacae 3.9.2 Assumptions and Initial Conditions XV-11 XV-11 3.9.3 Comments 3.9.4 Results XV-11 3 '0 Flow Controller Decrease Flow Malfunction XV-12 3.10.1 Objectives XV-12 3.10 ' Assumptions and Initial Conditions XV-12 XV-12 3.10.3 Comments 3.10 ' Results XV-12 3 '1 Inadvertent Actuation of Solenoid Relief Valve One XV-12
- 3. 11 1 Objectives XV-12 Initial Conditions
~
- 3. 11.2 Assumptions and XV-12
- 3. 11. 3 Comments XV-13
- 3. 11. 4 Results XV-13 3 ~ 12 Safety Valve Actuation (Overpressurization Analysis) XV-13 3.12.1 Objectives XV-13 3.12.2 Assumptions and Initial Conditions XV-13 3.12.3 Comments XV-14 3.12.4 Results XV-14 3 '3 Feedwater Controller Malfunction (Zero Demand) XV-15 3.13 ' Objectives XV-15 3.13.2 Assumptions and Initial Conditions XV-15 XV-15 3.13.3 Comments 3.13.4 Results XV-15
- 3. 14 Turbine Trip with Partial Bypass (Low Power) XV-16 3.14.1 Objectives XV-16 3.14.2 Assumptions and Initial Conditions XV-16 XV-16 3.14.3 Comments 3 '4.4 Results XV-16 3.15 Turbine Trip with Partial Bypass (Full Power) XV-17 3.15.1 Objectives XV-17 3.15.2 Assumptions and Initial Conditions ~
XV-17 XV-17 3.15.3 Comments 3.15.4 Results XV-17 3.16 Inadvertent Actuation of One Bypass Valve XV-18 3.16.1 Objectives XV-18 3.16.2 Assumptions and Initial Conditions XV-18 XV-18 3.16.3 Comments 3.16.4 Results XV-18
- 3. 17 One Feedwater Pump Trip and Restart XV-18 3.17 ' Objectives XV-18 3.17.2 Assumptions and Initial Conditions XV-18 XV-19 3.17.3 Comments UFSAR Revision 14 Xxiii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae
- 3. 17.4 Results XV-19
- 3. 18 Loss of Main Condenser Vacuum XV-19
- 3. 19 Loss of Electrical Load (Generator Trip) XV-19
- 3. 19. 1 Objectives XV-19 3.19.2 Assumptions and Initial Conditions XV-19 XV-20 3.19.3 Comments 3.19 ' Results XV-20 3.20 Loss of Auxiliary Power XV-20 3.20.1 Objectives XV-20 3.20.2 Assumptions and Initial Conditions XV-20 XV-20 3.20.3 Comments 3.20.4 Results XV-20 3.21 Pressure Regulator Malfunction XV-21 3.21.1 Objectives XV-21 3.21.2 Assumptions and Initial Conditions XV-21 3.21.3 Comments XV-21 3.21.4 Results XV-21 3.22 Instrument Air Failure XV-22 3.22.1 Objectives XV-22 3.22 ' Assumptions and Initial Conditions XV-22 XV-22 3.22.3 Comments 3.22.4 Results XV-22 3.23 Dc Power Interruptions XV-26 3.23.1 Objectives XV-26 3.23.2 Assumptions and Initial Conditions XV-26 XV-26 3.23.3 Comments 3.23.4 Results XV-26 3.24 Failure of One Diesel Generator to Start XV-27 3.24.1 Objectives XV-27 3.24.2 Assumptions and Initial Conditions XV-27 XV-27 3.24.3 Comments 3.24.4 Results XV-27 3.25 Power Bus Loss of Voltage XV-27 3.25.1 Objectives XV-27 3.25.2 Assumptions and Initial Conditions XV-27 XV-28 3.25.3 Comments 3.25.4 Results XV-28 C. STANDBY SAFEGUARDS ANALYSIS XV-29 1.0 Main Steam Line Break Outside the Drywell XV-29 1.1 Identification of Causes XV-29 1.2 Accident Analysis XV-29 1.2.1 Valve Closure Initiation XV-30 1.2 ' Feedwater Flow XV-30 1.2.3 Core Shutdown XV-30 1.2.4 Mixture Level XV-30 UFSAR Revision 14 xxiv June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectio Title Pacae 1.2.5 Subcooled Liquid XV-30 1.2.6 System Pressure and Steam-Water Mass XV-3 1 1.2.7 Mixture Impact Forces XV-3 1 1.2.8 Core Internal Forces XV-31 1.3 Radiological Effects XV-31 1.3.1 Radioactivity Releases XV-32 1' ' Meteorology and Dose Rates XV-32 1' ' Comparison with Regulatory Guide 1.5 XV-33 2.0 Loss-of-Coolant Accident XV-34 2.1 Introduction XV-34 2.2 Input to Analysis XV-35 2.2.1 Operational and ECCS Input Parameters XV-35 2.2.2 Single Failure Study on ECCS Manually-Controlled Electrically-Operated Valves XV-35 2.2.3 Single Failure Basis XV-35 2.2.4 Pipe Whip Basis XV-3 6 2.3 Deleted XV-36 2.4 Appendix K LOCA Performance Analysis XV-3 6 2.4. 1 Computer Codes XV-3 6 2' ' Description of Model Changes XV-37 2.4.3 Analysis Procedure XV-37 2.4.3.1 BWR/2 Generic Analysis XV-37 2.4.3.2 Unit 1 Specific Analysis Break Spectrum Evaluation XV-38 2.4.4 Analysis Results XV-38 3.0 Refueling Accident XV-40 3' Identification of Causes XV-40 3 ' Accident Analysis XV-41 3 ' Radiological Effects XV-44 3.3 ' Fission Product Releases XV-44 3.3.2 Meteorology and Dose Rates XV-45 3'3 ' Comparison to Regulatory Guide 1.25 XV-45 4.0 Control Rod Drop Accident XV-45 4.1 Identification of Causes XV-45 4.2 Accident Analysis XV-46 4.3 Designed Safeguards XV-46 4.4 Procedural Safeguards XV-47 4.5 Radiological Effects XV-47 4.5.1 Fission Product Releases XV-48 4.5.2 Meteorology and Dose Rates XV-50 5.0 Containment Design Basis Accident XV-50 5.1 Original Recirculation Line Rupture Analysis With Core Spray XV-50 5.1.1 Purpose XV-50 UFSAR Revision 14 xxv June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic ~itic Pacae
- 5. 1.2 Analysis Method and Assumptions XV-51 5.1.3 Core Heat Buildup XV-51 5.1.4 Core Spray System XV-52 5.1.5 Containment Pressure Immediately Following Blowdown XV-53 5.1.6 Containment Spray XV-54 5 1.7 Blowdown Effects on Core Components XV-55 5 '.8
~
5.1.8.1 Radiological Effects Fission Product Releases XV-56 XV-56 5 '.8.2 Meteorology and Dose Rates Original Containment Design Basis XV-59 5.2 Accident Analysis Without Core Spray XV-59 5.2. 1 Purpose XV-59 5.2.2 Core Heatup XV-59 5.2.3 Containment Response XV-60 5.2.4 Fission Product Release from the Fuel XV-61 5.2.5 Fission Product Release from the Reactor and Containment XV-61 5.2.6 Meteorology and Dose Rates XV-61 5.3 Design Basis Reconstitution Suppression Chamber Heatup Analysis XV-61 5.3.1 Introduction XV-61 5.3.2 Input to Analysis XV-62 5.3.3 DBR Suppression Chamber Heatup Analysis XV-63 5.3.3.1 Computer Codes XV-63 5.3.3.2 Analysis Methods XV-63 5.3.3.3 Analysis Results for Containment Spray Design Basis Assumptions XV-64 5.3.3.4 Analysis Results for EOP Operation Assumptions XV-65 5.3.4 Conclusions XV-66 6.0 New Fuel Bundle Loading Error Analysis XV-66 6.1 Identification of Causes XV-66 6.2 Accident Analysis XV-67 6.3 Safety Requirements XV-67 7.0 Meteorological Models Used in Accident Analyses XV-68 7 ' Ground Releases XV-68 7 ' Stack Releases XV-68 7.3 Variability XV-69 7.4 Exfiltration XV-70 7.5 Ground Deposition XV-76 7.6 Thyroid Dose XV-77 7.7 Whole Body Dose XV-77 UFSAR Revision 14 xxvi June 1996
Nine Mile Point Unit 1 FSAR TABLE. OF CONTENTS (Cont'd.) Section Title Pacae D. REFERENCES XV-79 SECTION XVI SPECIAL TOPICAL REPORTS XVI-1 A. REACTOR VESSEL XVI-1 1~0 Applicability of Formal Codes and Pertinent Certifications XVI-1 2.0 Design Analysis XVI-2 2.1 Code Approval Analysis XVI-2 2.2 Steady-State Analysis XVI-3 2.2.1 Basis for Determining Stresses XVI-3 2.3 Pipe Reaction Calculations XVI-4 2.4 Earthquake Loading Criteria and Analysis XVI-4 2.4.1 Seismic Analysis for Core Shroud Repair Modification XVI-5 2.5 Reactor Vessel Support Stress Design Criteria and Analysis XVI-5 2.6 Strain Safety Margin for Reactor Vessels XVI-7 2.6.1 Introduction XVI-7 2.6.2 Strain Margin XVI-8 2.6.3 Failure Probability XVI-9 2.6.4 Results of Probability Analysis XVI-11 2.6.5 Conclusions XVI-11 2.7 Components Required for Safe Reactor Shutdown XVI-11 2.7 ' Design Basis Load Combinations XVI-12 2.7.2 Expected Stress and Deformation XVI-12 2.7 ' ' Recirculation Line Break XVI-12 2' ' ' Steam Line Break XVI-13 2.7.2.3 Earthquake Loadings XVI-13 2.7.3 Stresses and Deformations at Which the Component is Unable to Function and Margin of Safety XVI-14 2.7.3. 1 Recirculation Line Break XVI-14 2.7.3.2 Steam Line Break XVI-15 2.8 Safety Margins Against Ductile Fracture XVI-17 3.0 Inspection and Test Report Summary XVI-18 3.1 Materials XVI-18 3.2 Fabrication and Inspection XVI-18 4.0 Surveillance Provisions XVI-20 4.1 Coupon Surveillance Program XVI-20 4.2 Periodic Inspection XVI-21 5.0 Core Shroud Stabilizer Design Description XVI-21 UFSAR Revision 14 xxvii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~sectic Title Pacae B. PRESSURE SUPPRESSION CONTAINMENT XVI-22 1.0 Applicability of Formal Codes and Pertinent Certifications XVI-22 2.0 Design Analysis XVI-23 2.1 Code Approval Calculations Under Rated Conditions XVI-23 2 ' Ultimate Capability Under Accident Conditions XVI-23 2.3 Capability to Withstand Internal Missiles and Jet Forces XVI-23 2.4 Flooding Capabilities of the Containment XVI-24 2.5 Drywell Air Gap XVI-25 2.5.1 Tests and Inspections XVI-26 2.6 Biological Shield Wall XVI-26 2.7 Compatibility of Dynamic Deformations Occurring in the Drywell, Torus, and Connecting Vent Pipes XVI-28 2.8 Containment Penetrations XVI-30 2.8.1 Classification of Penetrations XVI-30 2.8.2 Design Bases XVI-30 2.8 ' Method of Stress Analysis XVI-31 2.8.4 Leak Test Capability XVI-31 2.8.5 Fatigue Design XVI-31 2.8.6 Material Specification XVI-32 2.8.7 Applicable Codes XVI-32 2.8.8 Jet and Reaction Loads XVI-33 2.9 Drywell Shear Resistance Capability and Support Skirt Junction Stresses XVI-33 3.0 Inspection and Test Report Summary XVI-34 3.1 Fabrication and Inspection XVI-34 3.2 Tests Conducted XVI-34 3' Discussion of Results XVI-36 3~3 ~ 1 Results XVI-36 3.3.2 Effect of Various Transients XVI-36 3.3.2-1 Ambient Temperature and Solar Heating of Shell XVI-36 3.3.2.2 Thermal Lag Through Reference Chamber Wall XVI-37 3.3.2.3 Condensation in Reference Chamber XVI-37 3.3.2.4 Volume Changes Due to Thermal Transients XVI-37 3 '.2.5 Overpressure Test Plate Stresses XVI-38 C. ENGINEERED SAFEGUARDS XVI-39 1.0 Seismic Analysis and Stress Report XVI-39 1.1 Introduction XVI-39 1' Mathematical Model XVI-40 UFSAR Revision 14 XXViii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~eectic Title Pacae 1.3 Method of Analysis XVI-40 1.3.1 Flexibility or Influence XVI-41 Coefficient Matrix 1.3'2 Normal Mode Frequencies and Mode Shapes XVI-41 1.3 ~ 3 The Seismic Spectrum Values XVI-42 1.3 ~ 4 Dynamic Modal Loads XVI-43 1.3.5 Modal Response Quantities XVI-43 1.3.6 The Combined Response Quantities XVI-43 1' ' Basic Criteria for Analysis XVI-44 1.4 Discussion of Results XVI-44 2.0 Containment Spray System XVI-45 2.1 Design Adequacy at Rated Conditions XVI-45 2.1 ~ 1 General XVI-45 2.1.2 Condensation and Heat Removal Mechanisms XVI-45 2.1 3
~ Mechanical Design XVI-50
- 2. 1.4 Loss-of-Coolant Accident XVI-51 2.2 Summary of Test Results XVI-52 2.2 ' Spray Tests Conducted XVI-52 D. I DES GN OF STRUCTURES g COMPONENTS I XVI-53 EQUIPMENT, AND SYSTEMS 1.0 Classification and Seismic Criteria XVI-53 1~1 Design Techniques XVI-55 1~1~1 Structures XVI-55 1.1 2
~ Systems and Components XVI-58 1.2 Pipe Supports XVI-59 1.3 Seismic Exposure Assumptions XVI-60 2.0 Plant Design for Protection Against Postulated Piping Failures in High-Energy Lines XVI-61 2~1 Inside Primary Containment XVI-61 2.1 ~ 1 Containment. Integrity Analysis XVI-61 2.1.1.1 Fluid Forces XVI-62
- 2. 1. 1. 2 Impact Velocities and Effects XVI-62 2.1 ~ 2 Systems Affected by Line Break XVI-63 2.1 ~ 3 Engineered Safeguards Protection XVI-67 2.2 Outside Primary Containment XVI-69 3.0 Building Separation Analysis XVI-69 4.0 Tornado Protection XVI-69 E. EXHIBITS XVI-72 F. CONTAINMENT DESIGN REVIEW XVI-110 G. USAGE OF CODES/STANDARDS FOR STRUCTURAL STEEL AND CONCRETE XVI-121 UFSAR Revision 14 xxix June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae H. REFERENCES XVI-122 SECTION XVII ORIGINAL ENVIRONMENTAL STUDIES XVII-1 A. METEOROLOGY XVII-1 1.0 General XVII-1 2.0 Synoptic Meteorological Factors XVII-2 3.0 Micrometeorology XVII-2 3.1 Wind Patterns XVII-2 3.1 ' 200-Foot Wind Roses XVII-2 3.1.2 Estimates of Winds at the 350-Foot Level XVII-2 3 ' ' Comparison Between Tower and Satellite Winds XVII-16 3.2 Lapse Rate Distributions XVII-19 3.3 Turbulence Classes XVII-19 3.4 Dispersion Parameters XVII-19 3.4.1 Changes in Dispersion Parameters XVII-39 4.0 Applications to Release Problems XVII-45 4.1 Concentrations from a Ground-Level Source XVII-46 4.2 Concentrations from an Elevated Source XVII-53 4.3 Radial Concentrations XVII-55 4.3.1 Monthly and Annual Sector Concentrations XVII-55 4 ' Least Favorable Concentrations Over an Extended Period XVII-83 4.4. 1 Ground-Level Release XVII-83 4.4.2 Elevated Release XVII-86 4.5 Mean Annual Sector Deposition XVII-87 4.6 Dose Rates from a Plume of Gamma Emitters XVII-90 4.6.1 RADOS Program XVII-90 4.6.2 Centerline Dose Rates XVII-91 4.6.3 Sector Dose Rates XVII-100 4.7 Concentrations from a Major Steam Line Break XVII-103 5.0 Conclusions XVII-106 B. LIMNOLOGY XVII-107 1' Introduction XVII-107 2.0 Summary Report of Cruises XVII-107 3.0 Dilution of Station Effluent in Selected Areas XVII-109 3.1 Dilution of Effluent at the Lake Surface Above the Discharge XVII-109 3.2 Dilution of Effluent at the Site Boundaries XVII-114 UFSAR Revision 14 XXX June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic Title Pacae 3~2~1 General XVII-114 3.2.2 Dilution of Effluent at the Eastern Site Boundary XVII-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 XVII-123 3.3 ' Tilting of the Isothermal Planes and Subsequent Dilution XVII-123 3 ' ' Dilution as a Function of Current Velocity XVII-124 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 XVII-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 XVII-130 C. EARTH SCIENCES XVII-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' Seismological Conditions XVII-142 4.5 Conclusion XVII-142 SECTION XVIII HUMAN FACTORS ENGINEERING/SAFETY PARAMETER DISPLAY SYSTEM XVIII-1 A. DETAILED CONTROL ROOM DESIGN REVIEW XVIII-1 1.0 General XVIII-1 UFSAR Revision 14 xxxi June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) ~Sectic Title Pacae 2.0 Planning Requirements for the DCRDR XVIII-1 3.0 DCRDR Review Process XVIII-2 3' Operator Survey XVIII-2 3.2 Historical Review XVIII-2 3.3 Task Analysis XVIII-3 3.4 Control Room Inventory XVIII-3 3.5 Control Room Survey XVIII-3 3.6 Verification of Task Performance Capabilities XVIII-3 3.7 Validation of Control Room Functions XVIII-4 3.8 Compilation of Discrepancy Findings XVIII-4 4.0 Assessment and Implementation XVIII-4 4.1 Assessment XVIII-4 4.2 Implementation XVIII-5 4.2. 1 Integrated Cosmetic Package XVIII-5 4.2.2 Functional Fixes XVIII-6 5.0 Reporting XVIII-6 6.0 Continuing Human Factors Program XVIII-6 6.1 Fix Verifications XVIII-7 6.2 Multidisciplinary Review Team Assessments XVIII-7 6.3 Human Factors Manual for Future Design Change XVIII-7 6.4 Outstanding Human Factors Items XVIII-7 7.0 References XVIII-8 B. SAFETY PARAMETER DISPLAY SYSTEM XVIII-10 1.0 Introduction to the Safety Parameter Display System XVIII-10 2.0 System Description XVIII-10 3.0 Role of the SPDS XVIII-11 4.0 Human Factors Engineering Guidelines XVIII-11 5.0 Human Factors Engineering Principles Applied to the SPDS Design XVIII-11 5.1 NUREG-0737, Supplement 1, Section 4.1.a XVIII-12 5.1.1 Concise Display XVIII-12
- 5. 1.2 Criteria Plant Variables XVIII-12 5.1.3 Rapid and Reliable Determination of Safety Status XVIII-12 5.1.4 Aid to Control Room Personnel XVIII-12 5.2 NUREG-0737, Supplement 1, Section 4.1.b XVIII-13 5.2.1 Convenient Location XVIII-13 5.2.2 Continuous Display XVIII-13 UFSAR Revision 14 xxxii June 1996
Nine Mile Point Unit 1 FSAR TABLE OF CONTENTS (Cont'd.) Section Title Pacae 5.3 NUREG-0737, Supplement 1, Section 4.1.c XVIII-13 5.3.1 Procedures and Training XVIII-13 5.3.2 Isolation of SPDS from Safety-Related Systems XVIII-13 5.4 NUREG-0737, Supplement 1, Section 4.1.e XVIII-14 5.4.1 Incorporation of Accepted Human Factors Engineering Principles XVIII-14 5.4.2 Information Can be Readily Perceived and Comprehended XVIII-14 5.5 NUREG-0737, Supplement 1, Section 4.1.f, Sufficient Information XVIII-15 6.0 Procedures XVIII-15 6.1 Operating Procedures XVIII-15 6.2 Surveillance Procedures XVIII-15 7.0 References XVIII-16 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 14 XXXiii June 1996
Nine Mile Point Unit 1 FSAR LIST OF TABLES Table ~Nu ber Title II-1 1980 Population and Population Density for Towns and Cities Within 12 Miles of Nine Mile Point Unit 1 II-2 Cities Within a 50-mile Radius of the Station With Populations over 10,000 II-3 Regional Agricultural Use II-4 Regional Agricultural Statistics Cattle and Milk Production II-5 Industrial Firms Within 8 km (5 mi) of Unit 1 II-6 Public Utilities in Oswego County II-7 Public Water Supply Data for Locations Within an Approximate 30-Mile Radius II-8 Recreational Areas in the Region V-1 Reactor Coolant System Data V-2 Operating Cycles and Transient Analysis Results V-3 Fatigue Resistance Analysis V-4 Codes for Systems Connected to the Reactor Coolant System V-5 Time to Automatic Blowdown VI-1 Drywell Penetrations VI-2 Suppression Chamber Penetrations VZ-3a Reactor Coolant System Isolation Valves VI-3b Primary Containment Isolation Valves Lines Entering Free Space of the Containment VI-4 Seismic Design Criteria for Isolation Valves VI-5 Initial Tests Prior to Station Operation VII-1 Performance Tests VIIZ-1 Association Between Primary Safety Functions and Emergency Operating Procedures VIII-2 List of EOP Key Parameters VIII-3 Type and Instrument Category for Unit 1 RG 1.97 Variables ZX-1 Magnitude and Duty Cycle of Major Station Battery Loads XII-1 Flows and Activities of Major Sources of Gaseous Activity XII-2 Quantities and Activities of Liquid Radioactive Wastes XII-3 Annual Solid Waste Accumulation and Activity XII-4 Liquid Waste Disposal System Major Components XII-5 Solid Waste Disposal System Major Components XII-6 Occupancy Times UFSAR Revision 14 xxxiv June 1996
Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.) Table Number Title XII-7 Gamma Energy Groups XII-8 Area Radiation Monitor Detector Locations XIII-1 ANSI Standard Cross-Reference Unit 1 XV-1 Transients Considered XV-2 Trip Points for Protective Functions XV-3 Table Deleted XV-4 Instrument Air Failure XV-5 Blowdown Rates XV-6 Iodine Concentrations (pCi/gm) XV-7 Fractional Concentrations in Clouds XV-8 Main Steam Line Break Accident Doses XV-.9 Significant Input Parameters to the Loss-of-Coolant Accident Analysis XV-9A Core Spray System Flow Performance Assumed in LOCA Analysis XV-10 ECCS Single Valve Failure Analysis XV-ll Single Failures Considered in LOCA Analysis XV-12 Table Deleted XV-13 Table Deleted XV-14 Table Deleted XV-15 Table Deleted XV-16 Table Deleted XV-17 Table Deleted XV-18 Table Deleted XV-19 Table Deleted XV-20 Table Deleted XV-21 Table Deleted XV-21A Analysis Assumptions For Nine Mile Point 1 Calculations XV-21B Table Deleted XV-21C Table Deleted XV-21D Table Deleted XV-21E Table Deleted XV-22 Reactor Building Airborne Fission Product Inventory (curies) XV-23 Stack Discharge Rates (curies/sec) XV-24 Fuel Handling Accident Doses (REM) XV-25 Fission Product Release Assumptions XV-26 Atmospheric Dispersion and Dose Conversion Factors XV-27 Effect on Dose of Factors Used in the Calculations XV-28 Noble Gas Release XV-29 Halogen Release XV-29a Wetting of Fuel Cladding by Core Spray XV-29b Airborne Drywell Fission Product Inventory (curies) UFSAR Revision 14 xxxv June 1996
Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.) Table Number Title XV-29c Reactor Building Airborne Fission Product Inventory (curies) XV-29d Stack Discharge Rates (curies/sec) XV-30 Airborne Drywell Fission Product Inventory (curies) XV-31 Reactor Building Airborne Fission Product Inventory (curies) XV-32 Stack Discharge Rates (curies/sec) XV-32a Significant Input Parameters to the DBR Containment Suppression Chamber Heatup Analysis XV-3 3 Downwind Ground Concentrations XV-34 Maximum Ground Concentrations XV-35 Diversity Factors for Ground Concentrations XV-3 6 Reactor Building Leakage Paths XVI-1 Code Calculation Summary XVI-2 Steady-State (1004 Full Power Normal Operation) Pertinent Stresses or Stress Intensities XVZ-3 List of Reactions for Reactor Vessel Nozzles XVI-4 Effect of Value of Initial Failure Probability XVI-5 Single Transient Event for Reactor Pressure Vessel XVI-6 Postulated Events XVZ-7 Maximum Strains from Postulated Events XVI-8 Core Structure Analysis Recirculation Line Break XVI-9 Core Structure Analysis Steam Line Break XVZ-9a Core Shroud Repair Design Supporting Documentation XVI-10 Drywell Jet and Missile Hazard Analysis Data XVI-11 Drywell Jet and Missile Hazard Analysis Results XVI-12 Stress Due to Drywell Flooding XVI-13 Allowable Weld Shear Stress XVI-14 Leak Rate Test Results XVI-15 Overpressure Test Plate Stresses XVI-16 Stress Summary XVI-17 Heat Transfer Coefficients as a Function of Drop Diameter XVI-18 Heat Transfer Coefficient as a Function of Pressure XVZ-19 Relationship Between Particle Size and Type of Spray Pattern XVI-20 Allowable Stresses for Floor Slabs, Beams, Columns, Walls, Foundations, etc. XVI-21 Allowable Stresses for Structural Steel XVI-22 Allowable Stresses Reactor Vessel Concrete Pedestal XVI-23 Drywell Analyzed Design Load Combinations XVI-24 Suppression Chamber Analyzed Design Load Combinations UFSAR Revision 14 xxxvi June 1996
Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.) Table Number Title XVI-25 ACI Code 505 Allowable Stresses and Actual Stresses for Concrete Ventilation Stack XVI-26 Allowable Stresses for Concrete Slabs, Walls, Beams, Structural Steel, and Concrete Block Walls XVI-27 System Load Combinations XVI-28 High-Energy Systems Inside Containment XVI-29 High-Energy Systems Outside Containment XVI-30 Systems Which May Be Affected by Pipe Whip XVI-31 Capability to Resist Wind Pressure and Wind Velocity XVII-1 Dispersion and Associated Meteorological Parameters XVII-2 Relation of Satellite and Nine Mile Point Winds XVII-3 Frequency of Occurrence of Lapse Rates 1963 and 1964 XVII-4 Relation Between Wind Direction Range and Turbulence Classes XVII-5 Stack Characteristics XVII-6 Distribution of Turbulence Classes By Sectors XVII-7 Sector Concentrations 1963-64 Sector A Elev. 350 XVII-8 Sector Concentrations 1963-64 Sector B Elev. 350 XVII-9 Sector Concentrations 1963-64 Sector C Elev. 350 XVII-10 Sector Concentrations 1963-64 Sector D, Elev. 350 XVII-11 Sector Concentrations 1963-64 Sector D~ Elev. 350 XVII-12 Sector Concentrations 1963-64 Sector E Elev. 350 XVII-13 Sector Concentrations 1963-64 Sector F Elev. 350 XVII-14 Sector Concentrations 1963-64 Sector G Elev. 350 XVII-15 Sector Concentrations 1963-64 Sector A Ground Height XVII-16 Sector Concentrations 1963-64 Sector B Ground Height XVII-17 Sector Concentrations 1963-64 Sector C Ground Height XVII-18 Sector Concentrations 1963-64 Sector D, Ground Height XVII-19 Sector Concentrations 1963-64 Sector Dz Ground Height XVII-20 Sector Concentrations 1963-64 Sector E Ground Height UFSAR Revision 14 XXXVii June 1996
Nine Mile Point Unit 1 FSAR LIST OF TABLES (Cont'd.) Table Number Title XVII-21 Sector Concentrations 1963-64 Sector F Ground Height XVII-22 Sector Concentrations 1963-64 Sector G Ground Height XVII-23 Estimates of the Least Favorable 30 Days in 100 Years XVII-24 Concentrations in the Least Favorable Calendar Month 1963-64 XVII-25 Annual Average Sector Deposition Rates (Vg = 0.5 cm/sec) XVII-26 Annual Average Sector Deposition Rates (Vg = 2.5 cm/sec) XVII-27 Principal Radionuclides in Gaseous Waste Release XVII-28 Correction Factors to Obtain Adjusted Centerline Dose Rates for Sector Estimates XVII-29 Annual Average Gamma Dose Rates XVII-30 Dilution Calculation for Eastward Currents Based on Water Availability XVIII-1 SPDS Parameter Set UFSAR Revision 14 xxxviii June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES Figure Numine Title Piping, Instrument and Equipment Symbols II-1 Station Location II-2 Area Map II-3 Site Topography II-4 Population Distribution Within a 12 Mile Radius of the Station II-5 Counties and Towns Within 12 Miles of the Station II-6 1980 Population Distribution Within a 50 Mile Radius of the Station III-1 Plot Plan III-2 Station Floor Plan Elevation 225-6 III-3 Station Floor Plan Elevations 237-0 and 250-0 III-4 Station Floor Plan Elevation 261-0 III-5 Station Floor Plan Elevations 277-0 and 281-0 III-6 Station Floor Plan Elevations 281-0 and 291-0 III-7 Station Floor Plan Elevations 298-0 and 300-0 III-8 Station Floor Plan Elevations 317-6 and 318-0 III-9 Station Floor Plan Elevations 320-0, 333-8, 340-0 and 369-0 III-10 Section Between Column Rows 7 and 8 III-11 Section Between Column Rows 12 and 14 III-12 Turbine Building Ventilation System III-13 Laboratory and Radiation Protection Facility Ventilation System IZZ-14 Control Room Ventilation System III-15 Waste Disposal Building Ventilation System III-16 Waste Disposal Building Extension Ventilation System III-17 Off Gas Building Ventilation System III-18 Technical Support Center Ventilation System III-19 Circulating Water Channels Under Screen and Pump House Normal Operation III-20 Circulating Water Channels Under Screen and Pump House Special Operations III-21 Intake and Discharge Tunnels Plan and Profile III-22 Stack Plan and Elevation III-23 Stack Failure Critical Directions IV-1 Limiting Power/Flow Line (Typical) IV-2 Figure Deleted IV-3 Figure Deleted IV-4 Typical Control Rod Isometric IV-5 Figure Deleted IV-6 Control Rod Drive and Hydraulic System IV-7 Control Rod Drive Assembly IV-8 Typical Control Rod to Drive Coupling Isometric UFSAR Revision 14 xxxix June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title IV-9 Reactor Vessel Isometric V-1 Reactor Emergency Coolant System V-2 Reactor Vessel Nozzle Location V-3 Reactor Vessel Support V-4 Figure Deleted V-5 Pressure Vessel Embrittlement Trend V-6 Figure Deleted V-7 Figure Deleted V-8 Emergency Condenser Supply Isolation Valves (Typical of 2) VI-1 Drywell and Suppression Chamber VI-2 Electrical Penetrations High Voltage VI-3 Electrical Penetrations Low Voltage VI-4 Pipe Penetrations. Hot VI-4a Clamshell Expansion Joint VI-5 Typical Penetration For Instrument Lines VI-6 Reactor Building Dynamic Analysis Acceleration East-West Direction VI-7 Reactor Building Dynamic Analysis Deflections East-West Direction VI-8 Reactor Building Dynamic Analysis Elevation vs. Building Shear East-West Direction VI-9 Reactor Building Dynamic Analysis Elevation vs. Building Moment East-West Direction VI-10 Reactor Building Dynamic Analysis Acceleration North-South Direction VI-11 Reactor Building Dynamic Analysis Deflections North-South Direction VI-12 Reactor Building Dynamic Analysis Elevation vs. Building Shear North-South Direction VI-13 Reactor Building Dynamic Analysis Elevation vs. Building Moment North-South Direction VI-14 Reactor Support Dynamic Analysis Elevation vs. Acceleration VI-15 Reactor Support Dynamic Analysis Elevation vs. Deflection VI-16 Reactor Support Dynamic Analysis Elevation vs. Shear VI-17 Reactor Support Dynamic Analysis Elevation vs. Moment VI-18 Typical Door Seals VI-19 Details of Reactor Building Air Locks VI-20 Instrument Line Isolation Valve Arrangement VI-21 Typical Flow Check Valve VI-22 Isolation Valve System VI-23 Drywell Cooling System UFSAR Revision 14 xl June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title VI-24 Reactor Building Ventilation System VII-1 Core Spray System VII-2 Core Spray Sparger Flow, Per Sparger, for One Core Spray Pump and One Topping Pump VII-3 Containment Spray System VIX-4 Figure Deleted VII-5 Figure Deleted VII-6 Liquid Poison System VII-7 Minimum Allowable Solution Temperature VII-8 Figure Deleted VII-9 Typical Control Rod Velocity Limiter VII-10 Control Rod Housing Support VII-11 Hydrogen Flammability Limits VXI-12 Combustible Gas Control System VII-13 H~-O, Sampling System VII-14 Hydrogen and Oxygen Concentrations in Containment Following Loss of Coolant Accident VII-15 Nitrogen Added by Containment Atmospheric Dilution Operation Following Loss of Coolant Accident VII-16 Containment Pressure with Containment Atmospheric Dilution Operation Zero Containment Leakage VII-17 Feedwater Delivery Capability (Shaft Driven Pump) to Time After Turbine Trip for 1000 psig Reactor Pressure and 1.0 Inch HG ABS Exhaust Pressure VIII-1 Protective System Function VIII-2 Reactor Protection System Elementary Diagram VIII-3 Protective System Typical Sensor Arrangement VIII-4 Recirculation Flow and Turbine Control VIII-5 Neutron Monitoring Instrument Ranges VIII-6 Source Range Monitor (SRM) VIII-7 SRM Detector Location VIII-8 Intermediate Range Monitor (IRM) VIII-9 IRM Core Location VIII-10 LPRM Location Within Core Lattice VIII-11 LPRM and APRM Core Location VIII-12 Local Power Range Monitor (LPRM) and Average Power Range Monitors (APRM) VIII-13 APRM System Typical VIXI-14 Trip Logic for APRM Scram and Rod Block VIII-15 Traversing In-Core Probe VIII-16 Rod Pattern During Startup VIII-17 Radial Power Distribution for Control Rod Pattern Shown in Figure VXII-16 VIII-18 Distance from Worst Control Rod to Nearest Active IRM Monitor UFSAR Revision xli June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title VIII-19 Measured Response Time of Intermediate Range Safety Instrumentation VIII-20 Envelope of Maximum APRM Deviation by Flow Control Reduction in Power VIII-21 Envelope of Maximum APRM Deviation for APRM Tracking With On Units Control Rod Withdrawal VIII-22 Main Steam Line Radiation Monitor VIII-23 Reactor Building Ventilation Radiation Monitor VIII-24 Offgas System Radiation Monitor VIII-25 Emergency Condenser Vent Radiation Monitor VIII-26 Stack Effluent and Liquid Effluent Radiation Monitors VIII-27 Containment Spray Heat Exchanger Raw Water Effluent Radiation Monitor VIII-28 Containment Atmospheric Monitoring System VIII-29 Rod Worth Minimizer IX-1 A.C. Station Power Distribution IX-2 Control and Instrument Power IX-3 Trays Below Elevation 261 IX-4 Trays Below Elevation 277 IX-5 Trays Below Elevation 300 IX-6 Diesel Generator Loading Following Loss-of-Coolant Accident IX-7 Diesel Generator Loading for Orderly Shutdown X-1 Reactor Shutdown Cooling System X-2 Reactor Cleanup System X-3 Control Rod Drive Hydraulic System X-4 Reactor Building Closed Loop Cooling System X-5 Turbine Building Closed Loop Cooling System X-6 Service Water System X-7 Decay Heat Generation, Q vs. Days After Reactor Shutdown ', X-8 Spent Fuel Storage Pool Filtering and Cooling System X-9 Breathing, Instrument, and Service Air X-10 Reactor Refueling System Pictorial X-11 Cask Drop Protection System XI-1 Steam Flow and Reheater Ventilation System XI-2 Extraction Steam Flow XI-3 Main Condenser Air Removal and Off Gas System XI-4 Circulating Water System XI-5 Condensate Flow XI-6 Condensate Transfer System XI-7 Feedwater Flow System UFSAR Revision 14 xiii June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title XII-1 Radioactive Waste Disposal System XIII-1 NMPC Upper Management Nuclear 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-2 Figure Deleted XV-3 Plant Response to Loss of 100 F Feedwater Hea ting XV-4 Figure Deleted XV-5 Figure Deleted XV-6 Figure Deleted XV-7 Figure Deleted XV-8 Startup of Cold Recirculation Loop Partial Power XV-9 Recirculation Pump Trips (1 Pump) XV-10 Recirculation Pump Trips (5 Pumps) XV-11 Recirculation Pump Stall XV-12 Flow Controller Malfunction (Increased Flow) XV-13 Flow Controller Malfunction Decreasing Flow XV-14 Inadvertent Actuation of One Solenoid Relief Valve XV-15 Figure Deleted XV-16 Figure Deleted XV-17 Feedwater Controller Malfunction Zero Flow XV-18 Turbine Trip With Partial Bypass Intermediate Power XV-19 Turbine Trip With Partial Bypass XV-20 Inadvertent Actuation of One Bypass Valve XV-21 One Feedwater Pump Trip and Restart XV-22 Loss of Electrical Load XV-23 Loss of Auxiliary Power XV-24 Pressure Regulator Malfunction XV-25 Main Steam Line Break Coolant Loss XV-26 Figure Deleted XV-27 Figure Deleted XV-28 Figure Deleted XV-29 Figure Deleted XV-30 Figure Deleted XV-31 Figure Deleted XV-32 Figure Deleted XV-33 Figure Deleted XV-34 Figure Deleted XV-35 Figure Deleted XV-36 Figure Deleted XV-37 Figure Deleted XV-38 Figure Deleted UFSAR Revision 14 xliii June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title XV-39 Figure Deleted XV-40 Figure Deleted XV-41 Figure Deleted XV-42 Figure Deleted XV-43 Figure Deleted XV-44 Figure Deleted XV-45 Figure Deleted XV-46 Figure Deleted XV-47 Figure Deleted XV-48 Figure Deleted XV-49 Figure Deleted XV-50 Figure Deleted XV-51 Figure Deleted XV-52 Figure Deleted XV-53 Figure Deleted XV-54 Figure Deleted XV-55 Figure Deleted XV-56 Figure Deleted XV-56A Figure Deleted XV-56B Figure Deleted XV-56C Figure Deleted XV-56D Loss-of-Coolant Accident With Core Spray Cladding Temperature XV-56E Loss-of-Coolant Accident Drywell Pressure XV-56F Loss-of-Coolant Accident Suppression Chamber Pressure XV-56G Loss-of-Coolant Accident Containment Temperature With Core Spray XV-56H Loss-of-Coolant Accident Clad Perforation With Core Spray XV-57 Containment Design Basis Clad Temperature Response Without Core Spray XV-58 Containment Design Basis Metal-Water Reaction XV-59 Containment Design Basis Clad Perforation Without Core Spray XV-60 Containment Design Basis Containment Temperature Without Core Spray XV-60a DBR Analysis Suppression Pool and Wetwell Airspace Temperature Response Containment Spray Design Basis Assumption XV-60b DBR Analysis Suppression Pool and Wetwell Airspace Temperature Response EOP Operation Assumptions XV-61 Reactor Building Model XV-62 Exfiltration vs. Wind Speed Northerly Wind XV-63 Reactor Building Differential Pressure XV-64 Exfiltration vs. Wind
- Speed Southerly Wind XV-65 Reactor Building Isometric XV-66 Reactor Building Corner Sections UFSAR Revision 14 xliv June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title XV-67 Reactor Building Roof Sections XV-68 Reactor Building Panel to Concrete Sections XV-69 Reactor Building Expansion Joint Sections XV-70 Reactor Building Exfiltration Northerly Wind XV-71 Reactor Building Exfiltration Southerly Wind XV-72 Reactor Building Differential Pressure XVI-1 Seismic Analysis of Reactor Vessel Geometric and Lumped Mass Representation XVI-2 Reactor Support Dynamic Analysis Elevation vs. Moment XVI-3 Reactor Support Dynamic Analysis Elevation vs. Shear XVI-4 Reactor Support Dynamic Analysis Elevation vs. Deflection XVI-5 Reactor Support Dynamic Analysis Elevation vs. Acceleration XVI-6 Figure Deleted XVI-7 Figure Deleted XVI-8 Figure Deleted XVI-9 Reactor Vessel Support Structure Stress Summary XVI-10 Thermal Analysis XVI-11 Failure Probability Density Function XVI-12 Addition Strains Past 44 Required to Exceed Defined Safety Margin XVI-12a Shroud Welds XVI-12b Core Shroud Stabilizers XVI-13 Loss of Coolant Accident Containment Pressure No Core or Containment Sprays XVI-14 Figure Deleted XVI-15 Drywell to Concrete Air Gap XVI-16 Typical Penetrations XVI-17 Biological Shield Wall Construction Details XVI-18 Vent Pipe and Suppression Chamber XVI-19 Primary Containment Support and Anchorage XVI-20 Seal Details Drywell Shell Steel and Adjacent Concrete XVI-21 Drywell Sliding Acceleration, Shear, and Moment XVI-22 Shear Resistance Capability Inside Drywell XVI-23 Shear Resistance Capability Outside Drywell XVI-24 Drywell Support Skirt Junction Stresses XVI-25 Point Location for Containment Spray System Piping Heat Exchanger to Drywell XVI-26 Comparison of Static and Dynamic Stresses (PSI) Seismic Conditions Containment Spray System Heat Exchanger to Drywell XVI-27 Conduction in a Droplet XVI-28 Loss of Coolant Accident Containment Pressure UFSAR Revision 14 xlv June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title XVI-29 Loss of Coolant Accident Containment Pressure XVI-30 Nozzle Spray Test Pressure Drop of 80 psig XVI-31 Nozzle Spray Test Pressure Drop of 80 psig XVI-32 Nozzle Spray Test Pressure Drop of 30 psig XVI-33 Nozzle Spray Test Pressure Drop of 30 psig XVI-34 Seismic Analysis Reactor Building XVI-35 Dynamic Analysis Drywell XVI-36 Reactor Support Structure Seismic XVI-37 Seismic Analysis Waste Building XVI-38 Seismic Analysis Screenhouse XVI-39 Seismic Analysis Turbine Building (North of Row C) XVI-40 Seismic Analysis Turbine Building (South of Row C) XVI-41 Seismic Analysis Concrete Ventilation Stack XVI-42 Reactor Building Mathematical Model (North-South) XVI-43 Reactor Support Structure Seismic XVI-44 Reactor Support Structure Reactor Building XVI-45 Reactor Support Structure Reactor Building and Seismic XVI-46 Plan of Building XVI-47 Wall Section 1 XVI-48 Wall Section 1 Detail "A" XVI-49 Wall Section 1 Detail "B" XVI-50 Wall Section 1 Detail "C" XVI-51 Wall Section 1 Detail "D" XVI-52 Wall Section 1 Detail "E" XVI-53 Wall Section 2 XVI-54 Wall Section 3 XVI-55 Wall Section 3A Details XVI-56 Wall Section 4 XVI-57 Wall Section 4 Detail 1 XVI-58 Wall Section 4 Detail 2 XVI-59 Wall Section 5 XVI-60 Wall Section 6 XVI-61 Wall Section 7 XVII-1 Average Wind Roses for January '63-'64 XVII-2 Average Wind Roses for February '63-'64 XVII-3 Average Wind Roses for March '63-'64 XVII-4 Average Wind Roses for April '63-'64 XVII-5 Average Wind Roses for May '63-'64 XVII-6 Average Wind Roses for June '63-'64 XVII-7 Average Wind Roses for July '63-'64 XVII-8 Average Wind Roses for August '63-'64 XVII-9 Average Wind Roses for September '63-'64 XVII-10 Average Wind Roses for October '63-'64 XVII-11 Average Wind Roses for November '63-'64 UFSAR Revision 14 xlvi June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure Number Title XVII-12 Average Wind Roses for December '63-'64 XVII-13 Average Wind Roses for '63-'64 XVII-14 Average Diurnal Lapse Rate January '63-'64, February '63-'64 XVII-15 Average Diurnal Lapse Rate March '63-'64, April
'63-'64 XVII-16 Average Diurnal Lapse Rate May '63-'64, June i63-'64 XVII-17 Average Diurnal Lapse Rate July '63-'64, August
'63- 64 XVII-18 Average Diurnal Lapse Rate September '63-'64, October '63-'64 XVII-19 Average Diurnal Lapse Rate November '63-'64, December '62-'63 XVII-20 Lapse Rates by Wind Speed and Turbulence Classes for January '63-'64 XVII-21 Lapse Rates by Wind Speed and Turbulence Classes for February '63-'64 XVII-22 Lapse Rates by Wind Speed and Turbulence Classes for March '63- 64 XVII-23 Lapse Rates by Wind Speed and Turbulence Classes for April '63-'64 XVII-24 Lapse Rates by Wind Speed and Turbulence Classes for May '63-'64 XVII-25 Lapse Rates by Wind Speed and Turbulence Classes for June '63-'64 XVII-26 Lapse Rates by Wind Speed and Turbulence Classes for July '63-'64 XVII-27 Lapse Rates by Wind Speed and Turbulence Classes for August '63-'64 XVII-28 Lapse Rates by Wind Speed and Turbulence Classes for September '63-'64 XVII-29 Lapse Rates by Wind Speed and Turbulence Classes for October '63-'64 XVII-30 Lapse Rates by Wind Speed and Turbulence Classes for November '63-'64 XVII-31 Lapse Rates by Wind Speed and Turbulence Classes for December '63-'64 XVII-32 Sector Map XVII-33 Centerline Concentrations Turbulence Class I XVII-34 Centerline Concentrations Turbulence Class Class II III XVII-35 Centerline Concentrations Turbulence XVII-36 Centerline Concentrations Turbulence Class IV XVII-37 Centerline Concentrations Turbulence Class II Becoming Class IV at 2 km and Class II at 23 km XVII-38 Centerline Concentrations Turbulence Class IV Becoming Class II at 16 km UFSAR Revision 14 xlvii June 1996
Nine Mile Point Unit 1 FSAR LIST OF FIGURES (Cont'd.) Figure gum~be Title XVII-39 Centerline Concentrations Turbulence Class IV Becoming Class II at 2 km XVII-40 Radial Concentrations Turbulence Class I XVII-41 Radial Concentrations Turbulence Class II XVII-42 Radial Concentrations Turbulence Class III XVII-43 Radial Concentrations Turbulence Class IV XVII-44 Radial Concentrations Turbulence Class II Becoming Class IV at 2 km and Class II at 23 km XVII-45 Radial Concentrations Turbulence Class IV Becoming Class II at 16 km XVII-46 Radial Concentrations Turbulence Class IV Becoming Class II at 2 km XVII-47 Centerline Gamma Dose Rates Turbulence Class I XVII-48 Centerline Gamma Dose Rates Turbulence Class II XVII-49 Centerline Gamma Dose Rates Turbulence Class ZII XVII-50 Centerline Gamma Dose Rates Turbulence Class IV XVII-51 Centerline Gamma Dose Rates Turbulence Class II Becoming Class IV at 2 km and Class II at 23 km XVII-52 Centerline Gamma Dose Rates Turbulence Class IV Becoming Class II at 16 km XVII-53 Centerline Gamma Dose Rates Turbulence Class IV Becoming Class II at, 2 km XVI1-54 Assumed Concentration and Dose Rate Distributions Close to the Elevated Source XVII-55 Gamma Dose Rate as a Function of ay at 1 km From the Source XVII-56 Southeastern Lake Ontario XVII-57 Dilution of Rising Plume XVII-58 Estimated Lake Currents at Cooling Water Discharge XVII-59 Temperature Profiles in an Eastward Current at the Oswego City Water Intake XVII-60 Subsurface Section Plot Plan XVII-61 Log of Boring (Boring CB-1) XVII-62 Log of Boring (Boring CB-2) XVII-63 Log of Boring (Boring CB-3) XVII-64 Log of Boring (Boring CB-4) XVII-65 Attenuation Curves UFSAR Revision 14 xlviii June 1996
Nine Mile Point Unit 1 FSAR SECTION I INTRODUCTION AND
SUMMARY
This report is submitted in accordance with 10 CFR Part 50.71(e)for entitled "Periodic Updating of Final Safety Analysis Reports" Niagara Mohawk Power Corporation's (NMPC) Nine Mile Point Nuclear Station Unit 1 (Unit 1). The Station is located on the southeast shore of Lake Ontario, in Oswego County, New York, 7 mi northeast of the city of Oswego. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR A. PRINCIPAL DESIGN CRITERIA The following paragraphs describing the principal design criteria are oriented toward the twenty-seven criteria issued by the United States Atomic Energy Commission (USAEC).+ 1.0 General The Station is intended as a high load factor generating facility to be operated as an integral part of the NMPC 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. 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 Buildings 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. 3.0 Reactor 1~ A direct-cycle boiling water system reactor (BWR), 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 MWt. 2 ~ The reactor is fueled with slightly enriched uranium dioxide contained in Zircaloy clad fuel rods described UFSAR Revision 14 I-2 June 1996
Nine Mile Point Unit 1 FSAR in Section IV. Selected fuel rods also incorporate small amounts of gadolinium as burnable poison. I, To avoid fuel'damage, the minimum critical power ratio k 3 (MCPR) is maintained greater than the safety limit CPR.
~
4 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.
- 5. 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.
- 6. 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. cooling capability is also available from the Additional high-pressure coolant injection (HPCI) system and the fire protection system. Redundant and independent core spray systems are provided to cool the core in the event of a loss-of-coolant accident (LOCA). 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 LOCA will be limited to less than 1 percent of the Zircaloy clad. 7 ~ 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. UFSAR Revision 14 I-3 June 1996
Nine Mile Point Unit 1 FSAR 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 (CRD) hydraulic system is described in Section X-C. The force available to scram a control rod is approximately 3000 lb at the beginning of a scram stroke. This is well in excess of the 440-lb force required in the event of fuel channel pinching of the control rod blade during a LOCA, as discussed in Section XV. Even with scram accumulator failure a force of at least 1100 lb from reactor pressure acting alone is available with reactor pressures in excess of 800 psig. 8 ~ 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~ <0.97, at any time in the core life, independent of the control rod system capabilities.
- 9. A flow restrictor in the main steam line (MSL) limits coolant loss from the reactor vessel in the event of a MSL break (Section VII-F).
4.0 Reactor Vessel 1 ~ 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. A bypass system having a capacity of approximately 40 percent of turbine steam flow for the throttle valves wide open (VWO) condition partially mitigates the effects of sudden load rejection. This and other transients and maneuvers which have been analyzed are detailed in Section XV. 2 ~ Separate systems to prevent serious reactor coolant system (RCS) 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 RCS pressure relief devices is included in Section V-C. 3 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. UFSAR Revision 14 I-4 June 1996
Nine Mile Point Unit 1 FSAR 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' rod worth minimizer (RWM) (Section VIII-C). Power excursion analyses for control rod dropout accidents are included in Section XV. 4 The reactor vessel will not be substantially pressurized until the vessel wall temperature is in
~
excess of nil ductility transition temperature (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 (RG) 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 1~ The primary containment, including the drywell, pressure suppression chamber, and 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 (DER) or equivalent failure of any coolant pipe within the drywell. The primary containment is designed to accommodate the pressures following a LOCA 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 -20'F and is not expected to increase during the lifetime of the Station. These structures are described in Sections VI-A, B and C. Additional details, particularly those related to design and fabrication, are included in Section XVI. 2 ~ Provisions are made for the removal of heat from within the primary containment, for reasonable protection of the containment from fluid jets'r missiles and such other measures as may be necessary to maintain the integrity of the containment system as long as necessary following a LOCA. 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 UFSAR Revision 14 I-5 June 1996
Nine Mile Point Unit 1 FSAR 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. 3 ~ Provision is made for periodic integrated leakage rate tests (ILRT) 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 described in Section VI-F. 4 ~ The containment system and all other necessary engineered safeguards are designed and maintained such that, offsite doses resulting from postulated accidents are below the values stated in 10CFR100. The analysis results are detailed in Section XV.
- 5. Double isolation valves are provided on all lines directly entering the primary containment freespace or penetrating the primary containment and connected to the RCS. Periodic testing of these valves will assure their capability to isolate at all times. The isolation valve system is discussed in detail in Section VI-D.
- 6. 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 1 ~ 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. 2 ~ 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 on Figures VIII-6 UFSAR Revision 14 I-6 June 1996
Nine Mile Point Unit 1 FSAR and VIII-8, respectively, for the source range monitor (SRM) and intermediate range monitor (IRM) neutron instrumentation. In the power range, average power range monitor (APRM) instrumentation provides both control rod and recirculation flow control blocks, as shown on Figure VIII-14. Reactivity excursions involving the control rods are either prevented or their consequences substantially mitigated by a control RWM (Section VIII-C.4.0) which supplements procedural controls in avoiding patterns of high rod worths, a low power range monitor (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 CRD. 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. 3 ~ A reliable, dual-logic channel reactor protection system (RPS), 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 inotherone logic channel coupled with high pressure in the logic channel will result in a scram. The RPS circuitry fails in a direction to cause a reactor scram in the event of loss of power or loss of air supply to the scram solenoid valves. Periodic testing and calibration of individual subchannels is performed to assure system reliability. The ability of the RPS to safely terminate a variety of Station malfunctions is demonstrated in Section XV. UFSAR Revision 14 I-7 June 1996
Nine Mile Point Unit 1 FSAR 4 Redundant sensors and circuitry are provided for the actuation of all equipment required to function under
~
postaccident 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 and power are provided to assure a capability for prompt shutdown continued maintenance of the Station in a safe condition under all credible circumstances. These features are discussed in Section IX. 8.0 Radioactive Waste Disposal 1 ~ Gaseous, liquid and solid waste disposal facilities are designed so that discharge of effluents is in accordance with 10CFR20 and 10CFR50 Appendix I. The facility descriptions are given in Section XII-A while the development of appropriate limits is covered in Section II. 2 ~ Gaseous discharge from the Station is appropriately monitored, as discussed in Section VIII, and automatic isolation features are incorporated to maintain releases below the limits of 10CFR20 and 10CFR50 Appendix I. 9.0 Shielding and Access Control Radiation shielding and access control patterns are such that doses will be less than those specified in 10CFR20. These features are described in Section XII-B. 10.0 Fuel Handling and Storage Appropriate fuel handling and storage facilities which preclude accidental criticality and provide adequate cooling for spent fuel are described in Section X. UFSAR Revision 14 I-8 June 1996
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 Corporation Ownership Net Electrical Output 615 MW (Maximum) 2.0 Reactor Reference Rated Thermal 1850 MW Output Dome Pressure 1030 psig Turbine Inlet Pressure 956 psig Total Core Coolant 67.5 x 10'lb/hr Flow Rate Steam Flow Rate 7.32 x 10'lb/hr 3.0 Core Circumscribed Core 167.16 in Diameter Active Core 171.125 in Height + Assembly 4.0 Fuel Assembly Number of Fuel Assemblies 532 Fuel Rod Array SRLR+
Fuel Rod Pitch Reference 3 Cladding Material Reference 3 Fuel Material UO, and UO,-Gd,03 Active Fuel Length Reference 3 Cladding Outside Diameter Reference 3 Cladding Thickness Reference 3 Fuel Channel Material Reference 3 5.0 Control System Number of Movable Control 129 Rods Shape of Movable Control Cruciform Rods Pitch of Movable Control 12.0 in Rods Control Material in B4C 704 Theoretical Movable Control Rods Density; Hafnium Type of Control Drives Bottom Entry, Hydraulic Actuated UFSAR Revision 14 I-9 June 1996
Nine Mile Point Unit 1 FSAR Control of Reactor Output Movement of Control Rods and Variation of Coolant Flow Rate 6.0 Core Design and Operating Conditions Maximum Linear Heat Core Operating Limits Report Generation Rate Heat Transfer Surface Area Average Heat Flux Rated Power Initial Critical Power Core Operating Limits Report Ratio for Most Limiting Transients Core Average Void Fraction Coolant within Assemblies Core Average Exit Quality Coolant within Assemblies 7.0 Design Power Peaking Factor Total Peaking Factor GE8x8EB 2.90 GE11 2.94** 2 '2*** 8.0 Nuclear Design Data Average Initial Volume Reference 3 Metric Enrichment Beginning of Cycle 12 Core Effective Multiplication and Control System Worth-No Voids, 20C+ Uncontrolled 1.095 Fully Controlled 0 '49 Strongest Control 0 '82 Rod Out
- These parameters are recalculated for each reload because of their dependency on core composition and exposure. These calculated values are intermediate quantities that do not represent design requirements or operating limits and thus are not separately reported in the SRLR+.
Maximum total peaking factor for the portion of the bundle containing part length rods.
- Maximum total peaking factor for the region above the part length rods.
UFSAR Revision 14, I-10 June 1996
Nine Mile Point Unit 1 FSAR Standby Liquid Control System Capability: Shutdown Margin (dR) 20C Xenon Free SRLR~~ SRLR~> 9.0 Reactor Vessel Inside Diameter Internal Height 17 63 ft ft - 10 inin 9 Design Pressure 1250 psig at 575'F 10.0 Coolant Recirculation Loops Location of Recirculation Containment Drywell Loops Number of Recirculation Loops and Pumps Pipe Size 28 in 11.0 Primary 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 12.0 Secondary Containment Type Reinforced concrete and steel superstructure with metal siding Internal Design Pressure 40 lb/ft Design Leakage Rate 1004 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 Design Seismic Ground 0.11g Acceleration Sustained Wind Loading 125 mph, 300 not to ft above ground level hourly Control Room Shielding Dose exceed equivalent (based on 40-hr week) of maximum permissible quarterly dose specified in 10CFR20 UFSAR Revision 14 I-11 June 1996
Nine Mile Point Unit 1 FSAR 14.0 Station Electrical System 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 batteries One nonsafety 125-V dc battery system 15.0 Reactor Instrumentation System Location of Neutron In-core Monitor Sensors Ranges of Nuclear Instrumentation: Four Startup Range Source to 0.014 rated power and to Monitors 104 with chamber retraction Eight Intermediate Range 0.00034 to 104 rated power Monitors 120 Power Range Monitors 14 to 1254 rated power 16.0 Reactor Protection System Number of Channels in Reactor Protection System Number of Channels Required to Scram or Effect Other Protective Functions Number of Sensors per Monitored Variable in each Channel (Minimum for scram function) UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR C. IDENTIFICATION OF CONTRACTORS The General Electric Company (GE) was engaged to design, fabricate and deliver the nuclear steam supply system (NSSS), turbine generator, and other major elements and systems. GE also furnished the complete cor'e design and nuclear fuel supply for the initial core and is currently furnishing replacement cores. NMPC, 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 (SWEC) was engaged by NMPC to manage field construction. Currently, NMPC utilizes various contractors to assist in continuous Station modifications. UFSAR Revision 14 I-13 June 1996
Nine Mile Point Unit 1 FSAR D. 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 Unit 1 can be operated without endangering the health and safety of the public. UFSAR Revision 14 I-14 June 1996
Nine Mile Point Unit 1 FSAR E. REFERENCES
- 1. USAEC Press Release H-252, "General Design Criteria for Nuclear Power Plant Construction Permits," November 22, 1965.
2 ~ GENE 24A5157, Revision 0, "Supplemental Reload Licensing Report for NMPl, Reload 13, Cycle 12," January 1995.
- 3. GE Fuel Bundle Designs, General Electric Company Proprietary, NEDE-31152P, February 1993.
UFSAR Revision 14 I-15 June 1996
Nine Mile Point Unit 1 FSAR SECTION II STATION SITE AND ENVIRONMENT A. SITE DESCRIPTION 1.0 General The Nine Mile Point Nuclear Station Unit 1 (Unit 1), owned by Niagara Mohawk Power Corporation (NMPC), is located on the western portion of the Nine Mile Point promontory. Approximately 300 ft due east is Nine Mile Point Nuclear Station Unit 2 (Unit 2). The eastern portion of the promontory is comprised of the James A. FitzPatrick Nuclear Power Plant, owned by the New York Power Authority (NYPA). The site is on Lake Ontario in Oswego County, approximately 5 mi north-northeast of the nearest boundary of the city of Oswego. Figure II-1 shows the Station location on an outline map of the state of New York. It is 230 mi northwest of New York City, 143.5 mi east-northeast of Buffalo, and 36 mi north-northwest of Syracuse. Figure II-2 is a detailed map of the area within about 50 mi of the Station. 2.0 Physical Features Figure II-3 is a detailed site map showing Station location; an associated plot plan is presented as Figure III-1 of the following section. Station buildings are situated in the western quadrant of a 200-acre cleared area centrally located along the lakeshore. Site property consists of partially-wooded land formerly used almost exclusively for residential and recreational purposes. For many miles west, east, and south of the site the country is characterized by rolling terrain rising gently up from the lake. Grade elevation at the site is 10 ft above the record high lake level, while underlying rock structure is among the most structurally stable in the United States (U.S.) from the standpoint of tilting and folding. There is no record of wave activity, such as seiche or tsunami, of such a magnitude as to make inundation of the site likely. A shore protection dike composed of rock and the lake. fill from the excavation separates the buildings All elevations in this report refer to the United States Land Survey (USLS) 1935 data.
- 1. To convert elevations to 1955 International Great Lakes Data (IGLD 1955), subtract 0.375m (1.23 ft).
UFSAR Revision 14 II-1 June 1996
Nine Mile Point Unit 1 FSAR
- 2. To convert elevations to 1985 International Great Lakes Data (IGLD 1985), subtract 0.217m (0.71 ft).
Exclusion distances for the site are approximately 1 mi to the east, a mile to the southwest, and over a mile to the southern site boundary. 3.0 Property Use and Development There are no residences, agricultural or industrial developments (other than the James A. FitzPatrick Nuclear Power Plant) on the site; all former summer homes and farm buildings have been removed. Site boundaries and the former country road which traverses the site are posted as private property. The area immediately around the Station buildings is fenced, with building access controlled by Station security personnel. A visitors'nergy Information Center, manned by NMPC and NYPA personnel, and the Niagara Mohawk Nuclear Learning Center are located about 1,000 ft west of the Station, per Figure II-3. These installations may be reached by the public over private drives maintained by the company. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR B. DESCRIPTION OF AREA ADJACENT TO THE SITE 1.0 General The Station is located on the Lake Ontario coast in the town of Scriba in the north-central portion of Oswego County, approximately 5 mi north-northeast of the nearest boundary of the city of Oswego. 1.1 Population Population growth in the vicinity of the Station has been very slow, with the city of Oswego showing a decrease in population. The 1960 census enumerated 22,155 residents compared to approximately 19,793 people in 1980. However, county population increased from 86,118 in 1960 to 113,901 in 1980. The total 1980 population within 12 mi of the Station is estimated to be 46,349 (see Figure II-4). This area contains all or portions of one city and ten towns. Population and population density for the ten towns and one city within this area are shown in Table II-1. Counties and towns within this area are shown on Figure II-5. Transient population within 12 mi of the Station is limited due to the rural, undeveloped character of the area. There are, however, a number of school, industrial, and recreational facilities in the area that create small daily and seasonal changes in area populations. The population within a 50-mi area surrounding the Station was approximately 914,193 in 1980 (see Figure II-6). The city of Syracuse is the largest population center within this area, with a population of 170,105 in 1980. Table II-2 lists cities within this 50-mi radius with populations over 10,000. The 50-mi radius contains portions of three Canadian Census Divisions located in the province of Ontario: Prince Edward, Frontenac, and Addington/Lennox. The 1976 population counts totaled 22,559, 108,052, and 32,633, respectively. 2.0 Agriculture, Industrial and Recreational Use 2.1 Agricultural Use The area within a 50-mi radius of the site encompasses all or portions of ten New York counties: Cayuga, Jefferson, Lewis, Madison, Oneida, Onondaga, Ontario, Oswego, Seneca, and Wayne. Approximately 37 percent of the land within this ten-county region is used for agricultural production. Tables II-3 and II-4 present agricultural statistics for this ten-county region. 2.2 Industrial Use Several industrial establishments are located in Oswego County, with the Alcan Aluminum Corporation and the Independence UFSAR Revision 14 II-3 June 1996
Nine Mile Point Unit 1 FSAR Generation Plant operated by Sithe Energies USA being located nearest to the Station. The lakeshore east of Oswego is the most industrially developed area near the site. The cities of Fulton and Mexico are the only other industrial sites within 15 mi of the site. Two natural gas pipelines lie within 8 km of the Plant and the other plant; one pipeline supplies the Independence located supplies Indeck Energy. Both pipelines are on the north-south and east-west transmission line corridors. The major their industrial establishments in Oswego County, II-5 locations, and their principal products are listed in Tables and II-6. The nearest public water supply intake in Lake Ontario is located approximately 8 mi southwest of the Station location. This intake supplies the city of Oswego and Onondaga County. Data on these and other vicinity public water supplies are listed in Table II-7. Figure II-2 shows the locations of the communities listed. 2.3 Recreational Use Seventeen state parks and one national wildlife refuge are located within a 50-mi radius of the Station. Table II-8 identifies the state parks and their facilities, capacities, and visitor counts. The Montezuma National Wildlife Refuge is 44 located north of Cayuga Lake in Seneca County, approximately mi southwest of the Station. UFSAR Revision 14 II-4 June 1996
Nine Mile Point Unit 1 FSAR C.~ METEOROLOGY An original 2-yr study was performed to determine the site
~ ~
meteorological characteristics.
~
This study is presented in Section XVII-A. The meteorological monitoring system measures parameters to provide data that are representative of atmospheric conditions that exist at all gaseous effluent release points. Meteorological data is compiled for quarterly periods in accordance with the Technical Specifications. This data is used to provide information which may be used to develop atmospheric diffusion parameters to estimate potential radiation doses to the public resulting from actual routine or accidental releases of radioactive materials to the atmosphere. UFSAR Revision 14 II-5 June 1996
Nine Mile Point Unit 1 FSAR D. LIMNOLOGY A comprehensive research program, designed to monitor various parameters of the aquatic environment in the vicinity of Nine Mile Point, was begun in 1963. This detailed lake program was continued through 1978. Currently, an aquatic ecology study program (closely coordinated with James A. FitzPatrick Nuclear Power Plant) is conducted in the vicinity of Nine Mile Point on Lake Ontario to monitor the effects of plant operation with respect to selected ecological parameters, and to perform impingement studies on the traveling screens in the intake screenwell. This program is carried out and results reported in accordance with the station State Pollutant Discharge Elimination System (SPDES) Discharge Permit. UFSAR Revision 14 II-6 June 1996
Nine Mile Point Unit 1 FSAR E.~ EARTH SCIENCES A preconstruction evaluation of the geology, hydrology, and seismology of the Nine Mile Point promontory is presented in
~
Section XVII-C. Subsequent inspection of rock exposed during excavations for the reactor and cooling water tunnels allowed for a more detailed study of subsurface conditions. No faults were encountered and no unusual conditions were observed. The structures rest on a firm, almost impervious rock foundation. Station seismic design criteria were based upon a conservative evaluation of the maximum earthquake ground motion which might conceivably occur at the site. This condition was calculated by assuming that the worst shock ever observed within an effective range of the site might be located at, the closest position to the site at which an earthquake of any intensity occurred. The "maximum possible" shock assumed for Station structure acceleration calculations is of magnitude 7 at a 50-mi epicentral distance. Dames and Moore estimates that this shock will probably never occur unless unusual regional geologic changes take place. UFSAR Revision 14 II-7 June 1996
Nine Mile Point Unit 1 FSAR F. ENVIRONMENTAL RADIOLOGY Controlled releases of radioactive materials in liquid and gaseous effluents to the environment is part of normal Station operation. A Radiological Environmental Monitoring Program ensures that the release rates for all effluents are within the limits specified in 10CFR20 and the release of radioactive material above background to unrestricted areas conforms with Appendix I to 10CFR50. Comprehensive studies were originally conducted to establish the effluent emission rates which would produce the above limiting conditions in the uncontrolled environment. Currently, a Radiological Environmental Monitoring Program~, inclusive of Unit 1, is in operation. This program details the design objectives for control of liquid and gaseous wastes, including specifications for liquid and gaseous waste effluents, . and specifications for liquid and gaseous waste sampling and monitoring. An annual Environmental Operating Report and Semiannual Radioactive Effluent Release Reports are prepared and submitted in accordance with the reporting requirements in the Technical Specifications. UFSAR Revision 14 II-8 June 1996
Nine Mile Point Unit 1 FSAR G. REFERENCES
- 1. Nine Mile Point Nuclear Station "Technical Specifications
~ ~ ~
and Bases". UFSAR Revision 14 II-9 June 1996
Nine Mile Point Unit 1 FSAR TABLE II-1 1980 POPULATION AND POPULATION DENSITY FOR TOWNS AND CITIES WITHIN 12 MILES OF NINE MILE POINT UNIT 1 Population Density 1980 Po ulation Peo le Per S are Mile City of Oswego 19,793 2665.2 Oswego (town) 7,865 302.7 Granby 6,341 142.9 Richland 5,594 105.9 Scriba 5,455 137.0 Volney 5i358 119.1 Mexico 4,790 108.3 Hannibal 4,027 99 ' Palermo 3,253 81.8 New Haven 2,421 82.1 Minetto 1,905 325.0 UFSAR Revision 14 1 of 1 June 1996
Nine Mile Point Unit 1 FSAR TABLE II-2 CITIES WITHIN A 50-MILE RADIUS OF THE STATION WITH POPULATIONS OVER 10,000 Population
~Cit ~Count 1980 Census Newark Village Wayne 10/017 Clay Onondaga 52,838 Cicero Onondaga 23,689 Manlius Onondaga 28,489 Dewitt Onondaga 26,868 Syracuse Onondaga 170,105 Geddes Onondaga 18,528 Camillus Onondaga 24,333 Onondaga Onondaga 17,824 Van Buren Onondaga 12,585 Salina Onondaga 37,400 Fulton Oswego 13/312 Oswego Oswego 19,793 Oneida Madison 10,810 Rome Oneida 43,826 Watertown Jefferson 27,861 UFSAR Revision 14 1 of 1 June 1996
Nine Mile Point Unit 1 FSAR TABLE II-3 REGIONAL AGRICULTURAL USE Agricultural Corn Use (All Purposes) Wheat Fruit Totals County (square miles) (acres) (acres) (acres) (acres) Cayuga 560 84,002 11,999 395 96,396 Jefferson 847 42,501 499 43,000 Lewis 373 14,201 14,201 Madison 407 " 28,001 400 173 28,574 Oneida 612 35,601 1,401 222 37,224 Onondaga 336 45,002 4,900 1,097 50,999 Ontario 511 59,101 21,500 2,330 82,931 Oswego 267 13,200 11, 001 845 25,046 Seneca 299 31, 502 16,501 954 48,957 Wayne 40,499 5,001 25,125 70,625 Totals 4,630 393,610 73,202 31,141 497,953 SOURCE: NMP2 Environmental Report, Tables 2.2-9 and 2.2-10 UPSAR Revision 14 1 of 1 8une 1996
Nine Mile Point Unit 1 FSAR TABLE II-4 REGIONAL AGRICULTURAL STATISTICS - CATTLE AND MILK PRODUCTION Average Milk All Cattle and Calves Beef Cows Milk Cows Production/Cow (lb) Cayuga County 51,000 2,200 25,000 12,200 Jefferson County 84,000 2,600 44,000 11,100 Lewis County 59,000 600 32,500 12,300 Madison County 60,000 1,600 35,500 11,800 Oneida County 65,000 2,500 33,500 11,300 Onondaga County 32,500 2,500 17,000 13,200 Ontario County 33,000 1,600 11,500 11,900 Oswego County 25,500 2,300 11,500 11,400 Seneca County 11,500 1,000 4,300 11,200 Wayne County 19,000 1,800 8,500 10,400 Region 440,500 18,700 223,300 11,680 State 1,780,000 85,000 912,000 11,488 SOURCES: New York Crop Reporting Service, Cattle Inventory by County - 1980; Albany, NY, 1980
- 2. New York Crop Reporting Service, Milk Production 1978, Albany, NY. 1979
- 3. New York Crop Reporting Service, New York Agricultural Statistics - 1978, Albany, NY, 1979 UFSAR Revision 14 lof1 June 1996
Nine Mile Point Unit 1 FSAR TABLE II-5 INDUSTRIAL FIRMS WITHIN 8 KM (5 MI) OF UNIT 1 Distance/ Direction from Site Firm km Products Em lo ent Alcan Aluminum 4.5/SW Aluminum 1,000 Corporation sheet and plate James A. FitzPatrick (1/E Electrical 500 Nuclear Power Plant generation Nine Mile Point Adjacent Electrical 1, 100 Unit 2 to Unit 1 generation Sithe Energies USA 3.5/SW Electrical 75 Independence generation Generation Plant NOTE For complete listing of major industries in Oswego County, reference Oswego County Industrial Directory. UFSAR Revision 14 1 of 1 June 1996
Nine Mile Point Unit 1 FSAR TABLE II-6 PUBLIC UTILITIES IN OSWEGO COUNTY Location Service Niagara Mohawk Power Many sites Gas and Electric Corporation New York Telephone Many sites Communications Company Penn Central Railroad 'Shipping Oswego County Telephone Oswego Communications Company Alltel New York, Inc. Fulton Communications New York Power Authority Many sites Gas and Electric UFSAR Revision 14 1 of 1 June 1996
Nine Mile Point Unit 1 FSAR TABLE ZZ-7 PUBLIC WATER SUPPLY DATA FOR LOCATIONS WITHIN AN APPROXIMATE 30-MILE RADIUS Distance Average from Site Direction Output (miles) from Site Town (mgd) Source of Water 0-10 SW Onondaga (County) 36 Lake Ontario (intake at Oswego) SW Oswego 9 Lake Ontario ESE Mexico 0.5 Three wells> two 40-ft deep, one 38-ft deep 10-20 ENE Pulaski 0.3 Springs SSE Fulton 2 Twelve wells, 30- to 70-ft deep; two wells, 21-ft deep NE Sandy Creek 0.2 Two wells, 21-ft deep 20-30 SE Central Square 0.08 One well, 24-ft deep ENE Orwell Not available Spring SSE Phoenix 0.35 Two wells; one 25-ft deep, one 45-ft deep S Baldwinsville 1 Four wells; one 93-ft deep, three shallow wells SW Fairhaven 0.15 Spring; one well, 46-ft deep SSW Cato 0.033 Three wells; two 55-ft deep, one 70-ft deep SW Wolcott 0.220 Lake Ontario NE Adams 0.3 Springs SW Red Creek 0.03 Wells and springs SOURCE: Nine Mile Point Unit 2 PSAR UFSAR Revision 14 1 of 1 June 1996
Nine Mile Point Unit 1 FSAR TABLE II-8 RECREATIONAL AREAS IN THE REGION Distance and Total Visitor Direction Capacity Count from Unit (No. of (April 1979-Park (miles) County Acreage Activities/Facilities People) March 1980) Selkirk Shores 9.8 NE Oswego 980 Camping, picnicking, hiking, swimming 3, 646 305,000 Battle Island 10.5 S Oswego 235 Golfing, fishing, hiking 303 40,000 Frenchman Island 26.7 SE Oswego 26 Fishing, hiking, picnicking, boating 100 Fair Haven 18.3 SW Cayuga 845 Camping, picnicking, boating, fishing 6,247 352,000 Beach Southwick 19.1 NE Jefferson 472 Camping, picnicking, boating, 4,401 70,000 Beach fishing, swimming, hiking Westcott Beach 29.3 NE Jefferson 319 Camping, picnicking, boating, 4,494 72,000 fishing, swimming, hiking Long Point 36.0 NE Jefferson 23 Camping, picnicking, boating, 754 9,000 fishing, swimming Cedar Point 47.8 NE Jefferson Camping, picnicking, boating, 1,853 60,000 fishing, swimming Burnham Point 45.4 NE Jefferson 12 Camping, picnicking, boating, 553 15,000 fishing, swimming Whetstone Gulf 48.0 ENE Lewis 2,000 Camping, picnicking, swimming, hiking 1,981 28,000 Chittenango 47.2 ENE Madison 183 Camping, picnicking, hiking 699 115,000 Falls Verona Beach 41.9 SE Madison 1,735 Picnicking, swimming 4,374 305,000 Lock 23 21.6 SSE Onondaga Picnicking, boating 119 Brewerton Green Lakes 38.7 SSE Onondaga 1,101 Camping, picnicking, hiking, boating, 3,361 1, 015, 000 fishing, swimming Clark 39.1 SSE Onondaga 290 Picnicking, hiking, playground 1,255 356,000 Reservation UFSAR Revision 14 lof2 June 1996
Nine Mile Point Unit 1 FSAR TABLE IZ-8 (Cont'd.) Distance and Total Visitor Direction Capacity Count from Unit (No. of (April 1979-Park (miles) County Acreage Activities/Facilities People) March 1980) Cayuga Lake 45.7 SSW Seneca 135 Camping, picnicking, swimming, 3, 270 129,000 boating, playground Chimney Bluffs 30.8 WSW Wayne 597 Camping, picnicking, swimming, 1,036 30,000 boating, playground NOTE: All facxlxt es are seasonal (summer) Not available UFSAR Revision 14 2 of 2 June 1996
Nine Mile Point Unit 1 FSAR SECTION III BUILDINGS AND STRUCTURES The structural design of buildings and components is based on the maximum credible earthquake motion outlined in Volume II of the Preliminary Hazards Summary Report (PHSR). Specifically, this maximum motion consists of a magnitude 7 (Intensity IX) shock at an epicentral distance of 50 mi from the site. The maximum ground motion acceleration is 11 percent of gravity and the maximum response acceleration is 45 percent of gravity for oscillations in the period range of 0.2 to 0.3 sec. All critical structures for the Station were subjected to a dynamic response analysis for the determination of maximum stresses in the structure. Class I structures and components whose failure could cause significant release of radioactivity, or which are vital to safe shutdown and isolation of the reactor, were designed so that the probability of failure would approach zero when subjected to the maximum credible earthquake motion. (Acceleration response spectrum, Plate C-22, Section III, First Supplement to the PHSR.) Functional load stresses resulting from normal operation when combined with stresses due to earthquake accelerations are within the established working* stresses for the material involved in the structure or component. Primary load stresses, when combined with stresses due to temperature and pressure, together with stresses due to earthquake accelerations, are within applicable code or working* values. Class II structures and components were designed for stresses within the applicable codes relating to these structures and components when subjected to functional or operating loads. Stresses resulting from the combination of operating loads and earthquake loads or wind loads have been limited to stresses 33 1/3 percent above working* stresses in accordance with applicable codes. Class III structures and components are those of a service nature not essential for safe reactor shutdown and isolation, and failure of which would not result in significant release of radioactive materials. These structures were designed on the basis of applicable building codes with seismic and wind requirements. All major components in the Station were classified as above and analyzed to the appropriate degree. Vital fluid containers were analyzed and designed for hydrodynamic pressures resulting from earthquake motion. As a result of deflection determinations,
- Also see Section XVI, Subsection G.
UFSAR Revision 14 III-1 June 1996
Nine Mile Point Unit 1 FSAR provisions were made for relative motion between adjacent components and structures where damage might result from differential movement and impact stresses. A list of the is structures and components reviewed for seismic design contained on pages III-1, III-2 and III-3 of the First Supplement to the PHSR. Stresses in the various structural members were investigated after the earthquake analysis was completed to verify that stresses are in compliance with those specified in the conventional codes such as those of the American Institute of Steel Construction, American Concrete Institute, and other applicable codes such as the New York State Building Code. All major structures are founded on very substantial Oswego sandstone which exists on the site at an average of 11 ft below grade. This eliminates the potential problems of soil consolidation and differential settlement. Figure III-1 is a plot plan showing the relationship of structures. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR A. TURBINE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings Exterior loadings for wind, snow and ice used in the design of the turbine building meet all applicable codes as a minimum. The roof and its supporting structure are designed to withstand a loading of 40 psf of snow or ice. The walls and building structure are designed to withstand an external loading of 40 psf of surface area, which is approximately equivalent to a wind velocity of 125 mph at the 30-ft level. 1.2 Pressure Relief Design To prevent failure of the superstructure due to a steam line break, a wall area of 1800 ft has been attached with bolts that will fail due to an internal pressure of approximately 45 psf, thus relieving internal pressure. Wall or building structure failure would occur at an internal pressure in excess of 80 psf. 1.3 Seismic Design and Internal Loadings The turbine building is designed as a Class II structure. Components are either Class II or Class I, as outlined on pages - III-1, III-2 and III-3 of the First Supplement to the PHSR. An analysis of the turbine building resulted in the use of the following earthquake design coefficients for the major components. Com onent Percent Gravit Comment Feedwater heaters 16.0 20.5 (calculation Based on and drain cooler used: 20.0 horizontal specific support structures 10.0 vertical) dynamic analysis Turbine generator 23.4 N-S horizontal Based on foundation 26.7 E-W horizontal specific dynamic analysis Condenser support 11.0 horizontal Based on structure 5.5 vertical specific dynamic analysis For the following components, percent gravity was 20.0 horizontal and 10.0 vertical, based on the Uniform Building Code. UFSAR Revision 14 III-3 June 1996
Nine Mile Point Unit 1 FSAR Steel structure supporting emergency Class I condenser makeup water storage tanks and demineralized water storage tank, and condensate demineralizer (CND) Motor generator (MG) sets for reactor Class II recirculating pump motors 150/35-ton overhead traveling crane Class II Structural anchors supporting main Class I steam, offgas, etc., piping Anchor bolts and associated bases and Classes I frame for support of all tanks, & II filters and pumps as well as electrical equipment. (Power boards, control consoles, etc.) Supports for moisture separators and Class II reheaters Stresses resulting from the functional or operating loads are within applicable codes relating to these structures and components. Stresses resulting from the combination of operating loads and earthquake or wind loads have been limited in accordance with applicable codes to a 33 1/3-percent increase in allowable stresses*. The adjoining walls of the turbine and reactor building superstructures are structurally separated to provide for dissimilar deformations due to earthquake motion. 1.4 Heating and Ventilation Heating and ventilation is provided for equipment protection, personnel comfort and for controlling possible radioactivity release to the atmosphere. 1.5 Shielding and Access Control Shielding is provided around much of the equipment to limit dose rates, as described in Section XII. Normal access to the turbine building is provided through the administration building. 2.0 Structure Design The turbine building houses the power generation and allied equipment. The equipment arrangement and principal dimensions are shown on Figures III-2 through III-11.
- Also see Section XVI, Subsection G.
UFSAR Revision 14 III-4 June 1996
Nine Mile Point. Unit 1 FSAR 2.1 General Structural Features The poured-in-place reinforced concrete building substructure and turbine generator foundation are founded on firm Oswego sandstone 15 ft to 25 ft below grade. The maximum bearing pressure on the rock, as recommended by consultants, is 40 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. Some of the actual bearing pressures on the confine rock are as follows. Maximum Rock Structure Bearin Pressure Building column piers 27 tons/sq ft Crane column piers 20 tons/sq ft Walls below grade 13 tons/sq ft Turbine generator 24 tons/sq ft foundation The turbine generator foundation is isolated from the floors of the building to minimize transmission of vibration to the floors. This foundation is designed for stability under all conditions of loading, including vertical, horizontal and torque loads, and loads due to temperature changes, piping and seismic forces. Elastic deflection and vertical shortening of members and stresses resulting from such loading were taken into consideration. The turbine building superstructure consists of an enclosed structural steel frame. The lower 24 ft of building is covered with 8-in thick insulated precast concrete wall panels. From the 24-ft level to the roof, the building is enclosed with insulated metal wall panels made up of type FK 16 x 16 and FKX 12 x 12 metallic-coated interior liner elements, 1 1/2-in insulation with a minimum density of 2 1/2 pcf and 16 B & S gage F-2 porcelainized aluminum exterior face sheets, all manufactured by H. H. Robertson Company. The roof is covered with metal decking, insulation, and a 4-ply tar roofing material flashed at the parapet walls. Anrail overhead rolling door at the west end of the building provides car access into the building. 2.2 Heating and Ventilation System The turbine building ventilating system, shown on Figure III-12, is designed to provide filtered and heated air at an approximate rate of one change per hour, corresponding to 170,000 cfm. Two independent air supply systems are provided, each consisting of a fresh air intake, filter, electric heating unit, flow control damper, two fans, dampers and ductwork to distribute air to UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR various areas in the turbine building. Each fan system is capable of supplying one-half of the required air, and either of the two fans in each system is considered an installed spare. The air duct electrical heating units are automatically controlled to maintain the supply air temperature at the desired level. The exhaust air system consists of two full-capacity fans, with one fan considered an installed spare, and connecting ductwork designed to induce flow of air through areas of progressively higher contamination potential prior to final discharge to the stack. An air inlet is located in each room and at each piece of equipment or other place where radioactive contamination in the form of dust, gas or vapor could be released. Ducts from these areas lead to an exhaust air manifold with each duct having a manually set control damper. The radiation protection and laboratory facilities ventilating system, shown on Figure III-13, discharges directly to the turbine building exhaust duct. In case power to the turbine building ventilation system is lost, an alternate outside source of filtered and heated air is available to the laboratory area. This area includes the technician's office, instrument storage room, high level lab, low level lab, counting room, auxiliary counting room and instrument calibration room. shunt circuit draws air from the exhaust manifold and monitors A its airborne radioactivity. The circuit is located so that monitors building air conditions and not the exhaust from it equipment vents. High activity causes alarm in the Station control room. The exhaust system discharges into the plenum which also receives air from the containment and other buildings, as shown on Figure VI-24. Backflow from other systems to the turbine building is prevented by interlocks which require valves to be closed exhaust fans are not in operation. if the The turbine building atmosphere is automatically controlled at a negative pressure of about 0.1 in of water relative to the outside by modulating the flow control dampers on the air supply systems. This is to control release of contaminated air and prevent out-leakage. When the turbine building roof vents are opened during operation, the turbine building differential pressure may approach zero in localized areas. In such cases, supplemental monitoring is instituted to prevent an unmonitored release to the environment. Electrical heaters are provided in various areas of the building for auxiliary heat should the ventilation system not be in UFSAR Revision 14 III-6 June 1996
Nine Mile Point Unit 1 FSAR operation for any reason. Water-cooled heat exchanger cooling units are provided in areas surrounding the extraction heaters, moisture separators, condensate circulating pumps and reheaters to dissipate the radiant heat loss from this equipment and to maintain desired temperatures for personnel comfort and equipment protection. The cooling water is supplied from the turbine building closed loop cooling water (TBCLCW) system. 2.3 Smoke and Heat Removal Smoke and heat removal capability is provided for the three smoke zones on el 250 of the turbine building and the upper elevation of the turbine building. Twelve motor-operated vents are installed in the roof over the turbine generator, and five sidewall vents are installed in the wall at el 351. A fire which produces low heat but a large concentration of smoke will be vented through the roof and sidewall vents. This capability is provided by manual actuation of the motor-operated vents. High heat and high smoke fires will automatically open the roof vents when the fusible link trips. In addition, the railroad access door on el 261 will be remotely opened to assist in smoke purging. 2.4 Shielding and Access Control Personnel access into the turbine building is controlled from the administration building at el 248'-0". An elevator for operating personnel serves the entire seven floor levels in the turbine building and is located at H row between column lines 11 and 12 (Figures III-4 through III-9). Stairs are also provided alongside the personnel elevator to serve the seven floor levels. In addition to the main or full-height. stairs, stairs are provided at four locations at grade for accessibility to floors above grade, and at seven locations to serve floors below at el 250 and 237. Walls, floors and roofs around equipment containing radioactivity are designed to have concrete thicknesses which significantly reduce radiation levels, as discussed in Section XII. 3.0 Safety Analysis The turbine building walls are of noncombustible material consisting of poured-in-place concrete, precast concrete, or insulated metal panels. The turbine room internal roof also consists of noncombustible material. Metal decking spans the steel purlins and is covered with rigid insulation and 4-ply built-up roofing material. All floors are of noncombustible material: either poured concrete or steel grating. Pressure relief to prevent failure of the superstructure due to a steam line break has been provided in the metal wall siding on the north wall of the crane bay (column Row C). UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR A peripheral drain at the exterior of the building provides for the removal of groundwater seepage and discharges into a sump pit with pump at the low point of all the buildings (southwest exterior corner of the reactor building). A rock dike 1000-ft long at the shoreline protects the Station from lake wave action or possible ice accumulation. The dike is 2 ft higher than yard grade and is constructed of rock from the Station excavation. Large rocks face the lake side of the dike and have proven very effective in wave damping and as a barrier to floating ice. The turbine building grade floor at el 261 is 12 ft above maximum lake level (el 249). Poured-in-place concrete foundations enclose the turbine building below grade floor level, and preformed rubber water stops are incorporated in the concrete construction joints for watertightness. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR B. CONTROL ROOM The control room is located in the southeast corner of the turbine building at el 277. It is bounded by the administration building offices on the south and east, the turbine room on the west, and the control room break area, instrumentation and control (I&C) office area, and diesel building on the north. 1.0 Design Bases 1.1 Wind and Snow Loadings The wind and snow loadings for the control room are the same as for the turbine building. 1.2 Pressure Relief Design There are no special pressure relief requirements for the control room. 1.3 Seismic Design and Internal Loadings The structural design for the control room, as well as the auxiliary control room below at el 261, is Class I seismic based on the maximum credible earthquake motion outlined in the introduction to Section III. Components are also designed as Class I. The seismic analysis resulted in the application of acceleration factors of 20.0 percent gravity horizontal and 10.0 percent gravity vertical. These acceleration factors were calculated from the dynamic analysis of the turbine building. Although the control room is structurally a part of the turbine building, functional load stresses when combined with stresses due to earthquake loading are maintained within the established working stresses* for the structural material involved. 1.4 Heating and Ventilation Heating and air conditioning are provided for personnel comfort and instrument protection. The ventilating system also provides clean air to the control room following an accident. 1.5 Shielding and Access Control Normal access to the control room is provided from the administration building through security-controlled doors. Shielding is supplied to allow continuous occupancy during any reactor accident. The most limiting accidents are the main steam line (MSL) break accident and the loss-of-coolant accident (LOCA) without core spray, which are described in Section XV. As
- Also see Section XVI, Subsection G.
UFSAR Revision 14 III-9 June 1996
Nine Mile Point Unit 1 FSAR stated in the First Supplement to the PHSR, personnel in the control room would not receive more than the hourly equivalent of the maximum permissible quarterly radiation dose according to 10CFR20. In addition, the concentration of radioactive materials in the control room during all credible accidents would be within the limits for restricted areas given in Paragraph 20.103 and Table I, Appendix B of 10CFR20. If air outside the building is contaminated, the ventilating system will be controlled to assure that contamination within the control room is minimized and kept within the above limits, as shown in Section 3.0, following. 2.0 Structure Design Plans showing location and principal dimensions are shown on Figures III-4, III-5, and III-6. 2.1 General Structural Features The structural steel enclosing the control room and the auxiliary control room below is supported on concrete walls and concrete foundations bearing on and keyed into sound rock. Actual rock bearing pressures are less than one-third of the allowable working bearing pressure. Lateral earthquake forces or wind loads are transmitted to the concrete foundations by the combination of structural steel bracing and concrete walls. The control room walls, roof and floors are framed with structural steel. The west and north interior walls are 12-in solid reinforced concrete. The east wall is enclosed with insulated metal wall panels made up of FK-16 x 16 metallic-coated interior liner elements, 1 1/2-in insulation and 16 B 6 S gage F-2 porcelainized aluminum exterior face sheets, as manufactured by H. H. Robertson Company. The wall panel joints are sealed with a synthetic elastomer caulking material. This wall is3-in separated from the administration building extension by a rattle space. The south interior wall consists of 8-in concrete blocks laid with steel-reinforced mortar joints. An interior metal partition wall parallel to the south wall forms a 6'-6" corridor and is provided with windows for observing the control room operations from the corridor. The slab immediately above the control room at el 300 is pinned to the walls and provides radiation shielding, and consists of 8 1/2-in thick poured-in-place reinforced concrete supported on structural steel beam framing. Two-thirds of this slab area has a roof above at el 333 which is made up of 3-in deep metal decking, 2 in of insulation and a 5-ply roof with slag surface. The remaining third of the slab area provides a shielding roof over the control room and consists of the 8 1/2-in thick poured-in-place reinforced concrete slab to which is applied 1 1/2 in of rigid insulation and a 5-ply roof with slag surface. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR The control room floor is poured-in-place reinforced concrete on 14-gauge metal decking. The gross depth of the floor slab is 8 in and the average depth of concrete is 5 3/4 in. 2.2 Heating, Ventilation and Air Conditioning System The ventilation system shown on Figure III-14 is designed to provide air at a rate of approximately 16,300 cfm to the control room and auxiliary control room areas. Outside air enters the system through a louvered intake after which normal supply isolation damper, which is it passes through a interlocked with an emergency ventilation inlet damper. The air then passes into the outside air mix damper which is set at 100-percent open position. Outside air is needed to recoup air from leakage and losses. The air is then mixed with recirculated return air from the recirculation damper which is set at 12,750 cfm minimum. The total amount of air (16,300 cfm) will then pass through a two-element dust filter. Next, it passes through a cooling coil where it will be cooled, if necessary, to maintain the control room temperature at approximately 75 F. The cooled air enters the control room circulation fan for distribution to various areas through ducts. Air will circulate through the control room to the return ductwork for recirculation and mixing with additional outside air. In order to prevent infiltration of potentially contaminated air, doors are weatherstripped and penetrations are sealed to maintain a positive pressure of approximately one-sixteenth of an inch of water. In the event of outside air contamination, the normal supply dampers will be automatically closed, and upon a high radiation signal, the emergency inlet dampers will be opened. The one outside air will then flow through a 15-kW duct heater and then ventilation of fans. the two full-capacity control room emergency The design flow range for the control room emergency ventilation system is 2875 cfm + 10 percent. This is the air flow range determined to maintain a positive pressure of 0.0625 in W.G. It then passes through a high-efficiency particulate filter andwill then through a heated activated charcoal filter unit. This air then join the normal ductwork and enter the outside air mix damper to be circulated by the normal ventilation fan. Heating is provided by thermostatically-controlled ventilation duct heaters. Cooling is provided by two chiller units. Tests and inspections on the control room emergency ventilation filters are done in accordance with Technical Specifications. 2.3 Smoke and Heat Removal To assist in maintaining a habitable atmosphere in the control room and auxiliary control room, a smoke purge capability is provided from two independent fans, one 6000-cfm makeup fan and one 8000-cfm exhaust fan (Figure III-14). UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR 2.4 Shielding and Access Control Normal personnel access to the control room is provided by three controlled access doors all located on el 277. The north door opens into the control room break area, the south door opens into the administration building, and the west door opens into a corridor, giving access to the administration building at el 277 and also making available the stairway to el 261 of the administration building. In addition to the above, a stair is provided within the control room (northwest corner) down to the auxiliary control room on the ground floor, shown on Figure III-4. In case of a reactor accident, personnel access to or from the control room would be from the southerly extreme of all buildings and approximately 400 ft from the center of the reactor. The walls, roof and floors are designed to have concrete thicknesses which provide shielding during the design basis accident (DBA). 3.0 Safety Analysis The control room is designed for continuous occupancy by operating personnel during normal operating or accident conditions. Concrete shielding provided in the roof and floors above and in the walls facing the reactor building is more than sufficient to prevent dose rates from exceeding the hourly equivalent of the 10CFR20 quarterly radiation dose. Maintaining positive pressure inside the control room and regulating the filtered outside air supply prevents the concentration of radioactive materials from exceeding the limits of 10CFR20. In addition, supplied air respirators are available in the control room for use if necessary. Both normal and emergency lighting are provided in the control room together with communications, air conditioning, ventilation, heating and sanitary plumbing facilities. If normal electric power service is not available, provision has been made to power the cooling, ventilating and heating units from the emergency diesel generators. Building components and finish materials are noncombustible and combustible materials are not stored in the control room. The minimum distance of the control room to the centerline of the reactor is 330 ft and there are no direct connections from passageways, ventilating ducts or tube connections between the reactor building and the control room. The floor of the control room is 16 ft above yard grade and 28 ft above maximum lake level (el 249). Therefore, the possibility of flooding or inundation is incredible. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR C. WASTE DISPOSAL BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings Wind and snow loadings for the waste disposal building are the same as for the turbine building. 1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 1.3 Seismic Design and Internal Loadings The waste disposal building and major components within are designed as Class I structures. The analysis of stress levels used the following earthquake design coefficients. Percent Gravit Horizontal Vertical Elevations 225 and 229 11.0 5.5 Elevation 236-6 11.5 5' Elevations 246-6, 247 12.2 5.5 and 248 Elevation 261 17. 0 7.33 Elevation 277 (276-6) 30.7 7.33 Roof Elevation 289 30.7 7 '3 Exterior walls of the substructure are designed for an earth pressure at any depth equal to the depth in feet times 90 psf. The exterior walls of the substructure and the base slab are designed to resist hydrostatic pressure and uplift due to exterior flooding to el 249. Except where concentrated loading due to the handling and placement of equipment requires construction of greater strength, the substructure floors are designed for dead loads plus the following: UFSAR Revision 14 III-13 June 1996
Nine Mile Point Unit 1 FSAR Live Loads Elevations Pounds Per S Ft 225 and 229 Unlimited 236-6, 237 and 248 350 241 and 247 250 The grade floor at el 261, including the concrete shielding plugs which close hatchways over equipment in the substructure, is designed for a uniform live load of 450 psf; or in the loading area a concentrated loading pattern produced by an AASHO* H20 loading, or 1000 psf, whichever requires the stronger construction. 1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort, equipment protection and for controlling possible radioactivity release to the atmosphere. 1.5 Shielding and Access Control Shielding is provided around tanks and equipment to maintain dose rates as described in Section XII. Normal access to the waste disposal building is from the turbine building. 2.0 Structure Design Floor and roof plans, exterior elevations, sections showing interior walls, and architectural details of the building are shown on Figures III-2 through III-6 and Figure III-11. 2.1 General Structural Features The poured-in-place reinforced concrete building substructure is founded on firm Oswego sandstone. The maximum bearing pressure on the rock as recommended by consultants is 40 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. The building has a flat roof consisting of a cellular metal deck covered with insulation and a bitumen and felt roofing membrane. The exterior facing of the superstructure walls is of sheet metal, attached either to an exterior shielding wall or to insulated cellular sheet metal wall. The interior walls of the
- American Association of State Highway Officials.
UFSAR Revision 14 III-14 June 1996
Nine Mile Point Unit 1 FSAR substructure are of cast-in-place concrete and those for the superstructure are either cast-in-place or made of concrete masonry units. With minor exceptions, all structural floors are poured-in-place concrete slabs. The superstructure frame is of fabricated steel. The north section of the basement is divided into three levels. These floors are for the storing of solid radioactive waste in metal drums until it permanent disposal area. is suitable for offsite shipment to a Each of these storage areas is served by a pair of lifts for drums, one being located near each side of the building. The intermediate level floor elevation is for the storage of evaporator bottoms and filter sludge prior to solidification. The south section of the basement provides space for the temporary storage, pumping and processing of radioactive liquid waste as described in Section XII. The loading area for receiving empty waste drums and equipment as described in Section XII is located on el 261 (Figure III-4). The designed control for spilled liquid is to allow the fluid to seek a lower level and, thus, be accommodated by the sumps which contain the fluid, and pump it directly to storage tanks. All drainage sumps have smooth linings of steel plate with all joints welded. The waste drum filling area also has a drainage gutter lined with half of a steel pipe. These designs are to facilitate cleanup by preventing contaminated liquids from permeating the concrete shell of the sump pit or gutter. 2.2 Heating and Ventilation System The heating and ventilating system, shown on Figure III-15, is designed to supply filtered and heated air at approximately 9,000 cfm and exhaust it after filtration. This corresponds to about one change of air per hour. No air is discharged from the building except through the stack. The supply fans, exhaust fans and exhaust filters are provided with full-capacity backups. Either supply fan and either exhaust fan can then be used to operate the system while the other members of the pairs are on standby. Outside air is drawn into the system through a fixed louver housed above the roof of the building and protected by bird and insect screening. The air is drawn through a filter designed to remove dust, and an electric heater of 200-kW capacity. The heater is thermostatically controlled to warm the air to maintain at least 70 F in accessible areas. Beyond the heater section the supply duct is split with each half routed through a supply fan of 9,000 cfm capacity. Each fan is isolated in its section of duct by a butterfly valve damper on both inlet and discharge UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR sides. Beyond the fan discharge control dampers, the ducts rejoin into a common manifold from which supply ducts convey fresh air to various areas 'of the building. At or near the discharge point of each of these ducts, a manually set damper determines the fraction of air delivered at that particular point. The fresh air supply points are located where the rate of air contamination is lowest while the inlets to the exhaust ducts are located where the rate of contamination is likely to be the highest. An air outlet is located in each room and at each piece of equipment or other place where radioactive contamination in the form of dust, gas or vapor could be released. Ducts from these areas lead to an exhaust air manifold with each duct having a manually set control damper. shunt circuit draws air from the exhaust manifold and monitors is located so that it A its airborne radioactivity. The circuit monitors building air conditions and not the exhaust from equipment vents. High activity is alarmed in both the waste building control room and the Station main control room. Beyond this point, the exhaust duct divides into two full-sized parts, each of which contains a roughing filter followed by a high-efficiency filter and an exhaust fan as shown on Figure III-15. Butterfly valves in the ducts, before the filters, which between filters and fans, and following the fans determine of the alternate routes the exhaust will take and regulate the amount of air exhausted. From here on, the ducts are reunited and discharge to the plenum leading to the stack. Backflow from other systems is prevented by interlocks which require valves to be closed if the exhaust fans are not in operation. Each high-efficiency particulate filter in the exhaust system has a minimum removal efficiency of 99.97 percent based on the 0.3 micron "DOP" (dioctylphthalate smoke) test. Supplementing this exhauster system is a 300-cfm capacity auxiliary system, which exhausts air directly from the hydraulic filter baler through a roughing filter and a high-efficiency by means of a small exhauster fan, and discharges directly into the ventilation breaching. Also, a 5000-cfm capacity auxiliary a system exhausts directly from the drum filling area through roughing filter by means of a small exhauster fan, and discharges of the building ventilating system. to the exhaust duct Equipment vents and the sample Station hood discharge directly to the exhaust duct. Supplementing the heat supplied by the main intake air heater, small heating units are provided locally to maintain desired temperatures for comfort of personnel and protection of equipment. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR The ventilation system for the waste building extension is shown on Figure III-16. One of two full-capacity exhaust fans draws air at a rate of 5400 cfm from the waste building and distributes the air through ductwork to the various equipment rooms within the waste building extension. The air that passes through the system is discharged to the stack. 2.3 Shielding and Access Control Normal personnel access to the waste disposal building is from the turbine building through the waste disposal control room. Access doors from the turbine building are also located near the baler room. Access is also available through the truck loading bay located at the northeast, corner of the building. All access to the building is at grade level as shown on Figure III-4. All levels are accessible by steel stairways from the grade of floor and an emergency ladderway exit is provided for those parts the drum storage area which are remote from the stairs. Hatches are provided for access to equipment. Concrete thicknesses for both walls and floors are established to provide the degree of radiation shielding of radioactive waste adjacent to the shielded area. The reinforced concrete substructure completely isolates the basement and serves as shielding for adjoining basement areas. Each item or group of closely associated items of equipment is housed in a separate room, surrounded by concrete shielding walls of appropriate thickness to provide adequate protection to operating personnel as determined by the anticipated intensity of radiation and time duration of exposure. The waste disposal building control room is completely surrounded by shielding walls and with access so arranged that the room will be accessible at all times. 3.0 Safety Analysis The design and construction of the waste building has provided for all foreseeable conditions and loads. All structural material used is noncombustible and accumulation of combustible material is carefully avoided. As outlined in the detailed description of the structure, provision has been made that, should some unforeseen condition or accident release contaminated waste, the hazard would be localized and the size of the cleanup and decontamination job restricted. All tanks are made of ductile metal and all sump pits are lined so that these containers can be subjected to substantial distortion without rupture. The two rooms for the centrifuges on the grade floor are surrounded by heavy walls which serve a dual purpose by providing UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR both radiation and mechanical shielding. ln the extremely unlikely event that the centrifuge should suffer a mechanical failure, it would be contained within the room and prevent injury to operating personnel or damage to tanks, piping, pumps or other equipment outside the room. The substructure is massive reinforced concrete, not. subject to fracturing. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR D. OFFGAS BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings Exterior loadings for wind, snow and ice used in the design of the offgas building are the same as the turbine building. 1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 1.3 Seismic Design and Internal Loadings The offgas building is designed as a Class I structure. The analysis of stress levels used the following earthquake design coefficients. North-South East-West Elevation G G 289 37.2 32.0 276 19.3 24 ' 261 15.2 19.0 247 13.6 16.0 236 12.0 13.0 The live load design on the ground floor and intermediate subfloors is 300 psf. 1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort. 1.5 Shielding and Access Control Shielding is provided around tanks and equipment to maintain dose rates as described in Section XII. Normal access to the offgas building is from the turbine building. 2.0 Structure Design Floor and roof plans, exterior elevations, sections showing interior walls, and architectural details of the building are shown on Figures III-2 through III-9. 2.1 General Structural Features The substructure is constructed of cast-in-place reinforced concrete and is founded on firm Oswego sandstone. UFSAR Revision 14 III-19 June 1996
Nine Mile Point Unit 1 FSAR The maximum bearing pressure on the rock is 20 tons/sq ft. This results in a safety factor of 18 based on actual unconfined compressive strength tests on selected specimens of rock core extracted from test borings. The building has a built-up roof consisting of a cellular metal deck covered with insulation and asbestos felt and a gravel surface. The superstructure is structural steel frame with insulated exterior metal walls. The interior walls of the substructure are of cast-in-place concrete and those for the superstructure are concrete block with a 144-pcf density for shielding. With minor exceptions, all structural floors are poured-in-place concrete slabs. The basement is divided into two levels. El 229 houses the charcoal column tank room. Located on el 232 is the chiller system compressors and deicing water buffer tank rooms. The next floor is divided into three levels. The main level el 247 houses the three chiller rooms and equipment hatch. El 244'-9" houses the two preadsorber rooms, and at el 250 is grating surrounding the charcoal tanks. Normal personnel and equipment access from the turbine building is located on el 261. Also located on this level are equipment plugs, equipment hatch and stair openings to the levels below. 2.2 Heating and Ventilation System The heating and ventilation system is shown on Figure III-17. One of two exhaust fans with a full capacity of 6,000 cfm draws air at a rate of 5400 cfm from the turbine building and distributes the air through ductwork to the various equipment rooms within the offgas building. The air that passes through the system is discharged to the stack. 2.3 Shielding and Access Control Normal personnel access to the offgas building is from the turbine building. An access door from the waste disposal building is also provided. All access is located on grade level 261. All levels of the offgas building are accessible by steel are stairways from the grade floor. Equipment plugs and hatch provided for access to equipment. Concrete thicknesses for both walls and floors were established to provide adequate radiation shielding consistent with as low as reasonably achievable (ALARA) criteria. 3.0 Safety Analysis The design and construction of the offgas building has provided for all foreseeable conditions and loads. UFSAR Revision 14 III-20 June 1996
Nine Mile Point Unit 1 FSAR All walls, floors and roof are of noncombustible materials. Equipment is housed in rooms with walls, floors and shield walls appropriately designed to provide adequate shielding to meet ALARA criteria. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR E. NONCONTROLLED BUILDINGS 1.0 Administration Building The administration building is a one and two-story structure adjoining the turbine building on the south and east. 1.1 Design Bases 1.1.1 Wind and Snow Loadings The wind and snow loadings for the administration building are the same as for the turbine building. 1.1.2 Pressure Relief Design There are no special pressure relief requirements for the administration building. 1.1.3 Seismic Design and Internal Loadings The administration building is designed as a Class II and III structure. The original administration building was designed as a Class III structure with no special seismic criteria. The following design live loads were used in addition to the dead loads for the original administration building. Elevation 261 Store room and shop room 1000 psf Other Areas 150 psf Elevation 277 Office areas, including areas for office equipment and personnel, corridors, stairways and other related areas 125 psf The administration building extension is designed as a seismic Class II structure. A portion of the extension is located over the diesel generator rooms requiring an upgraded seismic classification. The extension is designed to accommodate the same seismic loads as the control room and diesel generator rooms. The criteria used for the administration building extension are:
- 1. Normal allowable stress* levels were used. (However, up to 1/3 overstress was permitted.)
- Also see Section XVI, Subsection G.
UFSAR Revision 14 III-22 June 1996
Nine Mile Point Unit 1 FSAR 2 ~ Horizontal north-south and east-west earthquakes were not combined but were considered separately. 3 ~ Vertical accelerations were assumed to be 1/2 of the horizontal. 4 ~ Accelerations and deflections caused by the earthquake are: North-South East-West Elevation O Q <o G 300 34. 0 30.0 277 19. 0 18.0 261 13. 0 13.0 250 12. 0 12.0 1.1.4 Heating, Cooling and Ventilation Heating, cooling and ventilation are provided for personnel comfort. 1.1.5 Shielding and Access Control
~ ~
No shielding is required.
~
1.2 Structure Design The administration building, shown on Figures III-3 through III-5, contains all the facilities required offora nuclear administrative and technical servicing functions required generating station. 1.2.1 General Structural Features The administration building is a steel-framed structure with cellular metal and concrete floors and exterior walls of insulated sandwich precast concrete slabs. The exterior walls of the administration building extension are metal siding. The exterior south and west walls of the women's locker room and the foam room are masonry walls. The building has three levels. The basement (el 248) houses the onsite Technical Support Center (TSC). The TSC meets the requirements of NUREG-0578. The layout of the TSC and its proximity to the control room is shown on Figure III-5. This level is also used for storage, additional office space, and entrance to the turbine building and personnel locker room. UFSAR Revision 14 III-23 June 1996
Nine Mile Point Unit 1 FSAR The ground floor (el 261) is divided into three parts. One of these is assigned to Station stores. The remaining two are assigned to shops. The balance of the ground floor contains an ante room and a foyer for the stairway and elevator to the general offices on the second floor. The room for equipment and materials which produce fire extinguishing foam is also in this area. On the upper level (el 277) are the stair, elevator lobby, restrooms, offices, conference rooms, and a satellite document control station. Document control, microfilming facilities, and the record storage facility, in accordance with ANSI N45.2.9-5(6), are located at Nine Mile Point Nuclear Station Unit 2 (Unit 2). 1.2.2 Heating, Ventilation and Air Conditioning Ventilation for the administration building and the administration building extension is provided as follows. One self-contained rooftop air conditioning unit, one supply fan, three exhaust fans, and associated ductwork and equipment provide ventilation to the original administration building. Five supply fans, associated ductwork and equipment supply air to the administration building extension. Individual heating and air conditioning units are provided throughout the original administration building and the administration building extension for personnel comfort. The onsite TSC located on el 248 is provided with an air filtering system which is housed in the charcoal filter building at el 261 (see Figure III-18). 1.2.3 Access Control Normal access to the administration building is provided by two doors located on the west side of the building. Three overhead doors are located on the south side of the building to provide access to the shops and stores at the 261 ft level. 1.3 Safety Analysis No radioactivity complications exist at any of the noncontrolled buildings. Fire hazard is low since construction is of fire-resistant, materials and each building has a minimum of combustibles. UFSAR Revision 14 III-24 June 1996
Nine Mile Point Unit 1 FSAR 2.0 Sewage Treatment Building The new sewage treatment facility (STF), which utilizes part of the existing STF, is located in the vicinity railroad track of spur no. 3 that was removed for construction, approximately 300 ft northwest of the turbine building and due west of the north end of the reactor building as shown on Figure III-1. The site was selected based on review of available areas outside the flood plain for a Unit 2 10,000-yr flood year flood (rain). The existing STF was modified to function as a raw sewage pump station and an equalization tank for the new STF. The control building for the new STF is located between and to the south of the circular extended aeration units. The control building houses a new laboratory, a motor control center (MCC), blower room, storage room, maintenance room and hypochlorite room, as well as an influent/effluent room. Normal access to the treatment units is from inside the control building's influent/effluent room. Maintenance and emergency access to the treatment unit may be from outside access doors on each tank. 2.1 Design Bases 2.1.1 Wind and Snow Loadings The wind loadings for the sewage treatment building are the same as for the turbine building. The snow loading for the building roof is 14 lb/ft~. 2.1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 2.1.3 Seismic Design and Internal Loadings The sewage treatment building is designed as a Class III structure with no special seismic criteria. The system conforms to state regulations for sewage systems. 2.1.4 Electrical Design In certain areas of the building, electrical components are protected by explosion-proof enclosures. 2.1.5 Fire and Explosive Gas Detection Automatic fire detection equipment is provided in the STF. The fire detection equipment actuates alarms on local fire panels in the STF which informs personnel of fire location. Automatic gas detection equipment is provided for chlorine, and for methan and other explosive gases. The detection equipment actuates an alarm bell and warning lights inside and outside the STF. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR Both systems are provided for personnel safety and equipment protection. 2.1.6 Heating and Ventilation Heating and ventilation is provided for equipment protection and personnel comfort in accordance with the required codes. 2.1.7 Shielding and Access Control Shielding is not required. 2.2 Structure Design 2.2.1 General Structural Features The sewage treatment plant will provide secondary treatment and disinfection for a minimum flow of 10,000 gal/day and a peak flow of 240,000 gal/day. Wastewater flows by gravity from Nine Mile Point Nuclear Station Unit 1 (Unit 1) facilities, the Energy Information Center (EIC), the Nuclear Learning Center (NLC), and Unit 2 to the existing Unit 1 sewage treatment plant building and associated preliminary treatment facilities. After preliminary treatment, the flow is pumped to the extended aeration units. Flow through the remainder of the plant is by gravity. Discharge from the plant is through a 12-in outfall sewer to a drainage ditch leading to Lake Ontario. Flow measurement is available and is recorded on stripcharts. Raw sewage will pass through a comminutor to shred large solids. Two comminutors are provided, each capable of treating flows up to 300,000 gal/day. In the event of failure of both comminutors, a bypass hand-cleaned bar screen is provided to protect the raw sewage pumps from large solids. Raw sewage is then pumped to the new treatment facilities. Pumping after preliminary treatment minimizes the need for rock excavation for downstream treatment units. A 4-in and 6-in dual-force main is used to meet the anticipated flow range of 35,000 gal/day to 240,000 gal/day. A three-pump raw sewage station is utilized with two pumps operating and the third pump acting as an installed standby. Wastewater pumped to the new treatment facilities will enter a flow distribution structure and will be split equally by weirs to two extended aeration units. Each unit contains two equally-sized basins of 2800 cu ft, while affording maximum control and operational flexibility. At double outage design conditions, two units each with two basins of this size would provide an average hydraulic detention time of approximately 17 hr with an average organic loading of about 18 lb biological oxygen demand (BOD) per day per 1000 cu ft of tank volume. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR The aeration system for the activated sludge process is a coarse-bubble diffused air system. A total of three air blowers (including standby) are provided, having a total capacity of 700 scfm. These blowers will provide approximately 3200 cu ft of aeration air per pound. The mix liquor is then sent to the activated sludge settling tank where the sludge solids are separated. This produces a well-clarified effluent low in BOD and suspended solids. Each treatment unit. contains an 18-ft diameter clarifier with 12-ft side water depth. These tanks are center feed clarifiers with radial outward flow. At double outage design conditions, the tanks will have an overflow rate of 240 and 470 gal/day/sq ft at average peak flows, respectively. Scum is to be removed from the surface of the final settling tanks by a rotary wiper arm. Scum from the surface of the settling tank is drawn over a short inclined beach and is discharged to a scum trough. The scum is then flushed to a scum well from which it is air lifted to the aerated sludge holding tanks. To maintain the activated sludge in an active condition, final sludge is removed from the settling tanks continuously. Sludge withdrawn from the final settling tanks is returned to the aeration tanks at a rate to maintain a constant mixed liquor suspended solids and solids retention time in the aeration tanks and to avoid excessive sludge depths in the settling tanks. Return sludge air lifts are used to return sludge to the head of the aeration tank. Excess sludge solids will be wasted from the settling tanks and air lifted to aerated sludge holding tanks to be concentrated prior to sludge dewatering. Hypochlorite is used for disinfection of the final effluent at the new treatment facilities. Each treatment unit includes a separate chlorine contact zone of 170 cu ft which provides 15 min and contact at the peak flow of 240,000 gal/day. detention time Each treatment unit contains an aerated sludge holding tank of approximately 2000 cu these tanks provide in ft each. At double outage design flows, excess of 30 days sludge storage. Each treatment unit is furnished with an aluminum geodesic dome cover for winterization protection. Each dome is equipped with two skylights and one gravity vent to provide natural lighting and ventilation. The walls of the treatment units are extended to support the domes and provide a workable clear headroom height along the interior circumference of the treatment unit. The domes are designed to be removable as a complete unit. 2.2.2 Ventilation System The STF is air conditioned and electrically heated. Unit air conditioners in the lab room only and heating coils for ventilation air are located throughout the facility where required. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR 2.2.3 Access Control The equipment house has no windows except in certain doors and a lock on the door prevents access by unauthorized personnel. 3.0 Energy Information Center The EIC is a single-story flat-roofed structure located on a slight promontory 1000 ft west and slightly south of the Station (Figure III-1). 3.1 Design Bases 3.1.1 Wind and Snow Loadings Exterior loadings for wind, snow, and ice used in design of the EIC meet all applicable codes as a minimum. The roof and its supporting structure are designed to withstand a loading of 40 psf of snow or ice. The walls and building structure are designed to withstand an external or internal loading of 40 psf of surface area, which is approximately equivalent to a wind velocity of 125 mph at the 30-ft level. 3.1.2 Pressure Relief Design There are no special pressure relief requirements for the EIC. 3.1.3 Seismic Design and Internal Loadings The EIC and components are designed as Class III structures with no special seismic criteria. The following design live loads were used in addition to the dead loads: Live load on stairways and all public areas except restrooms 100 psf. Live load on all other floorareas including the classroom, offices and conference room 60 psf. Allowable bearing pressure on undisturbed soil foundations of 1.5 tons/sq ft. Stresses in steel construction are those allowed by the AISC 1963 Specifications for the Design, Fabrication and Erection of Structural Steel for Buildings when using ASTM A36 Structural Steel. Stresses in concrete construction are those allowed by the ACI 318-63 Standard for 3000 psi concrete with intermediate grade new billet steel A-15. UFSAR Revision 14 III-28 June 1996
Nine Mile Point Unit 1 FSAR 3.1.4 Heating and Ventilation Heating and ventilation is provided for personnel comfort. 3.1.5 Shielding and Access Control No radioactivity is contained in or near the building; therefore, no shielding is required. 3.2 Structure Design 3.2.1 General Structural Features As shown on Figure III-1, the principal part of the building is in the form of a regular hexagon with sides 56-ft long. A wing of irregular shape but approximately 96-ft long by 36-ft and 45 1/2-ft wide extends to the west. The lobby occupies the full width of the southwest portion of the principal part of the building. To the rear of the lobby are a small theater, a room for a model of the Station and a room for various exhibits. The building's core, central to these rooms, contains a storage room, a projection room for the theater and stairs for access to the basement. Public restrooms and a women's lounge are located in the wing and adjoin the lobby on the left. The wing also contains a classroom, a conference room, offices, a central corridor, an extension of the main lobby and three secondary entrances to the building. The EIC building has a structural steel frame resting on a concrete substructure. Its exterior curtain walls are of concrete block with a veneer of native stone, trimmed with redwood, and well insulated. Interior walls are plastered metal or gypsum lath on steel studding. The roof is comprised of a bituminous waterproofing membrane on rigid insulation which is carried by metal roof decking rolled and open web steel joist purlins, which are in turn supported by steel girders and fascia beams. A concrete slab, hexagonally shaped in plan, about 30 ft in diameter and 4-in thick is centrally located on the roof to serve as a platform for the air conditioning condensers. 3.2.2 Heating and Ventilation System The EIC is air conditioned and electrically heated. Compressors, heat exchangers, heating coils for ventilation air and other mechanical equipment are located in equipment rooms in the basement. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR 3.2.3 Access Control Access to the EIC is from a separate road than that leading to the rest of the Station. Each room to which the public will be admitted has doors of ample width to the rooms adjoining on either side and, in addition, the theater and the model roomthese each has its own exit door to the outside of the building. All provide ample egress from any area for any conceivable emergency. UFSAR Revision 14 III-30 June 1996
Nine Mile Point Unit 1 FSAR F. SCREENHOUSE, INTAKE AND DISCHARGE TUNNELS 1.0 Screenhouse The screenhouse adjoins the north wall of the reactor and turbine buildings and its superstructure is completely isolated from the reactor building. 1.1 Design Basis 1.1.1 Wind and Snow Loadings The wind and snow loadings for the screenhouse are the same as for the turbine building. 1.1.2 Pressure Relief Design There are no special pressure relief requirements for the screenhouse. 1.1.3 Seismic Design and Internal Loadings The screenhouse substructure has been designed to conform to the requirements for a Class I structure while loaded with any possible combination of filled and unwatered conditions of the channels located in this substructure. The superstructure is designed as a Class II structure as discussed on Page III-3 of the First Supplement to the PHSR. The seismic analysis resulted in the application of acceleration factors of 20.0 percent gravity horizontal and 10.0 percent gravity vertical. 1.1.4 Heating and Ventilation No heating, cooling or ventilation is provided for the screenhouse. 1.1.5 Shielding and Access Control No shielding is required. Normal access to the screenhouse is through the turbine building. 1.2 Structure Design The superstructure of the screenhouse is of framed structural steel supported on a reinforced concrete substructure which is founded on rock. The building has a flat roof consisting of cellular metal decking covered with insulation and a tar and felt roofing membrane. The two bays of the east wall, which are a continuation of an east wall of the turbine auxiliaries building extension, are of the same insulated sheet metal construction. The balance of the exterior wall, about 7/8 of the total, is of 8-in internally-insulated precast concrete panels corresponding with those in the base of the reactor building walls. Wall and UFSAR Revision 14 III-31 June 1996
Nine Mile Point Unit 1 FSAR roofing material and construction are identical with those used for the reactor and turbine buildings. The screenhouse substructure comprises channels for the flow of very large quantities of raw lake water, gates and stop logs for control of the flow, racks and screens for cleaning the water and pumps. The water channels are shown schematically on Figures III-19 and III-20. Five plain vertical gates near the north end of the substructure separate the channels from the tunnels. Gates A and B separate the intake tunnel from the forebay. Gate C separates the discharge channel from the discharge tunnel; gate E separates the discharge channel from the intake tunnel; and gate D separates the forebay from the discharge tunnel. Each of gates A, B, C, and D has a dedicated electric motor-driven hoist for raising, lowering, and maintaining position of the gates. Gate E is operated using a hydraulic ram system. Normal circulation is provided by opening gates A, B, and C with gates D and E closed. Reversed flow through the tunnels is obtained by closing gates A, B and C with gates D and E open. Tempering (partial recycle flow) is obtained by partially opening gate E with all other gates set for normal operation. The forebay and the secondary forebay are connected by three parallel cool water channels, in each of which are located trash racks, rack rakes and traveling screens to remove trash, water plants and fish from the water. Each of these channels has provisions for stop logs at each end so that any one of them may be segregated and unwatered for maintenance work without shutting down the Station. On the floor above the secondary forebay are mounted four containment spray raw water pumps and two emergency service water (ESW) pumps with a strainer for each. Also on this floor and above each of the three cool water channels are the screen wash pumps. Adjacent to the secondary forebay, on its south side and separated from it by channels fitted with stop log guides, are inlet chambers for the two circulating water pumps which provide water to the main condensers. By means of stop logs, either of these chambers can be isolated for unwatering and work on the corresponding pump. A lateral branch leads off to the east from the secondary forebay. Three chambers off this branch, separated from it by sluice gates, supply water to each of two service water pumps with strainers and a pair of fire pumps. One of these fire pumps is driven by an electric motor, the other by a diesel engine. The screenhouse is also equipped with a floor-operated electric overhead traveling bridge crane. This crane serves the various functions of placing and removing stop logs, and servicing the trash racks, rack rakes and traveling screens, maintenance of the two circulating water pumps and all pumps mounted above the secondary forebay. The service water pumps, their strainers, and the fire pumps are serviced for maintenance work by overhead beam runs, trolleys and hoists. UFSAR Revision 14 III-32 June 1996
Nine Mile Point Unit 1 FSAR 2.0 Intake and Discharge Tunnels As shown on Figure III-21, water is drawn from the bottom of Lake Ontario about two-tenths of a mile offshore and returned to the lake about one-tenth of a mile offshore. 2.1 Design Bases The water intake and discharge tunnels are designed to conform to the requirements for Class II structures. The intake and concrete-lined bores through solid rock. discharge tunnels are As such, they are highly rigid structures with extremely small natural periods of vibration and a seismic response of only 11 percent of gravity regardless of the damping factor. 2.2 Structure Design Water is admitted to the intake tunnel through a bellmouth-shaped inlet. The inlet is surmounted by a hexagonally-shaped ft guard structure of concrete, the top of which is about 6 above the lake bottom and 14 ft below the lowest anticipated lake level. The structure is covered by a roof of sheet piling supported on steel beams, and each of the six sides has a water inlet about 5-ft high by 10-ft wide, with the latter openings water guarded by to be galvanized steel racks. This design provides for drawn equally from all directions with a minimum of disturbance and with no vortex at the lake surface, and guards against. the entrance of unmanageable flotsam to the circulating water system (CWS) . water drops through a vertical concrete-lined shaft to a The concrete-lined tunnel concrete-lined in the rock, through which vertical shaft under the it flows to the forebay in the foot of a screenhouse. The foot of this shaft contains a sand trap to catch and store any lake-bottom sand which may wash over the sills of the inlet structure. The top of the shaft has a bell-mouthed discharge. Water is returned to the lake at a point about one-tenth of a mile offshore through a bell-mouthed outlet surmounted by a hexagonal-shaped discharge structure of concrete. The top of this structure is about 4 ft above lake bottom and 8 the 1/2 ft below the lowest anticipated lake level. The geometry of structure closely resembles the inlet structure, although reduced in size. The six exit ports are about 3 ft high by 7 1/3 ft wide. The discharge 'tunnel from the screenhouse is identical in cross-section with the intake tunnel. The vertical shaft connecting the discharge tunnel with the discharge channel under the screenhouse also has a sand trap at its foot. Water is discharged directly to the vertical discharge shaft. A submerged diffuser in the vertical shaft ensures a good dilution before discharge to the lake. Samples are drawn at a lower point in the shaft. UFSAR Revision 14 III-33 June 1996
Nine Mile Point Unit 1 FSAR 3.0 Safety Analysis The selection and arrangement of equipment and components of the screenhouse and circulating water tunnels is based on the knowledge gained over many years of experience in the design, construction and operation of such facilities for coal-fired steam-electric stations. All components of the system which might possibly be subject to unscheduled outage, and by such outage affect the operability of the Station, are duplicated. In the case of the duplicate fire pumps, the prime movers are also totally independent. The gates are simple and rugged in construction, and their operation is simple and straightforward, with the possibility of inadvertent erroneous operation cut to a minimum. The pump suctions are amply submerged below the lowest low water surface elevation of the lake surface adjusted for the friction and velocity drops in the supply tunnel and channels. The supply of water by direct gravity from the lake is inexhaustible. The main portion of the superstructure, a single-story structure elastic frame of one bay width, has a relatively long natural period of vibration, and being bolted has a comparatively high damping factor. As a result, the dynamic loads which could be applied to it by wind pressure and also operation of the crane are more critical than those due to the seismic loading. Thus, while no dynamic analysis of the framing was required or made, it is quite probable that the building superstructure meets Class I conditions instead of only Class II, as specified in the First Supplement to the PHSR. Shearing forces in the walls and in the bottom chord plane of the roof truss system are resisted by systems of diagonal bracing. The sizes of the members of these systems were governed by detail and minimum allowable slenderness rather than by calculated forces, which resulted in excess strength being available in the system. UFSAR Revision 14 III-34 June 1996
Nine Mile Point Unit 1 FSAR G. STACK The stack is a freestanding reinforced-concrete chimney, 350-ft high, located 100 ft east of the northeast corner of the reactor building. 1.0 Design Bases 1.1 General The height of the stack and the velocity of discharge are to provide a high degree of dilution for routine or accidental Station effluents. This is discussed on Page IV-8 of the First Supplement to the PHSR. 1.2 Wind Loading Analysis shows that the loads due to seismic action are considerably greater than those which would be exerted by the velocity of wind for which the other Class I structures are designed: 125 mph at the 30-ft level. Since this is true for all levels of the stack (wind velocities and pressures varying according to elevation aboveground), lateral loads due to seismic forces govern the design. 1.3 Seismic Design The design and construction of the stack meet the seismic requirements of a Class I structure. Seismic forces applied are those obtained from the velocity and acceleration response spectra included in the First Supplement of the PHSR for a ground motion acceleration factor of 11 percent of gravity (Plate C-22). 1.4 Shielding and Access Control Shielding is required for the offgas and gland seal exhaust piping. Access is provided for inspection and maintenance during shutdown. 2.0 Structure Design The general features of the stack, including its principal dimensions, are shown on Figure III-22. It is a tapered monolithic reinforced-concrete tube resting on a massive concrete base which extends to sound rock. From this base it rises it through the turbine auxiliaries building extension from which is completely isolated structurally. The top of the stack is at el 611, or 212 ft 6 in above the top of the reactor building, the next highest structure in the Station. After filtration, all Station ventilation exhaust which is radioactively contaminated is brought to the stack through UFSAR Revision 14 III-35 June 1996
Nine Mile Point. Unit 1 FSAR breaching, which is connected above the roof of the surrounding building. Two pipes, 6 in and 12 in in diameter, bring radioactively contaminated gases and vapors from the turbinestack shaft seals and from the condenser. These pipes enter the below the grade floor and turn up through encasing concrete to a terminal point at el 335, which is 20 to the stack. At ft this above the top of the point turbulence is breaching entrance high, which ensures best mixing and dilution of the contaminated gases. An >>Isokinetic Probe" gas sampler is located within the stack with its orifices at el 535, or 76 ft below the top of the stack. This device is supported by a beam which spans the interior of the stack and cantilevers outside to facilitate withdrawal of the device for cleaning and maintenance. An opening is provided in the stack wall through which the device is installed. This opening is a 16-in diameter pipe sleeve with its outer end closed by a blind flange. A smaller adjoining opening makes to it possible visually to measure the gas velocity profile in the stack or inspect the probe without withdrawing it. The probe is connected to monitoring equipment located near the base of the stack by tubing which descends inside the stack. Access to the interior of the stack is through an airtight door from the basement of the surrounding building. Exterior access to the top of the stack and to four external platforms is from the roof of the building by means of a guarded ladder. At the probe level a small platform provides access and working area. Three other platforms completely surround the stack which provide access for external maintenance and painting of the stack. The stack is protected by four lightning rods and down conductors which are interconnected at the top, middle and bottom of the stack, then connected to the Station grounding grid. The structural reinforcing steel, platforms and ladder are in turn grounded by attachment to this system. The top of the stack is, in effect, an 8-ft 6-in inside diameter nozzle. For normal gas flows of 216,000 cfm, the corresponding velocity of the discharge jet is 63 fps. This relatively high velocity assures that the turbulence generated will thoroughly mix, dilute and disperse the discharged gas even at times of low wind velocity. 3.0 Safety Analysis 3.1 Radiology If during normal operation the stack were to be inoperative, there would be no serious radiological consequences for a period of time depending on the level of activity being released. If the stack were to remain inoperative for a significant length of time, the reactor would be shut down to prevent exceeding 10CFR20 UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR limits. Exfiltration cases involving an inoperative stack are discussed in Section XV. 3.2 Stack Failure Analysis In the event that portions,~of the stack strike the plant, structural analysis indicated that the stack would topple with approximately the upper 3/4 (280 ft) intact. As a structural element the stack is weak in circumferential bending. This means that the stack cross-section would flatten to out-of-round or oval when it it struck, spread the load over a larger area than had remained circular, and absorb energy in doing so. Since the stack is strong longitudinally, span from girder to girder. it would tend to span openings or The consequences of the stack striking the plant have been evaluated by what is believed to be the three most critical directions (see Figure III-23).
- 1. Southwest, striking the reactor building
- 2. South, striking the diesel generator building
- 3. Northwest, striking the screen and pump house 3.2.1 Reactor Building A considerable amount of energy would be absorbed as the stack fell through the braced walls, the roof trusses and the crane considerations taken into account, it is girders. With the above unlikely that the stack would penetrate the bottom of the fuel pool or the shield plugs over the reactor. The worst conditions would occur damaged.
if one or both of the emergency cooling systems were Since the emergency cooling return lines are equipped with check valves, the only flow path would be out the supply lines to the emergency cooling system. The isolation valves in this line will automatically close on high flow in the line. High temperature in the vicinity of the line and high radiation are alarmed in the control room, resulting in manual closure of the isolation valves. Because of the angular separation between the diesel generator and the reactor building, the diesel area would not be affected by failure of the stack in the direction of the reactor building. The battery room is outside the reach of the stack regardless of the direction in which the stack is assumed to fall. Should they be needed, all sources of electric power remain available to safeguard systems. Adequate protection is therefore afforded in this case. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR 3.2.2 Diesel Generator Building Failure of the stack in the southerly direction could damage the diesel generators. Since the control room is 350 ft from the stack and the upper 3/4 of the stack is approximately 280 ft, it is highly improbable that the control room would be damaged. If would not be damaged. Normal sources power building failure were in the southerly direction, the reactor would of electric be available to conduct a safe shutdown. 3.2.3 Screen and Pump House If the stack fell due north, the diesel fire pumps, the diesel generator cooling water pumps, and associated piping systems could become inoperative. If the raw stack fell within the northwest quadrant, the containment water, circulating water and service water pumps, as spray well as the lines from the diesel fire pumps, could be damaged. However, safe shutdown could still be afforded by use of the normal supplies of electric power and the emergency cooling system. UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR H. SECURITY BUILDING AND SECURITY BUILDING ANNEX The security building and security building annex are located on the southwest corner of the Station security perimeter. See Figure III-1. The principal function of these buildings is to monitor controlled ingress and egress of personnel and equipment to the Station security perimeter. Administrative offices are contained within these buildings for support of the duties associated with Station security. Because of the nature of this subject, a detailed description of these buildings will not be discussed in this document. For additional information regarding this subject, refer to the Station security plan. UFSAR Revision 14 III-39 June 1996
Nine Mile Point Unit 1 FSAR I. RADWASTE SOLIDIFICATION AND STORAGE BUILDING 1.0 Design Bases 1.1 Wind and Snow Loadings Wind and snow loadings for the radwaste solidification and storage building (RSSB) are designed to meet or exceed those of the waste disposal building. 1.2 Pressure Relief Design There are no special pressure relief requirements for this building. 1.3 Seismic Design and Internal Loadings+ The foundation mat, structural walls, columns, floors and roof of the RSSB are classified as primary structural elements. All primary structural elements are seismically designed to withstand the effects of an operating basis earthquake (OBE) in accordance with Regulatory Guide (RG) 1.143. Secondary structure elements, including platforms, catwalks, pipe supports, equipment and vessel supports, and internal masonry walls, are classified as nonseismic-resistant items and are designed by conventional method. 1.4 Heating, Ventilation and Air Conditioning+ The heating, ventilation and air conditioning (HVAC) and chilled water systems are designed for the following primary functional requirements: heat, ventilate and air condition the RSSB; remove airborne particulates from the RSSB atmosphere; prevent unfiltered exfiltration of airborne radioactivity from the building; prevent infiltration of airborne radioactivity into the RSSB control room and electrical room; control and provide a means for monitoring (via the main stack) the release of airborne radioactivity via the ventilation exhaust system; minimize the effects on the facility and its occupants from releases of radioactivity into the RSSB atmosphere; collect and filter air displaced via the vents from all RSSB tanks containing radioactive fluids; continuously purge the RSSB of truck exhaust fumes and other hazardous gases to ensure safe occupancy at all times. 1.5 Shielding and Access Control@ Shielding is designed to limit radiation levels on the building exterior, in the control room, in the electrical room, stairwells, and the passageway to the truck bays. Access to the exterior of the RSSB is controlled by access to the protected area, which is controlled by Nuclear Security. Normal UFSAR Revision 14 III-40 June 1996
Nine Mile Point Unit 1 FSAR access to the building interior is via the waste building extension. Two exterior rollup doors allow access for vehicles to the two truck bays. Four exterior doors are normally locked and provide emergency egress. 2.0 Structure and Design Floor and roof plans and sections showing interior walls are shown on Figures III-3 through III-8. 2.1 General Structural Features<'> The RSSB is located to the east of, and is adjacent to, the existing offgas building, waste disposal building, and waste building extension of Unit 1. The arrangement of the RSSB can be considered as follows: process, handling and storage areas. This section is rectangular in shape and approximately 277 ft long below grade, 330 ft long above grade (north-south), and 61 ft wide (east-west). The majority of the primary structural components are reinforced concrete. The foundation mat is generally founded on top of bedrock. The finish grade and truck entrance and exit openings are at el 261'-0". The roof elevation is located at el 301'-2 1/2", with the material handling crane running longitudinally underneath the roof at el 292'-6 1/2". With the exception of a few feet around the perimeter, the crane can service the entire interior area of this section. Those portions of the RSSB which are classified as seismic-resistant elements are designed to maintain their structural integrity during and after all credible design loading phenomena, including OBE. Those items which are classified as seismic-resistant elements are the foundation base mat, structural concrete walls, floors and roof. Nonseismic-resistant structural elements are designed to maintain their structural function for all anticipated, credible design loading conditions encountered during construction, testing, operation, and maintenance of the facility. Those compartments containing with large tanks (over 2,000 steel to contain 1.5 gal) of radioactive liquids are lined tank volumes in the event of a tank rupture during a seismic event. During normal operation, maintenance, and loading and unloading operations, the structure provides sufficient environmental isolation to ensure that the exposure of plant operating personnel and the general public to radiation is ALARA. 2.2 Heating, Ventilation and Air'Conditioning+ Fresh air is filtered and conditioned and supplied to the control and electrical rooms, which are maintained at a slightly positive pressure with respect to other areas of the RSSB and the adjoining radwaste building. Air from other portions of the RSSB is not recirculated back to these areas. Air is recirculated within the RSSB and is processed through a filter system prior to reconditioning and redistribution. The recirculation filter UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR system is comprised of the following primary filtration components:
- 1. Prefilters to remove larger particles to reduce dust loading on the high-efficiency particulate air (HEPA) filters.
- 2. HEPA filters with an individual efficiency of at least 99.97 percent.
All RSSB ventilation exhaustinto air is processed through a filter train prior to discharging the stack. The filter is comprised of the following primary filtration elements:
- 1. Prefilter to remove larger particles to reduce loading of the HEPA filters.
- 2. HEPA filters with an individual efficiency of at least 99.97 percent.
- 3. Two carbon adsorber sections for the removal of radioactive iodine from the exhaust stream.
Final HEPA filters with an individual efficiency of at least 99.97 percent. Air flow through the process areas of thetoward RSSB is from areas of low radioactive contamination potential areas with increasingly higher contamination potential. Air from the two truck bays is ducted to the ventilation exhaust system rather than returned to .the recirculating atmospheric cleanup system to prevent recirculation of truck exhaust fumes in the RSSB. The RSSB atmosphere is continuously purged (10,250 cfm) with clean outside air by operation of the fresh air supply and ventilation exhaust systems. Purge air from the process areas of the RSSB replaces the air drawn from the truck bays such that the entire building is purged via the exhaust from the truck bays. Radioactive tank vents are piped directly into the exhaust system upstream of the filter. Heating coils (electrical), cooling of the filter (chilled water), and fans are located downstream components to protect them from radioactive contamination. Supplemental heating is provided for the control and electrical rooms by duct heaters. Stair towers are provided with space heaters. Chilled water is produced in one of two 100-percent capacity water chillers and circulated by one of two 100-percent capacity chilled water pumps. Single failure of any one fan, heating coil or cooling coil may result in operating variations from the design basisi however, the overall effect with regard to the health and safety of the building occupants or the public will not be compromised. Fresh air inlet and ventilation exhaust penetrations through the RSSB outer walls are each fitted with two series mounted dampers designed to withstand a minimum of 3 psi pressure differential resulting from severe weather pressure are for conditions. All design and specification requirements UFSAR Revision 14 June 1996
Nine Mile Point Unit 1 FSAR nonseismic, nonnuclear safety-related systems and components. Instrumentation and control systems are provided to achieve required space temperature conditions and to maintain air flow requirements to provide acceptable building and process area pressure relationships. Relative humidity is not controlled, although it is maintained at reasonable levels by the HVAC system. All operating control functions are automatic. Temperature control systems in the fresh air supply and recirculating atmospheric cleanup systems are independent. Air flow control systems in the fresh air supply system and the exhaust ventilation system include interlock provisions to maintain pressure relationships upon de-energizing an exhaust or supply fan. Air flow controls of the recirculating atmospheric cleanup system are independent of the other systems. Redundant temperature sensing and control loops are provided in the fresh air supply and recirculating atmospheric cleanup system. Local instruments and remote indication and/or annunciation are provided. 2.3 Shielding and Access Control~> The RSSB is designed to minimize exposure to plant personnel and the public by its location and design. The RSSB is located within the protected area and is heavily shielded by reinforced concrete. 3.0 Use The RSSB was constructed with the specific intent of providing onsite storage of low-level radioactive waste (LLW). The need to store LLW onsite is the result of the federal Low-Level Radioactive Waste Policy Act as amended in 1985, which initiated the process by which the three existing LLW disposal sites (Barnwell, SC; Beatty, NV; and Hanford, WA) would no longer be required to receive LLW. Although originally designed to store Unit 1 LLW, the RSSB is capable of providing interim storage of LLW produced at both Unit 1 and Unit 2. From a technical standpoint, the storage of Unit 2 waste at Unit 1 is considered acceptable based on the following: 1 ~ The isotopic library to be considered is essentially the same for both units; 2 ~ The isotopic distributions for the two units are similar; however, since Unit 2 is a zinc injection plant, the distribution is more heavily weighted toward Zn-65, while Unit 1 is more heavily weighted toward Co-60. The net impact on interim storage in the RSSB is not significant since the shielding has been designed assuming the more limiting Co-60 levels of Unit 1;
- 3. The selective storage of the high-activity LLW from both units in the RSSB (and the low-activity LLW at UFSAR Revision 14 III-43 June 1996
Nine Mile Point Unit 1 FSAR Unit 2) creates the potential for the storage of greater average activity concentration in the building, although not greater volume. However, since the RSSB was designed assuming the storage of incinerated resins which represent a bounding activity concentration, the building design is considered adequate for the combined storage from both units; 4 ~ Total activity in the RSSB will ultimately be controlled per the Site radiation protection program to ensure that both onsite and offsite dose and dose rate limits are maintained; and
- 5. The transfer of by-product material between Unit 1 and Unit 2 will be conducted in accordance with approved radiation protection implementing procedures.
Radioactive piping is routed through a shielded pipe tunnel and in shielded areas to limit exposure. Major pieces of equipment that can be significant sources of radiation exposure are each provided with a separate shielded cubicle. The storage vaults are shielded with 48 in of concrete in the storage zone (below crane). The roof is 24-in thick. The tank cubicles are shielded by 36 in of concrete. The east-west. truck bay is equipped with a retracting shield door in the ceiling which mitigates albedo radiation in the truck bay from the storage vaults. The low-level storage room and the process equipment cubicle are equipped with sliding shield doors. Access is controlled administratively by the Unit 1 Radiation Protection Program. Physical control of high radiation areas is maintained in accordance with Technical Specifications. UFSAR Revision 14 III-44 June 1996
Nine Mile Point Unit 1 FSAR J. REFERENCES
- 1. Catalytic, Inc., Project No. 36700, System Description for Radwaste Solidification and Storage Building, Procedure No.
601 Revision 1, February 26, 1981. 2 ~ Catalytic, Inc., Project No. 36700, System Description for Heating Ventilating and Air Conditioning (HVAC) and, Chilled Water Systems, Procedure No. 204, 204.1 Revision 1, February 10, 1981.
- 3. Catalytic, Inc., Project No. 36700, System Description for Radiation Protection, Procedure No. 603 Revision 0, October 14, 1981.
UFSAR Revision 14 III-45 June 1996}}